ference onn Re Co se Ma Sydneyn liag 5–9 November 2007 t i eme ctive U nt and Effe Organized by the International Atomic Energy Agency (IAEA) Hosted by the Government of Australia through the Australian Radiation Protection and Nuclear Safety Agency (ARPANSA) IAEA web site www.iaea.org International zation h Reactors arc fe Sa LIST OF EXTENDED SYNOPSES Synopses no. IAEA-CN- Session/Synopses Title Main Author 156/ Session 1: Welcome & Opening Addresses Session 2: Experience with the Code of Conduct on the Safety of Research Reactors Where to with the Code of Conduct on the Safety of S-1 Research Reactors? Loy, J.G. Feedback from the Regional Meetings on the Application of the Code of Conduct and Updating of the IAEA Programme S-69 on Research Reactors Safety Abou Yehia, H. Effectiveness of Safety Regulation at Russian Research S-3 Reactors in Compliance with International Practice Sapozhnikov, A.I. Some Operational Aspects of BRR's Management System S-4 with respect to the Code of Conduct Tozser, S. Session 3: Sustainable Utilization and Strategies Utilization Programmes For Low And Medium Flux U-11 Research Reactors Chaplot, S.L. A Corrosion Monitoring Programme for Research Reactor F-3 Spent Fuel Basins Ramanathan, L. The IEA-R1 Research Reactor: 50 years of operating experience and utilization for research, teaching and U-40 radioisotopes production Saxena, R.N. U-43 The Utilization of 10mW Research Reactor in Tashkent Yuldashev, B. The 30 kW Research Reactor Facility in Ghana: Past, U-13 Present and Future Programmes Nyarko, B.J.B. The Halden Reactor Project: Experience gained in U-18 international research Beere, W.H.Å. The utilization of research reactors and related facilities for the supporting of innovative power reactor and related fuel U-35 cycle Itoh, M.I. RACE-T Experimental Activities – An overview of the subcritical measurements preliminary to the accelerator U-9 coupling experiment Rosa, R. Session 4: Safety Management and Operational Safety Improvements in the Management of Safety in Research Reactor Operation through Appropriate Application of S-5 Selected Power Reactor Good Practices Voth, M.H. S-41 Safety Management at NRG; Petten Boogaard, J.P. Safety Management and Effective Utilization of Indian S-52 Research Reactors Apsara, Cirus and Dhruva Shukla, D.K. Ageing Management of Pakistan Research Reactor-1 S-45 (PARR-1) Latif, M. S-40 The Safety Reassessment of Research Reactors in France Bignan, G. S-50 Compatibility of safety and security Repussard, J.R. Synopses no. IAEA-CN- Session/Synopses Title Main Author 156/ Operational Safety Experience at 14 MW TRIGA Research S-25 Reactor from INR Pitesti, Romania Ciocanescu, M. Ljubljana TRIGA Mark II: 40 Years of Successful S-34 Operation Peršič, A. Ageing Management Programme of Kartini Reactor for S-39 Safe Operation Nitiswati, S. Periodic Safety Review Management for French Research S-51 Reactors - TSO Approach Couturier, J. Experience of IPEN-CNEN/SP in the Execution of the First Phase of the Safety Culture Enhancement Programme at S-57 IEA-R1 Brazilian Research Reactor Vieira Neto, A S. Development of Safety Performance Indicators for S-58 HANARO Wu, J.S Emergency intervention plan for 14 MW TRIGA - Pitesti S-46 Research Reactor Margeanu, S. Session 5: Fast Flux Research Reactors Analysis of sodium cooled fast reactors operation and U-3 consequences for future reactor design and operation Guidez, J. Two Decades of Operating Experience with the Fast U-31 Breeder Test Reactor Gopala Iyengar, S. Budget uncertainty and minimum detectable concentrations using relative, absolute and k0 –IAEA standardization for U-5 the INAA laboratory of the ETRR-2 Khalil, M.Y. Usage of the BOR-60 reactor for investigation of advanced U-20 fuel cycles and materials Bychkov, A.V. Session 6: Research Reactor and Network Corporation Developing Research Reactor Coalitions and Centres of U-25 Excellence Goldman, I.N. IAEA’s Subprogramme on Research Reactors: Technology U-44 and Non-Proliferation Adelfang, P. Research Reactor Utilization in the Mediterranean Region - the Experience of Montenegro and Possibilities for Future U-46 Cooperation Jovanovic, S. Role of the Oarai Branch as a facility open for university U-16 researchers in utilization of research reactors Shikama, T.T. Session 7: Regulatory Aspects and Experience with Current Research Reactor Issues Including Safety Aspects of Core Conversion Canadian Experience in Implementing Modern Regulations S-29 to Existing Research Reactors Alwani, A. The Role of Regulatory Authority in Safe Operation of S-31 Research Reactors Mikulski, A.T. Causal Factors Guide For The Evaluation Of Accidents In S-33 Research Reactors Perrin, C.D. Synopses no. IAEA-CN- Session/Synopses Title Main Author 156/ Contribution of Research Reactors to the Programmes for U-30 Research and Technological Development on the Safety Couturier, J. Review of the ANSTO Application for a Facility Licence to Operate the OPAL Research Reactor in Australia: Case S-59 study review of operational readiness Ward, J.S. S-38 Research Reactors in Germany: An Overview Schneider, M. S-42 The French approach for the regulation of research reactors Conte, D. S-30 MARIA research reactor conversion to LEU fuel Krzysztoszek, G. Session 8: Specific Utilization Applications Experience with different methods for on- and off-line detection of small releases of fission products from fuel U-2 elements at the HOR Delorme, T.V. Neutronic Analysis for the Fission Mo-99 Production by U-7 Irradiation of a LEU Target at RECH-1 Reactor Medel, J. Root Cause Analysis of Swelling Problem in Kartini U-4 Reactor Syarip, S. Technique of testing the VVER-1000 high burnup fuel rods U-23 in the MIR reactor at the design basis RIA parameters Alekseev, A.V. Realization of the IBR-2 Research Reactor Modernization U-41 Program Vinogradov, A.V. Session 9: Decommissioning and Waste Management DE-3 Overview of Research Reactor Decommissioning Rowling, J. The transition from a research reactor in operation to a DE-2 facility undergoing decommissioning Nielsen, K.H. Carbon steel construction removal from spent fuel storage WM-2 pool Pešic, M. Session 10: Core Safety and Utilization Parameters Status Report on Preparation of IAEA Guidelines for S-36 Qualification of Research Reactor Fuels Snelgrove, J.L. Simulation of Flow Behavior in the HANARO Reactor S-13 Pool by Using the MARS Code Park, C. Experimental Measurements for Plate Temperatures of MTR Fuel Elements at Sudden Loss-of Flow Accident and S-27 Comparision with Computed Results Sevdik, B. S-60 Contract Performance Demonstration Tests in the OPAL Hergenreder, D.F. Session 11: Programmes for the Minimization of the Use of HEU Status of the United States Department of Energy's Nuclear F-8 Fuel Return Programmes Dickerson, S.L. Conversion of Research and Test Reactors to Low Enriched F-11 Uranium Fuel: Technical Overview and Programme Status Roglans-Ribas, J. Wachs, D.M. F-12 High Density Fuel Development for Research Reactors Lemoine, P. Synopses no. IAEA-CN- Session/Synopses Title Main Author 156/ Measuring Progress in Reactor Conversion and HEU Minimization Towards 2020 – the Case of HEU-fuelled S-35 Research Facilities Reistad, O.C. Session 12: New Research Reactor Projects S-21 Commissioning the new OPAL Research Reactor Irwin, A. S-23 Reactor PIK status of construction Konoplev, K.A. Licensing of the OPAL Reactor during construction and S-37 commissioning Summerfield, M. S-24 Concept for a new research reactor in Ukraine Lobach, Y. Poster Presentations: Safety S-9 Safety analysis of a 1-MW pool-type research reactor Hamidouche, T. Safety Core Parameters Prediction in Research Reactors using Artificial Neural Networks: A Comparative Study of S-12 various Learning Algorithms Mazrou, H. Determining Degradation of Sensors used for Fuel Assemblies Cooling Temperature measurement by the S-18 Analysis of Thermal Fluctuations Noise Saadi, S. Physical calculations for the design of irradiation device and nuclear heating calculation in silicon ingot at Es-salam S-20 research reactor Widad, T. S-43 Upgrading I&C for the Es Slam Research Reactor Djaroum, B. U-48 Training and Qualification of Reactor Operating Personnel Kulak, R. Experimental Heat Transfer Analysis of the IPR-R1 TRIGA S-26 Reactor Mesquita, A. Appraisal for the Operation Safety of SPRR-300 in recent U-1 years Chen, W. Calculations and Measurements at the Training Reactor S-14 VR-1 Rataj, J. Improvements of the Training Reactor VR-1 – Only Way S-54 how to be Attractive for Students and Scientists Sklenka, L. Study of temperature effects on ULYSSE reactor for S-44 training and qualification of operating personnel Foulon, F. Monte Carlo Simulation of GRR-1 Core and Neutron S-17 Irradiation Positions Using MCNP Stamatelatos, I.E. S-28 Radiological characterization of the GRR-1 pool Tzika, F. The Next 20 Years Operation of the 36 Years Old S-71 Hungarian Training Reactor Aszódi, A. Validation of the Monte Carlo Method of MVP Code on the S-15 First Criticality of Indonesian Multipurpose Reactor Sembiring, T.M. Neutronic Analysis of RSG-GAS Silicide core with S-16 Uranium density of 4.8 gr/cc Setiyanto Fault Position Estimation Using Color Segmentation Approach to the Thermal Infrared Image of a Cabling S-48 Network in the Nuclear Reactor Nugroho, D.H. Water Chemistry Surveillance for Multi Purpose Reactor S-56 30MW GA Siwabessy, Indonesia Sunaryo, G.R. Synopses no. IAEA-CN- Session/Synopses Title Main Author 156/ Assessment of qualification of RSG-GAS Reactor operator U-51 in Indonesia Pandi, L.Y. Developing an Ultrasonic NDE System for a Research S-49 Reactor Tank Perets, Y. Neural networks application to CRDM and thermohydraulic S-70 data validation Sepielli M. Investigation of JRR-3 control rod worth changed with burn U-32 up of follower fuel elements Hosoya, T. Project to Replace the Control and Protection System at the S-72 WWR-K Research Reactor Chakrov, P. Modeling of Thermal Hydraulics Behaviour in Reactor S-11 Core Mohamed, K.N. PUSPATI TRIGA Reactor: 25 Years of Safe Operation and S-67 Strategies for Ensuring Safety and Security Masood, Z. Core Calculation of 1MW PUSPATI TRIGA Reactor (RTP) using Continuous Energy Method of Monte Carlo S-68 MVP Code System Abdul Karim, J. Operational Experience and Programmes for Optimal S-10 Utilization of the Nigeria Research Reactor-1 Jonah, S.A. Regulatory control of the Nigerian Research Reactor S-32 (NIRR-1) Ogharandukun, M.O. Comparative dose calculation for TRIGA HEU and LEU S-47 fuel in nuclear accident situations Margeanu, S. Safety Analysis of MNSR Reactor during Reactivity S-8 Insertion Accident Using the Validated Code PARET Hainoun, A. Fuel management methodology upgrade of Thai Research S-19 Reactor (TRR-1/M1) using SRAC computer code Tippayakul, C. Poster Presentations: Utilization Operating Experience Utilization Programmes of the BAEC U-54 3 MW TRIGA Mark-II Research Reactor in Bangladesh Haque, M. Studies on environmental pollution in Bangladesh using U-55 reactor based neutron activation analysis technique Abdul Latif, S. Improvement on Sensitivity for the Track - Etch Neutron U-8 Radiography Pugliesi, R. U-10 New Perspectives for the TRIGA IPR-R1 Research Reactor Soares Leal, A. First experiments in the IPEN-CNEN/SP PSD Neutron U-15 Powder Diffractometer Parente, C.B.R. Future utilization of the Research Reactor IRT IN SOFIA U-45 after its reconstruction Ilieva, K.D. U-38 Utilization of the LVR-15 Research Reactor at Rež Marek, M. U-49 Utilization of the VR-1 Training Reactor Matejka, K. The 250 kW FiR 1 TRIGA Research Reactor - International Role in Boron Neutron Capture Therapy (BNCT) and Regional Role in Isotope Production, Education and U-28 Training Auterinen, I. Chemical characterization of early fine-ware pottery by neutron activation analysis: analytical and statistical U-6 approaches to production and trade Kilikoglou, V. U-56 Feasibility Study of I125 Brachytherapy in Indonesia Soentono, T.M. U-17 Characterization of Airborne Particulate Matter at Urban Santoso, M. Synopses no. IAEA-CN- Session/Synopses Title Main Author 156/ and Rural Area in Bandung and Lembang Indonesia using Instrumental Neutron Activation Analysis Characterization of a Neutron Collimator for Neutron U-14 Radiography Applications Palomba, M. Optimizing Conditions Suited for Stress Determinations in U-33 Q-Space Focusing Configurations Ionita, I. U-34 SANS facility at the Pitesti 14MW TRIGA Reactor Ionita, I. RIAR Capabilities in Support of the Innovative Nuclear U-19 Technologies Bychkov, A.V. U-21 SM Reactor After Core Modernization Klinov, A. Set of Investigations of HFR Fuel Rods in Justification of U-22 their Serviceability and Safe Operation Tsykanov, V.A. U-12 Utilization of irradiation holes in HANARO Lee, C.S. U-27 Design and Installation of Fuel Test Loop in HANARO Ahn, S.H. Design Characteristics of Cold Neutron Source in U-42 HANARO Wu, S.I. Pure Commercial Gold Foils As Neutron Flux Monitor U-47 Neutron Self-Shielding Assessment Helal, W. Implementation of TRR-1/M1 for Thailand’s Nuclear U-50 Engineering Program Nilsuwankosit, S. Poster Presentations: Fuel and Waste Management F-4 Converting HIFAR to Low Enriched Uranium Fuel Storr, G.J. The regulatory role of the Australian Radiation Protection and Nuclear Safety Agency in relation to spent fuel arising F-6 from research reactors in Australia Sarkar, S. Design and construction of a Decay Pool at IAN-R1 F-7 Research Reactor Sarta Fuentes, J.A. Safety Assessment for Decommissioning of Research DE-1 Reactors Kaulard, J. Further Development in Characterization of Radioactive Waste Drums by Non Destructive Gamma Spectrometry at WM-4 GRR-1 Savidou, A. Al-Clad Spent Nuclear Fuel Corrosion Studies at Magurele F-1 site, Romania Dragolici, A.C. F-2 Vinca site preparation for spent fuel shipment Pešic, M. Experimental study of systematic errors of gamma WM-5 technique for assay of radioactive waste drums Tran, Dung Session 2: Experience with the Code of Conduct on the Safety of Research Reactors Synopses no. IAEA-CN- Synopses Title Main Author 156/ Where to with the Code of Conduct on the Safety of S-1 Research Reactors? Loy, J.G. Feedback from the Regional Meetings on the Application of the Code of Conduct and Updating of S-69 the IAEA Programme on Research Reactors Safety Abou Yehia, H. Effectiveness of Safety Regulation at Russian Research Reactors in Compliance with International S-3 Practice Sapozhnikov, A.I. Some Operational Aspects of BRR's Management S-4 System with respect to the Code of Conduct Tozser, S. IAEA-CN-156/S-1 Where to with the Code of Condtu ocn the Safety of Research Reactors? JohnL oy Chief Executive Officer , AustralianR adiation Protection and Nuclear Sayf et Agency (ARPANSA) The Code of Conduct on the Safety ofs Reaerch Reactors was adopted by the Board of Governors of the IAEA and endoerds by the General Conference in 2004. The development of the Code took place over several years and followed letters to the Director General on research reactor safety from the International Nuclear Safety Advisory Group. The Code is a non-binding internationalg alel instrument designed to ‘serve as guidance to States for, inter alia, the dloepvement and harmonization of policies, laws and regulations on the safety of researecahc tors.’ It contains guidance on best practice directed to the Sta teo, the regulatory body and thoe operating organization. As it is non-binding in nature, the Co ddeoes not itself include a mechanism for implementation based upon the process ofi cratifon and the participation in formal review meetings that implements theo nCvention on Nuclear Safety and the Joint Convention. Nonetheless, processes fofor rmination exchange and a form of peer review are being considered by inteereds tMember States. In 2006, the General Conference supported a propo sthaal t periodic meetingsb e organised to discuss application of the Code in Member Setsa.t It looked forward to discussion of implementation of the Code includinagt this International Conference. The non-binding status of the Code and cthoen sequently more informal nature of mechanisms for implementation may be an advantage in allowing for a graded approach to the different tysp eof research reactors, thesitra tus and the safety issues they face. IAEA-CN-156/S-69 Feedback from the Regional Meetings on the Application of the Code of Conduct and Updating of the IAEAP rogramme on Reaserch Reactors Safety H. Abou Yehia and A.M. Shokr Research Reactors Safety Sectionv,i sDioi n of Nuclear Installation Safety, International Atomic Energy Agency The information related to the safety statusre osfe arch reactors worldwide were collected by the IAEA through a number of channels. Theinscelu de the regional meetings on application of the Code of Conduct on thSe afety of Research Reactors, safety review missions, evaluation of the INSARR (Integrated Saf eAtyssessment of Research Reactors) mission reports, and reports collected via the Incnitd eReporting System of Research Reactors (IRSRR). The analysis of this information helped identifying issues that present safety concern and need an increased attention. Three regional meetings on application oef tChode of Conduct werhee ld by the IAEA in 2006 and 2007. The results of the self-assessments made by the participating Member States (MS) showed that there is a common need to: ‚ Improve the capabilities of regulatory bodiens t he review and sasessment of safety submittals; ‚ Increase the attention to commissioninogf modifications and experiments. In particular, safety aspects of core conversion need to be taken into account in a more effective manner; ‚ Perform site re-evaluation for existing resceha reactors as a part of maintaining their safety assessment up to date; ‚ Develop comprehensive emergency plansd establish prepeadrness and response capabilities at the national level; ‚ Improve the safety culture in operating onrigza tions and address human factors issues in all phases of resecahr reactors lifetime; ‚ Improve the capability to prepare the stya fdeocumentation for decommissioning, and establish criteria for research reactors in extended shutdown and for release from regulatory control for decommssiioned research reactors. The analyses of the results of the IAEA sa freetyview missions wer euseful in identifying common safety issues and ntdres, although the observations m aadned the issues raised by these missions are specific to the problems of einadcivhidual reactor. The results of the recent missions indicated out-of-date safety documeonnta (tsi afety analysis report, operational limits and conditions, emergency plans, etc.), whicehd n teo be updated too ncform the actual status of the facility. These missions showed also the need to: ‚ Improve the role and responsibilities of the safety committees in many operating organizations; ‚ Establish integrated management systems/quality assurance programmes; ‚ Put in place a clear strategy for the manmaegnet of radioactive waste generated from research reactors. IAEA-CN-156/S-69 The analysis of the incidents reported to the IRSRR showed that ageing of components is one of the most important root causes of the innctids,e representing more than 50 %. Two-thirds of the operating research reactors are over 3a0rs y oeld. Although many of these old reactors have been refurbished to comply with today’s safety requirements, ageing of systems and components, including obsolescen of the instrumentation ando nctrol systems, is considered to be an important safety issue. In addition to the safety issues mentioned aeb, ofuvture challenges include development of self-assessment capabilities for safety rev iienw all MS and coordination with other international organizations hanvgi activities regarding research reactors safety. The IAEA programme on research reactsoarsfe ty was developed taking in atoccount these safety issues as well as the future challenges. It entails the following four projects: 1. Enhancing the safety of research reactTorhsis: projec tis focusing mainly on promoting the application of the Code of Conducatn d providing assistance for its effective implementation and completing the corpus of croemhpensive safety standards that support the application of the Code. The project givesio rpitry to the improvement of the regulatory supervision of research rearcst oand self-assessment capabilities of MS, conduct of safety review missions and assistance in impletmateionn of the recommendations given by these missions. The other activities under this prot jiencclude assisting ind eveloping systematic ageing management programmes and decomminssgi opnlai ns, increasing awareness of safety of experiments and modifications includincgo re conversion projects, and establishing integrated management systems. 2. Monitor ing and safety enhancemt eonf research reactors under agreemeTnhte:r e are 35 research reactors in 27 M uSnder Project and Supply Aregement with the IAEA. In accordance with this project, safety review moinssi are conducted at these reactors to ensure their compliance with the requirements of the IA EsaAfety standards. This project will ensure the continuous operation of the follow-up sys tfeomr monitoring the safety of these research reactors, which is based on collection and anisa loyfs data on safety performance indicators and the dissemination othf e operating experience. 3. Foster ing international exchange of infoartmion on research reactors safety aspects: The objective of this project is to facilita thee improvement of operating practices and the exchange of information on events with lensss olearned. Under th isproject, IAEA will continue operating the IRSRRco, nducting reviews of incidesn tand assessment of Safety Review Mission reports to idetifny significant issues, and goarnizing meetings for the exchange of operating experience feedbaTchke. other important aticvities include the development of a Research Reactor Infaotrimon Network (RRIN) and the conduct of Coordinated Research Projects (CRPs) one scutsb jrelated to research reactors safety. 4. Assisting on safety aspects relatteod protection against sabotage for research reactors: This project aims at enhancing thaew areness and understanding the synergy between safety and security and improving prtoiotenc of research reactors against sabotage, including development of the regulatory haourtity’s capabilities on methodologies for assessment and improvement of the synergy between safety and security. In the final paper, the safety status of rese arercahctors including common safety issues and trends, feedback from the regional meetings on the appolicna otif the Code of Conduct, and details of the IAEA programmes on research reactors safety will be presented along with discussions. IAEA-CN-156/S-3 Effectiveness of Safety Regulation at Russian Research Reactors in Compliance with International Practice Alexander Sapozhnikov Federal Environmental, Industrial and Nuclear Supervision Service of Russia (Rostechnadzor), Moscow, Russia E-mail address of main author: sai@gan.ru th Russian research reactors base was created in 50-80 years of XX century to develop fundamental and applied sciences in the sphere of physics, power engineering, and science of materials, biology and medicine. After the USSR collapse the changing of Russian economy was attended with decreasing of state financial support for safety ensuring and development of nuclear research facilities 1 (NRF) and reduction of their amount. Nevertheless the Russian fleet of NRF remains the largest in the world and posed more than 20% of worldwide quantity. Nowadays the Russian atomic industry plans to start commissioning of new units of nuclear power plants on the basis of evolutionary designs and innovating research of power reactors, implementation of “small” and regional nuclear power facilities, problem-solving of closed fuel cycle, spend fuel management, and utilization of radioactive wastes. The high flux beam research reactor is being in commissioning to develop fundamental neutron investigation. Russia supports the conversion of the research reactors situated in development countries to reduced enrichment fuel. Thereupon the lay down development of nuclear science and technology should demand efficient utilization and modernization of Russian research reactors multitude. Russian state regulatory body for nuclear and radiation safety has to be ready to internal and external challenges and its influence shall become stronger on the basis of priorities of safety in problem-solving of the use of atomic energy. There were a few stages of improvement and harmonization of the Russian nuclear regulatory system on the basis of IAEA safety standards and transferred Western European safety principles and practice. Rostechnadzor carries out major functions of the regulatory body in environment, different sphere of industry and nuclear activity. The nuclear and radiation safety regulation in Russia is based on: development of the legislative framework including international agreements, federal laws in sphere of the use of atomic energy, national safety standards and regulations of executive bodies; realization of the licensing system of activities in sphere of the use of atomic energy and system of permissions for personnel; fulfillment of conformance evaluation of equipment and systems; 1 NRF - will be interpreted structures and complexes with civil research nuclear reactors, critical and subcritical nuclear assembles, which have been designed for utilization of neutrons and ionizing radiation for research purposes. IAEA-CN-156/S-3 improvement of inspection programs and promotion of effectiveness of enforcement measures. The legislative and regulatory infrastructure in the sphere of nuclear and radiation safety of NRF in accordance with fundamental safety principles and regulatory approach of IAEA especially under the Code of Conduct on the Safety of Research Reactors is considered. The information on current status of Russian NRF and their state of nuclear and radiation safety is provided. The results of NRF licensing are reviewed as in-process improvement of safety during lifetime of the facility commensurate with grade of facility hazard and nuclear activity. The qualitative indicators of nuclear regulatory efficiency are based on model of quality management being developed by international organizations. The quantitative performance indicators of safety regulation at NRF are appraised in framework of methodology based on final indicators of achievement the plan targets of the regulatory body. With regard to analysis result of current state of safety at Russian NRF the issues of further improvement and harmonization of regulatory system and enhancement of safety at NRF are posed in view of regulatory approach of IAEA and advanced practices of Western European countries. IAEA-CN-156/S-4 Some Operational Aspects of BRR’s Management System with respect to the Code of Conduct S. TQzsér1), F. Gajdo1s), K. Késmárky1), G. Tóth1) 1) MTA KFKI Atomic Energy Research Institute (AEKI), Budapest, Hungary E-mail address of main author: tozser@sunserv.kfki.hu The Code of Conduct (CoC) provides a basis for the self-samsesenst of all aspects of man- agement practice. In line with this idea the evaluation of the BRR’s (Budapest Research Reac- tor) past 15-year management practice with respect to the CoC has resulted in good feedback, which has supported the applied practices and also highlighted deficiencies or weaknesses. The period of self-assessment was an obvious choice as it encompasses the time since the full-scale reconstruction of the reactor (which was completed in 1992); that, more over, repre- sents a sufficient amount of time for drawing useful conclusions, summarhisein ge xpt eri- ences, and showing trends. In this paper, from the perspective of operationa otirogna,n sizome aspects of management practice are highlighted. These include: operation and utilization is- sues; conformity and safety reviews; and safety culture focusing on human factors. The paper also highlights how this management practice relates to and supports nuclear safety and dem- onstrates this safety to the public. It is hoped that the highlighted aspects of management sys- tem described here may serve as a general methodology for the RR community. Operation and utilization practic Te.he BRR has adopted a range of management practices since the time of first criticality in 1959. The development of this practicei nwfluaes nced in a number of ways. Among others self-improvement methodology, regulatqouryir erme ents and the sense of practicality also played a significant role. One of the practical matters concerned the operation and utilization practice that arose during the two-year period following the po- litical changes in Hungary in 1990. This was a period of uncertainty during which it was es- sential to demonstrate the necessity of the BRR. For this reason a consortium, the Budapest Neutron Centre (BNC), was founded by four academic institutes to coordinate the reactor utilization, but first of all help to win public support for reactor start-up. Following the reactor start up, as the BNC put into operation the facilities around the reactor and started to manage the utilization strategy, it became obvious that the BNC could effectively represent the user in- terests, thereby leaving the reactor management to focus on safe reactor operation. Now this management system is highlighted as being one of the best operational practices. [1]. Conformity and safety review sD.uring the last 15 years, two conformity reviews (CRs) and one periodic safety review (PSR) have been conducted at the BRR. The CRs were passed fol- lowing the development of Nuclear Safety Regulations (NSRs) in 1997 and then in 2005. The CRs essentially consisted of a number of comparison assessments, where the conformity of the design requirements and the implementation of operation requirements prescribed in the NSRs were reviewed and evaluated on the existing reactor structures and the applied opera- tion practices including the status survey of mandatory documentation. Regarding the PSR, the Hungarian legal system obliges operators to prepare a safety review every decade. As the BRR’s operation licence was issued in 1993 the operating organization was obliged to conduct a PSR in 2002-2003, which ensured a complex overall review of the BRR, treating the reactor as a complex system whilst taking into account its service life. The generasl opphhilyo of the PSR was to assess the condition of the reactor structures with respect to the 10-year operation record, and evaluate all operational experiences including event records. [2]. IAEA-CN-156/S-4 Safety culture focusing on human facto Srsa.fety has been given the highest priority at the reactor since the time of the first criticality. This safety-committed approach formed a deep- rooted nuclear culture at the BRR that has been inherited down through thaet iognesn eorf employees. As it is essential, the importance of safety is understood both by the top reactor management and by all staff involved in the operation organization. During the PSR in the frame of examining human factors, thorough and painstaking reviews were performed involv- ing all safety issues, including all organization and administrative factloartsin gre to safety. The review statements certified that the 10-year service life of the reactor was safe, there were no violations of the OLCs, and that the operation and maintenance practices met with both local and Licensing Authority’s regulations. However due to some safety increasing measures following the PSR, at present the accident conditions from the perspective of human factors have already been analysed in the final SAR. A new version of the QA program was also is- sued which classifies the human function as a safety critical barrier and stipulates the ex- pected safety policy. That is to say each staff member is committed to implemepnotinlicgy a of safety awareness, which should be based on perception and prevention. Demonstrate the safe tIyt .has been obvious since the PSR that there is an increasing demand by society for a nuclear facility not only to be safe, but also to be seen to be safe. Thus the BRR’s operating organization continuously strives to demonstrate the safety. There are many traditional (regular reporting, open door policy, etc.) and modern practices (internet technolo- gies) that ensure near continuous access to the reactor indicators and events. On the basis of general experience in this field it can be stated that to meet the demand (demonstrate the safety) the common approach of any public information system is to increase the transparency and traceability of the activities. The public information must not only contea inin ftohrmation necessary to justify the conformity and demonstrate the safety, but mussot idn oa n easily accessible manner. Based on feedback from the public, the most important considerations are that the information be authentic, coherent and comparable with previous years. It must also clearly report unscheduled events and provide comparison to well-known pointers. To meet this public demand one of the best comparisons to make is to demonosntrfaotrem city to op- eration regulations and records. Another strong comparison to make is between the safety indicators of the reactor and the appropriate aspects of the IAEA’s standards and recommen- dations, and (above all) the IAEA’s ‘Code of Conduct on the Safety of Research Reactors’. In summarizing the general management practice with respect to the CoC two important statements can be made. Firstly: the CoC does not contradict the everyday practices (includ- ing the legislative and safety standards and regulation system). Secondly: the CoC clarifies the duties and responsibilities of all, be they organizations, regulaotodriey sb, or individuals (regardless of their position relative to the management hierarchy). Hence, the CoC acts as a compass that harmonizes and directs the safety approach taken at all levels, from the top to the implementation. It may be said therefore that it shapes the unity of content and form of a RR’s safety. [1] S. TQzsér: Full-scale reconstruction and upgrade of the BRR. IAEA Training Meet- ing/Workshop on "Modernization and Refurbishment of Research Reactors". 9 - 11 Oc- tober 2006, Delft University of Technology, Delft, The Netherlands. [2] S. TQzsér, J. Gadó, K. Késmárky, I. Vidovszky: Periodic Safety Review of the Budapest Research Reactor. International Conference on Topical Issues in Nuclear Installation Safety: Continuous Improvement of Nuclear Safety in a Changing World (IAEA-CN- 120/4), pp. 186-190. Beijing, China, 18–22 October 2004. Session 3: Sustainable Utilization and Strategies Synopses no. IAEA-CN- Synopses Title Main Author 156/ Utilization Programmes For Low And Medium Flux U-11 Research Reactors Chaplot, S. L. A Corrosion Monitoring Programme for Research F-3 Reactor Spent Fuel Basins Ramanathan, L. The IEA-R1 Research Reactor: 50 years of operating experience and utilization for research, teaching and U-40 radioisotopes production Saxena, R.N. The Utilization of 10mW Research Reactor in U-43 Tashkent Yuldashev, B. The 30 kW Research Reactor Facility in Ghana: Past, U-13 Present and Future Programmes Nyarko, B.J.B. The Halden Reactor Project: Experience gained in U-18 international research Beere, W.H.Å. The utilization of research reactors and related facilities for the supporting of innovative power U-35 reactor and related fuel cycle Itoh, M.I. RACE-T Experimental Activities – An overview of the subcritical measurements preliminary to the U-9 accelerator coupling experiment Rosa, R. IAEA-CN-156/U-11 Utilization Programmes For Low And Medium Flux Research Reactors S. L. Chaplot Solid State Physics Division, Bhabha Atomic Research Centre, Mumbai 400085, India chaplot@barc.gov.in We consider various major utilization programmes that may be pursued at any research 12 2 reactor and discuss relevance to low flux (10 neutrons/cm /s) and medium flux reactors up 14 2 to 1x10 neutrons/cm /s. Low-flux reactors inherently have low radiation field and much simplicity, and therefore, despite their limitations of flux, enable greater flexibility in designing new applications and testing of new ideas. These could generate valuable experience that could form the basis for advanced applications at higher flux reactors. 1. Neutron beam research Thermal neutrons have been used to investigate various nano-scale structural, dynamical and magnetic properties of materials of scientific interest and technological importance. A variety of specific techniques are used such as diffraction, small-angle scattering, reflectometry, neutron depolarization, and inelastic and quasi-elastic scattering. The main advantages of using neutrons over much brighter photon beams are that the former provide (i) a sharp contrast between isotopes or between neighbouring atoms in the periodic table, (ii) high energy resolution of sub-meV, (iii) investigation up to large wave-vector transfer needed for liquids and amorphous structures, (iv) intense magnetic probe, (v) characterization of bulk samples and (vi) particularly large cross-section of scattering from hydrogen atom. All the techniques may be fruitfully exploited at medium-flux facilities. However, experiments at low-flux facilities might be limited to using diffraction techniques on samples of relatively simple crystal and magnetic structures. The experimental results are usually of fundamental scientific interest, and are expected to lead to newer ideas and possible applications. 2. Neutron beam applications These involve characterization of various materials for their properties at microscopic or mesoscopic length scale. Typical examples are (i) residual stress analysis; (ii) surface and interface characterization including corrosion in magnetic and nonmagnetic multilayers; (iii) hydrogen sensing including environment around the hydrogen atoms; (iv) size and shape of large molecular systems such as hydrocarbons, micells, polymers and proteins and (v) voids in bulk materials such as steel, ceramic, coal and cement. These applications, usually possible at medium-flux facilities, are expected to lead to improved materials and processes. 3. Neutron radiography This technique is complementary to gamma-ray and x-ray radiography. Specifically large engineering systems can be characterized by neutron radiography due to bulk penetrability of thermal and epithermal neutrons and their sensitivity to hydrogen. Of particular interest are investigations of internal defects in thick samples. Special techniques, such as hydrogen-sensitive epithermal neutron radiography, could detect rather low concentration of hydrogen atoms. Real-time images may be observed by dynamic neutron radiography. These applications are well suited at both low and medium-flux reactors. IAEA-CN-156/U-11 4. Radioisotope production A large variety of radioisotopes with different half-lives for applications to industry, health care, education and research, and other societal benefits can be produced by neutron absorption in reactors. These applications are well suited at medium-flux facilities. 5. Medical applications such as boron neutron capture therapy (BNCT) While such applications are indeed highly desirable, they require large infrastructure involving nuclear and medical professionals. 6. Industrial applications such as neutron transmutation doping. Neutron irradiation can provide excellent uniformity of dopants by transmutation in large size ingots of silicon etc. 7. Neutron activation analysis (NAA) and prompt gamma neutron activation analysis (PGNAA) These form major programmes at low and medium-flux reactors for trace-level detection of many elements in a variety of samples of geological, nuclear and other materials. 8. Research and development support to nuclear energy programme Research reactors may be used for testing of fuel and other power reactor components, neutron irradiation and radiation damage studies. 9. Human resource development for nuclear energy programme 10. Education and training at university level in basic sciences and technology Most of the above programmes have relevance to both low and medium flux reactors though the scales may differ. At low-flux facilities the focus would be more on NAA and radiography while crystal structure investigations and some isotope production are feasible. Experiences at several medium-flux facilities show that preliminary and even complete research investigations are often possible. These could form the basis for advanced research at high flux facilities. In India a national facility for neutron beam research is operated at the research reactor 14 2 Dhruva (10 neutrons/cm /s). It includes single-crystal and powder diffractometers, a polarization analysis spectrometer, inelastic and quasi-elastic scattering spectrometers in the reactor hall, and small-angle scattering instruments and a polarized neutron reflectometer in the neutron-guide laboratory. A diffractometer for residual-stress measurements is being built. In addition a neutron radiography facility and a detector development laboratory are 12 2 located at APSARA reactor (10 neutrons/cm /s). All the instruments including the detectors and electronics have been developed within India. The National facility is utilized in collaboration with various universities and other institutions. Various applications at the reactor facilities in India include neutron beam research and applications; neutron radiography; neutron activation analysis, support to nuclear energy programme, radioisotope production; as well as education and training, manpower development and teaching. IAEA-CN-156/F-3 A Corrosion Monitoring Programme for Research Reactor Spent Fuel Basins. L.Ramanathan1, R. Haddad2 and P.Adelfang3 1. IPEN, Brazil, 2. CNEA, Argentina, 3. IAEA, Austria. The main reason for conducting a corrosion surveillance or corrosion monitoring programme (CMP) at a RR facility or at a spent fuel wet-storage facility is to evaluate the effect of the prevailing water parameters at either facility on the corrosion of the spent fuel cladding and/or of other structural materials. A programme of this nature gives an insight into the extent of corrosion of the metallic materials. It is well known that good quality water is essential to prevent corrosion in a spent fuel basin. However, certain water parameters like conductivity, chloride ion content and some other ions, in quantities well below levels of concern, have a synergistic effect on the pitting corrosion behavior of aluminium alloys. Hence, maintenance of water parameters within specified limits is not reason enough for complacency about corrosion of fuel cladding. A well-planned and executed CMP would give the RR or spent fuel basin manager an insight into the state of fuel cladding and/or metallic structural materials in terms of corrosion. A CMP involves the exposure of test coupons to the RR or basin water for a pre-determined period followed by its evaluation to detect for corrosion. The CMP also involves the determination of water parameters at periodic intervals. Thus, the RR or basin manager is informed of any transients in water parameters, (which often go unperceived in the absence of a CMP) and its effects, if any, on the corrosion of the coupons, and consequently on that of the fuel cladding and other structural materials. A CMP involves 3 stages, the planning stage, the execution stage and the action stage. The planning stage includes: (1) listing metallic materials that are exposed to the RR or spent fuel basin water; (2) specifying their composition, microstructure, the heat treated condition and surface condition; (3) selecting the materials for the programme; (4) specifying the duration of the programme, if any; (5) specifying the frequency of corrosion monitoring; (6) specifying the frequency for monitoring water parameters; (7) specifying the location within the RR or spent fuel basin to place the coupons; (8) determining availability of sufficient materials in the desired states, a manufacturer or supplier adequately equipped to manufacture the test coupons and racks, all accessories required to identify, introduce, follow-up and withdraw the rack of test coupons; (9) determining laboratory facilities for water analysis, sediment analysis, test coupon evaluation; (10) specifying dimensions of the coupon, the insulating separator and the rack; (11) specifying configuration of the coupons with respect to nature of corrosion – crevice and bimetallic; (12) specifying if settled solids need to be evaluated in terms of quantity and composition, and if so, aspects mentioned in sub-items 7-9. The execution stage involves all actions related to preparing the test coupons and racks, conducting the corrosion programme and evaluation of the coupons. A test protocol was elaborated by the authors during the IAEA sponsored Coordinated Research Project (CRP) on ‘Corrosion of Research reactor Aluminium Clad Spent Fuel in Water’ containing all details relate to the execution stage. IAEA-CN-156/F-3 The action stage involves: (1) correlation of the results of coupon corrosion evaluation and the water parameters; (2) correlation of the results of coupon corrosion with that of the spent fuel cladding and/or other structural components; (3) taking of appropriate actions, if necessary, to alter water parameters, reduce extent of settled solids and verify fuel cladding surface integrity. Guidelines to be followed during the three stages of a corrosion surveillance programme for RR spent fuel basins will be presented and discussed. IAEA-CN-156/U-40 THE IEA-R1 RESEARCH REACTOR: 50 YEARS OF OPERATING EXPERIENCE AND UTILIZATION FO R RESEARCH, TEACHING AND RADIOISOTOPES PRODUCTION Rajendra N. Saxena Research Reactor Center Instituto de Pesquisas Energéticas e Nucleares (IPNE/CNEN-SP) São Paulo, Brazil ABSTRACT This paper describes almost 50 years of operatxinpge reience and utilization of the IEA-R1 research reactor for research, teaching randdio isotopes production. The current and future program of upgrading the reactor is adlessocribed. IEA-R1 research reacto r at the Instituto de Pesquisas Energéticas e Nucle(IaPreEsN ), Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximumw epro rating of 5 MWth. It is being used for basic and applied research in the nucalneadr neutron related sciences, for the production of radioisotopes for medical and indiuasl trapplications, and for providing services of neutron activation analysis, real timn eu tron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swiminmg pool reactor, with light water as the coolant and moderator, and graphite and buemry llai s reflectors. The reactor was commissioned on September 16, 1957 and achieve fdir sit scriticality. It is currently operating at 3.5 MWth with a 64-hour cycle per w.e eOk riginally charged with 93% enriched U-Al fuel elements it currently uses 20n%ri cehed uranium (U3O8-Al and U3Si2- Al) fuel that is produced and fabricated at IPEN. In the early sixties, IPEN produce1d31 I, 32P, 198Au, 24Na, 35S, 51Cr and labeled compounds for medical use. In the year 1980, pcrtoiodnu of 99mTc generator kits from the fission 99Mo imported from Canada was started. This prodounc tis continuously increasing, with the current rate of about 16,00i 0o fC 99mTC per year. The 99 mTc generator kits, with activities varying from 250 mCi to 2,0 0m0Ci, are distributed to more than 260 hospitals and clinics in Brazil. Several radiophacrmeutical products based o13n1I , 32P , 51Cr and1 53Sm are also produced. These radioisotopes amo umnot rte than US$15 million in sales per year for IPEN. During the past few years, a concerted effort heaesn bmade in order to upgrade the reactor power to 5 MWth. One of the reasons foisr dthecision was to produc9e9M o at IPEN, thus minimizing the importation co satnd reliance on on lyone or two world suppliers. The reactor cycle will be graduallyr einacsed to 120 hours per week continuous operation. The reactor modernization plans inc ltuhdee following during 2003-2009: • Pool water treatment and purification system. • Replacement of four reactor control elements. • Replacement of one of the primary heat exchanger. IAEA-CN-156/U-40 • Installation of a new core grid plate. • Replacement of the reactor control panel. The other infrastructure modernizations include f othlleowing: • Modernization of the reactor fuel element fabriocna tfiacility in order to increase the production capacity to 15-183 SUi2 type fuel elements per year. • Optimization of radiochemical facilities to proc e99sMso using the gel process. • Development of an effective project for spent fmuealn agement and storage. It is anticipated that these programs will asshuere s tafe and sustainable operation of the IEA-R1 reactor for several more years, to pcroed uimportant primary radioisotopes 99Mo, 125I, 131I, 153Sm and 192Ir. The production of these isotopes will resunlt lei ss dependence on the world supply, and reduce impi onrt actosts. Routine radiation monitoring is carried out at 25 different locatio ins the reactor building. This regular monitoring practice has helped in maintaining tqhuei vealent dose of reactor workers below the established limits. Following the guidelines IAofEA an ALARA program is being implanted at the reactor and associated laborast,o wriheich handle radioactive materials. In order to achieve the goals of modernization raenadc tor aging management, IPEN undertook the following tasks under a technical pceoroation program funded by the International Atomic Energy Agency (IAEA) and Brlaiazni government agencies. • Replacement of some electrical and refrigeratiosnt esmys; radiometric analysis system for water and air samples; reactor contnroslt rui mentation; radiation monitoring equipment. • Neutron flux mapping facility using self-poweredu ntreon detectors (SPNDs). • Improved computational facility for neutronic callactuions. • A highly radioactive sample handling facility. • Training of personnel engaged in electrical and hmaencical maintenance, water chemistry, and irradiation services. • Installation of a continuous vibration monitorinygs tsem for rotating machinery. Currently, all aspects of dealing with fuel elem efanbtrication, fuel transportation, isotope processing, and spent fuel storage arel ehda nbdy IPEN at the site. Spent fuel assemblies are visually inspected routinely usindge urwater cameras. Seeping analysis is performed if there is an indication of fission purocdt release in the pool water. IPEN has an on-going project with IAEA on spent fuel managem. e nTt he reactor modernization program is slated for completion by 2009. IAEA-CN-156/U-43 THE UTILIZATION OF 10Mw RESEARCH REACTOR IN TASHKENT B. YULDASHEV* Stanford University, Stanford, CA 94305-6165, USA U.SALIKHBAEV, A.DOSIMBAEV, S. BAYTELESOV, Yu.KOBLIK, P.PIKUL Institute of Nuclear PhysicsU,l ughbek, Tashkent, 702132, Uzbekistan M. ABDUKAYMOV, P CHISTYAKOV Radiopreparat Enterprise of IN UPl,ughbek, Tashkent, Uzbekistan ABSTRACT We present the short review of basic apnpdl iaed research as well as data on the mass production of reactor isotopes with use 1o0f Mw water-water research reactor of the Institute of Nuclear Physics of Uzbekistan Academy of Sciences in Tashkent. Despite the relatively long time of operaotni (since 1959) the several percotjs on the modernization of machine have been done. The reactor operates more than 5000 hours a year and serves also for elemental analysis (neutron acttiiovna), radiochemistryr,a diation hardness and fission products studies awsell as for changing thper operties of optical and semiconductor materials. Until 1997 reactors woaperating with use of highly enriched fuel (90% of enrichment) and starting fr othme middle of 1997 it hsa been converted to use 36% enrichment fuel. In the second hoaf l2f 007 the preparatory works on the full conversion to the 19.7% enrichment fuheol usld be completed. We also present the results of our experience sinending highly enriched spefnute l back to country-origin (Russia) for first time in last 16 years. The external one-arm spectrometer of secroyn fdisasion products made it possible a study of properties (the chargen, gaular and momentum distribountis) of fission products which revealed quite intereinstg features inconsistent with msoe standard models of fusion. The special method has been developecda rtory out elemental analysis which once applied, for example, to pure metals allotwo sd etermine impurities with concentration up to 10-10 % (for almost 60 elements simultaneloyu).s The reactor is also using for changing the properties of msoe materials like semiconductors, ceramics and natural crystals bringing their quality to market demands. The main activity of our reactor is thmea ss production of isotopes for medical and scientific needs. In particul,a trhe isotopes like Tc-99m (fmro irradiation of Mo wires), P-32, P-33, I-125, I-131, S-35, Au-186, Sr-89,5 F5e a-nd some others are producing at the level of hundreds and thousands Cuarineds, in addition, e rady-to-use products like Tc99m generators for medical cliniacnsd labeled compounds for research are delivering to customers in many countries.e T hhigh chemical purity and specific activity are making those products competitive at international market. ----------------------------------------------------------------- * On leave of absence from the Instituotfe N uclear Physics, Tashkent, Uzbekistan IAEA-CN-156/U-13 Τηε 30 κΩ Ρεσεαρχη Ρεαχτορ Φαχιλιτψ ιν Γηανα: Παστ, Πρεσεντ ανδ Φυτυρε Προγραµµεσ Β.ϑ.Β. Νψαρκο, Ε.Η.Κ. Ακαηο, Ψ. Σερφορ−Αρµαη Γηανα Ατοµιχ Ενεργψ Χοµµισσιον Π.Ο. Βοξ ΛΓ 80 Λεγον−Αχχρα Γηανα Χελλ: 233−21−7406076 Φαξ: 233−21−400807 Ε−µαιλ: βηβνψαρκο≅ψαηοο.χο.υκ, β.νψαρκο≅γαεχγη.οργ Τηε Γηανα Ρεσεαρχη Ρεαχτορ−1 (ΓΗΑΡΡ−1) ισ α σµαλλ, σιµπλε, ρελιαβλε ανδ σαφε ρεαχτορ δεσιγν ανδ χονστρυχτεδ βψ Χηινα Ινστιτυτε οφ Ατοµιχ Ενεργψ (ΧΙΑΕ). ΓΗΑΡΡ−1 αδοπτσ τηε ποολ−τανκ στρυχτυρε ανδ εµπλοψσ ηιγηλψ ενριχηεδ υρανιυµ ασ φυελ, λιγητ ωατερ ασ µοδερατορ ανδ χοολαντ, µεταλ βερψλλιυµ ασ ρεφλεχτορσ. Τηε ρεαχτορ ισ χοολεδ βψ νατυραλ χονϖεντιον. Τηε ρατεδ µαξιµυµ τηερµαλ ποωερ οφ ΓΗΑΡΡ−1 ισ 30 κΩ; τηε 12 −2 −1 χορρεσπονδινγ νευτρον φλυξ ισ 1.0ξ10 χµ σ . Τηε ρεφυελινγ µοδε οφ τηε ρεαχτορ ισ το τοταλλψ χηανγε τηε ολδ χορε ωιτη α νεω ονε, τηε λιφετιµε βεινγ µορε τηαν τεν ψεαρσ. Σινχε τηε χοµµενχεµεντ οφ οπερατιον οφ τηε λοω−φλυξ µινιατυρε νευτρον σουρχε ρεαχτορ (ΜΝΣΡ) ιν 1995, α σιγνιφιχαντ νυµβερ οφ ρεσεαρχη ανδ δεϖελοπµεντ ιν τηε φιελδ οφ νευτρον αχτιϖατιον αναλψσισ ηαϖε τακεν πλαχε. ∆υρινγ ιτσ 12 ψεαρσ οφ οπερατιον, αφτερ τηε φιρστ χριτιχαλιτψ, τηε ρεαχτορ ηασ βεεν υσεδ ασ α νευτρον σουρχε φορ ρεσεαρχη, τεαχηινγ ανδ τραινινγ το συππορτ σεϖεραλ γραδυατε ανδ ποστ γραδυατε χαρεερσ φορ στυδεντσ φροµ υνιϖερσιτιεσ ιν Γηανα ανδ τηε Ωεστ Αφριχαν συβ−ρεγιον. Οωινγ το τηε σταβλε φλυξ οφ τηε ρεαχτορ ανδ ραπιδ προλιφερατιον ιν υτιλιζατιον, σεϖεραλ αναλψτιχαλ τεχηνιθυεσ ηαϖε βεεν δεϖελοπεδ. Ασ α νατιοναλ νευτρον σουρχε ρεαχτορ φαχιλιτψ, Γηανα’σ ΜΝΣΡ αλσο κνοων ασ ΓΗΑΡΡ−1 ισ νοω συχχεσσφυλλψ υτιλιζεδ ιν ϖαριουσ αρεασ οφ νευτρον αχτιϖατιον αναλψσισ (ΝΑΑ), τεαχηινγ, ρεσεαρχη ανδ τραινινγ. Τηε ΓΗΑΡΡ−1 αππλιχατιον ιν νευτρον αχτιϖατιον αναλψσισ ινχλυδεδ: (ι) Φοοδ αναλψσισ; (ιι) Ηεαϖψ µεταλσ δετερµινατιον ιν ενϖιρονµενταλ σαµπλεσ; (ιιι) ∆ετερµινατιον οφ µαϕορ, µινορ ανδ τραχε ελεµεντσ ιν γεολογιχαλ σαµπλεσ; (ιϖ) Ανδ µινεραλ προσπεχτινγ αµονγ οτηερσ. Τηε εδυχατιοναλ προγραµµεσ ιν πλαχε ατ τηε χεντερ αρε τεαχηινγ ανδ λεαρνινγ ιν νυχλεαρ ενγινεερινγ, νυχλεαρ πηψσιχσ, νυχλεαρ ανδ ραδιοχηεµιστρψ ανδ οτηερ ρελατεδ φιελδσ. Τηε παπερ ωιλλ φοχυσ ον τηε παστ ανδ χυρρεντ στατυσ οφ ΓΗΑΡΡ−1 ωιτη ρεσπεχτ το υτιλιζατιον ανδ µαναγεµεντ ανδ φυτυρε προγραµµεσ το ενηανχε ιτσ υσεσ ιν τηε φιελδσ οφ τεαχηινγ, ρεσεαρχη ανδ τραινινγ. Τηε πρεσεντ ανδ φυτυρε χολλαβορατιον προγραµµεσ ωιτη νατιοναλ ινστιτυτιονσ ανδ τεχηνιχαλ χοοπερατιον αµονγ σοµε ΑΦΡΑ µεµβερ στατεσ σηαλλ βε γιϖεν. IAEA-CN-156/U-18 The Halden Reactor Project: Expercien gained in international research William Beere1 ) 1) Institutt for energiteknikk, Norway william.beere@hrp.n o The Norwegian government has with thseu ccessful completion of the June 2006 “Symposium and Technical Workshop on Minimisation of Highly Enriched Uranium in the Civilian Nuclear Sector” [1] confirmed its engamgent for minimizing the risk related to the continued use of fissile material in genernadl aHEU in particular. One suggested concept in reducing risk associated with fissile materrieaql uired for research is the concentration of research efforts in shared facilities. The increased global demand for nuclear energy also points towards the need for first class rese afarchilities. Pooling resources makes sense as a way to maximise research efforts. But how should a shared facility work and function? The OECD-Halden Reactor Project [i2s] a good example of operation of a shared research facility which has been in operation for nearly 50 yse. aArn overview of how this cooperation has worked and how the research facility has gnro awnd developed will be presented in this paper. First a look at the historical perspective.e T ihnitial idea in the 50’sfo r the building of the Halden Reactor was that Norway wanted experience in building nuclear reactors for use in ship propulsion. It soon became clear that Nor wcoauyld not afford this project alone, even the close cooperation between Norway andN theeth erlands was not enough to complete the project. Finally in 1958 when the reactor woapse ned the project had become international with the Institute for Atom Energy (Now lclead the Institute for Energy Technology) being the host. FIG. 1. Cutout view of Halden React or. FIG. 2.Control room simulation and observation gallery, MTO-Laboratorie s Then as now the research was organised3 iyne ar programmes. Countries participating provide financial support, are part of the esrting committee which decides on the content of the research. Visiting research scientists are also sent to Halden to gain experience and to participate in the operation of the experimse. nWt hich and how many countries are members may change. There is no obligation to ren- joaifter a 3 year period has finished. This IAEA-CN-156/U-18 flexibility is probably one of the main reaso nthsat this organisation form has remained unaltered in now almost 50 years. It has alseoa nmt that the research program is constantly changing so as to be as up to date as possible. Another factor contributing to the projectsc cseuss is the design of the reactor. The Halden reactor is a boiling heavy watreera ctor typically operated at 23flC5 . The neutron flux level is typical of commercial power reactors. W hmatakes the Halden Reactor unique is a combination of instrumentation, availability and versatility. With two operational periods each year a ctyapl iavailability of 50% is achieved. This combined with many experimental positions more than compensates for the moderate flux level. In-pile instrumentation has been undevr edloepment since the start of the reactor in 1958. To date almost all physical parametcearsn be measured; fuel temperature, stack elongation; cladding outer temperature, diamerotedr ;i nternal pressure, to name but a few. The versatility of the reactor is also importt;a sneveral different conditions can be simulated simultaneously: BWR, PWR, PHWR. All with their own unique water chemistry, temperature and pressure. Power ramps can be performre din dfoividual fuel channels without affecting the whole reactor; and recently a LOCA test was performed in realistic conditions. In the beginning of the 80’s it was recogniseda t threactor research should not be solely concentrated on research reactors. Other aspects of commercial reactor operation were of equal importance, such as computer tecohgnyo,l human factors and psychology. This is where the dynamic nature of the projects organoisna tpi roved its usefulness, with an increased emphasis being placed on these new areas in the 1982-84 three year programme. To date these new areas in safety, Man-TechnolOogrgy-anisation (MTO) represent 40% of the projects activities and include areas such asr:a otopre procedures, computerised surveillance, visualisation techniques, use of virtual reya, liatdvanced communications, software reliability, and human reliability in power station environments. In addition to providing training, experiencned a countless amounts of experimental data for the projects member countries, the Halden Pr ohjeacst also been extremely beneficial for the hosting country Norway. Despite the facta tt hthe Norwegian nuclear industry did not materialise there have been several bietsn effor conventional Norwegian industry (in particularly petroleum) and has lead to sreavl ecommercialisation successes. The most recent being in electronic energy trading and cablel t faduetection. In addition the Institute provides commercial services based on the reacatonrd associated technology with turnover comparable to the research project. In short we believe that the Halden Percotj is a good example of how to operate an international research project. The key ingrnetdsi ebeing flexible project membership, active participation of project members, and ao nsctantly evolving research programme. The research data obtained at Halden is vitalt hfoer safe operation of many nuclear power plants as well as helping the industry gain economsaicv ings. The hosting of the Halden Project has also been beneficial for Norway and the Institute. [1] “Minimization of Highly Enriched Uranium (HEU) in the Civilian Nuclear Sector”, Oslo 2006,h ttp://www.nrpa.no/symposium/ [2] “The Halden Reactor Project”, http://www.ife.no/hrp/index_html?set_language=en&cl=en IAEA-CN-156/U-35 4002 Nar ita-cho, Oarai-machi, Higashi-ibaraki-gun, Ibaraki-ken, Japan, 311-1393 e-mail; masahiko.ito@jaea.go.jp Over 30% of electricity in Japan was produced by the existing power reactor ”LWR”. Nuclear power has an advantage to reduce the emission of greenhouse gas “CO2”. In near future, the LWR are utilized to produce the electricity and it is planned to extend the lifetime to go up the economic efficiency and availability. To meet this technical request, the experimental resolution for the ageing degradation of materials under neutron environment and upgrading the fuel performance are so important issues. From the viewpoint of energy resource and the reduction of environmental burdens, the innovative nuclear reactor systems are investigated in the worldwide, such as generation IV systems and INPRO (International Project on Innovative Nuclear Reactors and Fuel Cycles). In Japan, the demonstration fast breeder reactor is expected to start the operation around 2025. As a key for the realization of the fast breeder reactor system, it is important to improve the fuel performance and to develop the new materials. For this purpose, it is necessary to investigate the fuel behavior to make clear the fuel stability limit and safety margins. The role of research reactor is to figure out the degradation of materials and fuel performance, include the development of new reactor materials. The important thing for the research of fuel performance using the research reactor should be collocated with the fuel fabrication facility for the irradiation test, post-irradiation examination facility, and so on. In Oarai Research and Development Center, there are some research reactors and post-irradiation examination facilities to investigate the fuels and materials for innovative power reactor, such as Fast Reactor, High Temperature Reactor. That is, the research reactors as the thermal neutron irradiation field are Japan Material Test Reactor (JMTR) and High Temperature Test Reactor (HTTR). The Japan Experimental Fast Reactor “JOYO” is utilized as the fast neutron irradiation field. IAEA-CN-156/U-35 The JMTR has been used as the irradiation facility for the research the fuels and materials of the existing power reactor and the fundamental research of materials. The JOYO has been used for the research and development of the fuels and materials of the fast breeder reactor and the fundamental research of new materials development for the fusion reactor. The many type of irradiation devices to meet the objectives of irradiation experiments are provided include the instrumented vehicle for JMTR and JOYO. In addition, there are four post-irradiation examination (PIE) facilities and a fuel specimen fabrication facility in Oarai Research and Development Center. Three PIE facilities are to treat the Plutonium fuels. And also it is possible to fabricate the fuel pellets remotely. Furthermore, the reassembling of irradiation vehicles in the hot cell to reload into the reactor is possible. From the view point of the sophisticated irradiation technology, how to contribute in the field of irradiation effects of reactor materials for fission and fusion reactor is under discussion. Thus, it is possible not only to fabricate the fuel specimens, examine the fuels and materials after irradiation, but also to treat and storage the radioactive waste. Besides, the research center of universities for the irradiation behavior of materials is located in the same site. Hence, Oarai Research and Development Center is able to support the research of irradiation behavior for reactor materials and upgrading the reactor fuels, effectively and efficiently. IAEA-CN-156/U-9 RACE-T Exper imental Activities An overview of the subcr itical maesurements preliminary to the accelerator coupling exper iment Roberto Ros a ENEA, Via Anguillarese, 301 0,0060 Roma, Italy, robrteo.rosa@casaccia.enea.it Mario Carta, Mario Palomba, Stefano Mo -n EtiNEA As it is well known, the concept of accelerator-driven systems (ADS) can provide a solution for the nuclear waste management issue by burning minor actinides. The subcritical n ature of ADS makes them safer regarding prompt criticality accidents. However, the question of the reactivity control of ADS remains a main connc esrince their power is inversely proportional to their reactivity level. The objective of the European Integra Pterdoject EUROTRANS of the EURATOM 6th Framework Program is to bring answers to the high level nuclear waste transmutation in ADS. The EUROTRANS experimental activities have been joined into the ECATS domain, namely Experiment on the Coupling of an Accelerator, a spallation Target and a Sub-bclraitnickael t. In the period between February 2004 – February 2006, in view of the past TRADE project and the subsequent RACE experiment planned in the frame of the EUROTRANS Integrated Program of the t6h European Framework Program, different experimental activities were carried out in the 1 MW TRIGA reactor operated by EN EinA his Casaccia Research Center near Rome. In that time the reactor operation was mainl ys uinb-critical conditions, with the exception of the critical reference core assessment at the beginning of the campaign wherxeim thuem m a power was 50 W. The critical reference configuration was characterized by spectrum index measurements and flux distribution as well. Measurements of spectrum indices (ratio of various fission rates) in different core pso sition were performed usin2g3 5U, 238U and2 37Np miniature (diameter 1.5 mm) fission chambers for radial and axial measurements on locations situated on the main diagonal in the reference configuration, and a complementary measurement performed to check a dissymmetry effect on the radial profile. These measurements were rea wlizitehd the use of a special fuel pin that will be described. The determination of the neutron flux inside tThRIGA core at a critical configuration with a reactor power of 10 watts was performed with the well-known activation technique. A pneumatic Fast Rabbit device was settled down inside thaect orer hall in order to send gold and indium samples into the reactor. Those samples werdei airtread with and without cadmium covers for the assessment of axial and radial flux distributions for both thermal and epithermal neutrons. The cadmium ratios were also calculated. Results deduced from gold and indium sample measurements were compared. From the aluminutmer imala of the irradiate dshuttles, an estimate of the fast neutron flux was derived. IAEA-CN-156/U-9 The RACE-T experiments allowed improving the knowledge of the experimental techniques for absolute reactivity calibration at either startup or shutdown phases of accr-edlerirvaetno systems. Various experimental techniques for assessing a subcritical level were inter-compared through three different subcritical configurations SC0, SC2 and SC3, about -0n.5d, --63 a dollars, respectively. The area-ratio method based of the use of a pulsed noeuurtcroen a pspears as the most performing. When the reactivity estteim isa expressed in dorll aunit, the uncertainties obtained with the area-ratio mtheod were less tha1n% for any subcritical configuration. The sensitivity to measurement location was about slightly more than 1% and always less than 4%. It is noteworthy that the source jerk technique using a transient caused by the pulseds noeuurctreo n shutdown provides results in good agreementht wthiose obtained from the area-ratio technique. In the ADS projects, one of the most important aspects needed for approval of the demonstrator is the experimental verification of the simulation. The determinatihoen noef utron spectrum is particularly interesting (i.e. neutron flux as a function of the ne euntreorgy) for different configurations of the sub-critical device. As well known, the neutron flux in ADS consists of neutrons produced via spallation reactions in the target and ffirsosmio nthse multiplying blanket.U nfortunately the neutron spectra can bneo tmeasured using only one type of detector. To cover the complete energy range of the produced neutrons a tnroenw dneetuector concept based on Micromegas technology, widely used in particles or nuclear physics experiments. One of the main qualities neededth faotr detector is its high resistance to the radiation. A test with sealed prototype placed inside an empty rod of the reactor has been performed in-core in the RACE-T experimental cpamign: the main outcomes will be illustrated. In order to corroborate the outcomes from the previous MUSE experiment, and to i mprove our understanding about the experimental methodologies needed to monitor the subcritical level in ADS, a common calculation benchmark is activated, in the Frame of the IAEAA-CnaRlyPt ical and Experimental Benchmark Analyses of Accelerator Driven Systems, (aAimDiSn)g to evaluate the spatial/energy effects to abpeplied to some selected reivaictyt determination measurements obtained in the frame of the RACE-T experimental campaign. The rationale and the status of the benchmark activities will be illustrated. Acknowledgement s The authors appreciate the efforts and suppoarltl othf e scientists and institutions involved in EUROTRANS and the presented rwk,o as well as the financial support of the European Commission through the contract FI6W-CT-2004-516520. Session 4: Safety Management and Operational Safety Synopses no. IAEA-CN- Synopses Title Main Author 156/ Improvements in the Management of Safety in Research Reactor Operation through Appropriate S-5 Application of Selected Power Reactor Good Practices Voth, M.H. S-41 Safety Management at NRG; Petten Boogaard, J.P. Safety Management and Effective Utilization of Indian S-52 Research Reactors Apsara, Cirus and Dhruva Shukla, D.K. Ageing Management of Pakistan Research Reactor-1 S-45 (PARR-1) Latif, M. The Safety Reassessment of Research Reactors in S-40 France Bignan, G. S-50 Compatibility of safety and security Repussard, J.R. Operational Safety Experience at 14 MW TRIGA S-25 Research Reactor from INR Pitesti, Romania Ciocanescu, M. Ljubljana TRIGA Mark II: 40 Years of Successful S-34 Operation Peršič, A. Ageing Management Programme of Kartini Reactor S-39 for Safe Operation Nitiswati, S. Periodic Safety Review Management for French S-51 Research Reactors - TSO Approach Couturier, J. Experience of IPEN-CNEN/SP in the Execution of the First Phase of the Safety Culture Enhancement S-57 Programme at IEA-R1 Brazilian Research Reactor Vieira Neto, A.S. Development of Safety Performance Indicators for S-58 HANARO Wu, J.S. Emergency intervention plan for 14 MW TRIGA - S-46 Pitesti Research Reactor Margeanu, S. IAEA-CN-156/S-5 Improvements in the Mnaagement of Safety in Research Reactor Operation through Appropr iate Application of Seleecdt Power Reactor Good Practices Marcus H. Voth1 ) 1) United States NucleaRr egulatory Commission E-mail address of main authomr: hv@nrc.gov Research reactor managers are incnregalys iimplementing improvements in their management of safety through the apploicna toi f good practices oriingally developed as power reactor programs. This paper considers wtoa syeslect practices teomulate, effectively incorporate them into a research reactor prmog arand evaluate their cornibtution to safety. Relative to research reactors, power roera pcrtograms look relatively homogeneous when considering source terms, stored energy, pcorwee r density, operating cycles, plant systems and staff sizes. They have potential hazcaornds equences that require effective safety management programs. Finally, power reacgtoernse rate a stream of revenue to fund these programs. The power reactor community hams bcioned their resources with the homogeneity of their challenge to create impressive tsya mfeanagement tools, many of which can be effectively implemented in the research reoar ctommunity. However, not all programs can be effectively implementeidn all research reactors. A number of power reactor programs are anal yinze tdhe paper with consideration of their effective implementation and potentciaoln tribution to research reactor. IAEA-CN-156/S-41 Safety Management at NRG-Petten 1), J.P. Boogaard 1) NRG (Nuclear Research and consultancy Group); PO box 25; Petten; Netherlands E-mail address of author: Boogaard@NRG-NL.COM NRG (Nuclear Research and consultancy Group) develops knowledge, products and processes for safe applications of nuclear technology on behalf of energy, environment and health. NRG has acquired more than 45 years of experience in the nuclear field. Amongst others this includes experience in the operation of a wide range of nuclear facilities, namely: • Operation of the 45 MW High Flux Reactor for materials testing, (medical) isotope production and BNCT; • Operation of a 30 kW Argonaut Low Flux Reactor for training and biological applications; • Operation of Hot Cells for materials testing and isotope separation; • Overall responsibility for a Molybdenum Production Facility • Operation of a Decontamination and Waste Treatment facility for both internal and external services. Management of safety As a part of NRG’s mission statement safety has the highest priority. In respect of nuclear operations, safety requirements have an over-riding priority above of the production demands. To as-sure this mission statement is satisfied outstanding safety management is of paramount importance. Overall safety management is achieved through a combination of a safety- management system and the company safety culture. At NRG, safety management is an Systems & Procedures integral part of operational practices, with V Safety Management Behaviiour the safety procedures and instructions I Code off conductt fully incorporated into NRG’s integrated S Safe I operation management system. The basis for our B Skills & Competences Safe safety culture are the shared values and I Espoused Vallues,, operation shared basic assumptions which are t L At ttiittude Safe formalized in our code of conduct. The operation I visible characteristics of our safety Mission & Policy T management range from our Mission and Safety CulltureY Shared vallues,, Basiic Policy, through the skills and Assumpttiions competences of our staff to the most FIG. 1. Visibility of safety culture versus Safety management visible characteristics the written (based on Schein and Frischknecht) procedures and instructions of our management system. The characteristics of a safety culture are less visible but are based on the shared values and basic assumptions of an organisation. The attitude of the staff is mostly based on the espoused values within an organisation. The visible aspects of the safety culture are the behaviour of the staff and the formal code of conduct of an organisation. IAEA-CN-156/S-41 NRG’s integrated management system NRG developed a management system which satisfies NEN-ISO-9001:2000 as well as NEN- ISO-17025:2000. Where necessary our Management System has been further developed to fulfill additional requirements of the Dutch Nuclear Safety Requirements NVR-1.3, which are based on IAEA safety requirements DS 338 and the corresponding safety guides. These adaptations mostly concern issues of: management review, control of Policy modifications of installations, operations and maintenance of the nuclear installations, control of Company experimental and testing programmes and registration, Profile storage and transport of fissile and radioactive material. The ISO-14001 standard has been taken as the basis for Company our environmental management procedures, which are procedures also integrated in our Management System. The structure of our Management System is presented in Product group documents figure 2. The corporate procedures consist of flowcharts defining the tasks and responsibilities of all relevant processes and apply to all product groups. Where FIG 2. Structure of NRG’s necessary independent control- or review activities and management system quality registrations are prescribed. Every product group has a set of specific procedures and instructions. These documents provide a more detailed description of the specific product group related processes such as detailed operational, maintenance, calibration and testing instructions. NRG’s management system is based on the skills and competences of our staff. NRG’s organisation improvement programme Since organisation culture aspects are essential layers of the defence in depth NRG launched in 2003 an extended organisation improvement programme. This programme has the objective to transforming NRG from a technocratic institute into a human oriented learning organisation. This programme started with a dedicated conscious management development training of the senior management, followed by conscious management development training for the second management layer and young potentials. As a part of the improvement programme NRG developed also a new vision and long-term business strategy, emphasising safety as key business element, to be shared and borne by all company staff. Individual responsibility, open communi-cation with mutual respect and trust, a learning attitude Design Organisation Culture of management and staff, positive engagement and a blame free culture are Risk the key elements which are also incorporated into our code of conduct. An open learning attitude is pro-actively being stimulated by top management encouraging an environment in which all safety issues and the behaviour of all staff are being discussed openly. This approach Procedures Training workerFIG 3. Safety culture aspects at defence in depth forms the basis for our safety culture in which in addition to our customers the public, environment and the competent authorities are seen as our stake-holders too. Through continuous improvement NRG strives to be an excellent organisation with an excellent safety culture. IAEA-CN-156/S-52 Safety Management and Effective Utilization of Indian Research Reactors APSARA, CIRUS and DHRUVA D.K.Shukla Bhabha Atomic Research Centre (BARC), Mumbai, India dkshukla@barc.gov.in Research Reactors Apsara, Cirus and Dhruva are located at Bhabha Atomic Research Centre, Mumbai, India. Apsara, a swimming pool type reactor with maximum design power of 1MWth is in operation since 1956. Cirus (40MWth) and Dhruva (100MWth), both tank type reactors were commissioned in 1960 and 1985 respectively. These research reactors have played very important role in developing India’s nuclear energy programme and in establishing the safety basis for the related activities. Due to diversity in their design, operating power levels and utilization aspects, a slightly flexible and graded approach is necessary for managing safety in these research reactors. While the major safety aspect for research reactors are similar to that of power reactors, special attention is necessary on account of several irradiations and experiments that are carried out in these facilities. Several good operating practices like special work permits for maintenance activities, issuance of written valve slips for effecting valve status changes, special procedures for non- routine activities, pile irradiation requests for irradiation of samples, close attention to chemistry of fluid systems, prompt reporting of faults and incidents etc. have been practiced over the years in these reactors. Following good operating practices meticulously results in development of good operating culture which in turn provides right environment for developing safety culture. It is well known that training is essential for development of good operating culture. Strong emphasis is therefore laid on formal training and qualification of personnel in research reactors right from their inception. These practices led to development of a strong safety culture wherein all plant personnel are conscious about safety importance of their actions and are proactive about maintaining and enhancing safety. In fact, research reactors laid the foundation of a strong safety environment in reactor operations and this trickled down to the nuclear power plants also. Track record of safety of research reactors in India has been excellent. Root cause analysis of every incident is carried out and appropriate corrective measures are implemented. Detailed surveillance and In Service Inspection programmes have been evolved based on specific reactor design and operating experience and are adhered to. Ageing of old reactors has been managed by systematic assessments and refurbishing actions. The refurbishing outage has been also utilized for making several safety upgrades to meet present day requirements, as in the case of refurbishment of the Cirus reactor. Safety improvements have been made on a continuing basis based on operating experience and new knowledge. At times, these improvements have gone beyond the requirements of design and safety analysis giving credence to the slogan AHARA “Safety – As High As Reasonably Achievable”. For ensuring continued safety during the operating life of the research reactors a well planned safety management system is in place. There exists a well defined hierarchical structure and line of communication among operating organization, regulatory agency, health and safety organization, maintenance and services organization and various safety committees / groups, IAEA-CN-156/S-52 quality groups, and experimenters, to facilitate smooth functioning of the operational activities of the research reactors at BARC. The paper would outline the safety management system practiced in these research reactors. All the three research reactors have been well utilised for basic and applied research, neutron radiography, nuclear detectors testing, radioisotope production, material testing and human resource training and development. The National Facility for Neutron Beam Research (NFNBR) has been created at BARC during early nineties to cater to the needs of the Indian scientific community. Scientists from, universities and national laboratories also use these facilities in research reactors through collaborative research projects. Many of these collaborations are being supported by University Grant Commission - DAE Consortium for Scientific Research (UGC-DAE CSR), Board of Research in Nuclear Sciences (BRNS), and other agencies. Besides conventional uses, these research reactors have been utilised for conducting various engineering experiments also. A refurbishment and upgradation plan has been drawn up for Apsara which has completed 50 years of service. The refurbishment would incorporate reactor design modifications to obtain enhanced neutron flux and a larger irradiation volume. Reactor core would be converted to a LEU Silicide fuel core. The paper would describe some important utilisation of research reactors Apsara, Cirus and Dhruva, modifications incorporated and up gradation proposed to enhance utilisation. Our unique experience in developing ISI programme for Dhruva and Cirus will also be covered in brief . References: [1] S.K.Sharma, S.K.Agarwal and D.K.Shukla, “Safety of research reactors- An overview”, First national conference on nuclear technology (NRT-1), Mumbai, India (2002). [2] Bhabha Atomic Research Centre, “Operating Experience & Utilisation of High Flux Research Reactors at Trombay”, BARC brochure, Mumbai, India ( 1990 ). [3] D.K.Shukla, “Development of Surveillance and In Service Inspection Programme for Indian Research Reactors Cirus and Dhruva”, ICONE-14, USA (2006). IAEA-CN-156/S-45 Ageing Management of Pakistan Research React-o1r (PARR-1) Mujahid Latif 1), Showket Perve 1z), Muhammad Isra 1r) 1) Nuclear Engineering Divisio n Pakistan Institute of Nuclear iSecnce and Technology (PINSTEC H) Nilore, Islamabad Pakista n Email: mujahid@pinstech.org.p k Introduction to PARR -1 Pakistan Research Reac-1to (rPARR-1), a swimming pool, MTR type research reactor went critical on 21 December 1965 and attained full power of 5 MW on 22 June 1966 w %it h 93 Highly Enriched Uranium (HEU) fuel. The reoarc twas shut down in 1990 for core conversion to commercially availab0.1 MeV). As may be required, temperature is stabilized through changing thermal resistance in the heat supply circuit or intensifying heat removal with liquid boiling metal. These devices make it possible to provide a predetermined azimuthal temperature non-uniformity and temperature non-uniformity along the height. These experimental devices can be used for irradiation of fuel elements down to 15 mm in diameter, which are enclosed in different arrays (triangular, square etc.) and environments (sodium, lithium, lead, different gases etc.). The available pilot reprocessing facility of irradiated fuel and production facilities of fuel and fuel elements for fast reactors make it possible to develop a closed fuel cycle. The Institute performs examinations and irradiation tests of refabricated fuel with involvement of minor actinides and long-lived fission products in the closed fuel cycle. The efficient operation of the BOR-60 reactor as a burner of minor actinides as well as power-grade and weapon-grade plutonium has been demonstrated. Taken together these factors allow for solving important tasks on reduction in value of fuel factor, decreasing the amount of radioactive waste and improving the ecological situation in the nuclear power engineering. The BOR-60 reactor made it possible to obtain a wide diversity of experimental data on irradiation of oxide, metal, ceramic-metal, carbide and nitride fuel compositions for reactors of different applications, in particular for fast reactors with sodium coolant. Regularities of gas release, deformation and structure formation were studied. The obtained results made it possible to justify the usage of fuel elements for fast reactors BN-350 and BN-600 as well as other types of reactors. Uranium pellet fuel was used in the reactor since 1969 and then in 1981 it was converted to vibropac mixed oxide fuel with the use of power-grade and even weapon-grade plutonium in the last years. So it is possible to perform a large-batch testing of different types of fuel. A special loop-capsule was used for vibropac fuel testing under emergency conditions until fuel element failure. These tests proved the possibility of fuel testing not only under steady state and transient but emergency conditions also. The results obtained during examination of different fuel compositions form the basis for fuel cycle development of advanced fast reactors with the enhanced safety. Among them are the BREST-OD-300 reactor with the nitride fuel and other advanced reactors. The use of carbide and IAEA-CN-156/U-20 nitride fuel that possess high fissile atom density and high thermal conductivity as opposed to the oxide one allows for increasing thermal loads and safety indexes significantly. A loop device was used for performance of the first test runs of the BREST fuel elements in lead at the irradiation test parameters similar to the design ones. A burnup of 0.5 % h.a. was achieved. Currently the same fuel elements are being irradiated as a part of fuel assemblies cooled with sodium to enhance representativeness of the results. The achieved fuel burnup is 2.2 % h.a. and damage dose - 23 d.p.a. in steel. Nowadays the reactor continuous its successful operation and thus provides for performance of a large scope of scientific research. The Institute holds a license for its operation till 2010. Work is being done to extend the service life of the reactor till 2015. In doing so there are preconditions for continuation and expansion of activities in the field of fuel cycle justification of advanced reactors. Session 6: Research Reactor and Network Corporation Synopses no. IAEA-CN- Synopses Title Main Author 156/ Developing Research Reactor Coalitions and Centres U-25 of Excellence Goldman, I.N. IAEA’s Subprogramme on Research Reactors: U-44 Technology and Non-Proliferation Adelfang, P. Research Reactor Utilization in the Mediterranean Region - the Experience of Montenegro and U-46 Possibilities for Future Cooperation Jovanovic, S. Role of the Oarai Branch as a facility open for university researchers in utilization of research U-16 reactors Shikama, T.T. IAEA-CN-156/U-25 DEVELOPING RESEARCH REACTOR COALITIONS AND CENTRES OF EXCELLENCE IRA N. GOLDMAN1), PABLO ADELFANG1), KEVIN ALLDRED2), NIGEL MOTE2) 1)Department of Nuclear Energy, International Atomic Energy Agency (IAEA) Wagramer Strasse – 5, P.O. Box 100, A-1400 Vienna, Austria 2)Alldred Associates, LLC, 322 Kent Road, New Milford, CT 06776, USA 3)International Nuclear Consultants, Inc., 415 Mikasa Drive, Alpharetta, GA 30022, USA E-mail address of main author:I.goldman@iaea.org Research reactors continue to play a key role in the development of peaceful uses of atomic energy. They are used for a variety of purposes such as education and training, production of medical and industrial isotopes, non-destructive testing, analytical studies, modification of materials, for research in physics, biology and materials science, and in support of nuclear power programmes. The IAEA Research Reactor Data Base lists about 250 operational research reactors worldwide, many of which have been operating for more than 40 years. Through both statistical and anecdotal evidence, it is clear that many of these reactors are underutilized, face critical issues related to sustainability, and must make important decisions concerning future operation. These challenges are occurring in the context of increased concerns over global non-proliferation and nuclear material security, due to which research reactor operators are coming under increased pressure to substantially improve physical security and convert to the use of low enriched uranium (LEU) fuel. Thus, there is a complex environment for research reactors, and one in which underutilized and therefore likely poorly funded facilities invoke particular concern. Many research reactors are challenged to generate sufficient income to offset operational costs, often in a context of declining political and/or public support. Many research reactor operators have limited access to potential customers for their services and are not familiar with the business planning concepts needed to secure additional commercial revenues or governmental or international programme funding. This not only results in reduced income for the facilities involved, but sometimes also in research reactor services priced below full cost, preventing recovery of back-end costs and creating unsustainable market norms. Parochial attitudes and competitive behaviour restrict information sharing, dissemination of best practices, and mutual support that could otherwise result in a coordinated approach to market development, building upon strengths of various facilities. Moreover, belief that the markets for research reactor products and services are a “zero-sum” game, with market gains by one research reactor coming at the expense of another facility, result in a general lack of openness within the research reactor community. Yet there is evidence to suggest that the market for research reactor services is supply limited, rather than demand limited. A number of factors limit the ability of research reactors to expand their user base and to generate new sources of revenue: IAEA-CN-156/U-25 • Many potential customers do not know how, or where, to contact the research reactor community, and have only limited knowledge or awareness of the range of research reactor services, equipment and locations available. • The standards of quality control and quality assurance between research reactors are not uniform, impede business development, and may result in a lack of confidence in service reliability. As a consequence, customers need to conduct due diligence for each facility to be used, reducing the enthusiasm and financial rationale for developing additional sources of supply. • Transport of radionuclides is becoming increasingly difficult, with examples of shipments held in customs, prevented from leaving the country of origin or from entering the customer destination, and requires specific expertise and experience to manage this issue. In order to address the complex of issues related to sustainability, security, and non- proliferation aspects of research reactors, and to promote international and regional cooperation, the IAEA is initiating the Research Reactor Coalitions and Centres of Excellence initiative. This activity is supported by a two-year grant from the Nuclear Threat Initiative, Inc. (NTI), and by a 2007-2008 IAEA Technical Cooperation Project, “Enhancement of the Sustainability of Research Reactors and their Safe Operation Through Regional Cooperation, Networking, and Coalitions” (RER/4/029). These two activities will work in an integrated manner, along with other relevant national and regional IAEA Technical Cooperation projects and complementary IAEA regular and extra-budgetary funded programme activities in research reactor utilization, safety, security, and the fuel cycle. These activities were endorsed by the IAEA Board of Governors in its March 2007 meeting which encouraged regional cooperation and networking among research reactors. The aim of this initiative will be to establish a pilot project involving the formation of at least one voluntary, subscription-based, self-financed coalition of research reactor operators (possibly including other participants, sponsors, etc.), which may serve as a model for the establishment of additional coalitions. IAEA-CN-156/U-44 IAEA’s Subprogramme on Research Reactors: Technology and Non-Proliferation P. Adelfang1), S.K. Paranjpe2), I.N. Goldman1), A.J. Soares1), E.E. Bradley1) 1) Research Reactors Unit, Division of Nuclear Fuel Cycle and Waste Technology, International Atomic Energy Agency (IAEA), Vienna, Austria 2) Physics Section, Division of Physical and Chemical Sciences, International Atomic Energy Agency (IAEA), Vienna, Austria E-mail address of main author: P.adelfang@iaea.org For nuclear research and technology development to continue to advance, research reactors (RRs) must be safely and reliably operated, adequately utilized, refurbished when necessary, provided with adequate proliferation-resistant fuel cycle services and safely decommissioned at the end of life. The IAEA has established its competence in the area of RRs with a long history of assistance to Member States in improving their utilization, by taking the lead in the development of norms and codes of good practice for all aspects of the nuclear fuel cycle and in the planning and implementation of decommissioning. The IAEA Subprogramme on RRs is formulated to cover a broad range of RR issues and to promote the continued development of scientific research and technological development using RRs. Member States look to the IAEA for coordination of the worldwide effort in this area and for help in solving specific problems. The IAEA coordinates and implements an array of activities that together provide broad support for RRs. As with other aspects of nuclear technology, RR activities within the IAEA are spread through diverse groups in different Departments. To ensure a common approach a Cross-Cutting Coordination Group on Research Reactors (CCCGRR) has been established, with representatives from all departments actively supporting RR activities. Utilization and application activities are generally lead from within the Department of Nuclear Applications (NA). With respect to RRs, NA is primarily carrying out IAEA activities to assist and advise Member States in assessing their needs for research and development in the nuclear sciences, as well in supporting their activities in specific fields. Safety and Security aspects of RRs are handled by the Department of Nuclear Safety and Security (NS). The technological, fuel cycle and operational aspects of RR management are supported by the Department of Nuclear Energy (NE). NE is primarily working to support RR organizations in their pursuit of often diverse strategic objectives within the context of modern RR operational constraints. Today RR operating organizations must overcome challenges such as the ongoing management of ageing facilities, pressures for increase vigilance with respect to non proliferation, and shrinking resources (financial as well as human) while fulfilling an expanding role in support of nuclear technology development. IAEA-CN-156/U-44 In addition, the Department of Nuclear Safeguards is responsible for the control of the fissile material for RR and the Department of Technical Cooperation (TC) supports RR activities for the principal benefit of RRs in developing countries. TC is subsequently supported by NA, NS, and NE who assist in the development and implementation of relevant TC projects within their specific fields of expertise. The Subprogramme on RRs is under IAEA’s Programme D on Nuclear Science. Implementation of the IAEA Subprogramme on RRs (IAEA code D.2) is shared between NE and NA while separate subprogrammes, managed by NS, deal with RR safety and security. In this paper, only the activities managed by NE and NA under the subprogramme on RRs are presented, including a complete description of the ongoing projects and planned activities for the years 2008-2009. Special emphasis is put on new international collaborative undertakings, like the IAEA’s Technical Working Group on RRs. [1] IAEA, “The Agency’s Programme and Budget 2006–2007”, Printed by the International Atomic Energy Agency, July 2005. [2] IAEA, “The Agency’s Draft Programme and Budget 2008–2009”. Submission to the Board of Governors, February 2007. [3] I. N. Goldman, P. Adelfang and I. G. Ritchie, “IAEA Activities Related to Research Reactor Fuel Conversion and Spent Fuel Return Programmes”. Proceedings of the XXVI International Meeting on Reduced Enrichment for Research and Test Reactors, Vienna, Austria, November 7-12, 2004. [4] P. Adelfang and I. N. Goldman, “Latest IAEA Activities Related to Research Reactor Conversion and Fuel Return Programmes”. Proceedings of the XXVII International Meeting on Reduced Enrichment for Research and Test Reactors, Boston, USA, November 6-10, 2005. [5] I. N. Goldman, N. Ramamoorthy, and P. Adelfang, “The IAEA Coordinated Research Project: Production of Mo-99 Using LEU Fission or Neutron Activation. Proceedings of the XXVII International Meeting on Reduced Enrichment for Research and Test Reactors, Boston, USA, November 6-10, 2005. [6] P. Adelfang, I. N. Goldman, A. J. Soares and E. E. Bradley, “Status and Progress of IAEA Activities on Research Reactor Conversion and Spent Fuel Return Programmes in the Years 2005-2006”. Proceedings of the XXVIII International Meeting on Reduced Enrichment for Research and Test Reactors, Cape Town, South Africa, October 29 - November 2, 2006. [7] I. N. Goldman, N. Ramamoorthy, and P. Adelfang, “Progress in the IAEA Coordinated Research Project: Production of Mo-99 Using LEU Fission or Neutron Activation”. Proceedings of the XXVIII International Meeting on Reduced Enrichment for Research and Test Reactors, Cape Town, South Africa, October 29 - November 2, 2006. IAEA-CN-156/U-46 Research reactor utilization tinhe Mediter ranean region The experiecne of Montenegro and possibilities for future coopera tion S. Jovanovic University of Montenegro, Faculty of Sciences, Department of Physics, P.O.Box 211, Cetinjski put bb, MNE-81000 Podgorica, Montenegro, and Centre for Eco-Toxicological Research of Montenegro, Department of Radiation Protection and Monitoring, P.O.Box 374, Put Radomira Ivanovica 2, MNE-81000 Podgorica, Montenegro Being a small developing country, Montenegroh aisr dly advanced when research in nuclear sciences is concerned. The latter is limitedt wto institutions: Centre for Eco-Toxicological Research of Montenegro, Department ofd iRataion Protection and Monitoring (CETI) and University of Montenegro, Faculty of ScienceDse,p artment of Physics (FS), both in Podgorica, the capital. Strange enough, evtehno ugh there has never been ae arercsh reactor in the country, there is quite some tradition and experience wRiRth utilization, particularl ywith reactor neutron activation analysis (RNAA). Namlye, already from the years 70’s onwards, physicists from FS were educated&trained in prominent nucleasre raerch centers in former Yugoslavia (Vinca NRI near Belgrade; Jozef StefaNnR I, Ljubljana, Rudjer Boskovic NRI, Zagreb) and abroad (University of Ghent NR IB, elgium; Dubna NRI near MoscowU;n iversity of Charlottesville RR, USA; NIST, Gaithersburg, USA; Rez NRI eanr Prague; Saclay NRI near Paris, Forschunszentrum Garching near Munchen, Germany, etc.). Quite extensive interantional scientific ocoperation existed on rviaous RNAA topics under national and international resehr cprojects: environmental potlliuon, geological, pedological and metallurgical studies, reactor unteron flux characteriztaion, RNAA nuclear dat astandardization, k0-RNAA method development, semiconductodre tector gamma-spectrometry software optimization, true coincidence effects, c.e t- the output being counted by hundreds of publications, reports, studies, Ph.D. theses, Seotcm. e of them are listed below [1-11]. In addition, ANGLE software for semiconductor tedcetor gamma-efficiency calculations, developed by nuclear spectrometry group at Uthneiv ersity in Podgorica, Montenegro, is nowdays in use in numerous gamma spemcetrtory (including NAA)l aboratories all around. Unfortunately, these activities practically colslaepd in the 90’s, following long lasting political and economical crisis in the country. Recentlyrm, feor research cooperoanti links are being re- established and new ones created, to mentisotn thjue two in the Balkan region involving RR utilization: with “Jozef Stefan” Institute,L jubljana, Slovenia, Environmental Sciences Department, k0-RNAA group (topic: gammpae-sctrometry software upgrading) and “Demokritos” Research Institute, RR Laboorrayt, Athens, Greece (topic: bulky samples RNAA characterization). There is also a formal cootpioenra agreement with Nuecal r Research Institute of the Bulgarian Academy of Sciences, Sao, fiBulgaria (topic: environmental pollution measurements and monitoring, howeverR nRo utilization included up to now). IAEA-CN-156/U-46 Experience of Montenegro in research reacuttoilriz ation through interantional cooperation, in particular in the Balkan/Mediterranean regiwona,s presented during IAEA Technical Meeting on Strategic Planning for Sustainability-Meditenrera n Region - Research Reactor Utilization, Vienna, March 19-22, 2007. On t hoiscassion Monteneg rwoas pointed out as a good example of how a “no-RR country” could incorporate weinll to RR community. It was subsequently proposed that Montenegro coordinates the activities on initiafotirnmgu/ lating a regional (mediterranean) IAEA project on unteron activation analysis seclted topics. The latter will include i.a. preparation of reference materials, irradiation under controlled conditions, measurements, software assessment and idnatetarp retation. Applictaions in areas like Environment, Geology, Health, Agriculture & Curlatul Heritage are particularly emphasised. REFERENCES [1] A.SIMONITS, S.JOVANOVIC, F.DE CORTE, L.MOENS, J.HOSTE, "A method for experimental determination of effective resonance energies related to (n,gamma) reactions", J.Radioanal.Nucl.Chem., 821/16 9(1-197894.) [2] S.JOVANOVIC, F.DE CORTE, L.MOENS, A.SIMONITS, J.HOSTE, "Some elucidations to the concept of the effective resonance energy Er", J.Radioanal.Nucl.Chem., 82/2 (1984) 379-383. [3] F.DE CORTE, S.JOVANOVIC, AS.IMONITS, L.MOENS, J.HOSTE, "Determination of the essential flux parameters for (n,gamma) activation analysis - a new methodology applieeda ctoto r r Thetis (Belgium)", Atomkernenergie - Kerntechnic, 44 (1984) 641-647. [4] D.DRAGOVIC, P.VUKOTIC, S.JOVANOVIC, "Lithological types and distribution of rare earth elements in white bauxite deposits of Montenegro", Travaux ICSOBA, JAZU, 14-15 (1985) 77-82. [5] S.JOVANOVIC, F.DE CORTE, A.SIMONITS, L.MOENS, P.VUKOTIC, J.HOSTE, "The effective resonance energy as a parameter in (n,gamma) activation analysi sr ewaicthtor neutrons", J.Radinoaal.Nucl.Chem., Articles, 113 (1987) 177-185. [6] B.SMODIS, R.JACIMOVIC, S.JOVANOVIC, P.STEGNA RP,.VUKOTIC, "Efficiency characterisation of HPGe detectors for use in the k0-method of neutron activation analysis", Vestn.Slov. Kem. Drus.9, 8385)/4 3 (917-408. [7] S.JOVANOVIC, B.SMODIS, R.JACIMOVIC, P.VUKOTIC, P.STEGNAR, "True coincidence corrections and related peak-to-total ratio measurements of HPGe detectors for use in the k0-method of NAA", Vestn.Slov.Kem.Drus., 35/4 (1988) 409-424. [8] S.JOVANOVIC, B.SMODIS, R.JACIMOVIC, P.VUKOTIC, P.STEGNAR, "Neutron flux variability at the TRIGA Mark II reactor, Ljubljana, as a parameter whaepnp lying the k0-method of ANA", J.Radioanal.Nucl.Chem., Letters, 135 (1989) 59-65. [9] F.DE CORTE, A.SIMONITS, F.BELLEMANS, M.C.FREITAS, S.JOVANOVIC,B.SMODIS, G.ERDTMANN, H.PETRI, A.DE WISPERAERE, "Recent advances in the k0-standardization of neutron activation analysis: extensions, applications, prospects", J.Radioanal.Nucl. Chem., Articles, 169 (1993) 125-158. [10] S. JOVANOVIC, F. CARROT, C D. ESCHAMPS, N. DESCHAMPS, P.VUKTOIC, "A study of the air pollution in the surroundings of an aluminium smelter, usingip heyptic and lithophytic lichens", Journal of Trace and Microprobe Techniques, 13(4), 463-471 (1995) [11] S.JOVANOVIC, A.DLABAC, N.MIHALJEVIC, P.VUKOTIC, "ANGLE - a PC-code for semiconductor dete ctor efficiency calculations", J.Radioanal.Nucl.Chem., 218 (1997) 13-20 IAEA-CN-156/U-16 Role of the Oarai Branch as a facility open for university researchers in utilization of research reactors Tatsuo Shikama and Minoru Narui* Institute for Materials Research, Tohoku University 2-1-1 Katahira, Aobaku, Sendai, 980-8577 Japan The Oarai Branch, Institute for Materials Research, Tohoku University Naritacho, Oarai, Higashiibariagkun, Ibarakiken, 311-1313 Japan Correspondence; Tatsuo Shikama, Shikama@imr.tohoku.ac.jp Research reactors incarnate an invaluatbhlea ter for materials irradiation, from fundamental points of views, namely, itsla rteively homogeneous fludxi stributions, long and stable operations, controllabilityo f irradiation atmospheres, and an electronic-excitation-and-nuclear-displacemernatt io relevant to actual application conditions. In the meantime, researcheisli zuintg fission reactors are expensive and time&manpower consuming, which could befo ardfable only for a long-term and large scale project. The Oarai Branch of Instit ufoter Materials Research in Tohoku University (Hereafter denoted as the Oarai Branch) bheaesn acting as a codoinr ator for university researchers’ utilizing fisosni reactors for their fundamnteal studies, under the close collaboration with Japan Atomic Energy Agen (cJyAEA). Related research fields extend into a variety of topics such as cosmocloagl iand geological age teimsation, detection of a trace amount of actinide elements ing hhlyi pure semi-conductors, materials issues related with safety and extension of e lifof water cooledp ower reactors, and developments of materials for advanceda crteors including a nxet generation fission reactors and nuclear fusion reactors. The pwapilel re port a role of the Oarai Branch in utilizing research and test retoarcs and will describe preste sntatus of university-related research activities utilizinge ractors. A variety of researecrhs in universities have a variety of interests in utilizing research reactors such as Japan Materials Testing Reactor (JMTR; water cooled mixed spectra rear)c taond JOYO (sodium cooled high flux fast reactor). Universit yresearchers of about 2000-3000 nm daays are visiting the Oarai Branch every year to carry o tuhteir own research projects. For utilization of research reactors fourn fdamental studies, somirera diation techniques and intimate linkages between reactors and iprroasdti ation facilities(PIFs) are needed to develop. The JAEA is planning to estahb lias comprehensive framework for advanced reactor-irradiation-studies in the Oaraie a,r whose details will be described in a separate paper. In this framework, the OaBrrain ch is planning top lay an unique role not only for convenience of university researcsh beurt also for some large-scale national IAEA-CN-156/U-16 projects which will be mangaed mainly by the JAEA andn di ustries. Typical Examples are prompt transportation of irradiatede csipmens to PIFs, and prompt implementation of post irradiation examinations (PIEs)h.o Srtening of an iteration period from specimen preparation to implementation of PIEs is enstisael for the university researchers, where education of post-graduate students in an deedf iperiod is one among major targets. To realize this, legitimate and propmt transfer of irradiated sepcimens from the JAEA to the Oarai Branch is established with neededft wsaore and hardware. Examples will be a development of shuttle irradiation rig fothre JOYO irradiation. Other roles which the Oarai Branch tries to take an initiativee adrevelopment advanceadn alyzing instruments for studies of nano-scale evuotilon of radiation inducedm icrostructural changes and development of advanced irradiation technsiq. uFeinally, it is essential for fundamental studies of university researchers to conitrrroal diation conditions such as a temperature and an atmosphere, independently ofr ea ctor operating mode and to monitor irradiation conditions and som peroperty changes of matersia ul nder irradiation in-situ. By 1980s, no-instrumented rigs were dominfaonr ts imple irradiation of materials with possible neutron fluences. In late 19,8 0its was seriously recognized through accumulated studies that well croonllted reactor irradiation iisnevitable to make reactor irradiation studies compatible with advan cmedaterial science. Since then, specially designed irradiation rigs were developed. One ammonogs t important topics is a temperature control being independeonft a reactor operation mode. Usually, temperatures of irradiated specimens wseerreio usly influenced bya nuclear heating (a gamma-ray dose rate), which is dependoen t a reactor power. So, an irradiation temperature will vary when a reactor power ievsa, especially at transient periods of reactor operations such as a startup ansdh uatd own. It was cleayr l revealed that the variation of irradiation temperature wilrle sult in very compcliated evolution of radiation induced microstrutucres, which could not be rationally understood from fundamental points of view. In-situ measureenmts of some properties of irradiated materials under reactor irradiation will be another topic. Electrical and optical measurements were realized in JMTR, weh reardiation induced ecl trical conductivity (RIC) of ceramic insulators were systaetmically measured and radiation induced luminescence from ceramics was measurerodu tghh radiation resistant optical fibers during JMTR irradiation. Now, the JMTR is under refurbishing and a new framework is under survey for its utilization more effective and more conventi for customers. The Oarai Branch is planning to reorganize its role in utilizings reearch reactors for university researchers under close collaboration with JAEA and university-research-network for utilizing research reactors. Session 7: Regulatory Aspects and Experience with Current Research Reactor Issues Including Safety Aspects of Core Conversion Synopses no. IAEA-CN- Synopses Title Main Author 156/ Canadian Experience in Implementing Modern S-29 Regulations to Existing Research Reactors Alwani, A. The Role of Regulatory Authority in Safe Operation S-31 of Research Reactors Mikulski, A.T. Causal Factors Guide For The Evaluation Of S-33 Accidents In Research Reactors Perrin, C.D. Contribution of Research Reactors to the Programmes for Research and Technological Development on the U-30 Safety Couturier, J. Review of the ANSTO Application for a Facility Licence to Operate the OPAL Research Reactor in S-59 Australia: Case study review of operational readiness Ward, J.S. S-38 Research Reactors in Germany: An Overview Schneider, M. The French approach for the regulation of research S-42 reactors Conte, D. S-30 MARIA research reactor conversion to LEU fuel Krzysztoszek, G. IA E A-CN-156/S-29 Canadian Exper ience in Implementing Modern Regulations to Existing Research Reactors A. Alwani Canadian Nuclear Safety Commission, Ottawa, Canada abdul.alwani@cnsc-ccsn.gc. ca Research reactors in Canada are regu lbayte tdhe Canadian Nuclear Safety Commission (CNSC). The CNSC is an independent agency of the Government of Canada and operates in a transparent manner. The mission of the CNSC is to regulate the use of nuclear energy and materials to protect health, safety, security and the environment and to respect Canada's international commitments on the peaceful use of nuclear energy. Currently, there are seven research reactors in operation in Canada: The National Research Universal (NRU) reactor, 135 MW; McMaster Neuaclr Reactor (MNR), 5 MW, and five Safe Low-Power Kritical Experiment (SLOWPOKE) reactors, 20 kW. This is in addition to two Multipurpose Applied Physics Lattice Experiment (MAPLE) reactors, 10 MW which are in the commissioning stage. The operating research reactors in Canada ah alovneg history of operation and established safety track records. The NRU was build in 1957, the MNR in 1959 and the SLOWPOKEs in the seventies. The Atomic Energy Control Board, the predecessor of the CNSC, was the national regulatory body in Canada since 1946 and regulated the research reactors among other nuclear facilities and materials for decades. In May 2000, the Nuclear Safety and Control Act replaced the Atomic Energy Control Act of 1946 with a modern statute that reflects public expectations for the regulation of nuclear energy. The new Act provides for more explicit and effective regulation of nuclear activities. It ensures high standards in the areas of health, safety, security and protection of the environment, sustaining our environment and for ensuring a modern regulatory regime to meet the needs of the 21st century. While the licensing regime for nuclear facilities including research reactors continues with the new Act, several changes occurred to the requirements and the process. Both the licensee and the regulator made particular effortsa tcoe f the challenges of defining the implications of the new requirements, managing the expectations, and bringing the licensed activities up to compliance with the new standards. One main aspect of the modern regulatory practice of the CNSC is to focus on programs to be in place to address safety areas. Examples of these areas are training, criticality safety, and fire protection. These aspects have always been subject to a continuous regulatory oversight. However, the CNSC now expects the research reactor licensees to have formalized documented programs in place to address these aspects. Several regulatory guidance and standard documents have been published since 2000 and which provide recommended or approved methods to formalize the safety area programs. IAEA-CN-156/S-29 Another change with the modern regime is the requirement that research reactor licensees, among other large facilities’ licensees, establish financial guarantees to ensure that costs of decommissioning their nuclear facilities, here research reactors, are borne by the licensees and not by the taxpayers. Quality Management in another area of new requirements formally established with the new Act and the regulations pursuant to the Act. The requirements to have a quality management system in place are now standard to all research reactor licensees. Associated with it are incident investigation process and capabilities as well as human factors engineering reviews. The implementation of the modern regime presented a number of challenges to the research reactor facilities. These facilities have usually small organizations and fewer resources to devote to address the new demands. Also, with the mature operations and the established good performance and safety records, the new requirements may be perceived as an added burden without additional safety benefits. The CNSC approach is characterized by the following: ‚ Risk Informed Approach: Many CNSC requirements were originally designed for nuclear power reactors. They have been tailored to research reactor to take into account the smaller risks imposed by the research reactor facilities. The graded risk approach has been used to define both the applicability of specific requirements and the acceptance criteria. ‚ Transitional Periods: The CNSC recognizes the need for transitional periods for rolling in and implementing the various new requirements. Transitional periods for full compliance with each requirement aimrep osed after careful assessment of the priority and feasibility of the required action. This is to ensure that the upgrade to the current standard is done within a reasonable time frame and also effectively. ‚ Compliance Promotion: As expected in any change, there is a need to communicate fully with the licensees and provide the rationale for each new requirement. The notion that issuance of a new requirement by the CNSC does not mean that what was safe yesterday is not safe today, is explained fully to the licensees. All is aimed toward a common goal of ensuring that the Canadian worker, public and environment are protected from undue risk from research reactors by adapting the best safety standards. IAEA-CN-156/S-31 The Role of Regulatory Author ity iSn afe Operation of Research Reactors A.T. Mikulski National Atomic Energy Agency (PAA), Warsaw (Poland) E-mail address: mikulski@paa.gov.pl To be presented at International Conference on Research Reactors: Safe Management and Effective Utilization 5-9 November 2007, Sydney, Australia EXTENDED SYNOPSIS The safe and secure operation of research reactors depends on several factors. One of them is the activity of local (country) regulatory authitoy.r It is not enough to follow the well known international guidance, e.g. the IAEA Codoef Conduct for research reactors, Safety Guidance, etc. but training and experience sopf einctors contributes significantly in this field. The activity of regulatory activity should be concentrated in many directions, as follows: (1) issuing licences for reactor operation, (2) giving permission for technological or procedural changes in operating instructions, (3) requiring and/or authorizing technical modifications, (4) verification of reports submitted by operating staff, (5) performing own analysis also by invited specialists or experts. The paper summarises the activity of Polish regulatory authority in this field during the last five years and is referred to the MARIA resceha rreactor. The reactor is used mainly for isotope production and experiments in nuclear phsy. s Iict is in operation since more than 30 years and undergoes systematic improvemeInt tis. o perated on a one-week long cycles with two longer breaks per year for regular overhamualsd e by reactor staff. Regulatory authority requires detailed plan of the work to be pemrfeodr and verifies results before starting of new period of reactor operation. The reactor undergoes several modification and technical improvements. They are prepared by reactor staff and approved by regulatory authority. In case of more complicated improvements or requiring detailed analysisp, eecsially numerical calculations, they are analysed by independent experts, mainly f rtoemchnical universities. The most important examples of such improvements are: (a) transformation from HEU to MEU fuel r(ofm 80 to 36% enrichment, which were performed in years 1999-2003 and required installation of a special fuel assembly equipped with thermocouples in order ptoe rformed verification of thermo-hydraulic calculations), (b) preparation for transformation to LEU fueel s(ls than 20% enrichment, to be started in year 2007), (c) agreement for decrease number of heat exchrsa ning efuel channel cooling system (from 4 to 3 exchangers in a case when one is leaking), IAEA-CN-156/S-31 (d) restoring of cooling pipeso cnnecting two fuel channels (dismantled many years ago in order to place a special test loop), (e) modernization of dosimetric system, (f) installation of new system for measurement of technological parameters, (g) preparation of new locations in a reactor ec ofor isotope irradiation in higher neutron flux, (h) improving of cooling conditions for a natul rcaonvection during a decay heat removal from fuel elelemnts, etc. The other role of regulatory authority iusg sgesting technological changes and improvements in safety equipment, to be done in future: masodernization of vibration monitoring system, development of more friendly procedures fosru vailisation of information stored by signalling system, etc. The authority has some fundr st hfois activity, which are coming from a state budget and granted once a year. According to Polish Atomic Law, reactor operator is obliged to present quarterly reports of reactor operation and send an immediate incoatifon in case of any abnormal situation, including shortening of operation cycle. Theresep orts are carefully analysed and in many cases used as suggestions for improvemeinn tso perating instruction or installing new equipment. The reactor operator is obligeds utob mit the actual version of a Safety Analysis Report before applying for a new operting licenwchei,c h is granted for period specified in an appl;ication and this period was equal to five years in the past. All these problems will be described in det ahial,ving in mind an experience of almost 30 years of the MARIA reactor operation in Pnodla and may be of interest for regulatory authorities in other countries. IAEA-CN-156/S-33 Causal Factors Guide For The Evaluation Of Accidents In Research Reactors 1) Carlos Dante Perrin 1) Research Reactors and Critical Assemblies, Nuclear Regulatory Authority, Argentina E-mail address of main author: cperrin@sede.arn.gov.ar ABSTRACT In the field of radiological and nuclear safety, the Nuclear Regulatory Authority (ARN) of Argentina controls three research reactors and three critical assemblies, by means of evaluations, audits and inspections, in order to assure the fulfillment of the requirements established in the Licenses, in the regulatory standards and in the mandatory documentation in general. From the Nuclear Regulatory Authority point of view, within the general process of research reactors safety management, the management of operating experience plays an outstanding roll. In this aspect the ARN has established specific requisites in the Operation Licences in relation to the communication, evaluation, investigation of causes, and adoption of corrective measures, for the happened events. From the experience collected in the analysis of the reports sent by the operators it has been verified some weakness in relation to the methodology of analysis of events and in the determination of the causal factors. In such a sense, with the purpose to establish a help for the analysts and to homogenize the treatment of the events, two reference guides were designed: a guide for the evaluation of events and another with a grid of causal factors This paper describes the main aspects of the operating management system for research reactors and critical assemblies in Argentina, and the guides developed for the event analysis and determination of causal factors. IAEA-CN-156/U-30 Contr ibution of research reactors t thoe programmes for research and technological development on the safety Jean Couturie1r) and al . 1) Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Fontenay-aux-Roses, France E-mail address of main author: jean.couturier@irs n.fr The most significant current use of nuclear roerasc tis for the generation of electrical power, for instance with pressurized water react(oPrsW R) which constitute the most commercial type of reactor. Nevertheless, since the begnign noif nuclear activities, important applications of nuclear reactors are research applicationnclsu ding: the providing of neutrons source for the basic research (beamline experimentse) , dtehvelopment of nuclear technology, and of new types of reactors, for example to bring solutions of the radioactive waste management. Moreover, nuclear reactors are usedr ftoraining, nuclear propulsion, production of radioisotopes for medicine and industry. Therefore, the research reactors constitumtea jor equipments to the scientific and technological research. Research is of benfoerf itb oth reactors currently in operation and for future reactors as the phenomena occurrdinugr ing a severe accident are the same. Nevertheless, if severe accidents were notn t ainkteo account in the initial design of nuclear power plants currently in operation, improvemse nhtave been made in France in terms of prevention and limitation of consequences. The objective of the present article is to illutsetr athe contribution of the research reactors to the programmes for research and technological development led to increase the safety of the nuclear power plants, more particularly in tfhieeld of the accidents oifn sertion of reactivity (RIA), loss of primary coolant (LOCA) and ptiarl or total melting of the core; indeed, the physical phenomena occurring during a sevaecrcei dent are extremely complex. Research aims to better understand these phenomena and to reduce the associated uncertainties in order to assess the extent to which the state oqfu iarecd knowledge can be used to make reliable predictions. The examples will concern accidethnatst may occur in pressurized water reactors and in sodium fast reactors (SFR). Some given examples will show the fundamael nimt portance of experiments in research reactors to establish and support the safety demonstrations; indeed, such in-pile tests: ‚ allow to characterize badly known physical phenomena, ‚ can put in evidence some not foreseen phenomena, ‚ contributei n fine to the development and in the global validation of the codes used in the safety demonstrations, such as ASTEC (Adecncit Source Term Evaluation Code) used to simulate an entire accident from the initniagt ievent to the possible radionuclide release outside the containment. IAEA-CN-156/S-59 Regulatory Review Methods: Revie owf the ANSTO Application for a Facility Licence to Operate eth OPAL Research Reactor in Australia: Case study revwie of operational readiness Vince Diamond, Rhonda Evans and John Ward Australian Radiation Protectioann d Nuclear Safety Agency Email of main author: john.ward@arpansa.gov.au This paper will examine the review mheotds used by assessors of the Australian Radiation Protection and Nuclear SafAetgye ncy in advising the CEO of ARPANSA on the application by the Australian cNleuar Science and Technology Organisation (ANSTO) for a facility licence authorising it to operate the OPAL reactor. In particular it will focus on an aspect ofa tth review that considered the operational readiness demonstrated by ANS TinO relation to the OPALe ractor as an aspect of the review of safety of the OPAL reactor. The review of operational arediness examined the extent to which the managerial, procedural and administrative controlsa tt hwere proposed for the reactor were appropriate and the extent to which t hdeeymonstrated underlying support for the operation of the react.o r The review utilised guidance from Inteartnional Atomic Energy Agency documents on the operations of research reactors and also an internal ARPANSA regulatory guide that has been based on a review rorfe cnut international best practice in nuclear safety in this area. The review in paurtliacr focused on the adequacy and maturity of the information on matters such as effeec tcivontrol, safety management, radiation protection, radioactive waste managemelntitm, uate disposal, security and emergency planning in relation to the OPAL reactor. The paper will focus in particular on the manner in which each of these plans and arrangements were reviewed and ther aolvl eimportance that was placed on this review in the regulatory decision makipnrgo cess undertaken by ARPANSA. Key themes that will also be explored in the paper are: ‚ The development and implementation of ARPANSA regulatory guidance, its basis in key international guidance and its use in undertaking regulatory reviews; ‚ The importance of key safety management plans and the review of those plans in establishing regulatory confidence in the operator. ‚ The importance of open and questioning safety management policy and procedures, with the development osfe at of safety performance indicators (SPI) that can provide the regulatobroyd y with assurance of nuclear safety. ‚ The challenge of assurance for the regulator when systems are new and untested, and staffing profiles are under development Extended Synopsis IAIEAAE-CAN-C-1N56-/1S5-368/ Research Reactors in Germany: An Overview M. Schneider, Federal Office for Radiation Protection, Salzgitter, Germany E-mail address: m.schneider@bfs.de In Germany, substantial experience was gained in the field of research reactors during the last five decades. In this paper, an overview about the legislative and regulatory framework in Germany is given particularly with respect to research reactors, as well as a survey of the plant and licensing status of the facilities in Germany. A comprehensive legislative and regulatory framework is established to govern the safety of nuclear installations in Germany. The various nuclear safety regulations are structured hierarchically. At the top level, the Atomic Energy Act comprises the general national regulations for the safety of nuclear installations in Germany and constitutes the basis for the associated ordinances, e.g. the Radiation Protection Ordinance or the Nuclear Licensing Procedure Ordinance. The Atomic Energy Act and the ordinances are applicable to all kind of nuclear installations, and therefore are applied to nuclear power reactors as well as to research reactors in a common approach. After the amendment of the Atomic Energy Act in 2002, one of its purposes is to phase out the use of nuclear energy in a controlled manner. In fact, this is in force for nuclear power reactors for commercial generation of electricity, but not for research reactors. Below the legal level, the safety provisions and regulations of the Atomic Energy Act and its ordinances are put into concrete terms by general administrative provisions, by regulatory guidelines, by safety standards of the Nuclear Safety Standards Commission (KTA), by recommendations from the Reactor Safety Commission (RSK) and the Commission on Radiological Protection (SSK), and by conventional technical standards (e.g. DIN, ISO, IEC). The nuclear safety regulations concerning nuclear installations are in compliance with the international accepted safety standards, e.g. the “Safety Fundamentals” of the IAEA. The regulatory guidelines and the KTA Safety Standards are mainly developed for nuclear power plants. In practice, they are applied in analogy or with some interpretation for research reactors, in accordance with the potential hazards of the specific research reactor by means of a graded approach. There are only very few regulations implemented specifically for research reactors, e.g. two regulatory guidelines relating to the technical qualification of research reactor personnel or one KTA Safety Standard relating the monitoring of the discharge of radioactive substances from research reactors. Moreover, some recommendations from the RSK regarding specific licensing procedures of individual research reactor facilities have been made. In 2004, the IAEA Board of Governors approved the “Code on Conduct on the Safety of Research Reactors”. To comply with its recommendations, the German Federal Government included research reactors in the report for the third review meeting in April 2005 [1] for the “Convention on Nuclear Safety”. According to the Atomic Energy Act, a licence is required for the construction, operation or commissioning of research reactors. The licensing procedure and the continuous regulatory IAEA-CN-156/S-38 supervision of the facilities lies within the responsibility of the individual Federal States (“Länder”). To preserve the legal uniformity for the entire territory of the Federal Republic of Germany, the Federal Government supervises the licensing and supervisory activities of the “Länder”-authorities regarding lawfulness and expediency. In Germany, a total of 46 research reactors were built and operated. In the meanwhile, most of them are in decommissioning or have already been dismantled completely. Concerning the design, there is, or has been, a very broad range of different types of research reactors. The variety of facilities includes large pool or tank reactors with a thermal power of several tens of megawatt as well as small educational reactors with a thermal power in the order of only hundred milliwatts. At present, 13 research reactors are still in operation in Germany. The newest facility, the high flux neutron source FRM-II, became critical for the first time in March 2004 and started routine operation in April 2005. As an example, an overview about the licensing procedure and the specific plant characteristics of this facility will be given in the presentation. The actual decommissioning and dismantling projects comprise 9 research reactors. Furthermore, 24 facilities have been dismantled completely and the sites are released from regulatory control. Examples for the regulatory procedure of decommissioning projects will be given also in the presentation. [1] Federal Ministry for the Environment, Nature Conservation and Nuclear Safety, Convention on Nuclear Safety, Report by the Government of the Federal Republic of Germany for the Third Review Meeting in April 2005, Bonn (2004) IAEA-CN-156/S-42 The French approach for the regulation of research reactors D. Conte, A. Chevallier Autorité de sûreté nucléaire , Paris, France E-mail address of main author: dorothee.conte@asn.fr There are several types of research reactor currently in operation in France. Their usage includes neutronic studies, technological irradiations, neutron beam utilisation, safety research and teaching purposes. Most of these were built during the 60s and because they are all different each type presents particular hazards. The French Nuclear Safety Authority (ASN) works to ensure that the regulatory framework maintains a high level of safety for research and experimental activities. After a description of French research reactors and their hazards and operation this paper will summarise the French regulatory framework and its evolutions over the last few years. For instance, the law of June 13th 2006 on Transparency and Security in Nuclear Field is now the most important piece of French legislation in the field of nuclear safety. It builds upon the high requirements of the existing regulatory framework and, sets a new base for the control of nuclear activities and facilities. Also it creates an independent nuclear safety authority (ASN) and includes the principle of periodic (10 yearly) safety reviews for all nuclear facilities. The safety analysis and methods of control of research reactors safety have tended to become more and more similar in France to those of power reactors. For instance, even if a graduated approach has always been used, the safety analysis approach applied to operating conditions on research reactors is the same as that used for power reactors. Moreover, design codes used are often the same. The French regulatory framework is applicable to all nuclear facilities, including research reactors. To comply with the law and this regulatory framework, each facility must deliver a safety analysis report (SAR) which determines its particular operating limits. This analysis is assessed by ASN and its technical support organization, IRSN (Radioprotection and Nuclear Safety Institute). To ensure that the licensee assume all its responsibilities and, to allow the necessary flexibility in the ever changing operations of research reactors, ASN has utilised the principle of internal authorizations. Even if the operation is not explicitly described in the SAR, the licensee has the opportunity, under certain conditions, to self authorize the operation provided that it is of minor safety significance and is bounded by another operation in the SAR. This principle has been applied to the use of experimental devices in specific conditions. Since its introduction in 2002 the internal authorization process has generated a significant amount of plant operations information which has been subject to review and feedback by both the licensee and ASN. ASN is about to extend its application to core management the long stop periods, especially for refurbishment in research reactors. This paper then deals with the feedback provided over the last few years from events reported at French nuclear research reactors. There has been a significant increase in reported events between 1999 and 2006. Last year (2006) 29 events were reported on research reactors in France. Most of these were rated at level 0 and are minor events but this relatively high number is probably due to the ageing of many facilities. Nevertheless, it’s also the result of the introduction of new criteria for declaration for instance when an automatic shutdown IAEA-CN-156/S-42 occurs. However this increase in the number of events provides the opportunity for analysis to acquire more knowledge about research reactors operation and also an opportunity to tackle the subject of the ageing of those facilities. Research reactors are essential support tools for the nuclear industry and for the design of the next generations of power reactors. Thus it is important to keep these facilities operational. To conclude, this paper give ASN's perspectives for research reactors for the coming years and also highlight the regulatory challenges associated with keeping those facilities operational with a high level of safety. The regulation of the projects RJH (Jules Horowitz Reactor), th ITER (International Thermonuclear Experimental Reactor) or the next prototype of 4 generation reactor will be one of those challenges. IAEA-CN-156/S-30 Mar ia research reactocro nversion to LEU fuel Grzegorz Krzysztoszek Institute of Atomic Energy, Otwock-Vwierk, Poland E-mail addressg: krzysz@cyf.gov.pl The multipurpose high flux research reactor MARIA is a water and beryllium moderated reactor of a pool type with graphite reflector and pressurised fuel channels. The main areas of reactor application is production of radioisotopes and physics research with usage of a neutron beam from horizontal channels. Conversion programme of MARIA research reactor is closely associated with the “Global Threat Reduction Initiative” (GTRI). The MARIA reactor is a Russian-designed research reactor that operates with Russian highly enriched uranium (HEU) fuel. The reactor initially was operated with 80% enriched HEU and converted to 36% HEU in 1999. The programme objective is to convert the current core to LEU fuel. The proposed silicide fuel (U3Si2) with a 3 density of 4,8 g/cm has been qualified to achieving very high burnups. The assembly design proposed for use in the MARIA reactor will be unique, however, it needs to be qualified for the MARIA operating conditions by irradiating two lead test assemblies in the MARIA reactor. The Institute of Atomic Energy (IAE) have prepared a neutron-physics and thermo-hydraulic analysis [1], [2], [3]. Also the IAE with cooperation of the BR-2 reactor specialist elaborated the final fuel specification for fuel procurement contract. This document contains the specification and detailed design of fuel element, mechanical, chemical and thermal constraints, materials specification, design of the laminated fuel as well as drawings and tubes specification. Under the IAEA bidding process the technical evaluation team’s conclusion was that the CERCA and INVAP are qualified to manufacture U3Si2 fuel elements with flat or 3 slightly curved plates at densities of 4.8 g/cm . The first stage of LEU fuel qualification will include testing of mockup assemblies. IAE plans to perform hydraulic testing to measure the pressure drop through the assembly and thermal extention of tubes. The scope of work required for the preparation of the Safety Analysis Report (SAR) for the insertion and irradiation of lead test assemblies (LTA) in the irradiation of lead test assemblies (LTA) in the conversion to LEU was developed. The following transients analysis and specification of limits are required: 1. Decrease of core cooling capability 2. Deterioration of the cooling feasibility by the secondary circuit 3. Insertion of Positive Reactivity and Power fluctuation 4. Failures of the core structural components or experimental equipment 5. Partial Meltdown of core 6. Safety Limits for forced flow of coolant in primary circuits 7. Safety Limits at natural Convection of water in the fuel channel IAEA-CN-156/S-30 8. Safety Thresholds at forced flow of coolant in the primary circuits. The main stage for LEU fuel qualification will be irradiation of the LTAs in the MARIA core. Irradiation of the LTAs to the burn-up for qualification (~ 60%) is expected and it will take about 18 months. During the time of irradiation there will continuous monitoring of LTAa by means of the Radiation Monitoring System (RMS) to detect a fission product release based on delayed neutron measurements. On achieving the required burn-up the LTA’s will be discharged from the reactor core and subjected to the examination in the storage pool. Within the PIE the gaps between individual fuel tubes excluding the possible swelling of the fuel elements will be measured. The affirmative results of the aforementioned examination will allow for accomplishing the core conversion of the MARIA reactor by successive replacement of the fuel channels with high enrichment by the fuel channels with low enrichment. [1] Andrzejewski, K. et al., „Neutron-physics characteristics of the U3Si2 fule with 19.7 % enrichment”, Report IEA Nr: B-12/2005 [2] Andrzejewski, K. et al., „Comparision of neutron-physics characteristics for two types (MR-6 80%/350 and 36%/540) of reactor MARIA fuel and considered U3Si2 fuel elements with 19,75 % enrichment”, Report IEA Nr: B-23/2005 [3] Bykowski, W. et al., „Thermal characteristics of heat exchangers system of MARIA reactor fuel channels primary cooling circuit”, Report A IAE 124/A Session 8: Specific Utilization Applications Synopses no. IAEA-CN- Synopses Title Main Author 156/ Experience with different methods for on- and off-line detection of small releases of fission products from U-2 fuel elements at the HOR Delorme, T.V. Neutronic Analysis for the Fission Mo-99 Production U-7 by Irradiation of a LEU Target at RECH-1 Reactor Medel, J. Root Cause Analysis of Swelling Problem in Kartini U-4 Reactor Syarip, S. Technique of testing the VVER-1000 high burnup fuel rods in the MIR reactor at the design basis RIA U-23 parameters Alekseev, A.V. Realization of the IBR-2 Research Reactor U-41 Modernization Programme Vinogradov, A.V. IAEA-CN-156/U-2 Exper ience with different methods fonr -o and off-line detection of small releases of fission products from fuel elements at the HOR T.V. Delorme, A.C. Groenewegen, A. van der Kooij, W.J.C. Okx Reactor Institute Delft (RID), Delft, The Netherlands E-mail address of main author: T.V.Delorme@tudel ft.nl During the HEU to LEU-fuel conversion othf e Hoger Onderwijs Reactor (HOR) we encountered four cases in which the fission product concentration in the pool water increased significantly. Although this increase never caused any exceeding of intervention levels as they are mentioned in our permit and safety analyses, the ALARA principle urged us to analyse the situation in more detail. All cases have been related to a failure of the cladding of four elements respectively, all from the first batch of LEU fuel. Once an element has been identified to have a defect, it was removed from operation. The first two LEU cases are described in [1]. In this paper we present the strategy we used to handle such cases. Already before having these cases, differentet cdteion techniques were used for on- and off- line determination of (increased) activity within the reactor pool and containment, such as: ‚ Automated dose-rate measurement in containment; ‚ Direct integral measurement of deposits in ion-bed exchanger; ‚ Cooling water delayed neutron activity detection; ‚ Off-line analyses of ion-bed exchanger; ‚ Off-line analyses of pool water at the end of a production week; ‚ Direct measurement of air-borne activity concentration just above the reactor pool; ‚ Measuring primary cooling water delayed neutron activity; ‚ Contamination check at the exit of the containment. In our effort to obtain fast and reliable incadtiors for the reactor core performance, we developed the following instruments in addition to the listed techniques: ‚ A wet sipping device (NIP) to determine the fission product ‘leak rate’ of a separated, isolated element at suspense. Furthermore the idea is to set a reference level for every element and to detect minor anomalies in the ‘leak rate’ at an early stage by regularly re- measure it during the lifetime. ‚ A model to relate the pool water activity with the mean core fission product ‘leak rate’. This approach turns out to be a sensitive and reliable indicator even at very short reactor operation times as a few hours. Understanding the different factors from this analyses and combining it with the merits of the above mentioned direct air system led to the development of a new device. ‚ The ‘RID-cascade’, which measures on-line the air-borne activity concentration as it builds up from a direct coolant water-air interface. Important aspects of the device are: high efficiency for fission or decay products, easy maintainability and easy to operate. IAEA-CN-156/U-2 The paper will give an overview of the different methods for on-line and off-line detection of activity measurement devices used at the HOR. The newly installed techniques with their pro’s and con’s are discussed in more detail and a precise description of the ‘RID-cascade’ is given with the gained experience of this new instrument. [1] Verkooijen, A.H.M. and J.W. de Vries, “Experience with MTR fuel at the HOR”, Kerntechnik 69 (2004) 3. IAEA-CN-156/U-7 Neutronic Analysis for the Fission Mo-99 Production by Irradiation of a LEU Target at RECH-1 Reactor Jorge Medel and Gonzalo Torres Chilean Nuclear Energy Commission Santiago, Chile e-mail: jmedel@cchen.cl ABSTRACT The RECH-1 is a 5 MW open pool type research reactor, cooled and moderated by light water, reflected by beryllium, using MTR type fuel with low enriched uranium. The reactor is operated by the Chilean Nuclear Energy Commission and is located at La Reina Nuclear Center. The LEU fuel elements were built by the Chilean Fuel Fabrication Plant with a 3 uranium density of 3.4 g/cm . The technical specifications of these fuel elements were developed by the Chilean Manufacturer based on the original HEU assembly and approved by the reactor operator. The present core configuration, the first one with LEU fuel elements only, was configured in May of 2006 and has 32 fuel elements containing U3Si2-Al. 99 The Chilean Nuclear Energy Commission has initiated studies to produce fission Mo by the irradiation of a low enrichment uranium target at RECH-1 reactor, thanks to the IAEA Coordinated Research Project (CRP) “Developing Techniques for Small Scale Indigenous Molybdenum 99 Production using LEU Fission or Neutron Activation”, supported by ANL, USA. This target is made of a foil of 13 gram LEU metallic uranium inserted between two concentric aluminium cylinders. Results for the neutronic and activity calculations are presented, taking into account different irradiation and decay times. A criticality safety analysis for the storage of the originating radioactive wastes of the process was performed using MCNP code. The neutronic calculations were carried out for the present core configuration of RECH-1 reactor using WIMS and CITATION codes, supposing that the target would be introduced in the reactor grid position of the maximum thermal neutron flux. Reactivity change due to target loading has been examined to confirm the reactor safety. The mean neutron thermal flux and the power generated for the target were calculated. With the target power, the fission product activities have been calculated using ORIGEN-S code from the SCALE-4.4a system, taking into account different irradiation and decay times. At the end of the irradiation, the LEU foil target will be allowed to decay in the reactor pool. Prior to reactor start up, the irradiated target will be removed from the reactor pool and transported to the hot cell for disassembly. IAEA-CN-156/U-4 Root Cause Analysis of Swelling ProblemK inar tini Reactor Syarip 1), Sri Nitiswati 1), Y. Sardjono 1), Tegas Sutond 1o), T.W Tjiptono 1), Paul Stather 2s), R. Blevins2 ), Kevin Thorogood2 ) 1) National Nuclear Energy Agency (BATAN), Yogyakarta, Indonesia. 2) ANSTO, Structural Integrity Group, New Illawarra Road, PMB 1 Menai, NSW, Australia E-mail address of main author: syarip@batan.go.id Kartini reactor is an open pool type reacotof rT RIGA Mark II family, with aluminium pool liner of 6 mm thickness. The reactor has been in operation since 1979. In 2001 the pool was emptied of fuel and control system to ena ab lecomplete inspection of the pool liner after more than 20 years of service. Several NDTth modes had been used in the inspection i.e.: a comprehensive visual examination, hardness survey, dye penetrant examination, ultrasonic thickness survey, and replication foefatures of interest. In gernael the inspection revealed that the pool liner was in good condition and the res oufl ttshe thickness and hardness survey were consistent with the service history. Three aroefa isn terest were observed; a small area with apparent thinning, a small crack that was aneadl yass an original manufacturing defect, and there were two areas of swelling (bulges) eorbvesd under the thermal column and these were assigned as S1 and S2 for identification purposes. New inspection equipment was acquired byT BAAN through the IAEA TC Project No. INS 9022, and the features re-examined2 0in0 3, 2004, 2005 and 2006. The bulges had increased in size over this period. Visual inspections tbhye video equipment and replication indicate that the swelling parts observed in 2001 had gnr oinw size. In 2004 the dimension of S1 and S2 were observed to have 7.72 mm and 7 mm of height, and 1326 a5n md m1083 mm2 of area, respectively. While in 2005, the heights of aSn1d S2 were seen to increase to 7.78 mm and 7.56 mm, and the areas to be 1389 2m amnd 1839 mm2 respectively. Recent examination (Sep. 2006) showed that the size of bulgest ivrellay constant, and the peak of the bulges appears to contain tears (cracks). Therefore, it is wise to undertea ka review of the above features, starting with a root cause analysis related to the above swelling prob. leRmoot cause analysis (RCA) is a method to identify the root cause of an event or adveresned tsr associated with corrective actions to a set of events. The RCA result shows that probable root cause of swelling are as follows: ‚ It is probable that the seal on the cover plate in the service pool has deteriorated and allowed water to enter both the thermal column and the space between the aluminium reactor pool liner and the concrete. The wr awteill also saturate the concrete and has the potential to corrode the steel reinforcetm celonse to the surface of the concrete. It is believed that water leakage from tsheer vice pool has entered the area behind the aluminium pool liner and has saturated the concrete. • It is believed that carbon steel reinforcemt celnose to the inner surface of the reactor block has corroded. The expanding corrospiorond uct (rust) has the forced layer of concrete covering the steel reinforcement and subsequently pushing the aluminium pool liner inwards, causing the swelling. • There are two issues: the mechanism tfhoer creation of the swellings or bulges dominated by Iron corrosion, and the potential for corrosion of aluminium dominated by the pH of water in contact with theu amlinium. As the evidence for this condition IAEA-CN-156/U-4 were: the apparent corrosion of steeiln froercement behind aluminium liner and formation of Ca(CO3) on outside of concr ereteactor block indicating water saturation of the concrete block. (see Fig. 1.) Fig. 1. CaCO3 forming on the outer surface of the reactor block near the service pool The condition is advancing and will most probarbelys ult in a loss of pool liner integrity, this may or may not result in a loss of pool wadteure to the concrete backing behind the pool liner. The situation is not urgent at this timb ec ause a loss of cooling accident (LOCA) is not credible from the defects observed. A sloewa klage of pool water may result. The issue is predominantly one of maintenance not safetyis. Ivt ery probable that water is leaking from the sealing plate in the bulk shielding facil(itsye rvice pool) and has saturated the area behind the liner and probably the thermal column. T phrisovides an environment highly susceptible to corrosion for these components and it can be reasonably expected that accelerated pool liner corrosion will occur. Based on the above analysis, the remedial actions that can be considered are as follows: • The pool liner would be patched with a wel dpeadnel that would allow the features we have observed to keep growing in a harmwleasys. It is expected that they will reach a finite size as the iron is consumed by corrosion. • The area behind the pool liner would be d roieut, this should limit further growth of the bulges by improving the corrosion conditions behind the pool liner. • The bulk storage facility would be lined w isthtainless steel (the potential to use the thermal column could be preserved for future use). This would remove what is thought to be the root cause of the bulges and keep the reactor block dry. [1] SYARIP, PAUL STATHERS, Comprehensive In-Service Inspection Of Kartini Reactor Tank Liner And Fietnss For Continued Operatio, nProceedings of International Conference on Researceha cRtor Utilization, Safety, Decommissioning, Fuel and Waste Management, Sagnoti,a Chile, 10-14 Nov. 2003, STI/PUB/1212, ISBN 92-0-113904-7, © IAEA June 2005. [2] YONG H. LEE, Root Cause Analysis Focused to the Human Error Investiga, tions paper presented at theth I4AEA IRSRR Technical Meeting, Daejon, Korea, May 16-20, 2005. [3] SYARIP, WIDI SETIAWAN, Ageing Investigation and Upgrading of Components/ Systems of Kartini Research Rea,c toPrroceedings of the 1997 Workshop on the Utilization of Research Reactors, JAERI-Conf 98-015, October 1998. [4] http://www.alu-info.dk/html/alulib/modul/A00155.ht m: Aluminium in contact with concrete (Topic 1104 0) IAEA-CN-156/U-23 Technique of testing the VVER-1000 high burnup fuel rods in the MIR reactor at the design basis RIA parameters A.V. Alekseev, I.V. Kiseleva, V.N. Shulimov (FSUE”SSC RIAR”) International Conference on Research Reactors:Safe Management and Effective Utilization, Australia,Sydney, 5-9 November 2007 To study the VVER-1000 fuel behavior under design basis RIA accident conditions, a technique was developed and implemented in the MIR reactor for conducting reactor dynamic experiments with simulation of parameters responsible for thermomechanical state of fuel rods. A power pulse of the experimental assembly is formed in the reactor channel operating at constant power when the absorber shield (that surrounds fuel rods in the initial state) is removed from the assembly. Hafnium is used as the absorber allowing long-term operation at temperature of 0 400 ... 500 C. According to the nuclear safety conditions pulse formation must be accompanied by slight and negative reactivity addition. A reactivity compensator that moves synchronously with the absorber shield was introduced to reduce the effect on the core. The absorber shielding and reactivity compensator are designed as a combined shielding device. In order to move the shielding device at a high speed (up to 200 mm per sec), a fast acting hydraulic drive was developed. It uses the potential pressure energy of the loop primary circuit. The driving force for the shielding device movement appears when the space above the hydraulic cylinder piston contacts with the atmosphere. To limit and regulate the pressure drop across the hydraulic cylinder piston, an original device is applied, in which the throttling effect is achieved when two preliminary separated medium flows meet each other. A three-element fragment of the VVER-1000 fuel assembly including two high burnup fuel rods is used as an experimental fuel assembly. Total length of experimental fuel rods is 230 mm and the fuel column length is 200 mm. The specified fuel rod parameters –fuel column temperature and fuel enthalpy – are determined from the pulse amplitude and time of radiation exposure at the maximum power. A precise experiment control system was developed to set this time. The method for defining parameters, which determine the thermomechanical fuel state, is based on processing the experimental curve of temperature change in the fuel center and readings of other measuring sensors installed in the experimental fuel assembly and primary circuit of the loop facility under stationary and transient conditions. The rate of sensor signals measurement under dynamic conditions is equal to 100 Hz. The diagram below demonstrates the calculation results for distribution of the fuel rod linear power and fuel center temperature as a function of time. IAEA-CN-156/U-23 2200 900 2000 800 1 1800 700 1600 2 600 3 1400 4 500 5 1200 6 400 7 1000 8 300 9 800 200 LP 600 100 400 0 0 1 2 3 4 5 6 7 8 9 10 Time, s Diagram – Temperature of the fuel column center at different height (the fuel column is marked starting from its bottom and then in each 2.5 cm) and change of the linear power (LP) in section 4 Separation of the experimental fuel rods into short parts for calculations enables to obtain fragments of fuel rods with different pulse parameters within one experiment (amplitude ranges from 2.63 to 4.25 and half-width – from 0,8 s to 2.2 s). As a result it considerably increases the value of a single experiment. Parts 50 mm long are considered most appropriate. Temperature, Üで LP, W/ïm IAEA-CN-156/U-41 Realization of the IBR-2 Resear cRheactor Modernization Program A.V.Vinogradov1), V.D.Ananiev1), E.P.Shabali1n), I.T.Tretjakov2) 1)Frank Laboratory of Neutron Physics, Joinst tIitnute for Nuclear Research (JINR), Dubna, Moscow region, Russia 2)Research and Development Institute of Power Engineering (RDIPE), Moscow, Russia E-mail address of main authoar:l vin@nf.jinr.ru The pulsed fast reactor IBR-2 is a pulsed troera ocf periodic action (pulsed reactor) and its original difference from other reactors cont s isin mechanical reaticvity modulation with a movable reflector (MR). The movable reflecto ra i scomplex mechanical system with a total mass up to 60 t providing for reliable operation t hoef two parts, which determine reactivity modulation: the main movable reflector (MMRan) d the additional movable reflector (AMR). The MMR and AMR rotors rotate in the samere dcition with differentv elocities. When both reflectors coincide near the reactor zone, a p opwuelsre is generatedT.h e factors determining the duration of a fast neutron pulse are fnaesut tron lifetime, configuration and rotation velocity of the rotors [1]. The IBR-2 reactor was put intoop eration in February 1984. Aptr esent an average power of the reactor is 1,5 MW and pulse repetition rias t5e Hz. Due to its pulse power, equal to 1500 MW, IBR-2 possesses the highest in the wd oprlulsed thermal neutron flux for beam investigations, which is 1106 n/cm²©s. Pulse duration is 24o5s for fast neutrons and 35o0s for thermal neutrons (behind the 4 cm thick water moderator) [2]. The IBR-2 reactor is used principally for abme studies in solid-state physics (solids and liquids), biology, and material science. Thxep erience of the IBR-2 reactor operation proved it to be a rather effective neutron source, whh fiocr many applicationsis as good as the best sources, based on proton accelerators. More othver ,development of neutron experiment technique and application of modern develoepnmts at the IBR-2 reactor have shown that neutron flux magnitude is ofuf ndamental importance for higehff iciency of a pulsed source. At the same time pulse duration can be diffe rienn dt ifferent experiments. This methodical conclusion can be essential for further doepvemlent of neutron sources throughout the world. Operating experience of IBR-2 is especially imrtapnot at present when the interest is aroused in pulsed neutron sources w iltahrge pulse duration [3]. The IBR-2 reactor possesses a record neuftlruoxn and yet is a very economic and rather cheap machine. Its construction cost is ~20 inMc$luding the cost obf uildings, and operating expenses are 1 M$/year. Due to low averagwee pr,o activation of the equipment and burning up of the zone go on slowlyU. nder the established operatirnegg ime of 2500 hours per year for physical experiments, the service life iso uatb 20 years for the zone and 7 years for the movable reflector. Accounting for the time undeirs tohperating regimeo, ne can see that the service life of the main units othf e reactor should end in 2002. With this account it was considered approtper iato replace units of the reactor where necessary after 2002. The fundingff icdui lties, however, made us to revise the established operating regime to slow down the wearing ou tI BoRf -2 remaining service life, so that the reactor will have its rated resource exhteadu sby 2007. In this connection the reactor modernization program has elaborated foer ptheriod up to 2010. The present concept of the IBR-2 reactor modernization involves rcyainr g out work including development, manufacturing and installation othfe reactor equipment. At the same time, accounting for the experience of reactor operation and physicala rrecshe, the given concept contains a number of IAEA-CN-156/U-41 novel technical solutions that substalnlyti aimprove operation and physical reactor characteristics, which permits one to assert that actually in the process of modernization a new IBR-2M reactor i sbeing created [4]. In the report the following basic themes are stated: general descrimptaioinn ,c haracteristics and current state of the rearc. toAlso the realization of the IBR-2 reactor modernization program and its main directions, terms, ress, uflitnancing, postnat aml odernization IBR-2 parameters are given. [1] E.P.SHABALIN / Fast Pulsed and But rRseactor, Pergamon Press, Oxford, 1979 [2] V.D.ANANIEV, A.V.VINOGRADOV, The IBR-2 pulsed research reactor: status report. Proceedings of PANS-II Secondt eIrn ational Seminar "Advanced Pulsed Neutron Sources: Physics of/at AdvancPeudls ed Neutron Sources", June 14-17, 1994, Dubna, Russia. [3] V.L.AKSENOV / Update on pulsed reactoIBr R-2 at Dubna, Phisica B 174 (1991), North-Holland Publishing Company [4] V.V.KHMELSHCHIKOV, I.T.TRETJAKOV, A.A.PORTNOV, A.V.VINOGRADOV Modernization and reconstruction of ageings Rsiuan research reactors as the method to extend their operational life (IAEA-SM36- 0/42). Symposium “Research Reactor Utilization, Safety and Management”, 6–10 September 1999, Lisbon, Portugal Session 9: Decommissioning and Waste Management Synopses no. IAEA-CN- Synopses Title Main Author 156/ DE-3 Overview of Research Reactor Decommissioning Rowling, J. The transition from a research reactor in operation to a DE-2 facility undergoing decommissioning Nielsen, K.H. Carbon steel construction removal from spent fuel WM-2 storage pool Pešic, M. IAEA-CN-156/DE-3 Overview of Research Reactor Decommissioning SYNOPSIS John Rowling ANSTO john.rowling@ansto.gov.au The purpose of this paper is to present a wide view of Research Reactor decommissioning. The focus will include historical terms, showing progress, trends, and state-of the-art strategies for remaining and emerging issues. Some of the issues relate directly to the decommissioning program planned by the Australian Nuclear Science and Technology Organisation (ANSTO). ANSTO has operated HIFAR, the 10MW research reactor since 1958. HIFAR was shutdown at the beginning of this year after 49 years of successful operation. The reactor provided neutron beams for research purposes, radioisotopes for medical and industrial use and irradiated silicon for the semiconductor industry. In addition, although ANSTO has successfully mothballed MOATA, a small Argonaut type reactor, there are still substantial planning requirements to progress MOATA’s final decommissioning program. ANSTO faces a number of challenges heading into the decommissioning of HIFAR. These include: the establishment of a modern decommissioning strategy in the absence of a long-term waste management facility or waste acceptance criteria for the material generated by the decommissioning. The impact of closure of the facility on staff morale and retention of key staff, plus the regulatory requirements as well as the Organisation’s needs are now the major issues for the Decommissioning Team. These challenges are compounded by competition for skilled resources now required by the full power operation of the newly commissioned research reactor (OPAL) at the same site and to meet the needs of the researchers, isotope production and the silicon customers. The technical problems are now not insurmountable for decommissioning; however the areas of policy, planning, timing, costs, waste disposal, safety criteria and regulatory aspects need further development. The industry now must demonstrate its maturity and keep up with safety, environmental and regulatory requirements, with pressure of change in political perceptions and expectations. There are 832 research reactors, of which there are 287 operating, 524 either shutdown or decommissioned plus 205 that are greater than 40 years old. There are 90 reactors planned or progressing to unrestricted use or safe enclosure (11%). There are 233 with a current status of decommissioning completed as “in unrestricted use” or “safe enclosure” (28%). This is based on data from the IAEA at May 2006 from Technical Report Series No. 446. The Community have an expectation of the nuclear industry to clean up their backyard. This is much easier for the nuclear power industry where the income from electrical tariffs make provision for decommissioning, such as in France. The nuclear research reactor industry has supported the power industry by its research but has been left out of the funding equation in most cases. In Australia there is no nuclear power industry as coal-fired power stations dominate electrical power production. IAEA-CN-156/DE-3 Fortunately the federal government will fund the decommissioning of MOATA and HIFAR. Two years ago we commenced the planning for the decommissioning of HIFAR, this included the application for funding. HIFAR was performing better over its last five years than it did over its previous 40 odd years. However the plant and equipment was old. The processes were manual and time consuming - hence the reason to construct OPAL. It was determined that ANSTO had four choices for HIFAR after shutdown: 1) do nothing, 2) immediate dismantling 3) deferred dismantling, and 4) entombment. 1 & 4 were perfectly reasonable alternatives. The site now had OPAL for perhaps another 40 years. The preferred option was 2) immediate dismantling, taking advantage of the expertise of the existing highly skilled staff but this required a national waste repository. The only good option then left was 3) deferred dismantling. If immediate action was taken to dismantle the reactor block there would be significant cost implications in the storage and double handling of activated waste. The fuel and heavy water have been removed. This work was performed under the existing operating licence. With this strategy in place we have prepared an application to the regulator for a “Control or Possess” licence for a period up to ten years. It is planned that during this period we can remove all non-active components and shrink the footprint of the HIFAR facility as a whole to just the reactor building shell and the auxiliary plant room. Economic sustainability is not an issue as ANSTO is located in the outer metropolitan area of Sydney. ANSTO has operated HIFAR for years and there is a well developed fuel and waste management policy. The next major issue for ANSTO is the active waste release/clearance criteria taking into account the newly proposed waste repository. There is no agreed location for the repository and no waste acceptance levels set yet. Planning for decommissioning did start some 2 years ago using the above deferred dismantling strategy and similarly we have followed the IAEA guidelines for a structured approach consisting of generic tasks namely: a) Preparatory work and planning for shutdown; b) Final shutdown; c) Removal of radioactive sources (including liquids); d) Radiological characterisation (delayed until prior to application for decommissioning licence); e) Decontamination and inactive dismantling; f) Refurbishing of essential systems (HVAC and Electrical Supply) g) Demolition of structures and buildings (Under P or C licence); h) Surveillance and maintenance (Through whole project); i) Final dismantling of reactor block and building j) Waste Management k) Site clearance and release (Planned for 2017-18) During this total process substantial time will be given to the collection and recording of all data. The operating records have been collected and analysed to capture past events that may have significance in the decommissioning work. “Lessons learnt” for me have been such items as staffing the decommissioning team, maintenance of a good safety culture during final stages of operation, working towards regulatory approval for decommissioning and strategies for knowledge retention. IAEA-CN-156/DE-2 The transition from a research reacitno ro peration to a facility undergoing decommissioning Kirsten Hjerrild Nielsen Danish Decommissioning, DK-4000 Roskilde, Denmark E-mail address: khn@dekom .dk Danish Decommissioning (DD) was establishaesd an institution under The Ministry of Science, Technology and Innovation based uproens oal ution passed by Parliament in March 2003. The task is to decommission all the nucflaecairli ties at Risø National Laboratory (Risø) to the state of “Green Field” where all buinilgds, equipment and materials that cannot be decontaminated below established clearancel sl eavre removed, and to maintain the nuclear facilities until these have been fully decommissioned. The six nuclear facilities to be decommissioneed: aTrhree research reactors - DR 1, DR 2 and DR 3, Hot Cells, Fuel Fabrication and a WaMstaen agement Plant. Figure 1 below shows the location of the nuclear facilities at Risø: FIG. 1. Location of the nuclear facilities at Risø. FIG. 2. Schedule. IAEA-CN-156/DE-2 The time frame for decommissioning the nucleairl itfiaecs at Risø is 11-20 years (DD expects to have completed the task by 2018). The time schedule is shown above in Figure 2. DR 1 has been decommissioned in 2004-2005 and the hall and surrnogu nadrieas were released for unrestricted use in January 2006. The decommissioning of2 D isR ongoing and clearance measurements will be done in 2008. Reactor DR 3 – which this paper will be focusing on – was a 10 MW heavy water cooled and moderated research reactor of a design similar to the British "PLUTO" type. Originally, DR 3 was built as a Materials Testing Reactor, but ended as a Multipurpose Research Reactor. With a cold neutron source, six three-axis spectrometers and a small-angle neutron scatter instrument DR 3 was appointed a Large European Beam Facility and these neutron beam instruments were used intensively by recsheears from Risø and from the other EEC- countries. The main production activities w eNreutron Transmutation Doping of Silicon (NTD), isotope production and activation anaisly. sBy the end of its operating period, DR 3 supplied 1/3 of the world market of NTD. DR 3 was in operation from 1960 until 2000 in a 4-week-cycle with 23 days ofn ucounsti operation and 5 days of shut down. After thnea lf ishut down the fuel elements have been removed and shipped to the US. The heavy whaates rb een stored in drums and later exported to Canada. Characterization of the reactor block has been done in 2005-2006 and characterization of DR 3 storage is carried out in 2006-2007. Most of the auxiliary systems outside the roera cbtuilding have been removed by now (e.g. the secondary and tertiary cooling systems) moor dified (e.g. the power supply). Inside the containment the removal of auxiliary systems are ongoing as part time taosnkgs wailth decommissioning of other nuclear facilities at sthite and are to be finished by the end of 2011. Planning the decommissioning of theea crtor block has started and the work is to be carried out in 2012-2016. Originally, the DR 3 reactor was suppo steod be running until 2006, but in 2000 it was decided to close all the nuclear facilities at Risø. Therefore no planning of the decommissioning had been done in advance thaen do rganization was not prepared for the new purpose. During operation a large poafr tt he work was well known – operating and maintaining the reactor – according to thefi ndeed running and shut down periods. After the final shut down of DR 3 the type of work agdrually changed from routine tasks into more project-based. The organization has undergone major changes in order to adapt to the new conditions. This paper describes the transition from otpinegra a research reactor to be decommissioning a nuclear facility and deals with the change of the organization from bedienpga art ment of Risø to be an independent institution. The “treading wa pteerr”iod form right after the decision to finally shut down DR 3 was made until it was decided how to get on with the decommissioning will be discussed. The relations to the authorities, who wase rleittle prepared for the new situation as we were, will be described along with the issue of preparing the staff for the change – retraining and replacing a large part. The current statusD Dof – how did we manage the decommissioning of the two small reactors – and how are we prinegp atrhe decommissioning of the largest reactor, DR 3? How do we maintain nuclear knowleding eh ouse when key personnel retire and how do we keep the staff until 2018? These are some of the issues, which will be covered in this paper. IAEA-CN-156/WM-2 Carbon Steel Construction Removal from Spent Fuel Storage Pool M. Peši5, O. Šoti5, S. Pavlovi5, V. Ljubenov, A. Nikoli5 Vin7a Institute of Nuclear Sciences, Belgrade, Serbia E-mail address of main author: mpesic@vin.bg.ac.yu Carbon steel structure in basin no. 4 of the spent nuclear fuel pool the RA research reactor at Vinca Institute of nuclear sciences, Belgrade, Serbia was constructed in 1959 for spent fuel assembly management. It consists of a robust metal construction with so called ‘working table’ and two sets of two connected coaxial tubes. The construction is shown in Figure 1. Fig. 1. Carbon steel construction with tubes The construction was the main source of corrosion process products in the pool filled with stagnant tap water. Inappropriate water chemical parameters [1] and carbon steel corrosion particles have also initiated corrosion of aluminium cladding of spent fuel elements stored in aluminium storage barrels. Fission products (137Cs nuclide) leakage to the pool water was detected. Also this construction was the main source of sludge creation that was deposited to basins bottom walls, the construction itself and spent fuel storage containers. For that reason, a complex technological project was initiated, in 2004, with the IAEA TCP assistance and Russian R&D company OEC NIKIMT to design and manufacture a proper equipment and develop safe technological process to remove (under water) the carbon steel construction from the pool. After long and careful preparation processes, facility modification and upgrading many elements of radiation protection systems, including training RA reactor operational staff in procedures prepared and at mock up and after thorough review of the Safety analysis report of the whole operation, the Serbian regulatory authority issued approval for the operation in November 2007. IAEA-CN-156/WM-2 The removal of the structure from the spent nuclear fuel pool was carried out from mid- November 2006 to March 2007. The main structure was safely cut in few big pieces (due to unforeseen high gamma dose rate of the deposits at the construction further cutting to smaller pieces was abandoned). Tubes were safely cut in two pieces for the same reason. All removed elements were packed in new designed metal storage containers and temporary and stored at the Vinca RAW storage facility area, until commission of the new designed Waste Processing Facility (foreseen at the end of 2008). Figure 2 shows operators’ actions on removal of the biggest part of metal construction, ‘working table’ and two inner tubes, all heavy corroded and their storage at metal containers. Fig. 2 Removal of the biggest part of metal construction, ‘working table’ and two inner tubes Acknowledgement Authors and the all members of the Vin7a VIND SNF/RadProt/RAW team acknowledge to the IAEA TCP Department (Contract SCG4003-890905), IAEA experts from NFCWT /NFCM and DNIS/RRS Sections, Ministry of Science and Environmental Protection of the Republic Serbia, and the experts of the OEC NIKIMT, Obninsk, from the Russian Federation for their support and engagements during the whole operation. References [1] PEŠI4, M., et all, “Study of Corrosion of Aluminium Alloys of Nuclear Purity in Ordinary Water” – Part I and Part II, Nuclear Technology & Radiation Protection, Vol. XIX, No.2, pp. 77-93 (2004) and Vol. XX, No.2, pp. 45-60 (2005) Session 10: Core Safety and Utilization Parameters Synopses no. IAEA-CN- Synopses Title Main Author 156/ Status Report on Preparation of IAEA Guidelines for S-36 Qualification of Research Reactor Fuels Snelgrove, J.L. Simulation of Flow Behavior in the HANARO Reactor S-13 Pool by Using the MARS Code Park, C. Experimental Measurements for Plate Temperatures of MTR Fuel Elements at Sudden Loss-of Flow Accident S-27 and Comparision with Computed Results Sevdik, B. Contract Performance Demonstration Tests in the S-60 OPAL Hergenreder, D.F. ANNEX B IAIEAAE-CAN-C-1N56-/1S5-366/ Status Report on Preparation of IAEA Guidelines for Qualification of Research Reactor Fuels 1) James L. Snelgrove 1) Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 USA E-mail address of main author: jimsnelgrove@anl.gov Because development of high-density research reactor nuclear fuels (RRNF) is being pursued in many countries and because no comprehensive document addressing the rationale of 1 qualification of RRNF has been published, the IAEA is carrying out a project to produce a set of guidelines for RRNF qualification. These guidelines are intended to provide points of reference for the type, quality, and completeness of the information to be generated in order to ensure the acceptable performance of high density LEU fuels to be used and / or retrofitted in existing and new research reactors. It is anticipated that this guidelines document will be in the final editing stages by the time of the Sydney Conference. This paper provides an overview of the content and conclusions of this document. The Guidelines Document addresses basic definitions; approaches and processes relevant to the qualification of research reactor nuclear fuel and defines essential information required for the licensing and use of fuels in research reactors. The most basic definition is that fuel qualification is a process carried out both by a fuel developer and a fuel manufacturer. The fuel developer must demonstrate the acceptable irradiation performance under conditions, 235 including fission rate and U burnup, that exceed any conditions that might exist during normal operation of a reactor using the fuel and in geometric configurations used in the reactors that are candidates to use the new fuel. The fuel manufacturer must demonstrate that it can reliably and consistently manufacture fuel assemblies acceptable to the customer. The information that should be provided by the fuel developer for fuel qualification falls naturally into five groups: (1) basic fuel properties, (2) as-fabricated fuel meat properties, (3) fuel meat irradiation properties, (4) fuel element (plate, tube, or rod) properties, and (5) fuel assembly properties. This information provides the basis for fuel element and fuel assembly design and the basis for licensing the use of such fuel assemblies in research reactor cores. It should be noted that the fuel developer should endeavor to obtain as broad a range of data as possible at each step of development, within constraints of schedule and funding, in order to maximize the applications for which the fuel can ultimately be licensed. The ultimate proof that a fuel is qualified from an irradiation-performance perspective is the approval by a national regulatory authority for use of the fuel in a research reactor. Manufacturing process qualification occurs after the fuel has been developed and is limited to activities performed to demonstrate that the process and product meet specified requirements. However, in most cases, the manufacturer has participated in the fuel development process. At the very least, it is generally expected that the manufacturer will have produced fuel 1 Current Guidelines Document Committee Members: P. Adelfang (IAEA); H. Taboata (Argentina, CNEA); P. Lemoine (France, CEA); D. S. Sears (Canada, AECL); C. Jarousse (France, CERCA); A. Enin (Russia, NCCP); N. Arkhangelsky (Russia, Rosatom); C.-K. Kim (South Korea, KAERI); J. L. Snelgrove, T. Totev (USA-ANL); M. Nilles (USA-BWXT); D. Wachs, D. E. Keiser (USA-INL) IAEA-CN-156/S-36 elements and fuel assemblies for final irradiation-behavior qualification tests. Nevertheless, ultimately, manufacturer qualification is determined by the customer, guided by requirements of the particular country and its national regulatory authority. In contrast to the usual one-time irradiation-behavior qualification for a particular reactor or reactor type, manufacturer qualification may be required more than once if the manufacturer produces fuel for a particular reactor only sporadically. In summary, qualification of RRNF is a complex process involving the fuel developer, the fuel manufacturer, the fuel customer (usually the reactor operating organization), and the national regulatory authority. This new IAEA Guidelines Document is expected to clarify the process for all of these parties and, thus, lead to a more-efficient qualification process in the future. IAEA-CN-156/S-13 Simulation of Flow Behavior in the HANARO Reactor Pool by Using the MARS Code C. Park*, H. Kim, B.D. Jeong Korea Atomic Energy Research Institute (KAERI), Daejeon, Korea cpark@kaeri.re.k r Synopses Generally speaking, no robust computer cosduecsh as RELAP5, RETRAN, and MARS are available for the analysis of a research reactor, and it is highly recommended that these computer codes developed for a transient ana olyfs pisower reactors baep plied carefully after assessing their applicability to a specific research reactor. The MARS code [1] is a realistic system transient analysis code that can beed ufosr the simulation of a wide variety of PWR system transients. This code is a unifiedrs ivoen of a 1-D reactor system analysis code, RELAP5/MOD3 and a 3-D reactor vessel aynsaisl code, COBRA-TF coupled with a 3-D reactor kinetics code, MASTERn da a containment code, CONTEMP. TS4ome assessment calculations [2] have shown that the MARS c ocdaen also be applied to the thermal-hydraulic analysis of research reactors with careful evaluations. The HANARO is an open-tank-in-pool type resceha reactor of 30 MWth [3], where a 3-D analysis of its flow behaviour is often nescseary. One of major areas concerned for the 3-D flow behaviours in HANARO is reactor pool werhe the bypass flow may rises up to the pool surface. In this paper the simulation results of the flow behaviour in the HANARO reactor pool by using the MARS code are described. The measurement and prediction by the CFD code [4] for the flow behaviour in theH ANARO reactor pool shows that a counter clockwise circulating flow exists in the entire reactor pool if one looks into the pool from the top, as shown in Fig. 1. It may bec aused by an unsymmetrical pouring of the bypass flow into the bottom of the reactor pool. A large swirl flow rises up slowly toward the pool surface, but it soon turns downwda art near the bottom of the hot water layer and then fitlo ws downward to the core through the chimney. FIG. 1 Flow Behavior in the HANARO Reactor Pool IAEA-CN-156/S-13 The dimensional flow behaviours in thHeA NARO reactor pool mentioned above have been simulated by using the 3-D model of MthAe RS code in order to inquire about the applicability of the MARS code to suchc a se. The nodalization of the HANARO reactor pool for the MARS code simulation is shno win Fig. 2. The reactor pool was modelled by using the MARS code multi-dimensionmalo del. The numbers of Z, R, anごd cells are 19, 8 and 8, respectively. So, the total number ofu vmoels for the reactor pool are 1216. The reactor assembly part (green colour) is modelled as a 1-D part. The reactor pool surface is connected to a time dependant volume as boundary condoitifo tnh e atmosphere. Other components such as the piping and pumps are modelled with proper components in the code. FIG. 2 Nodalization of the Reactor Pool and System of the HANARO The results showed reasonable predictifoonrs t he 3D flow behaviour in the HANARO pool. This capability may be useful to predtihcet effect of 3D flow phenomena on a core during a flow reversal transient in research reactors. References [1] J.J. Jeong, W.J. Lee, B.D. Chung, “Simulatoiof na main steam line break accident using a coupled system thermal-hydraulics, eteh-rdimensional reactor kinetics, and hot channel analysis code”, Annals of Nuclear Energy 33, 2006 [2] C. Park, Validation Calculations for thep pAlication of MARS Code to the Safety Analysis of Research ReactoKrsA, ERI/TR-3259/2006, Technical Report, 2006. [3] http://hanaro.kaeri.re.kr/english/index.html [4] H. Kim and G.Y. Han, “Flow Charaecrtistics of HANARO Reactor Pool”,t h9 IGORR, Sydney, Australia, 2003. IAEA-CN-156/S-27 Experimental Measurements for Plate Temperatures of MTR Fuel Elements at Sudden Loss-of Flow Accident and Comparison with Computed Results 1) 2) 1) 1) Bülent Sevdik , Tanzer Türker , Sinan Taylan , Onur Uzonur , 1) ÇNAEM, Reactor Division, Istanbul, Turkey 2) ÇNAEM Nuclear Engineering Division, Istanbul, Turkey E-mail address of main author: bulent.sevdik@taek.gov.tr The aim of this study is to generate experimental data to be used for sensitivity analysis and assessment calculations on the thermal-hydraulic codes written for Loss of Flow Accident (LOFA) analysis of research reactor with MTR- type (plate type) fuel elements. In an open pool research reactor with downward coolant flow, an accident such as shaft breaks between pump and flywheel can lead to sudden loss of flow. Flow reversal occurs in many MTR-type research reactors during transient after shut-down (flow scram). It can be mismatched with the decay heat removal by natural convection when the coolant flow rate becomes so low. A critical heat flux may occur even at a low heat flux. It is essential to demonstrate that such a sudden loss of flow does not lead to excessive temperature increase at the fuel plate and consecutively not cause fuel melting. In the TR-2 Research Reactor, Sudden Loss of Flow Accident which happens by shaft break between pump and flywheel is simulated by closing the core outlet valve. There is a butterfly type valve at the core outlet and total closing time is about 30 seconds but effective flow decrease occurs in 10 seconds. Beginning of the flow decrease by closing of the valve was accepted as time zero and the decrease of flow rate versus time was plotted. An instrumented fuel element, which has five thermocouples along the vertical length of the fuel plate, was used for the experiments. This fuel element was placed in a certain position of the TR-2 core and the plate temperatures were measured and recorded by a digital recorder. At the same time primary flow rate and core pressure drop (∆P) at the grid plate were measured and recorded. Power distributions of the fuel elements in the reactor core were determined experimentally by using copper wire activation technique. According to the existing core loading of the TR-2 Reactor, core positions, U-235 weights, relative neutron flux distributions and power distributions of the fuel elements were given as the inputs of the calculations to simulate the loss of flow accident at the computer codes. Four experiments were performed for sudden loss of flow accident. Three of these experiments were repeated with different initial conditions such as different core inlet temperatures and different reactor operation times at 5 MW nominal power level and 750 3 m /h primary flow rate. The reactor was shutdown by flow scam at these three experiments. Experiment I: The TR-2 Reactor had been operated at 5 MW for 5 minutes. The 3 o primary flow rate was 750m /h and core inlet (pool) temperature was 23 C. Primary coolant 3 flow rate began to decrease from 750m /h (time zero) by closing of the core outlet valve. The primary flow scram happened after 7 seconds and natural convection flappers opened after 25 seconds. Plate temperatures and the decrease in primary flow rate were recorded for 1 second time intervals. IAEA-CN-156/S-27 Experiment II: The TR-2 Reactor had been operated at 5 MW power for 5 minutes. 3 o The primary flow rate was 750m /h and core inlet (pool) temperature was 17 C. Primary 3 coolant flow rate began to decrease from 750m /h (time zero) by closing of the core outlet valve. The primary flow scram happened after 9 seconds and natural convection flappers opened after 35 seconds. Plate temperatures and the decrease in primary flow rate were recorded for 1 second time intervals. Experiment III: The TR-2 Reactor had been operated at 5 MW power for 2 hours. 3 o The primary flow rate was 750m /h and core inlet (pool) temperature was 29.5 C. Primary 3 coolant flow rate began to decrease from 750m /h (time zero) by closing of the core outlet valve. The primary flow scram happened after 9 seconds and natural convection flappers opened after 28 seconds. Plate temperatures and the decrease in primary flow rate were recorded for 1 second time intervals. The sudden loss of flow accidents performed in these three experiments were simulated and analyzed by using PARET computer code. The measured plate temperatures were less than the calculated plate temperatures. The differences may come from the decay heat calculation or modeling of the natural convection for narrow, rectangular flow channels at PARET code. To investigate the source of these differences a fourth experiment was performed at 50kW power level, which allows the reactor to operate with natural convection without primary flow scram. Experiment IV: The TR-2 Reactor had been operated at 50kW power for 5 minutes. 3 o The primary flow rate was 750m /h and core inlet (pool) temperature was 19 C. Primary 3 coolant flow rate began to decrease from 750m /h (time zero) by closing of the core outlet valve. The primary flow scram not occurred (50kW neutronic power) and natural convection flappers opened after 35 seconds. Plate temperatures and the decrease in primary flow rate were recorded for 1 second time intervals. The results of the experimental measurements were given in graphics (plate temperatures and flow rates versus time). Computational results obtained from PARET code were compared with the experimental values. In conclusion, it was determined that the residual decay heat calculation and the natural convection modeling for narrow rectangular channels of PARET code are not quite realistic. PARET uses the decay heat generation rate within the reactor core based on the standard fission-product decay heat curve for uranium-fuelled thermal reactors published by the American Nuclear Society as a proposed standard (ANS-5.1/N18.6). ANS decay heat curve is too pessimistic. For the calculation of the residual decay heat, special attention must be given to the fact that “Approximately one-half of the decay heat is due to gamma radiation with energies in the range of about 0.2-2.0 MeV. The e-folding length for 1 MeV gamma radiation in aluminum is 6 cm, i.e., the source gamma intensity is attenuated to 1/e of its original value after passing through 6 cm of aluminum. This corresponds to a penetration of 40-50 fuel plates and hence shows that a significant portion of a given level of fuel element decay heat will be deposited at the outside of the fuel element”. Realistic calculations can be done for LOFA and LOCA analysis by using the half of the values of decay heat generation curve of ANS. IAEA-CN-156/S-60 IAEA-CN-156/S-60 Session 11: Programmes for the Minimization of the Use of HEU Synopses no. IAEA-CN- Synopses Title Main Author 156/ Status of the United States Department of Energy's F-8 Nuclear Fuel Return Programmes Dickerson, S.L. Conversion of Research and Test Reactors to Low Enriched Uranium Fuel: Technical Overview and F-11 Programme Status Roglans-Ribas, J. High Density Fuel Development for Research Wachs, D.M. F-12 Reactors Lemoine, P. Measuring Progress in Reactor Conversion and HEU Minimization Towards 2020 – the Case of HEU- S-35 fuelled Research Facilities Reistad, O.C. IAEA-CN-156/F-8 IAEA-CN-156/F-11 Conversion of Research and Test Reactors to Low Enriched Uranium Fuel: Technical Overview and Program Status1 J. Roglans-Ribas Nuclear Engineering Division, Argonne National Laboratory 9700 So. Cass Avenue, Argonne, IL 60439 USA roglans@anl.gov Nuclear research and test reactors worldwide have been in operation for over 60 years. Many 235 of these facilities operate with high enriched uranium (HEU – U enrichment  20%) fuel. In response to increased worries over the potential use of HEU from research reactors in the manufacturing of nuclear weapons, the U.S Department of Energy (DOE) initiated a program – the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel in research reactors by converting them to low enriched uranium (LEU) fuel. The reactor conversion program was initially focused on U.S.-supplied reactors, but in the early 1990s it expanded and began to collaborate with Russian institutes with the objective of converting Russian-supplied reactors to the use of LEU fuel. Increased security concerns in recent years have led to the establishment of the Global Threat Reduction Initiative (GTRI) by the U.S. DOE’s National Nuclear Security Administration. The overall GTRI objectives include securing radiological and fissile materials. A follow up conference for the International GTRI partnership held at the IAEA in September 2004 [1] established the framework for international collaborations in meeting the goals of the program. As an integral part of the GTRI, the Conversion Program has accelerated the schedules and plans for conversion of additional research reactors operating with HEU. A total of 129 reactors are included in the scope of the Program. The major technical activities of the Conversion Program include: (1) the development of advanced LEU fuels; (2) conversion analysis and conversion support; and (3) technology 99 development for the production of Molybdenum-99 (Mo ) with LEU targets. The key factor in enabling the conversion of a research reactor lies in the availability of a fuel 235 with much greater uranium content, to compensate for the reduction in the content of U in the LEU material. Several high density LEU fuels have been developed. The RERTR program developed the uranium disilicide dispersion fuel. General Atomics developed LEU fuel for TRIGA reactors. Oxide tube-type LEU fuels for the conversion of Russian-supplied reactors have also been qualified for use in several reactors. An accelerated fuel development and qualification program focusing on UMo alloy fuel is currently underway with the objective of qualifying very high density LEU fuels to enable the conversion of high performance research reactors that are not convertible with the existing qualified fuels. 1 The submitted manuscript has been created by UChicago Argonne, LLC, Operator of Argonne National Laboratory “Argonne”). Argonne, a U.S. Department of Energy Office of Science laboratory, is operated under Contract No. DE-AC02- 06CH11357. The U.S. Government retains for itself, and others acting on its behalf, a paid-up nonexclusive, irrevocable worldwide license in said article to reproduce, prepare derivative works, distribute copies to the public, and perform publicly and display publicly, by or on behalf of the Government. IAEA-CN-156/F-11 The conversion analysis and support activity provides the required analytical and design evaluations to support the program. Since the inception of the RERTR program, analysis methods and codes have been developed specifically for the analysis of research reactors. The methods and codes are currently evolving to incorporate the latest tools and data and have been validated with experimental data. Conversion analysis in general includes three major tasks: - Feasibility studies to determine suitable LEU fuel assembly designs for each reactor. The design of the fuel assembly may be an iterative process with the fuel development activities. - Operational and safety analysis, necessary to demonstrate that the transition from HEU to LEU fuel can be done safely and without interrupting normal operations - Resolution of regulatory issues to obtain regulatory approval for the conversion to LEU fuel. It must be demonstrated that all safety requirements are met. 99 The programmatic objective of the technology development for Mo program element is to 99 eliminate the use of HEU targets in production of Mo . The program intends to accomplish this objective through the development of LEU targets and chemical processing methods that do not significantly impact the isotope production yields, costs and waste treatment and disposal with respect to current production with HEU targets. Since the inception of the Conversion Program, 51 of the 129 reactors have been converted to LEU fuel or have shutdown prior to conversion. The current goal is to convert the remaining 78 reactors in the list of candidates by the year 2018. Of the 78 remaining research reactors within the scope of the Conversion Program, 50 can be converted with existing LEU fuels, the high density UMo fuel under development will allow the conversion of 19 additional reactors, and the remaining 9 reactors require further analysis. A key element of the Conversion Program is the coordination with multiple organizations, ranging from facility operators to regulatory bodies, and significantly, the International Atomic Energy Agency (IAEA). The IAEA supports the objectives of the Conversion Program and is currently leading coordinated research projects for conversion analysis of 99 Miniature Neutron Source Reactors and technology development for Mo production, in addition to country-specific conversion support projects under the Technical Cooperation Department. The paper will provide a more detailed overview of the status of the program, the technical challenges and accomplishments, and the role of international collaborations in the accomplishment of the Conversion Program objectives. 1. IAEA, Global Threat Reduction Initiative, International Partners’ Conference, Summary of the Proceedings and Findings of the Conference, Vienna, Austria, September 18-19, 2004, available at IAEA Web site, www.iaea.org. IAEA-CN-156/F-12 High density fuel development for Research Reactors 1) 2) Patrick Lemoine , Dan Wachs 1) CEA, CEN-Saclay, 91191, Gif sur Yvette, France 2) Idaho National Laboratory, P.O Box 2528 , Idaho Falls, ID 83415, USA E-mail addresses of the authors: patrick-marie.lemoine@cea.fr daniel.wachs@inl.gov An international effort to develop, qualify, and license high and very high density fuels has been underway for several years within the framework of multiple national RERTR programs. The current state of development is the result of significant contributions from many laboratories, specifically CNEA in Argentina, AECL in Canada, CEA in France, TUM in Germany, KAERI in Korea, VNIIM, RDIPE, IPPE, NCCP and RIARR in Russia, ANL and INL in USA. These programs are mainly engaged on UMo dispersion fuels with densities from 6 to 9 gU/cm3 (high density fuel) and UMo monolithic fuel with density as high as 16 gU/cm3 (very high density fuel) The limitations discovered under severe (high power) operating conditions during tests on first generation high density fuel, postponed initial plans and schedules for qualification tests and the subsequent licensing program [1], [2], [3]. These failures, firstly observed in French FUTURE experiment (see Fig. 1 & 2), and confirmed in the US miniplates and Russian tubes, are clearly attributed to the excessive formation of the interaction product and the inability of this compound to retain fission gas in the form of stable bubbles. Fig.1: FUTURE metallography in the pillowed region (plate transverse section) Fig.2: FUTURE metallographies with (UMo)Alx interaction product around the UMo particles, lenticular shape porosities at the interfaces with Al matrix, and final meat decohesion Thus, it was clear that the only way to improve UMo dispersion fuel behavior for medium to high power operating conditions was to drastically reduce the formation of the interaction compound. To solve the problem, full scientific and experimental programs were launched to understand the mechanism of UMo/Al interaction, develop and test in-pile solutions to avoid an inappropriate behavior under irradiation: 1. Thermodynamic studies to determine, through the U-Mo-Al ternary system, the phase’s equilibriums at different temperatures (CEA, CNEA, ANL) IAEA-CN-156/F-12 2. Metallurgical studies on diffusion couples to get a better understanding of the mechanisms of the UMo/Al interaction phenomena and underline the parameters which could prevent this reaction (KAERI, INL, CNEA, CEA, VNIINM) 3. Numerical simulation studies on effect of some additives in particles or matrix (CNEA, ANL) 4. High energy ion irradiation, typically 100 MeV Xenon, in order to simulate the damage created by fission products (TUM, CEA, INL). 5. In reactor tests on promising solutions (INL, ANL, CEA, KAERI, RIARR) The in-pile irradiation programs are specifically designed to evaluate and optimize the irradiation behaviour of UMo dispersion fuels (high density fuel) as well as UMo monolithic (very high density fuel). This experimental work will culminate with irradiation tests on full size plates that will demonstrate the integrated package that includes both manufacturing and irradiation aspects, as has been previously demonstrated by the French UMo Group A number of modifications to the fuel and/or the matrix aluminium, based on available literature and experience, have been evaluated specifically in US RERTR Experiments [4] and will be used to help select fuel compositions for future full size plate testing. The most promising remedies for dispersion fuel appears to be an addition of silicon to the matrix or the oxidation of the particles without any additive in Al matrix [5]. For monolithic fuel, tests are in progress to improve cohesion between foil and clad during and after irradiation. Promising solutions include the application of a silicon rich layer to the interface between the cladding and fuel, application of a zirconium diffusion barrier to the interface, or by using Zy cladding co-rolled with the UMo foil [6]. The final paper will give the details of this entire program and will provide an updated schedule for qualification and licensing of the different solutions. [1] J.L. Snelgrove, P. Lemoine, N. Arkhangelsky, P. Adelfang, “ Qualification and Licensing of U- Mo fuel”, Proceedings of the 7 th Int. Topical Meeting Research Reactor Fuel Management, Aix en Provence, France, March 9-12, 2003. [2] P. Lemoine, J.L. Snelgrove, N. Arkhangelsky, L. Alvarez, “UMo dispersion fuelk: Results and status of Qualification Program”, Proceedings of the 8 th Int. Topical Meeting Research Reactor Fuel Management, March 21-24 2004, München, Germany. [3] J.L. Snelgrove, P. Lemoine, L. Alvarez, “High density UMo fuels – Latest results and reoriented Qualification Programs”, Proceedings of the 9 th Int. Topical Meeting Research Reactor Fuel Management, April 10-13 2005, Budapest, Hungary. [4] M. Meyer & al “Progress in the RERTR Fuel Development Program”, Pro ceedings of the 10th Int. Topical Meeting Research Reactor Fuel Management, 30th March -3rd May 2006, Sofia, Bulgaria. [5] S. Dubois, J. Noirot, J.M. Gatt, M. Ripert, P. Boulcourt, P. Lemoine, “Compr ehensive overview on French Experiments in Irradiation Tests and PIE on High density UMo/Al dispersion fuel”, Proceedings of the 11th Int. Topical Meeting Research Reactor Fuel Management, March 11-15, 2007, Lyon, France [6] D. Wachs, P. Lemoine, C. Jarousse, “Status on the LEU fuel development”, Proceedings o f the 11th Int. Topical Meeting Research Reactor Fuel Management, March 11-15, 2007, Lyon, France IAEA-CN-156/S-35 Measur ing Progress in Reactor Conversion and HEU Minimization Towards 2020 – the Case of HEU-fuelled Research Facilities Ole Reistad and Styrkaar Hustveit Norwegian Radiation Protection Authority Grini Naeringspark 13 1353 Oesteraas Norway Ole.Reistad@nrpa.n, oStyrkaar.Hustveit@nrpa.n o The primary impediment that prevents nuclear peroralitfion is the lack of access to fissile materials. Thus, a recognized objective internationally has been to minimize the use of HEU and reduce the number of locations with HEU present. Yet, nearing the 30 year anniversary of this objective, the number of HEU-fuelled research facilities in operatrieomn ains high, HEU is still being used in large quantities, and significant quantities of HEU is still to be found in a large number of unsecured locations worldwide. This paper identifies the smt iomportant indicators for measuring progress for the historical and future national and internoantai l efforts for research reactor conversion and decommissioning of vulnerable facilities. Figure 1. Categories of HEU-fuelled Reeasrch Facilities in Operation 1978 - 201 22 0 80 Critical assemblies 70 Fast reactors 60 Pulsed reactors 50 Steady-state reactors (<0.25 MW) 40 Steady-state reactors (1 MW > x >= 0.25 MW) 30 Steady-state reactors (2 MW > x >=1 MW) 20 Steady-state reactors (10 MW > x >2= MW) 10 Steady-state reactors (>=10 MW) 0 78 1 4 7 0 3 6 9 2 5 8 1 4 7 0 19 19 8 98 98 99 99 99 9 0 0 0 1 1 1 21 1 1 1 1 19 20 20 20 20 20 20 20 The primary impediment that prevents nuclear peroralitfion is the lack of access to fissile materials. Thus, a recognized objective internationally has been to minimize the use of HEU and reduce the number of locations with HEU present. Yet, nearing the 30 year anniversary of this objective, the number of HEU-fuelled research facilities in operatrieomn ains high, HEU is still being used in large quantities, and significant quantities of HEU is still to be found in a large number of unsecured locations worldwide. This paper identifies the smt iomportant indicators for measuring progress for the IAEA-CN-156/S-35 historical and future national and internoantai l efforts for research reactor conversion and decommissioning of vulnerable facilities. The present concern for HEU could be traced ba cthke t oInternational Fuel Cycle Evaluation (INFCE) initiative.3 The INFCE report recognized that there w oevrer 140 HEU-fuelled research reactors with significant power-output (between 10 kWt and 250 MWt) in operation in more than 35 countries, with nominal power in excess of 1700 MW annually, cuomnisng each year more than 1200 kg U-235.After re-establishing this baseline measurement, pthaipser assesses that similar figures for 2006 are 1150 MW, 72 facilities, primarily as part of the steady-state reactor categories given in Figure 1, with an HEU consumption in excess of 700 kg. In additiothne, potential for conversion of other types of facilities not included in the INFCE study, among orsth peulsed facilities and various types of critical assemblies, are discussed. When includinlgl HaEU-fuelled research facilities converted, commissioned or decommissioned after 1978; moaren t3h10 facilities are identified and considered when identifying reductions in HEU consumptio(en. g. converted vs. shutdown facilities). This assessment includes facilities within and outsidee sthcope of current international programs, in particular the Global Threat Reduction Initiative (GTRI) and the program for Reduced Enrichment for Research and Test Reactors (RERT4 RIn) .total, when also includin Hg EU-fuelled isotope production facilities, more than 150 HEU-fuelled facilities striell main in operation today. Additional measures for assessing the overall risk and need for conversio and, dinition to the different categories of material as suggested in InfCirc 224, Rev. 4, then considering also average core inventory and HEU consumption together with the fuel burn- uapre, suggested as part of this paper. Figure 2. HEU-fuelled Research Facilities ine Oraption in Different Global Regions 1978 - 2020 300 250 200 Europe China 150 US Russia and NIS Other 100 50 0 78 80 82 84 86 8 0 2 4 69 9 9 9 9 98 99 99 99 99 99 8 00 02 04 06 08 10 12 14 16 80 0 0 0 0 0 0 0 0 01 02 0 1 1 1 1 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 The main conclusion is that a more compreheen saipvproach to HEU minimization and conversion is needed to achieve a real HEU cl-eoaunt with respect to all types oref search facilities. While the idea of a HEU global cleanout indeed hgaasi ned increased political atteonnti, there is no clear and unified international policy and agreed measures on progarneds sc ompletion. Real progress in all relevant areas may not be expected until a higher degreoen osfe cnsus amongst all states on the final objectives is established. Again, the IAEA – maintaining atne rinational rather than a national focus on the HEU clean-out and minimization activities – seems the obvious forum for information exchange as well as IAEA-CN-156/S-35 technical support, coordination, and cooperation. On the political levelw, eredn ceonversion interest may be accomplished by a second International FCuyecle Evaluation initiative. Particular support should be given to the current initiatives sponsobrye dth e GTRI; for example the RERTR program is instrumental in assessing gaps and issues which incereeda sed attention. This paper suggests, based on the findings above, that a another GTRI globanlf ecroence should take place to facilitate the needed interaction and attention to prioritize the next steps and address urgent needs – for example in relation to spent fuel from HEU-fuelled facilities. 1 The statistics presented in this extended synopsis is primarily based different electronic and prionntesd o vf ers the IAEA Research Reactor Databawsew w( .iaea.org/RRDB). Specific entries on individual facilities may have been included on the basis of documentation worked out of individuals and/ or nationpaelt ecnotm authorities; however, this is thoroughly discussedt hine final paper in relation to each figure. 2 For research reactors, the power levels for categorizing research reactors suggested in International Atomic Energy AgencyT, he Application of Research React,o IrAsEA-TECDOC-1234, August 2001 has been applied in this paper. 3 INFCE, Advanced Fuel Cycle and Reactor Conc,e Rptesport of INFCE Working Group 8, IAEA, 1980, p. 137: “The trade in and widespread use of highly enriched uranium and the production of fissile matenrsiatiltsu tceo proliferation risks with which INFCE is concerned. Plifreoration resistance can be cinreased by: 1. Enrichment reduction preferably to 20% or less which is internation arellcyognized to be a fully adequate isotopic barrier to weapons usability o2f3 5U; 2. Reduction of stockpiles of highly enriched uranium; 3. Reduction of the annual production of fissile materials in research reactors, although attainment of weapobnlse- umsaterial would require spent fuel reprocessing. For example, for s oremsearch reactors fuelled with natural uranium the proliferation resistance might be improved by utilizing slightly enriched uranium, which reduces the annual plutonium production. ” 4 Global Threat Reduction Initiative, GTRI StrategPicla n 2007, January 2007 (as accessed April 17, 2007 http://www.nnsa.doe.gov/na-20/dso/Gc TRI_Strategic_Plan_2007.p) dafnd Dr. Armando TravelliF, uel Issues – Replacement of HE,U Presentation at the IAEA ScientifiFc orum 2004, September 2004 (as accessed http://www-pub.iaea.org/MCTD/Meetings/PDFplus/200/g4csfSess3-Travelli.pd).f Session 12: New Research Reactor Projects Synopses no. IAEA-CN- Synopses Title Main Author 156/ S-21 Commissioning the new OPAL Research Reactor Irwin, A. S-23 Reactor PIK status of construction Konoplev, K.A. Licensing of the OPAL Reactor during construction S-37 and commissioning Summerfield, M. S-24 Concept for a new research reactor in Ukraine Lobach, Y. IAEA-CN-156/S-21 OPAL: Commissioning a New Research Reactor Tony Irwin 1), Nestor De Lorenz2o) 1) ANSTO, Lucas Heights, NSW, Australia 2) INVAP, Lucas Heights, NSW, Australia E-mail address of main author: arr@ansto.gov .au In 1997, the Australian Federal Government decidoe dco tmmence the process of replacing ANSTO’s 10 MW HIFAR reactor, which had operated sei n1c958, with a new 20 MW multi- purpose reactor for radioisotope production, irradiateiornv icses and neutron beam research. The contract which included design, construction caonmdmissioning was awarded to INVAP in 2000. INVAP designed an open pool type reactorU, LfuEelled, light water cooled with a heavy water reflector surrounding the core. The designd inecsl ua cold neutron source. ANSTO established the organisation for the projienc ta ccordance with the IAEA Safety Guide for Commissioning of Research Reactors [1j]o. iAnt ANSTO-INVAP Commissioning Management Grou pensured resources were available and authorize ds ttahrt of each commissioning stage. The ANSTOR eplacement Research Reactor Project G rowuaps responsible for all contractual matters, overseetihneg construction, participating in pre- commissioning and acting as the focus for the slicinegn work. ANSTO also established a Commissioning Operations Gro utpo take part in commissioning. This group formede th nucleus of the final Reactor Operating organisation. INVAP design activities were based in Bariloche, eAnrtgina but throughout construction and commissioning, INVAP maintained a strong presenc Ae uinstralia, arranging for appropriate staff to be brought to site to support each stagAelt.h o ugh INVAP was responsible for commissioning, it was agreed during pre-commissigo ntinhat ANSTO Commissioning Operations Group would be involved in all commisnsiniog activities. In this way the group gained valuable experience, and by the time of lfouaedl ing was actually carrying out all operations, under the supervision of INVAP. In adodni tito training by INVAP and plant experience, an extensive classroom training progirnacmlu, ding training on a simulator, was completed before fuel loading. Establishment of ao dg o perating safety culture was of fundamental importance. The Operating organisation is based on the IAEAe tSya Gf uide [2]. ANSTO decided to maintain HIFAR operating 24/7 during the commissiniogn and early operation of the new OPAL reactor. There was not enough existing sta ffo ptoerate two reactors, but some experienced operations staff were able to be redle afsrom HIFAR to join the OPAL organisation. One of the success of the project two acso mbine these experienced staff with ten new young engineering/science graduates, pe wrsiothn sexperience of other reactors, staff seconded from other divisions on site and contract sta ffof,r mto an effective operating team. Stage A commissioning was completed in 74 daysa rinly e2006 without any major problems. All systems were operated with dummy fuel assems binlie the core to obtain the correct flow characteristics. IAEA-CN-156/S-21 In Stage B1 (5 days) , 14 of the 16 LEU fuel assemblies wered l oaandde the reactor taken critical for the first time on 12 August 2006. The main issurein dg Stage B1 commissioning was spurious trips from the nucleonics instrumentation t od uneoise. This problem was solved by close attention to earthing, connections and cable sincgre. en In Stage B2 (25 days), the core loading was completed and yt hneu ckle onics and reactivity parameters of the core measured at powers up to 400 kW. Moinboler mprs with the nucleonics instrumentation were solved by instrument tuning. Stage C commenced in October 2006 and the reactor power cwreaass iend in stages until the full power 20 MW was achieved on 3 November 2006. A probletmh twhei core outlet temperature measurement was resolved by a simple moidoinfi ctoa tthe primary cooling flow around the temperature detector. The cold neutron sourclde ncootu be tested in cold mode during this stage, but a problem with the turbine has now beeseonlv red and cold testing completed. The secondary cooling tower performance, aglhth aoduequate for full power operations, did not meet the design specification for ext rweemaether conditions in Australia and further system tuning has been necessary. During December, a degradation of the heavy water in thec troerf lweas observed. Tests identified minor leaks in a non-structural seal weld bentw tehe reflector vessel and the neutron beam tubes. As of April 2007, the reactor was stilla otipnegr at full power whilst INVAP proposals for repair were being discussed. In addition to producing thousands of project documents,A IPN Vproduced the design operations and maintenance manuals. Each manual has bieewne dre tvo incorporate commissioning experience. ANSTO established the OPAL Beuss inManagement System (BMS), obtained ISO 9001/ISO 4001 accreditation and the IPN VmAanuals are now being incorporated into this BMS. The Operating Limits and Conditions (OLC) are the basisa ffoer osperation of the facility. In drafting the OLCs, ANSTO used IAEA guidance and also an IAEvAie Rwe Team examined the OLCs in detail. [1] INTERNATIONAL ATOMIC ENERGY AGENCY, Draft Safety Gudie Commissioning of Research Reactors, NS-G-4.1 (NS 259), V ie nna [2] INTERNATIONAL ATOMIC ENERGY AGENCY, Draft Safety Gudie The Operating Organisation and the Recruitment, Training and Qfiucatlion of Personnel for Research Reactors, DS 325, Vienna [3] INTERNATIONAL ATOMIC ENERGY AGENCY, Draft Safety Gudie Operational Limits and Conditions for Research Reactors, NS-G-4.3 (NS), 2V6ie1nna IAEA-CN-156/S-23 REACTOR PIK STATUS OF CONSTRUCTION K.A. KONOPLEV Neutron Physics Department, Petersburg Nuclear Physics Institute, Orlova Roscha, 188300, Gatchina, Russia Report gives description of the current status of reactor PIK that is under construction at Petersburg Nuclear Physics Institute. Powerful research reactor has thermal neutron flux 5.1015n/cm2sec. Plans of complete construction work and putting reactor PIK in operation are considered. Second part of report gives results of critical experiments supported reactor first criticality program. Last years the rector construction has been financed in poor level less then 200 million rubles annually. We are waiting remarkable increase and prepare three construction phase for finalize design. The equipment for the first phase is mounted on about 90%. Practically all subsystems of the primary coolant circuit are completed. Water test on this systems and primary coolant circuit itself begins. Electrical and instrumentation systems also are in test. Fire control system and physical protection systems for the first phase partly are in operation. On the rector full-scale critical facility startup core and steps for increasing power are examined. The PIK reactor features high experimental potential not only due to the high intensity neutron beam, which is approximately by one order higher than that at the existing medium- power reactors, but due to availability of the sources of hot, cold and ultracold neutrons as well. Therefore, as compared to the research reactors created in 50-60s, the PIK reactor will give unique opportunities for extending the neutron beam research activities conducted currently in Russia, as well as for launching new researches that are technically impossible at the moment. Choice of Area - New research reactor projects. Type of presentation – Oral presentation. IAEA-CN-156/S-37 LICENSING OF THE OPAL REACTOR DURING CONSTRUCTION AND COMMISSIONING Mark Summerfield1) 1) ANSTO, Lucas Heights, NSW, Australia E-mail address of main author: msx@ansto.go v.au This paper presents a description of the licensaicntgiv ities associated with the construction and commissioning of the Australian Nuclear Scie nacned Technology Organisation’s (ANSTO) OPAL reactor. It addresses the Construc tLioicnence, the interface between ANSTO, INVAP (the contractor with responsibility for dgens iand construction of the facility) and the Australian Radiation Protection and Nuc leSarfety Agency (ARPANSA, the Australian nuclear regulator) during the constrounc toi f OPAL, specific licensing issues that have arisen during the construction and commisnsgio npirocess, and the Operating Licence Application. Particular emphasis will be given t htoe way in which the licensing process is integrated into the overall project program athned lessons learnt that may be of benefit to other licensees and regulators. 1. The OPAL Construction Licence The Application for the Facility Licence, Construocnt i Authorisation was submitted to ARPANSA in May 2001. This licence was required cinc oardance with the ARPANS Act [1] and Regulations [2] in order to commence councstitor n of OPAL. Following an extensive review process, the CEO of ARPANSA gra nthted Facility Licence, Construction Authorisation in April 2002. It incorporated 18 Leincce Conditions, some of which had significant impacts on the construction and commissio onfin OgPAL, as described below. 1.1 Licence Condition 4.6: Construction of Items Important to Safety ARPANS Regulation 54 states thaTth “e holder of a licence, or a person covered by a liec,e nc must not construct an item that is important forf estya, and that is identified in a safety analysis report, as part of the construction of an tcrolled facility, unless the CEO has given the holder, or the person, approval to construct the ”i.t e m It defined “items important for safety” as all thSea fety Category 1 and Safety Category 2 structures, systems and components. This licence condition had a significant impact tohne construction of OPAL, due to the number of items that required such approval. During the course of the construction of OPAL, o1ve2r0 individual submissions were made to ARPANSA under Licence Condition 4.6. The prepioanra tof such a large number of submissions by ANSTO and their subsequent reviewe avnaldu ation by ARPANSA imposed a significant workload on both organisations. 1.2 Licence Condition 4.7: Commissioning of Items Important to Safety Licence Condition 4.7 is an extension of Licence Condition, a4n.6d states that the approval of the CEO of ARPANSA was required to commission inidduival items important for safety as part of the cold commissioning of OPAL. However,t hine light of the experience gained in the implementation of Licence Condition 4.6, the CoEfO A RPANSA subsequently revised IAEA-CN-156/S-37 Licence Condition 4.7 such that it was split intoo twparts, one covering the overall arrangements for cold commissioning and the othdeern tifying specific systems and components for which detailed information about cthoemmissioning tests was required to be provided to ARPANSA. As a result, instead of the 120 plus submissionse ru nLdicence Condition 4.6, there was a single submission of the overall arrangements hfoer ctold commissioning, together with the formal submission of specific pre-commissioning acnodmmissioning test procedures as identified by the CEO of ARPANSA. A total of 24 tse swere specified by ARPANSA, including those of the first and second shutdowsnt esmys, the containment permeability, and the establishment of natural circulation cooling. 2. Licensing issues arising during construction As can be fully expected with a project of thise t,y pa number of licensing issues arose during the course of construction and cold commissioninf gO PoAL. The most significant with respect to their impact on the project were: • Discovery of a geological fault during site excavations • Concrete cracking in the reactor building basement • Reactor pool heavy water penetration cut-outs • Repairs to reactor pool 3. The OPAL Operating Licence The Application for the Facility Licence, OperatinAgu thorisation was submitted to ARPANSA in September 2004. It was sub-divided into 5 par ftos llaosws: Part A: General information on the purpose and locatione o Rf tehactor Facility Part B: The plans and arrangements for managing safety Rofe tahcetor Facility Part C: The Safety Analysis Report (SAR) for thea cRtoer Facility, together with associated safety and licensing documentation Part D: The Operational Limits and Conditions (OLCs) for R theeactor Facility Part E: The plans and arrangements for hot commissionin Rge tahcetor Facility The Application was supported by a large volumed ofc umentation, including detailed engineering and analysis reports supporting the ,S AthRe plant design, operation and maintenance manuals prepared by INVAP and the OPBAusLi ness Management System (BMS). Also considered as part of the supportingu dmoecntation to the Application were all the submissions made under Licence Condition 4.6. The CEO of ARPANSA granted the Facility Licence, Oraptieng Authorisation in July 2006. This operating licence authorised ANSTO to load l, fupeerform hot commissioning and operate OPAL. In addition to the Licence Conditi oinhserent in the ARPANS Act and Regulations, it also identified six Licence Condnitsio covering periodic safety reviews, periodic security reviews, safety culture, quartererlpy orting, discharge authorisations and an index of licensing documentation. [1] Australian Radiation Safety and Nuclear Safety (ARPANSt) 1 A9c98, [2] Australian Radiation Safety and Nuclear Safety (ARPANSg)u Rlaetions, 1999 IAEA-CN-156/S - 24 Concept for a new resear crehactor in Ukraine I.N.Vishnevsky, V.V.Davidovsky, E.U.Grinik, M.V.Lysenko, P.G.Litovchenko, Yu.N.Lobach, V.N.Makarovsky, V.N.Pavlovich, E.V.Svarichevskaya, V.V.Trishin, V.N.Shevel Institute for nuclear research of NASU, Kiev, Ukraine E-mail address of main autholro: bach@kinr.kiev.u a Currently in Ukraine there is only one high-power research reactor – the WWR-M – in operation. The final shutdown of this reac tiosr expected in 2015 and, therefore, the construction of a new research reactor is uncduerre nt consideration. The first stage was to develop a Concept for a new multi-purpose research nuclear reactor. The Concept substantiates the need for a neswea rech reactor in Ukraine and determines the optimum alternative. The Concept definese tbhasic requirements for the reactor’s design, construction and operation at an appropriatet ys alefevel. The Concept involves considerations about the reactor type and relevant technpicaarla meters, the reactor’s main uses, possible locations, and the necessary scientific and technical infrastructure. The following prerequisites underly the Concept: / the new reactor is intended as a replacement for the existing WWR-M reactor; / the new reactor will be designed and constetrdu cwith the aim of satisfying current and future needs of the State as a power source of neutrons; / it will have multi-purpose use for both fundamental and applied investigations; / the new research reactor will be an integral part of the Ukrainian nuclear power industry; / the design will follow modern internationale ntrds for increased reactor utilization whilst maintaining all nuclear and radiation safety requirements; / the benefits of construction to the domestic industry and economy; / the new research reactor should be the core installation of a new national nuclear centre. The main aim of a national nuclear centre baasreodu nd the new research reactor will be an ef- fective infrastructure to investigate and develop technology in the following areas: / nuclear power (primarily, testing of structu mraal terials for the power reactors under neutron irradiation; testing of new materials, heart-rciears, fuels, devices and operational regimes); / fundamental science (neutron physics and enaurc sl pectroscopy; high-temperature semicon- ductors; biology, chemistry, medicine, ecology etc); / applied nuclear technologies (radionuclideo dpurction for medicine and industry; silicon doping; neutron-activation analysis; steritlizoan technologies; modification and strengthen- ing of polymers, metals and alloys etc); / personnel training and retraining. The reactor is expected to be a pool-type wthitehrmal power in the range of 20-30 MW and having an averaged neutron flux about 2.0·-140.014 n©cm-2©s-1. Low-enriched uranium fuel will be used for the reactor operation. Reactor taencdh nological systems will be located in the containment. The following laboratories and complexes are planned around the reactor: / laboratory for the radionuclide production; / laboratory for the neutron doping of silicon and another materials; / laboratory for the neutron-activation analysis; IAEA-CN-156/S-24 / material testing complex; / medical complex. The selection of the reactor site will be madcec ording to established procedures at the design stage. The suburbs of the town of Slavutichth ien Kiev region are considered now as the most preferable location; however, some othera cpels should be considered too. Preliminary estimates are that the area of reactor siiltle b we about 40-50 hectares. Taking into account the duration of the reactor’s construction, operation and decommissioning, the term of site use will be about 100 years. Implementation of the Concept tasks is foreseen during 2007-2015. Poster Presentations: Safety Synopses no. IAEA-CN- Synopses Title Main Author 156/ S-9 Safety analysis of a 1-mw pool-type research reactor Hamidouche, T. Safety Core Parameters Prediction in Research Reactors using Artificial Neural Networks: A S-12 Comparative Study of various Learning Algorithms Mazrou, H. Determining Degradation of Sensors used for Fuel Assemblies Cooling Temperature measurement by the S-18 Analysis of Thermal Fluctuations Noise Saadi, S. Physical calculations for the design of irradiation device and nuclear heating calculation in silicon ingot S-20 at Es-salam research reactor Widad, T. S-43 Upgrading I&C for the Es Slam Research Reactor Djaroum, B. Training and Qualification of Reactor Operating U-48 Personnel Kulak, R. Experimental Heat Transfer Analysis of the IPR-R1 S-26 TRIGA Reactor Mesquita, A. Appraisal for the Operation Safety of SPRR-300 in U-1 recent years Chen, W. Calculations and Measurements at the Training S-14 Reactor VR-1 Rataj, J. Improvements of the Training Reactor VR-1 – Only S-54 Way how to be Attractive for Students and Scientists Sklenka, L. Study of temperature effects on ULYSSE reactor for S-44 training and qualification of operating personnel Foulon, F. Monte Carlo Simulation of GRR-1 Core and Neutron S-17 Irradiation Positions Using MCNP Stamatelatos, I.E. S-28 Radiological characterization of the GRR-1 pool Tzika, F. The Next 20 Years Operation of the 36 Years Old S-71 Hungarian Training Reactor Aszódi, A. Validation of the Monte Carlo Method of MVP Code on the First Criticality of Indonesian Multipurpose S-15 Reactor Sembiring, T.M. Neutronic Analysis of RSG-GAS Silicide core with S-16 Uranium density of 4.8 gr/cc Setiyanto Fault Position Estimation Using Color Segmentation Approach to the Thermal Infrared Image of a Cabling S-48 Network in the Nuclear Reactor Nugroho, D.H. Water Chemistry Surveillance for Multi Purpose S-56 Reactor 30MW GA Siwabessy, Indonesia Sunaryo, G.R. Assessment of qualification of RSG-GAS Reactor U-51 operator in INDONESIA Pandi, L.Y. Developing an Ultrasonic NDE System for a Research S-49 Reactor Tank Perets, Y. Neural networks application to CRDM and S-70 thermohydraulic data validation Sepielli M. Investigation of JRR-3 control rod worth changed with U-32 burn up of follower fuel elements Hosoya, T Project to Replace the Control and Protection System S-72 at the WWR-K Research Reactor Chakrov, P. Modeling of Thermal Hydraulics Behaviour in Reactor S-11 Core Mohamed, K.N. PUSPATI TRIGA Reactor: 25 Years of Safe Operation and Strategies for Ensuring Safety and S-67 Security Z. Masood Core Calculation of 1MW PUSPATI TRIGA Reactor (RTP) using Continuous Energy Method of Monte S-68 Carlo MVP Code System Abdul Karim, J. Operational Experience and Programmes for Optimal S-10 Utilization of the Nigeria Research Reactor-1 Jonah, S.A. Regulatory control of the Nigerian Research Reactor S-32 (NIRR-1) Ogharandukun, M. O. Comparative dose calculation for TRIGA HEU and S-47 LEU fuel in nuclear accident situations Margeanu, S. Safety Analysis of MNSR Reactor during Reactivity S-8 Insertion Accident Using the Validated Code PARET Hainoun, A. Fuel management methodology upgrade of Thai Research Reactor (TRR-1/M1) using SRAC computer S-19 code Tippayakul, C. IAEA-CN-156/S-9 IAEA-CN-156/S-9 IAEA-CN-156/S-9 IAEA-CN-156/S-9 IAEA-CN-156/S-12 IAEA-CN-156/S-12 IAEA-CN-156/S-18 Determining Degradation of Sensors used for Fuel Assemblies Cooling Temperature measurement by the Analysis of Thermal Fluctuations Noise S.Saadi, A.Benali. Nuclear Research Center of Birine (CRNB), BP 180 Ain-Oussera 17200, Djelfa, Algeria. E-mail address of main author: saadisdz@yaho o.fr 1.Abstract: This work describes opportunities for continuous passive surnvceeil lwa ith noise analysis as the methodology, it is aimed to detect sensor degradation througho nrsees ptime factor calculation, to determine whether sensors have the response timiens thweit hplant safety analysis definitions, to ensure that the safety limits are ncoet eedxed. The noise analysis techniques provide the capability for on-line monitoring of the rnesep otime of the installed sensors. 2. Introduction: Reactor control is usually accomplished by thileis aution of sensors. The aim of these sensors is to reflect the exact behaviour of the physical procesdse ru snurveillance; and to allow the user to ensure a well functioning of the plant under control. The characteristics and properties of the sensors being used are critica lw fahcetnoersver safety and high efficiency are required, especially when dealing with sophisticated plants such as nuclear plants. In fact, the achievement of these two requirements necse sas ictaotnetinuous passive surveillance of sensors, however, this is traditioancacllyo mplished by periodic tests perturbing the sensors and this direct sensor testing requires asn taoc caell sthe sensors. The direct sensor testing, as it is shown in figure 1, is usually performed by agp ap lykinown signal to the sensor under test to determine the sensor response time faosrm aa pnecer testing factor. Drawbacks of this method are: 1 - It requires an access to each sensor; therefore it must be performedh ew rheeanc tor is at a cold shutdown. 2 - It does not provide response time at normal operating connsd, itsio it does not yield the in-service response time of the sensor. 3 - Since sensors are continuously subject to degradations in their perform ita insc reesq,uired that the frequency of periodic tests must be increased, which is not caplrlya ctaind economically efficient. Therefore, the need for a passive on-line methro ds efnosor performance testing is more than a requirement. The noise analysis technique, which will be presented in detauilr inw ork, provides the means for implementing a passive sensor surveillance, that yields thev iinc-es erersponse time, on the one hand, and that overcomes the drawbacks of the direct technique, on the other hand. Known input Sensor Output test signal under test signal Comparison of input and output signals Determina ti o n o f r e sp onse time Figure 1: Traditional (direct sensor testing) IAEA-CN-156/S-18 3. Noise analysis technique: Sensor response time testing is performed in the reactor tom dineet ewrhether sensors have the response times assumed in the plant safety analysis. Since re stipmoen sdeegradations are known to occur, these tests must be performed periodically utore e tnhsat safety limits are not exceeded. Furthermore, since the response time of some sensors denp ethned sp roocess condition and installation, the tests must be performed amt anl oorperating conditions to yield the in-service response time of the sensor. The application of the noise analysis technique to response time estimsa stihoonw in in the block diagram of figure(2). Plant under control: Se nsor display Signa l + noise Response time Validation Identification estimation Testing if response time is within safety If out off limits: sensor performance limits degradation Figure 2: On-line sensor surveillacne using noise analysis techniques. 4. Results: A program was established and initial tests were performed toa etev alolung term performance of resistance temperature detectors RTD of the type used in our research reactor to measure the difference in temperature in the fuel assemblies. The effort addressed th oef eafgfeincgt on RTD calibration accuracy and response time. Sensors have been simbuy lastmeda ll random heat generators. For pressure transmitters, we used a hydrauli c greanmeprator in the laboratory. The in situation response time testing of FOXBORrOce fobalance pressure sensors involves tricking the sensor, through the removal ainidtia rteing of power to the sensor, into the simulation of a pressure perturbation. The responsie nttr aof the sensor is measured, and the ramp response calculated. This transient is identical to oantee dg ebnye ar step change in pressure. This test can be completed in ten minutehs efr ocomn trol room with the reactor operating at full power. [1] Candy, J.V. “Signal processing: theo dmern approach”. Newyork: McGraw Hill. 1988. [2] Hashemian, h.m. Thie.A.J, and Updyaya, B.R “Reactor sensor surveillance using noise analysis”. In nuclear science engineering.96.98.102. 1988. [3] Kellaf.A.”Mesure du temps de répondses thermo sondes”. Rapport interne, CDSE. 1991. [4] Jenkins, G.M. Box, G, E, and P.”Timer isees analysis: forecasntgi and control”. San Francisco: Holden-Day. 1970. [5] Parzen.E.“An approach to time series". California. Stanford University (pages 951 to 989). 1961. [6] Steven. M.K.Standy.L.M.”Spectrum analy:s ais modern approach”. In proceedings of the IEEE. Vol 69.NO.1N1.OV. 1981. [7] Tokhi. M.O. Leith.R.R.”Active noisec ontrol”. Oxford :clarendon press. 1992. IAEA-CN-156/S-20 PHYSICAL CALCULATIONS FOR THE DESIGN OF IRRADIATION DEVICE AND NUCLEAR HEATING CALCULATION IN SILICON INGOTS AT ES SALAM RESEARCH REACTOR W. TITOUCHE 1), T. ZIDI 2), M. SALHI 1) (1) Centre de Recherche Nucléaire de Birine, BP180, Ain oussera 17200, W. DJELFA, Algeria (2) Commissariat à l'Energie Atomique (COMENA) kouid_wid@hotmail.com t/itouche_ikhlas@yahoo. f r Abstract To insure homogeneity of the neutron flux through Silicon irradiated at a Es-salam research reactor an irradiation device have to be conceived. The main criteria of its design are to fulfil all the necessary safety restrictions and recommendations while inserted into the core of the reactor. The excess of reactivity has to be also evaluated as well as nuclear heating in irradiated material. This work consist on the evaluation of excess reactivity misleads by the insertion of the device in the reactor core. To do so, the reactor core has been divided into several homogeneous regions and a cartesian three dimensional XYZ model of the reactor core was performed using the three dimension diffusion code CITATION [1, 4]. The fewgroup library was generated by WIMS/D4[2,4]. The obtained excess reactivity of the irradiation device is negative, and with reg ards to neutronic consideration some positive reactivity is needed to make up for its insertion. This can be achieved through appropriate choice of the control rods insertions. As for nuclear heating in the Silicon both neutron and photons contribution have been considered. The 1D transport code ANISN [3, 4] was used to perform the heating calculations. We have choice Pl= 3 order of scattering, S8 quadratic order and a distributed source in 207 energy group, 171 neutron group and 36 photon group to perform the calculations. As it was expected photon heating is predominant on neutron part. This will allow us to neglect the neutron contribution to the heating and performing the design of the necessary cooling system. IAEA-CN-156/S-20 In this paper, the adopted methodology is specified and the obtained results will allow us to decide on the cooling system and the rate of coolant to be fulfilled, also the necaecstisoanrsy to maintain the reactor critical can be drawn. Keywords: Silicon, Irradiation device, Excess reactivity, Nuclear heating. References [1] Askew and al., A General Description of the Lattice Code WIMS, J. Brit. Nucl. Eng. Soc. 5, Oct., (1966). [2] Fowler and al. CITATION, A Nuclear Reactor Core Analysis Code, ORNL-TM-2496, 1972. [3]. ANISN/PC Manual, D.Kent Parsons, ORNL-RSIC-CCC-514. [4]. MTR_PC v2.6 package, IAEA-1356 IAEA-CN-156/S-43 Upgrading I&C for the Es Salam Research Reactor B. Djaroum, S. Ait Mohammed, Y. Tireche, A. Benali, D. Merrouche Centre de Recherche Nucléaire de Birine, BP 180 Ain Oussera 17200, Algeria. E-mail address: djaroum_bakiri@yahoo.fr The Es Salam research reactor was built at the nuclear research centre of Birine, 36 kilometers from Ain Oussera, Algeria. Es Salam, whose construction started in 1987, reached criticality for the first time on February 1992. The load was gradually increased until the reactor reaches its full power during performance tests in July 1992. The Es Salam is a multi-purpose research reactor intended for the production of the radio elements, material tests, education and training. It serves also as a source for neutron beams used by chemists, biologists, metallurgists and physicists for fundamental research and applications. The Es Salam reactor is a tank type reactor of a nominal power of 15 MW(th) 14 and of a neutron flux of 2.10 (n/cm2-s). It is moderated and cooled by heavy water and has graphite reflector. Es Salam offers experimental possibilities in the fields of neutron spectrometry, production and materials studies. The main task of the Es Salam Reactor Operation Division is to ensure a safe and reliable operation of the reactor. With this objective in hand, the Operation Division is responsible for updating the safety measures and conditions in the installation for the reactor equipments and systems. Due to the increased demand for experiences and to the ageing effects, modification and modernization of some safety items become necessary in comparison with state of the art installations. Furthermore, the technological advances and the development and introduction of new instruments, components and systems increased the need for modification. In this paper, we present the new I&C system which will be used for the Es Salam research reactor. The Es Salam I&C System At present, the Es Salam is still operated with the original I&C system which employs a technology from the 80’s. This original, aged and analog, I&C system used discrete- component, solid-state devices and is no longer state-of-the-art and is in need of renovation and upgrade. A schematic diagram of the existing I&C system is shown in FIG. 1 below. I &C upgrading The existing analog I&C system becomes increasingly more difficult to maintain and to upgrade because of obsolete, ageing equipment and spare part procurement problems. It has been thought necessary to modernize and replace it with a digital control system which improves reliability, offers greater flexibility, increased data availability and enhanced features such as automatic self-testing and diagnostics. The I&C modernization project for the Es Salam research reactor will be developed and implemented using distributed computer control with data communication networking possibilities. A modernization program is thus in preparation and will be launched soon. IAEA-CN-156/S-43 The Replacement I&C system (see FIG. 2 below) includes the following equipment and systems: Digital reactor protective system, Digital power regulation system, Digital monitoring system, Fault diagnosis system, and Operator Station. FIG. 1. Schematic Diagram of the Existing I&C System. FIG. 2. General Layout of the New Digital I&C System. [1] INTERNATIONAL ATOMIC ENERGY AGENCY, Management of Research Reactor Ageing, IAEA-TECDOC-792, IAEA, Vienna (1995). [2] INTERNATIONAL ATOMIC ENERGY AGENCY, Modern Instrumentation and Control for Nuclear Power Plants, Technical Reports Series No. 387, IAEA, Vienna (1999). [3] INTERNATIONAL ATOMIC ENERGY AGENCY, Int’l Conf., Safety of Research Reactors, Topical Issues Paper No.4, 3-6 September 2001, IAEA, Vienna. [4] Operating Drawing of MHWRR, Instrumentation and Control System. [5] Rapport sur la durée de vie des centrales nucléaires et les nouveaux types de réacteurs par MM. Christian Bataille et Claude Birraux, Paris le 07 Novembre 2002. [6] ARAPAKOTA (D.), XING (A.). - Advanced Control and Operator Interface Systems for CANDU 9 Fuel Handling System. Conférence annuelle, Société Nucléaire Canadienne, 18-21 oct. 1988. IAEA-CN-156/U-48 Training and Qualification of Reactor Operating sPoenrnel R Kulak, P Walsh, D Vittorio Richard.kulak@ansto.gov.au Australian Nuclear Science and Technology Organisation PMB 1 Menai NSW 2234 Sydney, Australia Synopsis INTRODUCTION The training and qualification of OPAL Reactor Operations pers oinsn beal sed on IAEA guidelines which recommend a Systematic approach to tgra. in Tinhis involves performance and task based training which when used correctatlby lisehses and maintains the competency and qualifications of staff. OPAL oopre rtarat ining is broadly grouped into classroom theory training, simulator training aancdti cparl plant training. The training is carefully structured to ensure tha t stehlected candidate fulfils not only his personal ambitions but also the organisationsd sn. e eOur discussions will outline this approach ROLES OPAL Operations shifts consist of a Shift Manager, Reactor Operantdo ras a Utilisation Operator. The Shift Manager position is occupied by a professional engoinre secri entist with extensive experience in the reactor operator role. The Reacpteora tOor and Utilisation Operator position is generally occupied by a techllny icaompetent person have a trade/post trade qualifications. The responsibilities of each of these roles aims to achievere qthueir ed breadth of activities to enable safe and efficient operation of the orer aacntd utilisation of the irradiation facilities. RECRUITMENT The recruitment of Shift Managers and Reactor Operators f ocro tmhemissioning and initial Operation of OPAL commenced in 2004, three years prior hteo t commencement of commissioning. The first Reactor Operatonri ntgra ci ourse took place over 6 months in 2006, and full 24 hour shift manning commenced in July 2007, prior to fuel loading. TRAINING PROGRAM The training program is comprised of an appropriate combinationl aosfs rcoom instruction, simulator training, practical hands on sessions and suepde rsveislf-study. In addition to this, fundamental safety principles are also incaotrepdo r into the training to promote and improve safety culture and good work practices. There are six basic components to the training which are dc aorruiet in a staged process: IAEA-CN-156/U-48 1. Reactor Fundamentals 2. Classroom Design and Operations Theory 3. Simulator Training 4. Practical Training 5. Ongoing Training consisting of Safety Culture awareness, goodk wor practices and facility update training 6. Shift Manager rotation Shift Managers undergo additional detailed training in the areasm oefr gency response, safety analysis and application of Operating Limits and Con.d itions Training is assessed by written examinations, panel intervieimwus,la stor assessment and completion of competency based checklists. Training needs are identified in order to prepare training plancsh w ahrie incorporated into the Quality Management System. The prime objective ionfi ntrga is to ensure the safe and efficient operation of the reactor. IAEA-CN-156/S-26 Experimental Heat Transfer Analysis of the IPR-R1 TRIGA Reactor Amir Zacarias Mesquita Nuclear Technology Development Center (CDTN), Belo Horizontez,i lB ra E-mail address: amir@cdtn.br The heat generated by nuclear fission is transferred from lfeumele ents to the cooling system through the fuel-to-cladding gap and the cladding to coolant interfaches .f uTel thermal conductivity and the heat transfer coefficient from the claddingh et oc ot olant were evaluated experimentally. A correlation for the gap conductance betweenu ethl ea nf d the cladding was also presented. As the reactor core power increases, ther ahnesafet rt regime from the fuel cladding to the coolant changes from single-phase natural convection otoo lseudb ncucleate boiling. Results indicated that subcooled boiling occurs at the cladudrinfagc es in the central channels of the reactor core at power levels above approximatkeWly .6 0 The IPR-R1 TRIGA Nuclear Research Reactor (Fig. 1), ilnesdt aal t Nuclear Technology Development Center (CDTN), is a pool type reactor cooled by n actiurcraulation, and having as fuel an alloy of zirconium hydride and uranium enriched at 202%35U in. The core contains 59 aluminum-clad fuel elements and 5 stainless steel-clad feumele enlts. One of these steel- clad fuel elements is instrumented in the center with threme othceoruples (Fig. 1). FIG 1. Core upper view with the instrumented fuel element in ring B and instrumented fuel element scheme The objective of the thermal and hydrodynamic projects of the re aisc ttoor sremove the heat safely, without producing excessive temperature in the fuel enltesm. eThe regions of the reactor core where boiling occurs at various power levels c adne tbeermined from the heat transfer coefficient data. IAEA-CN-156/S-26 The thermal conductivity k() of the metallic alloys is mainly a function of temperatuInre . nuclear fuels, this relationship is more complicated becak uasles o becomes a function of irradiation as a result of change in the chemical and physicapl ocsoitmion (porosity changes due to temperature and fission products). Many factors affectu ethl et hfermal conductivity. The major factors are temperature, porosity, oxygen to mteotmal raatio, PuO2 content, pellet cracking, and burnup. The second largest resistance to heat conducthtieo nfu ienl rod is due to the gap. Several correlations exist [1] to evaluate its valupeo wine r reactors fuels, which use mainly uranium oxide. The only reference found to TRIGA reactorls i sfu Geeneral Atomic [2] that recommends the use of three hypotheses for the heat rt rcaonesfffeicient in the gap. The heat transfer coefficienht )( is a property not only of the system but also depends on the fluid properties. The determination oh f is a complex process that depends on the thermal conductivity, density, viscosity, velocity, dimensions and specifaict . hAell these parameters are temperature-dependent and change when heat is being transrofemrr ethde f heated wall to the fluid. An operational computer program and a data acquisition andl spirgoncaessing system were developed as part of this research project [3l]o wto oanl line monitoring of the operational parameters. Subcooled pool boiling occurs above approximately 60 kW on the cladding siunr fathce central channels of the IPR-R1 TRIGA core. However, the high thraenastfer coefficient due to subcooled boiling causes the cladding temperature be quite uniform maolosnt gof the active fuel rod region and do not increase very much with the repaocwtoer r. The IPR-R1 TRIGA Reactor normally operates in the range from 100 kW untial xai mum of 250 kW. On these power levels the heat transfer regime between the ucrlfaadc es and the coolant is subcooled nucleate boiling in the hottest fuel element. Boiling rhaenastf et r is usually the most efficient heat transfer pattern in nuclear reactors coreA [n4o].t her important aspect of the reactor operation safety is that it is far from the occurr eonf ceritical heat flux [5]. [1] TODREAS N.E., KAZIMI M.S., “Nuclear Systems I: Therml Haydraulic Fundaments”, Hemisphere Publishing Corporation, New York, (1990). [2] GENERAL ATOMIC, “Safeguards Summary Report for the New YoUrnki versity TRIGA Mark I Reactor”. (GA-9864). San Diego, (1970). [3] MESQUITA A. Z., “Experimental Investigation on Temperatures trDibiutions in a Research Nuclear Reactor TRIGA IPR-R1”, Ph.D thesis, Usniidvaedre Estadual de Campinas, São Paulo, (in Portuguese), (2005). [4] Duderstadt J.J, Hamilton L.J., “Nuclear Reactor AnalysJiso”h, n Wiley & Sons, Inc. New York, (1976). [5] MESQUITA A.Z., REZENDE H.C., “Experimental Prediction tohfe Critical Heat Flux on the IPR-R1 TRIGA Nuclear Reactor”. Proceeding ondf W3 orld Triga Users Conference, Belo Horizonte, August 22 to 25, (2006). IAEA-CN-156/U-1 Appraisal for the Operation Safety of SPRR-300 in Recent Years CHEN Wei Institute of Nuclear Physics and Chemistry, China chenwei_roy@yahoo.com SPRR-300 reached first criticality in 1979. During its 27 years of safe operations, the reactor has never been renovated comprehensivelyt.h Se oc ondition of the devices and the reactor must be analyzed to appraise the safety of the operation. A management networks for the safety athned quality of the reactor operation were built within the operation department. A geneorault line for quality of SPRR-300 during operation stage went into effect in 2004. A set of filesv eh abeen written or rewritten to ensure that the outline is operated smoothly. The rules and raetgiounl s of the reactor are insisted on. The emergency scheme and its execution programs had been established around 2002. Emergency practices were held in 2003, 2004, and 2006. Throthuigsh s eries of practices, the abilities for emergency treatment of the staff have been boosted continually. The minimalist technical transformation tohfe reactor had only been done once around the year of 2002. The failure ratio of the reac gtores down year by year. Before 2002, 22 out of the 31 scrams occurred without even one wnagr nsiignal showed at the same time. Since the minimalist technical transformation finished,a t hkind of scram has not occurred any more. The reactor resumed normal operation in 2003t.h iIsn year, there were 27 scrams, in which nine were caused by staff faults. 18 scrainm 2s0 03 were cause by devices breakdowns, while most of them were originated from malfutinocns of the old sets of power protection instruments. There were 14 scrams in 20a0n4d 2005. Five of them were caused by staff faults. More than half of the nine scramcasu sed by devices breakdowns were related to the electricity net lapse. It is reasonable that there are small quantities of manual failure. Every time before and after operation, the group of technical managemwenotu ld organize some kind of meeting to discuss the problem that may appear or hapvpe ared. With the continued fostering of safety culture, the operation quality of the staff insh eanced constantly. To those old reactors like SPRR-300, devices are the key of the reascatofer ty although some suitable modifications have already been done. In this period, ususaollmy e new kinds of device failure would turn up. So it is a must that the quality of thef fs tsahould be enhanced steadily to solve every accident correctly and resolutely. There were 12 cases in 2003 that the power liseusp fpor ionization chambers gave alarms of lower voltage because the voltage of local power grid was too low. The reactor was removed of automatic operation afterwards becausem oaflf unction of the power regulate system. To solve this kind of hidden peril, the institute nmaaging the reactor began to work together with local power supply bureau to stabilize the vgoelt aduring every periodo f operation. Also in 2003, the second regulate rod system workbendo ramal two times. Its lifting speed was much lower than that of normal. Itnh e first case, the reason was that the contact of its speed-adjust electric resistance was poor. In another caser,e tahseo n was short-circuit between the turns of the stator coil of the servo motor. Once in 2004 the drop-fuse of one transformer outside the reactor building struck arc. The reactor wfaosrc ed to be shut down temporarily so the transformer could be disconnected to be repaired later. IAEA-CN-156/U-1 The bad installation of the second motor for the ventilation of the coolant tower was one of the inherent defects in SPRR-300. The retarder lapsed eventually. Thorough repair of the retarder had been accomplished in 2005. As a result, the most of threats to the norompael ration of the reactor come from instability of power grids outside the reactor building. Becatuhsee inherent safety of the reactor and the developing of the ability for accident treatmenith win the reactor staff, instability of power grids can only influences the usages of the orer atcot a certain extent. It will not produce an effect on the safety of the reactor. Through maintaining and repairing in accordanciteh wcertain cases apart from overhauling at regular intervals for the devices, all systeomf sth e reactor are kept in good condition, which meets the demand to guarantee the safety orfe tahceto r. So long as the regulations are abided by, the training of the staff ist rengthened over and over, and the safety culture is fostered continually, the safety and the quality of the operation can be ensured. IAEA-CN-156/S-14 Calculations and Measuremenatts t he Training Reactor VR-1 Jan Rataj, Hubomír Sklenka, Karel MatEjka Czech Technical University, Prague, Czech Republic rataj30@seznam.cz The Paper presents basic information about verifying calculations of the experimental measurements at theVR-1 reactor and their evaluation. The training reactor VR-1 has been in operation since 1990. The VR-1 reactor is a pool-type light-water reactor based on low enriched uranium fuel IRT-4M with maximum thermal power 1kWth. The reactor is successfully used for training the students of Czech universities and preparing the experts for the Czech nuclear programme. The reactor VR-1 can be used for basic research, when respecting small power of the reactor. The calculation preparations and verifications of the experiments are integral part of the VR-1 reactor operation. The neutronics calculations and analyses are very often necessary condition to gain the authorization to experiments realization at the reactor VR-1. Deterministic codes WIMS and DIFER were used formerly for the neutronics calculations of the VR-1 reactor at the Department of Nuclear Reactor. Since 1998 neutronic calculations are performed by the statistical code MCNP. The Department of Nuclear Reactors currently works with the MCNP version 5. For the last nine years the MCNP model of the training reactor VR-1 has been permanently adjusted and improved. Now the MCNP model of the VR-1 reactor contains all important parts. The main reactor components in the model are: ‚ fuel assemblies IRT-4M (4-tube, 6-tube and 8-tube assemblies), ‚ 6-tube assemblies with control rods, ‚ fuel dummies, ‚ vertical and horizontal channels, ‚ external neutron source AmBe type, ‚ core grid, ‚ beryllium and graphite reflectors. MCNP model of typical VR-1 reactor core configuration is shown on Fig. 1. The main part of MCNP calculations is focused on different parameters and characteristics of the new core configurations at the VR-1 reactor, e.g. control rods worth calculation, determination of reactivity excess and calculations of neutron flux distribution. In Table 1 calculated and measured values of control rod worth are given. A lot of MCNP simulations are focused on verification of neutron activation analysis experiments or determination of delayed neutrons parameters. The Department of Nuclear Reactors made detailed criticality calculations for core configurations with LEU fuel using MCNP code. These calculations were the necessary condition to gain the authorization to operating the training reactor VR-1 with the LEU fuel IRT-4M. The department of Nuclear Rectors has already performed calculating verifications of 15 core configurations at the VR-1 reactor. The calculated results prove good agreement with experimental results. IAEA-CN-156/S-14 Fig. 1 MCNP graphical output - mdoel of the VR-1 reactor co re control rod MCNP Experiment worth [%] deviation worth [%] deviation B1 1.50 0.0848 1.37 0.0967 B2 2.11 0.0861 1.89 0.0606 B3 1.64 0.0878 1.46 0.0789 E1 1.22 0.0811 1.00 0.0482 E2 0.96 0.0901 0.93 0.0642 R1 0.46 0.0888 0.42 0.0601 R2 0.84 0.0849 0.90 0.0838 Table 1 Calculated and measured control rods worth [1] Rataj, J.: Criticality Calculations and Transient Analyses for the VR-1 Reactor with IRT- 4M LEU Fuel Assemblies, 2005 International Meeting on Reduced Enrichment for Research and Test Reactors, Boston, Massachusetts, 6-10 November 2005. [2] Skoda, R.: Sparrow Flies Lower, 2005 International Meeting on Reduced Enrichment for Research and Test Reactors, Boston Massachusetts, 6-10 November 2005 [3] Rataj, J. – MatEjka, K.: Schedule of the basic critical experiment with the core configuration C1 at the training reactor VR-1, CTU, Prague, July 2005. [4] Sklenka, H. - MatEjka, K.: The First Critical Experiment with a LEU Russian Fuel IRT- 4M at the Training Reactor VR-1, 2005 International Meeting on Reduced Enrichment for Research and Test Reactors, Boston, Massachusetts, 6-10 November 2005 IAEA-CN-156/S-54 Improvements of the Training Reactor VR-1 – Only Way how to be Attractive for Students and Scientists Lubomír Sklenka, Martin Kropík, Karel MatEjka Czech Technical University in Prague, Czech Republic E-mail address of main author: sklenka@troja.fjfi.cvut.cz The VR-1 Reactor is operated for training of university students and nuclear power plant personnel, R&D, and information services for non-military nuclear energy use. During last four years a large number of improvements has been achieved. The paper describes four of them which are the most important for the operation from the safety-point of view. 1. Conversion of the reactor and operation with LEU fuel During the autumn of 2005 VR-1 reactor has been successfully converted from the Russian IRT-3M fuel containing HEU (enrichment 36 % 235U) to the Russian IRT-4M fuel containing LEU (enrichment: 19.7 % 235U). This task was completed within the scope of the RERTR programme, that was initiated by the US Department of Energy, consistent with the global non-proliferation policy goal of minimizing the use of highly-enriched uranium in civil programmes worldwide. The fuel swap involved a direct cooperation of DOE and the RERTR programme, a NZCHK (Russian fuel supplier), IAEA, and a SOSNY (Russian company, which was repatriating the HEU fuel IRT-3M) [3]. During the first critical experiment with LEU fuel and the testing regime all criticality and neutronics characteristics of the LEU core were measured; hence the necessary conditions to license the core were fulfilled. Then the other reactor’s operational characteristics were measured, such as thermal neutron flux at different positions in the core or reflector. Also the reactor’s dynamics was studied, namely responses to positive, periodic and negative reactivity changes [4]. All the results confirm theoretical calculations when was expected that the reactor would behave equally with the new LEU fuel practically in a same way as with the HEU fuel [1]. The measurements confirmed the difference, that was expected in the calculations, of a lower neutron flux density with the LEU fuel compared with the HEU fuel. Training and teaching was already started during the testing regime for the undergraduate and graduate students. All standards experiments were implemented and the effects of the new LEU fuel were closely watched [1]. 2. Upgrade of the control and safety system The reactor was put into operation in the 1990. The original reactor I&C seemed to be obsolete and their replacement (upgrade) was being carried out. The upgrade is being done gradually during holidays in order not to disturb the reactor utilization during teaching and training. The new I&C substantially improves the reactor safety, the comfort of the reactor operation, and facilitates the maintenance. The I&C upgrades started in 2001 [2] with the human-machine interface, continued in 2002 with the control rod motors, drives and safety circuits upgrade. The control system upgrade followed in 2003 [2]. The next upgrade stage was the independent power protection system upgrade in 2004/05. The last stage is going to be the upgrade of the operational power measuring system in autumn 2007. The upgrades bring the reactor I&C to the top conditions and enable a IAEA-CN-156/S-54 prolongation of its functionality and maintainability for at least next 10 years. The new protection system consists of the operational power measuring (OPM) and the independent power protection (IPP) system [5]. The new OPM system is a full power range system that receives signals from fission chambers. The signals are evaluated according to the reactor power either in the pulse or current mode. The current mode utilizes DC current and Campbell techniques. The new IPP system working in the two highest power range decades receives signals from boron chambers and evaluates them in the pulse mode. The computers of both systems calculate the reactor power and power rate, compare them with the safety limits, and if they are exceeded, the safety action is initiated. There are 4 independent channels (3 in active mode) in each system and evaluated in the 2 out of 3 logic. The OPM and IPP systems are diverse different types and location of chambers, completely different hardware, software algorithms, development tools (PharLap based tools and Keil µVision 2 system) and teems for software manufacturing. Quality assurance, configuration management, verification and validation accompanied the software development [5]. 3. Upgrade of the radiation monitoring system The original radiation monitoring system STADOS was developed in the mid-80s. The new monitoring system RMS was implemented at the reactor in 2004. New RMS system is fully computerised state-of-art system connected with reactor information system. RMS system consists of 2 neutron and 10 gamma detectors, aerosol detector, external gamma detector located at the roof, control unit, archive system, user friendly LCD touch screen for control with all necessary information and connection with reactor console. 4. Upgrade of the physical protection system After the reactor conversion the nuclear materials category was decreased to the no. 3, but the physical protection system corresponding to the category no. 2 has not been changed. The new comfort and more reliable physical protection system has been installed at the reactor in 2006. The system respects all changes at the reactor hall during I&C upgrade and allows to protect the reactor hall at higher level using new technologies in the alarm communication and display between the physical protection system, reactor staff and respond forces. [1] Rataj, J. - Sklenka, H. - MatEjka, K.: Calculations and Measurements at the Training Reactor VR-1. International Conference on Research Reactors: Safe Management and Effective Utilization – extended synopsis, Sydney, Australia2007 [2] Sklenka, H. - MatEjka, K. - Kropík, M.: Improving of Safe Operation and New Open Policy at VR-1 Reactor. International Conference on Research Reactor Utilization, Safety, Decommissioning, Fuel and Waste Management, Santiago, Chile, 2003, Contributed Papers (Cd-rom). [3] Škoda, R.: 1st Full Conversion of Russian Design HEU Fuel: Sparrow Flies Lower. International Symposium on Minimalisation of Highly Enriched Uranium in the Civilian Nuclear Sector, Oslo, 2006 [4] Sklenka, H. - Škoda, R. - Rataj, J. - MatEjka, K.: LEU experience on the Sparrow training nuclear reactor. The RERTR-2006 International Meeting On Reduced Enrichment For Research And Test Reactors, Cape Town, South Africa, 2006 [5] Kropík, M. - MatEjka, K. – JuUí7ková, M.: New I&C at VR-1 Training Reactor. 15th International Conference on Nuclear Engineering, Nagoya, Japan, 2007 IAEA-CN-156/S-44 Study of temperature effects on ULYSSE reactor for training and qualification of operating personnel F. Foulon, G. Badeau, M. Dubois, J. Safieh CEA/Saclay, INSTN/UEIN, 91191 Gif-sur-Yvette, France Part of the French Atomic Energy Commission, the National Institute for Nuclear Science and Technology (INSTN) provides both academic and professional courses in all disciplines related to nuclear energy applications, including the physics and the operation of nuclear reactors. Theoretical courses and training courses on simulators are completed by experimental work on a training reactor which ensure a practical and comprehensive understanding of the reactor operation. For the training and qualification of reactor personnel and regulators, this global approach that includes experimental work contribute to the improvement of the safety of the reactor operation. We report here experiments conducted in this framework on the ULYSSE training reactor which is a 100 kW Argonaute type reactor. The main experiment developed here concerns the study of the temperature effects and the determination of temperature coefficients. ULYSSE reactor is a water moderated and cooled, graphite reflected system, using enriched MTR fuel elements. The core has an internal diameter of 90 cm. 24 fuel elements form an annular crown around a graphite reflector. Thus, both water and graphite play a important role in temperature effects. The global temperature coefficient αglobal corresponds mainly to the sum of the water αwater and the graphite αgraphite temperature coefficients. The experiment is carried out by maintaining the reactor critical at low power and adjusting the position of the regulation rod while the temperature of water is intentionally changed through the use of the different elements that constitute the primary and secondary water circuits. During the experiment, the water temperature in the core and the position of the regulation rod are measured as a function of time. Figure 1 shows the evolution of the rod position and of the water temperature during the experiment. Using the calibration curve of the regulation rod, the evolution of core reactivity as a function of time is determined. The core reactivity varies as the inverse of the rod position. The evolution of graphite temperature is extrapolated from the water temperature measurements, showing evidence of the thermal inertia of graphite. The analysis of the experimental results, give an insight on the temperature effects. At first, one can observe that the core reactivity varies as the inverse of water temperature, i.e. α < 0 . An increase of the reactor power and thus an increase of the water temperature produces a decrease of the core reactivity, which in turn produces a decrease of the water temperature. This shows that, by conception, the reactor is sub moderated and thus autoregulates when the power is increased, participating to the safe regulation of the reactor. Using the recorded data, one can then successively calculate the global, the water and the graphite temperature coefficients, i.e about - 2 , - 10 and + 8 pcm/°C, respectively, showing that α can take either positive or negative values depending on the physical IAEA-CN-156/S-44 effect inducing the temperature effect. The negative value of α in water is related to dilatation effect that reduces neutron moderation, while the positive value in graphite is related to the decrease of the capture cross section of graphite σ with the neutron kinetic energy ( σ ≈ 1/ T 1/2 ). The experiment also shows that the apparent temperature coefficient that is "lived" by the operator who regulates the reactor is not constant through the time. Indeed, when the water temperature rapidly change, while the graphite temperature is constant due to its inertia, the apparent α is close to αwater, i.e - 10 pcm/°C. This contribute to a rapid change in the core reactivity with time that needs a rapid variation of the control rod position. But, when heat transfer contribute to an equilibrium in water and graphite temperature, then the apparent α is close to αglobal, i.e - 2 pcm/°C. Thus, what is important, is not only the value of the temperature coefficients, but also the physical effects to which they are related and the kinetics of this effects. From the understanding of these temperature effects, empathise will then be given on there impact on the safety of reactor operation, particularly in the occurrence of an accidental situation. 36 450 34 Water temperature Rod position 400 32 30 350 28 300 26 24 250 22 20 200 18 150 16 14 100 -200 0 200 400 600 800 1000 1200 1400 1600 TIME (s) WATER TEMPERATURE (°C) ROD POSITION (mm) IAEA-CN-156/S-17 Monte Carlo Simulation of GRR-1 Core and Neutron Irradiation Positions Using MCNP I. E. Stamatelatos and F. Tzika Institute of Nuclear Technology and Radiation Protection, NCSR “Demokritos”, Aghia Paraskevi, Attiki, Greece E-mail address of main author: ion@ipta.demokritos.gr Prediction of neutron flux at the irradiation devices of a research reactor facility is essential for the design and evaluation of experiments involving material irradiations. The Monte Carlo technique offers significant advantages, since the complex geometrical configuration of the reactor core and irradiation devices can be modeled in detail. A model of the Greek Research Reactor (GRR-1) core configuration was developed using the Monte Carlo code MCNP with continuous energy neutron cross-section data evaluations from ENDF/B-VI library [1]. The model includes detailed geometrical representation of the fuel and control assemblies, beryllium reflectors, graphite pile and irradiation devices [2]. In this work, the MCNP model of GRR-1 core was applied in order to predict the neutron field characteristics at (i) a special in-core irradiation element at core lattice position D-4 (ii) irradiation devices used for neutron activation analysis and radioisotope production (iii) the graphite thermal neutron column assembly The MCNP estimated neutron flux was compared with measurements using gold foils with and without cadmium covers to obtain the thermal and epithermal neutron flux in the irradiation positions. A good agreement between calculated and experimental results was observed. Since experimental determination of the flux characteristics at the reactor irradiation devices is time and labor consuming, the developed model is a useful tool in predicting neutron flux in samples for purposes of designing and evaluating experiments involving material irradiations. Moreover, it was proven particularly useful in the time period of gradual conversion of GRR-1 core from HEU to LEU fuel since it enabled estimation of the neutron field characteristics at the irradiation devices under the actual mixed core configuration conditions. [1] BRIESTMEIER J (Ed.), MCNP – A General Monte Carlo N-particle Transport Code (Version 4C), RSIC CCC-700, Oak Ridge National Laboratory (1997). [2] STAMATELATOS IE et al, “Monte Carlo simulation of the Greek Research Reactor neutron irradiation facilities”, Nucl. Instr. and Meth. in Phys. Res. B (2007), doi: 10.1016/j.nimb.2007.04.074. IAEA-CN-156/S-28 Radiological characterization of the GRR-1 pool F. Tzika, A. Savidou, I. E. Stamatelatos Institute of Nuclear Technology and Radiation Protection, NCSR “Demokritos”, Aghia Paraskevi, Attiki, Greece E-mail address of main author: faidra@ipta.demokritos.gr GRR-1 is a 5MW open pool type research reactor with MTR-type fuel elements cooled and moderated by light water with beryllium reflectors at the two opposing sides of the core. A graphite thermal neutron column is adjusted to one side of the core. Six radial horizontal beam tubes are available, of which three contain in-pile collimators for neutron scattering instruments. The reactor is currently out of operation for inspection and refurbishment purposes. The core has been dismantled and the fuel elements are stored in the used fuel storage tank. The GRR-1 inspection and refurbishment plan involves inspection and eventually replacement of the reactor’s primary cooling circuit. The health physics procedures to be implemented during inspection of the main water outlet are divided in three stages: a) pool dose rate survey from pool top, b) pool drainage by decreasing water level in steps and c) inspection of the water main outlet. Purpose of the present work is the evaluation of the gamma radiation fields inside and around the pool during the above procedures. The Monte Carlo neutron and photon transport computer code MCNP5 was employed in order to asses the activation of materials in the pool and predict gamma radiation levels during pool drainage. The code enabled simulation of the reactor core, pool and activated components within the pool i.e. experimental tubes with in pile collimators, beryllium blocks, reactivity control rods, lead shields and irradiation rigs. The actual photon source in the reactor pool depends upon alloy, impurity content, neutron spectrum, reactor operation history, and decay time. The MCNP predicted gamma dose rate results were calibrated against experimental measurements performed using a submersible GM detector. The results of this study were used to derive the gamma dose rate profile inside and around the reactor pool. The dose rate profile data will be used in order to set the radiation protection requirements, such as implementation of certain procedures or manufacture and placement of special shielding, during the inspection and refurbishment operations. IAEA-CN-156/S-71 The Next 20 Years Operati oonf the 36 Years Old Hungar ian Training Reactor Attila Aszódi 1) 1) Budapest University of Technology ando Encomics, Institute of Nuclear Techniques, Budapest, Hungary E-mail: aszodi@reak.bme. hu The four units of the Paks NPP generate 3o8f %th e Hungarian electricity production, which properly indicates the stressed importance ofle naur cenergy in Hungary. One decade prior to the construction of the Paks NPP, in the e a6r0ly’s, the country had begun preparations for domesticating nuclear technology and the preliminary steps of construction of the training reactor had been taken. The construction finished in 1971 at the Technical University of Budapest, first criticality was reached in May, 1971. The training reactor is a pool-type reac twoirth 100 kW nominal power, operating with Russian-made EK-10 fuel assemblies of uranium enriched to 10%, and based on Hungarian reactor design. The training reactor has been successfully serving Hungarian nuclear expert education, and technical education of physictisetasc, hers and engineers in the last 36 years. Hungary prepares for extending the designt imlife of the four VVER-440/213 type units; in that case they will finis hoperation between 2032 and 20 3D7i.scussion on possible new nuclear units in Hungary was recently commenced. Lifetime extension and the construction of nuenwi ts require continuous expert supply. For this purpose the technical possibility of furrt hoeperation of the training reactor provides a good base, but numerous renewals are to brfeo rpmeed and the generation-change of the operating and educational personne tlh oef training reactor is needed. The colleagues participated in the start-up of the reactor and assured the safe operation of the reactor during the last 36 years, are now eredt ior r preparing to retire. Their replacement requires a special program, which has been pdla naned was implemented in the last 3 years. In the most important operating positions tnhuem ber of employees has been doubled in 2 years, so that the young colleagues can work together with the experienced ones day-by-day. This ensures the most effective transfer of special operational and maintenance knowledge. Up to know the young experts have successfullya ionbetd their licenses for reactor operation, and they are able to operate the reactor alone. For a more effective technic kanl owledge transfer a modern, computer-based 3D CAD model has been developed based on the 40-45 yeadr sd roalwings. This has been proved to be effective from more viewpoints: - differences between the old drawings athned implementation have been revealed; - the young employees could learn reactonr sctruction in a more effective manner through the problems arisen during dtheev elopment of the 3D CAD model; - there are up-to-date drawings, which are may continuously be updated according future modifications of the reactor. A long-term technical development plan was aolusotlined to ensure eth extended operation of the reactor. IAEA-CN-156/S-71 In the last 3 years the following main aocntsi were made at the training reactor: - Renovation of one part of the ventilation system; - Renewal of the pneumatic rabbit system; - Restoration of the classrmoo in the reactor building; - Application of state-of-the-art 3D reaocrpt hysical and thermohyadur lical simulations in the safety analysis; - Building of a new document archive for do land new safety reports, procedures, manuals, research reports and books; - Development and use of a modern intrasniteet serving as general information source for the reactor operators and the collective of the institute. - Actions for the conservation of collective knowledge. The capacity utilization of the training reactor is quite high due to the different education programs. The reactor participates in theu ceadtion of engineering-physicists, energetic, mechanical, electrical and chemical engineTehrse.r e are about 2 to 3 thousands of secondary school students visiting the reactor annually, which helps the teachers in the basic nuclear education. We are participating also in the ENEN (Eopuer an Nuclear EducatioNne twork Association), aiming at the integration of eth European nuclear educatioTnh. e first and most successful course of the ENEN is thEeu gene Wigner Course for aRcetor Physics Experimen, ttshe main emphasis of which is to perform reactor phy seixcpseriments to enhance the knowledge of the students in nuclear engineerinagn d reactor safety. The cou rise based on 3 research- and training reactors in three different countr i(eVsienna – Austria, Prague – Czech Republic, Budapest – Hungary). The 21 days course was started in 2003. In the last 4 years 58 participants from 12 countries accomplished the course. The participants are mainly MSc students, but PhD students and young retxsp cean be found among them as well. In 2006 and 2007 the Periodic Safety Review (PSR) was carried out for the training reactor. Based on this PSR the Hungarian nuclear saafuethyo rity is to issue the operating license for further 10 years. During the PSR the safet yt hoef reactor has been re-evaluated, a new 3D neutron kinetics code has been developned , maodern Computational Fluid Dynamics (CFD) methods have been applied for the safety yasnisa lof the special EK-10 fuel (aluminium- coated, filled with magnesium and uranium-doex imixture) for LOCA and RIA transients. The paper will describe among others theo vaeb mentioned actions in human resource management and knowledge management, and t haels onew safety analysis methods which were applied during the recent Periodic Sa Rfeetyview of the Hungarian Training Reactor. IAEA-CN-156/S-15 Validation of the Monte Carlo Method of MVP Code on the First Criticality of Indonesian Multipurpose Reac1t)o r Tagor Malem Sembirin2g), Surian Pinem2) and Setiyant2o,3) Extended abstract The validation research works in BATANe a frocused using Monte Carlo method with recent nuclear data on theex perimental results. In this par,p tehe validation results of Monte Carlo method of MVP code on the first critiictya lexperimental ofI ndonesia Multipurpose Reactor (RSG GAS reactor) are presented.e MThVP code is a veocrtized and continuous energy Monte Carlo code developed by Japaonm Aict Energy Agency (JAEA). The objective this paper is to show the acracucy of the code using rece nutclear data of JEF-3.0, JENDL- 3.3 and ENDF/B-VI.8. The final go aolf this research is to eu sthe code as a in-core fuel management code since the code hmaso dau le of burn-up caulclation (MVP-BURN). RSG GAS reactor is a beryllium (Be)-recflted, light-water-moderated and -cooled, 30 MWth (max.) multipurpose reactor. Presenthlye, reactor uses MTR-type LEU (19.75 w/o) silicide fuel (U3Si2-Al) elements (FEs). On the 10 1x0 core grid positions there are 40 standard FEs (each consisting of 21 fuel pla teeisg)h,t control elements (CEs, each consisting of 15 fuel plates) initially loaded with 250 and 178.263 g5 U respectively, Beryllium reflector elements and other irradiation facilities. isT hfuel loading corresponds to a uranium meat density of 2.96 g/cm3. The equilibrium core is achieved through so tmraensition cores wit hsmaller core and lower power. In the first transition core, aFlEl s and CEs were fresh oxide fuel with same uranium meat density of 2.96 g/3c.m Some experiments wercea rried out in the core, including criticality experiments, as a part coofmmissioning activity. Since the first core using the fresh fuels, so the core can be used a bench mark core to validate the accuracy of a selected code. In the criticality experimenthtse,r e were some types of experiment have been performed as follows: - first criticality by adding FEs and CEs - criticality condition by adding thBe eryllium reflector elements - criticality condition by adjustment vaorui s positions of six control rods - excess reactivity and total control rod worth In this paper, those criticality experimaeln tresults are compared to the calculated results by using the MVP code. The calculartesdu lts showed that eth calculated results IAEA-CN-156/S-15 using the selected nuclear data are very clo steh et oexperimental results. For example, the calculated core excess reactivity using JENDL-3.3 is in the range of 8F.1k7/k %– 8.35 %Fk/k. The calculated resuislt very close to thex eperimental result of 8.41 F%k/k. It can be concluded that the MVP codieth w JENDL-3.3 and ENDF/B-VI.8 nuclear data can be applied for the MTR type reactor with bulky Beryllium reflector. 1) Submitted to International Conference on Research Reactors: Safe Management and Effective Utilization, Sydney, Australia, 5-9 November 2007 2) Center for Reactor Technology and Nuclear Safety – BATAN, Kawasan Puspiptek Gd. No. 80 Serpong, Tangerang -15310 INDONESIA 3) Will present the paper; E-mail: setia@batan.go.id IAEA-CN-156/S-16 Neutronic Analysis of RSG GAS Silicide Core with Uranium Density of 4,8 g/cm3 1) Setiyanto2,3), Surian Pinem2), T.M. Sembiring2) and Lily Suparlina2) Extended Abstract Fuel conversion program of et hRSG GAS multipurpose reac tiosr to convert the fuel from oxide, U3O8-Al to silicide, U3Si2-Al. The aim of the programis to gain longer operation cycle by having, which is technically possible for silicide fuel, a higher density. Since 1999, a core conversion program of the RSG GAS reafcrtoomr oxide to silicide fuel with the same fuel density of 2.96 g U/c3m has been started and t haet end of year 200a2l l silicide core was achieved. However the conversion study ofn gu shi igher uranium density fuel in the RSG GAS reactor is still on-going. The study isc ufosed on silicide fuel with density of 3.55 gU/cm3, 4.8 gU/cm3 and 5.2 gU/cm3. Previous research showed that 3.55 g3 /cumranium density can extend the operation cycle length from 615 MWD to 900 MWD signcifaintly without any mtaerial and core configuration change. Th ipsaper presents an advcaend research on the 4.8 g /3c muranium density utilization on RSG-GAS core. Thrise search focused on the neutronic aspect, including reactivity balance, poewr peaking factor and kinetic parameters. The calculations were performed using computer codesM WSI/D4, Batan-2/3 DIFF and Batn-EQUIL-2D. During the design, the following constraints are used: 1. No major modification to the reactor balance of plant, shieldingre, cmoain structural components, and civil buildings may be made. 2. The number as well as the performance of irradiation positions and facilities must be maintained. 3. The existing one-stucko-dr reactivity margin and thermhayl-draulic safety requirements must be fulfilled. 4. Maximum discharge burn-up is limited to 70 % for licensing purposes. The calculations showed that the reactor cycle length can be extended until 1400 MWD with the maximum radial power peakinfagc tor 1.31 less that the limit value of 1.4. However, the two safety rodssh ould be added to higher thheu stdown margin (one stuck rod criteria). The reactivity balance and the oinr-ec fuel management of the core are also presented in this paper. For kinetic parameter, thcea lculated delayed neutrona cfrtion, delayed neutron decay IAEA-CN-156/S-16 constant and average neutron life time are 7.03256 -3x, 71.08520 x10-2 s-1 and 55.49o s respectively. Those parameters are not chasnignei,f icantly, compared with the existing core. It is concluded that the proposed equilibrium silicide of 4.8 gU3 /ccomre configuration could be applied in RSG-GAS operation, safely. 1) Submitted to International Conference on Research Reactors: Safe Management and Effective Utilization, Sydney, Australia, 5-9 November 2007 2) Center for Reactor Technology and Nuclear Safety – BATAN, Kawasan Puspiptek Gd. No. 80 Serpong, Tangerang -15310 INDONESIA 3) Will present the paper; E-mail: setia@batan.go.id IAEA-CN-156/S-48 Fault Position Estimation Using Color Segmentation Approach to the Thermal Infrared Image of a Cabling Network in the Nuclear Reactor D.H. Nugrohoa), A. Satmokoa), R. Himawan a), B.C. Dewib), S.M. Isab) a) National Nuclear Energy Agency (BATAN) – Indonesia b) Tarumanagara University - Indonesia e-mail address of main author: djokohn@batan.go.id Currently, predictive maintenance has become an indispensable part of many plants maintenance planning strategies which contribute to the ageing management of a nuclear reactor. It uses projected data or trends from condition monitoring techniques to determine the trouble-free service life of equipment, thereby eliminating unscheduled shutdowns. These techniques monitor deterioration, processing conditions, and specific events that precede the development of equipment failures. One of the predictive maintenance techniques utilizes thermal infrared imagers. This technique detects hot spots in electrical equipment in the plants by scanning the object such as substations, distribution lighting panels, or electrical motors. Problems are easily and quickly eliminated before they cause system failure. Thermal infrared imagers are detector and lens combinations that give a visual representation of infrared energy emitted by all objects. Fault position in a system is estimated in this research using color segmentation approach to a thermal infrared image of a cabling network which can be implemented for ageing management in a nuclear reactor. Image segmentation is a process to convert the real image into some certain area black and white[1]. Implementation of color segmentation for fault diagnosis in the nuclear reactor cabling network is conducted in this research utilizing image from thermal infrared imagers. In this case, the operator should determine the color in infrared image which indicate the faulty. Image input is first converted to RGB (Red, Green, Blue) image model and then converted to CMYK (Cyan, Magenta, Yellow, Key for Black) image model. Assume that the yellow color in the image represented the faulty, the CMYK image model is then diagnosed using color segmentation model to estimate the fault position in a cabling network. The system diagram block is represented in the Figure 1. Conversion Color Image data CMYK Image Output : (real data) RGB to CMYK M o d e l dat a Segmentation Fault position Figure 1. Diagram Block of the Fault Diagnosis System If ko is output value, and the input value is represented by chromatization value ki, then the color segmentation represented in the equation (1) would become binary image which has 2 value as such 1 (white) if the chromatization value greater than threshold (τ), and 0 (black) if the chromatization value is same or less than the threshold (τ), 1 IAEA-CN-156/S-48 = 1 if ki > τ (1)   = 0 if ki < τ Based on the segmented binary image then the fault position can be estimated using equation (2) as follows [2] : (13)                     and     (2)       Input data image, output image, fault location center in cartesian coordinate and fault indication judgement are provided in the Tabel 1. Table 1. Fault Diagnosis Image Input Image Output Image Fault Location (x,y) Fault Indication number 1 (83,75) occured 2 (114,83) occured 3 (24,42) occured It can be concluded from experimental results provided in the Table 1 that the color segmentation approach from input image capable to estimate the location of faulty center in the whole picture of three sheets image acquired, if assumed that the fault is represented by yellow color in the input image. REFERENCES [1]. LIU, Jianging and YANG, Yee-Hong. “Multiresolution Color Image Segmentation”. IEEE TRANSACTION ON PATTERN ANALYSIS AND MACHINE INTELLIGENCE. Vol. XVI, number 7, July,1994. [2]. USMAN, Ahmad. “Pengolahan Citra Digital & Teknik Pemrogramannya”. Yogyakarta: Graha Ilmu, 2005. (in Indonesian) 2 IAEA-CN-156/S-56 WATER CHEMISTRY SURVEILLANCE FOR MULTI PURPOSE REACTOR 30MW GA SIWABESSY, INDONESIA Geni Rina Sunary1o), Sr iyono1), Diyah Er lina Lestar2i) 1)Center for Reactor Technology and Nuclear Safety, BATAN, Bldg. 80, Puspiptek Areag, ,S e rpon Tangerang 15310, Indonesia 2)Center for Multi Purpose Reactor, BATAN, Bldg. 30, Puspiptek Area, Serpong, Tangerang 15310, Indonesia E-mail address of the main author : genirina@batan.go.id There are three research reactors in Indoan, easlli operated by BATAN. The major and the largest one is the 30 MW Multi Purposes eRaerch Reactor (MPR) G.A. Siwabessy located in Serpong, about 30 km from Jakarta. It is o apnen-pool type reacrt ocooled and moderated by light water and fueled by 19.75% enriche3dO U8Al/U3Si2Al yielding averaged neutron flux of 1x1014 n/cm2 s. The reactor construction began 1 i9n83 and completed in July 1987 when the reactor reached its firsti tcicrally, however, its full poweor f 30 MW was reached in March 1992. The reactor cooling system compriseims aprry and secondary systems. The primary coolant is subjected to gamma and neutirrorand iation by the core, producing radiolysis species, such as oxidator2 Oand H2O2 which induce oxidation reactions of dissolving substances within them. This causes corro rseioanctions in the reactor tank components and other undesirable effects like forming compnodu deposits inside the cooling system, in particular, inside the heat cehxanger cores, thereby adveyrs ealffecting its performance and life-span. As such, it is imperative to mitonr, control and maintain the primary water chemistry to the required specifications aaltl times in order toe nsure longevity and performance of the reactor system. The seacroyn dcooling water, on the other hand, is not subjected to irradiaotin, and its water chemistry is aidm eat suppressing corrosion and algae growth. A special surveillance program has been apredp and put into practice to study the material behaviour when subjected to therr ecnut water chemistry regime and to acquire experimental data upon which BATN can effectively improve the primary water chemistry. The main objectives of thseu rveillance program are to: (i) establish standard procedures for ocsoirorn monitoring and surveillance; and (ii) provide technical guidelines for coi nuted wet storage of spent fuels. Experiments have started with clear undersitnagn dof the current water chemistry profile, followed by the manufacture of a series of aopprirate coupons to sdtuy the effect of the primary water on homogeny, galvanic and crevciocrer osion. The test coupons were made of the same grade of materials as used aflourm inium fuel claddnig and reactor tank, and stainless steel fuel storage racks. As shoinw Fni gure 1, the coupons amrea de of a series of discs (h95 mm with h15 mm centre hole; 1 mm thick for stainless steel and 5.5 mm thick for aluminium) put together as an assembly usinsgta ain less steel centre pin and insulated with a ceramic bushing and 5 mm thick spacers betwee nd itshcs. The coupons are also assembled in two orientations to allow veticral and horizontal installations itnhe pool (Fig. 2). The first batch of six coupon assemblies were strategically positioned in rtvhiece s epool in January 2007 and are due to be withdrawn in stagesin fsopre ction after 1, 2, and 3 years of exposure. IAEA-CN-156/S-56 SS Al2 Al3 SS-Al2 SS-Al3 A2l-Al3 Al2-Al2 Al3-Al3 SS-SS Ceramic SS 304 Figure 1. Coupon Assembly.(SS=SS 304, Al2=Aluummin for fuel cladding, Al3=Aluminum for tank) (a) Vertical (b) Horizontal Figure 2. Coupon Orientations [1] IAEA, TRS 418, ‘Corrosion of ResearcRhe actor Aluminum Clad Spent Fuel in Water’. [2] IAEA TECDOC 927, ‘Influence of WateCr hemistry on Fuel Cladding Behaviour’. [3] Diyah Erlina Lestari, Geni Rina SunaryYou, si Eko Yulianto, Sentot Alibasyah H. and Setyo Budi Utomo, ‘G.A. Siwabessy Resceha Rr eactor Water Chemistry’, Module B2, Water Chemistry Course I, Serpong, Indonesia, 2004. IAEA-CN-156/U-51 ASSESSMENT OF QUALIFICATION OF RSG-GAS REACTOR OPERATOR IN INDONESIA LY Pandi1), DC Sinaga2)and B. Isa3 ) 1)Center for Regulatory Syesmt and Technology of Nucleianrs tallations and Materials 2)Directorate for Nuclear Instaltliaons and Materials Regulation 3)Directorate for Nuclear Instlaltions and Materials Licensing Nuclear Energy Regulatory Agency Email addressp: .liliana@bapeten.go.id Abstract The reactor operator and supervisor who opetrhaet er eactors have important tasks to ensure the safe operation of the facilities. For thaats roen, personnel who are cinh arge to operate the reactors, both operators ansdu pervisors, shall possess wionrgk license issued by Bapeten according to the Act No. 10/1997. To implement the Act No. 10/1997 especialllya trieng to the personnewl ho operates a nuclear reactor in Indonesia, Bapeten has enacted the guidance which is required by operator/supervisor to fulfill the requirements obtaining license to nuclear installation and ionization radiation utilizaotin installation, such as Govemrnent Regulation No. 64/2000 a nd Decree of Bapeten Chairman No. 17/1999. In this paper, the objective of the assessmoef ntht e operators and supervisors of RSG-GAS Reactor is performed to identify that operatinga onrizations (in that ca,s feor Center for Multi Purpose Reactor, the operating organoizna tihas managed the RSG-GAS Reactor) has implemented the Bapeten provisions and requirnetsm. eThis paper describes the qualification for operators and supervisors of RSG-GAS Roera acnt d the assessment of the operators and supervisors of RSG-GAS Reactor. 1. INTRODUCTION Indonesia has three research reactorsR iS.eG. -GAS Reactor at Srpeong, TRIGA 2000 Reactor at Bandung and Kartini Reactor at YogyakarFtao.r operating nuclear reactor, it is needed some reactor personnel. Nuclear installatione raotping organizations shall ensure that the designated reactor personnel iflulinlfg the requirements to opetera nuclear reactor according to the Act No. 10 Year 1997 Article 19 clause). (I1t stated that any employee who operates a nuclear reactor and other nucleinasrt allation and in the installations that use ionizing radiation sources shall possse a license.[1]. 2. REGULATION The training, qualification nad working license are baseodn the following regulations: - The Act No. 10 year 1997 on Nuclear Energy - Government Regulation No. 64 Year 2000 ocne Lnising of Nuclear Energy Utilization IAEA-CN-156/U-51 - Government Regulation No. 43 Ye2a0r0 6 on Licensing of Nuclear Reactor - Decree of Bapeten Chairman No.06 /BKAa-PETEN/V-99 on The Construction and Operation of Nuclear Reactor - Decree of Bapeten Chairman No. 17/Ka-Btaepne/IX-99 on The requirement to Obtain a Working License for Personnel in NucleInasr tallation And Radiation Installations - Decree of Bapeten Chairman No. 04P/Ka-Beanp/eI-t03 on The Guidacne on the Training of Nuclear Reactor Operators and supervisors. According to Chapter 3 of GovernmenRte gulation No. 64 Year 2000, the operating organization shall have qualified personnerl fnouclear energy utilization [2], and the personnel operating or control the reactor opoenra otir personnel rlaeted to safety shall obtain working license issued Bapeten before pthersonnel performs the duty. [2, 3, 4]. In Decree of Bapeten Chairman No.06 /Ka-PBEATEN/V-99 Article 10 is governed that [4]: 1. The expert and the personnel who act a ansu clear reactor operator shall have a working license from Bapeten. 2. The working license is issued by Bapne atefter the personnel has passed the exam. 3. The working license is issued for some period and to be renewed. According to Decree of Bapeten Chairmaon. 0N6 /Ka-BAPETEN/V-99 Article 24 clause 2.c., the operating organization shall establish thaein itnr g programme for reaocrt personnel, and in Decree of Bapeten Chairman .N 1o7/Ka-Bapeten/IX-99 in Chaprt e5 is governed that any reactor should be operated by some capables kainll dp ersonnel, at least consisting of [5]: a. one operator b. one supervisor c. one radiation protection officer d. one maintenance personnel Article 6 stipulates that the personnel shall h tahvee training and pass the examination to proof her/his qualifications. The Decr eoef Bapeten Chairman conclusd tehat reactopr ersonnel shall perform the training in thaec creditation organiziaotn/institution. He/she has carried out an examination by Bapeten for a working license. 3. TRAINING AND QUALIFICATION A. Preliminary Training Programme (Qualification Programme) Preliminary training programmseh all provide the understanding othne basic principles of nuclear technology, nuclear safety, radina tioprotection, reactor design and reactor utilizations, etc. fora pplication in field. That training for operators and supervisors ls choanl tain an adequatee tohretical and practical knowledge on function, location and operation m ofd ereactor, and also shall emphasize to the importance of limiting and condition of opteioran, and the consequences of violation the limit, and the safety consequences of fault procedures. IAEA-CN-156/U-51 B. Programme of Qualification Examination After passed the training examination, the candidate of operator and supervisor shall be examined and evaluated in tfhinea l training before examination by BAPETEN for a working license suitable for his professionalism. The method of evaluatiobny Bapeten consists of: 1. Written examination 2. Oral examination (can be done in the cloars sa s a part of praictal examination in facility), and or 3. Practical examination on normal and emergency operation condition. C. Requalification Programme Requalification programme is sbead on the systematic apprmoxaition to ensure that the knowledge, skill and attitude oofperating personnel can bee sper rved. If necessary his skill and knowledge can be increased. Any personnel h atavsinkg in the field of reactor safety shall perform this programme. The programme is ceadr roi ut every two years. The requalification programme includes the selected topics frpormel iminary training, uspporting important tasks for reactor safety operation, infrequent anfdfi cduilt task. The programme shall also include periodical training and case studies in egmeenrcy conditions and evnts by operating organization. D. Programme of requalification Examination Operators and supervisors which will renewe thworking licensing shall take part in programme of requalification examination. Tphreo gramme of requalifaiction examination is conducted by Bapeten. The method of evaluation on the requalification examination bpye tBean is the same with the method of evaluation of the qualification examination. E. Promotion The promotion of their occupations is usedb eto p rovided in all facitlies. When the operator is assured to be capable of a reactor suispoerr,v his/her experiences, leadership and communication capability shall be taken into oaucnct. He/she as a new professional worker shall be trained to increase his/her knowledge and skill. IAEA-CN-156/U-51 4. THE REQUIREMENTS OF QUALIFICATION EXAMINATION The requirements of operator and supervisor examination are following : a. The operator shall be graduated at lefraosmt scientific or engineering senior high school. The supervisor shall be graduaatet dl east from Bachelor of science or engineering. He/she should haveo tywear experiences in nuclear field. b. In general, the operator and superv sishoarll have good health and physical condition. Therefore it is assumed that there is non utincaipated failure or fault resulting in the hazard for the health and safety publics, for example: epilepsy, psyche, diabetes mellitus, hypertension, heart, unconsciousneasrs ,o r eye disability and other mental and physical condition. 5. APPLICATION OF WORKING LICENSE The licensing applicant shall fitlhl e application form and enclose: i. Documents of applicant abilities on thmee thods of safe reactor operating. For supervisor, the applicant aslhl understand his/her joabn d responsibilities. These documents include a certificate or recommendation from the examination team. The certificate or recommendation consists othf e: detail course, cosuer hours, start-up and shut down experiences. ii. Physician recommendation from his/her medical check up. 6. WORKING LICENSE Bapeten will give a working license for a tpicairpant passing the exam. The working license is one of significant documents presented capabilities as a reactor operator or supervisor. The working licenses hase ebn valid for two years. The expired license will be come into eff,e cift the licensee renews his/her license. The renewal application must be enclosed as following: i. Documents on applicant exrpieences includnig total hour. ii.A letter of recommendation from operatinogrg anization stated atht the applicant has implemented his/her working very well. iii.Medical check up results. iv.A certificate of requalification training. The duration of working cliense can be renewed if: a. The operator or supervisor has good heanltdh pahysical condition so that there is no unanticipated failure or faltu resulting in the hzaard on the health and safety publics. IAEA-CN-156/U-51 b. The operator/supervisor has performed hist/ahsekr actively or andh e/she is able to continue his tasks. c. He/she has passed the examination arerda nbgy Bapeten. For participant not passing this exam, he/she may takee tehxamination at another occasion. Reactor operator or supervisliocre nse can be revoked if : a. The operator or supervisor has not a good health and physical condition. The requirements in section 4.b.s h naot yet been fulfilled. b. His/her fault can cause any accident increasing the radiation hazards and contamination for personnel, member of publics and environment. 7. ASSESSMENT OF QUALIFICATION FOR RSG-GAS REACTOR OPERATOR AND SUPERVISOR A. Method of Assessment The assessment of RSG-GAS operatorsd’ saunpervisors’ qualification has performed by comparing the RSG-GAS operators’ and supeorvsi’s licensing documen twsith the DCB No. 17/ 1999 and DCB No. 04P/2003. B. Result and analysis/assessment RSG-GAS reactor is research reactor in oInedsia, and has potential radiation hazard, therefore to operate it, it is neededm seo qualified operators and supervisors. The requirements of the operator and superv ilsicoernse has been established by Decree of Bapeten Chairman No. 17/Ka-Bapeten/IX-99r. iFdoentifying whether the RSG-GAS operator and supervisor have fulfilled the qualificatio in accordance with the provisions issued by Bapeten, it is needed to assess the qcuatliofins of RSG-GAS reactor operators and supervisors. The background education requirements to beea act or r operator or supeisrvor is according to the examination requirements of sec. 5.a. In this case, RSG-GAS reactor operators and supervisors graduated at let afsrom scientific or engienering senior high schools and Bachelors of science or engineering, and nhga veixperience in nuclear field minimum of 10 years. The requalification for operator and superv iswoilrl renew the working license. The training requalification programme is performed periodically by Center for Multi Purpose Reactor (CMPR) for minimum 2 years. According to sec. 5.b. the reqeumir ent of physician and healctho ndition, CMPR is to perform medical check up for the operator and superv cisaonrdidates. The operator and supervisor may renew the working license. In table 1 can be shown thre sults of assessment ofe thqualification for operator and supervisor to fulfill the provisions/requirements issued by Bapeten. IAEA-CN-156/U-51 TABLE 1: ASSESSMENT RESULT OFT HE QUALIFICATION RSG-GAS REACTOR OPERATOR AND SUPERVISOR No. Provision/Requirement Regulation AssessmReenstu lt 1. a. The operator shall be graduated at DBC No. 17/1999a. Education for operators : least from scientific or engineering Annex II. 2.1 generally scientific or senior high school and experienced engineering senior high in nuclear filed minimum 2 years school : 22 persons with b. The supervisor shall be graduated at experience between 2- 14 least from Bachelor of science or years engineering. He/she should have two b. Supervisor education: year experiences in nuclear field scientific or engineering senior high school : 6 persons and Bachelor of science or engineering 4 persons with experience more than 10 years. 2. having good health and physical DBC No. 17/1999The physician and health condition. Therefore it is assumed that Annex II.2.1. condition of RSG-GAS there is no unanticipated failure or fault reactor operators and resulting in the hazard for the health and supervisors are good safety publics, for example: epilepsy, condition, is proved by the psyche, diabetes mellitus, hypertension, medical check up reports. heart, unconsciousness, ear or eye disability and other mental and physical condition. 3. Attending the preliminary training DBC No. 17/1999 Operators and supervisors programme for the operator and article 6 have followed the supervisor candidates and the DCB No. 04P/03 qualification (preliminary requalification programme for the programme) and operator and supervoirs which will renew requalification programmed the working licensing programmed by by CMPR, is proved by operating organization. training certificates. 4. Performing the qualification and DBC No. 17/1999 Passing the examination, requalification examination arranged byA rticle 6, 10, proving by the working Bapeten. Annex II.2.1. and license issued by Bapeten. 5.1 DCB No. 04P/03 . 8. CONCLUSION Reactor personnel operating nuacrl ereactor shall fulfill ther equirements in regulation, because they will determine whether the troera coperation is safe or not. One of the requirements is working license. This lice nwseill be submitted after the personnel has attended the training by the accreditation trnagin oi rganization/instittuion and has passed the examination by Bapeten. From Table 1 can be concludd ethat RSG-GAS reactor opaetirng organization –CMPR has complied the Bapeten provisions for arrangineg qthualification and reqaulification training programme for the operators and supeorrvsi.s The RSG-GAS reactor operators and supervisors have followed the qualification arenqdu alification trainings and examinations. IAEA-CN-156/U-51 9. REFERENCES [1] Act No 10 Year 1997 on Nculear Energy, Jakarta (1997). [2] Government Regulation No. 64 Yea2r0 00 on Nuclear Energy Utilization Licensing, Jakarta(2000). [3] Government Regulation No4.3 Year 2006 on The Licensing Nofuclear Reactor, Jakarta (2006). [4] Decree of Bapeten Chairman No.06 a/-KBAPETEN/V-99 on The Construction and Operation of Nuclear Reactor, Jakarta (1999). [5] Decree of Bapeten Chairman No. 17/Ka-Betaepn/IX-99 on The requirement to Obtain a Working License for Personnel in Nuclear Ianllsattion And RadiationIn stallations, Jakarta (1999). [6] Decree of Bapeten Chairman No. 04P/Kap-eBtaen/I-03 on Guidance on training of nuclear reactor operator and supervisor, Jakarta (1999). [7] Licensing Document o Rf SG-GAS Reactor Operators and Supervisors. IAEA-CN-156/S-49 Developing an Ultrasonic NDE System for a Research Reactor Tank Y. Perets, E. Brosh, M. Ghelman, N. Moran, Y. Drymer, Y. Kadmon, S. Dahan Nuclear Research Center – Negev (NRCN), Beer-Sheva, Israel E-mail address of main author: yaronprts@gmail.com Ultrasonic testing is one of the established tools for routine in-service inspection of reactor tanks [1]. As part of the preventive maintenance of the IRR2 reactor, an ultrasonic scanning system was developed for the inspection of the reactor tank wall. Here, we present the main features of the special equipment developed for this task. In addition, we describe the procedure used for validating the inspection method. The inspection will be done from inside the tank, using the immersion technique [2], when the tank is empty of fuel and control rods but filled with coolant/ moderator water. The construction of the mechanical apparatus that was designed for the task is sketched in figure 1a. The ultrasonic scan-head is carried by a custom-designed manipulator that is inserted in the center of the tank. The manipulator has two motion axes that are the elevation Z and the rotation Φ. The scanning will be done in three stages: 1. A preliminary, low resolution (1 mm X 4 mm), scan of the whole tank wall using two 5 MHz straight transducers and six pairs of 2.25 MHz angular transducers operated in both Pulse-Echo and the Pitch-Catch modes [2]. The aim of this stage is to detect areas in the tank that may contain flaws. 2. A high resolution (1 mm X 1 mm) scanning of suspect areas, detected in stage 1, using the same scan-head as in stage1. 3. A local scan for the evaluation of suspect- flaws that may be detected in stages 1 and 2. This stage is performed using a 10 MHz straight transducer and 3 pairs of 2.25 – 10 MHz angular transducers. The scan-head used has an additional degree of freedom ω. In order to enable the thorough evaluation of the results, all the acquired A-scan signals are saved on a 2 TByte RAID system. A dedicated software package was developed for the processing of this huge amount of data. The validation of the design was carried out in two stages. First, the ultrasonic technique was calibrated on specimens with known flaws, using a laboratory-scale ultrasonic scanner. In the second stage, the whole inspection system was tested in a full-sized mock-up. In this mock-up installation, critical procedures such as the insertion and the folding of the manipulator arm were practiced. Also the mock-up tank wall contained flaws that were unknown to the inspectors. In figure 1b, a sample C-scan of an area containing such validation-flaws is shown. Conclusions A special apparatus was developed for the ultrasonic scanning of a research reactor tank wall, the operation of which was practiced using a full-scale mock-up. The inspection technique was validated using a variety of flaws that were unknown to the operators. IAEA-CN-156/S-49 Motor Z Φ b ω a FIG. 1. a) The special manipulator designed for scanning the reactor tank. The manipulator is inserted vertically through a central hole and folded to scanning position. b) A C-scan of a defect- containing area in the mock-up. [1] ASME Boiler and Pressure Vessel Code, Section XI. [2] ASM Handbook, Vol. 17 Nondestructive Evaluation and Quality Control Section: Methods of Nondestructive Evaluation, Ultrasonic Inspection IAEA-CN-156/S-70 Neural Networks application to CRDM and thermo-hydraulic data validation Authors: M. Sepiell1i, M. Palomba2, A. Ratto1, R. Rosa2, M. Bernabucc3i, 1ENEA – FPN 2ENEA – FPN-FISON 3ENEA – Guest Reference: Ing. Massimo Sepielli – ENEA – FPNV –ia Anguillarese, 301 – 00123 Rome - Italy The activity is aimed at validating data taoibned by reactor measurements through soft- computing models based on neural networks (NN). Application is made on the CRDM rod position athned fuel temperature which are calculated by NN-based algorithms and compared with the dreaatal coming from sensors. The bias is used to train the NNs for improving the performance in the next experiments. General Descr iption of the Reac t or The reactor is a typical light-water reactort hw ian annular graphite reflector cooled by natural convection, wthi a power of 1MW. Reactor Cor e The core is placed at the bottom of the 6m.2-5h-igh open tank with 2-m diameter. The core has a cylindrical configuration. In total there are 91 locations in the core, whciachn be filled either by fuel elements or other components like control rods, a neutron sourcraed, iar tion channels, e tcT.he core lattice ha s an annular but not periodic structure. Reactor instrumentatio n Core temperatur eis measured by 20 thermocoupilteu asted above and under the core (see Fig.1). Fuel temperatur eis measured in two fuel eleme nintstrumented with thermocouples. Neural Network application In this first application we try to validate the position of the control rod. To do this we use a neural network that elaborates the measuretad odna the thermo-fluid-dynamic system of the reactor, in parallel with the control system. The used net is feed-forward 3-layerst wnoerk trained using second-order algorithms (Levenberg-Marquart). It is constituted from two nets, Fig.1 that woinrk p arallel: the first used for the validation of the data (trained initially thorugh the use of reactor data set); the second one used for the training that update its parameters everye t imoccur an error given from the difference between the exact value of the first nentd athe measured value from the instruments (validated from the operator). The new parameters of training will come therefore transferred to the first net. IAEA-CN-156/S-70 Fig.1 Validation network Fig.2 Neural netw ork The net’s geometry is the following: inpluaty er 35 neurons; hidden layer 22 neurons; output layer 2 neurons Fig.2. Input data: core temperature (20 thermoceo)u;p fluel temperature; water temperature. Output data: rod position (2 channels). The outcome of training is displayed below Fig.3 Rod position channels Fig.4 Rod position error In this second application weo wnt validate the value of the fu teelmperature. To do this we use a neural network trained with tmhe asured power and fuel temperature. The used net is feed-forwa4rd-l ayers network trained usinfgir st-order algorithms (steepest descent). The outcome of training is displayed below Fig.5 Experimental data (blu,e ) Fig.6 Percentage rror Simulated data (pink). IAEA-CN-156/U-32 Investigation of JRR-3 control rod worth changed with burn up of follower fuel elements T. Hosoya1 ) Y. Murayama1) T. Kato1) 1) Japan Atomic Energy Agency (JAEA), Tokai, Naka, Ibaraki, Japan E-mail address of main authohro: soya.toshiaki@jaea.go.jp JRR-3 (Japan Research Reactor No.3) is a swimming pool type research retahc tthoer rwmial power of 20MW. The core is composed of 26 standard fuel elements, 6 control rods (Sa-1, Sa-2, S-1, S-2, R-1, R-2), beryllium reflectors, and vertical irradiation holehso wasn sin Fig.1. Control rod is composed of follower type fuel and hafnium neutron absorber. Length of an operation cycle is 26 days, and seven operation cycles are conducted for a year. The control rod worth (CR worth) has been measured in an annual maintenance period by inverse kinetics method (IK method). The CR worth is used as important and basic data for management of reactor operation, such as prediction of control rod position at stadrt -up an estimation of an irradiation sample reactivity and excess reactivity chanitghe db uwrn-up of fuel elements. However, because the control rod worth would be affected by thge cohf an burn up of fuel, the reliability of the measured CR worth would be degraded when the burn-up would be different. In particular, the difference in the CR worth would be en larged before and after the fuel exchange. For these reasons, it must be necegssr asrpy toh e CR worth precisely to operate the reactor safely. Table 1 shows the CR worth measured in 7 years. It is shown that the maximuemn cdeif fienr the total worth is about 3.1%k/k and would be caused by the difference in burn-up. In order to evaluate the difference, the CR worth in different burn-ups are calcula steimd ublyating the measurement with IK method, and compared with the measured one. This report hseh ows t results of evaluation and also proposes a convenient method to estimate the hC Rw itwho rt higher reliability. Whole core calculation is carried out by the Monteo C Caorlde MCNP5. IAEA-CN-156/U-32 Fig.1 Horizontal Cross Section of JRR-3 Core Table 1 Control Rod Worth of JRR-3 Measured with IK method Control Rod R-1 R-2 S-1 S-2 Sa-1 Sa-2 Total Fiscal Year 2000 3.949 3.942 4.140 4.237 4.818 5.017 26.103 2001 3.344 3.263 3.393 3.553 4.665 4.768 22.986 2002 3.562 3.552 3.752 3.764 4.219 4.319 23.168 2003 3.490 3.500 3.858 3.806 4.327 4.196 23.177 2004 3.755 3.947 3.752 3.663 4.029 4.127 23.273 2005 3.728 3.671 3.768 3.941 4.371 4.490 23.969 2006 3.890 3.792 3.820 3.950 4.242 4.319 24.013 (Unit: %k/k) [1] M. Takeuchi, et al., “The development of the Measurement Technique of the Condt rol Ro Worth Source”: JAERI-Tech 2000-054 (2000) [2] X-5 Monte Carlo Team, “MCNP — A General Monte Carlo N-Particle Tran sCpodrte, Version 5” LA-CP-03-0245 (2003 ) IAEA-CN-156/S-72 Project to Replace the Control and Protection System at the WWR-K Research Reactor P. Chakrov1 ), F. Arinkin 1), Sh. Gizatulin1 ), R. Schultz2 ), N. Mote 3), K. Alldred 3) 1) Institute of Nuclear Physsic, NNC RK, Almaty, Kazakhstan 2) Nuclear Threat Initiative, Washington, USA 3) International NucleaEr nterprise Group, USA E-mail address of main author: chakrov@inp .kz The WWR-K is a tank-type, light water moderda taend cooled multipurpose research reactor with a nominal power rating of 6 MWn ad a peak thermal neutron flux of over1 41 n0/cm2s. The reactor is operated by the Institute of Neuacr lPhysics of the National Nuclear Center of the Republic of Kazakhstan (INP), and is locaatet dA latau, near Almaty, the largest city in Kazakhstan. The reactor was commissioned 6in7 1 w9ith a power rating of 10 MW, then shut down between 1988 and 1998 while seismic saufpegtyr ades were completed. Currently, INP is undertaking the planning and technical wor ka ltlow the reactor to be converted from HEU fuel with a U-235 content of 36% to LEU fuweilt h a U-235 content of 19.7%. The reactor has an important role in Kazakhstan, and ios dpurcing isotopes for medic anl d industrial uses, providing material testing servic easnd neutron activation analysis. Replacement of reactor control and protectione smys (tCPS) is a part of the wider program of the reactor modifications related to its cornsvioen from HEU to LEU fuel, supported by the US Department of Energy, the Nuclear Th rIenaittiative and the Kazakhstan government. The program includes development ao fnew LEU fuel assembly sdiegn and re-configuration of the reactor core, with change in number paonsdi tions of the contrl orods, which requires modification of the CPS. Furthermore, replaceenmt of the existing instrumentation, some of which is 40 years old, will improve reactor stayf,e bring the CPS up to current international standards, and provide an upgraded control interface. The existing CPS at the WWR-K relies to rag lea extent on old designs of detectors and electronic equipment that are no longer in prcotidoun and the system as a whole does not meet the necessary standards for prolonged operati othne o rfeactor. Consequently, the Kazakhstan Atomic Energy Committee (KAEC), the nucleraerg ulatory body of Kazakhstan has mandated that the project for conversion of the WWR-K u tsoe LEU fuel must include replacement of the CPS. By using the supposretr vices availablfer om the IAEA for procurement of the replacement CPS, INP will ensure that thwe nseystem meets current safety standards. Organizationally, the project will take full adnvtage of the assistance available from the IAEA Technical Cooperation proagmr for the procurement of equipment and services for the replacement of the CPS at tWheW R-K reactor. As a first estp, INP is developing a draft technical requirements document for the neswt esmy . IAEA technical and procurement staff will work closely with the team from INP, awsill the reactor designr eand the funding parties in reviewing the technical requirements document for completeness and compliance with applicable safety regulations and standaTrdhse. completed technical requirements document will be submitted to the KAEC, for approvaAlf.t er approval by KAEC, it will be formally submitted to the IAEA along with a request foor cpurrement, to initiate a standard tender IAEA-CN-156/S-72 procedure. The contract between IAEA aned sthelected supplier of the replacement CPS is expected to be signed in late 2008. Replacement of the CPS inclusd tehe following activities: - Replacement of the reactor power monitoring, automatic control and emergency shutdown systems based on in-core neutron flux measurements; - Installation of a new modulfeo r collection and processing of data from instruments measuring non-nuclear parameters. The seidsmetiecc tion instrumentiaon will be replaced, although the existing instrumentation will beta riened for measuremte onf other non-nuclear parameters; - The existing control rod drive motorsil lw be retained but the control rod position monitoring instruments will be replaced annde aw control system will be installed. The control rod position indicators on the rearc ctontrol desk will also be replaced; - The control desk in the main control roomnd a all reactor operations and emergency system displays will be replaced; - The control desk in the stand-by control rowomill be replaced (including all safety system equipment, neutron flux and rod positiionnd icators and operational controls); - All audible and visual warning systems will be replaced; - A new data management system will be insdta, ltleo provide upgraded collection, storage and processing of information on operationth oef reactor and ancillary equipment, plus improved control room data readouts; - Provision of an auto-diagnostic system tov pidre pre-startup testing of the relevant instrumentation and monitoring equipment atensdt ing of the serviceability of operating equipment. Both to ensure maximum reliability and tom spilify maintenance, the replacement CPS will, to the extent possible, be based on standnadruds itrial controllers. Performance of the power level and rate of power increase control systems and the emergency protection system will be tested under reactor opernagt iconditions during commissioning of the new CPS. Procurement and installation of the replacement CPS will include the following steps: - development of the technicraelq uirements document for the new system compatible with other core modifications, - development of the system design for the new system, - equipment manufacturing and factory acceptance tests, - installation of the nwe CPS at the reactor, - commissioning of the new CPS and staff training. These activities will meet the regulationesc, htnical and safety norms and standards of Kazakhstan, IAEA, IEC and other relevant bodthieast set safety standards for the operation of research reactors and replacement of recshe raeractor control and protection systems. Modeling of Thermal Hydraulics Behaviour in Reactor Core Of Reaktor TRIGA PUSPATI (RTP) Khairol Nizam Mohamed, Mohammad Suhaimi Kassim, Zarina Masood Reactor Facility, Division of Technical Support Malaysian Nuclear Agency ( Nuclear Malaysia ) Bangi 43000, Kajang, SELANGOR, MALAYSIA Email: khairol@nuclearmalaysia.gov.m y Abstract Reactor TRIGA PUSPATI ( RTP ) in Malaysian Nuclear Ancgye (Nuclear Malaysia) is the one and only research reactor in Malaysia and had bede ne xucslusively for research and development ( R&D ), training for reactor opesr aatonrd education purposes. The RTP is a 1 MWt pool type reactor with natural cotinovne cooling system and pulsing capability up to 1200 MWt. It went critical 8o nJ u2ne 1982 and the core configuration has been changed twelve times to Tdhaete c. ore is a mixed type using 20% enriched U-ZrH fuel element containing 8.5, 12 and 20watn%iu mur. This paper will discuss the modeling of thermal-hydraulics beohuar vini reactor core of RTP using computer code namely PARET. The results of thceu lactaiol n that were carried out at RTP are modelled and temperature profil ethse o fthermal hydraulics data at different locations and power levels are developed. As a comparison to the thermal hydraulics calculation using EPTA, Ran experiment were carried out at several different locations and powverl sl ein the reactor core for temperature profile in the core to compare the result oebdta firnom PARET. Finally, an overall analysis of the result of PARET calculation and emxpeenrtial measurement were exhibited in this paper. Keywords : TRIGA, PARET, thermal-hydraulics behaviour IAEA-CN-156/S -67 PUSPATI TRIGA Reactor : 25 Years of Safe Operation and Straetesg fior Ensur ing Safety and Secur ity Z. Masood, M.S. Kassim Reactor Facility, Technical Support vDisiion, Malaysian Nuclear Agency Bangi, 43000 Kajang, Selangor, MALAYSIA zarina@nuclearmalaysia.gov.my The PUSPATI TRIGA Reactor (RTP) is a pool-ty1p eM W research reactor which has been in operation since 28 June 1982. During its 25 yea rssa foef operation, no incidents as listed in the unusual event reporting cateigeos rhave been reported. A pragmatic approach to safety has ensured that the reactor is operated within o iptserating limits and controls, while regular maintenance has been carried out accordinmg atnou facturer’s recommnedations. In addition to this, several reactor systems or componehnatvse been repaired, refurbished or replaced over the years, in order to maintain the reactor integrity. However, ageing systems and components especially the instrumentation aonndt rcol systems are of epsrsing concern. This has lead to the initiation of a project to uapdger the present analog control console. An in- service inspection has also been initiated wthiteh assistance with the non-destructive testing group in Malaysian Nuclear Agency. An important element in ensuring reactorf estya is the personnel involved in the reactor operation and maintenance. Previously, insuefnfict ipersonnel were allocated to the reactor facility however this has cahnged since early 2006 wheree tnhumber of personnel has more than doubled. Retraining and upgrading of threre cnut reactor operators are underway and new reactor operators are also being trained. The last safety analysis report (SAR) for RwTPas written in 1983, a eyar after the reactor started its operation. Hence, in 2006 and 20th0e7r,e was a concerted effort to update the SAR and it has been submitted to the regulatory body in mid 2007 as part of the requirement for the issuance of the operation license of RTP. In 2004 RTP was awarded the ISO9001:2000 for riatsd iar tion services fro neutron activation analysis. However, in odrer to be in line witht he practice in the nuecal r industry, a quality assurance programme (QAP) was developed in 2006 and implemented in mid 2007. An assessment of the safety culture at RTPs cwaarried out by an international peer review team in early 2006. In response to the recommendations of this peer review, several changes and improvements to the facility and procedures were implemented. Subsequently, the peer review revisited the facility and commendoend the improvement and prompt implementation of the majority of the recommendations. The recent emphasis on the security of research reactors world wide has lead to the reassessment of the physical protection systnedm e amergency preparedness at the reactor site. Improvements are being implemented and systems are being enhanced. In conclusion, the strategies taken to enhanec es athfety and security of the RTP in line with international practice will assure the safe opoenra toi f the last 25 yesa ris continued in the future. IAEA-CN-156/S -68 Core Calculation of 1MW PUSPATI TRIGA Reactor (RTP) using Continuous Energy Method of Monte Car lo MVP Code System A. K. Julia, Z. Masoo d Reactor Facility Unit, Malaysian Nuclear Agency, 43000 Kajang, Selangor, Malaysia julia@nuclearmalaysia.gov.m y The RTP is a light-water moderated and p-toyople TRIGA MARK II reactor with power capacity of 1MWt. It was built in 1979 andtt aained the first criticality on 28 June 1982. The RTP was designed mainly for neutron actitoivna analysis, small angle neutron scattering, neutron radiography, radioisotope productioend,u cation and training purposes. It uses standard TRIGA fuel developed by Genl erAatomic in which the zirconium hydride moderator is homogenously combined with enric uhreadnium. It has a cylindrical core with which possibility of locating 127 of fuel eleme.n t sBoth of the coolant and moderator uses light water system and the reflector is mad eh iogfh purity graphite. Because of its relatively small power, it uses natural convection for its licnogo system. To ensure the integrity of the core, fuel shuffling have been carrieodu t several times. Until now, there were 12 configurations of the core, the most rececnhta nge being in July 2006. This paper will describe the RTP core calculation using the Monte Carlo MVP code system. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation in order to have an accurate afansdt Monte Carlo simulation of neutron and photon transport problems. The MVP MonCtea rlo code calculation is based on the continuous energy method. This code is acbalep of adopting an accurate physics model, geometry description and variance reduction technique. When compared to the conventional scalar method, this code could achieve hri gchoemputation speed by up to a factor of 10 on the vector super-computer. The RTP core has been modelled using cylri ngdeeometry along the z-coordinate geometry with the MVP code system while its materciarlo ss section data is calculated beforehand. The JENDL3.3 data library was used in twheh ole calculation. The objectives of the calculation are to calculate the multiplication factor valueesff) ,( kfission density and flux distribution from the tally data. The calctuiolan also gives control rod worth value and comparison with experimental data was made to evaluate the safety of the r eactor. IAEA-CN-156/S-10 Operational Experience and Programmes for Optimal Utilization of the Nigeria Research Reactor-1 S.A. Jonah, G.I. Balogun, A.I Obi, Y.A. Ahmed, M. Kyari, B.Nkom Reactor Engineering Section, Centre for Energy Research and Training, Ahmadu Bello Univerisity, Zaria, Nigeria A.A. Mati, I. Yusuf Engineering and Instrumentation Design Section, Centre for Energy Research and Training, Ahmadu Bello Univerisity, Zaria, Nigeria E-mail address of main author: jonahsa2001@yahoo.com Abstract The Nigeria Research Reactor-1 (NIRR-1) is the nation’s first nuclear reactor and it is sited at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria. It is a Miniature Neutron Source Reactor (MNSR) that attained criticality on February 03, 2004 and was licensed to operate at a maximum power of 31 kW three days a week in June 01, 2004. This presentation enumerates the measures put in place to ensure safe operation and adequate maintenance regime as well as the strategic plans for optimal utilization of the reactor. Some of these measures, which bothers on safe operation and sustainable maintenance culture that have been implemented include: strict adherence to the periodic preventive maintenance routines; standard procedures for pre- startup, startup and shut down procedures; provision of a quick access to reactor top to facilitate rapid response in case of emergency, especially in the case of rod-stuck incident. Similarly, on the basis of experience gained since the commissioning vis-à-vis the neutron flux spectrum characteristics of the MNSRs, experimental protocols are presented for the analysis of elements producing short-lived, medium-lived and long- lived activation products in geologic materials with negligible nuclear interferences especially for the analysis of Al in the presence of Si. Furthermore, research and development activities in core physics analysis and thermal hydraulics with regards to conversion from the current HEU core to a LEU core under the aegis of the IAEA Coordinated Research Project entitled “Conversion of MNSR to LEU” are outlined. IAEA-CN-156/S-32 IAEA-CN-156/S-32 IAEA-CN-156/S-32 IAEA-CN-156/S-47 Comparative dose calculation for TRIGA HEU and LEU fuel in nuclear accident situations Sorin Margeanu, Cristina Alice Margeanu, Marin Ciocanescu, Constantin Paunoiu Institute for Nuclear Research Pitesti, Romania E-mail address of main author: margeanu@sc n.ro INR-Pitesti TRIGA research reactor is basically a pool type reactor with a special design in order to fulfill the requirements for material testing, power reactor fuel and nuclear safety studies. The dual-core concept involves the operation of a TRIGA-SSR high-flux, steady-state research and material testing reactor at one end of a large pool, and the independent operation of an annular-core pulsing reactor (TRIGA-ACPR) at the other end of the pool. The steady- state reactor is used for long term testing of power reactor fuel components (pellets, pins, subassemblies and fuel assemblies), and the annular-core pulsing reactor is used for transient testing of power reactor fuel specimens. The safety evaluation involved a several design basis accidents [1]: single-pin cladding failure in water, 25-pin fuel bundle failure in water and in air, accidental reactivity insertions, loss of flow from main coolant pump accidents, and interaction between the two cores within the tank. One fission product release category were analyzed - a design basis release. The failure of a single fuel pin classing (due to material deficiencies) with the consequent release of fission products is an event that has a small but significant probability and, over the life of the core, it could be expected that such a failure could occur during normal operation. For the design basis release has been assumed to be no retention of volatile fission products in the fuel- moderator material. This failure was analyzed using the following assumptions: a) The fuel pins that fail have operated at an average power density of twice as great as the average power density in the core; b) The core has operated continuously for a total of 7700 MWd; c) For the anticipated release, experimentally determined fraction of volatile products released from the fuel material [2] will be about 6.3E-04; d) For the design basis release, there is no retention of volatile fission products in the fuel- moderator material; e) 100% of the noble gases in the fuel-clad gap are released from the fuel bundle and, subsequently, are transferred directly to the reactor hall. 25% of the halogens are released from the fuel bundle (with the remainder assumed to plate-out on the relatively cool clad). As regarding the halogens that escape for the anticipated (single-pin failure) release, 10% are assumed to form organic compounds which escape in the pool water. Only 1% of the balance is undissolved in the pool water and appears in the reactor hall air. The net halogen release to the reactor room and potentially outside is 2.725%. All other fission products remain in the pool or are otherwise unable to escape from the reactor room. For the design basis release (in air) fully 25% of the halogens are released to the reactor hall. All other fission products are assumed unable to escape from the reactor room because of plate-out on cool surfaces [2]; f) The release height is assumed to be 50 m above the ground level; g) For this case, the fuel bundle is in air. The assumption for this case, that there is no retention of fission products in the fuel, implies that the fuel involved is at temperatures IAEA-CN-156/S-47 approaching the melting point of U-ZrH. That is inconceivable because the fuel bundle is in water. For training purposes, and to exercise our ability to conduct Level-3 PSA studies, a severe accident scenario involving 14-MW INR-TRIGA research reactor has been developed. In this scenario is assumed that a large part of the reactor hall roof or a heavy object escaped from the crane hook is dropped over the 14-MW TRIGA-SSR core, resulting in mechanical damage of the core. It is assumed, also, that no core melting is occurring, but only fuel- cladding rupture being involved for several 25-pins fuel bundles. The accident was analyzed using the following assumptions: a) The core has operated continuously for a total of 7700 MWd; b) The affected fraction of the core is 45% c) For the release, experimentally determined fraction of volatile products released from the fuel material [2] will be about 6.3E-04; d) There is no retention of volatile fission products in the fuel-moderator material; e) 100% of the noble gases in the fuel-clad gap are released from the fuel bundle and, subsequently, are transferred directly to the reactor hall. 25% of the halogens are released from the fuel bundle (with the remainder assumed to plate-out on the relatively cool clad). As regarding the released halogens, 10% are assumed to form organic compounds which escape in the pool water and 1% appears in the reactor hall air. There is no retention of fission products in the considered fuel (as shown above), and due to the high temperature of fuel a 2.0E-02 release fraction was assumed for other fission products than noble gases and halogens; f) During the accident evolution, an emergency ventilation system occurs and charcoal traps are not available, so no fission products will be retained by traps; g) The release height is assumed to be 60 m above the ground level; h) The radiological consequences assessment has been performed with PC-COSYMA computer code [3], considering a site specific meteorological file (being generally known that the dispersion model -MUSEMET- used in PC-COSYMA version of the computer code have a bug for single stability class type calculation, for the D stability class). By using a meteorological file with hourly changes in wind direction for each hour at each phase option considered, the arise of bug is avoided; i) The calculation was performed by considering both possibilities regarding countermeasures: no countermeasures for the general public and the environment are taken, and countermeasures are implemented considering site specific intervention parameters [4, 5]; j) As regarding the calculations, both deterministic and probabilistic approaches have been used. The evaluation of the radiological consequences considers both early and late consequences. [1] Design and safety evaluation of INR 14-MW TRIGA research reactor, Gulf General Atomic Report (1974). [2] Foushee, F. C., and Peters R. H., “Summary of TRIGA Fuel Fission Products Release Experiments”, Gulf Energy & Environmental Systems Report Gulf-EES-A10801 (1971). [3] PC COSYMA (Version 2): An accident consequence assessment package for use on a PC, Commission of the European Communities, EUR 16239, Brussels, 1996 [4] “On-site Emergency Intrevention Plan”, Pite3ti (2007). [5] Margeanu S., “Level 3 - PSA with PC-COSYMA for TRIGA reactor”, in Proceedings of TRIGA Conference, 2000. IAEA-CN-156/S-8 Safety Analysis of MNSR Reactodru r ing Reactivity Inser tion Accident Using the Validated Code PARET A. Hainoun, F. Alhabit Reactor Safety Division, Nuclear Engineering Department, AECS, Damascus, Syria E-mail: A.Hainoun@gmx.ne, tor ahainoun@aec.org.sy In the frame work of the IAEA's CRP proje (cJt7.10.10) on "Safety significance of postulated initiating events for various types of researecha ctors and assessmenta onfa lytical tools" the Syrian team contributed in the assessmen tc oomf putational codes related to the safety analysis of research reactors [1]. Duritnhge project implementation the codes PARET and MERSAT have been tested, modified and fvi edri regarding specific phenomena related to safety analysis of research reactors [2]t.h Ien framework of this contribution the code PARET has been applied to model ther ec o f the Syrian MNSR reactor. The code analysis includes the simulation of steady state operation and oau pg rof selected reactivity insertion accident (RIA) including the design basis accidents dnega lwi ith the insertion of total available excess reactivity. For this purpose a PARET input model for tchoer e of MNSR reactor has been developed enabling the simulation of neruotn kinetic and thermal hydracu liof reactor core including reactivity feedback effects. The neutron ktiicn emodel depends on thpeo int kinetic with 15 groups of delayed neutrons including photo neut roofn bseryllium reflector. In this regard the effect of photo neutron on the dynamic behoauvr ihas been analysed through two additional calculations. In the first the yield of photo untreons was neglected completely and in the second its share was addedt htoe sixth group of delayed nerount s. In the thermal hydraulic model the fuel elements witth eir cooling channels were sdtriibuted to 4 different groups with various radial power factso.r The pressure loss factors ffroicr tion, flow direction change, expansion and contraction were esattiemd using suitable approaches. Figure 1 presents the evolution of relative rera pctoower after a step retaivcity change of 1 mk starting from 20% of reactor nominal poweFri.g ure 2 shows the developments of average core temperature following a complete withdroafw r eactor control rode from the cold core condition corresponding tao reactivity insertion of about 6 mk . The results of these RIA simulation show goode aegmrent with the experimental data. Thus, it can be concluded that the code PARET psoss sgeood ability to model the expected thermal hydraulic and neutron dynamic phenomena. Pualartricly the results nidicate the correct simulation of inherent safety features of MSRN reactor resulting fro mthe strong reactivity feedback effects of coolant temperaet urnder natural circulation conditions. 1/2 IAEA-CN-156/S-8 Fig.1. Relative power distribution after a step cretivaity change starting from the initial power le voef l 20% of nominal reactor power. Fig.2.Evolution of relative average core tempteurae following a complete withdraw of reactor control rode. [1] Hainoun, A., Gazi, N., Alhaibt, F. 2006. Safety Significant of PIE for Research Reactors and Assessment of Analytical Tools, 3rd RCM Meeting, IAEA Vienna. [2] Hainoun, A., Gazi, N., Alhabit, F. 2007. Mofidciation and Validation of the Natural Heat Convection and Subcooled Void FormatioMno dels in the Code PARET, (to be published). IAEA-CN-156/S-19 Fuel management methodology upgrade of Thai Research Reactor (TRR-1/M1) using SRAC computer code C. Tippayakul, D. Saengchantr Thailand Institute of Nuclear Technology, Bangkok, Thailand E-mail address of main author: chanatipt@hotmail.com Thailand Institute of Nuclear Technology (TINT) is currently responsible for the nuclear research reactor called “TRR-1/M1” which is located in Bangkok. The existing fuel management tool for TRR-1/M1 is a computer program called TRIGAP [1] which was developed in Slovenia during the 80’s. Although TRIGAP is capable of calculating reactor parameters such as core excess reactivity or neutron fluxes, this tool has several drawbacks. Since TRIGAP only models the spatial distribution of neutrons in cylindrical geometry, the TRR-1/M1 core, which is formed in hexagonal lattices, needs to be homogenized into cylindrical rings. As a result, TRIGAP is unable to provide pin-wise data such as normalized power distribution of the reactor. To overcome this, an upgrade to the existing methodology is proposed. The upgraded methodology is actually similar to the existing methodology that it is executed in 2 steps. However, both steps are performed by more advanced computer programs collectively packaged into one system called “SRAC” [2] which has been developed in Japan since 1978. The first step, which is the group cross section generation, is performed by the PIJ module of SRAC system utilizing 2D collision probability method. Typically, the group cross section generation is performed using infinite arrays of 2D lattice models corresponding to unique lattice regions of the reactor. In essence, each 2D lattice model represents an axial node which has the same material throughout axial direction. For TRR-1/M1, there are three types of rods: fuel element, fuel follower control rod and air follower control rod. Each type of the fuel rods is divided axially into different types of 2D lattice models. Furthermore, the water lattice model is created to represent the lattice filled completely with water. To generate the group cross sections, these 2D lattices are classified into two model types, i.e., fuel type lattice and non-fuel type lattice. The fuel type lattice is performed as it is in the group cross section generation while the non-fuel type lattice is performed in a special color-set model to simulate the environment effects from surrounding fuel rods. Moreover, the group cross sections of the fuel type lattice are generated at various burnup points from fresh to very high burnup and they are also generated at different power levels which correspond to different equilibrium temperatures. Following the group cross section generation step, the reactor calculation step is performed. The upgraded methodology uses the COREBN module of the SRAC system to perform the reactor core calculation. The COREBN module has a capacity of performing burnup calculation and modeling hexagonal lattice of TRR-1/M1 core. To validate the upgraded methodology, the group cross sections of different lattices needed for the reactor core calculation were generated by the PIJ module of the SRAC system and the reactor core calculations of TRR-1/M1 core number 1 and 2 were performed afterwards. The Keff of the “all-rods-out” models of the reactor cores were derived and the excess reactivity was calculated by (Keff -1)/( Beta*Keff) where Beta is fraction of delayed neutrons (0.007). Table I presents the core excess reactivity results of TRR-1/M1 core number 1 and 2 with the comparison against operation data from the operation log book. IAEA-CN-156/S-19 Table I: Core excess reactivity results of TRR-1/M1 core number 1 and 2 Model Keff by SRAC code Calculated core excess Operation core excess reactivity by SRAC code reactivity Core #1 1.05809 7.84$ 7.43$ Core #2 1.05252 7.13$ 6.87$ As it can be seen from Table I, the excess reactivity calculated by SRAC system agrees well with the operation data when considering that the operation value has inherently some amount of measurement uncertainty. There seems to be a bias of around 0.40$ between the calculated results and the operation data. Also, the change of core excess reactivity as a function of power level is obtained from series of reactor core calculations at various equilibrium temperatures. Fig. 1 presents the changes of core excess reactivity as a function of power level of TRR-1/M1 core number 1. 7.00 6.00 5.00 4.00 3.00 2.00 1.00 0.00 0 200 400 600 800 1000 1200 1400 1600 1800 2000 Power level (kW) Fig. 1: Core excess reactivity of TRR-1/M1 core #1 as a function of power level (kW) From Fig. 1, the core excess reactivity decreases quite linearly as a function of power level as expected. This result confirms the negative temperature feedback of the TRR-1/M1. It can be concluded that the upgraded methodology is functioning well and it can be used routinely as a fuel management strategy for TRR-1/M1. [1] I. Mele, M. Ravnik, “TRIGAP – A computer programme for research reactor calculations”, 1985 [2] K. Okumura, T. Kugo, K Kaneko and K. Tsuchihashi, “SRAC (Ver. 2002); The comprehensive neutronics calculation system”, 2002 Core excess reactivity ($) Poster Presentations: Utilization Synopses no. IAEA-CN- Synopses Title Main Author 156/ Operating Experience Utilization Programmes of the BAEC 3 MW TRIGA Mark-II Research Reactor in U-54 Bangladesh Haque, M. Studies on environmental pollution in Bangladesh using reactor based neutron activation analysis U-55 technique Abdul Latif, S. Improvement on Sensitivity for the Track - Etch U-8 Neutron Radiography Pugliesi, R. New Perspectives for the TRIGA IPR-R1 Research U-10 Reactor Soares Leal, A. First experiments in the IPEN-CNEN/SP PSD Neutron U-15 Powder Diffractometer Parente, C.B.R. Future utilization of the Research Reactor IRT IN U-45 SOFIA after its reconstruction Ilieva, K.D. U-38 Utilization of the LVR-15 Research Reactor at Rež Marek, M. U-49 Utilization of the VR-1 Training Reactor Matejka, K. The 250 kW FiR 1 TRIGA Research Reactor - International Role in Boron Neutron Capture Therapy (BNCT) and Regional Role in Isotope Production, U-28 Education and Training Auterinen, I. Chemical characterization of early fine-ware pottery by neutron activation analysis: analytical and U-6 statistical approaches to production and trade Kilikoglou, V. U-56 Feasibility Study of I125 Brachytherapy in Indonesia Soentono, T.M. Characterization of Airborne Particulate Matter at Urban and Rural Area in Bandung and Lembang Indonesia using Instrumental Neutron Activation U-17 Analysis Santoso, M. Characterization of a Neutron Collimator for Neutron U-14 Radiography Applications Palomba, M. Optimizing Conditions Suited for Stress U-33 Determinations in Q-Space Focusing Configurations Ionita, I. U-34 SANS facility at the Pitesti 14MW TRIGA Reactor Ionita, I. RIAR Capabilities in Support of the Innovative U-19 Nuclear Technologies Bychkov, A.V. U-21 SM Reactor After Core Modernization Klinov, A. Set of Investigations of HFR Fuel Rods in Justification U-22 of their Serviceability and Safe Operation Tsykanov, A.V. U-12 Utilization of irradiation holes in HANARO Lee, C.S. Design and Installation of Fuel Test Loop in U-27 HANARO Ahn, S.H. Design Characteristics of Cold Neutron Source in U-42 HANARO Wu, S.I. Pure Commercial Gold Foils As Neutron Flux Monitor U-47 Neutron Self-Shielding Assessment Helal, W. Implementation of TRR-1/M1 for Thailand’s Nuclear U-50 Engineering Program Nilsuwankosit, S. IAEA-C N-156/U-54 Operating Exper ience and Utilizanti oProgrammes of the BAEC 3 MW TRIGA Mark-II Researc hReactor of Banglades h M. Monzurul Haque, M. A. Malek Soner , P.K. Saha, M.A. Salam and M.A. Zulquarnain Reactor Operation and Maintenance UAntito, mic Energy Research Establishment Bangladesh Atomic Energy Commission,n Gakabari, Savar, Dhaka, Bangladesh. E-mail: romu@dhaka.ne; tmmhaque_2000@yahoo.com Synopsis The 3 MW TRIGA Mark-II research reacr tof Bangladesh Atomic Energy Commission (BAEC) has been operating since September 11948,6 . The reactor is used for radioisotope production 1(31I, 99mTc, 46Sc), various R&D activities, manpower training and education. The reactor has been operated suscfcuellsy since it’s commissionig with the exception of a few reportable incidents. Of these, the decay tanak alege incident of 1997 icso nsidered to be the most significant one. As a result of this idnecni t, reactor operation at full power remained suspended for about 4 years. Hovwere, the reactor operation wasn ctionued during tihs period at a power level of 250 kW to cater the needs of various R & D groups, which required lower neutron flux for their experiments. This was dme apossible by establishing a temporary by pass connection across the decay tank using local technol oTghye. reactor was made operational again at full power after successful replacemoef ntth e damaged decay tank in August 2001. At that time, several modifications of the reactorl icnogo system along with its associated structures were also implemented and then necessasrtyin tge and commissioning otfhe newly installed component/equipment were carried out. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in Mrcah 2002 when a pyrex vial containing 50g of TeO2 powder got melted inside the DCT. The vwials melted due to high heat generation on its surface while the reactor was operated for 8 hoaut r3s MW for trial production of Iodine-131 (131I). A Wet Central Thimble (WCT) was usteod r eplace the damaged DCT in June 2002 such that the reactor operation could be resum Tehde. WCT was again replaced by a new DCT in June 2003 such that radioisotoppreo duction could be continued. The facility has so far been used to ntr auip a total of 27 presonnel including several foreign nationals to the level oSfenior Reactor Operator (SRaOn)d Reactor Operator (RO). The reactor is operated 4 days a week at a p olweveerl of 3 MW for production of Iodine-131. During the other one weekday, the reactor iesr aotped at lower power levels (250 – 500 kW) to cater the needs of NAA and NR groups. At prestheen tt otal burn-up of theco re stands at about 484 Megawatt Days. BAEC has a plan to incretahsee p roduction of Iodine-131 to install more dry tubes in the core so as mtoe et the total demand of RI ine t hcountry. There is also plan to develop unused experimental facilities such as, thermal column and radial beam ports for strengthening the R& D activitsie around the reactor. A total 1o0f 93 irradiation requests (IRs) have been catered so far for different reacutsoers . The total amount of RI produced stands at about 4000 GBq. The total amount of burn--fuepl is about 11607 MWh. Efforts are on to undertake an ADP project so as c toonvert the analog console andC I &system of the reactor into digital one. The paper summarizes the troera coperation, maintenance experiences and utilization programmes focusing on troubleshootirnegc,t ification, modification, RI production, various R&D activities and training proagmr being conducted at the facility. Keywords: Reactor, Dry Central Thimble (DC TW),et Central Thimble (WCT), demineralize water, irradiation request (IR), pyrex v,ia lTeO2 powder, radioisotope, burn up. IAEA-C N-156/U-54 References [1] "2nd World TRIGA Conference ViennaA, ustria, 15-19 September 2004" Operation Experience with the 3 MW TRIAG Mark-II Research Reactor oBfangladesh, M. S. Islam, M. M. Haque, M. A. Salam, M. M. Rahman, MR. I. Khandokar, M. A. Sardar, P. K. Saha, A. Haque, M. A. Malek Soner, M. M. UddinS,. M. S. Hossain and M. A. Zulquarnain. [2] "Operation and Maintenance Status oef t3h MW TRIGA Mark-II Research Reactor of Bangladesh"- Md. Monzurul Haque, M.A.MaleSko nar, P.K. Saha, M.A. Salam, M. Ali Zulquarnain, International seminar on nuclseafre ty, operation and maintenance of nuclear facility course, RADA, Japan, Sep. 26-Oct. 07, 2005. [3] "Strengthening Operational Safety throuMgho dification of CoolingS ystem and Upgrading Safety Analysis Report of BAEC Research Rtoera"c- M. A. Zulquarnain, M. M. Haque, M. A. Salam, P. K. Saha, A. Haque, M. M. Uddin, and M. A. M. So Ineter,rnational seminar on nuclear safety, safety analycsoisu rse, RADA, Japan, Dec.05-17, 2005. [4] Safety Analysis Report for the BAEC M3 W TRIGA Mk-II Research Reactor at AERE, Savar, April 2006. IAEA-CN-156/U-55 Studies on environmental pollution in Bangladesh using reactor based neutron activation analysis technique Sk. A. Latif, S. M. Hossain, M.S. Uddin, M.A. Hafiz and M.A. Islam Institute of Nuclear Science & Technology (INST), Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. E-mail: sklatif1967@yahoo.com Environmental and health related problems have become a major global concern in the recent years. Bangladesh is now facing a serious problem about arsenic (As) and chromium (Cr), which contaminate our environment. Arsenic exposure is a potential health risk to local populations in most of the parts of Bangladesh. Hazaribagh is a densely populated area of Dhaka city in Bangladesh, where about 149 leather-processing industries are in operating and discharge a lot of solid and liquid waste directly to the low-lying areas, river and natural canals without proper treatment. These tanneries follow the practice of chrome tanning. In this practice the leather takes only 50-60% of the applied chromium and the remaining is discharged as waste. The pollution load emanating from tanneries is directly affecting surface water, ground water, soil and air. A number of people have been affected directly and indirectly with chromium of wastes from tannery industries. Environmental research using instrumental neutron activation analysis (INAA) for the determination of trace and ultra-trace element pollutants has a great potential in relation to human health. The total element concentration has been traditionally used to assess environmental impact and health risk of the element. We have a 3MW TRIGA Mark-II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. Our interest is to study environmental pollution due to As and Cr through distribution in environment over Bangladesh. Particularly, this work was undertaken for determining As content in water, soil and herbal plants, and Cr-content in soil of tannery industrial areas as a part of our systematic studies. For Cr-determination, thirty soil samples were collected at several depths of different locations (L-1, L-2 ......) from both of Hazaribagh tannery area and the Tatuljhura proposed tannery area. The samples were dried in an oven at a temperature of about 70° C until they attained constant weight followed by ground. Two sets of samples for Hazaribagh and Tatuljhura were prepared for separate irradiations. The samples and standards (IAEA Soil-7 & SL-1 and NIST Coal Fly Ash (CFA) 1633b) were irradiated with thermal neutrons of ~ 2 12 -2 -1 ×10 n cm sec for 3 hours at 250 kW in Lazy Susan of the 3 MW TRIGA Mark-やや research reactor. The activities of irradiated samples were measured by HPGe gamma-ray spectroscopy system coupled with computer based S100 MCA acquisition software. Arsenic levels in water, soil and herbal samples collected from 10 individual locations of Sonargaon in Narayanganj district were determined by INAA technique. These samples were irradiated at the same irradiation conditions as chromium. Quality control (QC) test is performed to investigate the reliability of the analysis by measuring chromium concentration in certified reference materials CFA, SL-1 and Soil-7 relative to primary standard. The experimental results are varied with the certified values within 6グ. The deviation was achieved within the uncertainties quoted with the certified values. Therefore, INAA is an efficient method to determine chromium in soil samples. IAEA-CN-156/U-55 50 51 The Cr-element was identified via the Cr(n,i) Cr reaction. The gamma ray emitted from 51 Cr at 320.1 keV was not interfered from other short-lived radioactive nuclides since the samples were measured after long cooling time. In Hagaribagh, the Cr-content in the range of 880 to 33550 ppm was found in surface soils and it was 71-90 ppm for Tatuljhura. For Hagaribagh, the Cr-concentration decreases with the increasing of depth upto 180 cm, and then scattered results were found as shown in Fig.1. The Cr-concentrations of soils in Hagaribagh are rather high and in the most cases these are above permissible level. The Cr- concentration of Tatuljhora was in the range of 50.87 to 93.63 ppm, which is below the permissible level reported in worldwide (Sandia Corporation, 2000). Arsenic was detected in each of ten soil samples in the concentration range of 0.63-11.35 ppm, where only one was under permissible level (2 ppm). Five water samples out of ten contain arsenic above permissible level (0.05 ppm). Surprisingly, arsenic was also found 1.0E+05 Hazaribagh (L-1) Hazaribagh (L-2) 1.0E+04 Hazaribagh (L-3) 1.0E+03 1.0E+02 1.0E+01 1.0E+00 0 50 100 150 200 250 300 Depth (cm) Fig. 1 Chromium content in soil samploefs the tannery area in Hazariba gh considerably high in some herbal samples. The arsenic was not found in some herbal samples corresponding to the water samples where arsenic was not detected. Therefore, the presence of arsenic in ayurvedic herbal medicine may be done through the contamination of herbal plants with arsenic contaminated water and soil. The obtained results will play important role to create public awareness on contamination with As and Cr. This study reports baseline data for the proposed tannery industry in Tatuljhura that will help to assess the level of contamination when the tannery industries will discharge their waste in the environment. Reference 1. Sandia Corporation 2000. Chromium Background Soil Levels http://www.state.ma.us/dept/eesector/gs/gc/na/ chromiumsoillevels.html Cr-content og.g-1 IAEA-CN-156/U-8 Improvement on Sensitivity for thTer ack - Etch Neutron Radiography R. Pugliesi, F.Pugliesi, V.Siacni, M.A.Stanojev Pereira Instituto de Pesquisas Energceatsi e Nucleares (IPEN-CNEN/SP) Av. Prof. Lineu Prestes 2242 Ciddea Universitária, Butantã CEP 05508-000 São Paulo-SP, Brazil E-mail address of main author: pugliesi@ipen .br The use of solid state nuclear track deotresc(tSSNTD) to record neutron radiography(NR) images is a well known technique. The radioghrya ips obtained by irradiating a sample in an uniform neutron beam and a converter screaens tfrorms the transmitted neutron intensity into ionizing radiation which is able to causem daages into the detector. Usually boron based converter screens are used and in this case paalprhtiacles and lithiumo ins cause the damages. By means of a chemical etching the latenmt adgaes are enlarged and are called tracks and, they form a two-dimensional image which viiss ible by the naked eye. One of the main restrictions to the employment of track detesc tfoorr neutron radiography is the low-intrinsic optical contrast achieved in the recorded iem athgat leads to a poosre nsitivity to discern thickness changes of the materials. The sievnitys itis determined by measuring the light transmitted through the radiographed image and for such purpose, conventional analog optical microphotometers are typically employed. In the present work a digital system, cotninsgis of a photo enlarger, video camera, capture frame grabber and a computer has been employed to measure the light transmitted and a significant improvement in the sensitivity was aecvheid. The light intensyi tis evaluated in a 8 bit gray level scale. The track detector CR-39,o 5m0 0thick was used to record the images and the chemical etching was performed ain KOH(30%) aqueous solution at a constant temperature of 7o0C [1]. The sensitivity has been deteinrmed for two materials, Plexiglas and iron. The samples are step wedges with thiscskense varying from 2 mm to 12 mm which have been radiographed in a NR facility installaetd t he radial beam-hole 08 of the 5MW IEA-R1 Nuclear Research Reactor of the IPEN-CNSEPN /[2]. The sample-detector set has been irradiated at a neutron exposure of E = 49xn1/c0m2 and the detector etched for 25 minutes. For these conditions the recorded image exhitbhiets h ighest intrinsic optical contrast in the detector[3]. The sensitivity expressed as the minimum detectable thickness, is numerically evaluated by: Fx = -F(GL)/C.....................(1) where “F(GL)“ is the minimal discernible gray leveinl tensity in the image and “C” is a constant depending of the sample and of the digital system. The FIG 1 shows the behavior of the gray levinetle nsity(including thed etector background) as functions of the sample thicknesses. For the present digital sFy(sGteLm) = 2.5 and the obtained results for the sensitivity are shownT ainb le I. The uncertainties in the results have been determined by the standarodp pargation method applied to (1). IAEA-CN-156/U-8 160 140 Plexiglas 120 100 Iron 80 60 0 2 4 6 8 10 thickness(mm) FIG. 1: Behavior of the gray level inteitniess as functions of the sample thickness. Fx(mm) Plexiglas Iron Digital system 0.26‒0.01 0.50‒0.02 Microphotometer 0.47‒0.01 0.75‒0.03 TABLE 1. COMPARISON OF THE SENSITIVITY VALUES. In order to compare the potential of the pernets digital system theT able 1 shows also the values of the sensitivity as evaluatedc,c oarding to the samep rocedures, by using a microphotometer [3] to analyzee t hlight transmission. As can bse en the sensitivity provided by the digital system overcomes the one provided by the microphotometer, demonstrating that the former can be an important too li mtoprove the neutron radiography technique. The rapid data acquisition is a lsaon important characteristic othfis system. Each gray level intensity and its corresopnding uncertainty are evaluated bye raavging the intenitsies of about 1700 individual pixels in an area corresponding to about 02.4 o cf mthe image. This procedure takes some few seconds. For the microphoteorm tehte reading procedure takes about 30 minutes in an area approximately 200 times smaller. [1]. Pugliesi, R.; Pereira, M.A.S.; Moraes, AM.P. .V., 1999. Characteristsic of the Solid State Nuclear Track Detector CR-3fo9r Neutron Radiography Purposes. Applied Radiation and Isotopes, 50(2), 375-380. [2]. Pugliesi, R., 2001. Progresosf Neutron Radiography at eth IEA-R1 Nuclear Research Reactor (last 15 years). Menegti of the Advisory Group for Dve lopment and Application of Neutron Radiography. Int. Atomic Energy eAngcy(IAEA) Vienna, Austria 1-4, October [3]. Pereira S. M. A., 2000. Employment tohfe Polycarbonates Makrofol-DE and CR-39 for Neutron Radiography. M.Sc. Thesis. Nucr leEanergy National Commission. IPEN-CNEN/SP BRASIL. gray level IAEA-CN-156/U-10 New Perspectives for the TRIGA IPR-R1 Research Reactor A.S.Leal, R.C.O. Sebastião, R.R. Rodrigues Nuclear Technology Development /CDTN – Brazilian Nuclear Energy Commission/CNEN Rua Professor Mário Werneck s/n - 30123-970, Belo Horizonte, MG, Brazil E-mail address of main author: asleal@cdtn.br The TRIGA IPR-R1 CDTN´s research reactor is of the type MARK I, which core is below the floor level, as it is shown in Figure 1. It is operating since 1960 and the main activities have been the neutron activation analysis and the training of nuclear power plant operators [1]. Since 2001, new projects of utilization of the reactor were initiated as the production of some special labeled molecules to be used in medical purposes the improvement of color of Brazilian gemstones by neutron irradiation [2,3]. More recently, CDTN started a project to evaluate the possibility of obtain a characterized neutron beam from a vertical tube from the TRIGA's core. The availability of a neutron beam with appropriate characteristics as intensity, spectrum and collimating, will enlarge the possibilities de application in these fields and also will open interesting perspectives of new applications such as the study of DNA structures by neutron irradiation and neutrongraphy [4]. This work presents the recent results of these new applications, about gemstones, special activated molecules and also the preliminary simulated results of the neutron extractor of the using the MCNP code. Figure 1 - View of the well of the TRIGA IPR MARK 1 reactor [5]. IAEA-CN-156/U-10 [1] Maretti-Júnior, F., Fernandes, M. P., Oliveira P.F., Amorim, V. A., 1999. Proc. 7th Meeting of the International Group on Research Reactors (IGOOR), (1999) San Carlos de Bariloche, Argentina. [2] Leal, A.S., Krambrock K., Ribeiro, L.G.M., Menezes, M.Â.B.C., Vermaercke, P., Sneyers, L., Nucl. Instr. and Meth. A, Study of neutron irradiation-induced colors in Brazilian topaz, accepted to publish (2007). [3] Leal, A.S., Carvalho Jr., A.D., Abrantes, F.M., Menezes, M.A.B.C., et.al., Production of the radioactive antitumoral cisplatin, Applied Radiation and Isotopes, 64, (2006), 178-181. [4] Gual MR, Rodriguez O, Guzman F, Deppman A, Neto JDTA, et al.,. Study of neutron- DNA interaction at the IPEN BNCT research facility, Brazilian Journal of Physics, 34, (2004) 901-903. [5] Dalle, H. M., Ph.D Thesis, UNICAMP/São Paulo/Brazil, (2005). IAEA-CN-156/U-15 First exper iments in the IPE-NCNEN/SP PSD Neutron Powder Diffractometer* C. B. R. Parentae, V. L. Mazzocchai, J . Mestnik-Filhoa, Y. P. Mascarenhab s a Instituto de Pesquisas Energéticas e Nucleares (IPEN-CNEN/SP), São Paulo, SP, Brazil b Instituto de Física da Universidade de São Paulo (IFSCar), São Carlos, SP, Brazil E-mail address of main author:cparente@ipen.br The neutron powder diffractometer, recently installed at the IEA-R1 3.5 MW research reactor, is equipped with a position sensitive detector (PSD) [1]. The PSD spans 20° of a neutron powder diffraction pattern. An extended pattern can be obtained by measuring intensities in 20° segments in a 2し angular interval ranging from 5 to 125°. At a take-off angle of 84°, a focusing Si monochromator [2], installed in the instrument, can be easily positioned in order that the following wavelengths become available: 1.111, 1.399, 1.667 and 2.191 Å. A rotating-oscillating collimator (ROC) was also installed in the new instrument. The ROC eliminates parasitic scattering from furnace or cryorefrigerator heat shields in the vicinity of the sample. It also makes the PSD less sensitive to the ambient background. The new diffractometer can measure a neutron powder diffraction pattern in a matter of ten to twenty hours with a good statistics. (See reference [3] for more details about this instrument). Figure 1 is a photograph of the IPEN-CNEN/SP PSD neutron powder diffractometer. FIG. 1. The IPEN-CNEN/SP PSD Neutron PowDdeiffrr actometer. (Photo by Marcello Vitorino). After calibration of the instrument, a series of experiments were done in order to determine its operational characteristics. In all experiments, wavelength used was そ = 1.399 Å. Samples used for this purpose were powdered Ni, Si and Al2O3. Figure 2 is a comparison between two neutron powder patterns for Ni measured with the old and the new diffractometers. The old * Instrument financed by Fundação de Amparo à Pesquisa do Estado de São Paulo (FAPESP) under Project no. 95/05173-0. IAEA-CN-156/U-15 diffractometer (commissioned in 1966) was a single detector (BF3)/single wavelength (そ = 1.103 Å) instrument. It was installed at the same place where is now installed the new one. Observing Figure 2, it is worthwhile to note the improvement in resolution of peaks when using the new instrument particularly for large values of 2し. Another point to be highlighted in the comparison is that, although having four times the number of intensity points (for a same angular 2し interval), the new pattern took ca. one third of the time spent in the measurement of the old one. Several other patterns of many different compounds as, for example, Fe2O3 (magnetite), CeO2, SrTi0.65Fe0.35O3-h and BaY2F8:Nd were also obtained with the new diffractometer. Utilization of the IPEN-CNEN/SP PSD neutron powder diffractometer is open to brazilian and other latin-american scientific and technological communities. 20 15 10 5 0 10 20 30 40 50 60 70 80 90 100 110 120 130 2s (degrees) FIG.2. Comparison between neutron powder pattefornr sN i measured with the old (above) and the new (below) instruments. [1] BERLINER, R. et al., “A large area position sensitive neutron detector”, Nucl. Instr. and Meth. 185 (1981) 481. [2] POPOVICI, M., YELON, W. B., “A high performance focusing silicon monochromator”, J. Neutron Research 5 (1997) 227-239. [3] PARENTE, C. B. R. et al., “The New IPEN-CNEN/SP Neutron Diffractometer”, Proceedings of the International Conference on Research Reactor Utilization, Safety, Decommissioning, Fuel and Waste Management, IAEA, Santiago, Chile (2003). Contributed Papers and Posters (CD-ROM). 3 Intensity (10 neutrons) - 111 - 002 - 022 - 113 - 222 - 004 IAEA-CN-156/U-45 FUTURE UTILIZATION OF THE RESEARCH REACTOR IRT IN SOFIA AFTER ITS RECONSTRUCTION Krassimira Ilieva, Tihomi Apostolov, Sergey Belousov Institute for Nuclear Research and Nuclear Energy of Bulgarian Academy of Science, Tzarigradsko shossee 72, Sofia 1784, Bulgaria The research reactor IRT-2000 (IRT) in Sofia to the Institute for Nuclear Research and Nuclear Energy (INRNE) was built and put into operation in 1962. It was temporarily shut down in 1989 for improvement. The reconstruction of the IRT is being carried out under the decision of the Council of Ministers of Republic of Bulgaria from 2001. The strategy for sustainable utilization considers the IRT as a national base and aims to satisfy the society needs for: ö" education of students and training of graduated physicists and engineers in the field of nuclear science and nuclear energy, ö" implementation of applied methods and research, ö" development and preservation of nuclear science, skills, and knowledge. The IRT Technical Design is being in process of elaboration. The IRT will be reconstructed into a reactor: ‚" of thermal power 200 kW; ‚" with low enriched fuel, with uranium-235 enrichment below 20% in accordance with the current requirements of the security of transportation and storage of nuclear and other radioactive materials which are vulnerable to theft by terrorists; ‚" with ten vertical and seven horizontal experimental channels which will supply maximal fast neutron flux about 3. 1012 n/cm2s, and maximal thermal flux about 8.1012 n/cm2s; ‚" with channel which will supply epithermal neutron flux about 0,9.109 n/cm2s suitable for medical Boron Neutron Capture Therapy (BNCT) application. The INRNE together with the Technical University in Sofia have proposed to the Ministry of Education a new programme for education of students in nuclear energy. The Nuclear Energy course will be obligatory for obtaining the Master of Science Degree of the Technical University in Sofia. The educational classes refer: types of research reactors, main characteristics and design of the reconstructed IRT, safety assuring and licensing, reactor physics and thermo-hydraulic characteristics determination, accident analyses, fresh and spent fuel management, radioactive waste management and governmental categorization norms and rules. Acquaintance with calculational codes as the MCNP code for neutron transport and criticality calculations, WIMS-ANL code – for preparing of neutron cross sections for diffusion calculation, REBUS code for the fuel burn depth calculation, SCALE code system for spent fuel transport and storage devices safety assessment, PLTEMP/ANL code for calculation of thermo-hydraulic steady-state, and RELAP5 code – for transient operation, etc. is planned too. Preliminary acquaintance with the neutron activation analysis and BNCT is included in the educational programme. The classes will be held in the INRNE and the reconstructed IRT will be used for carrying out specific training exercises on the reactor: reactor start, manual and automatically control, control rod calibration, delayed neutron group measurements, sub-critical multiplication/shutdown margin measurements, excess reactivity and shutdown margin measurements; reactor-physics measurements of static and kinetic reactor parameters, reactor dosimetry, measurements of the spent fuel characteristics in the hot laboratory, radiological characterization survey - alfa, beta and gamma measurement techniques, contamination measurement, etc. IAEA-CN-156/U-45 The reconstruction of the IRT includes an arragement for a BNCT facility. Preliminary neutron transport calculations for BNCT channel regarding the geometry and material composition design have been carried out (Fig.1). Feasibility studies within the national network of the Medical University in Sofia, the National Centre of Radiobiology and Radiation Protection, the Institute of Experimental Pathology and Parasitology and Institute of Electronics of the Bulgarian Academy of Sciences, and the Faculty of Physics of Sofia University are carried out. Contacts with institutes, experienced in BNCT as EC JRC, Petten, the Netherlands, VTT, Finland and NRI- Rez, the Czech Republic, were established. Human, social and economical results due to the BNCT for patients from Balkan region are expected. Besides the financial support of the Bulgarian government the IRT has the IAEA support through the project BUL/4/014 “Refurbishment of the Research Reactor” and the support of the US Department of Energy in the frame of the RERTR program. The reconstructed IRT is a basis for keeping up specialists with researcher’s approach and skills who are able to give adequate responses to the challenges of complex modern technologies and the associated environmental problems. The reactor will be used for production of isotopes needed for medical therapy and diagnostics; it will be the neutron source in element activation analysis having a number of applications in industrial production, medicine, chemistry, criminology, etc. Nuclear energy has a strategic place within the structure of the country’s energy system. A new nuclear power plant Belene with two reactors of 1000 MeV will be built. The extremely high requirements regarding nuclear safety call for the availability of scientific and technical potential, and for an adequate culture of safe use of nuclear energy. The acquired scientific experience and qualification in reactor operation is a basis for participation of the country in the international cooperation within the European structures. In that aspect, the operation and use of the IRT brings economic benefits for the country. Figure 1. The BNCT beam tube model: 1. Vessel of Channel; 2. Filter 3. Lead Shielding; 4. Collimator; 5. Lead Shielding of Channel; 6. Concrete. IAEA-CN-156/U-38 Utilization of the LVR-15 Research Reactor at Řež M. Marek, J. Kysela, Nuclear Research Institute Řez, plc, 250 68 Řez, Czech Republic The LVR-15 research reactor commenced operation in 1957 as a multipurpose source of neutrons for basic research at horizontal channels and user-oriented research at mostly vertical loop channels and rigs as well. Since 1957 the reactor has undergone two reconstructions. During the last one in 1989 all the reactor components and systems were replaced, including the reactor vessel. The LVR-15 is a tank type reactor (Fig. 1) and currently uses IRT-2M fuel of 36 wt.% 235U enrichment manufactured by the NZCHK Company in Novosibirsk, Russia. The fuel features limit the output reactor power to 10 MW. 18 2 18 2 The thermal and fast neutron flux reach up to 1.5 x 10 n/m s and 2.5 x 10 n/m s, respectively. Due to the nature of the reactor use, the reactor working cycle is 21 days and the number of the cycles is 8- 10 per year. Core c onfig uratio n: Kxxxx Date 2.6.2003 A B C D E F G H Le gend Fuel IRT-2M 10    (Standard) HSSI - IAR Fuel IRT-2M 9    control rod-out Fuel IRT-2M 8 Be Be Be Be Be Be Be    control rod-in 7 Be Be     Be Be Be Be - full Be with 6 Be         channel φ 24 mm Water 5 Be       dis placement Air 4 Be    Be  dis placement 3 Be       Be  P ipe pos t 2 Be Be    Be CT 20 1  Be Be Be  Fig. 1 LVR-15 research reactor; example of LVR-15 core (axial cross section) Reactor Use For Material Research The advantage of the reactor arrangement results from flexible diameter of irradiation channels, good access to the upper parts of the channels, and the fact that the core can be refueled without outage of the irradiation facility. The main fields of the reactor utilization are neutron beam research including BNCT, fuel and material irradiation tests, and radioisotopes and silicon production. LVR-15 special reactor features in the field of material research can be summarized: − Core and irradiation channel size flexibility − Irradiation rigs for irradiation of small (ring, tensile) to large (1CT, 2CT) specimens − Five big loops with specialized mechanically loaded or heated irradiation channels − Water chemistry and dosimetry control ensuring the conditions in testing facilities to be as close as to the conditions in power plants. Other, more limited uses are in the areas of medicinal and industrial radioisotope manufacturing, production of radiation doped silicon and development of boron neutron capture therapy. The reactor is equipped with hot cells for post-irradiation sample manipulation, disassembling and assembling of core channels. The LVR-15 reactor specializes – due to its output and achievable neutron fluxes – in the study of combined effects of radiation and ambient media on materials. The reactor is equipped with experimental facilities such as loops and rigs, which permit exposure under conditions corresponding to those in power reactors. The generally utilized procedure is that the material is pre-irradiated in rigs IAEA-CN-156/U-38 and then is further exposed in loops enabling also the simulation of the thermal flux or physical stresses. Irradiation rigs permit the exposure starting from small samples (ring, tensile) up to very large samples (1CT, 2CT). Total five loops simulating either PWR or BWR conditions in various irradiation channels and other specialized facilities are in operation in the reactor. Fuel cycle Reactor has joined the Russian Research Reactor Fuel Return (RRRFR) initiative to be converted from HEU to LEU. The present fresh fuel stock consists of 86 fuel elements that are sufficient for the reactor operation till the end of 2010 year. The spent fuel FA’s are stored either in the reactor pools (113 pieces of IRT-2M, 36 wt.%, 80wt.%)or in the NRI interim storage (240 pieces of IRT-2M, 80 wt.%) or in dry barrels (206 pieces of EK- 10, 10 wt.%). At present the transportation of spent fuel to Russia is being prepared. Spent fuel will be shipped to Russia in 9/2007 in the VPVR casks (Fig.2) that were designed for all the Russian fuel types used in the LVR-15 similar research reactors of the Russian origin round the world. The cask is licensed in the Czech Republic and Russia as well. Conclusions The LVR-15 reactor is an important facility, which supports and contributes to research of nuclear materials and water chemistry. Experience that has been achieved operating the reactor during the last 50 years can be now transferred to the new irradiation facility designs including those which performs the research for Generation IV reactors, e.g. such as reactors cooled by high-temperature helium or water with supercritical parameters. Fig.2 SKODA VPVR/M cask References 1. J. Kysela, O. Erben, R. Všolák, M. Zmítko, In-pile irradiation research at NRI Řež for corrosion and material testing; NUCLEON 1995, Řež, Czech Republic 2. J. Kysela, M. Zmítko, R. Všolák, Loop capabilities at Řež for water chemistry and corrosion control of cladding and in-core components, Technical Committee Meeting on Water Chemistry and Corrosion Control of Cladding and Primary Circuit Components; 28 September - 2 October, 1998, Hluboká nad Vltavou, Czech Republic 3. J. Kysela, M. Zmítko, J. Šrank, R. Všolák, M. Ernestová, O. Erben, Overview of loop’s facilities for in-core materials and water chemistry testing; JAIF International Conference on Water Chemistry in Nuclear Power Plants, 13-16 October, 1998, Kashiwazaki, Japan 4. M. Marek, P. Novosad, P. Kotnour, M. Picek, SKODA VPVR/M Cask for Spent Nuclear Fuel from Research Reactors, TM on Specific Application of Research Reactor, IAEA, Viena, 29.5- 2.6.2006 IAEA-CN-156/U-49 Utilization of the VR-1 Training Reactor Karel Matejka, Hubomír Sklenka, Jan Rataj Czech Technical University, Prague, Czech Republic matej@troja.fjfi.cvut.cz The Paper presents basic information about utilization of the training reactor at Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, signed VR-1 and named VRABEC (which means “Sparrow”). The reactor has been used very efficiently especially for education of university students and specialists in favour of the Czech nuclear programme for more than 15 years. It is the only reactor of this type in the Czech Republic. Therefore, students from several Czech technical universities and also from universities in Central Europe participate on its use. The operator and main user of the VR-1 Reactor is the Czech Technical University in Prague. The VR-1 Reactor is well equipped for education and training not only by the experimental facility itself but also by carefully developed training methods. These are divided into several basic areas. Typical examples of them are as follows: ‚ Nuclear reactor control (start-up, operation, power changes, shut down), ‚ Reaching the critical state and critical parameters measurement, ‚ Dynamic experiments (periodical reactivity changes, delayed neutrons examination, examination of “bubbly boiling” influence on reactivity), ‚ Reactivity measurement and control rods calibration, ‚ Measurement of spatial distribution of neutron flux density, ‚ Neutron detection, ‚ Simulation of selected operational states of WWER type of power reactors, ‚ Neutron activation analysis. The education experiments can be combined into training courses attended by students according to their study specialization. The methods enable to choose an appropriate level of student participation in task completion and result evaluation. The training programme of university students at VR-1 Reactor covers overall information on nuclear safety, radiation protection, emergency preparedness, and physical protection principles. Every year, approximately 250 university students undergo training at VR-1 Reactor. Their stay at reactor site means an enormous benefit for their study process. The Czech Republic has a well-developed system of university nuclear education. The technical universities are cooperating in CENEN association (See Figure) In addition to the education of university students, the utilization of the VR-1 Reactor covers selected parts of training of specialists under the Czech nuclear programme and it also covers a supplemental research programme. The research programme is limited by a relatively small reactor power, however, a large variety of experiments and tests is still possible to be performed (from neutron detector calibration to verification of selected components developed for the transmutation technologies). The visits to the VR-1 Reactor are very popular. Part of the visit is a performance of reactor operation. Overall experience with VR-1 Reactor utilization is very good. There is a big call for reactor training. Detailed instructions and report forms for record and evaluation of measured values are available for every exercise. The forms usually exist in digital shape. The operation organization and reactor utilization is ruled by safety culture principles. IAEA-CN-156/U-49 Figure: A Scheme Diagram of Cooperation in CENEN (Czech Nuclear Education Network) Framework. [1] Matejka,K. – Sklenka,L.: Extensive Utilization of Training Reactor VR-1, In International Conference on Research Reactor Utilization, Safety, Decommissioning, Fuel and Waste Management, Wien: 2003, vol.1, p.88-89 [2] Rataj, J. – MatEjka, K.: Schedule of the basic critical experiment with the core configuration C1 at the training reactor VR-1, CTU, Prague, July 2005. [3] Sklenka, H. - MatEjka, K.: The First Critical Experiment with a LEU Russian Fuel IRT-4M at the Training Reactor VR-1, 2005 International Meeting on Reduced Enrichment for Research and Test Reactors, Boston, Massachusetts, 6-10 November 2005 [4] Rataj, J. - MatEjka, K.: Use of Training Reactor VR-1 for Specialists Preparation In st Research Reactors in the 21 Century (CD-ROM). Moskva: Research and Development Institute for Power Engineering, 2006 IAEA-CN-156/U-28 The 250 kW FiR 1 TRIGA Research Reactor - nal Role in Boron Nuetron Capture Therapy (BNCT) and Regional Role in Isotope Production, Education and Training Iiro Auterinen, Seppo Salmenhaara VTT Technical Research Centre of Finland, Espoo, Finland E-mail address of main author: Iiro.Auterinen@v tt.fi The Finnish Triga reactor, FiR 1, has beine no peration since March 1962. The reactor has been delivered by General Atomics, USA. The reactor belonged first to the Department of Technical Physics at the Helsinki University of Technology. The activities of the reactor were defined as training in nuclear technology, recshe aarnd production of radioactive isotopes. In 1972 the reactor was transferred under the admraintiiosnt of the Technical Research Centre of Finland (VTT). From its early days the reactor created vteilers aresearch to support both the national nuclear program as well as generally the industrnyd ahealth care sector. The volume of neutron activation analysis was impressive in the 7a0n’sd 80’s when the reactor was operated close to daily only for activation analysis. Now other analysis methods have nearly totally replaced neutron activation analysis. In isotope prodounc tai small research reactor is competitive only in producing short lived isotopes for local markets. Boron neutron capture therapy (BNCT) treantmtse dominate the current utilization of the reactor: three days per week are for BNCT puerps oasnd the rest for other purposes such as isotope production, neutron activation analysis and education. In the 1990’s a BNCT treatment facility was bu ailtd the FiR 1 reactor. A special new neutron moderator material Fluental™ (Al+A3l+FLi) developed at VTT ensures the superior quality of the neutron beam. Also the treatment reonnvmi ent is of world top quality after a major renovation of the whole reactor building in 1997. The FiR 1 reactor has proven to be a reliable neutron source for the BNCT treatme ntos ;patient irradiations have been cancelled because of a malfunction of the reactor. Oovneer hundred patient irradiations have been performed at FiR 1 since May 1999, when the lsicee fnor patient treatment was granted to the responsible BNCT treatment organization, Boneca Corporation. Currently three clinical trial protocols for tumours in the brain as well ians the head and neck region are recruiting patients. FiR 1 reactor has made already now a major contribution in the research and development of BNCT. Significant breakthroughs have been avcehdie in the application of BNCT for cancer treatment. FiR 1 has an important internatli ornoale in the development of boron neutron capture therapy for cancer as it is one of fethwe facilities in the world providing this kind of treatments. The successes in the BNCT development have now created a demand for these treatments, although they are given on an experimental basis. Due to the BNCT project FiR 1 has become an important research and education unit for medical physics. Since the early 1990’s sevgeradl uate and postgraduate students from the IAEA-CN-156/U-28 medical physics program of the University ofl sHineki have been working at the FiR 1 BNCT facility. In research projects funded by thFein nish Academy, the Finnish Center for Technology Development (Tekes) and the EU the dosimetry, radiation transport modelling, treatment planning, prompt-gamma imaging aonthde r medical physics aspects of the BNCT have been studied and developed. Over ten eamciacd theses and dissertations have been produced in these projects, along with over hun dsrceiedntific publications. The research has been performed in close international collaboration with European, American as well as Japanese researchers. Also at Helsinki eUrnsiivty of technology masters theses works have been made in connection to the BNCT rese artc FhiR 1. The students aiming at the hospital physicist exam credit up to one year of rereqdu ihands-on experience when working at the FiR 1 BNCT facility. In addition to BNCT FiR 1 continues to suptp roersearch in radiopharmaceutical applications. In the 1980´s FiR 1 had a major role ine thFinnish radiopharmaceutical research and development. The spin-off company established at the reactor at that time, MAP Medical Technologies, has been successful but relies onno wo ther sources for its radioisotopes, the accelerator at the Jyväskylä University Finin land and traditional international isotope producers. Now FiR 1 is utilised by researche rtsh ea tUniversity of Helsinki for studying the use of samarium as a tracer in pharmaceutical formulation development. Education and training play also a role at F1i Rin the form of university courses and training of nuclear industry personnel. Helsinki Univietyr sof Technology has a right to use the reactor for its purposes. Yearly there are at letawsot courses for technical physics and energy technology students in reactor and neutron ipchsy sthat utilize the reactor and the BNCT facility. Reactor physics demonstrations aarels o organized for the students of the Lappeenranta Technical University. FiR 1 is utilized also in the continuing ecdaution and training of the personnel at nuclear power companies, both in Finland and in eSdwen, and other organisations connected to nuclear power. These are typically one day nisnitvee courses with hands on exercises, or demonstrations and excursions in connection to longer lecture courses. The main part of the workintgim e has been reserved for the work at the BNCT irradiation facility. Still we have also for 20% - 40% of thtime e the possibility to irradiate samples either to produce some short-lived isotopes or ton deou tron activation analysis. Main isotopes for tracer studies produced in the reactor 2a4rNea , 82Br and 140La. The incomes from the production of isotopes are 15 % to 20 % of the turnover. The reactor is operated by four reactor operaatonrds five shift supervisors. All of them are part time operators or supervisors in addition to their work as research scientists or research engineers. This amount of operators and supersv iesnosrures that the reactor is easy to keep in operation during normal working hours and also during exceptional hours. IAEA-CN-156/U-6 Chemical character isation of ear lyn efi-ware pottery by neutron activation analysis: analytical and statisticaapl proaches to production and trade Vassilis Kilikoglou, Anno Hein and Peter M. Day Institute of Material Science, N.C.S .‘RD.emokritos;, Aghia Paraskevi 153 10, Greece kilikog@ims.demokritos.gr Neutron activation analysis (NAA) is a well-ebsltiashed method for the chemical analysis of ancient pottery, particaurlly in the pursuit of provenancsetu dies. The significance of NAA in such research is its eminent suitability ine tmheasurement of those trace elements which are characteristic in geochemical terms of clsaoyu rces used in ceramic production. Since the chemical variation between different clay sousr cgeenerally can be considered to be larger than natural variation within a particulara yc l source, pottery prodctuion sites, or even workshops that use a specific clay pasctea,n be identified by a so-called ‘chemical fingerprint’. At the N.C.S.R. “Demokritos”, NAA has been euds for more than 30y ears for the routine analysis of ceramics. Approximately 130 megx,a ctly weighed, of the powdered and dried sample are placed in a polyethylene vial, whic th eisn heat-sealed. The vials are irradiated for 45 min in the “Demokritos” swimming pool retaocr, at a thermal neutron flux of about 6©1013n©cm-2s-1, in batches of 10. Usually two batchaerse irradiated together, each of them including eight samples with unknown comiptiosn and two standard reference materials (SRM). Seven days after irradiation the vials are placed in the sample changer and counted for one hour with a Gei- detector covering the energy ranogfe 8 0-1600 keV. Thec oncentrations of As, Ca, K, La, Lu, Na, Sb, Sm, U and Ybe adretermined by this count. Three weeks after irradiation, samples and standards are counted again for two hours each and the concentrations of Ba, Ce, Co, Cr, Cs, Eu, HFfe, ,N d, Ni, Rb, Sc, Ta, Tb, Th, , Zn and Zr are determined. The principle standard employe dS OisIL-7, issued by the International Atomic Energy Agency. Calibration teshtsa ve been carried out agati nSsL-1 (IAEA), SRM 679 (brick clay-NIST), SRM 2711 (Montana soil- NISTa) nd in-house standards used in other laboratories. The applications and potentials of NAA in aarcehological ceramic studies are demonstrated by two case studies recently completed at N.C.S.R. “Demokritos”: those of Late Neolithic ‘black-on-red’ potterya nd Early Bronze Age ‘sauceboatbso’,t h from the prehistoric Aegean. We highlight these case studies as they invothlvee d etection of provenance and trade of fine ware pottery, which is difficult o address with mineralogicatel chniques. Fine pottery is relatively uncommon in these Neolithic andr lEy aBronze Age assemblages, and finding their origin is a matter of priority. ‘Black-on-red’ painted pottery iosne of the most characteris tciceramic types, which appears during the final phases of the Late NeolithicN ionr thern Greece. The largest quantities of this kind of pottery have been found in Eastern Macedonia and Thrace. However, significant amounts have also been excavated in Cle natnrad West Macedonia, but also, in lesser amounts, in the upper Strymon Valley, in Sho-uwtest Bulgaria, and in Thessaly. The importance of this ware has been appreciate tdh eb yresearchers of the Neolithic Period in the Aegean, using it as a diagnostic element foer rtehlative dating, as wll eas for the cultural IAEA-CN-156/U-6 attribution of different ceramc i assemblages. The purpose ooufr study was to search for production patterns for this ceramic group. More specifically , extoamine questions concerning the degree of standardisatio np rionduction, the location of production centres, indications of the scea lof production, along with the diffenrteiation of ceramic recipes within the group, which could reflect variioant (geographic or other) ipno ttery traditions. Altogether 198 samples were selected for NAA in an attempt to answer tehseti oqnus produced by the stylistic study of this material. Statistical evaluation of the chemical compositions of the ‘black-on-red’ pottery samples analysed froEmas tern Macedonia result eind the identification of four distinct groups, each of which represents a distinct production area. The pottery production within each of the four geogracpahl i areas had commonc htenological and, in a broader sense, stylistic chararcistteics. It was found that thper oducts of individual production centres were exchanged beyond their regio np rofduction and that the products of one particular centre were found as widely distributed as the island of Thasos and the plain of Thessaly. This demonstrates not only that eth easrly fine wares we rperoduced in separate centres which shared technological characteris btiucts ,also that they aicvtely participated in extensive exchange systems. The second case study concerns specialised shapes from the Early Bronze Age of the Aegean, known as ‘sauceboats’. These distinctive ve fsosreml s appear during the middle of the third millennium BC, as a crucial component oifn dkring sets which accompanied the introduction in the Aegean of true, flat-based ‘table-war e Ds’e. corated in three broad ways: dark-on-light painted, yellow-blue mottled and a glossya cbkl slip known as ‘urrfni is’, the different sauceboat types have beeni bautttred to two main areas opfr oduction: the Helladic mainland (in the case of the urfirnis and yellow/blue mottled) and the Cycladic islands (in the case of the dark-on-light painted). A large number soafuceboat samples were taken from various sites on the mainland, throughout the Cycladest,h oen A sia Minor coast and from the island of Crete. NAA was able to distinguish a nbuemr of chemical groups which represented the widely distributed products of specific cterens of production. The picture was far more complex than anticipated, with the production uorfirnis sauceboats on both the Cycladic island of Melos and on Crete. While the prcotdsu of a West Cretan centre have a limited distribution, those manufactur eodn Melos were found on the iAas Minor coast, as well as Northern Crete and throughout the islands. Ye-lbloluwe mottled sauceboats from the area of Attika on the mainland and the island of Kweae re also found to share their centre of production. In this case, NAA demonstrates siutsit ability to provided etailed information about the origin of important fine warewsh, ose mineralogical studhya d proved problematic. We use these examples to demonstrate ftfheec tieveness of NAA in teh study of fineware pottery groups, making an important croibnuttion to ceramic studies through their characterisation according to ctrea element concentrations, which prove far more distinct than petrographic or mineralogical. Itnh is effort, compared to othechr emical analytical techniques applied to ceramics, NAA still porvides advantages in terms soafm ple preparation, precision and accuracy. IAEA-CN-156/U-56 FEASIBILITY STUDY OF 125I BRACHYTHERAPY IN INDONESIA1 Tri Murni Soentono 1) 1) Centre for Technology of Radiation, Safety and Metrology National Nuclear Energy Agency of Indonesia (BATAN) Cinere, Pasar Jumat, Jakarta Selatan, Indonesia Telephone no.: 62 816787769 Telefax no.: 62 21 7803805 Email address: murni@batan.go.id ABSTRACT Cancer is a term for diseases in which abnormal cells divide without control (1). Many cancer treatments such as chemotherapy or immunotherapy, as well as radiotherapy, are intended to kill tumor cells. Brachytherapy can be performed by using tiny titanium cylinders which contain a small amount of radioactive material such as radioisotopes I125. These seeds are used as implants for prostate cancer or breast cancer. This Study is to develop and produce innovative radioisotopic products as brachytherapy seeds. Research reactor in Serpong Indonesia has been producing radioisotopes I125 from the mid year 2005 through 4 time irradiation. From these radioisotopes the research would continue to produce I125 brachystherapy. This study is focused on the technology producing of I125 brachystherapy production, for Indonesian people who has a cancer. BACKGROUND 1. Research reactor in Serpong has been producing radioisotopes I125 2. Brachytherapy I125 has low risk of radiation exposure for the pasien, health care workers and those around them (1) CANCER CASES IN INDONESIA The statistic from Health Department of Indonesian Government showed that 13,2 million peoples or 6 % of Indonesian population 2005 (around 241.973.879) has cancer desease. 65 % is too late to go to the hospital for diagnosis and treatment. A number of them frightened of operations. The cancer desease ranks fifth in the list of mortality after Hearth attack, Stroke, Lung and Diarrhea. The relative frequency of cancer desease in Indonesia is 23,66 % in age group of less than 40 years old and 22,7% in that of more than 45 years olds. (2) TECHNOLOGY OF BRACHYTHERAPY I125 PRODUCTION Research Reactor in serpong has been producing radioisotopes I125 from the mid year 2005 through 4 time irradiation of 124 Xe gas. From these radioisotopes the research would continue to produce I125 brachystherapy. IAEA-CN-156/U-56 TABEL 1: The Activity of 125I Product in 2005 No dates Activities 1 13 June 2005 9.5 Ci 2 5 July 2005 9.8 Ci 3 25 July 2005 17.2 Ci 4 12 December 2005 9.4 Ci Brachytherapy is internal radiation therapy using an implant of radioactive material such as 125I, 103Pd and 90Y placed directly into or near the tumor. The size of this brachytherapy is 4.5 mm long and 0.8 mm in diameter. The benefit of placing sources near or inside the tumor is that they are able to deliver a cell-killing radiation dose to the tumor while sparing the healthy surrounding tissue. In terms of efficacy, cost and maintaining the patient's quality of life. Technology of I125 brachytherapy production consists of two methods. First, Ion implantation to manufacture the radioisotopes I125 provides the seed, this method is more difficult than the second because this seeds are tiny and the dispensing of radioisotopes I125 to the seeds is buried has to be done the hotcell. In the second methods pure isotopes of non-radioactive Xenon-124 gas beneath the surface of the seed core. This method provides safety to the operator and easy for dispensing but leaking analysis of gas before irradiation must be done. Figure 1: brachystherapy seeds (3) CONCLUSION The research for producing I125 brachytherapy needs to bee continued to decrease people fear for operation. The I125 brachytherapy enables patient to leave hospital on the same day as that of admission. REFERENCE 1. INTERNATIONAL ATOMIC ENERGY AGENCY” Program of Action for Cancer Therapy”, 2007 2. Siswono Kamis, http://www.suaramerdeka.com 9 Maret, 2006 3. Implant Sciences Corporation “ I-PlantTM 125 Brachytherapy Seeds“, 2007 IAEA-CN-156/U-17 Characterization of Airb orne Particulate Matter at Urban and Rural Area in Bandung and Lembang Indonesia using Instrumental Neutron Activation Analysis Muhayatun Santoso, Achmad Hidayat, Diah Dwiana Lestiani Center of Nuclear Technology for Materials and Radiometry National Nuclear Energy Agency, Indonesia E-mail address of main authohr:a yat@bdg.centrin.net. id Samples of fine and coarse friaocnts of airborne particulate mttear were collected twice a week for 24 hours in Center of Nuclear Technology for Materials and Radiometry Bandung (urban) and Meteorological and Geophysical Agency tisotna Lembang (rural) from January 2004 to December 2004. The samples werelle coted using a Gent stackeidlt efr sampler in two fractions of < 2.5 µm fine and 2.5 – 1µ0m coarse sizes. The samples and synthetic standards were analyzed for elemental concentrations by inmsteruntal neutron activaotni analysis (INAA) at Bandung TRIGA 2000 reactor. Black carbon concetinotnras were determined using Smoke Stain Reflectometer. The synthetic standards werep aprred by pipetting 100u L of the solution containing one or more elements into a polyeetnheyl vial. Irradiations wre carried out under two experimental conditions. Short idrriation of 5 min at the pneuma ttircansfer tube with thermal neutron flux of approximately 1130 n cm-2 s-1were used to determine the short half-life elements, cooled for approximately 60 s, and counted for 300s (live time) in the Radiometry Analysis Technique Laboratory. For determining the muemd-ilong half-life elements, the samples were then irradiated again in the fixed idrriation system with a neutron flux of 1130 n cm-2 s-1 for 48 – 60 hours, cooled for 1 - 2 days and counted3 f0o0r0 s or more dependinogn their activities by a gamma spectrometer with coaxial detector codu ptole Integrated Signal Processor and System MCA Card both from APTEC. The identificioant of radioisotopes was carried out by characterizing its half-lifea nd its gamma-ray energies. The NIST’s standard reference tmeraial SRM 1648 airborne parutilcate matter were analyzed in the same experimental conditions used in thmep slea analysis as method validation to evaluate the precision and accuracy of the results. Trehseu lt of SRM 1648 analyses mostly have good agreement with the value quoted in the NIS Tce’srtificate (see TableI) . The radioisotopes measured in this study were related to 22 elem aesn tfsollows : C, Na, Al, V, Mn, Br, Cl, I, Cr, Fe, Zn, Sc, Sb, Co, La, Mg, Sm, K, Ca, Ti, As, and Cses.e (FIG.1). Most of the element concentrations in Bandung (urbaanr)e higher than in Lembang r(arul). These data of elements can be used to characterization of pollutsaonut rces by correlating the relationship among these elements, such as Al, Ca and Ti for sobilla, ck carbon and K for wood burning source. The correlation between those crustal elements Aal inasgt Ti and Ca of coarse fractions both at Bandung and Lembang show how well these elemceonrtrse lated to each other and demonstrates that they are basically lraeted to the one source, soil. The bkl accarbon versus Br of fine faction at Bandung plot also show good correlation, that tahreey related to motor vehicles source. This result associated with the number of veehsic lregistered in Bandung approximately increased 30% compared to the last two years. IAEA-CN-156/U-17 Table I. Elemental concentration obtained for NIST SRM 1648 Element Unit Thisw ork Certificatev alue (NIST value)c Xa ± SDb Al % 3.43 ± 0.12 3.42 ± 0.11 As mg/kg 124 ± 8.5 115 ± 10 Br mg/kg 445 ±1 4 500 Cl % 0.45 ±0 .03 0.45 Co mg/kg 18 ±2 .4 18 Cr mg/kg 378 ± 84 403 ± 12 Cs mg/kg 3.67 ±0 .13 3 Fe % 3.98 ± 0.38 3.91 ± 0.10 I mg/kg 19 ±2 20 La mg/kg 37 ±6 42 Mg % 1.3 ± 0.3 0.8 Mn mg/kg 788 ± 27 786 ± 17 Na % 0.403 ± 0.031 0.425 ± 0.002 Sb mg/kg 45.8 ±2 .2 45 Sc mg/kg 6.74 ±0 .28 7 Sm mg/kg 4.5 ±0 .85 4.4 Ti % 0.41 ± 0.05 0.40 V mg/kg 126 ± 7 127 ± 7 Zn % 0.418 ± 0.024 0.476 ± 0.014 a mean value b one standard deviation c NIST does not provide uncertainties for uncertified elem ents 10000 Bandung 1000 Lembang 100 10 1 0.1 0.01 Al As Br C Ca Cl Co Cr Cs Fe I K La Mg Mn Na Sb Sc Sm Ti V Zn Element FIG.1. Annual average of elemental concentration of APM in Bandung and Lembang INAA as one of nuclear techniques utilizings erearch reactor provides a very good, simultaneous, multi-elemental analysis methodfosr airborne particulate studi.e Isn particular, INAA is known to be a reliable analysis technique with lowte dcetion limits and effective for a large number of samples. The ability to provide elemental concentration information on all of the most significant elements enables statistical techniques to be applied in the data and determine the source contribution of the pollutant. Analysis the baoirrne particulate samples collected in Bandung and Lembang were reported to demonstrate the adavgaen ot f INAA method, espeaclily to get a better understanding about the condition of atmphoesre in Indonesia due to air pollution. Concentration (ng/m3) IAEA-CN-156/U-14 Character ization of a Neutron Collimator for Neutron Radiography Applications. 1) 1) M.Palomba , R.Rosa 1) ENEA – Triga RC-1, Rome, Italy E-mail address of main auth:o rpalomba@casaccia.enea .it TRIGA Mark II reactor of ENEA’s Casaccia research Center (in Italy named RC-1) reached first criticality in 1960 and, after an upgrade completed in 1967, its power was increased till 1 MW. At this power level it is still running, mainly for short mean life time radioisotopes production (for medical purposes) and neutron radiography. In the year 2003 started a program for a new neutron radiography/tomography facility. This facility will utilize the “Tangential Channel” of the Triga RC-1 reactor. After some preliminary studies and the design, a suitable collimator was installed in the reactor channel. This paper will show the technical description of the facility, some theoretical studies on the collimator design and the experimental characterization of the collimator. Several configurations of the internal filters were arranged and analyzed experimentally. For the collimator characterization, the Kobayashi method [1] was used in order to evaluate the L/D ratio, the geometrical resolution and the divergence angle. A map of the spatial distribution of the neutron flux, け-ratio and Cadmium-ratio was also analyzed and optimized. A neutron collimator is a device that focalize the neutron beam towards an imaging plane “modifying” the neutron emission from the neutron source into a parallel or slightly divergent beam. Usually, a neutron beam, could be classified in three different categories: ‚ Radial : when the collimator starts radially from the neutron source; ‚ Tangential : when the collimator lies tangentially from the neutron source; ‚ Beam Guide : when the collimator, starting from the neutron source, has a complex shape that “drives” the neutron beam. The facility installed in the Triga RC-1 Reactor is a Slightly Divergent Tangential Collimator (see Fig.1). Fig. 1. Triga RC-1 Tangential Channel Collimator The main feature of this particular type of device is that the け emission from the beam is less than in the case of a similar collimator installed radially from the core of the reactor. The main design parameters of such a type of collimator are: filters (け/n), neutron entry hole diameter, absorbing walls and filling gas. The main features of the device (depending from the optimization of the design parameters) are: Neutron Flux, Cadmium Ratio, け/n Ratio, Effective Beam Diameter and L/D Ratio, Background Components. IAEA-CN-156/U-14 The Neutron Beam/Collimator characterization was performed by a campaign of け/n measures: Space/Energy distribution measurements by Activation Foils, Cadmium Ratio Measurements by Activation Foils and Cadmium Capsules, け measurements by TLD devices and L/D Ratio measurements by Special Equipment and Imaging Plates (Kobayashi Method). The spatial distribution of the neutron flux was performed using some gold foils and the optimization of the channel performances was done by analyzing three different conditions: with or without a Bismute filter or with a Graphite filter. After the optimization of the Neutron Flux, the collimator was characterized using the Kobayashi Method in order to define the L/D Ratio and the Effective Beam Diameter. In Fig 2 is shown the special device (2 aluminum plates with a Cadmium plate inside) used to perform the measure. Fig. 2. Kobayashi Method Sample Device The image of the Sample Device, projected on an Imaging Plate placed at a proper distance from it, was analyzed and the calculations from which the L/D Parameter is derived, are reported and discussed. The final design optimization, the future improvements of the collimator and the design of a new concept of a Neutron/け shutter are reported and discussed too. [1] Kobayashi H., Wakao H., “Accurate measurement of L, D, and L/D for divergent collimators”, ASTM E 803-81 1981. IAEA-CN-156/U-33 Optimizing Conditions Suited for Stress Determinations in Q-Space Focusing Configurations I.Ionita Institute for Nuclear Research Pitesti, Romania E-mail address of main author: ionionita@lycos.com During the last decade a new concept of high-resolution focusing configuration has been developed, using Q-space focusing effects, which proved to be an alternative to the existing conventional configuration. If such a focusing configuration is to be used for stress determinations, special problems arise. As, for a given scattering angle, thin peaks are obtained only for a certain position of the sample, severe limitations appears in choosing the scattering angle or in amount of information possible to be obtained. A convenient solution is to limit the dimensions of the sample zone for which strain determinations are realised. That means not only the use of corresponding diaphragms but also getting real-space focusing at sample position. In conclusion if strain determinations are to be realised using Q-space focusing configuration both real space focusing at sample position and the phase-space focusing getting thin diffraction peaks must be obtained. The corresponding focusing conditions are deduced in this paper for three neutron diffractometer configurations: crystal diffractometer, 2 crystals diffractometer and time-of-flight diffractoemeter using steady-state neutron source. For this extended abstract only 1 crystal diffractoemeter configuration is considered. 1. Crystal neutron diffractometer 1.1 1 crystal monochromator unit The considered configuration is given in fig.1. The Bragg constraints are given by: 2l γ 0 + γ1 = m sign (θm + χm ) (1) Rm The configuration geometry gives: L0γ 0 = lm sin (θm + χm ) − l0 L1γ1 = ls cos (θ s + χ s ) + lm sin (θm − χm ) (2) L2γ 2 = ld − ls cos (θ s − χ s ) where θm and χm are the monochromator Bragg angle and the cutting angle respectively, li (i= 0, m, s, d) are the widths in the horizontal plane of the source, monochromator sample and detector respectively , Rm is the radius of curvature of the monochromator and Li (i=0, 1, 2) are the distances between source and monochromator, between monochromator and sample and between sample and detector respectively; γi are angular variables in the horizontal plane, representing deviations from the most probable values and sign(α) = abs(α)/α. IAEA-CN-156/U-33 Fig.1 Experimental setup for an one crystal diffractometer If lseff is the effective sample width (the part of the sample irradiated by the neutrons) in the direction normal to the beam incident on the sample, using (1) and (2) one obtains: 2L L  L lseff = l 1 m  sign (θm + χm ) − sin (θm − χm ) − 1 sin (θm + χ 1 m ) + l0 (3)  Rm L0  L0 The minimum value of lseff, i.e. the minimum width of the beam at the sample position, is obtained cancelling the lm coefficient. It is obtained: 2L1sign (θm + χm ) Rm = (4) sin (θ Lm − χ 1m ) + sin (θm + χL m )0 The relation (4) is the real space focusing condition giving the monochromator radius of curvature for which the contribution to the beam width at the sample position given by the monochromator length is compensated. If we want to have a thin diffraction line for the 2θ value of the scattering angle, the “Q space” focusing condition should be fulfilled, [1], [2], [3] also: a 2L1sign (θm + χm ) tanθ R = with a = − s m (5) 2a −1 sin (θm − χm ) tanθm The value given by the (4) condition depends only on θm and χm while that given by (5) depends on θm , χm and θs. In order to have the best conditions for the stress measurements, both (4) and (5) have to be fulfilled. These two conditions are satisfied, for a given value of θm, only for a certain value of the scattering angle given by: tan (θ ) tanθ ms = (6) L sin (θm + χm ) 1− 1 L0 sin (θm − χm ) This work has been partially supported by IAEA Vienna under the contract ROM/13579. IAEA-CN-156/U-34 SANS FACILITY AT THE PITESTI 14MW TRIGA REACTOR 1) I.Ionita , B.Grabcev 2) 2) 5), S.Todireanu , G.Popescu , F.Constantin 3) , M.Mincu1), E.Anghel1), V.Shvetsov4), A.Datcu 1) 1) Institute for Nuclear Research – Pitesti, Romania 2) National Institute of Materials Physics (NIMP) Bucharest, Romania 3) National Institute of Physics and Nuclear Engineering Bhcharest, Romania 4) Joint Institute for Nuclear Research Dubna, Russian Federation 5) National College Al. Odobescu Pitesti, Romania E-mail address of main author: ionionita@lycos.com At the present time, an important not yet fully exploited potentiality is represented by the SANS instruments existent at lower power reactors and reactors in developing countries even if they are, generally, endowed with a simpler equipment and are characterized by the lack of infrastructure to maintain and repair high technology accessories. The application of SANS at lower power reactors and in developing countries nevertheless is possible in well selected topics where only a restricted Q range is required, when scattering power is expected to be sufficiently high or when the sample size can be increased at the expense of resolution. The need for the installation of a new SANS facility at the Triga Reactor of the Institute of Nuclear Researches in Pitesti, Romania become actual especially after the shutting down of the VVRS Reactor from Bucharest. Experimental SANS configurations suited for a steady state reactor Any SANS configuration is formed by a monochromator unit, an analyser unit and a detecting system. In the most used instruments the analyser unit is a simple one given by the detecting system itself. The monochromator unit can by formed by one or two single crystal. For the two crystals system the spatial extension around the beam of the experimental setting is decreased significantly while the luminosity is lowered because of the second refection. A quite common configuration has a crystal monocromator and a crystal analyser the sample being positioned between monochromaror and analyser. The analyser crystal is rotated around an axis normal to the beam direction and a “rocking” curve I(θ) is recorded, representing the angular distribution os the scattered neutrons. For the case of a mechanical monochromator the monochromatic beam has the direction of the beam channel. The selection of the optimum configuration; The selection criteria definition For SANS the recorded curve is I(Q), Q = k i-k f. As the dimensions of the components are finite, a given number of counts correspond not to a certain value Q, but to Q+∆Q. The value of ∆θ is given by the overall collimation while ∆λ is given only by the monochromator. We shall choose as selection criterion the value of the monochromatic beam intensity at sample corresponding to a given value of ∆λ/λ. Another criterion is the instrument extension around the beam channel; the involved sum of money to realize the instrument is important as well. For the case of crystal monocromator unit, that with one crystal is the most desirable though the instrument is somewhat extended from the beam channel axis. The SANS instrument with double crystals monochromator is significantly less extended while that with crystal monocromator and analyser allows for the very low Q values determinations. For both of them, because of the two reflections, the luminosity is rather poor. These instruments could be preferred for the case of large flux reactors as neutron source. The SANS configuration with mechanical monochromator has significantly increased luminosity, in comparison with those having crystal monocromator units, while the spatial extension of the instrument is quite reasonable. Tacking account that TRIGA reactor existing IAEA-CN-156/U-34 at INR Pitesti is a medium flux reactor and that the available dimension for the SANS instrument is severely limited by the dimensions of the room where the instrument has to be installed, this experimental configuration has been chosen as the most suited for the situation existing in our institute. For usual collimation values of about 30 minutes, and for an inclination angle of the monochromator axis of about 2-3 degrees, ∆λ/λ is about 20-30%, i.e. quite reasonable value. sample width may be fixed between 10 mm and 20 mm. The o minimum value of the scattering vector is Qmin = 0.005 −1 A while the maximal value is 0 −1 Qmax = 0.5 A . The relative error is ∆Q / Qmin = 0.5 . In the case of our SANS instrument a 0 0 monochromatic neutron beam with 1.5 A ≤ λ ≤ 5 A is produced by a mechanical velocity selector with helical slots. The distance between sample and detectors plane is (5.2 m ). Mechanical Detecting System Shielding SSR Radial Monochromator Channel Fig.1 The instrument layout The SANS instrument (fig. 1) has the following components: The mechanical monochromator, shielding, the Bi filter, the sample table and holder, the detecting system 3 formed by two rows of 40 He detectors each, 2 flux monitors, the beam stop. Paraffine blocks and recipients with water form the shielding. The flux monitors are positioned one before the sample (in front of the monocromator window) the other in front of the beam stop. The sample can be rotated using a step by step motor. A cadmium slit system actioned by a step by step motor allows for the determination of the monochromatic beam center at the beam stop position. IAEA-CN-156/U-19 RIAR Capabilities in Support of the Innovative Nuclear Technologies A.V. Bychkov, V.A. Tzykanov, A.V. Klinov, M.V. Kormilitsyn Federal State Unitary Enterprise “State Scientific Center – Research Institute of Atomic Reactors”, Dimitrovgrad-10, 433510, Russia, ぎ-mail: bav@niiar.ru The RIAR installations capabilities for provide for solving problems of nuclear power engineering development, safety ensuring, environmental acceptability, efficiency and effectiveness are described in the report. RIAR reactor complex including high-flux research reactor SM-3, research fast reactor BOR-60, research material testing reactor MIR and water pool-cooled research reactor RBT (see table 1). Table 1 Research reactors operated in RIAR Power, Reactor Name Type Refurbishment MW start-up SM-3 Vessel-type 100 (50) 1961 1974, 1992 MIR-M1 Channel-type 100 1966 1975 BOR-60 fast (liquid-metal) 60 1969 RBT-6 Pool-type 6 1975 RBど-10/1 Pool-type 10 1983 RBど-10/2 Pool-type 10 1984 All RIAR research nuclear reactors provided with experimental devices and in-pile examination techniques to solve pressing tasks of nuclear power engineering and reactor material science with regard to the reactors of different application and examinations in support of new technologies and engineering solutions for modern and advanced reactors of a new generation. RIAR have got the unique material science complex also. The Complex based on hot cells and it is the largest material science complex in Russia: in its 3 buildings there are about 50 shielded chambers and over 100 heavy and medium- weight shielded boxes. Fuel element and FA non-destructive examinations, some fuel and absorbing element fragments as well as other specimens with emergency parameters are studied in one building. Material science investigations on samples, fuel and absorbing element fragments cut out from the standard or pilot products with taking into consideration non-destructive examination results are carried out in another building. In the third building there are cells and shielding box chains equipped to manufacture the experimental and pilot fuel elements with different fuel compositions. In the report the examples of RIAR international collaboration and proposals on development of international collaboration on the basis of the research nuclear reactors are described. IAEA-CN-156/U-21 SM Reactor After Core Modernization A.V.Klinov, ん.P.Malkov, A.L. Petelin, M.N.Svyatkin, V.A.Starkov, V.A.Tsykanov International Conference on Research Reactors: ”Safety Management and Efficient Usage” Sidney, Australia, November 5-9, 2007 The modernization purpose was to make possible the irradiation testing of structure materials for power water fission and fusion reactors at a damage rate of ‡ 20 dpa, helium accumulation rate - > 1000 appm per year and fast neutron dose - 100 dpa and more. At the same time, conditions must be created for simultaneous irradiation of representative sample arrays in water environment with necessary chemical parameters at a temperature regulated in 0 the range 100-300 C and pressure up to 17 MPa by using irradiation rigs well-instrumented with devices for on-line parameters monitoring, and temperature and neutron spectrum control. Actually, it was necessary, in addition to the available irradiation positions, to provide places for loop and capsule channels of large diameters, which replaced some part of fuel without significant changes of the reactor design and operational procedures. In the process of work on the modernization project a new fuel assembly design was developed that had an experimental channel ̋24.5mm in the fuel rods array, and the possibility was substantiated to place up to four such assemblies in the reactor core. Instead of two positions for fuel assemblies, two positions for loop or capsule channels ̋ 69mm were provided. Reactivity losses were compensated by the replacement of stainless steel wrappers with those of 235 zirconium alloys and by increase of U content in the fuel rods by 20%. This required mastering the fabrication process of the modified fuel rod and experimental verification of its serviceability 2 at a thermal flux density up to 15 MW/m on its surface, as well as of the serviceability of three types of fuel assemblies with such fuel rods, including the assemblies with the experimental channels. In the course of post irradiation examinations the data on fuel rod swelling as a function of burn-ups were obtained and deformation and failure temperatures at different burnups were determined. The fuel rods showed to be characterized by the sufficient serviceability under normal operation conditions and to have the necessary temperature margin before initiation of deformation and failure under overheating. In January 2005, during the routine fuel reloading without reactor shutdown the standard 235 fuel assemblies were replaced by the fuel assemblies with the increased U content. The reactivity characteristics of the reactor core were found to be slightly changed that is explained by the great effect of fuel self-screening. The fuel cycle parameters were noticeably ameliorated. Fuel utilization performance before and after modernization Parameter Value of 2004 Value of 2005 Energy production, MW-d 21900 21600 Number of spent fuel assemblies 101 77 Fuel assembly consumption per 1000 MW-d 4.61 3.56 235 U mass in the spent fuel assemblies,kg 92.6 82.9 Characteristics of a new configuration of the core ensure the requirements for operation and experiment conditions. IAEA-CN-156/U-21 Comparative core characteristics before and after modernization (experiment) Parameter Value before Value after modernization modernization Reactivity margin, def 10.6‒1.0 10.7‒1.0 Control rod efficiency, def 12.7‒0.7 13.2‒0.8 Temperature coefficient of reactivity under 0 -2 -2 nominal conditions, def/C -(1.8‒1.0)©10 -(2.0‒0.15)©10 Rate of reactivity loss as a function of fuel -3 -3 burn-up, def/MW-d (6.1‒0.5)©10 (5.5‒0.4)©10 Power coefficient of reactivity for “hot”, “poisoned” state, def/MW-d -(4.0‒1.3)© -3 -310 -(4.7‒0.3)©10 Simplicity and possibility to rapidly introduce the development results are accompanied by certain losses of neutron flux density in several reflector channels due to their shielding by the new ones loaded instead of fuel assemblies at the boundary between the core and reflector. The investigations and developments performed on the second stage of the modernization include, along with other activities, the development of a new fuel rod with small parasitic neutron absorption and of a fuel assembly with rod-type burnable absorbers, as well as the core rearrangement that allows us not only to compensate the disadvantages of the first stage but to obtain additional advantages too. IAEA-CN-156/U-22 Set of Investigations of HFR Fuel Rods in Justification of Their Serviceability and Safe Operation TsykanovV.ん., Chechetkina Z.I., Khudyakov ん.ん., Strizhenok ぜ.N., YakovlevV.V., Novoselov ん.ぎ., Shishin V.Yu., Starkov V.ん., Fedoseev V.ぎ.,( FSUE SSC RIAR, Russia). RIAR perform works on the upgrading of the SM high-flux research reactor core. The main objective of these works is to improve effectiveness of the reactor utilization by increasing a scope of experiments with a high density of neutrons intended for isotopes accumulation, irradiation of structural and other materials, as well as for the performance of different kinds of experiments. The objective may be achieved by the arrangement of irradiation channels within the SM reactor core. 235 The resulting reactivity losses can be compensated in two ways: by increasing of U loading into the fuel assemblies or using new fuel rods with a less cross-section of neutron absorption. At the initial upgrading stage it was decided to use standard fuel rod designs with the uranium loading increase by 20 percent in order to compensate the reactivity losses. Standard fuel rods of the SM reactor are made on the basis of dispersion fuel composition UO2 – Cu, while the fuel rod claddings are made of steel EI-847. The fuel rod section is cross- shaped. The SM standard fuel rods are successfully operated for 40 years. They have shown high 2 serviceability and operation safety at a high density of the heat flux from the surface of 15 MW/m However, current commissioning of the new or modified nuclear fuel provides for a performance of the required set of tests and investigations in justification of its serviceability. In this connection, RIAR implement a program for a study of serviceability and safe operation of the modified SM fuel rods with the increased uranium loading. A set of works presented in this paper incorporates the results of calculations and investigations of the experimental fuel rods tested under specially performed reactor experiments with a simulation of a wide range of neutron-physical and thermal-physical irradiation parameters on the fuel rods. Simulated parameters of the reactor tests of the fuel rods changed within the 14 3 15 3 following ranges: fission density (4.42*10 1/cm – 1.08*10 1/cm ), thermal flux density (6.62 2 2 MW/m – 15.7 MW/m ), temperature (295ºで – 582ºで), and concentration of the fission products up 3 to 1,1g FP/cm . Reactor test conditions different in temperatures and thermal loadings allowed to obtain detailed data on the behavior of fuel rods in general and of their components in particular. The paper describes a connection between neutron-physical and thermal-physical parameters of the reactor tests and a change in the properties of the fuel rod components. The following regularities of the change in macro- and microstructures of the fuel column were revealed: further sintering and its influence on the porosity in the fuel particles and matrix, migration of voids and fuel particles in the fuel columns. Peculiarities of the radial and local swelling of the fuel rods, as well as the swelling at different fuel rod elevations, were revealed. Dependencies of the radial swelling on the accumulation of fission products, fission density, heat flux density and testing temperature were plotted. The dependencies allow for a qualitative evaluation of the fuel rod swelling at different stages of operation and different values of the thermal-physical parameters, as well as for a selection of safe parameters subject to the ultimate swelling of the fuel rods and technical conditions. IAEA-CN-156/U-12 Utilization of Ir radiation Holes in HANARO Choong-Sung Lee, Guk-Hoon Ahn, In-Cheol Lim Korea Atomic Energy Research Institute (KAERI), Daejeon, Korea cslee1@kaeri.re.k r Since 1996, HANARO has been widely used fodri oraisotope production, material and fuel irradiation tests, beam application researncdh naeutron activation analysis. Seven irradiation holes are provided in the core and twehnotyle s including two NTDN(eutron Transmutation Doping) holes, large hole and NAA holes arec altoed in the reflector tank. The fuel and material irradiation tests that require lonrrga di iation time are performed by using the seven holes where the forced circulation of coroew fl exists. But the targets to produce RI(Radio Isotope) with the short half life are mainly idrriated in the holes at reflector tank where the natural convection of the pool water is available. In the initial stage of normal operation, thteil izuation of irradiation holes was not active because the development of capsules for irtriandgi amaterial, RI targets and test fuel was delayed. The capsule for RI production at ehso lin reflector tank was developed easily because the coolant flows by natural convectiont .i nB ucase of the capsusl eusing holes in the core, a locking device should be providedp troe vent an inadvertent removal during power operation and fixing device to reduce the vibratoiof nt he guide tube for instrument such as thermo couple and SPND by coolant flown. dA it should be confirmed for any kind of irradiation tests to safely remove the hienadtu ced by fission events or gamma heating during tests. The development of these capsulesd em tahe utilization of the irradiation holes vigorously. Especially the development of tihnest rument capsule has accelerated utilization of material and fuel irradiation. Figure 1 showes sthtatus of utilization of the irradiation holes in the core by year. 100 90 OR6 OR5 80 OR4 70 OR3 IR2 60 IR1 50 CT 40 30 20 10 0 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006 Year FIG. 1. The utilization status of irradiation holes in core by y. ear Utilization of irradiation in the core% ( IAEA-CN-156/U-12 In 1996, the first irradiation stet at HANARO started with teh HANARO test fuel containing 3 rods as the HANARO fuel qualification programt h igh power that was required to resolve a conditional licensing prerequisite. During dleovpeing the capsule for material and RI, the test fuels manufactured for localization tohfe HANARO fuel had been irradiated in the irradiation holes in the core until 1998. Thev edleopment of RI capsule for producing Ir-192 and the instrument capsule for material rtesqtu iring the high fast and thermal flux condition made the irradiation holes in the center area of the core used continuously. Another reason for increasing the utilization HoAf NARO is the fuel irradiation tests to develop the new fuels. In 1996, the first LEUS Ui 3fuel produced by atomization process was fabricated into a mini-assembly. After irratdioian test of this one, the full-length bundle test was performed. From the results of theirsrea diation tests, HANARO fuel manufacturing facility was completed on May 2004 and lomcaal nufacturing of theH ANARO fuel started. A qualification program for a rod type fuel aotfo mized U-Mo was initiated in 2000. The first and second irradiation tests were carried ino u2t001 and 2003. The third test started in 2006 and will be finished in 2007. The U-Mo fuerlr aidiation tests will contribute to new fuel development for research reactor. KAERI haese nb developing a small and medium reactor, SMART for electricity generation and sea water desalination. U-Zr fuel was adopted for the SMART reactor. The first irradtiiaon test of U-Zr fuel with three 8.1w/o enriched fuels and the second test with three 8痩.9 10.0 w/o fuels were compledt.e The third test had been performed with 19.75w/o fuels from 2004 to 2006. aA psower reactor fuel, the large grained UO2 pellet has been developed aiming hbiguhr n-up in KAERI. Two test assemblies are loaded in the upper and lower positions of as cualep. The test assembly of the upper position was unloaded for PIE (Post irradiation examtioinna) and the lower position assembly will be irradiated by the burn-up of 70MWD/kgU. KAREI has been studying DUPIC fuel through international co-operative research toge twheitrh CANADA, U.S.A. and IAEA. Up to the present, irradiation tests of DUPIC (Direct eU osf spent PWR fuels in CANDU reactors) fuel have been conducted 6 times. Through the DUiPrrIaCd iation tests, thermal behavior of DUPIC pellet was analyzed and the technologry r efomote assembling and handling has been developed. Following the experiences of the fuel irradiatitoens t, FTL (Fuel Test Loop) has been installed in the core to extend the utilization of thee lf uirradiation. This facility will be used for irradiation of PWR and CANDU fuel pellets unrd tehe environments of power reactor with high pressure and high temperature anildl owperate from 2008 after commissioning. The several irradiation holes at the reflector tank naorte used yet. It is required to strengthen the technology for supporting users and to enlarge the utilization area. IAEA-CN-156/U-27 Design and Installation of Fuel Test Loop in HANARO Sung Ho Ahn, Su Ki Park, Dae Young C Bhio, ng Sik Sim, Kook Nam Park, Chung Young Lee, Hark Rho Kim HANARO Utilization Technology Development Division, Korea Atomic Energy Research Institute (KAERI), Daejeon, KOREA shahn2@kaeri.re.kr FTL (Fuel Test Loop) is a test facility whi ccould conduct fuel irradiation test at HANARO reactor [1]. The fuels can be tested thine IR1 irradiation hole of HANARO under the commercial power plant operating conditions. Tohnec ceptual design of the FTL started at the end of 2001 and the detailed design was fiendis bhy March 2004. The installation of FTL was finished successfully on March 2007. Them cmoissioning of the FTL will be conducted by January 2008. The FTL would be used for thed irartaion test of high burn-up PWR fuels from January 2008. The design characteristics aned inthstallation of the FTL facility are introduced in this paper. The FTL is composed of an IPS (In-Pile teestc tSion) and an OPS (Out Pile system). Fig. 1 shows the schematic diagram of the FTL. The iIsP tSo be loaded intoth e IR-1 position in the HANARO core. This implies that the environemnt around the IPS is subjected to a high neutron flux (Thermal neutron flux : 1.·2 1014 n/cm2/sec, Fast neutron flux : 1.·6 1014 n/cm2/sec). The IPS can accommodate up to 3 poifn sfu el and has instruments such as thermocouple, LVDT and SPND to measure theel fpuerformances during the test. The IPS is composed of IPS head, outer pressure vesselr, pinrneessure vessel, flow divider and test fuel carrier. Inlet nozzle and outlet nozzle for the m caoinoling water located in the IPS head and insulated from the HANARO pool. Neon gas is fdil liento the gap between the outer pressure vessel and inner pressure vessel to ins uthlaet eIPS from the HANARO pool. A flow divider divides the outlet cooling water from the inlet lcinogo water. The test fuel carrier is composed of a fuel carrier support stem (with 6 slots foer thhot cooling water injection), fuel carrier leg (3 legs are arranged through the fl1 a2n0gles) and a fuel carrier head. Accumulator B Accumulator A Safety Relief Valve HANARO Pool Safety Injection Valve Isolation Valve PCV IPS Vent Vessel Main ICW ValveCooler HX Isolation Valve Cooling Tower Pressurizer FCV FCV LMP System Main Heater Waste Disposal Tank Main Pump HANARO Reactor FIG. 1. Schematic diagram of FT L. IAEA-CN-156/U-27 The OPS contains pressurizer, cooler, pumhepa, ter and purification system which are necessary to maintain the proper fluid condit.i oIns addition, the OPS contains a engineered safety system that could safely shutdownh b HoAt NARO and the FTL if an accident occurs. The FTL simulates the irradiation conditionst hoef commercial power plants such as pressure, temperature and neutron flux levels to conducr tt hfoe irradiation and thermo hydraulic tests. The FTL coolant is supplied to the IPS aet trhequired temperature, pressure and flow conditions that are consistent with the teset l.f uThe nuclear heat added within the IPS is removed by the main circulating water ceoro. l The main circulating pump provides the motive power to circulate the FTL coolant winit hthe loop. After pump discharge, an in-line heater provides the capability to increase tematpuerer for startup and for positive temperature control. A pressurizer is provided to establaisnhd maintain the coolant pressure to the test fuel type. A purification and de-gasificatiosny stem is provided to maintain the coolant inventory and the chemistry conditions. Theem ergency cooling system is provided to maintain the experimental fuel cooling ine thevent of the anticipated operational occurrence or the design basis accidents. The OPS compso naeren tlocated in FTL room 1 (safety related components), room 2 (non-safety related pconments) and the FTL control room which are dedicated rooms for the FTL. The control system for FTL operation is divdid iento the safety control system and the non- safety control system [2]. The safety control system is used for controlling of the safety related FTL process systems and shutdown the HANARO reactor against the abnormal operating conditions. The non-safety control sys cteomnsisted of a computer control system controls the non-safety related process syst eTmhse. application fields of the FTL are as follows. - Nuclear fuel irradiation behavior test tahte operating condition of the commercial power plant. - Fuel burn-up and mechanical integrity verification. - Irradiation data generation for the analysis model - Technical improvement of design and fabrication for the advanced fuel development. - Fuel rod irradiation test for performance verification. [1] CHI D.Y., et. al., “Evaluation of the fuel stte loop room for HELB loads”, Journal of Korea Society of Mechanical Technology, Vol.6 (1), p. 67, 2004. [2] AHN S. H., et al, “Instrumentation and contrsoyls tem design of fuel test loop facility”, Proceeding of the Korea Nuclear Society Autumn Meeting, Korea (2004). IAEA-CN-156/U-42 Design Character istics of Cold Neutron Source in HANARO Sang Ik Wu, Young Ki Kim, Kye Hong Lee, Hark Rho Kim, In Cheol Lim Korea Atomic Energy Research Institute (KAERI), Daejeon, Korea E-mail address of main author: siwu@kaeri.re .kr The HANARO has ̌ªªœfoperated for 12 years since its inli tciariticality in February of 1995. The reactor power has been gradually incre atos e3d0 MWth through out its service period. In order to enhance the utilization capacoityf HANARO, a CNS (Cold Neutron Source) development project has been underwayc es i2n003. As of now, the detailed design of the CNS has been completed. The overall design concept of the CNS is to ensure that the reactor safety systems and the on-site personnel eaqnudip ment are not adversely affected by the hydrogen-oxygen reaction from the CNS. The sa dfetsyign criteria of the CNS are a defence- in-depth approach that provides several mse taon avoid any accidental contact between the hydrogen in the system and the air. Therefore, the principles of a conservatism, simplicity, redundancy, fail-safe design, and passive saffeeatytu res are included to design it with an enhanced safely and efficiency. According t htoe safety classification based on ANSI N51.1 [1], the HANARO reactor assembly is classified as safety class 3. A vacuum chamber, which will be installed into the refleoctr tank of the reactor, is the highest safety class of the CNS components. As the vacuum should maintiatsin i ntegrity in the case of a hydrogen-oxygen reaction, it is defined as an ultimate pres sbuoruendary in accordancei twh the ASME Sec. III NB code requirements; So, thef estay class of the vacuum chamber is specified as safety class 3. Every component except the vacuum chamber is classified as a non-nuclear safety class. The HANARO CNS adopts the liquid hydrogeans a moderator. The liquid hydrogen contained in the moderator cell evaporadtuees to a gamma heating. The hydrogen vaporizes up to the condenser, where it is re-liquefieedn t hit returns down to the moderator cell. This thermo-siphon loop can only be established unad veerr y low temperature environment, which requires a method for a thermal insulation. Tehfoerre, the processing system of the CNS basically consists of a Hydrogen System, aac uVum System, a Gas Blanketing System, and a Helium Refrigeration System. The Hydrogen Seyms t(HS) consists of an In-Pile-Assembly (IPA) connected to the hydrogen buffer tank through adequate piping, a metal hydride unit, and a valve manifold. The IPA consists of a vacuum chamber, a moderate cell, a heat exchanger and a cryogenic transfer tube. Th ew HasS designed with the closed loop concept to avoid a direct venting or its pressure rfe. lTiehe HS is completely surrounded by blanketing gases to avoid any accidental contact waitihr or water from outside the system. The blanketing gas will be helium or nitrogen denpdeing on the installation position. A part of the system in the reactor pool is filled with ane rint helium gas and the other part is filled with nitrogen gas. The HRS is to cool down and lfiqy uthee gaseous hydrogen to a sub-cooled state in the condenser in order to establish ear mtho-siphon. The HRS has two different operating capacities for both the CNS and the DeuteriumS C, wNhich will be added in the future. The HRS is being designed in accordance with the following operating conditions, 1500 watts at 14 K for the CNS and 2000 watts at 19 K for the Deuterium CNS. IAEA-CN-156/U-42 The Vacuum System (VS) is to act as therimnsaul lation for the cryogenic part of the IPA and act as a safety barrier against an irruptio nl iqoufids and / or gase from the outside. The thermal insulation is of relevance to the poermrfance of the IPA cooling system process. The VS consists of a pumping station, valvaensd gauges, and connecting pipes. The vacuum level for the cryogenic insulation aslhl be at least lower than -15 0torr, which will be achieved by means of two vacuum pumping sets. Oneth oef pumping sets is in operation while the other is on standby. All of eth VS are installed in a vacuum box filled with nitrogen blanketing gas. The discharging gas from vthaecu um pump is collected in a gas collection tank and then the collected gas is released into the reactor hall in the case of the hydrogen contact not being higher than 3.5% of theta lt ovolume. Fig. 1 shows a final schematic drawing came from the detailed design of CNS in HANARO. This paper describes the design charactitcesri sof a CNS in HANARO. The detailed design based on the safety criteria and user’s requirements has been completed and it will provide the basis by which the manufacturing, insntagl,li and commissioning of the CNS will proceed. All of the CNS equipment and systems excfeopr t the IPA will be installed in August 2008 and it will start its commissioning of the end 2o0f08. The IPA will be installed in the first half of 2009. It is expected that the CNS will be available from the beginning of 2010. w̃ vfı•̨ ›“f„“«„fi‹“„⁄»¶„ ›“f”…••†̂ ›ªsfififi ›ªsfifi ‰⁄§……‡fƒ¶́ ·xsfififi S ›ªsfi ‹⁄” ⁄·⁄†̂¯“„ ·xf”…••†̂ fı•̨ ‹⁄” ›̂¤„¶‹“·f ¶xf⁄·⁄†̂¯“„ ⁄·⁄†̂¯“„ ƒ¶́ ⁄·⁄†̂¯“„ fi•⁄ „́f›⁄†† ·xsfi ‡“»⁄†f ›̂¤„fi¤“f…·fi» ‹⁄”f §¶††“§»fi¶·f»⁄·‒ ›xfƒ…««“„f»⁄·‒ ·xsfifi ›xf”…••†̂ FIG. 1. ”̶Œªø̨°º̶f¤¬̨³ºœØfߣf§·”fºœf›⁄·⁄„¶ [1] “Nuclear Safety Criteria for the Design oSftationary Pressurized Water Reactor Plants”, ANSI N51.1 [2] S.I. Wu, Y.K. Kim, and Y.J. Kim, ‘SafetDy esign Criteria for Facilities of Cold Neutron Source in HANARO’, HANARO Workshop 2004, May, 2004 [3] Y.K. Kim, S.I. Wu, and Y.J. Kim, ‘Overall Design Concept for the Systems and Facilities of a Cold Neutron Source HinA NARO’, Proceeding of the International Symposium on Research Reactor and Nroenu tScience in Commemoration of thet h1 0 Anniversary of HANARO, April, 2005 S vfı•̨ w{vfı•̨ S IAEA-CN-156/U-47 PURE COMMERCIAL GOLD FOILS AS NEUTRON FLUX MONITOR: NEUTRON SELF-SHIELDING ASSESSMENT Haddad Kh., H. Haj-Hassan, W. Helal Nuclear Engineering Dept. Atomic Energy Commission Of Syria P.O. Box 6091 Damascus Syria khhaddad@aec.org.sy 1. EXPERIMENTAL Certified activation monitors such as IRMM-530R Al-0.1% Au are used to determine neutron flux [1]. Considering the high cost and other complications, pure (99.9 %) commercial gold foils was introduced as an alternative material. The minimum available thickness of the commercial gold foil was 0.1 mm. According to that the principal problems which we must solve are: results repeatability investigation and determination of the neutron self-shielding factors for both thermal and epithermal neutrons for the prepared foils. Two groups of pure commercial gold foils of 2 mm diameter were prepared. The first one was of 0.1 mm thickness and was used to investigate the results repeatability. The second group was five couples of samples of thicknesses ranges from (0.1 to 0.5) mm and was used to determine the neutron self-shielding factors for both thermal and epithermal neutrons. All samples were irradiated in the same conditions in the inner irradiation sites of MNSR. The specific saturated activities resulting from thermal (th) and epithermal (epi) activations were determined experimentally [2]. Table1 shows the results repeatability. Table 1: The results repeatability bare Cd-covered d, mm n* 1, Bq/nuclide stdev% n* 2, Bq/nuclide stdev% 0.1 15 7.62E-11 5.64 6 1.55E-11 5.23 0.2 3 6.73E-11 7.67 3 1.10E-11 8.93 0.3 3 6.12E-11 11.72 2 9.04E-12 0.25 0.4 3 5.34E-11 10.44 3 8.57E-12 8.86 0.5 3 5.37E-11 7.19 3 7.62E-12 8.28 * - test frequency The second group foils were divided into two subgroups. The first subgroup foils were irradiated bare and the second subgroup foils were irradiated in the cadmium boxes. Induced specific activities were measured. The neutron self-shielding factor G for a foil was determined (Fig. 1). IAEA-CN-156/U-47 1.0 epi th 0.5 0.0 2 s g/cm 0 0.5 1 1.5 2 2.5 3 3.5 4 Fig.1: The curves of the neutron self-shielding factors versus the foil thickness for thermal and epithermal range. The resulted values of the thermal and epithermal neutron fluxes were compared with the corresponding values resulted by using certified activation monitors (IRMM-530R Al-0.1%). Table 2: Comparison between the results of certified activation detector and in-house one. detector  th er. %  e er. % standard 9.21E+11 2.5 5.86E+10 1.0 in-house 9.27E+11 3.1 6.31E+10 3.6 2. CONCLUSION An assessment of pure commercial (99.9 %) gold foils as neutron flux monitor was performed. A thin foils of pure commercial gold were prepared as an in-house reference material for neutron flux measurement. The assessed foils are available commercially and its cost is much less than the certified ones. Determination of the neutron self-shielding factors in these foils for both thermal and epithermal neutrons have been done experimentally. These foils show good results repeatability and good agreement with certified activation monitors. According to the well-known physical constants of the nuclide and its low cost comparing with certified foils, it can be used as an in-house reference monitor. The authors would like to thank Prof. I. Othman, the Director General of the Atomic Energy Commission of Syria, for his continuous support . REFERENCES 1. Ingelbrecht,-C.; Robouch,-P., BCR certified reference materials for reactor neutron dosmetry, Journal-of-Radioanalytical-and-Nuclear-Chemistry (Sep 2003) v. 257(3) p. 649-652 2. Neutron Fluence Measurements, IAEA, Technical Reports Series No. 107, Vienna 1970. G Design Character istics of Cold Neutron Source in HANARO Sang Ik Wu, Young Ki Kim, Kye Hong Lee, Hark Rho Kim, In Cheol Lim Korea Atomic Energy Research Institute (KAERI), Daejeon, Korea E-mail address of main author: siwu@kaeri.re .kr The HANARO has ̌ªªœfoperated for 12 years since its inli tciariticality in February of 1995. The reactor power has been gradually incre atos e3d0 MWth through out its service period. In order to enhance the utilization capacoityf HANARO, a CNS (Cold Neutron Source) development project has been underwayc es i2n003. As of now, the detailed design of the CNS has been completed. The overall design concept of the CNS is to ensure that the reactor safety systems and the on-site personnel eaqnudip ment are not adversely affected by the hydrogen-oxygen reaction from the CNS. The sa dfetsyign criteria of the CNS are a defence- in-depth approach that provides several mse taon avoid any accidental contact between the hydrogen in the system and the air. Therefore, the principles of a conservatism, simplicity, redundancy, fail-safe design, and passive saffeeatytu res are included to design it with an enhanced safely and efficiency. According t htoe safety classification based on ANSI N51.1 [1], the HANARO reactor assembly is classified as safety class 3. A vacuum chamber, which will be installed into the refleoctr tank of the reactor, is the highest safety class of the CNS components. As the vacuum should maintiatsin i ntegrity in the case of a hydrogen-oxygen reaction, it is defined as an ultimate pres sbuoruendary in accordancei twh the ASME Sec. III NB code requirements; So, thef estay class of the vacuum chamber is specified as safety class 3. Every component except the vacuum chamber is classified as a non-nuclear safety class. The HANARO CNS adopts the liquid hydrogeans a moderator. The liquid hydrogen contained in the moderator cell evaporadtuees to a gamma heating. The hydrogen vaporizes up to the condenser, where it is re-liquefieedn t hit returns down to the moderator cell. This thermo-siphon loop can only be established unad veerr y low temperature environment, which requires a method for a thermal insulation. Tehfoerre, the processing system of the CNS basically consists of a Hydrogen System, aac uVum System, a Gas Blanketing System, and a Helium Refrigeration System. The Hydrogen Seyms t(HS) consists of an In-Pile-Assembly (IPA) connected to the hydrogen buffer tank through adequate piping, a metal hydride unit, and a valve manifold. The IPA consists of a vacuum chamber, a moderate cell, a heat exchanger and a cryogenic transfer tube. Th ew HasS designed with the closed loop concept to avoid a direct venting or its pressure rfe. lTiehe HS is completely surrounded by blanketing gases to avoid any accidental contact waitihr or water from outside the system. The blanketing gas will be helium or nitrogen denpdeing on the installation position. A part of the system in the reactor pool is filled with ane rint helium gas and the other part is filled with nitrogen gas. The HRS is to cool down and lfiqy uthee gaseous hydrogen to a sub-cooled state in the condenser in order to establish ear mtho-siphon. The HRS has two different operating capacities for both the CNS and the DeuteriumS C, wNhich will be added in the future. The HRS is being designed in accordance with the following operating conditions, 1500 watts at 14 K for the CNS and 2000 watts at 19 K for the Deuterium CNS. Poster Presentations: Fuel and Waste Management Synopses no. IAEA-CN- Synopses Title Main Author 156/ F-4 Converting HIFAR to Low Enriched Uranium Fuel Storr, G.J. The regulatory role of the Australian Radiation Protection and Nuclear Safety Agency in relation to F-6 spent fuel arising from research reactors in Australia Sarkar, S. Design and construction of a Decay Pool at IAN-R1 F-7 Research Reactor Sarta Fuentes, J.A. Safety Assessment for Decommissioning of Research DE-1 Reactors Kaulard, J. Further Development in Characterization of Radioactive Waste Drums by Non Destructive Gamma WM-4 Spectrometry at GRR-1 Savidou, A. Al-Clad Spent Nuclear Fuel Corrosion Studies at F-1 Magurele site, Romania Dragolici, A.C. F-2 Vinca site preparation for spent fuel shipment Pešic, M. Experimental study of systematic errors of gamma WM-5 technique for assay of radioactive waste drums Tran, Dung IAEA-CN-156/F-4 Converting HIFAR to Low Enriched Uranium fuel 1) 1) (1) Greg Storr , David Vittorio , Rodney Hall 1) ANSTO, Lucas Heights, NSW, Australia E-mail address of main author: gjs@ansto.gov.au The Australian Nuclear Science and Technology Organisation (ANSTO) began operating the High Flux Australian Reactor (HIFAR) in 1958, a DIDO-class research reactor operated at a th thermal power of 10 MW. On 30 January 2007, after more than 49 years of successful and safe operation HIFAR was finally shutdown. Since that time all the fuel has been successfully removed from the reactor containment building. HIFAR was primarily used for neutron scattering science, service irradiations and isotope production. Over the nearly 50-year operating life of HIFAR a variety of fuel designs have been used. After the 1970s fuel enrichment was reduced in stages from over 90 percent to 19.75% in 2006. The reactor core consisted of 25 fuel elements with uranium-aluminium alloy fuel sections, arranged in concentric tubes. HIFAR was moderated and cooled by heavy water, and the coolant contained within an aluminium tank, which in turn was surrounded by a graphite reflector and concrete biological shielding. Reactor control and shutdown were achieved with six europium tipped cadmium control blades, which moved as a bank between the rows of fuel elements. Two cadmium shutdown rods provided additional shutdown capacity. In May 2006 the HIFAR reactor was fully converted to Low Enriched Uranium fuel. The conversion commenced in October 2004. The LEU fuel was procured from RISO National Laboratory in Denmark, was originally made for use in the DR3 reactor, and was modified to be compatible with HIFAR. This type of fuel was used safely in DR3 before its closure. A safety analysis report for the approval and use of the LEU fuel which was prepared well in advance of loading the fuel into HIFAR, provided detailed analyses of issues important to reactor and general fuel safety, including, criticality safety outside the reactor, reactor physics, eversafe times, thermal hydraulics and accident analyses. Many of the issues studied for LEU fuel reanalysed operational and accident conditions that had been previously analysed for HEU fuel. In most cases the conclusions provided in each analysis demonstrated there was little difference in behaviour between HEU fuel and LEU fuel in HIFAR under operational and accident conditions. However, there was one significant difference between HEU and LEU fuel as it was shown that in general eversafe times for LEU fuel are greater than for HEU fuel. Consequently, procedures were modified for some operations to ensure compliance with safe heat limits. The paper will present the process undertaken for the conversion of HIFAR, including the development of the safety case, requirements for regulatory approvals, and results from the conversion program. IAEA-CN-156/F-6 The regulatory role of the Australian Radiation Protection and Nuclear Safety Agency in relation to spent fuel ar ising from research reactors in Australia S Sarkar Samir.Sarkar@arpansa.gov.au This paper will describe the elemenatnsd performance of ARPANSA’s regulatory management of spent fuel arising ins Atrualia, with particular emphasis on the experience of ensuring compliance with Cthoed e of Practice Code of Practice for Safe Transport of Radioactive Materials r einlation to in land surface transport of spent fuel within Australia. The Australian Radiation Protection and Neuacrl Safety Agency is the regulatory authority for Commonwealth entities, suacsh the Australian Nuclear Science and Technology Organisation (ANST)O, who operate nuclear inaslltations in Australia.. Nuclear installations that operate undAeRrP ANSA facility licence include research reactors and plants for the storage and mgeamnaent of research reactor fuel. ANSTO is the only operator of nuclear installations in Australia. The Australian Radiation Protection and Nuc lSearfety Agency is also the competent Authority for inland surface transport. ARPANSA has adopted the IAEA Safety Regulations for Safe Transport of Radioivaec tMaterials domestically in the form of the ARPANSA Code of Practice for Safe Tsrapnort of Radioactive Materials (RPS 2). As the competent authority ARPANSA apopver s the shipment and design of a new cask, validate original certificate apypinlg the requirements of the RPS 2. ARPANSA’s regulatory oversight of compnlicae with the requirements of its own legislation and the requiremtesn of the Code emphasisaess urance of safety in the operation of nuclear installations ande tshhipment of spent fuel is achieved principally by prior assessment ofe t hoperator/consignors safety case, and by compliance monitoring through regular repogrt i(nquarterly and annually), as well as planned and reactive inspections. Durineg othperating life of tehse facilities for several decades there have been no incsid wehnitch have had off-site or significant on-site, consequences. This paper will examine that experiencned ain particular focus on the regulatory experience of oversight of recent shipment of spent fuel arising from the HIFAR reactor, including the methods for reviewing requests fporo avpal and the issues that have emerged. IAEA-CN-156/F-7 IAEA-CN-156/F-7 IAEA-CN-156/DE-1 Safety Assessment for Decommissioning of Research Reactors International Project on Evaluation and Demonstration of Safety during Decommissioning of Nuclear Facilities (DeSa) 1) 2) 3) J. Kaulard , S. Thierfeldt , B. Batandjieva 1) Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Germany 2) Brenk Systemplanung GmbH, Germany 3) International Atomic Energy Agency (IAEA), Vienna, Austria E-mail address of main author: joerg.kaulard@grs.de Decommissioning is the final stage in the life cycle of each nuclear facility, including research reactors. In the past, there was a tendency to address related aspects at a very late period of operation of the facility or even after final shut down stage. While this tendency has been changing especially in case of nuclear power plants and research reactors, the international community recommended to IAEA in the Berlin Conference in 2002 to stipulate an early addressing of decommissioning [1]. This emphasis took also into consideration the fact that worldwide there are over 500 research reactors and critical assembly units that will eventually require decommissioning. In addition, the international community recommended to IAEA to provide general assistance in development and review of safety assessments related to decommissioning. Accordingly, in its International Action Plan on Decommissioning of Nuclear Facilities, approved by the IAEA Board of Governors in 2004, IAEA reflected these recommendations [2] and initiated the international project on Evaluation and Demonstration of Safety during Decommissioning of Nuclear Facilities (DeSa Project) [3]. For the last three years about fifty experts from over thirty Member States have been working in the DeSa project on; (i) the establishment of a harmonized general safety assessment methodology for decommissioning; on (ii) the development of recommendations for a regulatory review process of such safety assessments; (iii) development of recommendations on the application of a graded approach to performance and review of safety assessment, which ensures that the extent of the safety assessment is commensurate which the risks posed by the facility and the proposed decommissioning activities, and finally (iv) the application of the assessment methodology, the regulatory review procedure and graded approach recommendations to three test cases with different complexities and hazard potentials – a nuclear power plant, a research reactor and a nuclear laboratory. This paper provides an overview on current status of the DeSa project activities, and their application to the development of the Research Reactor Test Case, including the presentation of preliminary lessons learned from applying the graded approach on a research reactor. The DeSa project results, including the outcomes of the review of the Research Reactor Test Case th by applying the review procedures are envisaged to be summarized at the 4 Joint DeSa meeting in October 2007, where the scope and objectives of a follow-up project will be also discussed. IAEA-CN-156/DE-1 [1] INTERNATIONAL ATOMIC ENERGY AGENCY, Safe Decommissioning for Nuclear Activities, Proceedings of an International Conference, Berlin, 14 – 18 October 2002 [2] INTERNATIONAL ATOMIC ENERGY AGENCY, International Action Plan on Decommissioning of Nuclear Facilities, BOG, June 2004 [3] INTERNATIONAL ATOMIC ENERGY AGENCY, International Project on Evaluation and Demonstration of Safety for Decommissioning of Nuclear Facilities (DeSa), http://www-ns.iaea.org/tech-areas/waste-safety/desa/start.asp IAEA-CN-156/WM-4 Further Development in Characterization of Radioactive Waste Drums by Non Destructive Gamma Spectrometry at GRR-1 A. Savidou, F. Tzika and I. E. Stamatelatos Institute of Nuclear Technology and Radiation Protection, NCSR “Demokritos”, Aghia Paraskevi, Attiki, Greece E-mail address of main author: savidou@ipta.demokritos.gr In the present work a non destructive technique based on gamma spectrometry and application of the Monte Carlo method for detector efficiency calibration, was used to assay radioactive content waste drums. TM Exploranium GR-130 miniSPEC portable gamma ray spectrometer was used to externally monitor 26 waste drums containing radioactivity of ion exchange resin waste from the water demineralization system of GRR-1 open pool-type research reactor facility at NCSR o “Demokritos”. A three 20 min consequent measurements (one every 120 of drum rotation) scheme was employed. The GR-130 gamma ray spectrometer was equipped with a 38 mm 137 diameter x 57 mm long NaI(Tl) scintillation detector of 7% resolution for Cs at 662 keV. Monte Carlo simulation of the drum and detector configuration was carried out using MCNP- 4C2 code and cross-section data from the ENDF-VI-b library. The code was used to perform numerical simulations taking into consideration the energy of the gamma ray emitter, the matrix material, the detector efficiency, the geometric configuration employed, the size of the drum, and the wall material and thickness of the drum. The detector was modelled as a cylinder of sodium iodide surrounded by an aluminum layer. The model geometry included a homogeneous distributed cylindrical volume source within a cylindrical iron drum. Runs were performed for cylindrical sources of diameter 55.6 cm and heights ranging between 60 cm and 40 cm, thus representing different drum loads. Three resin densities of 0.75, 0.84 and 0.92 -3 g cm were modelled. The NaI crystal active centre was positioned at 57 cm from the geometrical centre of the drum at the drum mid-height level with its axis vertical to the drum’s main axis of symmetry. The results of MNCP simulations of a point source positioned at the geometrical centre of an air filled drum were calibrated against experimental measurements performed under the same 60 137 152 conditions. Measurements were performed using Co, Cs and Eu standard point sources and appropriate adjustment factors for the MCNP calculations were derived for given photon energies. Efficiency curves in the energy range 60 to 1500 keV were predicted. A relative error of less than 5% was achieved in all simulated cases. Satisfactory agreement was observed by comparing the results of the non destructive method against analytical results of samples obtained from each drum. Fig. 1 and 2 show the activity concentrations in the waste drums as estimated by the non destructive drum assay, using efficiencies derived by the Monte Carlo method, and by the sample analysis technique, for 108m 60 Ag and Co radioisotopes, respectively. In these figures the z-scores representing the difference between the two measuring techniques in combined standard deviation units are also shown. IAEA-CN-156/WM-4 1.0 2 0.8 0.6 0 0.4 -2 0.2 linear fit x=y line 0.0 0.0 0.2 0.4 0.6 0.8 1.0 1 2 3 4 5 6 8 9 10 11 12 13 14 15 16 17 18 19 20 22 23 24 25 -1 Destructive (Bq g ) Drum number 108m FIG. 1. Ag activity concentration in the waste drums as estimated by the non-destructive drum assay and determined by the conventional gamma spectrometry technique. 3.5 3.0 2 2.5 2.0 0 1.5 1.0 -2 0.5 linear fit x=y line 0.0 0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 1 2 3 4 5 6 7 8 9 10111213141516171819202122232425 -1 Destructive (Bq g ) Drum number 60 FIG. 2. Co activity concentration in the waste drums as estimated by the non-destructive drum assay and determined by the conventional gamma spectrometry technique. The derivation of detector efficiency was performed assuming a homogeneous distribution of the source activity within the matrix material. Moreover, the MCNP model was used to examine the effect of inhomogeneities, variation of matrix material density and drum filling height on the accuracy of the technique and estimate the measurement bias. The simulations of the present study can easily be extended to model other materials and container types and sizes as well. Therefore, the technique can also be applied in other radiation protection applications such as biological radioactive waste from hospitals or other physical forms materials that can be met in radioactive waste. -1 Non-destructive (Bq g ) -1Non-destructive (Bq g ) Z-score Z-score IAEA-CN-156/ F-1 Al-Clad Spent Nuclear Fuel Corrosion Studies at Magurele site, Romania C. A. Dragolici 1), A. Zorliu 1), E. Neacsu1 ), M. Roth 2) 1) National Institute of R&D for Physics and Nuclear Engineering (IFIN-HH), Bucharest- Magurele, Romania 2) Nuclear Research Branch (SCN), Mioveni-Pitesti, Romania E-mail address of main author: adrag@nipne .ro This paper is an overview of the scientifnicv ei stigations of the IAEA Coordinated Research Project (CRP) on “Corrosion of Research Reactor Aluminium-Clad Spent Fuel in Water (Phase II)”, as carried out at IFIN-HH instit,u tMeagurele, Romania. The results of corrosion surveillance activities on racks containing coup(osnese Table I) immersed either in vertical or horizontal position in individul afuel storage basins are presented in this paper. IAEA instructions for coupon preparations and immersion [1] were strictly followed. TABLE I: ORDER AND POSITION OF COUPONS IN RACK. S No. IAEA order BASIN 2 BASIN 3 TOP TOP (Site order) ID # TOP (Site order) ID # CR1- Ceramic ring Ceramic ring Ceramic ring Coupon 1 Site specific alloy Site specific alloy 2-1 Site specific alloy 3-1 CR2- Ceramic ring Ceramic ring Ceramic ring Coupon 2 Site specific alloy Site specific alloy 2-2 Site specific alloy 3-2 Coupon 3 SS304 SS304 320 SS304 323 CR3- Ceramic ring Ceramic ring Ceramic ring Coupon 4 Szav-1 Szav-1 385 Szav-1 369 CR4- Ceramic ring Ceramic ring Ceramic ring Coupon 5 Szav-1 (P-O & S) Szav-1 (P-O & S) 316 Szav-1 (P-O & S) 304 CR5- Ceramic ring Ceramic ring Ceramic ring Coupon 6 Szav-1 Szav-1 339 Szav-1 370 Coupon 7 Szav-1 Szav-1 324 Szav-1 355 CR6- Ceramic ring Ceramic ring Ceramic ring Coupon 8 Szav-1 Szav-1 348 Szav-1 330 Coupon 9 SS304 SS304 346 SS304 339 CR7- Ceramic ring Ceramic ring Ceramic ring Coupon1 0 AA6061 AA6061 310 AA6061 304 CR8- Ceramic ring Ceramic ring Ceramic ring Coupon 11 AA6061 (P-O & S) AA6063 (P-O & S) 200 AA6063 (P-O & S) 209 CR9- Ceramic ring Ceramic ring Ceramic ring Coupon1 2 AA6061 AA6061 340 AA6061 319 Coupon1 3 AA6061 AA6063 246 AA6063 236 CR10- Ceramic ring Ceramic ring Ceramic ring Coupon 14 AA6061 AA6063 233 AA6063 (P-O & S) 216 Coupon1 5 SS304 SS304 324S S304 357 CR11- Ceramic ring Ceramic ring Ceramic ring BOTTOM BOTTOM BOTTOM P-O & S = pre-oxidized and scratched IAEA-CN-156/F-1 Specific procedures for coupon preparations efxoar mination and further analysis [2] were implemented for obtaining consistent results. Texhpee rimental investigations have been done on aluminium alloys (AlMg3, AA 6061, AA6063 and Szav-1) and stainless steel 304 coupons. Single coupons, bimetallic and crevice coupled h baeven investigated. Visual examinations, metallographic and ultrasonic investigations w[3e]r e used. Also, an image analysis software was applied to measure the surface areas eadff ebcyt corrosion. The experimental data on the most affected coupons by galvanic corrosion have been statistic analyzed. The sing3l e AlMg coupons have presen tead non-significant pitting corrosion, for SZAV-1, AA6063 the susceptibility to corrosion phenomenon is reduced, excepting AA6s0e6e1 F (ig. 1); FIG. 1. Coupon R2-02, pit aspe ct. The oxidized and scratched coupons are not affected by corrosion; The coupled coupons made from identical mrialt eare affected by crevice corrosion, the maximum pits depth being less than 0.1 mm; The bimetallic coupled discs present a galvanic corrosion, with material dissolution of extended areas (1.1%-1.7% of the surface); ftfheec te of weight reducing is more pronounced on this type of coupons. Also the pitting factor presents the highest value = 3.2; The most affected coupons are those immersed in Basin 3, in the coupled conditions. On horizontal rack, the appearance of pitticnogr rosion is clearly shown only at the first coupon owing to the deposit foreign particles f rtohme basin lids. These particles are cast iron and paint and also dust from the air carried by the ventilation system. [1] Test Protocol, IAEA CRP “Crorosion of Research Reactor Aluminium-Clad Spent Fuel in Water (Phase II)”. [2] ASTM – G1, Standard Recommended Prcaec tfior Preparing, Cleaning and Evaluating Corrosion Test Specimen. [3] ASTM – G46, Standard Recommended Pcraec tfior Examinations and Evaluation of Pitting Corrosion. IAEA-CN-156/F-2 Vinca Site Preparation for Spent Fuel Shipment M. Peši5, V. Ljubenov, O. Šoti5 Vinca Institute of Nuclear Sciences, Belgrade, Serbia Heavy water research reactor RA at theVinca Institute of nuclear sciences, Belgrade, Serbia was constructed according design developed in the Institute for Theoretical and Experimental Physics in ex-USSR in late 40es and mid-fifties of last century. The reactor at Vinca site was in operation from 1959 to 1984. It has operated using Russian origin TVR-S type fuel elements made of uranium metal (LEU) from 1959 to 1976. From 1976 for the reactor operation was used the same type fuel elements, but designed as uranium dioxide (HEU) dispersed in aluminium matrix. The reactor was temporary shut down in 1984 for refurbishment, but the operation was never started again. Serbian government has decided in 2002 to shut down the RA reactor permanently and to proceed with development of the decommissioning planning. At the same time the Government made decision to ship fresh and spent nuclear fuel (SNF) arose from the reactor operation back to the country of origin – Russian Federation within the frame of the Russian Research Reactor Fuel Return Program. IAEA has coordinating this Serbian SNF shipment program since 2003. A three-party (IAEA, Vinca and Russian Consortium Mayak/Sosny/Tenex) contract was signed at the IAEA on September 2006 for Vinca’s spent fuel repackaging and shipment. This paper covers the all activities on Vinca site initiated already in 2003 for SNF preparation and the RA facility and Vinca site modification within the frame of the contract. These activities include: ‚" Collection of the facility relevant documentation and converting to electronic form ‚" SNF identification and verification ‚" SNF condition assessment and investigation ‚" SNF nuclide inventory calculation ‚" Dose rate measurements from SNF storage containers and fuel elements for verification of calculation results ‚" SNF burn up measurements for verification of calculation results ‚" Monitoring of activity of 137Cs nuclide in pool water and in SNF storage containers ‚" Removal of sludge from water from SNF pool to increase water transparency ‚" Upgrading facility ventilation systems ‚" Upgrading monitoring systems for detection of emission of radionuclide from ventilation stack of the facility ‚" Examination (and possible increasing) of loading capacity of floor in reactor room and storage room ‚" Verification of operation of tools and equipment used for SNF management ‚" Verification of operation and upgrading of internal transport devices ‚" Preparation of RAW management strategy to meet requirements of SNF repackaging and transportation technology ‚" Preparation of basins of SNF storage pool, reactor core and reactor block for SNF repackaging IAEA-CN-156/F-2 ‚" Preparation of SNF room and basins for repackaging equipment and temporary storage of repackaged SNF ‚" Removal carbon steel structure from SNF water storage basin ‚" Preparation of the reactor room, SNF room and basins for loading repackaged SNF to transport casks ‚" Review of Institutes roads and examination for possible modification ‚" Upgrading radiation protection equipment and health physics instruments ‚" Increasing safety culture of operators through intensive courses and training on mock-ups Acknowledgement Authors and the all members of the Vin7a VIND SNF/RadProt/RAW team acknowledge to the IAEA experts for TCP/Europe Department, IAEA experts from NFCWT /NFCM and DNIS/RRS Sections, Ministry of Science and Environmental Protection of the Republic Serbia, and the experts of the Sosny/Mayak/Tenex Consortium from the Russian Federation for their support, efforts and engagements in this project. IAEA-CN-156/WM-5 EXPERIMENTAL STUDY OF SYSTEMATIC ERRORS OF GAMMA TECHNIQUE FOR ASSAY OF RADIOACTIVE WASTE DRUMS Tran Quoc Dung*, Nguyen Duc Thanh, Luu Anh Tuyen, Lo Thai Son Centre for Nuclear Techniques 217 Nguyen Trai St, D. 1, Hochiminh City, Vietnam Ngo Minh Triet Dalat University Phu Dong Thien Vuong St, Dalat City, Vietnam 1) Basic principle of measurement The operation of nuclear reactors results in the production of a considerable amount of radioactive low density waste, mainly consisting of organic materials which are usually stored in large sealed drums (208 l). The drums must be checked to satisfy regulations of radioactive waste management. The Segmented Gamma Scanner (SGS) is an traditional tool for the isotopic composition measurement and for determination of the activity level in gamma contaminated waste drums [1, 2]. The systematic error of this technique is still large because of: non-uniform distribution of radioactive source within the drums frequently causes the largest error [3,4]; non-uniform distribution of non-radioactive materials (matrix) [4,5]; particles size of the nuclear material, the lump effect, specially for uranium and plutonium assay [2,6,7,8]; the drum-to-detector distance [4]. The other measuring technique has been studied by Cesana [9] for assay of the drums containing low density waste, mainly consisting of organic materials such as contaminated paper, rags, protective clothing, shoes, etc. The measuring arrangement consists two identical detectors set at equal distances from two bases of drum. This technique was developed because of the reasons: first, the measure is usually limited to rather hard gamma rays emitted by Cs-137, Cs- 134, Co-60..., the mass-absorption coefficients are nearly independent of the atomic number of matrix, and the linear attenuation coefficients are very low (typically 0.01-0.03 cm-2) because of the low waste density (0.2-0.4 g/cm3). Therefore, the gamma attenuation in whole drum can be considered by means of on average linear attenuation coefficient, independent of the position in a drum; second, the number of drums to be examined is supposed to be very large, so that a detailed scanning by SGS is practically impossible; third, the measuring arrangement is very simple, so it can be used for any situation. However, in this measurement the drum must lie down. Thus, it spends time and is not convenient for use in practice. Moreover, the error is still very large (223 %) if the activity is distributed in the vertical large region. In order to overcome these disadvantages a modification to the measurement arrangement with the geometric coefficient for cylindrical sample was proposed [10]. The measuring principle is given in Figure 1. A detector is set perpendicularly to the drum axis at its middle point at equal distances from the curve face of drum. *Author for corresponding, e-mail: trandungquoc@gmail.com IAEA-CN-156/WM-5 R= 30 cm Gamma source 4 Detector 150 cm 1 2 3 L/2= 45cm Collimator Turning table Spectrometer Figure 1: Waste drum counting geometry. The drum is measured two times. In the second 0 measurement the drum is rotated 180 . The activity I in the drum can be determined as ( 1 2C I = 1 C2 ) (1) G α      with G = − µ1 × 0 832×R (2) S2 where C1, C2- the count rates of the detector in two measurements; S=K-0.823.R; K- distance from detector to the drum. This distance is larger the diameter of the drum several times; µ1=2/S+µ; α- coefficient, that is a function of the gamma-ray energy, intrinsic efficiency of detector and the effective distance K. µ - the linear attenuation coefficient in the waste mixture. 2 R- radius of drum; The values of α/K versus gamma ray energy can be determined by using appropriate standard source. 2) Experimental study and discussion The proposed coefficient G in formula (2) is based on the some approximations [9,10]. The distance between the centre of the drum and the detectors is about four times the half of the height of the drum. Thus, the geometric mean of the efficiencies of the sources at different positions of vertical axis can be considered as the same. When the height of the drum is 86 cm, K= 120 cm the maximum variation is less than 6 %. Therefore, the vertical count rate variation can be ignored. The systematic errors are caused mainly by the horizontal inhomogeneity of the source in the drum because the important assumption of this technique is that the activity is distributed in a small region of the drum volume. In order to develop the technique to measure the activity of the radioactive waste drums released during operation of Dalat Research Reactor, Dalat City, Vietnam, the experimental study are carried out to confirm the calculation results of the systematic errors and evaluate the performance of this technique in practice. The results are also basic for establishment of measuring procedure for assay the radwaste drums. In this paper, the results of homogenous matrix and point source in different positions in the drum are presented. IAEA-CN-156/WM-5 In the experiments, a standard drum was used. Two point sources Co-60 of 0.36 MBq and two point sources Cs-137 of 0.64 MBq were located at four specified radial positions in the different measurements as in Fig. 1. Many pieces of matrix with various shapes and densities are -1 made from clothing materials. The linear attenuation coefficients are in range 0.011- 0.031 cm corresponding to gamma rays emitted by Cs-137 and Co-60 sources. A High-purity germanium detector- Gem 50P4, and a standard digital spectrometer, model DSPEC.JR-312-Ortec, with GAMMAVISION 32 version 6.1 and COLEGRAM version 2.01 were used for detection and analysis of gamma spectra. The results are given in Table I. Here the experimental and calculated values of errors are presented as ratio of the experimental and calculated to true value of activity. The experimental results confirm prediction of the theoretical results of systematic errors of this technique. When the activity concentrates as a point source the systematic errors are minimum and independent on the attenuation coefficient of matrix as given in Table I.a. In these cases the errors depend only on the position of the sources. The distance between the centre of the drum and the detectors is equal to many times the half of the height of the drum. Thus, the geometric mean of the efficiencies of the sources at different positions of vertical axis can be considered as the same. When the height of drum is 86 cm, and detector- to –drum is 120 cm the maximum variation is less than 6 %. This is proved in case of the source in position 5. Therefore, the vertical count rate variation can be ignored. The errors are caused mainly by the horizontal inhomogeneity of the source in the drum as case of two sources with the same activity in position 1 and 3 ( see Table I.b). 3. Conclusion The experimental results confirm performance of this gamma technique for assay of radioactive waste in practice. The maximum systematic errors are not larger than about 70% -1 when the attenuation coefficients of matrix are in the range of 0.01-0.03 cm . The maximum error (~ 41%) of this method is larger than that (~ 20%) of SGS techniques. However its measurement apparatus is very simple, perform of measurement is faster, and it can be used for most of the situations encountered in practice. TABLE I a) Comparison of the experimental and calculated values for estimation of the systematic error when a point source in different positions. Case of study One source in position 1 One source in position 2 Gamma Linear Experimental Calculated value Experimental value Calculated value Energy attenuation value of error of error of error of error (kev) coefficient -1 (cm ) 661 0.016 0.97 ± 0.01   0.91 ± 0.02   661 0.024 0.92 ± 0.01   1.04 ± 0.02   661 0.031 0.92 ± 0.02   1.01 ± 0.02   1173 0.013 0.99 ± 0.01    0.90 ± 0.02    1173 0.014 0.99 ± 0.01    0.88 ± 0.02    1173 0.023 0.96 ± 0.01   0.91 ± 0.02   1332 0.011 0.99 ±0.01   0.90 ± 0.02   1332 0.013 1.00 ± 0.01    0.91 ± 0.02    1332 0.019 0.97 ± 0.01   0.91 ± 0.02    1332 0.021 0.96 ± 0.02    0.93 ± 0.02    IAEA-CN-156/WM-5 TABLE I (continued) b) Comparison of the experimental and calculated values for cases of large systematic errors. Case of study One source in position 5 Two sources in position 1 and 3 Gamma Linear Experimental Calculated Experimental Calculated Energy attenuation value of error value of error Value of error value of error (kev) coefficient -1 (cm ) 661 0.016 0.81 ± 0.02     1.15 ± 0.02     661 0.024 0.87 ± 0.02     1.27 ± 0.03     661 0.031 0.76 ± 0.02     1.38 ± 0.03     1173 0.013 0.82 ± 0.02     1.16 ± 0.02     1173 0.014 0.81 ± 0.02     1.15 ± 0.03     1173 0.023 0.78 ± 0.02     1.31 ± 0.03     1332 0.011 0.80 ± 0.02     1.14 ± 0.02     1332 0.013 0.81±0.02     1.25 ± 0.03     1332 0.019 0.90 ± 0.02     1.17 ± 0.03     1332 0.021 0.88 ± 0.02     1.30 ± 0.03     References 1) Bjork C. W . Proc. 3rd Int. Conf. on Facility Operation Safeguards Interface, Nov. 29-Dec. 4, 1987, San-Diego, California, p. 129 2) Sprinkle J. K and Hsue S. T. Proc. 3rd Int. Conf. on Facility Operation Safeguards Interface, Nov. 29-Dec. 4, 1987, San-Diego, California, p.188. 3) R. J. Estep, Trans. Am. Nucl. 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