Browsing by Author "Braoudakis, G"
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- ItemBenchmark consolidated results against experimental data from SPERT IV statics(International Atomic Energy Agency, 2019-08) Day, SE; Braoudakis, G; Wong, LThe IAEA CRP (CRP 1496) on ‘Benchmarking, against Experimental Data, of the Neutronic and Thermalhydraulic Computational Methods and Tools for Operation and Safety Analysis for Research Reactors’ provides a unique opportunity to benchmark and compare the accuracy and efficiency of both off-the-shelf and locally developed computational tools to a wide set of experimental research reactor benchmark analysis. In the scope of this project, various analysis groups have evaluated the SPERT IV benchmark analysis – consisting of a variety of commissioning experiments and multiple sets of Reactivity Insertion Accident (RIA) measurements. This report is focused on the commissioning experiments and associated measurements, referred to herein as the ‘Statics’ or neutronic section of the SPERT IV benchmark analysis. It summarizes and compares the analysis methodologies adopted, the code systems employed, and the simulation results generated by the various analysis groups. A comparison of the computational results to available experimental results is also provided in this report. © 2019 The Authors
- ItemCalculation of the core parameters measured during the commissioning of the OPAL Reactor(American Nuclear Society, 2010-05-12) Villarino, EA; Hergenreder, DF; Braoudakis, G; Ersez, TThe OPAL Research reactor is a multi-purpose open-pool type reactor. The nominal fission power of the reactor is 20 MW. It was commissioned during the second half of the year 2006. The reactor has several nuclear safety related design criteria that have to be experimentally verified during Stage B of the commissioning of the reactor. The present work presents the measurements carried out during the Stage B of the commissioning of the OPAL reactor, and the numerical verification of the calculated values using the design calculation methodology against these measured values. A brief description of the OPAL reactor, its commissioning plan, its nuclear safety related design criteria and the calculation and the experimental methodology are presented. The measured values and a comparison with the calculated is also given.
- ItemContract performance demonstration tests in the OPAL(International Atomic Energy Agency, 2007-11-08) Hergenreder, DF; Lecot, CA; Lovotti, O; Villarino, EA; Braoudakis, G; Ersez, TThis paper will describe the measurements and calculations that were done in the OPAL Reactor to demonstrate compliance against contractual Design Features and Performance Acceptance Criteria. The contract specifies several neutronic aspects to be fulfilled by the core, the irradiation and the beam facilities design, which have to be verified during the commissioning tests. Guaranteed flux values will be taken as being for equilibrium core conditions. The relationship between values measured during commissioning (First Core) and the guaranteed values is largely based on calculations. The calculated values are obtained modelling with full detail the measurement conditions using the INVAP traditional calculation lines: CITVAP and MCNP calculation lines. © The Authors
- ItemConversion of OPAL from U3Si2 to High Density U–Mo Fuel. Annex III(International Atomic Energy Agency, 2021-04) Braoudakis, G; Villarino, EAThe ongoing development of high density low enriched uranium (LEU) fuels to replace the current application of highly enriched uranium (HEU) fuels, in accordance with the non‑proliferation and security objectives of the Global Threat Reduction Initiative (GTRI), offers an opportunity to assess the impact of such fuels on performance and reactor core parameters. The performance of the current 4.8 gU/cm3 U3Si2 fuel in the Open Pool Australian Lightwater (OPAL) reactor is compared against that estimated for a variety of potential U–Mo high density fuels that range in density from 6.0 to 8.0 gU/cm3. The comparison includes cycle length, fuel usage, shutdown system performance, neutron fluxes in irradiation positions, kinetic parameters and reactivity feedback coefficients.High performance research reactors depend on a compact core design to maximize the neutron flux available for irradiation and beam facilities. The most readily available technology to achieve this goal is the use of HEU fuels. The extensive experience and demonstrable reliability and performance of such fuels makes them a clear choice for most of the high performance research reactors around the world. In addition, there are gains in fuel economy through maximization of fuel burnup and minimization of parasitic neutron absorption, which is present in some of the LEU fuels. In the spirit of GTRI, new high density LEU fuels are being developed with the intention of providing a non‑proliferation option while maintaining much of the performance expected from HEU fuels. One of the most promising high density LEU fuels is a uranium–molybdenum (U–Mo) alloy dispersion type fuel in an aluminium matrix. The addition of Mo stabilizes the uranium during irradiation and several different densities of uranium and alloy compositions are considered in this study, as no qualified fuel exists at this time.The current core design is optimized for the use of 4.8 gU/cm3 U3Si2 fuel. This study was performed while maintaining the core and fuel dimensions (specifically the fuel meat, fuel plate and coolant channel thicknesses) and only the fuel meat composition was modified. The estimated performance parameters were calculated using the same codes and methods that were used for the U3Si2 calculations. In this way a direct comparison can be made to assess the impact of using high density U–Mo fuel. © The Author
- ItemModelling of reusable target materials for the production of fission produced 99Mo using MCNP6.2 and CINDER90(Elsevier, 2021-10) Raposio, R; Braoudakis, G; Rosenfeld, AB; Thorogood, GJCurrent fission-based methods of 99Mo production require single use uranium targets which leads to spent uranium waste. This waste could be reduced if a target is developed that does not require dissolution so that it can be reused for multiple production runs. MCNP6.2 was used to model reusable targets of 20% and 1% enrichment for activity produced, target efficiency and burnup. The 1% enriched target was found to be much more efficient but had a lower activity produced compared to the 20% enriched target. The ideal target design for 99Mo production that optimises efficiency and reusability and reduces the self-shielding effect of UO2 was found to be a target that is made from 1% enriched UO2 with density as high as allowable for sufficient yields, efficient 99Mo extraction and having an irradiation time of 5 days, with the target able to be re-irradiated and re-processed 2–4 times. © 2021 Elsevier Ltd.
- ItemOn line neutron flux mapping in fuel coolant channels of a research reactor(IEEE, 2014-01-30) Barbot, L; Domergue, C; Villard, JF; Destouches, C; Braoudakis, G; Wassink, D; Sinclair, B; Osborn, JC; Wu, H; Blandin, C; Thévenin, M; Corre, G; Normand, SThis work deals with the on-line neutron flux mapping of the OPAL research reactor. A specific irradiation device has been set up to investigate fuel coolant channels using subminiature fission chambers to get thermal neutron flux profiles. Experimental results are compared to first neutronic calculations and show good agreement (C/E ~0.97). © 2022 IEEE
- ItemOPAL multicycle core benchmark using TRIPOLI4.10® and COCONEUT2.0(European Research Reactor Conference, RRFM, 2019-03-24) Decroocq, M; Bouret, C; Couybes, J; Privas, E; Gavoille, N; Koubbi, J; Manifacier, L; Braoudakis, GThe aim of the work performed by TechnicAtome is to benchmark its codes with data provided by ANSTO on the OPAL core. Calculation schemes used are Monte- Carlo TRIPOLI4.10® (with its depletion module) and deterministic COCONEUT2.0. After validating the models used on startup core configurations, we carry on the benchmarking process by performing the depletion calculations over cycles 7-13, as provided by ANSTO. Use of pre and post processing tools is highlighted, making the whole process easier to cross-check. COCONEUT2.0 in its homogeneous version and TRIPOLI4.10®, even though they are very different models, both show a similar increasing trend in the k-eff within each cycle, possibly due to an overestimation of cadmium wires burnup. On the other hand, COCONEUT2.0 in its semiheterogeneous version shows a rather flat k-eff behaviour within each cycle which could mean a better burnup calculation of both Cd and U-235. All three codes however show a trend between the cycles, consistent with that observed by other CRP participants, with an increase in the k-eff. Burnup calculations, both with COCONEUT2.0 and TRIPOLI4.10® are satisfactory.
- ItemOPAL nuclear reactor: experimental data(International Atomic Energy Agency, 2015-02) Braoudakis, GThe report provides technical details on the Open Pool Australian Light Water (OPAL) reactor core and immediate structure for analysis purposes. The goal of the report is to provide sufficient geometric and material data to build a computational neutronic model of the facility. © 2015 The Author
- ItemOPAL nuclear reactor: reactor specification(International Atomic Energy Agency, 2015-02) Braoudakis, GThe report provides technical details on the Open Pool Australian Light Water (OPAL) nuclear reactor core and immediate structure for analysis purposes. The goal of the report is to provide sufficient geometric and material data to build a computational neutronic model of the facility. © 2015 The Author
- ItemOPAL Reactor: calculation/experiment comparison of neutron flue mapping in fuel coolant channels(International Group On Research Reactors, 2013-10-13) Barbot, L; Domergue, C; Villard, JF; Destouches, C; Braoudakis, G; Wassink, D; Sinclair, B; Osborn, JC; Wu, HThe measurement and calculation of the neutron flux mapping of the OPAL research reactor are presented. Following an investigation of fuel coolant channels using sub-miniature fission chambers to measure thermal neutron flux profiles, neutronic calculations were performed. Comparison between calculation and measurement shows very good agreement. © The Authors
- ItemRadiation shielding design for neutron diffractometers assisted by Monte Carlo methods(Elsevier B. V., 2006-11-15) Osborn, JC; Ersez, T; Braoudakis, GMonte Carlo simulations may be used to model radiation shielding for neutron diffractometers. The use of the MCNP computer program to assess shielding for a diffractometer is discussed. A comparison is made of shielding requirements for radiation generated by several materials commonly used in neutron optical elements and beam stops, including lithium-6 based absorbers where the Monte Carlo method can model the effects of fast neutrons generated by this material. Crown copyright © 2006 Published by Elsevier B.V.
- ItemRadiation shielding design for neutron diffractometers assisted by Monte Carlo methods(The Bragg Institute, Australian Nuclear Science and Technology Organisation, 2005-11-27) Ersez, T; Braoudakis, G; Osborn, JCThe absorption and scattering of neutrons by neutron optical elements, beam stops and other components present significant radiation shielding challenges due to the generation of gamma radiation. In the case of neutron absorbers incorporating lithium- 6, fast neutrons are also generated. We show how Monte Carlo simulations using the MCNP computer code may be used to model the radiation fields produced by such components, thereby assisting in the choice of materials for shutters and other elements and assisting in the design of shielding. We discuss the use of these techniques to model instrument shielding bunkers, comprised principally of lead walls with boron-containing linings, for diffractometers at the OPAL Reactor, ANSTO. © 2005 The Authors
- ItemRadiation shielding for neutron guides(Elsevier B. V., 2006-11-15) Ersez, T; Braoudakis, G; Osborn, JCModels of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120 mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions. Crown Copyright © 2006 Published by Elsevier B.V.
- ItemRadiation shielding for neutron guides(The Institution of Engineers Australia, 2005-11-27) Ersez, T; Braoudakis, G; Osborn, JCModels of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions. © The Authors
- ItemTAIPAN – a spectrometer for inelastic neutron scattering at the replacement research reactor(Australian Institute of Physics, 2004-02-04) Hagen, ME; Horton, G; Moore, R; Braoudakis, G; Cussen, LDInelastic neutron scattering is widely used to study the lattice vibrational (phonon) and magnetic (spin wave and crystal field) excitations in condensed matter. In order to characterise such excitations comprehensively measurements on single crystal specimens are required and the most appropriate instrument for doing this at a steady state (reactor) neutron source is a three-axis spectrometer (TAS). We will describe the characteristics of the three axis spectrometer TAIPAN, which is currently under construction at the replacement research reactor, ANSTO and which will be available to the Australian scientific community in 2006.
- ItemThermal triple-axis spectometer at OPAL Reactor(Australian Institute of Physics, 2006-02-07) Danilkin, SA; Horton, G; Moore, R; Braoudakis, G; Hagen, MEInelastic neutron scattering is widely used to study the excitations such as phonon and magnon in condensed matter. A triple–axis spectrometer (TAS) is one of main instruments used in neutron scattering studies. TAS TAIPAN will be the first inelastic instrument at the Australian research reactor OPAL. It will be placed at a reactor face position and will use double focusing monochromators and analyser. In addition to the double-focusing regime, TAIPAN will have a standard mode of operation with Soller collimators providing high resolution. The instrument will use supermirror benders for polarization analysis.