Browsing by Author "Finlay, MR"

Now showing 1 - 7 of 7
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    Australian research reactors spent fuel management: the path to sustainability
    (European Nuclear Society, 2016-03-13) Finlay, MR; Miller, R; Dimitrovski, L; Domingo, X; Landau, P; Valery, J; Laloy, V
    Since the late 1950’s, ANSTO has successfully operated three research reactors in Australia: HIFAR (1958-2007), MOATA (1961-1995) and OPAL (2006- Specific strategies were developed and implemented for the management and disposition of spent fuel from HIFAR and MOATA. They included strategic considerations, technical options, fuel characteristics, storage capacity, operational constraints and associated implications. In addition, the operating licenses of the Australian reactors have required the identification of spent fuel disposition arrangements, i.e. the “deferment” strategy of storage indefinitely is not acceptable. Disposition then employed three routes with direct disposal in the USA under the US-DOE FRRSNFA Program and reprocessing in France by AREVA, and in the UK by the UKAEA. Both reprocessing routes included return of vitrified waste. ANSTO and AREVA have worked together since the late 1990’s on the disposition of uranium aluminide (UAlx) spent fuel from HIFAR. Today, ANSTO is committed to develop a lifetime strategy for management and disposition of uranium silicide (U3Si2) spent fuel from OPAL. AREVA’s ability to offer an integrated solution for storage, transport, reprocessing, waste return and long-term management, including addressing individual customer needs (type of fuel, timelines, quantities, final waste management strategy,...), has provided ANSTO with a viable spent fuel management strategy, for OPAL’s lifetime.
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    Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding
    (Elsevier, 2016-10-01) Pasqualini, EE; Robinson, AB; Porter, DL; Wachs, DM; Finlay, MR
    Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U–(7–10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry–4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction—either from fabrication or in-reactor testing—and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak). © 2016 Elsevier B.V.
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    Microstructural development in an irradiated monolithic LEU U-Mo fuel plate
    (American Nuclear Society, 2010-06-16) Keiser, DD; Jue, JF; Robinson, AB; Medvedev, P; Wachs, M; Finlay, MR
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    Microstructural development in irradiated U-7Mo/6061 Al alloy matrix dispersion fuel
    (Elsevier, 2009-09-01) Keiser, DD; Robinson, AB; Jue, JF; Medvedev, P; Wachs, DM; Finlay, MR
    A U-7Mo alloy/6061 Al alloy matrix mini-dispersion fuel plate was irradiated in the Advanced Test Reactor and then examined using optical metallography and scanning electron microscopy to characterize the developed microstructure. Results were compared to the microstructure of the as-fabricated dispersion fuel to identify changes that occurred during irradiation. The layer that formed on the surface of the fuel U-7Mo particles during fuel plate fabrication exhibits stable irradiation performance as a result of the 0.88 wt% Si present in the fuel meat matrix. During irradiation, the pre-formed interaction layer changed very little in thickness and composition. The overall irradiation performance of the fuel plate to moderate power and burnup was considered excellent. © 2009, Elsevier Ltd.
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    OPAL spent fuel management: status in 2019
    (European Research Reactor Conference, RRFM, 2019-03-24) Finlay, MR; Healy, M; Naidoo-Amegilo, P; Valery, J; Halle, L
    Since the late 1950’s, ANSTO has successfully operated three research reactors in Australia: HIFAR (1958-2007), MOATA (1961-1995) and OPAL (2006-present). ANSTO has demonstrated the safe and secure management of the spent fuel from those reactors. MOATA and HIFAR spent fuel was either reprocessed offshore in the UK or France and the vitrified waste returned to Australia or returned to the USA under the FRRSNFA Program. A strategy for the management of OPAL spent fuel was developed before construction started and has evolved since then. The strategy of spent fuel disposition is now established as offshore reprocessing and management of the returned vitrified wastes. To manage the large inventory of spent OPAL fuel generated, ANSTO has entered into a long term contract with Orano for the transportation and reprocessing of the OPAL spent fuel, with provisions included for the return of vitrified waste. The first transport of OPAL SNF to La Hague in France was performed in July 2018. Further transports are scheduled to be conducted at intervals of 6-7 years. This paper will provide an overview of the management of spent fuel in Australia and cover the preparation that facilitated the first transport of OPAL SNF to La Hague in 2018. It will address aspects such as processing considerations, regulatory and governmental approvals, operational planning and execution.
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    Post irradiation examination of monolithic mini-fuel plates from RERTR-6 and 7
    (European Nuclear Society, 2007-03-11) Finlay, MR; Wachs, DM; Robinson, AB; Hofman, GL
    Successful qualification of the monolithic fuel is required for the conversion of the high performance research reactors and significant effort is being devoted to its development. The RERTR-6 experiment was designed to irradiate the first monolithic fuel mini-plates at moderate power density to moderate burn-up. The follow-on experiment, RERTR-7, aimed to irradiate monolithic fuel mini-plates at very high power density to high burn-up. It contained monolithic mini-plates fabricated by friction stir welding (FSW) and transient liquid phase bonding (TLPB). The post-irradiation of RERTR-7 indicates that porosity formation is occurring at the interface between the foil and the cladding in FSW plates. No porosity was observed in TLPB plates. The observations of delamination of FSW plates correlate with the initial mechanical testing results. A number of developments are being pursued on the fabrication front to address some of the observations and should be implemented in RERTR-9.
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    Weld repair of creep damaged steels
    (New Zealand Welding Committee, 1996-02-04) Croker, ABL; Finlay, MR; Law, M
    This paper reviews a current three year Cooperative Research Centre project titled "Welding of Thermally Modified Structures" and describes some of the results achieved. The project was commenced in June 1993 with support from ANSTO, CSIRO, BHP, University of Wollongong and the CRC for Materials, Welding and Joining with the aim of quantifying the effects of performing repair welds on materials which have operated for extended periods at elevated temperature. Welding is increasingly used for performing repairs, replacements, retrofits and modifications to elevated temperature plant, however, the effects of these repairs on the ultimate life of a component are poorly understood. Details are presented of the three ex-service materials chosen for the project; a carbon steel and two low alloy steels. Initial results are also reported on the characterisation of repair welds by microscopy, toughness and creep testing and the use of finite element creep modelling to predict the behaviour of it repair welded joints during high temperature service.