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[en] The development of advanced thorium-based nuclear system raises new requirements on nuclear data. The multi-group data file of critical nuclides in the thorium-uranium recycle is the foundation of physical design, analysis and calculation of the reactor core. Based on authoritative nuclear data processing code NJOY, this paper obtains a WIMS format multi-group cross section data files through processing the ENDF/B-VII.1 evaluation nuclear data file, uses the specific update maintenance procedure WILLIE to get a WIMS format data file, and conducts a series of critical benchmarks on the data file using the multi-group reactor core calculation code WIMSD5B. The results show that the computed results of the WIMS file based on the processing of ENDF/B-VII.1 are basically the same as those of the latest WIMS-D file published on the websites of the 'WIMS-D' library updating project (WLUP) with higher accuracy and reliability than those of the shipped WIMS-D file of the WIMSD5B code. Furthermore, the average deviation of the new WIMS file performing in the validation of 16 thorium-uranium cycle benchmarks is 0.225 3% smaller than that of the old WIMS file. (authors)
[en] The technique of evaluation of the multiplication factors and reactivity characteristics of fast reactors operating in the mode of multiple refueling is presented. We describe and apply the calculation method. The results demonstrate that fuel composition comes into equilibrium concentration in the multiple refueling reactor operation mode. If initial loads were based on plutonium from spent fuel of thermal and fast power reactors, equilibrium was achieved with twice repeated refueling. For initial fuel loads based on highly enriched uranium nitride or uranium-plutonium nitride fuel with high enrichment of 239Pu, equilibrium is reached after 4-5 refuelings
[en] The paper deals with the evaluation of fuel handling in fast reactors in the closed nuclear fuel cycle. To solve these problems, the REPRORYV program code has been developed. It simulates the nuclide streams in out-of-reactor stages of the closed fuel cycle. Existed verified code JARFR is used for the calculation of neutron-physical characteristics of the core. Various options for nuclide streams are considered with a representative full-scale model of a fast reactor with sodium-cooled high-power. The changes of multiplication factor and mass of plutonium in the closed nuclear fuel cycle for a Russian BN-type fast reactor with use of the REPRORYV code are evaluated. Different scenarios of fuel reprocessing assuming the removing or retaining of actinides, taking into account various plutonium losses on reprocessing steps were considered. The questions of influence of the initial content of plutonium in nitride fuel (Pu-U-N) and the impact of the initial parameters to the possibility of fuel self-sufficient mode of the reactor during the whole period of its operation are studied
[en] In this meeting we will continue the established tradition to inform each other about the most important achievements of both projects and results of activities identified on last meeting which include areas of Proliferation Resistance, Economics, Safety, and Non-electric applications. We are also expecting discussion of new areas of potential cooperation such as Modelling and Simulation, SMRs, Institutional innovations, Advanced fuel cycles including Thorium, advanced materials, and Education and Training.
[en] Summary: • Development of the next version of G4ECONS is progressing well; • Comparable results obtained from the benchmarking of G4ECONS, HEEP and H2A for hydrogen production; • Completed benchmarking G4ECONS with NEST in collaboration with IAEA: –Once-through SCWR; – Two fast reactors with closed fuel cycle.
[en] Experimental reactor physics is an essential element of physics design of a nuclear reactor and plays an important role in the safe design and operation of nuclear reactors. Approximations in modelling the reactor using computer codes and the ‘uncertainty in the nuclear data’ that goes as input into these codes contribute to the uncertainty of the theoretically computed design parameters. Reactor physics experiments provide estimates of the uncertainty in the design by comparing the measured and computed values of these parameters. A thorium fuel cycle based advanced heavy water reactor (AHWR) is being designed in Reactor Physics Design Division, BARC. A zero power critical facility (CF) was commissioned to generate the experimental data for physics design validation of AHWR. A number of experiments were carried out in CF which includes the measurement of differential/integral parameters and various reaction rates. The covariance analysis of these measurement will be carried out to generate the relevant covariance matrices
[en] Literature on the thorium fuel cycle spanning eight decades from the 1940s to the 2010s is identified, categorized, and analyzed. The publications are divided among twelve topical categories, and overall thorium literature trends are evaluated using database analysis techniques. In total, 1449 publications are identified, with the most prevalent topics being Reprocessing and Waste Management, Molten Salt Reactors, Fuels, and Light Water Reactors. In aggregate, reactor-oriented categories (five in all) comprise 45.5% of publications. The US is the most dominant thorium-publishing nation with 916 publications, followed next by India with 82 and then eight other countries having 25 publications or more. National laboratories have contributed 45% of thorium publications, with roughly equal shares of the balance split between government agencies, universities, and corporations/companies. Oak Ridge National Laboratory in the US accounts for more than a quarter of all publications. Specialized criteria are developed and applied to identify some of the most important, or “keystone”, publications for each category. Across the different categories, and for the study of thorium fuel cycles overall, published research reached an intermediate peak in the 1970s followed by a sharp decline in the 1980s and 1990s; however, interest has been revived moving into the 21st century.
[en] Versatile computational tools with up to date capabilities are needed to assess current nuclear fuel cycles or the transition from the current status of the fuel cycle to the more advanced and sustainable ones. The TR-EVOL module, that is devoted to fuel cycle mass balance, simulates diverse nuclear power plants (PWR, SFR, ADS, etc.), having possibly different types of fuels (UO_2, MOX, etc.), and the associated fuel cycle facilities (enrichment, fuel fabrication, processing, interim storage, waste storage, geological disposal). This work is intended to cross check the new capabilities of the fuel cycle scenario code TR-EVOL.This process has been divided in 2 stages. The first stage is dedicated to check the improvements in the nuclear fuel mass balance estimation using the available data for the Spanish nuclear fuel cycle. The second stage has been focused in verifying the validity of the TR-EVOL economic module, comparing results to data published by the ARCAS EU project. A specific analysis was required to evaluate the back-end cost. Data published by the waste management responsible institutions was used for the validation of the methodology. Results were highly satisfactory for both stages. In particular, the economic assessment provides a difference smaller than 3% regarding results published by the ARCAS project (NRG estimations). Furthermore, concerning the back-end cost, results are highly acceptable (7% difference for a final disposal in a once-through scenario and around 11% for a final disposal in a reprocessing strategy) given the significant uncertainties involved in design concepts and related unit costs. (authors)
[en] Conclusion: • GIF Cost Estimating Methodology is available for use by all GIF projects. • Training for users is available at varying levels of detail. • The EMWG continues to monitor Methodology applications. • Small scale activities on additional GEN IV related case studies.
[en] Authors discuss the issues of protection of fast reactors and relevant nuclear fuel cycles from the proliferation of nuclear weapons using knowledge, technology and materials of nuclear energy in military programs. The features of the closed nuclear fuel cycle of fast reactors to maintain the global nonproliferation regime in comparison with the non-closed cycle of thermal reactors are also discussed
[ru]В статье обсуждаются вопросы защищенности быстрых реакторов и соответствующих ядерных топливных циклов от распространения ядерного оружия за счет использования в военных программах знаний, технологий и материалов атомной энергетики. Обсуждаются также особенности замкнутого ядерного топливного цикла быстрых реакторов по поддержанию глобального режима нераспространения