2023, Chereshkov, D. G., Тихомиров, Георгий Валентинович, Терновых, Михаил Юрьевич, Рыжков, Александр Александрович, Tikhomirov, G. V., Ternovykh, M. Y., Ryzhkov, A. A.

The new calculation code capabilities are applied in the current work as well as important fast reactor criticality parameters uncertainty assessment articles’ results based on different nuclear data libraries and covariance matrices. A comparative analysis of uncertainty estimations related to neutron reactions is presented for lead-cooled reactor models and sodium-cooled reactor models. For the models of advanced BN and BR fast reactors with three fuel types (UO2, MOX, MNUP), the multiplication factor uncertainty calculations are performed using 252-group covariance matrices based on ENDF/B-VII.1 library via the SCALE 6.2.4 code system. The main nuclear data uncertainty contributors in the multiplication factor are determined. Recommendations are formulated for improving the cross sections accuracy for several nuclides in order to provide more reliable results of fast reactor criticality calculations. Lead-cooled reactors have no operational history compared to light-water and sodium-cooled reactors. The experimental data insufficiency calls in the question about reliability of the simulation results and requires a comprehensive initial data uncertainty analysis for the neutron transport simulation. The obtained results support the idea that lead- and sodium-cooled reactors have close nuclear data sensitivity using one and the same computation tools, nuclear data libraries and fuel compositions. This makes it possible to use the accumulated data of benchmarks for sodium-cooled reactors in the safety determination of lead-cooled reactors.

2020, Pugachev, P. A., Saldikov, I., Takezawa, H. , Muramoto, T., Nishiyama, J. , Obara, T., Богданова, Екатерина Владимировна, Терновых, Михаил Юрьевич, Тихомиров, Георгий Валентинович, Смирнов, Антон Дмитриевич, Tikhomirov, G. V., Ternovykh, M. Y., Bogdanova, E. V., Smirnov, A. D.

Obninsk Institute for Nuclear Power Engineering, National Research Nuclear University 'MEPhI'. All rights reserved.After the accident at the Fukushima Daiichi nuclear power station, the attention of the scientific community is riveted on how the consequences are being eliminated. Removing corium - a resolidified mixture of nuclear fuel with other structural elements of the reactor - remains the most difficult task, the solution of which can take several decades. It is extremely important to exclude the occurrence of any emergency processes during the removal of corium. The purpose of this work was to solve a coordinated hydrodynamic and neutron-physical problem characterized by a large number of randomly oriented and irregularly located corium particles in water as part of the development of a benchmark for this class of problems. Monte Carlo- based precision codes were used to perform a neutronic analysis. The positions of particles with corium were obtained from the results of numerical simulation. The analysis results obtained using the codes involved showed good consistency for all the states considered. It was shown that modern neutronic codes based on the Monte Carlo method successfully cope with the geometric formation and solution of the problem with a nontrivial distribution of corium particles in water. The results of the study can be used to justify the safety of corium handling procedures, including its extraction from a damaged power unit.

2023, Рыжков, Александр Александрович, Ryzhkov, A. A., Ternovykh, M. Y., Tikhomirov, G. V., Терновых, Михаил Юрьевич, Тихомиров, Георгий Валентинович

Nuclear data are a main uncertainty source in neutron transport simulations making their consideration in reactor safety necessary. This arises anew with the state-of-the-art reactors known as Generation IV. Some of the reactors suggests providing the reactivity margin below the effective delayed neutron fraction excluding prompt criticality accidents, and the breeding ratio is the key factor in this. Consequently, assessing a degree of the breeding ratio accuracy is of interest. Therefore, in this work, the breeding ratio uncertainties are analyzed by performing a sensitivity and uncertainty analysis of the MET1000 and MOX3600 models with respect to nuclear data using SCALE. As a result, the breeding ratio uncertainties are obtained approximately equal to 2% as the main contributors are 239Pu(n, γ), 238U(n, γ), and 238U(n, n’ ). The uncertainty sources between the models are compared, and 16O preponderantly increases the total uncertainty not directly by its uncertainty but by its impact on the spectrum.

2023, Черешков, Д. Г., Тихомиров, Георгий Валентинович, Терновых, Михаил Юрьевич, Рыжков, Александр Александрович

В работе использованы новые возможности расчётных кодов и результаты публикаций по оценке неопределённостей важнейших нейтронно-физических характеристик реакторов на быстрых нейтронах на основе библиотек ядерных данных и ковариационных матриц. Представлен сравнительный анализ оценок, связанных с нейтронными реакциями, на моделях реакторов со свинцовым теплоносителем и натриевым теплоносителями. Для моделей перспективных быстрых реакторов типа БР и БН с тремя видами топлива (диоксид урана, МОКС и СНУП) выполнены расчёты неопределённостей коэффициента размножения на основе групповых ковариационных матриц библиотеки ENDF/B-VII.1 в программном коде SCALE 6.2.4. Определены основные источники неопределённостей коэффициента размножения. Сформулированы рекомендации по повышению точности сечений нуклидов для обеспечения более надежного расчёта критичности быстрых реакторов. У реакторов со свинцовым теплоносителем отсутствует столь же значительный опыт эксплуатации установки по сравнению с легководными и натриевыми реакторами. Недостаточность экспериментальных данных ставит под сомнение достоверность результатов расчётного моделирования и требует всестороннего анализа неопределённости исходных данных при моделировании. В работе полученными результатами поддерживается утверждение, что у свинцовых и натриевых реакторов чувствительность к ядерным данным близка при использовании одинаковых расчётных инструментов, библиотек данных и топливных композиций. Это позволяет использовать в обоснование безопасности свинцовых реакторов накопленные бенчмарки по натриевым реакторам.