Fuel debris criticality analysis technology using non-contact measurement method (Contract research); FY2021 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Tokyo Institute of Technology*

JAEA-Review 2022-043, 52 Pages, 2023/01

JAEA-Review-2022-043.pdf:3.48MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2021. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2021, this report summarizes the research results of the "Fuel debris criticality analysis technology using non-contact measurement method" conducted in FY2021. The purpose of research was to improve the fuel debris criticality analysis technology using non-contact measurement method by the development of the fuel debris criticality characteristics measurement system and the multi-region integral kinetic analysis code. It was performed by Tokyo Institute of Technology (Tokyo Tech), National Institute of Advanced Industrial Science and Technology (AIST), and National Research Nuclear University (MEPhI) as the first year of four years research project. For the criticality characteristic measurement systems to be developed by the Japanese and Russian sides, …

Sodium fire collaborative study progress; CNWG fiscal year 2022

Louie, D. L. Y.*; Aoyagi, Mitsuhiro

SAND2022-14235 (Internet), 29 Pages, 2022/10

https://doi.org/10.2172/1892376

This report discusses the progress on the collaboration between Sandia National Laboratories (SNL) and Japan Atomic Energy Agency (JAEA) on the sodium fire research in fiscal year 2022. First, the current sodium pool fire model in MELCOR is discussed. The associated sodium fire input requirements are also presented. The theoretical pool fire model improvement developed at SNL is discussed. A control function model has been developed from this improvement. Then, the validation study of the sodium pool fire model in MELCOR is described. To validate this pool fire model with the enhancement, JAEA F7-1 and F7-2 sodium pool fire experiments are used. The results of the calculation are discussed as well as suggestions for further model improvement. Finally, recommendations are made for new MELCOR simulations for next fiscal year, 2023.

Summaries of research and development activities by using supercomputer system of JAEA in FY2020 (April 1, 2020 - March 31, 2021)

HPC Technology Promotion Office

JAEA-Review 2021-022, 187 Pages, 2022/01

JAEA-Review-2021-022.pdf:10.11MB

Japan Atomic Energy Agency (JAEA) conducts research and development (R&D) in various fields related to nuclear power as a comprehensive institution of nuclear energy R&Ds, and utilizes computational science and technology in many activities. As shown in the fact that about 20 percent of papers published by JAEA are concerned with R&D using computational science, the supercomputer system of JAEA has become an important infrastructure to support computational science and technology. In FY2020, the system was used for R&D aiming to restore Fukushima (environmental recovery and nuclear installation decommissioning) as a priority issue, as well as for JAEA's major projects such as research and development of fast reactor cycle technology, research for safety improvement in the field of nuclear energy, and basic nuclear science and engineering research. This report presents a great number of R&D results accomplished by using the system in FY2020, as well as user support, operational records and overviews of the system, and so on.

Summaries of research and development activities by using supercomputer system of JAEA in FY2019 (April 1, 2019 - March 31, 2020)

HPC Technology Promotion Office

JAEA-Review 2020-021, 215 Pages, 2021/02

JAEA-Review-2020-021.pdf:13.11MB

Japan Atomic Energy Agency (JAEA) conducts research and development (R&D) in various fields related to nuclear power as a comprehensive institution of nuclear energy R&Ds, and utilizes computational science and technology in many activities. As shown in the fact that about 20 percent of papers published by JAEA are concerned with R&D using computational science, the supercomputer system of JAEA has become an important infrastructure to support computational science and technology. In FY2019, the system was used for R&D aiming to restore Fukushima (environmental recovery and nuclear installation decommissioning) as a priority issue, as well as for JAEA's major projects such as research and development of fast reactor cycle technology, research for safety improvement in the field of nuclear energy, and basic nuclear science and engineering research. This report presents a great number of R&D results accomplished by using the system in FY2019, as well as user support, operational records and overviews of the system, and so on.

Pressure resistance thickness of disposal containers for spent fuel direct disposal

Sugita, Yutaka; Taniguchi, Naoki; Makino, Hitoshi; Kanamaru, Shinichiro*; Okumura, Taisei*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(3), p.121 - 135, 2020/09

https://doi.org/10.3327/taesj.J19.007

A series of structural analysis of disposal containers for direct disposal of spent fuel was carried out to provide preliminary estimates of the required pressure resistance thickness of the disposal container. Disposal containers were designed to contain either 2, 3 or 4 spent fuel assemblies in linear, triangular or square arrangements, respectively. The required pressure resistance thickness was evaluated using separation distance of the housing space for each spent fuel assembly as a key model parameter to obtain the required thickness of the body and then the lid of the disposal container. This work also provides additional analytical technical knowledge, such as the validity of the setting of the stress evaluation line and the effect of the model length on the analysis. These can then be referred to and used again in the future as a basis for conducting similar evaluations under different conditions or proceeding with more detailed evaluations.

Summaries of research and development activities by using supercomputer system of JAEA in FY2018 (April 1, 2018 - March 31, 2019)

HPC Technology Promotion Office

JAEA-Review 2019-017, 182 Pages, 2020/01

JAEA-Review-2019-017.pdf:11.11MB

Japan Atomic Energy Agency (JAEA) conducts research and development (R&D) in various fields related to nuclear power as a comprehensive institution of nuclear energy R&Ds, and utilizes computational science and technology in many activities. As shown in the fact that about 20 percent of papers published by JAEA are concerned with R&D using computational science, the supercomputer system of JAEA has become an important infrastructure to support computational science and technology. In FY2018, the system was used for R&D aiming to restore Fukushima (environmental recovery and nuclear installation decommissioning) as a priority issue, as well as for JAEA's major projects such as research and development of fast reactor cycle technology, research for safety improvement in the field of nuclear energy, and basic nuclear science and engineering research. This report presents a great number of R&D results accomplished by using the system in FY2018, as well as user support, operational records and overviews of the system, and so on.

Sodium fire collaborative study progress; CNWG fiscal year 2019

Louie, D. L. Y.*; Uchibori, Akihiro

SAND2019-15043 (Internet), 35 Pages, 2019/12

https://doi.org/10.2172/1592872

This report describes the progress on the sodium fire research in fiscal year 2019 in the Civil Nuclear Energy Research and Development Working Group (CNWG). In this study, the validation study of the sodium pool fire model incorporated into the MELCOR code, which was originally developed for accident analysis in light water reactors, was carried out through the numerical analysis on the sodium pool fire experiment named F7-1. In this preliminary analysis, pool and atmosphere temperature went up to the same level with the measured results, while the unnatural behavior appeared in the latter half of the analysis. Based on this result, recommendations for improvement were made for a new analysis in next fiscal year, 2020.

Development of numerical analysis method for tube failure propagation under sodium-water reaction accident

Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11

Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium.

Three-dimensional numerical study on pool stratification behavior in molten corium-concrete interaction (MCCI) with MPS method

Li, X.; Sato, Ikken; Yamaji, Akifumi*; Duan, G.*

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07

Molten corium-concrete interaction (MCCI) is an important ex-vessel phenomenon that could happen during the late phase of a hypothetical severe accident in a light water reactor. In the present study, a three-dimensional (3-D) numerical study has been performed to simulate COMET-L3 test carried out by KIT with a stratified molten pool configuration of simulant materials with improved MPS method. The heat transfer between corium/crust/concrete was modeled with heat conduction between particles. Moreover, the potential influence of the siliceous aggregates was also investigated by setting up two different case studies since there was previous study indicating that siliceous aggregates in siliceous concrete might contribute to different axial and radial concrete ablation rates. The simulation results have indicated that metal melt as corium in MCCI can have completely different characteristics regarding concrete ablation pattern from that of oxidic corium, which needs to be taken into consideration when assessing the containment melt-through time in severe accident management.

A Simple and practical correction technique for reactivity worth of short-sized samples measured by critical-water-level method

Kitamura, Yasunori*; Fukushima, Masahiro

Nuclear Science and Engineering, 186(2), p.168 - 179, 2017/05

https://doi.org/10.1080/00295639.2016.1273017

An inconsistency between the reactivity worth of short-size samples measured by the critical-water-level (CWL) method and that conventionally analysed for validating the nuclear data and the nuclear calculation methods has been known. The present study investigated this inconsistency in terms of a simple theoretical framework and proposed a simple and practical technique for correcting the measured sample reactivity worth without making supplementary experiments. A series of Monte Carlo calculations that simulated typical sample reactivity worth measurement by the CWL method showed that this inconsistency is effectively reduced by the present correction technique.