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Nuclear Science Research Institute
JAEA-Review 2025-061, 183 Pages, 2026/03
JAEA-Review-2025-061.pdf:4.01MB
Nuclear Science Research Institute (NSRI) was composed of Planning and Management Department and six departments, namely Department of Operational Safety Administration, Department of Radiation Protection, Engineering Services Department, Department of Research Reactor and Tandem Accelerator, Department of Criticality and Hot Examination Technology, and Department of Decommissioning and Waste Management, and each department manages facilities and develops related technologies to achieve the "Medium- to Long-term Plan" successfully and effectively. On November 1, NSRI unified Department of Research Reactor and Tandem Accelerator, and Department of Criticality and Hot Examination Technology, newly organized Department of Research Infrastructure Technology Development. And, Planning and Management Department was reorganized to Promotion Office. Continuously, four research centers which are Advanced Science Research Center, Nuclear Science and Engineering Center, Nuclear Engineering Research Collaboration Center and Materials Sciences Research Center, belong to NSRI. In order to contribute to future research and development, and to promote management business, this annual report summarizes information on the activities of NSRI of JFY 2024 as well as the activity on research and development carried out by Collaborative Laboratories for Advanced Decommissioning Science, Nuclear Safety Research Center and activities of Nuclear Human Resource Development Center, using facilities of NSRI.
Fukuda, Kodai; Shiba, Shigeki*; Iwahashi, Daiki*; Gunji, Satoshi
Journal of Nuclear Science and Technology, 14 Pages, 2026/00
Times Cited Count：0Hamdani, A.; Soma, Shu; Abe, Satoshi; Shibamoto, Yasuteru
Nuclear Engineering and Technology, 9 Pages, 2026/00
Yoneyama, Kai; Nitta, Ayako; Tanaka, Yasuyuki; Kodaka, Noriyasu; Kikuchi, Riku; Sakano, Takuma; Furuse, Takahiro; Sato, Soichi; Sambongi, Mitsuru; Tanaka, Kosuke
JAEA-Technology 2025-008, 44 Pages, 2025/12
JAEA-Technology-2025-008.pdf:4.3MB
At the TEPCO's Fukushima Daiichi Nuclear Power Station (1F), an investigation inside the reactors has been carried out. In order to safely carry out the decommissioning work such as fuel debris retrieval and building demolition, it is important to estimate the contamination in primary containment vessel for control the decommissioning planning and the worker radiation exposure levels. Therefore, the analysis of the deposit inside the penetration for the 1F Unit 2 primary containment vessel ("X-6 penetration") was performed to clarify the components and activity. The smears from the deposit were used for the analysis. Non-destructive analysis such as gamma-ray spectrometry, X-ray Fluorescence (XRF) and Scanning Electron Microscope-Energy dispersive X-ray spectroscopy (SEM-EDX) for the smear-samples were performed to determine the gamma-nuclides and the morphology of elements in the deposit. Furthermore, in order to evaluate the nuclides and nuclide composition of the deposit in detail, the smear-samples were dissolved and the quantitative analysis of gamma-nuclides, Sr-90, alpha-nuclides in the dissolved solution were conducted. The results (non-destructive analysis and quantitative analysis) were compared with the results of samples collected at different locations in the X-6 penetration in 2020. In the gamma-ray spectrometry as non-destructive analysis where the smears were analyzed directly, Co-60, Sb-125, Cs-134, Cs-137, Eu-154, Eu-155 and Am-241 were detected. In XRF results, Fe originating from construction material was detected as a major element and small amount of U and Zr originating from the fuel and fuel cladding were also detected. In SEM-EDX results, O and Fe were found as a major element of the deposit and U particles coexisting with Fe, Si, Cr, Ni and Zr were also found. These results were consistent with the SEM-EDX results of the samples collected in 2020. In radioactivity analysis, quantitative values for gamma-nuclides (Co-60, Sb-125, Cs-134, Cs-137, Eu-154, Eu-155), Sr-90, Pu-238, Pu-239+240, Am-241, Cm-244, U-235 and U-238 were obtained. Using the results, the ratios of radioactivity based on Cs-137 and U-238 were calculated. Both sets of the ratios were compared to the calculated value of the Unit 2 fuel composition from ORIGEN.
Nakamura, Yuki*; Kojima, Yoshihiro*; Yamashita, Takuya; Shimomura, Kenta; Mizokami, Shinya
Journal of Nuclear Science and Technology, 62(12), p.1226 - 1230, 2025/12
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Yanagisawa, Hiroshi; Motome, Yuiko
JAEA-Research 2025-010, 197 Pages, 2025/11
JAEA-Research-2025-010.pdf:3.5MB
For understandings of nuclear criticality risks of TRIGA fuel rods and review of safety measures for handling them, nuclear criticality characteristics for infinite and finite heterogeneous lattice systems composed of the NSRR fuel rods were re-evaluated with the use of a detailed computational model for the fuel rod. The MVP version 3 code was used with the JENDL libraries including the latest version, JENDL-5, for the re-evaluation. As the criticality characteristics, variations of neutron multiplication factors of the infinite and water-reflected finite systems were examined in detail with parameters of the lattice pitch and density of moderator water. From the results of the re-evaluated criticality characteristics, the minimum critical number of fuel rods for the water-reflected hexagonal shaped lattice system was obtained to be 46.8 
0.2 using the JENDL-5 library. Moreover, the attainability of criticality without the water as moderator and reflector was examined because the zirconium hydride moderator and graphite reflector are equipped with the TRIGA fuel rod. It was found that the criticality is possible to be attained by 115.7 0.6 of the number of fuel rods, which is the smaller number of fuel rods than loaded in the NSRR standard core, even though no water exists.
Kawasaki, Nobuchika
JAEA-Review 2025-043, 74 Pages, 2025/10
JAEA-Review-2025-043.pdf:2.45MB
Russia is one of the most advanced countries in the civilian use of nuclear energy. However, understanding the internal mechanisms of its nuclear program remains difficult due to various reasons. Therefore, this study presents a historical overview of Russia's nuclear energy utilization, fuel supply, fuel manufacturing capabilities, and concepts regarding reprocessing and the nuclear fuel cycle. From this overview, insights have been extracted and analyzed. These insights are then organized under two strategic perspectives: "Strategic diversity and continuity in developments and demonstrations" and "Diversity in utilizations and deployments," with considerations of implications for Japan, as below. Russia's nuclear energy policy strategically utilizes a variety of reactor types and fuel cycle technologies to expand nuclear power generation both domestically and internationally. Currently, nuclear power, centered on light-water reactors (VVER series), accounts for about 20% of Russia's electricity supply, and there are plans to increase this share to 25% by 2045. A wide range of reactors, from large-scale to medium and small modular reactors, are being constructed in Russia. Russia is also actively developing fast reactor technologies, and focusing on the reprocessing and recycling of spent fuel. Internationally, VVER-1200 reactors are under construction in several countries, and cooperation with China is deepening in the field of fast reactors. Notably, Russia offers an integrated, or selectively customizable, package of nuclear technology services on the international stage. These include not only reactor deployment, but also fuel supply, reprocessing, waste management, and even the provision of radioisotopes. Rather than simply exporting products or technology, Russia fosters long-term relationships and trust by flexibly responding to the conditions and needs of partner countries. For this reason, Russia promotes the technology developments in advance within the country in areas anticipated for future overseas deployment. It carefully selects target technologies and services and systematically rolls them out. This flexible strategy, combining "technological diversity" and "strategic consistency", enables cooperation with countries across various geopolitical contexts. For Japan, this strategic approach offers valuable lessons on how to engage in comprehensive international nuclear cooperation, not merely through technology exports, but through integrated approaches that encompass the entire fuel cycle, and by combining elements such as fast reactors and RI supply.
Vauchy, R.; Horii, Yuta; Hirooka, Shun; Akashi, Masatoshi; Sunaoshi, Takeo*; Nakamichi, Shinya; Saito, Kosuke
Journal of Nuclear Materials, 616, p.156115_1 - 156115_16, 2025/10
Times Cited Count：1 Percentile：59.09(Materials Science, Multidisciplinary)Engineering Services Department, Nuclear Science Research Institute
JAEA-Review 2025-018, 83 Pages, 2025/09
JAEA-Review-2025-018.pdf:4.99MB
The Engineering Services Department is in charge of operation and maintenance of utility facilities (water distribution systems, electricity supply systems, steam generation systems and drain water systems etc.) in whole of the institute. And also is in charge of operation and maintenance of specific systems (power receiving and transforming facilities, an emergency electric power supply system, an air/liquid waste treatment system, a compressed air supply system) in nuclear reactor facilities, nuclear fuel material usage facilities and usual facilities or buildings. In addition, the department is in charge of maintenance of buildings, design and repair of electrical/mechanical equipment. This annual report describes summary of activities, operation and maintenance data and technical developments of the department carried out in JFY 2023. We hope that this report may help to future work.
Hamdani, A.; Soma, Shu; Abe, Satoshi; Shibamoto, Yasuteru
Progress in Nuclear Energy, 185, p.105771_1 - 105771_13, 2025/07
Times Cited Count：2 Percentile：80.51(Nuclear Science & Technology)Nuclear Science Research Institute
JAEA-Review 2024-058, 179 Pages, 2025/03
JAEA-Review-2024-058.pdf:7.42MB
Nuclear Science Research Institute (NSRI) is composed of Planning and Management Department and six departments, namely Department of Operational Safety Administration, Department of Radiation Protection, Engineering Services Department, Department of Research Reactor and Tandem Accelerator, Department of Criticality and Hot Examination Technology and Department of Decommissioning and Waste Management, and each department manages facilities and develops related technologies to achieve the "Medium- to Long-term Plan" successfully and effectively. And, four research centers which are Advanced Science Research Center, Nuclear Science and Engineering Center, Nuclear Engineering Research Collaboration Center and Materials Sciences Research Center, belong to NSRI. In order to contribute the future research and development and to promote management business, this annual report summarizes information on the activities of NSRI of JFY 2023 as well as the activity on research and development carried out by Collaborative Laboratories for Advanced Decommissioning Science, Nuclear Safety Research Center and activities of Nuclear Human Resource Development Center, using facilities of NSRI.
Nuclear Science Research Institute
JAEA-Review 2024-057, 178 Pages, 2025/03
JAEA-Review-2024-057.pdf:8.51MB
Nuclear Science Research Institute (NSRI) is composed of Planning and Management Department and six departments, namely Department of Operational Safety Administration, Department of Radiation Protection, Engineering Services Department, Department of Research Reactor and Tandem Accelerator, Department of Criticality and Hot Examination Technology and Department of Decommissioning and Waste Management, and each department manages facilities and develops related technologies to achieve the "Medium- to Long-term Plan" successfully and effectively. And, four research centers which are Advanced Science Research Center, Nuclear Science and Engineering Center, Nuclear Engineering Research Collaboration Center and Materials Sciences Research Center, belong to NSRI. In order to contribute the future research and development and to promote management business, this annual report summarizes information on the activities of NSRI of JFY 2022 as well as the activity on research and development carried out by Collaborative Laboratories for Advanced Decommissioning Science, Nuclear Safety Research Center and activities of Nuclear Human Resource Development Center, using facilities of NSRI.
Kaburagi, Masaaki; Miyamoto, Yuta; Mori, Norimasa; Iwai, Hiroki; Tezuka, Masashi; Kurosawa, Shunsuke*; Tagawa, Akihiro; Takasaki, Koji
Journal of Nuclear Science and Technology, 62(3), p.308 - 316, 2025/03
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Brumm, S.*; Gabrielli, F.*; Sanchez Espinoza, V.*; Stakhanova, A.*; Groudev, P.*; Petrova, P.*; Vryashkova, P.*; Ou, P.*; Zhang, W.*; Malkhasyan, A.*; et al.
Annals of Nuclear Energy, 211, p.110962_1 - 110962_16, 2025/02
Times Cited Count：12 Percentile：95.98(Nuclear Science & Technology)Katsumura, Kosuke*; Takagi, Junichi*; Miyahara, Naoya*; Uchida, Shunsuke*; Koma, Yoshikazu; Karasawa, Hidetoshi; Miwa, Shuhei; Satou, Yukihiko; Nagai, Haruyasu; Kurata, Masaki; et al.
Nihon Genshiryoku Gakkai-Shi ATOMO
, 67(2), p.128 - 132, 2025/02
no abstracts in English
Nagaya, Yasunobu
EPJ Nuclear Sciences & Technologies (Internet), 11, p.1_1 - 1_7, 2025/01
Japan Atomic Energy Agency (JAEA) has been developing a general-purpose continuous-energy Monte Carlo code MVP for nuclear reactor core analysis. Recently improvements to MVP have been focused on the development of an advanced neutronics/thermal-hydraulics coupling code. JAEA has also developed a new Monte Carlo solver Solomon for criticality safety analysis. Solomon aims to calculate the criticality of a damaged reactor core including fuel debris. This paper provides an overview of the capabilities and reviews recent applications of MVP and Solomon.
Calabrese, R.*; Hirooka, Shun
Progress in Nuclear Energy, 178, p.105516_1 - 105516_11, 2025/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Thermal creep is one of the key properties of mixed oxide (MOX) fuel for innovative fast reactors. Thermal creep of fuel affects markedly the interaction between the fuel and the cladding. A review of correlations available in the literature is presented. The effect of porosity, plutonium concentration, and stoichiometry are discussed also in the light of recent numerical results. Our analysis pointed out some inconsistencies concerning the modelling of the effect of porosity on diffusional creep and a re-evaluation of the effect of plutonium concentration. The discussion suggested that Evans's findings on the effect of stoichiometry should be better assessed as well as the level of increase in creep moving towards stoichiometry. Typical operating conditions of fast breeder reactors confirmed the need for an extension of porosity and temperature correlations' domains. Besides this, a new correlation based on a separate-effect approach has been proposed for fuel performance codes.
Seki, Akiyuki; Kondo, Yuki; Hashidate, Ryuta; Yoshikawa, Masanori; Yokoyama, Kenji; Takaya, Shigeru; Enuma, Yasuhiro; Hazama, Taira; Wakai, Takashi; Asayama, Tai
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 7 Pages, 2024/11
Yokoyama, Kenji; Hazama, Taira; Taninaka, Hiroshi; Oki, Shigeo
JAEA-Data/Code 2024-007, 41 Pages, 2024/10
JAEA-Data-Code-2024-007.pdf:1.1MB
The third version of the versatile reactor analysis code system, MARBLE3, has been developed. In the development of the former version of MARBLE, object-oriented scripting language Python (Python2) had been used and then the latest version of Python (Python3) was released. However, due to its backward incompatibility, MARBLE no longer worked with Python3. For this reason, MARBLE3 has been fully modified and maintained to work with Python3. In MARBLE3, newly developed analysis codes and newly proposed calculation methods were incorporated, and the user interface was extended and solvers were reimplemented for maintainability, extensibility, and flexibility. In MARBLE3, the three-dimensional hexagonal/triangular transport code MINISTRI Ver.7 (MINISTRI7) and the three-dimensional hexagonal/triangular diffusion code D-MINISTRI are available as the new analysis codes. These codes can be used in the neutronics analysis system SCHEME and the fast reactor burnup analysis system OPRHEUS, which are the subsystems of MARBLE. In addition, the user interface of CBG, a core analysis system embedded in MARBLE, was extended so that the diffusion and transport calculation solvers for the 2-dimensional RZ system of CBG can be used on SCHEME. On the other hand, MARBLE3 has extended the functionality of the burnup calculation solver so that it can use the numerical methods proposed in the papers on the improvement of the Chebyshev rational function approximation method and the minimax polynomial approximation method. From the viewpoint of maintainability, the point reactor kinetics solver POINTKINETICS, which was introduced in MARBLE2, has been newly reworked as the KINETICS solver in MARBLE3.
Pu
O
powder using master sintering curve theoryNakamichi, Shinya; Sunaoshi, Takeo*; Hirooka, Shun; Vauchy, R.; Murakami, Tatsutoshi
Journal of Nuclear Materials, 595, p.155072_1 - 155072_11, 2024/07
Times Cited Count：1 Percentile：23.55(Materials Science, Multidisciplinary)