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Center for Computational Science & e-Systems
JAEA-Evaluation 2026-003, 41 Pages, 2026/06
JAEA-Evaluation-2026-003.pdf:1.08MB
Research on advanced computational science for nuclear applications, based on "the plan to achieve the medium- and long-term goal of the Japan Atomic Energy Agency", has been performed by Center for Computational Science & e-Systems (CCSE), Japan Atomic Energy Agency. CCSE established a committee consisting of external experts and authorities which evaluates and advises toward the future research and development. This report summarizes the results of the R&D performed by CCSE in FY2024 (April 1st, 2024 - March 31st, 2025) and their evaluation by the committee.
-containing porous hydrogel via freeze-crosslinking for efficiency and salt-robust dye DecolorizationSugita, Tsuyoshi; Ueda, Yuki; Nakabe, Rintaro; Mori, Masanobu*; Nankawa, Takuya; Sekine, Yurina
Journal of Photochemistry and Photobiology A; Chemistry, 473, p.116773_1 - 116773_9, 2026/04
Times Cited Count：0 Percentile：0.00(Chemistry, Physical)We developed a WO-embedded hydrogel (WFG) by freeze-cross-linking that retained high activity even in the presence of coexisting salts. Confocal laser scanning microscope revealed interconnected channels < 200 
m. ensuring good water permeability, and contrast-matching small-angle neutron scattering showed that the secondary particle size of embedded WO (~300 nm) matched that in aqueous suspension. Under visible-light irradiation, WFG decolorized indigo carmine (INC) 1.5-fold increase in rate than suspended WO and 3.7-fold increase in rate than a WO-coated glass plate. Coexisting salts (NaNO, NaCl, Na
SO
, NaHPO) altered the decolorization efficiency; NaNO and NaSO enhanced, whereas Cl
and HPO suppressed the reaction, indicating that ionic strength and anion-species affect contact efficiency and charge transfer.
Kawazu, Ryohei
JAEA-Technology 2025-014, 48 Pages, 2026/02
JAEA-Technology-2025-014.pdf:2.19MB
The Japan Atomic Energy Agency (JAEA) conducts research and development in various fields related to nuclear energy as a comprehensive research and development organization for nuclear power. Computational science and technology are utilized in many of these research and development activities. The supercomputer system HPE SGI8600 (hereinafter referred to as the "supercomputer") was introduced in December 2020 as critical infrastructure to meet the increasing computational demands driven by advancements in technologies such as digital twins, machine learning, and big data processing. It has become indispensable for promoting research and development at JAEA. Improving the efficiency of job operations and program waiting times (hereinafter referred to as "job waiting times") on the supercomputer, which is an essential infrastructure supporting JAEA's computational science and technology, is useful for enhancing research and development efficiency. This report presents the results of the investigation into the changes in job waiting times following the integration of queue classes, which was implemented in fiscal year 2022 to efficiently utilize computational resources. It summarizes the process from the analysis of the supercomputer's usage information to the improvements made for the integration of queue classes and the improvement of job waiting times.
HPC Technology Promotion Office, Center for Computational Science & e-Systems
JAEA-Review 2025-044, 140 Pages, 2026/01
JAEA-Review-2025-044.pdf:8.77MB
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. Over the past 10 years or so, the publication of papers utilizing computational science and technology at JAEA has accounted for about 20 percent of the total publications each fiscal year. The supercomputer system of JAEA has become an important infrastructure to support computational science and technology. In FY2024, the system was utilized in R&D activities that were prioritized in the Fourth Medium- to Long-Term Plan, including contributing to carbon neutrality through the development of innovative technologies for improving safety, creating innovation by promoting diverse R&D related to nuclear science and technology, promoting R&D in response to the accident at TEPCO's Fukushima Daiichi Nuclear Power Station, steadily implementing technological developments for the treatment and disposal of high-level radioactive waste, and supporting nuclear safety regulatory administration and nuclear disaster prevention by promoting safety research for these purposes. This report presents a great number of R&D results accomplished by using the system in FY2024, as well as user support, operational records and overviews of the system, and so on.
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.
Irisawa, Eriko; Kato, Chiaki
Corrosion Science, 256, p.113173_1 - 113173_16, 2025/11
Times Cited Count：6 Percentile：77.29(Materials Science, Multidisciplinary)Onoda, Yuichi; Uchita, Masato*; Tokizaki, Minako*; Okazaki, Hitoshi*
Nuclear Technology, 211(11), p.2812 - 2831, 2025/11
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Collaborative Laboratories for Advanced Decommissioning Science; The University of Tokyo*
JAEA-Review 2025-014, 86 Pages, 2025/09
JAEA-Review-2025-014.pdf:9.38MB
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 FY2023. 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 "Novel mechanical manipulator for efficient fuel debris retrieval" conducted from FY2021 to FY2023. The present study aims to the development of a collision-tolerant robotic manipulator with the mechanical variable impedance actuators in an unknown environment. Another research target is the system architecture of an artificial intelligence-based control method for efficient exploration and decommissioning. In addition to conducting an investigation in the area deep inside the aperture, which has been difficult with conventional investigations, we aim to recover pebble-shaped fuel debris at the bottom of the pedestal using a gripper at the tip of the manipulator. We will establish a manipulator mechanism and remote-control system to cope with the environmental constraints inside the pedestal. This year focused on developing a manipulator navigation algorithm, evaluating control performance, and conducting on-site scenario demonstration experiments. Studies included building a driving model, designing a control algorithm, and assessing manipulator control performance, with comparisons to simulation models. Demonstration experiments were conducted at facilities like the JAEA Naraha Center for Remote Control Technology Development. The research was promoted in close collaboration with the UK team and external advisors, including bi-weekly meetings. At the end of the fiscal year, an oral presentation introducing the project was given at an international workshop organized by the University of Tokyo and Technical University of Munich (TUM).
Ishikawa, Akihisa; Segawa, Mariko; Toh, Yosuke; Watanabe, Kenichi*; Masuda, Akihiko*; Matsumoto, Tetsuro*; Yamazaki, Atsushi*; Yoshihashi, Sachiko*; Uritani, Akira*; Harano, Hideki*
Journal of Radiation Research (Internet), 66(5), p.563 - 569, 2025/09
Times Cited Count：1 Percentile：41.12(Biology)Takeshita, Kenji*; Okamura, Tomohiro*; Nakase, Masahiko*; Abe, Takumi; Nishihara, Kenji
Progress in Nuclear Science and Technology (Internet), 8, p.52 - 57, 2025/09
Motegi, Kosuke; Matsumoto, Toshinori; Shiotsu, Hiroyuki
Proceedings of 21st International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-21) (Internet), 10 Pages, 2025/08
Yanagisawa, Hiroshi; Motome, Yuiko
JAEA-Research 2025-001, 99 Pages, 2025/06
JAEA-Research-2025-001.pdf:1.98MB
The detailed computational models for nuclear criticality analyses on the first startup cores of NSRR (Nuclear Safety Research Reactor), which is categorized as a TRIGA-ACPR (Annular Core Pulse Reactor), were created for the purposes of deeper understandings of safety inspection data on the neutron absorber rod worths of reactivity and improvement of determination technique of the reactivity worths. The uncertainties in effective neutron multiplication factor (k
) propagated from errors in the geometry, material, and operation data for the present models were evaluated in detail by using the MVP version 3 code with the latest Japanese nuclear data library, JENDL-5, and the previous versions of JENDL libraries. As a result, the overall uncertainties in k for the present models were evaluated to be in the range of 0.0027 to 0.0029 
k. It is expected that the present models will be utilized as the benchmark on k for TRIGA-ACPR. Moreover, it is confirmed that the overall uncertainties were sufficiently smaller than the values of absorber rod worths determined in NSRR. Thus, it is also considered that the present models are applicable to further analyses on the absorber rod worths in NSRR.
Takeda, Takeshi
JAEA-Data/Code 2025-005, 106 Pages, 2025/06
JAEA-Data-Code-2025-005.pdf:2.93MB
JAEA has been creating input data for pressurized water reactor (PWR) analysis with RELAP5/MOD3.3 code, mainly based on design information for the four-loop PWR's Tsuruga Power Station Unit-2 as the reference reactor of the Large Scale Test Facility (LSTF). The cold leg large-break loss-of-coolant accident (LBLOCA) calculation in the flamework of the BEMUSE program is cited as a representative OECD/NEA activity related to the PWR analysis. The new regulatory requirements for PWRs in Japan include the event of loss of recirculation functions from emergency core cooling system (ECCS) in the cold leg LBLOCA. This event should be evaluated the effectiveness of measures against severe core damage. The input data for this study were made preparations to analyze the PWR LBLOCA, which is one of the design basis accidents that should be postulated in the safety design. This report describes the main features of the input data for the PWR LBLOCA analysis. The PWR model comprised a reactor vessel, pressurizer (PZR), hot legs, steam generators (SGs), SG secondary-side system, crossover legs, cold legs, and ECCS. A four-loop PWR was simulated by two loops in the LBLOCA calculation. Specifically, loop-A attached with the PZR corresponded to three loops, and loop-B mounted with the break was equal to one loop. The nodalization schemes of the PWR components were referred to those of the LSTF components. Moreover, interpretations were added to the main input data for the PWR LBLOCA analysis, and further information such as the basis for determining the input data was provided. In addition, transient analysis was performed employing the prepared input data for the loss of ECCS recirculation functions event. The present transient analysis was confirmed to be appropriate generally by comparing with the calculation in the previous study using the RELAP5/MOD3.3 code. Furthermore, sensitivity analyses were executed exploiting the RELAP5/MOD3.3 code to clarify the effects of a discharge coefficient through the break and water injection flow rate of the alternative recirculation on the fuel rod cladding surface temperature. This report explains the results of the sensitivity analyses within the defined ranges, which complement some of the content of the previous study's calculation for the loss of ECCS recirculation functions event.
Yamazaki, Dai; Maruyama, Ryuji; Aoki, Hiroyuki; Hanashima, Takayasu*; Akutsu-Suyama, Kazuhiro*; Miyata, Noboru*; Soyama, Kazuhiko
Quantum Beam Science (Internet), 9(2), p.20_1 - 20_12, 2025/06
Sonehara, Masateru; Okano, Yasushi; Uchibori, Akihiro; Oki, Hiroshi*
Journal of Nuclear Science and Technology, 62(5), p.403 - 414, 2025/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)For sodium-cooled fast reactors, understanding sodium combustion behaviour is crucial for managing sodium leakage accidents. In this study, we perform benchmark analyses of the Sandia National Laboratories (SNL) T3 experiment using the multi-dimensional thermal hydraulic code AQUA-SF. Conducted in an enclosed space with a large vessel volume of 100 m
and a sodium mass flow rate of 1 kg/s, the experiment highlighted the multi-dimensional effects of local temperature increase shortly after sodium injection. This study aims to extend the capabilities of AQUA-SF by focusing on the simulation of these multi-dimensional temperature variations, in particular the formation of high temperature regions at the bottom of the vessel. The proposed models include the temporary stopping of sodium droplet ignition and spray combustion of sodium splash on the floor. Furthermore, it has been shown that additional heat source near the floor is essential to enhance the reproduction of the high temperature region at the bottom. Therefore, case studies including sensitivity analyses of spray cone angle and prolonged combustion of droplets on the floor are conducted. This comprehensive approach provides valuable insights into the dynamics of sodium combustion and safety measures in sodium-cooled fast reactors.
Endo, Akira
ESI-News, 43(2), p.37 - 41, 2025/04
The International Commission on Radiation Units and Measurements (ICRU) published ICRU Report 95 in 2020, revising the operational quantities for external exposure. This article provides an overview of the developments in the discussions within the ICRU, the International Commission on Radiological Protection (ICRP), and experts from Japan, and explores the background and process that led the ICRU to revise the operational quantities, as well as future responses and challenges. The article aims to enhance the understanding of the experts of the new operational quantities and to contribute to their smooth implementation in the future.
Hamdani, A.; Soma, Shu; Abe, Satoshi; Shibamoto, Yasuteru
Progress in Nuclear Science and Technology (Internet), 7, p.53 - 59, 2025/03
Arai, Tsuyoshi*; Nakamura, Fumiya*; Abe, Ryoji*; Ueno, Fuga*; Seko, Noriaki*; Arai, Yoichi; Watanabe, So
Progress in Nuclear Science and Technology (Internet), 7, p.147 - 153, 2025/03
no abstracts in English
Tanigawa, Masafumi; Seya, Kazuhito*; Asakawa, Naoya*; Hayashi, Hiroyuki*; Horigome, Kazushi; Mukai, Yasunobu; Kitao, Takahiko; Nakamura, Hironobu; Henzlova, D.*; Swinhoe, M. T.*; et al.
JAEA-Technology 2024-014, 63 Pages, 2025/02
JAEA-Technology-2024-014.pdf:3.02MB
The liquid waste treatment process generated sludge items at the plutonium conversion development facility. They are highly heterogeneous and contain large amounts of impurities (Na, Fe, Ni etc.). Therefore, the sludge items have very large sampling uncertainty and so the total measurement uncertainty is very large (approximately 24%). The plutonium scrap multiplicity counter (PSMC) measurement technique for sludge items was developed by joint research between the Japan Atomic Energy Agency (JAEA) and Los Alamos National Laboratory (LANL). The technical validity for sludge items using the PSMC was evaluated using various types of sample measurements and Monte Carlo N-Particle transport code calculations. The PSMC measurement parameters were found to be valid for use with sludge items and the validity of multiplicity analysis was confirmed and demonstrated through comparisons with standard MOX powder and a standard sludge. As a result, the PSMC measurement values were shown to be consistent and reasonable and the large amount of impurity (Fe, Ni etc.) did not impact the results. Therefore, the measurement uncertainty of the improved nuclear material accountancy (NMA) procedure by combined PSMC and high-resolution gamma spectrometry was shown to be 6.5%. In addition, an acceptance test was conducted using PSMC/HRGS and IAEA benchmark equipment. Measured Pu mass by both equipment agrees within the measurement uncertainty of each method, and so the validity for Pu mass measurement by PSMC/HRGS was confirmed. The above results confirm the applicability of PSMC/HRGS as an additional NMA method for sludge and a newly designed NDA procedure based on this study is applied to sludge for NMA in PCDF.
Aihara, Jun; Ueta, Shohei; Honda, Masaki*; Kasahara, Seiji; Okamoto, Koji*
JAEA-Research 2024-012, 98 Pages, 2025/02
JAEA-Research-2024-012.pdf:32.24MB
Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for the purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO (PuO-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. In research project of Pu-burner HTGR carried out from fiscal year of 2014 to fiscal year of 2017, simulated CFPs were fabricated using Ce to simulate Pu. Moreover, simulated fuel compacts were fabricated using fabricated simulated CFPs. In this report, results of microstructural observation of CeO-YSZ and ZrC layer at each fabrication step are reported.
