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elik, Y.*; Stankovskiy, A.*; Iwamoto, Hiroki; Iwamoto, Yosuke; Van den Eynde, G.*
Annals of Nuclear Energy, 212, p.111048_1 - 111048_12, 2025/03
Fukuda, Kodai
Annals of Nuclear Energy, 208(1), p.110748_1 - 110748_10, 2024/12
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Brear, D. J.*; Kondo, Satoru; Sogabe, Joji; Tobita, Yoshiharu*; Kamiyama, Kenji
JAEA-Research 2024-009, 134 Pages, 2024/10
JAEA-Research-2024-009.pdf:2.45MB
The SIMMER-III/SIMMER-IV computer codes are being used for liquid-metal fast reactor (LMFR) core disruptive accident (CDA) analysis. The sequence of events predicted in a CDA is often influenced by the heat exchanges between LMFR materials, which are controlled by heat transfer coefficients (HTCs) in the respective materials. The mass transfer processes of melting and freezing, and vaporization and condensation are also controlled by HTCs. The complexities in determining HTCs in a multi-component and multi-phase system are the number of HTCs to be defined at binary contact areas of a fluid with other fluids and structure surfaces, and the modes of heat transfer taking into account different flow topologies representing flow regimes with and without structure. As a result, dozens of HTCs are evaluated in each mesh cell for the heat and mass transfer calculations. This report describes the role of HTCs in SIMMER-III/SIMMER-IV, the heat transfer correlations implemented and the calculation of HTCs in all topologies in multi-component, multi-phase flows. A complete description of the physical basis of HTCs and available experimental correlations is contained in Appendices to this report. The major achievement of the code assessment program conducted in parallel with code development is summarized with respect to HTC modeling to demonstrate that the coding is reliable and that the model is applicable to various multi-phase problems with and without reactor materials.
Center for Computational Science & e-Systems
JAEA-Evaluation 2024-001, 40 Pages, 2024/10
JAEA-Evaluation-2024-001.pdf:1.46MB
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 FY2023 (April 1st, 2023 - March 31st, 2024) and their evaluation by the committee.
Sato, Yuki
Applied Radiation and Isotopes, 212, p.111421_1 - 111421_8, 2024/10
Times Cited Count：0 Percentile：0.00(Chemistry, Inorganic & Nuclear)Zhang, Y.-J.*; Umeda, Takemasa*; Morooka, Satoshi; Harjo, S.; Miyamoto, Goro*; Furuhara, Tadashi*
Metallurgical and Materials Transactions A, 55(10), p.3921 - 3936, 2024/10
Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)
-CsOHRizaal, M.; Luu, V. N.; Nakajima, Kunihisa; Miwa, Shuhei
Proceedings of International Topical Workshop on Fukushima-Daiichi Decommissioning Research 2024 (FDR2024) (Internet), 4 Pages, 2024/10
Onoda, Yuichi; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa
Proceedings of Probabilistic Safety Assessment and Management & Asian Symposium on Risk Assessment and Management (PSAM17 & ASRAM2024) (Internet), 10 Pages, 2024/10
We developed the measures for improving resilience of the sodium-cooled fast reactor structure using the failure mitigation technology and evaluated the effectiveness of the measures. To prevent core damage in the event of an accident progressing to an ultra-high temperature state, both measures to prevent overpressure in the reactor vessel and measures to cool the reactor core are required. As a core cooling measure, we developed a core cooling concept that promotes radiant heat transfer from the reactor vessel and cools the containment vessel outer surface by natural convection named Containment Vessel Auxiliary Cooling System (CVACS). We developed a method to use the reduction rate of core damage frequency as an indicator for effectiveness of the measures for improving resilience. The core damage frequency was evaluated by calculating the core cooling performance using CVACS, reflecting the results of structural analysis and human reliability analysis. By implementing measures for improving resilience in addition to existing measures, the core damage frequency of Japan loop-type sodium-cooled fast reactor caused by LOHRS has been reduced to about one-hundredth of the previous level.
Collaborative Laboratories for Advanced Decommissioning Science; Tokyo Institute of Technology*
JAEA-Review 2024-022, 59 Pages, 2024/09
JAEA-Review-2024-022.pdf:4.27MB
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 FY2022. 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 FY2022, this report summarizes the research results of the "Investigation of effects of nano interfacial phenomena on dissolution aggregation of alpha nanoparticles by using micro nano technologies" conducted in FY2022. To ensure the safety of retrieval and storage management of nuclear fuel debris generated by the Fukushima Daiichi Nuclear Power Station accident, understanding of dissolution-denaturation behavior of the fuel debris alpha particles is one of the most crucial issues. This research aims to create novel microfluidic real-time measurement device for elucidating dissolution, aggregation, and denaturation processes of metal oxide nanoparticles under various solution environments, and clarify their nano-size and interfacial effects.
Tian, Q.*; Feng, L.*; Wu, C.*; Wen, J.*; Qiu, X.*; Tanaka, Kazuya; Onuki, Toshihiko*; Yu, Q.*
Journal of Colloid and Interface Science, 669, p.1006 - 1014, 2024/09
Times Cited Count：0 Percentile：0.00(Chemistry, Physical) at temperature range 170 - 290
CLuu, V. N.; Nakajima, Kunihisa
Nuclear Engineering and Design, 426, p.113402_1 - 113402_7, 2024/09
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Licensing Application Group, Fuels and Materials Department
JAEA-Testing 2024-002, 20 Pages, 2024/08
JAEA-Testing-2024-002.pdf:1.46MB
The contamination accident occurred at Plutonium Fuel Research Facility (PFRF) in Japan Atomic Energy Agency (JAEA) Oarai Research and Development Institute on June 6, 2017. During the work of opening the fuel storage container and checking the properties of the contents, the plastic bag that double-packed the inner container burst. The scattering of the fuels contaminated the work room and exposed the worker. The cause of the plastic bag burst was that the enclosed epoxy resin was decomposed by 
-rays and the internal pressure increased due to the generated hydrogen gas. The 54 storage containers containing plutonium held at PFRF also at risk of increasing internal pressure. Therefore, an opening inspection was conducted to confirm the contents of the storage container in the hot cell. In addition, the contents of storage containers that may generate gas were stabilized. We are planning to transport the fuel storage containers out to another facility for the decommission of PFRF. The other 9 storage containers include oxide raw material powder: Pu + 
U in excess of 220 g. In order to decrease to less than 220 g (the limit of transport cask), the metal inner containers in the storage container were taken out and repacked in another storage container. This report describes advance measures such as permit application and the details of about storage container opening inspection and metal inner container repacking.
Terada, Atsuhiko; Nagaishi, Ryuji
Journal of Nuclear Science and Technology, 61(8), p.1135 - 1154, 2024/08
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)In order to elucidate ventilation and exhaust of hydrogen leaked in a partially open space practically, the effects of outer wind on them were studied analytically by using a CFD code in the room of experimental Half-size Hallway model, which has a H
release hole on the bottom, one vent on the roof and another vent on the side: external air flowed in the room from the Door vent and then H was discharged outside from the Roof vent. The H concentration distribution in the room was divided into two layers at the height of Door vent, with a high concentration layer above it and a low concentration layer below it, forming a stratified interface. When the wind speed blown into the room increased, the combination of the Realizable k-e; turbulence model and the turbulence Schmidt number of 1.0 improved the reproducibility of the analysis results of H concentration distribution. The trial analysis suggested that the concern that wind would increase the indoor H concentration could be reduced by using the plate with a simple structure in which two plates were crossed on the Roof vent.
Ito, Tatsuya; Nagaishi, Ryuji; Kuwano, Ryo*
Nuclear Technology, 210(8), p.1427 - 1443, 2024/08
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)The retention of hydrogen (H) bubbles generated by water radiolysis was quantitatively studied in a high-viscous suspension of carbonate slurry consisting of a mixture of suspended solid (SS) of magnesium and calcium precipitates under strongly alkaline conditions, like the radioactive wastes discharged from the coagulation sedimentation (co-precipitation) process at the multinuclide removal equipment in the Fukushima Daiichi Nuclear Power Station. The H retention properties were evaluated in two types of carbonate slurry with different hydrophilicity: the hydrophilic "current type" and the hydrophobic "return type". Then, their properties were compared with those in another suspension of clay suspension of bentonite. From the comparison between the amounts of chemical adsorption and HO in the slurry, it was confirmed that HO molecules must be shared among the SS particles, and this sharing formed the structural viscosity in the slurry, different from that in the clay suspension where electrostatic bonding between the fine clay minerals forms the viscosity. The retention of H bubbles in (by) the slurry was evaluated from the difference in the amount of H observed with and without stirring the slurry after 
Co 
-irradiation. From the comparison of the retention properties of the hydrophilic slurry, the hydrophobic slurry, the clay suspension, and treated water, it was suggested that H2 bubbles were retained not only by the structural viscosity but also by the steric hindrance in the hydrophilic slurry.
Morita, Keisuke; Aoki, Takeshi; Shimizu, Atsushi; Sato, Hiroyuki
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 6 Pages, 2024/08
Department of Decommissioning and Waste Management
JAEA-Review 2024-004, 124 Pages, 2024/07
JAEA-Review-2024-004.pdf:3.39MB
This report describes the activities of Department of Decommissioning and Waste Management (DDWM) in Nuclear Science Research Institute (NSRI) in the period from April 1, 2022 to March 31, 2023. The report covers organization and missions of DDWM, outline and operation/maintenance of facilities which belong to DDWM, treatment and management of radioactive wastes, decommissioning activities, and related research and development activities which were conducted in DDWM. In FY2022 radioactive wastes generated from R&D activities in NSRI were treated safely. They were about 262 m of combustible solid wastes and 113 m
of noncombustible solid wastes and 203 m of liquid wastes. After adequate treatment, 527 waste packages (in 200 L-drum equivalent) were generated. The total amounts of accumulated waste packages were 122,925 as of the end of FY2022 due to efforts of the restitution of waste packages to the Japan Radioisotope Association and volume reduction treatments of the stored waste packages. Decommissioning activities were carried out for the JAEA's Reprocessing Test Facility (JRTF). As for the R&D activities, studies on radiochemical analyses of wastes for disposal were continued. In order to pass the conformity review on the New Regulatory Requirements for waste management facilities, the Approval of the design and construction method was applied sequentially for the Nuclear Regulation Authority. The ministry of the Environment and Tokai-mura office requested JAEA to dispose of the contaminated soil generated by the accident of the Fukushima Daiichi Nuclear Power Station. The monitoring work at the playground was conducted during this period.
Zhang, Z.*; Hattori, Takanori; Song, R.*; Yu, D.*; Mole, R.*; Chen, J.*; He, L.*; Zhang, Z.*; Li, B.*
Journal of Applied Physics, 136(3), p.035105_1 - 035105_8, 2024/07
Times Cited Count：0 Percentile：0.00(Physics, Applied)Solid-state refrigeration using barocaloric materials is environmentally friendly and highly efficient, making it a subject of global interest over the past decade. Here, we report giant barocaloric effects in sodium hexafluorophosphate (NaPF
) and sodium hexafluoroarsenate (NaAsF) that both undergo a cubic-to-rhombohedral phase transition near room temperature. We have determined that the low-temperature phase structure of NaPF is a rhombohedral structure with space group R
and NaAsF, i.e., F
, E
, and A
. The phase transition temperature varies with pressure at a rate of dT
/dP = 250 and 310 K/GPa for NaPF and NaAsF. The pressure-induced entropy changes of NaPF and NaAsF are determined to be around 45.2 and 35.6J kg
K, respectively. The saturation driving pressure is about 40 MPa. The pressure-dependent neutron powder diffraction suggests that the barocaloric effects are related to the pressure-induced cubic-to-rhombohedral phase transitions.
Nguyen, T.-D.*; Singh, C.*; Kim, Y. S.*; Han, J. H. *; Lee, D.-H.*; Lee, K.*; Harjo, S.; Lee, S. Y.*
Journal of Materials Research and Technology, 31, p.1547 - 1556, 2024/07
Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)Watanabe, Tomoaki; Yamane, Yuichi
Journal of Nuclear Science and Technology, 61(7), p.958 - 966, 2024/07
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)The total fission energy released in a criticality accident involving fissile solution boiling tends to be high because the relatively high fission power continues during boiling. Simulating fission power change correctly during boiling seems essential to estimate the total fission energy. Fission power during boiling changes depending on fissile concentration and volume as the solution evaporates. In this study, we investigated the effect of concentration and volume change on estimated total fission energy for a long time of boiling. We introduced a model calculating the evaporation of fissile solution into the modified quasi-steady-state method to simulate power change during boiling. Three CRAC experiments and the Idaho Chemical Processing Plant (ICPP) criticality accident in 1959 were analyzed. As a result, the calculated energy considering concentration and volume change during boiling reproduced the measured energy well.
Shikaze, Yoshiaki
Journal of Nuclear Science and Technology, 61(7), p.894 - 910, 2024/07
Times Cited Count：2 Percentile：41.04(Nuclear Science & Technology)Among the radioactive nuclides inside the nuclear reactor buildings emitted by the Fukushima Daiichi nuclear reactor accident, high-energy beta-ray sources, such as strontium-90 and yttrium-90, generate bremsstrahlung photons in the building materials, comprising the wall, floor, and interior structure. Therefore, evaluating the radiation dose of the bremsstrahlung to the workers in the nuclear reactor building is crucial for radiation protection. The precision of the evaluation calculation of the bremsstrahlung dose was investigated by comparing the Particle and Heavy Ion Transport code System (PHITS) and the GEometry ANd Tracking (GEANT4) simulation code results. In the calculation, behind various shielding plates (lead, copper, aluminum, glass, and polyethylene, with thicknesses ranging from 1.0 to 40 mm), the water cylinder was set as the evaluated material, the absorbed dose and the deposited energy spectrum by the bremsstrahlung photons were obtained, and the characteristics and differences for both simulation codes were investigated. In the comparison results of the deposited energy spectrum, the spectral shapes have consistent trends. In the energy range below several tens of keV, a peak is seen in the PHITS spectrum for the lead shielding material. In comparing the absorbed dose under various conditions of the shielding plate for generating bremsstrahlung photons, most results for both codes correlate within an 
10% difference for 2.280 MeV beta-ray sources and an 20% difference for 0.5459 MeV beta-ray sources, except for 30% for 20 mm thick lead. Although there were differences in some cases, the evaluation results of the two simulation codes were concluded to correlate well with the above precision.
