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Okamoto, Naritoshi; Komeno, Akira; Seya, Atsumasa; Inaba, Hideki*; Terakado, Shinichi*; Higuchi, Masashi*
JAEA-Data/Code 2025-022, 497 Pages, 2026/03
JAEA-Data-Code-2025-022.pdf:18.06MB
The Plutonium Fuel Third Development Laboratory of the Nuclear Fuel Cycle Engineering Laboratories has applied for a change of use permit (hereinafter referred to as "license") for plutonium fuel facilities. For the criticality safety design of gloveboxes and equipment/instruments handling mixed oxide (MOX), various criticality calculation codes are used. The most recent employs the 3D Monte Carlo calculation code KENO-V.a embedded in the SCALE 4.4 code system, along with the 27-group ENDF/B-IV neutron cross-section library. SCALE 4.4 was released by the Oak Ridge National Laboratory (ORNL) in the US in 1998, and has now been in use for 27 years. ORNL has continuously improved its functionality, with SCALE 6.3.2 released in 2024. When designing and constructing new MOX fuel facilities, it is desirable to obtain a license using criticality calculation codes based on the latest knowledge. However, it is necessary to verify that these codes have sufficient reliability. Therefore, in 2018, benchmark calculations were performed using the 252-group ENDF/B-VII.1 neutron cross-section library (v7-252n) for two versions of the criticality calculation sequences KENO-V.a and KENO-VI from SCALE 6.2.3, based on past criticality experimental setups. The estimated critical-limiting multiplication factor was calculated. The results indicate that these codes can be used with sufficient confidence for criticality safety design of MOX fuel facilities.
Miyahara, Shinya*; Koie, Ryusuke*; Uno, Masayoshi*; Kawaguchi, Munemichi*; Sato, Rika; Seino, Hiroshi
Nuclear Engineering and Design, 446(Part A), p.114523_1 - 114523_14, 2026/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)
vapor-liquid transfer model in the chemical behavior analysis code SCHERN for accident of evaporation to dryness by boiling of reprocessed high level liquid wasteYoshida, Kazuo; Hiyama, Mina*; Tamaki, Hitoshi
JAEA-Research 2025-011, 25 Pages, 2025/11
JAEA-Research-2025-011.pdf:2.15MB
An accident of evaporation to dryness by boiling of high-level radioactive liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (RuO) are released from the tanks with water and nitric-acid mixed vapor into the atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. RuO is expected to be absorbed chemically into water dissolving nitrous acid (HNO
). This behavior has been experimentally confirmed and plays an important role in the migration of Ru in the facility. A new model has been proposed as a chemical and physical absorption model based on the experimental results of the migration of RuO into nitric acid-water mixtures. In this study, to improve the analytical performance of SCHERN, these new analytical models have been incorporated and attempted to analyze the behavior of RuO in each phase. As a result, it has been observed a tendency that HNO in the liquid phase increases rapidly during the late boiling phase, when RuO release increases rapidly, and confirmed that this HNO concentration change significantly affects the subsequent migration behavior of RuO. These results indicate that it is essential to improve the analytical accuracy of the chemical behavior of HNO in each phase.
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.
Luu, V. N.; Taniguchi, Yoshinori; Udagawa, Yutaka; Katsuyama, Jinya
Nuclear Engineering and Design, 442, p.114222_1 - 114222_15, 2025/10
Times Cited Count：2 Percentile：80.51(Nuclear Science & Technology)Taniguchi, Yoshinori; Urano, Kenta; Mihara, Takeshi; Udagawa, Yutaka; Kakiuchi, Kazuo; Katsuyama, Jinya
Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), p.1292 - 1301, 2025/10
Ishida, Shinya; Tagami, Hirotaka*; Yamano, Hidemasa; Kubo, Shigenobu
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 11 Pages, 2025/09
Wozniak, N.*; Ohgama, Kazuya; Doda, Norihiro; Ota, Hirokazu*; Shemon, E.*; Feng, B.*; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; et al.
Progress in Nuclear Science and Technology (Internet), 8, p.165 - 170, 2025/09
To enhance the accuracy of the safety evaluations for sodium-cooled fast reactors, it is necessary to develop a method that can realistically evaluate the reactivity changes induced by core deformation. In this context, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using multiple ducts from a Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and experimental results demonstrated that the codes used by both countries were able to reasonably predict the thermal bowing of a row of assemblies in multiple duct configuration.
Ohgama, Kazuya; Doda, Norihiro; Ota, Hirokazu*; Wozniak, N.*; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; Ogata, Takanari*; Shemon, E.*; et al.
Progress in Nuclear Science and Technology (Internet), 8, p.160 - 164, 2025/09
To enhance the accuracy of the safety evaluations for sodium-cooled fast reactors, it is necessary to develop a method that can realistically evaluate the reactivity changes induced by core deformation. In this context, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using a single duct from a Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and experimental results demonstrated that the codes used by both countries were able to reasonably predict the axial distribution of horizontal duct displacement caused by thermal bowing as well as the contact load on the duct pad.
Yoshida, Kazuo; Hiyama, Mina*; Tamaki, Hitoshi
JAEA-Research 2025-003, 24 Pages, 2025/06
JAEA-Research-2025-003.pdf:2.06MB
An accident of evaporation to dryness by boiling of high-level radioactive liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (RuO) are released from the tanks with water and nitric-acid mixed vapor into the atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. RuO is expected to be absorbed chemically into water dissolving nitrous acid. Condensation of mixed vapor plays an important role for Ru transporting behavior in the facility building. The thermal-hydraulic behavior in the facility building is simulated with MELCOR code. The latent heat, which is a governing factor for vapor condensing behavior, has almost same value for nitric acid and water at the temperature range under 120 centigrade. Considering this thermal characteristic, it is assumed that the amount of nitric acid is substituted with mole-equivalent water in MELCOR simulation. Compensating modeling induced deviation by this assumption have been assembled with control function features of MELCOR. The comparison results have been described conducted between original simulation and modified simulation with compensating model in this report. It has been revealed that the total amount of pool water in the facility was as same as both simulations.
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.
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.
Tobita, Yoshiharu; Tagami, Hirotaka; Ishida, Shinya; Onoda, Yuichi; Sogabe, Joji; Okano, Yasushi
IAEA-TECDOC-2079, p.72 - 84, 2025/01
Since the fast reactor core is not in the maximum reactivity configuration, a hypothetical core disruptive accident could lead to the prompt criticality due to a change in the core material configuration, and the resulting energy generation has been one of the issues in fast reactor safety, and therefore appropriate measures are needed to mitigate and contain the effect of energy generated in the accident. In order to assess the effectiveness of these mitigative measures, a set of computer codes to analyze the event progressions and energy generation behavior in the ATWS of fast reactors have been developed, maintained, and improved under international collaboration in JAEA. Since the important physical phenomena, which govern the event progression, vary along with the progression of the accident, the whole accident process is divided into several phases in the analysis of accident, and dedicated analysis methods for each phase are provided to analyze the event progression in each phase. The organization and overview of these analysis methods are described in this paper. As a representative example of the validation approaches in applying these analysis methods to the reactor safety assessment in licensing procedure in Japan, the validation studies to confirm the applicability to reactor analysis of the SIMMER code for analyzing core material movement and reactor power, which is important to analyze the energy generation in the accident, are presented in the paper. The validation studies of the SIMMER code confirmed the applicability of SIMMER to the reactor analysis, while the critical phenomena that the effect of their uncertainty in the reactor analysis should be checked were also recognized.
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.
Yoshida, Kazuo; Hiyama, Mina*; Tamaki, Hitoshi
JAEA-Research 2024-007, 24 Pages, 2024/08
JAEA-Research-2024-007.pdf:2.1MB
An accident of evaporation to dryness by boiling of high-level radioactive liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into the atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. It has been observed experimentally that volatility of RuO is suppressed by HNO generated by HNO
radiolysis. The analysis of chemical reactions of NO
including HNO and HNO in the waste tank is essential to simulate of these phenomena. To resolve this issue, an analytical approach has been attempted to couple dynamically two computer codes SHAWED and SCHERN. The simulation of boiling behavior in the tank is conducted with SHAWED. SCHERN simulates chemical behaviors of HNO, HNO and NO in the tank. A programmatic coupling algorithm and a trial simulation of the accident are presented in this report.
Ono, Ayako; Okamoto, Kaoru*; Makino, Yasushi*; Hosokawa, Shigeo*; Yoshida, Hiroyuki
Proceedings of Specialist Workshop on Advanced Instrumentation and Measurement Techniques for Nuclear Reactor Thermal-Hydraulics and Severe Accidents (SWINTH-2024) (USB Flash Drive), 13 Pages, 2024/06
JAEA has been developing an advanced neutronic/thermal-hydraulics coupling simulation system. In the coupling simulation system, the detailed thermal-hydraulics codes based on an interface-capturing method (JUPITER or TPFIT) will be adopted to simulate thermal-hydraulics behavior in a fuel bundle. The experimental data and findings relating to the gas-liquid two-phase flow in a fuel bundle are especially required to validate JUPITER/TPFIT. In this study, we therefore develop a measurement method by combining Laser-Doppler Velocimetry (LDV) and photodiodes, which can access to a small flow channel such as a subchannel of a fuel bundle. The developed measurement method is validated by comparison with the measument by a electrical conductance probe. Finally, we obtain experimental data on local flow structures and interactions between gas and liquid phases. The developed measurement method is actually applied to an air-water dispersed bubbly flow to confirm its capability.
Ishida, Shinya; Tagami, Hirotaka; Okano, Yasushi; Yamano, Hidemasa; Kubo, Shigenobu; Tobita, Yoshiharu
Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 10 Pages, 2024/05
Murakami, Kenta*; Arai, Taku*; Yamada, Koji*; Momma, Kensuke*; Tsuji, Takashi*; Nakagawa, Nobuyuki*; Onizawa, Kunio
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 3 Pages, 2024/03
This paper studied the future vision of codes and standards in Japan by systematically comparing Japanese regulatory rules, standards, and industry guides related to long term operation with international safety standards, and confirmed that the Japanese standard system generally meets their recommendations. The recommendation for the future improvements of Japanese codes and standards were summarized into five items.
Taniguchi, Yoshinori; Mihara, Takeshi; Kakiuchi, Kazuo; Udagawa, Yutaka
Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01
Times Cited Count：2 Percentile：29.31(Nuclear Science & Technology)Nakamura, Izumi*; Otani, Akihito*; Okuda, Yukihiko; Watakabe, Tomoyoshi; Takito, Kiyotaka; Okuda, Takahiro; Shimazu, Ryuya*; Sakai, Michiya*; Shibutani, Tadahiro*; Shiratori, Masaki*
Dai-10-Kai Kozobutsu No Anzensei, Shinraisei Ni Kansuru Kokunai Shimpojiumu (JCOSSAR2023) Koen Rombunshu (Internet), p.143 - 149, 2023/10
In 2019, the JSME Code Case for seismic design of nuclear power plant piping systems was published. The Code Case provides the strain-based fatigue criteria and detailed inelastic response analysis procedure as an alternative design rule to the current seismic design, which is based on the stress evaluation by elastic response analysis. In 2022, it was approved to revise the Code Case with improving the cycle counting method for fatigue evaluation to the Rain flow method. In addition, the discussion to incorporate the elastic-plastic behavior of support structures is now in progress for the next revision of the Code Case. This paper discusses the contents and background of the 2022 revision, the progress of the next revision, and future tasks.
