Kinetic and thermodynamic controls on CsI-Mo gas-phase reactions under varying oxygen potentials

Shiotsu, Hiroyuki

Progress in Nuclear Energy, 195, p.106300_1 - 106300_11, 2026/05

https://doi.org/10.1016/j.pnucene.2026.106300

Benchmark analyses of criticality calculation on SCALE 6.2.3 code system

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.

A Novel kinetic model for dissolution and precipitation of oxide on stainless-steel surface in stagnant liquid sodium

Kawaguchi, Munemichi*; Ikeda, Asuka; Saito, Junichi

Annals of Nuclear Energy, 226, p.111880_1 - 111880_9, 2026/02

https://doi.org/10.1016/j.anucene.2025.111880

Collapse of Jahn-Teller phonons in LaSrMnO with weak magnetoresistance

Sterling, T. C.*; Savici, A. T.*; Kajimoto, Ryoichi; Ikeuchi, Kazuhiko*; Khan, N.*; Weber, F.*; Reznik, D.*

Communications Materials (Internet), 32 Pages, 2026/00

https://doi.org/10.1038/s43246-026-01139-4

Effect of repair welding location and number of welds performed on creep strength of Mod. 9Cr-1Mo steel welded joints

Yamashita, Hayato; Toyota, Kodai; Onizawa, Takashi; Yamamoto, Kenji*; Kubo, Koji*

Nihon Kikai Gakkai Rombunshu (Internet), 13 Pages, 2026/00

https://doi.org/10.1299/transjsme.25-00176

It is planned that Mod.9Cr-1Mo steel will be used as the structural material for the steam generator of the demonstration fast reactor. Creep strength of welded joints of Mod.9Cr-1Mo steel is lower than that of base metal at high temperature and long time. In addition, the creep strength of welded joints of Mod.9Cr-1Mo steel with repair welding is lower than that of without repair welding. In this study, the effects of the location and number of repair welds performed on the creep strength of welded joints of Mod.9Cr-1Mo steel were investigated, and a repair welding method was developed for the construction of the demonstration fast reactor. It was clarified that, regardless of the location and number of repair welds performed, the degradation in creep strength could be greatly reduced by performing the repair welding before post weld heat treatment (PWHT). Furthermore, it was found that the superposition of thermal history and the formation of a heat affected zone (HAZ) on the weld metal promote the coarsening of ferrite grains during creep and cause a slight reduction in creep strength. All repair welding was performed prior to PWHT, and a repair welding method was also developed to minimize thermal history superposition and weld metal HAZ formation.

Analysis of deposits inside "X-6 penetration" for the Unit 2 primary containment vessel at Fukushima Daiichi Nuclear Power Station

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.

A Methodology for the design of non-uniform core configurations in the modified STACY facility

Dechenaux, B.*; Brovchenko, M.*; Araki, Shohei; Gunji, Satoshi; Suyama, Kenya

Annals of Nuclear Energy, 223, p.111555_1 - 111555_11, 2025/12

https://doi.org/10.1016/j.anucene.2025.111555

Development of a prototype shielding-free radiation-resistant diamond neutron measurement system (Contract research); FY2023 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Hokkaido University*

JAEA-Review 2025-028, 66 Pages, 2025/11

JAEA-Review-2025-028.pdf:3.59MB

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 (1F), 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 FY2023, this report summarizes the research results of the "Development of a prototype shielding-free radiation-resistant diamond neutron measurement system" conducted in FY2023. The present study aims to develop a prototype of a shielding-free neutron measurement system for 1F. The system consists of diamond neutron detectors and radiation-resistant silicon integrated circuits, and has radiation resistance of more than 10 MGy and 4 MGy, respectively, at the component level in terms of integrated dose, and has a track record of stable operation under -ray dose rate environment of 1.5 kGy/h. Future applications are expected to include neutron detectors for debris investigation, criticality proximity monitoring monitors, and neutron detectors for dry tube investigation in pressure vessels. In this development, a prototype consisting of 100 diamond detector elements of 5 mm square will be developed to obtain system construction technology and to evaluate system performance. In addition, a subcriticality evaluation method will be developed. This development will lead to the completion of system development, development of the actual system in collaboration with the manufacturer, and introduction of the system into 1F decommissioning project.

Re-evaluation of nuclear criticality characteristics for infinite and finite heterogeneous lattice systems composed of uranium-zirconium hydride fuel rods used in the TRIGA annular core pulse reactor NSRR

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.

Challenge of novel hybrid-waste-solidification of mobile nuclei generated in Fukushima Nuclear Power Station and establishment of rational disposal concept and its safety assessment (Contract research); FY2023 Nuclear Energy Science & Technology and Human Resource Development Project

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

JAEA-Review 2025-016, 143 Pages, 2025/10

JAEA-Review-2025-016.pdf:10.71MB

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 (hereafter referred to "1F"), 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 "Challenge of novel hybrid-waste-solidification of mobile nuclei generated in Fukushima Nuclear Power Station and establishment of rational disposal concept and its safety assessment" conducted from FY2021 to FY2023. This study aims to establish the rational waste disposal concept of various wastes generated in 1F based on the hybrid-waste-solidification by the Hot Isostatic Press (HIP) method. The ceramics form with target elements, mainly iodine, which is challenging to immobilize, and Minor Actinides such as Am, an alpha emitter and heat source, are HIPed with well-studied materials such as SUS and zircaloy, which make the long-term stability evaluation and safety assessment possible. In 2024, the project's final year, we demonstrated the effectiveness of the hybrid solidification concept by linking all the sub-themes, from waste synthesis to disposal considerations. The compatibility of various wastes, such as ALPS, AREVA sediment wastes, AgI, waste silver adsorbent, ceria adsorbent, and iodine apatite, with metals and oxide matrices was investigated. which involves investigating the HIPed hybrid wastes after exploring the compatibility of various metals and oxide matrices using the rapid sintering method, spark plasma sintering (SPS), proposed in this project. It revealed that hybrid waste solidification with SUS matrix was superior for many wastes. Furthermore, we studied waste disposal concepts based on nuclide migration calculations. Finally, we could connect the waste fabrication to safety assessment for the first time, leading to finding an appropriate waste disposal scenario for 1F decommissioning.