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FCC hierarchical martensite transformation under dynamic impact in FeMnAlNiTi alloyLi, C.*; Fang, W.*; Yu, H. Y.*; Peng, T.*; Yao, Z. T.*; Liu, W. G.*; Zhang, X.*; Xu, P. G.; Yin, F.*
Materials Science & Engineering A, 892, p.146096_1 - 146096_11, 2024/02
Times Cited Count：0 Percentile：0.04(Nanoscience & Nanotechnology)Inamura, Yasuhiro
JAEA-Testing 2023-002, 80 Pages, 2023/12
JAEA-Testing-2023-002.pdf:2.43MB
"Utsusemi" is a suite of software used to process data obtained from measurements of neutron scattering experiments at the Materials and Life Science Experimental Facility (MLF), J-PARC. To directly obtain the physical quantities which scientists want to get from the data produced by instruments at MLF, many processes are required, such as creating histogram format data, easily visualizing and converting units and correcting intensity adapting the instrument conditions. "Utsusemi" software consists of many software components, many functions for data processing, graphical interface software for executing Utsusemi functions, data visualization applications, and so on. "Utsusemi" has already played an important role in data processing and has been widely employed in MLF beamlines. This document describes how to install the "Utsusemi" software on each operating system to be of help of instrument staff and users who want to process data by themselves. Installation of "Utsusemi" on Windows and macOS requires only general knowledge of working with PC applications according to this document.
Watanabe, So; Takahatake, Yoko; Ogi, Hiromichi*; Osugi, Takeshi; Taniguchi, Takumi; Sato, Junya; Arai, Tsuyoshi*; Kajinami, Akihiko*
Journal of Nuclear Materials, 585, p.154610_1 - 154610_6, 2023/11
Times Cited Count：0 Percentile：0.01(Materials Science, Multidisciplinary)Tada, Kenichi
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 8 Pages, 2023/10
The number of energy grids of the thermal neutron scattering law data has a large impact on the data size of a cross section file of continuous energy Monte Carlo calculation codes. The optimization of the number of energy grids is an effective way to reduce the data size. This study developed the linearization method of the thermal neutron scattering cross section to optimize the number of energy grids and the linearization function was implemented in the nuclear data processing code FRENDY. The linearization process which is used in the resonance reconstruction and the Doppler broadening was adopted. The criticality benchmarks which use ZrH as the moderator were calculated to estimate the impact of the difference of the energy grids on neutronics calculations. The calculation results indicate that the linearization of the thermal neutron scattering cross section improves the prediction accuracy of neutronics calculations.
Arai, Taiki*; Yoshigoe, Akitaka; Motohashi, Mitsuya*
Zairyo No Kagaku To Kogaku, 60(5), p.153 - 158, 2023/10
Si oxide films are currently widely used as insulating materials in electronic devices and biomaterials. The atomic bonding state of these films significantly influences the properties of each device, thus it is particularly necessary to understand and control the chemical bonding state between Si and O in the films. In this study, the Si oxide films formed by anodic oxidation on Si substrate surfaces in extremely low concentrations of HF solutions were analyzed by X-ray photoelectron spectroscopy mainly focusing on Si2p and F1s spectra. Although the HF concentration is in the order of ppm, the films contain percent order of F atoms, suggesting the formation of Si-F and Si-O-F bonds in the films. It was also found that the different depth profiles for F and O atoms was observed, indicating that the surface reaction processes seem to be different depending on each element.
Sato, Takumi; Otobe, Haruyoshi; Morishita, Kazuki; Marufuji, Takato; Ishikawa, Takashi; Fujishima, Tadatsune; Nakano, Tomoyuki
JAEA-Technology 2023-016, 41 Pages, 2023/09
JAEA-Technology-2023-016.pdf:2.74MB
This report summarizes the results of the stabilization treatments of post-experiment nuclear materials in Plutonium Fuel Research Facility (PFRF) from August 2018 to March 2021. Based on the management standards for nuclear materials enacted after the contamination accident that occurred at PFRF on June 6, 2017, the post-experiment nuclear materials containing plutonium (Pu): samples mixed with organic substances that cause an increase in internal pressure due to radiolysis (including X-ray diffraction samples mixed with epoxy resin and plutonium powder which caused contamination accidents), carbides and nitrides samples which is reactive in air, and chloride samples which may cause corrosion of storage containers, were selected as targets of the stabilization. The samples containing organic materials, carbides and nitrides were heated in an air flow at 650 
C and 950 C for 2 hours respectively to remove organic materials and convert uranium (U) and Pu into oxides. U and Pu chlorides in LiCl-KCl eutectic melt were reduced and extracted into liquid Cd metal by a reaction with lithium (Li) -cadmium (Cd) alloy and converted to U-Pu-Cd alloy at 500 C or higher. All of the samples were stabilized and stored at PFRF. We hope that the contents of this report will be utilized to consider methods for stabilizing post experiment nuclear materials at other nuclear fuel material usage facilities.
Nakazawa, Osamu; Takiya, Hiroaki; Murakami, Masashi; Donomae, Yasushi; Meguro, Yoshihiro
JAEA-Review 2023-012, 6 Pages, 2023/08
JAEA-Review-2023-012.pdf:0.93MB
The selection of back-end technology development issues to be prioritized and their schedule of the Japan Atomic Energy Agency (JAEA) have been put together as the "Strategic Roadmap for Back-end Technology Development." The results of questionnaires on development technologies (seeds) and technical issues (needs) within JAEA conducted in FY2022 were reflected in the selection. The issues were extracted from among those that match the seeds and needs, from the perspective of early implementation in the work front and the perspective of common issues, and nine themes were selected. We will build a cross-organizational implementation framework within JAEA and aim to implement the development results in the work front as well as social implementation.
Hayashi, Hirokazu; Tsubata, Yasuhiro; Sato, Takumi
Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 22(3), p.97 - 107, 2023/08
The Japan Atomic Energy Agency has chosen nitride fuel as the first candidate for the transmutation of long-lived minor actinides (MA) using accelerator-driven systems (ADS). The pyrochemical method has been considered for reprocessing spent MA nitride fuels, because their decay heat should be very large for aqueous reprocessing. This study was conducted to investigate the effect of decay heat on the pyrochemical reprocessing of MA nitride fuels. On the basis of the estimated decay heats and the temperature limits of the materials that are to be handled in pyrochemical reprocessing, quantities adequate for handling in argon gas atmosphere were evaluated. From these considerations, we proposed that an electrorefiner with a diameter of 26 cm comprising 12 cadmium (Cd) cathodes with a diameter of 4 cm is suitable. On the basis of the size of the electrorefiner, the number necessary to reprocess spent MA fuels from 1 ADS in 200 days was evaluated to be 25. Furthermore, the amount of Cd-actinides (An) alloy to produce An nitrides by the nitridation-distillation combined reaction process was proposed to be about one-quarter that of Cd-An cathode material. The evaluated sizes and required numbers of equipment support the feasibility of pyrochemical reprocessing for MA nitride fuels.
Nishino, Saki; Okada, Jumpei; Watanabe, Kazuki; Furuuchi, Yuta; Yokota, Satoru; Yada, Yuji; Kusaka, Shota; Morokado, Shiori; Nakamura, Yoshinobu
JAEA-Technology 2023-011, 39 Pages, 2023/06
JAEA-Technology-2023-011.pdf:2.51MB
Tokai Reprocessing Plant (TRP) which shifted to decommissioning phase in 2014 had nuclear fuel materials such as the spent fuel sheared powder, the diluted plutonium solution and the uranium solution in a part of the reprocessing main equipment because TRP intended to resume reprocessing operations when it suspended the operations in 2007. Therefore, we have planned to remove these nuclear materials in sequence as Flush-out before beginning the decommissioning, and conducted removal of the spent fuel sheared powder as the first stage. The spent fuel sheared powder that had accumulated in the cell of the Main Plant (MP) as a result of the spent fuel shearing process was recovered from the cell floor, the shearing machine and the distributor between April 2016 and April 2017 as part of maintenance. Removing the recovered spent fuel sheared powder was conducted between June 2022 and September 2022. In this work, the recovered powder was dissolved in nitric acid at the dissolver in a small amount in order to remove it safely and early, and the dissolved solution was sent to the highly radioactive waste storage tanks without separating uranium and plutonium. Then, the dissolved solution transfer route was rinsed with nitric acid and water. Although about 15 years had passed since previous process operations, the removing work was successfully completed without any equipment failure because of the organization of a system that combines veterans experienced the operation with young workers, careful equipment inspections, and worker education and training. Removing this powder was conducted after revising the decommissioning project and obtaining approval from the Nuclear Regulation Authority owing to operating a part of process equipment.
Watanabe, Kazuki; Kimura, Norimichi*; Okada, Jumpei; Furuuchi, Yuta; Kuwana, Hideharu*; Otani, Takehisa; Yokota, Satoru; Nakamura, Yoshinobu
JAEA-Technology 2023-010, 29 Pages, 2023/06
JAEA-Technology-2023-010.pdf:3.12MB
The Krypton Recovery Development Facility reached an intended technical target (krypton purity of over 90% and recovery rate of over 90%) by separation and rectification of krypton gas from receiving off-gas produced by the shearing and the dissolution process in the spent fuel reprocessing at the Tokai Reprocessing Plant (TRP) between 1988 and 2001. In addition, the feasibility of the technology was confirmed through immobilization test with ion-implantation in a small test vessel from 2000 to 2002, using a part of recovered krypton gas. As there were no intentions to use the remaining radioactive krypton gas in the krypton storage cylinders, we planned to release this gas by controlling the release amount from the main stack, and conducted it from February 14 to April 26, 2022. In this work, all the radioactive krypton gas in the cylinders (about 7.1
10
GBq) was released at the rate of 50 GBq/min or less lower than the maximum release rate from the main stuck stipulated in safety regulations (3.710
GBq/min). Then, the equipment used in the controlled release of radioactive krypton gas and the main process (all systems, including branch pipes connected to the main process) were cleaned with nitrogen gas. Although there were delays due to weather, we were able to complete the controlled release of radioactive krypton gas by the end of April 2022, as originally targeted without any problems such as equipment failure.
Yamamoto, Masahiko; Nishida, Naoki; Kobayashi, Daisuke; Nemoto, Ryo*; Hayashi, Hiroyuki*; Kitao, Takahiko; Kuno, Takehiko
JAEA-Technology 2023-004, 30 Pages, 2023/06
JAEA-Technology-2023-004.pdf:1.94MB
Glove-box gloves, that are used for handling nuclear fuel materials at the Tokai Reprocessing Plant (TRP) of the Japan Atomic Energy Agency, have an expiration date by internal rules. All gloves are replaced at a maximum of every 4-year. However, degrees of glove deterioration varies depending on its usage environment such as frequency, chemicals, and radiation dose. Therefore, physical properties such as tensile strength, elongation, hardness of gloves are measured and technical evaluation method for the glove life-time is established. It was found that gloves without any defects in its appearance have enough physical properties and satisfies the acceptance criteria values of new gloves. Thus, it was considered that the expired gloves could be used for total of 8-year, by adding 4-year of new glove life-time. In addition, the results of extrapolation by plotting the glove's physical properties versus the used years showed that the physical properties at 8-year is on the safer side than the reported physical properties of broken glove. Also, the data are not significantly different from the physical properties of the long-term storage glove (8 and 23 years). Based on these results, life-time of gloves at TRP is set to be 8-year. The frequency of glove inspections are not changed, and if any defects is found, the glove is promptly replaced. Thus, the risk related to glove usage is not increased. The cost of purchasing gloves, labor for glove replacement, and the amount of generated waste can be reduced by approximately 40%, respectively, resulting in more efficient and rationalized glove management.
Fukaya, Yuji; Maruyama, Takahiro; Goto, Minoru; Ohashi, Hirofumi; Higuchi, Hideaki
JAEA-Research 2023-002, 19 Pages, 2023/06
JAEA-Research-2023-002.pdf:1.48MB
A study on disposal of waste derived from commercial High Temperature Gas-cooled Reactor ("HTGR") has been performed. Because of significant difference between the reprocessing of Light Water Reactor ("LWR") and that of HTGR due to difference in structures of the fuel, adoptability of the laws relating to reprocessing waste disposal, which is enacted for LWR, to HTGR waste should be confirmed. Then, we compared the technologies and waste of reprocessing and evaluated radioactivity concentration in graphite waste by activation and contamination based on whole core burn-up calculation. As a result, it was found that SiC residue waste should be disposed of into a geological repository as 2nd class designated radioactive waste in the Designated Radioactive Waste Final Disposal Act (Act No.117 of 2000), by way of amendment of the applicable order, same as hull and end-piece of LWR, and graphite waste should be shallowly disposed of than geological disposal as 2nd class waste for pit disposal in the Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors (Act No.166 of 1957) same as a channel box of LWR.
Tada, Kenichi
Kaku Deta Nyusu (Internet), (135), p.1 - 10, 2023/06
This article summarizes presentations at the IAEA technical meeting on nuclear data processing. In this technical meeting, the current development status of nuclear data processing codes and comparisons of the processing results using these codes were presented.
Tada, Kenichi; Endo, Tomohiro*
EPJ Web of Conferences, 284, p.14013_1 - 14013_4, 2023/05
Times Cited Count：0 Percentile：0.21(Nuclear Science & Technology)The self-shielding effect in the unresolved resonance region has a large impact on the fast- and intermediate-spectrum reactors. The probability table method is widely used for continuous-energy Monte Carlo calculation codes to treat the effect. In this method, a table provides the probability distribution of the cross-section for a nuclide in the given energy grid points. The table is generated by averaging with a lot of "ladders" which represent pseudo resonance structures. Though many nuclear data processing codes require the number of ladders as an input parameter to generate the probability table, an optimal number of ladders has not been investigated. Our previous study revealed that the suitable number of ladders depends on the nuclide and its resonance parameters. This result indicates that it is very difficult for users to find the optimal number of ladders. We developed the calculation method of the statistical uncertainty for the probability table generation.
Tada, Kenichi
Robutsuri No Kenkyu (Internet), (75), 13 Pages, 2023/03
In addition to nuclear data processing, FRENDY has various functions such as editing nuclear data and plotting cross section data. This document introduces these functions.
Tada, Kenichi
Shahei Kaiseki No V&V Gaidorain Sakutei Ni Mukete, p.11 - 16, 2023/03
An overview of the nuclear data processing code FRENDY is introduced for shielding calculation code users who are not familiar with FRENDY. This paper explains the nuclear data processing flow in FRENDY, the purpose of use, input examples, verification and validation reports, and so on.
Nuclear Science and Engineering Center; Fuel Cycle Design Office; Plutonium Fuel Development Center; Nuclear Plant Innovation Promotion Office; Fast Reactor Cycle System Research and Development Center; J-PARC Center
JAEA-Review 2022-052, 342 Pages, 2023/02
JAEA-Review-2022-052.pdf:18.05MB
This report summarizes the current status and future plans of research and development (R&D) on partitioning and transmutation technology in Japan Atomic Energy Agency, focusing on the results during the 3rd Medium- to Long-term Plan period (FY 2015-2021). Regarding the partitioning technology, R&D of the solvent extraction method and the extraction chromatography method are described, and regarding the minor actinide containing fuel technology, R&D of the oxide fuel production using the simplified pellet method, the nitride fuel production using the external gelation method, and pyrochemical reprocessing of the nitride fuel were summarized. Regarding transmutation technology, R&D of technology using fast reactors and accelerator drive systems were summarized. Finally, the new facilities necessary for the future R&D were mentioned.
Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi
Annals of Nuclear Energy, 181, p.109534_1 - 109534_10, 2023/02
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Feasibility of reprocessing of High Temperature Gas-cooled Reactor (HTGR) spent fuel by existing Plutonium Uranium Redox EXtraction (PUREX) plant and technology has been investigated. The spent fuel dissolved solution includes approximately 3 times amount of uranium-235 and 1.5 times amount of protonium because of the 3 times higher burnup compared with that of Light Water Reactor (LWR). Then, the heavy metal of the spent fuel is planned to be diluted to 3.1 times by depleted uranium to satisfy the limitation of Rokkasho Reprocessing Plant (RRP) plant. In the present study, recoverability of uranium and plutonium with the dilution is confirmed by a simulation with a reprocessing process calculation code. Moreover, the case without the dilution from the economic perspective is investigated. As a result, the feasibility is confirmed without the dilution, and it is expected that the reprocessed amount is reduced to 1/3 compared with a diluted case even though the facility should be optimized from the perspective of mass flow and criticality.
(M = Gd, Np) by the reaction of MN with Pd and chlorination of MPd using cadmium chlorideHayashi, Hirokazu; Shibata, Hiroki; Sato, Takumi; Otobe, Haruyoshi
Journal of Radioanalytical and Nuclear Chemistry, 332(2), p.503 - 510, 2023/02
Times Cited Count：0 Percentile：0.01(Chemistry, Analytical)The formation of MPd (M = Gd, Np) by the reaction of MN with Pd at 1323 K in Ar gas flow was observed. Cubic AuCu
-type GdPd
(
= 0.4081 
0.0001 nm) and NpPd ( = 0.4081 0.0001 nm) were identified, respectively. The product obtained from the reaction of NpN with Pd contained additional phases including the hexagonal TiNi-type NpPd. Chlorination of the MPd (M = Gd, Np) samples was accomplished by the solid-state reaction using cadmium chloride at 673 K in a dynamic vacuum. Pd-rich solid solution phase saturated with Cd and an intermetallic compound PdCd were obtained as by-products of MCl formation.
Tada, Kenichi; Yamamoto, Akio*; Kunieda, Satoshi; Konno, Chikara; Kondo, Ryoichi; Endo, Tomohiro*; Chiba, Go*; Ono, Michitaka*; Tojo, Masayuki*
Journal of Nuclear Science and Technology, 10 Pages, 2023/00
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Nuclear data processing code is important to connect evaluated nuclear data libraries and radiation transport codes. The nuclear data processing code FRENDY version 1 was released in 2019 to generate ACE formatted cross section files with simple input data. After we released FRENDY version 1, many functions were developed, e.g., neutron multi-group cross section generation, explicit consideration of the resonance interference effect among different nuclides in a material, consideration of the resonance upscattering, ACE file perturbation, and modification of ENDF-6 formatted file. FRENDY version 2 was released including these new functions. It generates GENDF and MATXS formatted neutron multi-group cross section files from an ACE formatted cross section file or an evaluated nuclear data file. This paper explains the features of the new functions implemented in FRENDY version 2 and the verification of the neutron multigroup cross section generation function of this code.
