Periez, R.*; Brovchenko, I.*; Jung, K. T.*; Kim, K. O.*; Liptak, L.*; Little, A.*; Kobayashi, Takuya; Maderich, V.*; Min, B. I.*; Suh, K. S.*
Journal of Environmental Radioactivity, 261, p.107138_1 - 107138_8, 2023/05
Lagrangian models present several advantages over Eulerian models to simulate the transport of radionuclides in the aquatic environment in emergency situations. A radionuclide release is simulated as a number of particles whose trajectories are calculated along time and thus these models do not require a spatial discretization. In this paper we investigate the dependence of a Lagrangian model output with the grid spacing which is used to calculate concentrations from the final distribution of particles, with the number of particles in the simulation and with the interpolation schemes which are required because of the discrete nature of the water circulation data used to feed the model.
Katata, Genki*; Yamaguchi, Takashi*; Watanabe, Makoto*; Fukushima, Keitaro*; Nakayama, Masataka*; Nagano, Hirohiko*; Koarashi, Jun; Tateno, Ryunosuke*; Kubota, Tomohiro
Atmospheric Environment, 298, p.119640_1 - 119640_12, 2023/04
Dong, F.*; Chen, S.*; Demachi, Kazuyuki*; Yoshikawa, Masanori; Seki, Akiyuki; Takaya, Shigeru
Nuclear Engineering and Design, 404, p.112161_1 - 112161_15, 2023/04
Yamashita, Naoki; Irisawa, Eriko; Kato, Chiaki; Sano, Naruto; Tagami, Susumu
JAEA-Technology 2022-035, 29 Pages, 2023/03
In the treatment process of the current commercial reprocessing plant (Rokkasho Reprocessing Plant), the high-level liquid waste concentrator is the equipment that treats the most corrosive solution. In the high-level liquid waste concentrator, the extracted liquid waste after separation of uranium and plutonium is heated, concentrated, and reduced in volume. Therefore, the amount of gamma- rays emitted from fission products and the concentration of corrosive metal ion species such as neptunium-237 (Np) are the highest in the reprocessing process, and the amount of corrosion in the high-level liquid waste concentrate canner is expected to be large. In this study, in order to clarify the effect of gamma-rays on the corrosion reaction of stainless steel in nitric acid solutions containing Np from the electrochemical viewpoint, the corrosion test apparatus for heat transfer surfaces in an airtight concrete cell at the Waste Safety TEsting Facility (WASTEF) of Nuclear Science Research Institute was modified to enable electrochemical measurements under gamma-ray irradiation. The effect of gamma-rays on the corrosion reaction taking place on the stainless steel surface was discussed from the electrochemical test results obtained. As a result, changes in the immersion potentials of stainless steel and the polarization curves due to chemical species caused by radiolysis of gamma-ray irradiation were confirmed.
Sano, Naruto; Yamashita, Naoki; Hishino, Kazutoyo*; Tsukada, Manabu*; Sawauchi, Fumiya*; Otake, Yoshinori*; Ichise, Kenichi; Tagami, Susumu
JAEA-Technology 2022-034, 47 Pages, 2023/03
The Waste Safety Testing Facility (WASTEF) was established in 1982 as an experimental facility for long-term storage of solidified high-level radioactive waste generated in the reprocessing of spent light water reactor fuel and the subsequent safety assessment of geological disposal. It is a historic facility that started operation in 1982. This facility consists of 5 concrete cells, 1 lead cell, 6 glove boxes, and 7 hoods, and is a large-scale facility that can use nuclear fuel materials including uranium and plutonium and radioactive isotopes including TRU. In this facility, research and development requested by the research department is carried out in the Hot Material Examination Section. In addition, patrol inspections, self-inspections, etc. are also carried out as maintenance management based on safety regulations. This report summarizes the overview of WASTEF facilities, the results of operation, maintenance and management work in FY2021, and the future outlook.
Nuclear Science and Engineering Center
JAEA-Evaluation 2022-011, 34 Pages, 2023/03
Japan Atomic Energy Agency (hereinafter referred to as "JAEA") consults an assessment committee, "Evaluation Committee of Research Activities for Nuclear Science and Engineering" (hereinafter referred to as "Committee") for result and in-advance evaluation of "Nuclear Science and Engineering", in accordance with "General Guideline for the Evaluation of Government Research and Development (R&D) Activities" by Cabinet Office, Government of Japan, "Guideline for Evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology" and "Regulation on Conduct for Evaluation of R&D Activities" by the JAEA. In response to the JAEA's request, the Committee assessed the research program of the Nuclear Science and Engineering Center (hereinafter referred to as "NSEC). The Committee evaluated the management and research activities of the NSEC based on explanatory documents prepared by the NSEC, and oral presentations with questions-and-answers.
Nemoto, Yoshiyuki; Ishijima, Yasuhiro; Kondo, Keietsu; Fujimura, Yuki; Kaji, Yoshiyuki
Journal of Nuclear Materials, 575, p.154209_1 - 154209_19, 2023/03
Previous studies had shown that in certain conditions, the rate of oxidation of zirconium (Zr) based alloy fuel cladding is higher in air-steam mixtures than in dry air. In severe accidents in the spent fuel pool and in other air ingress accidents in nuclear power plants, the cladding is likely to be oxidized in an air-steam mixture, which makes it crucial to have an in-depth understanding of the nature of oxidation and its kinetics in that environment. Oxidation tests were conducted at 800C on Zircaloy-4 specimens in a mix of (air+steam) with various component ratios. Oxidation kinetics, details of the oxide layer, and hydrogen pick-up in the specimen were studied to investigate the mechanism of oxidation in each of these sets of conditions. Zirconium nitride precipitation in the oxide layer during the initial stages of the pre-breakaway oxidation stage and the widespread porous oxide growth on the cladding surface in the latter post-BA oxidation stage are related to the oxidation mechanism in the air-steam mixture. The differences in the mechanism of oxidation of the cladding in dry air and air-steam mixtures are discussed based on the experimental results.
Nagai, Yuki; Shinaoka, Hiroshi*
Journal of the Physical Society of Japan, 92(3), p.034703_1 - 034703_8, 2023/03
no abstracts in English
Constantini, J.-M.*; Ogawa, Tatsuhiko
Quantum Beam Science (Internet), 7(1), p.7_1 - 7_16, 2023/03
Sputtering, emission of constituent atoms or molecules of materials induced by irradiation, is regarded as one of standard engineering techniques. According to some experimental data, emission of atoms whose direction is anti-parallel to incident radiation momentum was found among the sputtered atoms. Based on the standard approach, the thermal-spike model, atoms are evaporated by equillibrated thermal canonical ensemble resulted in by heat propagation therefore emission must be isotropic. Inspired by the fact that ionizations induced by ion irradiation are arranged linearly along the ion path, and the electric repulsion force between the ionizations tend to be parallel to irradiation axis, we developed an alternative approach in this study to explain the anisotropic emission. Using the spatial configuration of the irradiation-induced positive ions calculated by track-structure calculation code RITRACKS, the momentum of ions driven by the electric force was calculated. The calculated result explains the inverse jet of ions in case of 1 MeV proton and 1 MeV/u carbon ion irradiation to water. Moreover, the calculated sputtering yield also agrees with earlier experimental data.
Lobzenko, I.; Wei, D.*; Itakura, Mitsuhiro; Shiihara, Yoshinori*; Tsuru, Tomohito
Results in Materials (Internet), 17, p.100364_1 - 100364_7, 2023/03
High-entropy alloys (HEAs) have received attention for their excellent mechanical and thermodynamic properties. A recent study revealed that Co-free face-centered cubic HEAs carried a potential to improve strength and ductility, which is of high importance for nuclear materials. Here, we implemented first-principles calculations to explore the fundamental mechanism of improving mechanical properties in Co-free HEA. We found that the local lattice distortion of Co-free HEA is more significant than that of the well-known Cantor alloy. In addition, the short-range order formation in Co-free HEA caused highly fluctuated stacking fault energy. Thus, the significant local lattice distortion and the non-uniform solid solution states composed of low- and high-stacking fault regions contribute to improving strength and ductility.
Yanagisawa, Hiroshi; Umeda, Miki; Motome, Yuiko; Murao, Hiroyuki
JAEA-Technology 2022-030, 80 Pages, 2023/02
Nuclear criticality benchmark analyses were carried out for TRIGA-type reactor systems in which uranium-zirconium hydride fuel rods are loaded by using the continuous-energy Monte Carlo code MVP with the evaluated nuclear data library JENDL-5. The analyses cover two sorts of benchmark data, the IEU-COMP-THERM-003 and IEU-COMP-THERM-013 in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook, and effective neutron multiplication factors, reactivity worths for control rods etc. were calculated by JENDL-5 in comparison with those by the previous version of JENDL. As the results, it was confirmed that the effective neutron multiplication factors obtained by JENDL-5 were 0.4 to 0.6% greater than those by JENDL-4.0, and that there were no significant differences in the calculated reactivity worths by between JENDL-5 and JENDL-4.0. Those results are considered to be helpful for the confirmation of calculation accuracy in the analyses on NSRR control rod worths, which are planned in the future.
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
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.
Tada, Kenichi; Yamamoto, Akio*; Kunieda, Satoshi; Nagaya, Yasunobu
JAEA-Data/Code 2022-009, 208 Pages, 2023/02
The nuclear data processing code has an important role to connect evaluated nuclear data libraries and neutronics calculation codes. Japan Atomic Energy Agency (JAEA) has developed the nuclear data processing code FRENDY since 2013 to generate cross section files from evaluated nuclear data libraries, such as JENDL, ENDF/B, JEFF, and TENDL. The first version of FRENDY was released in 2019. FRENDY version 1 generates ACE files which are used for continuous energy Monte Carlo codes such as PHITS, Serpent, and MCNP. FRENDY version 2 generates multi-group neutron cross-section files from ACE files. The other major improvements are as follows: (1) uncertainty quantification for the probability tables of the unresolved resonance cross-section; (2) perturbation of the ACE file for the uncertainty quantification using a continuous Monte Carlo code; (3) modification of the ENDF-6 formatted nuclear data file. This report describes an overview of the nuclear data processing methods and input instructions for FRENDY.
Li, W.*; Yamada, Shinya*; Hashimoto, Tadashi; Okumura, Takuma*; Hayakawa, Ryota*; Nitta, Kiyofumi*; Sekizawa, Oki*; Suga, Hiroki*; Uruga, Tomoya*; Ichinohe, Yuto*; et al.
Analytica Chimica Acta, 1240, p.340755_1 - 340755_9, 2023/02
no abstracts in English
Yabuuchi, Kiyohiro*; Suzudo, Tomoaki
Journal of Nuclear Materials, 574, p.154161_1 - 154161_6, 2023/02
In nuclear materials, irradiation defects cause degradation of mechanical properties. In these materials, the relationship between dislocations and voids is particularly important for mechanical strength. Although only spherical voids have been studied in the past, this study focuses on faceted voids, which are observed simultaneously with spherical voids. In the current study, molecular dynamics was used to analyze the effect of faceted voids in the irradiation hardening of pure iron. Specifically, we clarified the difference in obstacle strength and interaction processes between spherical voids and faceted voids, and that even faceted voids show differences in interaction depending on their crystallographic arrangement with dislocations.
Kinoshita, Norikazu*; Noto, Takuma*; Nakajima, Hitoshi*; Kosako, Kazuaki*; Kato, Takahiro*; Kuroiwa, Yoichi*; Kurabe, Misako*; Sasaki, Yuki*; Torii, Kazuyuki*; Maeda, Makoto; et al.
Journal of Radioanalytical and Nuclear Chemistry, 332(2), p.479 - 486, 2023/02
Kaku Deta Nyusu (Internet), (134), p.32 - 33, 2023/02
no abstracts in English
Kaku Deta Nyusu (Internet), (134), p.34 - 45, 2023/02
On the occasion of the 60th anniversary of the founding of the Sigma Committee in the Atomic Energy Society of Japan, we will attempt to record and preserve its history as a reference for future activities.
Chong, Y.*; Gholizadeh, R.*; Tsuru, Tomohito; Zhang, R.*; Inoue, Koji*; Gao, W.*; Godfrey, A.*; Mitsuhara, Masatoshi*; Morris, J. W. Jr.*; Minor, A. M.*; et al.
Nature Communications (Internet), 14, p.404_1 - 404_11, 2023/02
Interstitial oxygen embrittles titanium, particularly at cryogenic temperatures, which necessitates a stringent control of oxygen content in fabricating titanium and its alloys. Here, we propose a structural strategy, via grain refinement, to alleviate this problem. Compared to a coarse-grained counterpart that is extremely brittle at 77K, the uniform elongation of an ultrafine-grained (UFG) microstructure (grain size 2.0 m) in Ti-0.3wt.%O was successfully increased by an order of magnitude, maintaining an ultrahigh yield strength inherent to the UFG microstructure. This unique strength-ductility synergy in UFG Ti-0.3wt.%O was achieved via the combined effects of diluted grain boundary segregation of oxygen that helps to improve the grain boundary cohesive energy and enhanced dislocation activities that contribute to the excellent strain hardening ability. The present strategy could not only boost the potential applications of high strength Ti-O alloys at low temperatures, but could also be applied to other alloy systems, where interstitial solution hardening results into an undesirable loss of ductility.
Asahi, Yuichi; Onodera, Naoyuki; Hasegawa, Yuta; Shimokawabe, Takashi*; Shiba, Hayato*; Idomura, Yasuhiro
Boundary-Layer Meteorology, 34 Pages, 2023/01
We develop a Transformer-based deep learning model to predict the plume concentrations in the urban area under uniform flow conditions. Our model has two distinct input layers: Transformer layers for sequential data and convolutional layers in convolutional neural networks (CNNs) for image-like data. Our model can predict the plume concentration from realistically available data such as the time series monitoring data at a few observation stations and the building shapes and the source location. It is shown that the model can give reasonably accurate prediction with orders of magnitude faster than CFD simulations. It is also shown that the exactly same model can be applied to predict the source location, which also gives reasonable prediction accuracy.