Some considerations on the dependence to numerical schemes of Lagrangian radionuclide transport models for the aquatic environment

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

https://doi.org/10.1016/j.jenvrad.2023.107138

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.

Attention-based time series analysis for data-driven anomaly detection in nuclear power plants

Dong, F.*; Chen, S.*; Demachi, Kazuyuki*; Yoshikawa, Masanori; Seki, Akiyuki; Takaya, Shigeru

Nuclear Engineering and Design, 404, p.112161_1 - 112161_15, 2023/04

https://doi.org/10.1016/j.nucengdes.2023.112161

R&D and maintenance management of the WASTEF Facility (FY2021)

Sano, Naruto; Yamashita, Naoki; Hishino, Kazutoyo*; Tsukada, Manabu*; Sawauchi, Fumiya*; Otake, Yoshinori*; Ichise, Kenichi; Tagami, Susumu

JAEA-Technology 2022-034, 47 Pages, 2023/03

JAEA-Technology-2022-034.pdf:2.81MB

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.

Investigation of the oxidation behavior of Zircaloy-4 cladding in a mixture of air and steam

Nemoto, Yoshiyuki; Ishijima, Yasuhiro; Kondo, Keietsu; Fujimura, Yuki; Kaji, Yoshiyuki

Journal of Nuclear Materials, 575, p.154209_1 - 154209_19, 2023/03

https://doi.org/10.1016/j.jnucmat.2022.154209

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.

Coulomb spike modelling of ion sputtering of amorphous water ice

Constantini, J.-M.*; Ogawa, Tatsuhiko

Quantum Beam Science (Internet), 7(1), p.7_1 - 7_16, 2023/03

https://doi.org/10.3390/qubs7010007

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.

Nuclear criticality benchmark analyses on TRIGA-type reactor systems by using continuous-energy Monte Carlo code MVP with JENDL-5

Yanagisawa, Hiroshi; Umeda, Miki; Motome, Yuiko; Murao, Hiroyuki

JAEA-Technology 2022-030, 80 Pages, 2023/02

JAEA-Technology-2022-030.pdf:2.57MB

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 data processing code FRENDY version 2

Tada, Kenichi; Yamamoto, Akio*; Kunieda, Satoshi; Nagaya, Yasunobu

JAEA-Data/Code 2022-009, 208 Pages, 2023/02

JAEA-Data-Code-2022-009.pdf:3.87MB

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.

Interaction between an edge dislocation and faceted voids in body-centered cubic Fe

Yabuuchi, Kiyohiro*; Suzudo, Tomoaki

Journal of Nuclear Materials, 574, p.154161_1 - 154161_6, 2023/02

https://doi.org/10.1016/j.jnucmat.2022.154161

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.

60th Anniversary of the Sigma Committee; Thoughts on the 60th Anniversary of the Sigma Committee

Iwamoto, Osamu

Kaku Deta Nyusu (Internet), (134), p.32 - 33, 2023/02

no abstracts in English

Grain refinement in titanium prevents low temperature oxygen embrittlement

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

https://doi.org/10.1038/s41467-023-36030-0

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.