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Kamikawa, Yutaka; Suzuki, Makoto; Agake, Toshiki; Murakami, Takahiko; Morita, Yusuke; Shiina, Hidenori; Fukushima, Manabu; Hirane, Nobuhiko; Ouchi, Yasuhiro
JAEA-Technology 2023-030, 57 Pages, 2024/03
JAEA-Technology-2023-030.pdf:1.93MB
Owing to the publication of the latest data about aircraft crashes by Nuclear Regulation Authority (NRA), it was necessary to re-evaluate the probabilities of aircraft crashes for Nuclear Science Research Institute (NSRI). By using of the assessment method provided in "Regulatory Guide of the Assessment Standard for Probability of Airplane Crash on a Nuclear Power Reactor Facility", we re-evaluated the probabilities of aircraft crashes against the nuclear facilities in NSRI. As a result of the evaluations, the sum of the probabilities of aircraft crashes against Waste Treatment Facilities (maximum probability among all nuclear facilities in NSRI) is 5.68
10
(times/(reactor 
year)) which is lower than 10
(times/(reactor year)) that is the assessment criterion whether aircraft crashes is considered to be "anticipated external human induced events" in design basis or not.
Togawa, Orihiko; Hokama, Tomonori; Hiraoka, Hirokazu; Saito, Shota
JAEA-Research 2023-011, 78 Pages, 2024/03
JAEA-Research-2023-011.pdf:2.09MB
When radionuclides are released into the atmospheric environment at a nuclear emergency, protective measures such as evacuation and temporal relocation are carried out using motor vehicles such as private cars and buses to reduce radiation exposure to residents. To confirm conditions of contamination for the evacuated/relocated residents and the used motor vehicles, contamination inspection is conducted in the middle of the route from border areas of Nuclear Emergency Planning Zone to evacuation shelters. In the present inspection in Japan, a value of OIL4 = 40,000 cpm is used as decontamination criteria. For the details and derivation methods of the value, however, no official documents are found which give systematically detailed descriptions and explanation. It is also recognized that even few experts on nuclear emergencies can explain these subjects in detail as a whole. In order to explain scientifically and technically the OIL4 value of decontamination criteria used in contamination inspection in Japan, this report aims at investigating and estimating the deviation methods of OIL4, and examining and considering these results. To achieve the objectives, we show the bases for decontamination criteria, and investigate and estimate the derivation methods for limits of a surface contamination density corresponding to the generic criteria for each exposure pathway. Moreover, we give the OIL4 value some consideration and suggestions from a viewpoint of positioning and feature of OIL4 in Japan, and cautionary points at revising the value.
Nishi, Tsuyoshi*; Matsumoto, Saori*; Yamano, Hidemasa; Hayashi, Kiichiro*; Endo, Rie*; Bell
, M. R.*; Neubert, L.*; Volkova, O.*
Steel Research International, p.2300766_1 - 2300766_6, 2024/00
The density of Ni-based superalloys is measured using the maximum bubble pressure (MBP) method. The viscosity is evaluated using the oscillating crucible method. The surface tension is simultaneously measured using the MBP method.
, 
, and X-ray spectrum analysisOshima, Masumi*; Goto, Jun*; Hayakawa, Takehito*; Asai, Masato; Kin, Tadahiro*; Shinohara, Hirofumi*
Isotope News, (790), p.19 - 23, 2023/12
When analyzing samples that contain many radionuclides at various concentrations, such as radioactive waste or fuel debris, it is difficult to apply general spectrum analysis methods and is necessary to chemically separate each nuclide before quantifying it. The chemical separation is especially essential for analysis using a liquid scintillation counter (LSC). In this report, the authors explain the newly developed spectral determination method (SDM) in which the entire spectrum is fitted to quantify radioactivity of nuclides mixed in a sample. By applying the SDM to - and X-ray spectrum measured by LSC and -ray spectrum measured by Ge detector simultaneously, the authors demonstrated that radioactivity of 40 radionuclides mixed in a sample at concentrations varying by two orders could be quantified, which is useful to simplify chemical separation process in radionuclide quantification.
Tsuchiya, Harufumi; Toh, Yosuke; Ozu, Akira; Furutaka, Kazuyoshi; Kitatani, Fumito; Maeda, Makoto; Komeda, Masao
Journal of Nuclear Science and Technology, 60(11), p.1301 - 1312, 2023/11
Times Cited Count：1 Percentile：72.91(Nuclear Science & Technology)
Nb by activation method and half-life of 
Nb by mass analysisNakamura, Shoji; Shibahara, Yuji*; Endo, Shunsuke; Kimura, Atsushi
Journal of Nuclear Science and Technology, 60(11), p.1361 - 1371, 2023/11
Times Cited Count：1 Percentile：72.91(Nuclear Science & Technology)The thermal-neutron capture cross section (
) and resonance integral (I) for Nb among nuclides for decommissioning were measured by an activation method and the half-life of Nb by mass analysis. Niobium-93 samples were irradiated with a hydraulic conveyer installed in the research reactor in Institute for Integral Radiation and Nuclear Science, Kyoto University. Gold-aluminum, cobalt-aluminum alloy wires were used to monitor thermal-neutron fluxes and epi-thermal Westcott's indexes at an irradiation position. A 25-
m-thick gadolinium foil was used to sort out reactions ascribe to thermal-and epi-thermal neutrons. Its thickness provided a cut-off energy of 0.133 eV. In order to attenuate radioactivity of 
Ta due to impurities, the Nb samples were cooled for nearly 2 years. The induced radio activity in the monitors and Nb samples were measured by -ray spectroscopy. In analysis based on Westcott's convention, the and I values were derived as 1.11
0.04 barn and 10.50.6 barn, respectively. After the -ray measurements, mass analysis was applied to the Nb sample to obtain the reaction rate. By combining data obtained by both -ray spectroscopy and mass analysis, the half-life of Nb was derived as (2.000.15)10
years.
Maruyama, Shuhei; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 60(11), p.1372 - 1385, 2023/11
Times Cited Count：1 Percentile：72.91(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.
Nakamura, Shoji; Shibahara, Yuji*; Kimura, Atsushi; Endo, Shunsuke; Shizuma, Toshiyuki*
Journal of Nuclear Science and Technology, 60(9), p.1133 - 1142, 2023/09
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)In recent years, research has been advanced on lead-cooled fast reactors and accelerator drive systems, and it is required to improve the accuracy of the neutron capture cross section of Pb isotopes. Although 
Pb has a small natural abundance, it is of importance because it produces the long-lived radionuclide 
Pb (17.3 million years) by neutron capture reaction. However, it is difficult to measure its cross section by a conventional activation method using a nuclear reactor because the induced radioactivity of Pb is weak. Hence, the cross-section measurement was performed by applying mass spectrometry. This presentation gives the details of the experiment and the results obtained in the neutron capture cross-section measurement of Pb using mass spectroscopy.
Maeda, Makoto; Segawa, Mariko; Toh, Yosuke; Endo, Shunsuke; Nakamura, Shoji; Kimura, Atsushi
Journal of Radioanalytical and Nuclear Chemistry, 332(8), p.2995 - 2999, 2023/08
Times Cited Count：0 Percentile：0.01(Chemistry, Analytical)Kubo, Kotaro; Tanaka, Yoichi; Hakuta, Yuto*; Arake, Daisuke*; Uchiyama, Tomoaki*; Muramatsu, Ken
Mechanical Engineering Journal (Internet), 10(4), p.23-00051_1 - 23-00051_17, 2023/08
The significance of probabilistic risk assessments (PRAs) of nuclear power plants against external events was re-recognized after the Fukushima Daiichi Nuclear Power Plant accident. Regarding the seismic PRA, handling correlated failures of systems, components, and structures (SSCs) is very important because this type of failure negatively affects the redundancy of accident mitigation systems. The Japan Atomic Energy Research Institute initially developed a fault tree quantification methodology named the direct quantification of fault tree using Monte Carlo simulation (DQFM) to handle SSCs' correlated failures in detail and realistically. This methodology allows quantifying the top event occurrence probability by considering correlated uncertainties related to seismic responses and capacities with Monte Carlo sampling. The usefulness of DQFM has already been demonstrated. However, improving its computational efficiency would allow risk analysts to perform several analyses. Therefore, we applied quasi-Monte Carlo and importance sampling to the DQFM calculation of simplified seismic PRA and examined their effects. Specifically, the conditional core damage probability of a hypothetical pressurized water reactor was analyzed with some assumptions. Applying the quasi-Monte Carlo sampling accelerates the convergence of results at intermediate and high ground motion levels by an order of magnitude over Monte Carlo sampling. The application of importance sampling allows us to obtain a statistically significant result at a low ground motion level, which cannot be obtained through Monte Carlo and quasi-Monte Carlo sampling. These results indicate that these applications provide a notable acceleration of computation and raise the potential for the practical use of DQFM in risk-informed decision-making.
Kosaka, Wataru; Uchibori, Akihiro; Okano, Yasushi; Yanagisawa, Hideki*
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.1150 - 1163, 2023/08
The leakage of pressurized water from a steam generator (SG) and the progress after that are a key issue in the safety assessment or design of a SG in sodium-cooled fast reactor. The analysis code LEAP-III can evaluate a rate of water leakage during the long-term event progress, i.e., from the self-wastage initiated by an occurrence of a microscopic crack in a tube wall to the water leak detection and water/water-vapor blowdown. Since LEAP-III consists of semi-empirical formulae and one-dimensional equations of conservation, it has an advantage in short computation time. Thus, LEAP-III can facilitate the exploration of various new SG designs in the development of innovative reactors. However, there are several problems, such as an excessive conservative result in some case and the need for numerous experiments or preliminary analyses to determine tuning parameters of models in LEAP-III. Hence, we have developed a Lagrangian particle method code, which is characterized by a simpler computational principle and faster calculation. In this study, we have improved the existing particle pair search method for interparticle interaction in this code and developed an alternative model without the pair search. Through the trial analysis simulating in a tube bundle system, it was confirmed that new models reduced the computation time. In addition, it was shown that representative temperatures of the heat-transfer tubes evaluated by this particle method code, which is used to predict the tube failure in LEAP-III, were good agreement with that by SERAPHIM, which is a detailed mechanistic analysis method code.
Wakui, Takashi; Takagishi, Yoichi*; Futakawa, Masatoshi; Tanabe, Makoto*
Jikken Rikigaku, 23(2), p.168 - 174, 2023/06
Cavitation damage on the inner surface of the mercury target for the spallation neutron source occurs by proton bombarding in mercury. The prediction method of the cavitation damage using Monte Carlo simulations was suggested taking variability of the bubble core position and impact pressure distribution into account. The impact pressure distribution was estimated using the inverse analysis with Bayesian optimization was conducted with comparison between cavitation damage distribution obtained from experiment and the cumulative plastic strain distribution obtained from simulation. The average value and spread of maximum impact pressure estimated assuming the Gaussian distribution were 3.1 GPa and 1.2 m, respectively. Simulation results reproduced experimental results and it can be said that this evaluation method is useful.
Iwamoto, Toshihiro; Saito, Madoka*; Takahatake, Yoko; Watanabe, So; Watanabe, Masayuki; Naruse, Atsuki*; Tsukahara, Takehiko*
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 4 Pages, 2023/05
Applicability of temperature swing extraction technology employing monoamides was examined for uranium contaminated waste treatment procedure. Separation experiments on simulated target solution with three kinds of monoamides with different structure showed that Ce(IV) in the solution was selectively recovered by the temperature swing extraction operation. Based on the experiments, an appropriate monoamide for the procedure was selected.
Kurisaka, Kenichi
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
This study aims to understand a time trend of the occurrence rate of steam generator (SG) tube leak in the existing sodium-cooled fast reactors (SFRs) based on the observed data. The target on SFRs in the present paper is Phenix in France and BN600 in Russia. From the open literature review, we investigated the number of tube-to-tube plate weld, the number of tube-to-tube weld, heat transfer area of tube base metal, operating time of SGs, dates when SG tube leak occurred, leaked location, corrective action after tube leak such as replacement of leaked module. Based on these observed data, time to leak is estimated and then time trend of the occurrence rate of SG tube leak for each of the above-mentioned parts was quantitatively analyzed by the hazard plotting method. As a result, the rate of leak at tube-to-tube weld in Phenix shows increase with time due to probable cause of cyclic thermal stress in a short term. As for a long-term trend, the rate of tube leak in both Phenix and BN600 SGs indicated decrease with time probably thanks to improvement in welding and in SG operating condition and to removal of initial failure.
Yasue, Ayumu*; Kawakami, Mayu*; Kobayashi, Kensuke*; Kim, J.; Miyazu, Yuji*; Nishio, Yuhei*; Mukai, Tomohisa*; Morooka, Satoshi; Kanematsu, Manabu*
Quantum Beam Science (Internet), 7(2), p.15_1 - 15_14, 2023/05
Kubo, Kotaro; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Journal of Nuclear Science and Technology, 60(4), p.359 - 373, 2023/04
Times Cited Count：5 Percentile：78.52(Nuclear Science & Technology)Probabilistic risk assessment (PRA) is an essential approach to improving the safety of nuclear power plants. However, this method includes certain difficulties, such as modeling of combinations of multiple hazards. Seismic-induced flooding scenario includes several core damage sequences, i.e., core damage caused by earthquake, flooding, and combination of earthquake and flooding. The flooding fragility is time-dependent as the flooding water propagates from the water source such as a tank to compartments. Therefore, dynamic PRA should be used to perform a realistic risk analysis and quantification. This study analyzed the risk of seismic-induced flooding events by coupling seismic, flooding, and thermal-hydraulics simulations, considering the dependency between multiple hazards explicitly. For requirements of safety improvement, especially in light of the Fukushima Daiichi Nuclear Power Plant accident, sensitivity analysis was performed on the seismic capacity of systems, and the effectiveness of alternative steam generator injection by a portable pump was estimated. We demonstrate the use of this simulation-based dynamic PRA methodology to evaluate the risk induced by a combination of hazards.
Suzuki, Kenji*; Miura, Yasufumi*; Shiro, Ayumi*; Toyokawa, Hidenori*; Saji, Choji*; Shobu, Takahisa; Morooka, Satoshi
Zairyo, 72(4), p.316 - 323, 2023/04
Yokoyama, Kenji
EPJ Web of Conferences, 281, p.00004_1 - 00004_10, 2023/03
In Japan, development of adjusted nuclear data library for fast rector application based on the cross-section adjustment method has been conducted since the early 1990s. The adjusted library is called the unified cross-section set. The first version was developed in 1991 and is called ADJ91. Recently, the integral experimental data were further expanded to improve the design prediction accuracy of the core loaded with minor actinoids and/or degraded Pu. Using the additional integral experimental data, development of ADJ2017 was started in 2017. In 2022, the latest unified cross-section set AJD2017R was developed based on JENDL-4.0 by using 619 integral experimental data. An overview of the latest version with a review of previous ones will be shown. On the other hand, JENDL-5 was released in 2021. In the development of JENDL-5, some of the integral experimental data used in ADJ2017R were explicitly utilized in the nuclear data evaluation. However, this is not reflected in the covariance data. This situation needs to be considered when developing a unified cross-section set based on JENDL-5. Preliminary adjustment calculation based on JENDL-5 is performed using C/E (calculation/experiment) values simply evaluated by a sensitivity analysis. The preliminary results will be also discussed.
Kawamura, Sho; Kikuchi, Masanobu; Hosoya, Toshiaki
JAEA-Technology 2021-041, 103 Pages, 2023/02
JAEA-Technology-2021-041.pdf:8.7MB
In response to new regulatory standard for research and test reactor which is enforced December 2013, JRR-3 got license in November 2018 by formulate new design basis ground motion. After that we evaluated for insertion property of control rod using that new design basis ground motion, and that evaluation results were accepted as approval of the design and construction method by Nuclear Regulation Authority. Now, we re-evaluated to insertion property of control rod about neutron absorber and follower fuel element by time history response analysis method. In this report, it shows that new results have sufficiency of margin compared with the past results that are accepted as approval of the design and construction method.
