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Ichikawa, Tsubasa*; Hakoshima, Hideaki*; Inui, Koji*; Ito, Kosuke*; Matsuda, Ryo*; Mitarai, Kosuke*; Miyamoto, Koichi*; Mizukami, Wataru*; Mizuta, Kaoru*; Mori, Toshio*; et al.
Nature Reviews Physics (Internet), 6(6), p.345 - 347, 2024/06
Times Cited Count：19 Percentile：98.39(Physics, Applied)
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revealed by angle-resolved photoelectron spectroscopy and first-principles calculationSumida, Kazuki; Higaki, Sota*; Sato, Hitoshi*; Tsuru, Daichi*; Miyamoto, Koji*; Okuda, Taichi*; Kuroiwa, Yoshihiro*; Moriyoshi, Chikako*; Takase, Koichi*; Oguchi, Tamio*; et al.
Journal of the Physical Society of Japan, 92(8), p.084706_1 - 084706_6, 2023/08
Times Cited Count：1 Percentile：13.33(Physics, Multidisciplinary)Takamizawa, Hisashi; Lu, K.; Katsuyama, Jinya; Masaki, Koichi*; Miyamoto, Yuhei*; Li, Y.
JAEA-Data/Code 2022-006, 221 Pages, 2023/02
JAEA-Data-Code-2022-006.pdf:4.79MB
As a part of the structural integrity assessment research for aging light water reactor (LWR) components, a probabilistic fracture mechanics (PFM) analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed in Japan Atomic Energy Agency. The PASCAL code can evaluate failure probabilities and failure frequencies of core region in reactor pressure vessel (RPV) under transients by considering the uncertainties of influential parameters. The continuous development of the code aims to improve the reliability by introducing the analysis methodologies and functions base on the state-of-the-art knowledge in fracture mechanics and domestic data. In the first version of PASCAL, which was released in FY2000, the basic framework was developed for analyzing failure probabilities considering pressurized thermal shock events for RPVs in pressurized water reactors (PWRs). In PASCAL Ver. 2 released in FY 2006, analysis functions including the evaluation methods for embedded cracks and crack detection probability models for inspection were introduced. In PASCAL Ver. 3 released in FY 2010, functions considering weld-overlay cladding on the inner surface of RPV were introduced. In PASCAL Ver. 4 released in FY 2017, we improved several functions such as the stress intensity factor solutions, probabilistic fracture toughness evaluation models, and confidence level evaluation function by considering epistemic and aleatory uncertainties related to influential parameters. In addition, the probabilistic calculation method was also improved to speed up the failure probability calculations. To strengthen the practical applications of PFM methodology in Japan, PASCAL code has been improved since FY 2018 to enable PFM analyses of RPVs subjected to a broad range of transients corresponding to both PWRs and boiling water reactors, including pressurized thermal shock, low-temperature over pressure, and normal operational transients. In particular, the stress intensi
Kasugai, Yoshimi; Sato, Koichi; Takahashi, Kazutoshi*; Miyamoto, Yukihiro; Kai, Tetsuya; Harada, Masahide; Haga, Katsuhiro; Takada, Hiroshi
JPS Conference Proceedings (Internet), 33, p.011144_1 - 011144_6, 2021/03
A spallation neutron source with a mercury target has been in operation at the Materials and Life Science Experimental Facility of J-PARC since 2008. The target vessel made of stainless steel is required to be exchanged periodically due to radiation damage etc. In this presentation, tritium gas release observed in the first series of exchange work in 2011 and the analytical results will be shown.
Yamaguchi, Yoshihito; Mano, Akihiro; Katsuyama, Jinya; Masaki, Koichi*; Miyamoto, Yuhei*; Li, Y.
JAEA-Data/Code 2020-021, 176 Pages, 2021/02
JAEA-Data-Code-2020-021.pdf:5.26MB
In Japan Atomic Energy Agency, as a part of researches on the structural integrity assessment and seismic safety assessment of aged components in nuclear power plants, a probabilistic fracture mechanics (PFM) analysis code PASCAL-SP (PFM Analysis of Structural Components in Aging LWR - Stress Corrosion Cracking at Welded Joints of Piping) has been developed to evaluate failure probability of piping. The initial version was released in 2010, and after that, the evaluation targets have been expanded and analysis functions have been improved based on the state-of-the art technology. Now, it is released as Ver. 2.0. In the latest version, primary water stress corrosion cracking in the environment of Pressurized Water Reactor, nickel based alloy stress corrosion cracking in the environment of Boiling Water Reactor, and thermal embrittlement can be taken into account as target age-related degradation. Also, many analysis functions have been improved such as incorporations of the latest stress intensity factor solutions and uncertainty evaluation model of weld residual stress. Moreover, seismic fragility evaluation function has been developed by introducing evaluation methods including crack growth analysis model considering excessive cyclic loading due to large earthquake. Furthermore, confidence level evaluation function has been incorporated by considering the epistemic and aleatory uncertainties related to influence parameters in the probabilistic evaluation. This report provides the user's manual and analysis methodology of PASCAL-SP Ver. 2.0.
Te
Shikin, A. M.*; Estyunin, D. A.*; Klimovskikh, I. I.*; Filnov, S. O.*; Kumar, S.*; Schwier, E. F.*; Miyamoto, Koji*; Okuda, Taichi*; Kimura, Akio*; Kuroda, Kenta*; et al.
Scientific Reports (Internet), 10, p.13226_1 - 13226_13, 2020/08
Times Cited Count：86 Percentile：95.17(Multidisciplinary Sciences)Katsuyama, Jinya; Masaki, Koichi; Miyamoto, Yuhei*; Li, Y.
JAEA-Data/Code 2017-015, 229 Pages, 2018/03
JAEA-Data-Code-2017-015.pdf:5.8MBJAEA-Data-Code-2017-015(errata).pdf:0.15MB
As a part of the structural integrity research for aging light water reactor components, a probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed in JAEA. The PASCAL code can evaluate the conditional failure probabilities and failure frequencies for core region in reactor pressure vessels under the pressurized thermal shock events. In this study, we improved many functions such as the stress intensity factor solutions, the fracture toughness models, or confidence level evaluation function by considering epistemic and aleatory uncertainties related to influence parameters in the structural integrity assessment. We also developed the analysis module PASCAL-Manager which calculates the failure frequency for the entire core region taking into consideration the failure probabilities obtained from PACAL-RV. Based on these improvements, the new analysis code is upgraded to PASCAL Ver.4. This report provides the user's manual and theoretical background of PASCAL Ver.4.
Masaki, Koichi; Miyamoto, Yuhei*; Osakabe, Kazuya*; Uno, Shumpei*; Katsuyama, Jinya; Li, Y.
Proceedings of the ASME 2017 Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 7 Pages, 2017/07
A probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency (JAEA). PASCAL can evaluate failure frequencies of Japanese reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events based on domestic structural integrity assessment models and data of influence factors. In order to improve the engineering applicability of PFM to Japanese RPVs, we have performed verification of the PASCAL. In general, PFM code consists of many functions such as fracture mechanics evaluation functions, probabilistic evaluation functions including random variables sampling modules and probabilistic evaluation models, and so on. The verification of PFM code is basically difficult because it is impossible to confirm such functions through the comparison with experiments. When a PFM code is applied for evaluating failure frequencies of RPVs, verification methodology of the code should be clarified and it is important that verification results including the region and process of the verification of the code are indicated. In this paper, our activities of verification for PASCAL are presented. We firstly represent the overview and methodology of verification of PFM code, and then, some verification examples are provided. Through the verification activities, the applicability of PASCAL in structural integrity assessments for Japanese RPVs was confirmed with great confidence.
Masukawa, Fumihiro; Sato, Tatsuhiko; Sato, Koichi; Moriyasu, Hiroyuki; Nakazawa, Takashi; Miyamoto, Yukihiro; Nakashima, Hiroshi
Progress in Nuclear Science and Technology (Internet), 3, p.76 - 78, 2012/10
In high energy particle accelerator facilities, high energy neutron contribution to dose is the most interest as it has significant contribution to total dose, thus it determines the shielding design. Neutron dose rate were measured at Materials and Life Science Facility (MLF) of J-PARC, by two types of active detectors which can be applied in high energy neutron field: one is the modified type of ordinary rem meter, 
He propotional counter with polyethylene moderator added lead to improve high energy response, the other is the phoswitch-type scintillation detector, which consists of liquid organic scintillator BC501A coupled with ZnS(Ag) scintillation sheets doped with 
Li, developed as DARWIN (Dose monitoring system Applicable to various Radiations with WIde energy raNges) in JAEA. Neutron dose rate was measured on the pre-shield of the neutron beam line No.12 (BL12), which is located in the experimental hall No.1. The neutron dose rate by the modified rem meter was 
50 
Sv/h at the wall on 3 GeV-220 kW operation. Alhough the mercury target is well shielded to reduce the background radiation to the neutron experimental halls, such dose rate was observed as high energy neutrons from upstream beam loss caused by the muon target penetrate through proton beam tunnel. Neutron spectrum was also measured by DARWIN, showed that the high energy neutrons over 20 MeV have comparable contribution by those of 1 20 MeV.
Ito, Hideaki; Maeda, Shigetaka; Naito, Hiroyuki; Akiyama, Yoichi; Miyamoto, Kazuyuki; Ashida, Takashi; Noguchi, Koichi; Ito, Chikara; Aoyama, Takafumi
JAEA-Technology 2010-049, 129 Pages, 2011/03
JAEA-Technology-2010-049.pdf:6.99MB
The in-vessel gamma dose rate was measured in the experimental fast reactor Joyo to evaluate the activation of reactor structural components and the radiation exposure of the fiber scope used for in-vessel visual inspection. The measurement system, which requires a wide sensitivity range and high durability in a high-temperature environment, was specifically developed for use in the sodium cooled fast reactor. Using this system, the in-vessel gamma dose rate with cooling times of 450 and 720 days after reactor shutdown was measured in Joyo, which has been operated for 71,000 hours over approximately 30 years. The gamma dose rate was calculated using QAD-CGGP2 code with the gamma source intensity obtained by the ORIGEN2 code. The neutron flux used as input to the ORIGEN2 was evaluated by the Joyo dosimetry method. The ratio between the calculated and experimental values ranged from 1.1 to 2.4, confirming the accuracy of gamma dose rate and component activation calculation.
Matsuoka, Toshiyuki; Miyamoto, Tetsuo; Yabuuchi, Satoshi; Asamori, Koichi; Yamanaka, Yoshiaki*; Semba, Takeshi
Shadan Hojin Butsuri Tansa Gakkai Dai-121-Kai Gakujutsu Koenkai Koen Rombunshu, p.243 - 245, 2009/00
no abstracts in English
Osawa, Hideaki; Ota, Kunio; Hama, Katsuhiro; Sawada, Atsushi; Takeuchi, Shinji; Amano, Kenji; Saegusa, Hiromitsu; Matsuoka, Toshiyuki; Miyamoto, Tetsuo; Toyoda, Gakuji; et al.
JAEA-Research 2008-085, 742 Pages, 2008/11
JAEA-Research-2008-085-1.pdf:26.22MBJAEA-Research-2008-085-2.pdf:16.64MBJAEA-Research-2008-085-3.pdf:28.27MB
This report shows the results the project for the establishment of comprehensive site characterization technology, entrusted from Natural Resources and Energy Agency, Ministry of Economy, Trade and Industry Natural Resources and Energy Agency in 2007.
Oshima, Takeshi; Miyamoto, Haruki; Imaizumi, Mitsuru*; Morioka, Chiharu*; Kawakita, Shiro*; Shimazaki, Kazunori*; Kibe, Koichi*; Kawano, Katsuyasu*; Ito, Hisayoshi
JAEA-Review 2006-042, JAEA Takasaki Annual Report 2005, P. 6, 2007/02
no abstracts in English
Miyamoto, Haruki; Oshima, Takeshi; Imaizumi, Mitsuru*; Ito, Hisayoshi; Kibe, Koichi*; Kawano, Katsuyasu*
Proceedings of 2006 IEEE 4th World Conference on Photovoltaic Energy Conversion (WCPEC-4) (CD-ROM), p.1815 - 1817, 2006/05
no abstracts in English
Oshima, Takeshi; Miyamoto, Haruki; Imaizumi, Mitsuru*; Morioka, Chiharu*; Kawakita, Shiro*; Shimazaki, Kazunori*; Kibe, Koichi*; Kawano, Katsuyasu*; Ito, Hisayoshi
Proceedings of 2006 IEEE 4th World Conference on Photovoltaic Energy Conversion (WCPEC-4) (CD-ROM), p.1818 - 1821, 2006/05
no abstracts in English
Naito, Susumu*; Sano, Akira*; Izumi, Mikio*; Noda, Etsuo*; Hayashi, Kazuo*; Sato, Mitsuyoshi*; Suto, Osamu; Miyamoto, Yasuaki; Kondo, Shinichi*; Iinuma, Koichi*; et al.
Nuclear Science Symposium Conference Record, 2005 IEEE, Vol.1, p.495 - 499, 2005/10
An alpha radioactivity monitor (alpha-clearance monitor) is being developed to measure alpha radioactivity of uranium contaminated waste with large and complex surfaces. It uses an alpha radioactivity measurement method by collecting the air ionized by alpha particles using an air stream and by measuring its ion current. One of the major problems of the alpha-clearance monitor is the background ion current mainly by radon in atmosphere (700 fA), which is much larger than ion current of an alpha particle (several fA). It strongly influences on the detection limit of alpha radioactivity. In order to improve the detection limit, we developed a method to monitor the radon concentration inside the monitor without the additional radon detector. In addition, the influencing factors on the detection limit: ion mobility and ion recombination coefficient were measured and evaluated on various environmental conditions (temperature, humidity and aerosol density).
Ida, Mizuho*; Horiike, Hiroshi*; Akiba, Masato; Ezato, Koichiro; Iida, Toshiyuki*; Inoue, Shoji*; Miyamoto, Seiji*; Muroga, Takeo*; Nakamura, Hideo; Nakamura, Hiroshi*; et al.
Journal of Nuclear Materials, 307-311(Part2), p.1686 - 1690, 2002/12
Times Cited Count：6 Percentile：38.09(Materials Science, Multidisciplinary)no abstracts in English
Miyamoto, Yukihiro; Ikeno, Koichi; Akiyama, Shigenori*; Harada, Yasunori
JAERI-Tech 2002-086, 43 Pages, 2002/11
Description is given for the characteristic radiation environment for the High Intensity Proton Accelerator Facility and the design concept of the radiation control system of it. The facility is a large scale accelerator complex consisting of high energy proton accelerators carrying the highest beam intensity in the world and the related experimental facilities and therefore provides various issues relevant to the radiation environment. The present report describes the specifications for the radiation control system for the facility, determined in consideration of these characteristics.
Miyamoto, Yoshiaki; Ogawa, Masuro; Akino, Norio; Shiina, Yasuaki; Inagaki, Yoshiyuki; Shimizu, Saburo; Onuki, Kaoru; Ishiyama, Shintaro; Katanishi, Shoji; Takeda, Tetsuaki; et al.
JAERI-Review 2001-006, 147 Pages, 2001/03
JAERI-Review-2001-006.pdf:9.2MB
no abstracts in English
Masuyama, Koichi; Sato, Koichi; Yoshino, Koji; Tajima, Takahiro*; Seki, Kazunari; Kasugai, Yoshimi; Miyamoto, Yukihiro
no journal, ,
The mercury target vessel used in the Materials and Life Science Experimental Facility (MLF) in the Japan Accelerator Research Complex (J-PARC) needs to be replaced periodically. The replacement work, which is conducted in the hot cell, associates with release of radioactive materials produced via spallation reactions between protons and mercury. Therefore we monitored radioactivity of dust and air samples in the hot cell. By the monitoring, we newly found peculiar phenomena on the behavior of radioactive products, which is important for radiation control. In this work, we will show the monitoring data and the new observation.
