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Koyama, Shinichi; Ikeuchi, Hirotomo; Mitsugi, Takeshi; Maeda, Koji; Sasaki, Shinji; Onishi, Takashi; Tsai, T.-H.; Takano, Masahide; Fukaya, Hiroyuki; Nakamura, Satoshi; et al.
Hairo, Osensui, Shorisui Taisaku Jigyo Jimukyoku Homu Peji (Internet), 216 Pages, 2023/11
In FY 2021 and 2022, JAEA perfomed the subsidy program for "the Project of Decommissioning and Contaminated Water Management (Development of Analysis and Estimation Technology for Characterization of Fuel Debris (Development of Technologies for Enhanced Analysis Accuracy, Thermal Bahavior Estimation, and Simplified Analysis of Fuel Debris)" started in FY 2021. This presentation material summarized the results of the project, which will be available shortly on the website of Management Office for the Project of Decommissioning, Contaminated Water and Treated Water Management.
Koyama, Shinichi; Nakagiri, Toshio; Osaka, Masahiko; Yoshida, Hiroyuki; Kurata, Masaki; Ikeuchi, Hirotomo; Maeda, Koji; Sasaki, Shinji; Onishi, Takashi; Takano, Masahide; et al.
Hairo, Osensui Taisaku jigyo jimukyoku Homu Peji (Internet), 144 Pages, 2021/08
JAEA performed the subsidy program for the "Project of Decommissioning and Contaminated Water Management (Development of Analysis and Estimation Technology for Characterization of Fuel Debris (Development of Technologies for Enhanced Analysis Accuracy and Thermal Behavior Estimation of Fuel Debris))" in 2020JFY. This presentation summarized briefly the results of the project, which will be available shortly on the website of Management Office for the Project of Decommissioning and Contaminated Water Management.
Kotake, Shoji*; Chikazawa, Yoshitaka; Takaya, Shigeru; Otaka, Masahiko; Kubo, Shigenobu; Arai, Masanobu; Kunogi, Kosuke; Ito, Takaya*; Yamaguchi, Akira*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
A maintenance management required to prototype nuclear power reactors is proposed. Monitoring and control of sodium impurity and thermal transient are extremely important for sodium boundary maintenance for sodium-cooled fast reactors. At the fast stage of the prototype reactor Monju operation, degradation mechanism on the piping should be demonstrated based on operation experiences. Therefore inspection on a representative position for crack indication and pipe thickness is proposed. Due to less experience of SFR plants, early detection of boundary failure is considered. For a matured operation stage, when degradation mechanism is well demonstrated based on inspection data, inspection cycle could be extended. And for commercial reactors, maintenance without inspection will be established based on accumulated operation experiences including those of the prototype reactor Monju.
Arai, Masanobu; Kunogi, Kosuke; Aizawa, Kosuke; Chikazawa, Yoshitaka; Takaya, Shigeru; Kubo, Shigenobu; Kotake, Shoji*; Ito, Takaya*; Yamaguchi, Akira*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
Applications for maintenance program on piping support of prototype fast breeder reactor Monju are studied. Based on degradation mechanism, snubbers in Monju primary cooling system showed lifetime more than the plant lifetime of 30 years by experiments conservatively. For the first step during construction, visual inspection on accessible all supports could be available. In that visual inspection, mounting conditions and damages of all accessible supports could be monitored. One of major features of the Monju primary piping system is large thermal expansion due to large temperature difference between maintenance and operation conditions. Thanks to that large thermal expansion, integrity of the piping support could be monitored by measuring piping displacement. When technologies of piping displacement monitoring are matured in Monju, visual inspection on piping support could be shifted to piping displacement monitoring. At that stage, the visual inspection could be limited only on representative supports.
Tada, Hiroyuki*; Kumasaka, Hiroo*; Saito, Akira*; Nakaya, Atsushi*; Ishii, Takashi*; Fujita, Tomoo; Sugita, Yutaka; Nakama, Shigeo; Sanada, Masanori*
Doboku Gakkai Rombunshu, F2 (Chika Kukan Kenkyu) (Internet), 73(1), p.11 - 28, 2017/03
This study examined the mechanical characteristics of rock segments and backfill materials and analyzed the stability of the drift that is supported by the rock segments and gravel backfill. The results confirmed the technical aspects of the formation of the rock segments and the effectiveness of the planned efforts to further reduce the amount of cement used.
Fukaya, Masaaki*; Hata, Koji*; Akiyoshi, Kenji*; Sato, Shin*; Takeda, Nobufumi*; Miura, Norihiko*; Uyama, Masao*; Kaneda, Tsutomu*; Ueda, Tadashi*; Hara, Akira*; et al.
JAEA-Technology 2016-002, 195 Pages, 2016/03
JAEA-Technology-2016-002.pdf:46.3MB
JAEA-Technology-2016-002-appendix(CD-ROM).zip:16.11MB
The researches on examination of the plug applied to the future reflood test was conducted as a part of (5) development of technologies for restoration and/on reduction of the excavation damage relating to the engineering technology in the MIU (2014), specifically focused on (1) plug examination (e.g. functions, structure and material) and the quality control methods and (2) analytical evaluation of rock mass behavior around the plug through the reflood test. As the result, specifications of the plug were determined. These specifications should be able to meet requirements for the safety structure and surrounding rock mass against predicted maximum water pressure, temperature stress and seismic force, and for controlling the groundwater inflow, ensuring the access into the reflood gallery and the penetration performance of measurement cable. Also preliminary knowledge regarding the rock mass behavior around the plug after flooding the reflood gallery by installed plug was obtained.
Katata, Genki; Kajino, Mizuo*; Matsuda, Kazuhide*; Takahashi, Akira*; Nakaya, Ko*
Atmospheric Environment, 97, p.501 - 510, 2014/11
Times Cited Count：15 Percentile：37.24(Environmental Sciences)To investigate the impact of hygroscopic growth on dry deposition onto forest canopies, numerical simulations of PM2.5 sulfate deposition using a multi-layer atmosphere-SOiL-VEGetation model (SOLVEG) ware performed. The scheme of particle dry deposition in SOLVEG was extended for application to a broad-leaved forest. An aerosol hygroscopic model based on the widely used 
-K
hler theory was incorporated into the model to calculate water uptake by the aerosols. The model accurately reproduced essential turbulent exchange fluxes (momentum, heat, and water vapor) over the canopies and the soil temperature and moisture for a deciduous broad-leaved forest in central Japan. Temporal variations in the measured PM2.5 sulfate deposition velocity were generally reproduced by the model. By considering an increase in particle diameter due to hygroscopic growth, the prediction accuracy of the modeled deposition velocity under humid conditions was improved. Numerical experiments for varying aerosol size distributions and hygroscopic properties showed that the geometric mean diameter and hygroscopicity of particles have a large influence on hygroscopic growth levels. The results also suggested that the deposition velocity of wet particles increased due to hygroscopic growth when the relative humidity (RH) was approximately 50%, and that the velocity reached five times greater than that under dry conditions when RH exceeded 95%.
Tada, Hiroyuki*; Kumasaka, Hiroo*; Saito, Akira*; Nakaya, Atsushi*; Ishii, Takashi*; Sanada, Masanori; Noguchi, Akira*; Kishi, Hirokazu*; Nakama, Shigeo; Fujita, Tomoo
Dai-13-Kai Iwa No Rikigaku Kokunai Shimpojiumu Koen Rombunshu (CD-ROM), p.133 - 138, 2013/01
The authors have been developing methods for constructing tunnels using the minimum quantities of cement-type support materials in high-level radioactive waste disposal facilities and advancing research and development about the technical formation of rock segment using low alkaline mortar. In this study, the mechanical characteristic values concerning the rock segment and backfill materials were examined. The stability analysis of tunnel supported by the rock segment and backfilling with gravel were performed. Technical formation and effectiveness of the alternative supports planned for further reduction in cement influence was confirmed from a study result above-mentioned.
Wada, Ken*; Hyodo, Toshio*; Yagishita, Akira*; Ikeda, Mitsuo*; Osawa, Satoshi*; Shidara, Tetsuo*; Michishio, Koji*; Tachibana, Takayuki*; Nagashima, Yasuyuki*; Fukaya, Yuki; et al.
European Physical Journal D, 66(2), p.37 - 40, 2012/02
Times Cited Count：41 Percentile：85.00(Optics)no abstracts in English
Hayashi, Katsuhiko; Noguchi, Akira; Kishi, Hirokazu; Kabayashi, Yasushi*; Nakama, Shigeo; Fujita, Tomoo; Naito, Morimasa; Tada, Hiroyuki*; Kumasaka, Hiroo*; Goke, Mitsuo*; et al.
JAEA-Research 2010-057, 101 Pages, 2011/03
JAEA-Research-2010-057.pdf:7.47MB
Cement-type materials that are used for supports or grouting at high-level radioactive waste disposal facilities leach into the groundwater and create a highly alkaline environment. Of concern in highly alkaline environments are the alteration of bentonite used as buffers or backfill materials, and of surrounding rock mass, and the increased uncertainty regarding the provision of performance of the disposal system over a long period of time. In this study, to reduce the quantity of cement-type materials that cause highly alkaline environments, technical feasibility of the support structure including the materials which considered the long-term performance of the HLW disposal system are discussed by using knowledge and technology accumulated in JAEA and Shimizu Construction. Moreover, based on the results, the problems remained in the application to the future HLW disposal institution are summarized.
Shimizu, Ryo; Takaya, Akikazu; Shirozu, Hidetomo; Uchida, Naoki; Tanaka, Kosuke; Yamaguchi, Akira; Aoto, Kazumi; Watahiki, Masaru
Nihon Genshiryoku Gakkai Wabun Rombunshi, 4(3), p.203 - 212, 2005/09
It was occurred that process solution leaked from the steam jet in the acid recovery process of Tokai reprocessing plant. So we retrieved the jet from the cell and observed the inside. As a result, it was revealed that a defect by corrosion, which penetrated the inside of the structure, arose from the steam nozzle root in the suction room of the jet. To ascertain why such failure occurred, we carried out observation of metallurgical structure of the jet, an analysis of temperature distribution of the structure involving verification using mock-ups and evaluation of material by corrosion test. Consequently, the origin of the corrosion defect was pin hole which was occurred on the surface of the end of the material at the steam nozzle root, where nitric acid fluid was contacted, and then the defect progressed along the temperature of about 120 of the structure. Anisotropy of the material, possibly originating in the manufacturing process, was observed from the result of evaluation of material by corrosion test. Finally, we devised a new design of improved corrosion resistance on the result of the whole investigations.
Shirozu, Hidetomo; Shimizu, Ryo; Takaya, Akikazu; Uchida, Naoki; Tanaka, Kosuke; Yamaguchi, Akira; Aoto, Kazumi
Saikuru Kiko Giho, (21), p.41 - 55, 2003/00
The cause of the failure, which had occurred in the steam jet (273J202) in the acid recovery process of Tokai reprocessing plant, was investigated.
Raman, S.*; Nestor Jr., C. W.*; Ichihara, Akira; Trzhaskovskaya, M. B.*
Physical Review C, 66(4), p.044312_1 - 044312_23, 2002/10
Times Cited Count：54 Percentile：89.85(Physics, Nuclear)no abstracts in English
Kato, Takashi; Tsuji, Hiroshi; Ando, Toshinari; Takahashi, Yoshikazu; Nakajima, Hideo; Sugimoto, Makoto; Isono, Takaaki; Koizumi, Norikiyo; Kawano, Katsumi; Oshikiri, Masayuki*; et al.
Fusion Engineering and Design, 56-57, p.59 - 70, 2001/10
Times Cited Count：18 Percentile：75.38(Nuclear Science & Technology)no abstracts in English
Tsuji, Hiroshi; Okuno, Kiyoshi*; Thome, R.*; Salpietro, E.*; Egorov, S. A.*; Martovetsky, N.*; Ricci, M.*; Zanino, R.*; Zahn, G.*; Martinez, A.*; et al.
Nuclear Fusion, 41(5), p.645 - 651, 2001/05
Times Cited Count：58 Percentile：82.58(Physics, Fluids & Plasmas)no abstracts in English
Nishiyama, Yutaka; Fukaya, Kiyoshi; Onizawa, Kunio; Suzuki, Masahide; Nakamura, Terumi*; ; Sato, Akira*; Yoshida, Kazuo*
Small Specimen Test Techniques (ASTM STP 1329), 0, p.484 - 494, 1998/00
no abstracts in English
Nakamura, Terumi*; ; Yoshida, Kazuo*; Sato, Akira*; Onizawa, Kunio; Nishiyama, Yutaka; Fukaya, Kiyoshi; Suzuki, Masahide
Ishikawajima Harima Giho, 36(2), p.91 - 97, 1996/03
no abstracts in English
Nishiyama, Yutaka; Fukaya, Kiyoshi; Onizawa, Kunio; Suzuki, Masahide; Shibata, Katsuyuki; ; Nakamura, Terumi*; Sato, Akira*; Yoshida, Kazuo*
JAERI-Tech 96-003, 30 Pages, 1996/02
no abstracts in English
Onizawa, Kunio; Fukaya, Kiyoshi; Nishiyama, Yutaka; Suzuki, Masahide; Shibata, Katsuyuki; Kosaka, Atsuo; Kaihara, Shoichiro*; Nakamura, Terumi*; Sato, Akira*; Yoshida, Kazuo*; et al.
JAERI-Tech 94-017, 58 Pages, 1994/09
Regulatory codes require the surveillance test to evaluate the irradiation embrittlement of reactor pressure vessel during operation. However , it is anticipated that the number of those specimens is insufficient in case plant life is extended. Reconstitution techniques by electron beam weld, laser weld, arc stud weld as well as surface-activated joining (SAJ) have been investigated for the re-use of un-deformed parts from tested Charpy impact specimen. The important items for the reconstitution technique are to reduce the width of heat affected zone to maximize the material available, and to lower the maximum temperature of specimen during joining process to preclude the recovery of radiation damage. SAJ can be achieved from a removal pf surface contamination by rotating one-side specimen in vacuum with applying modest friction force. Therefore, SAJ method is expected to be suitable for specimen reconstitution in view of material heating and melting. This paper describes preliminary study to develop Charpy specimen reconstitution technique using reactor pressure vessel steel, A533B-1, by SAJ method. Test results showed that the SAJ method had a capability of joining affected zone less than 1.5mm in half width, and over-temperature region less than 3 mm in half width above reactor operating temperature during joining. It was also found that transition temperature from reconstituted Charpy specimen could be evaluated.
