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Hirako, Kazuhito; Sawazaki, Hiromasa; Goto, Takehiro
Hozengaku, 23(1), p.9 - 13, 2024/04
The decommissioning of the prototype fast breeder reactor Monju, which started in 2018, has completed the extraction and solidification of the secondary sodium and the removal of the fuel body, which are the main processes in the first stage of decommissioning, and has moved to the second stage of decommissioning from 2023. This report provides an overview of the Monju decommissioning programme and the first stage of Monju decommissioning as "Trends of Monju decommissioning, Part 1". The actions in the second stage of Monju decommissioning will be explained in the next issue, including the outline and the review of the performance maintenance facilities that will change with the progress of the decommissioning and the construction of the maintenance programme accordingly.
Hashidate, Ryuta; Yada, Hiroki; Takaya, Shigeru; Enuma, Yasuhiro
Hozengaku, 23(1), p.103 - 111, 2024/04
In this paper, we propose a new method for extracting design issues on maintenance. Maintenance periods might be prolonged due to design issues. In the proposed method, maintenance preconditions are extracted by organizing the design information. A maintenance schedule is created by using extracted maintenance preconditions. If the created maintenance schedule doesn't achieve target periods, design issues could be extracted from the viewpoint of maintenance precondition. A simple example using Monju design information is presented to illustrate the proposed method.
Aizawa, Kosuke
Hozengaku, 22(3), p.70 - 71, 2023/10
no abstracts in English
Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki
Hozengaku, 20(3), p.89 - 96, 2021/10
Hot sodium from the fuel assembly can mix with cold sodium from the control rod (CR) channel and the blanket assemblies at the bottom plate of the Upper Internal Structure (UIS) of Advanced-SFR. Temperature fluctuation due to mixing of the fluids at different temperature between the core outlet and cold channel may cause high cycle thermal fatigue on the structure around the bottom of UIS. A water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of UIS. We focused on the temperature fluctuations near the primary and backup control rod channels, and studied the countermeasure structure to mitigate the temperature fluctuation through temperature distribution and flow velocity distribution measurements. As a result, effectiveness of the countermeasure to mitigate the temperature fluctuation intensity was confirmed.
Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki
Hozengaku, 20(3), p.97 - 101, 2021/10
Focusing on the thermal striping phenomena that occurs at a bottom of the internal structure of an advanced sodium-cooled fast reactor (Advanced-SFR) that has been designed by the Japan Atomic Energy Agency, a water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of Upper Internal Structure (UIS). In the previous paper, we reported the effect of measures to mitigate temperature fluctuations around the control rod channels. In this paper, the same test section was used, and a water experiment was conducted to obtain the characteristics of temperature fluctuations around the radial blanket fuel assembly. And the shape of the Core Instrumentation Support Plate (CIP) was modified, and it was confirmed that it was highly effective in alleviating temperature fluctuations around the radial blanket fuel assembly.
Toyota, Kodai; Hashidate, Ryuta; Takahashi, Keita; Yada, Hiroki; Takaya, Shigeru
Hozengaku, 20(2), p.95 - 103, 2021/07
Morishita, Hideki*; Yoshida, Minoru*; Nishimura, Akihiko; Matsudaira, Masayuki*; Hirayama, Yoshiharu*; Sugano, Yuichi*
Hozengaku, 20(1), p.101 - 108, 2021/04
no abstracts in English
Tsukada, Takashi; Soma, Yasutaka
Hozengaku, 19(4), p.37 - 44, 2021/01
Corrosion Cracking phenomena in JPDR (Japan Power Demonstration Reactor) the first Japanese Light Water Reactor is reviewed. This review describes two major cracking failure. The first was found during inspection in 1966 as the cracking failure on weld-overlay cladding at the inner wall of the top head. A series of analysis showed that some of the cracks reached the base metal across the weld boundaries and further penetrated into the vessel wall. Significant depletion of ferrite content was detected in manually welded part considered to assisted the cracking. These inspection result in improvement of the welding procedure and no similar failures have been reported in Japanese reactor. This mode of failure gave rise to a new research field studying the corrosion fatigue behavior of low alloy steel because of importance to assess pressure boundary of the reactor. The experiment of JPDR also contributed to the establishment of international cooperation for studying EAC (environmentally assisted cracking). The second failure was found in 1972 near the welded part between stainless piping and safe end. The extensive research concluded that this failure was caused by Stress Corrosion Cracking.
Hashidate, Ryuta; Toyota, Kodai; Takahashi, Keita; Yada, Hiroki; Takaya, Shigeru
Hozengaku, 19(4), p.115 - 122, 2021/01
In order to improve both safety and economic efficiency of a nuclear power plant, it is necessary to realize rational maintenance based on characteristics of the plant. The prototype fast-breeder reactor in Japan, Monju, spent most of the year for the maintenance. Thus, it is important to identify causes of the prolonged maintenance of Monju and to investigate countermeasures for implementation of rational maintenance of next-generation fast reactors. In this study, the authors investigated the causes of the prolonged maintenance of Monju during reactor cold shutdown based on the plant schedule of Monju. In addition, we proposed the maintenance optimization idea for next-generation fast reactors to solve the revealed issues.
Hashikura, Yasuaki*; Ishijima, Yasuhiro; Nakahara, Masaumi; Sano, Yuichi; Ueno, Fumiyoshi; Abe, Hitoshi
Hozengaku, 19(3), p.95 - 102, 2020/10
A plutonium concentrator was selected, and constant load tensile tests with controlled applied potentials and electrochemical tests were conducted in nitric acid and sodium nitrate solutions. From the results, a map which shows the effect of nitric acid concentration to crack initiation potential was drawn. And, it was pointed out that not only the nitric acid but also the nitrate ion coordinated to the nitrate must be considered in evaluating the possibility of stress corrosion cracking.
Sakamoto, Naoki; Fujishima, Tadatsune; Mizukoshi, Yasutaka
Hozengaku, 19(2), p.125 - 126, 2020/07
The five post-irradiation examination facilities in JAEA's Oarai research and development institute have been operated for over 40 years in order to investigate the irradiation performance of fast reactor fuel materials. The equipment associated with these facilities has been managed to maintain secure from the problems occurred in the process of aging. Therefore, we established a safety assessment method for aging facilities in 2002, and we have been conducting maintenance management of facilities since then. In this study, improvement plans of the safety assessment method are considered in order to solve the issues detected as a result of analysis of past maintenance information.
Morooka, Satoshi
Hozengaku, 19(1), p.29 - 34, 2020/04
no abstracts in English
Suzuki, Hiroshi
Hozengaku, 19(1), p.24 - 28, 2020/04
no abstracts in English
Shimazaki, Yosuke; Yamazaki, Kazunori; Iigaki, Kazuhiko
Hozengaku, 18(1), p.16 - 20, 2019/04
no abstracts in English
Nishihara, Tetsuo; Shibata, Taiju; Inaba, Yoshitomo
Hozengaku, 18(1), p.30 - 34, 2019/04
We explain the current status of High Temperature Gas-cooled Reactor (HTGR) development in the world and international cooperation between Japan Atomic Energy Agency (JAEA) and these countries. We introduce the concept of Japanese HTGR technology deployment by using international cooperation.
Okamoto, Koji
Hozengaku, 17(4), P. 1, 2019/01
no abstracts in English
Komatsu, Kazumi*; Seki, Takeshi*; Naganawa, Akihiro*; Oka, Kiyoshi*; Nishimura, Akihiko
Hozengaku, 16(3), p.89 - 95, 2017/10
no abstracts in English
Kondo, Masahiro*; Yoshimoto, Tatsuya*; Ishikawa, Jun; Okamoto, Koji*
Hozengaku, 15(4), p.79 - 85, 2017/01
no abstracts in English
Takaya, Shigeru; Chikazawa, Yoshitaka; Hayashida, Kiichi; Tagawa, Akihiro; Kubo, Shigenobu; Yamashita, Atsushi
Hozengaku, 15(4), p.71 - 78, 2017/01
A maintenance management suitable to nuclear power plants (NPP) at R&D stage was discussed. Objectives of maintenance management of NPP at R&D stage was first clarified. Next, applicability of codes for maintenance management of commercial NPP to NPP at R&D stage was discussed. Then, requirements and consideration for maintenance management of NPP at R&D stage was proposed. Finally, an example that the proposal was applied to setting maintenance program of sodium-cooled fast reactor was presented.
Yada, Hiroki
Hozengaku, 15(3), P. 86, 2016/10
no abstracts in English
