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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
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.
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.
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
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.
Kondo, Masahiro*; Yoshimoto, Tatsuya*; Ishikawa, Jun; Okamoto, Koji*
Hozengaku, 15(4), p.79 - 85, 2017/01
no abstracts in English
Daido, Hiroyuki
Hozengaku, 15(3), p.20 - 25, 2016/10
Naraha Remote Technology Development Center is open for various users to contribute to recovery of the coast area of Fukushima as well as the decommissioning of Fukushima Daiichi Nuclear Power Station. The center is located within a distance of 20 km from the Fukushima Daiichi station. This is the first development center funded by the Government near the Fukushima Daiichi. Many people expect that the center plays a significant role to contribute to the decommissioning of Fukushima Daiichi and recovery of Fukushima area from the hazards. The author describe details of the facility and our plan.
Yada, Hiroki
Hozengaku, 15(3), P. 86, 2016/10
no abstracts in English
Takeda, Masayasu
Hozengaku, 15(2), p.31 - 34, 2016/07
no abstracts in English
Nemoto, Yoshiyuki; Oishi, Makoto; Ito, Masayasu; Kaji, Yoshiyuki; Keyakida, Satoshi*
Hozengaku, 14(4), p.83 - 90, 2016/01
Authors previously reported that magnetic data obtained by using Eddy current method and AC magnetization method showed correlation with the increase of susceptibility of the irradiation assisted stress corrosion cracking (IASCC) on neutron irradiated austenitic stainless alloy specimens. To discuss the mechanism of the correlation, microstructure observation was conducted on the irradiated specimen, and magnetic permalloy phase (FeNi
) formation along grain boundary was revealed in this work. From this result, the radiation induced magnetic phase formation along grain boundary seems to be a factor of the magnetic property change of the irradiated materials, and related to the correlation between magnetic data and IASCC susceptibility. In addition, sensor probe development was conducted in this work to obtain higher sensitivity and resolution. It was applied for magnetic measurement on type304 stainless steel irradiated up to different doses. In this case, magnetic ferrite phase was existed in the type304 stainless steel sample before irradiation therefore it was concerned that magnetic measurement on the irradiated ones would be disturbed by the magnetic signal from the pre-existing ferrite phase. In the magnetic measurements, increase of the magnetic data was clearly seen on the irradiated specimens. Thus, it was thought that the developed magnetic measurement technics can be applied for the irradiated austenite stainless steels which contain certain quantity of ferrite phase before irradiation.
