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Journal Articles

Water experiments on thermal striping phenomena at the core outlet of an advanced sodium-cooled fast reactor, 1; Proposal of countermeasures to mitigate temperature fluctuations around control rods

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.

Journal Articles

Water experiments on thermal striping phenomena at the core outlet of an advanced sodium-cooled fast reactor, 2; Proposal of countermeasures to mitigate temperature fluctuations around radial blanket fuel assemblies

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.

Journal Articles

Observation of vibration characteristics of a cylindrical water tank by a phase-shifted optical pulse interference sensor

Morishita, Hideki*; Yoshida, Minoru*; Nishimura, Akihiko; Matsudaira, Masayuki*; Hirayama, Yoshiharu*; Sugano, Yuichi*

Hozengaku, 20(1), p.101 - 108, 2021/04

no abstracts in English

Journal Articles

Proposal of maintenance rationalization for next-generation fast reactors based on the analysis of the prolonged maintenance of the prototype fast-breeder reactor in Japan, "Monju", 1; Analysis of plant schedule of "Monju" in cold shutdown

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.

Journal Articles

Concerning aging of nuclear fuel material use facilities Examination of measures to improve safety assessment methods

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.

Journal Articles

Application of neutron, 2; Materials evaluation using the Japan Proton Accelerator Research Complex (J-PARC)

Morooka, Satoshi

Hozengaku, 19(1), p.29 - 34, 2020/04

no abstracts in English

Journal Articles

Application of neutron, 1; Neutron stress measurement in the research reactor JRR-3

Suzuki, Hiroshi

Hozengaku, 19(1), p.24 - 28, 2020/04

no abstracts in English

Journal Articles

Maintenance management of HTTR (Characteristics and achievements of maintenance management)

Shimazaki, Yosuke; Yamazaki, Kazunori; Iigaki, Kazuhiko

Hozengaku, 18(1), p.16 - 20, 2019/04

no abstracts in English

Journal Articles

Trend of high temperature gas-cooled reactor development in the world, international cooperation and strategy

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.

Journal Articles

CLADS for Fukushima Daiichi decommissioning

Okamoto, Koji

Hozengaku, 17(4), P. 1, 2019/01

no abstracts in English

Journal Articles

Development of a laser processing head for the inspection and repair of damages inside of a half-inch pipe

Komatsu, Kazumi*; Seki, Takeshi*; Naganawa, Akihiro*; Oka, Kiyoshi*; Nishimura, Akihiko

Hozengaku, 16(3), p.89 - 95, 2017/10

no abstracts in English

Journal Articles

Proposal of maintenance management of nuclear power plants at R&D stage by taking account of their features

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.

Journal Articles

Effectiveness evaluation of filtered containment venting system using THALES-2

Kondo, Masahiro*; Yoshimoto, Tatsuya*; Ishikawa, Jun; Okamoto, Koji*

Hozengaku, 15(4), p.79 - 85, 2017/01

no abstracts in English

Journal Articles

Full operation of JAEA Naraha Remote Technology Development Center

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.

Journal Articles

A Report on the meeting; 2016 ASME Pressure Vessels & Piping Conference (PVP2016)

Yada, Hiroki

Hozengaku, 15(3), P. 86, 2016/10

no abstracts in English

Journal Articles

Current status of Neutron Sciences at Research Reactor JRR-3

Takeda, Masayasu

Hozengaku, 15(2), p.31 - 34, 2016/07

no abstracts in English

Journal Articles

Study on magnetic property change on neutron irradiated austenitic stainless steel

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$$_{3}$$) 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.

Journal Articles

Development of high temperature magnetic sensor

Takaya, Shigeru; Arakawa, Hisashi*; Keyakida, Satoshi*

Hozengaku, 14(3), p.81 - 87, 2015/10

A magnetic sensor which can be applied to measurement at elevated temperature was newly developed. It is a kind of flux gate magnetic sensor. Permendur was employed for a magnetic core instead of Permalloy which is commonly used because Permendur has much higher Curie point, about 1000$$^{circ}$$C. Heat resistant ceramic coating Cu wires were used for coils. External magnetic field is detected by shift of peak position of differential permeability during induction of triangular excitation current. A magnetic core has race track shape with a fine part to make peak position more clear and increase detectability. The output of the developed sensor showed good linearity with external magnetic field even at 500$$^{circ}$$C. Furthermore, the durability of the sensor was discussed, and it was shown that decrease in coil performance after some operation at elevated temperature seems to be a critical issue.

Journal Articles

A Study of cladding technology on tube wall surface by a hand-held laser torch

Terada, Takaya; Nishimura, Akihiko; Oka, Kiyoshi; Moriyama, Taku*; Matsuda, Hiroyasu*

Hozengaku, 13(4), p.87 - 94, 2015/01

55 (Records 1-20 displayed on this page)