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Nemoto, Yoshiyuki; Oishi, Makoto; Ito, Masayasu; Kaji, Yoshiyuki
Nihon Hozen Gakkai Dai-12-Kai Gakujutsu Koenkai Yoshishu, p.105 - 112, 2015/07
Authors previously reported that Eddy current method and AC magnetization method have potential to be applied for development of diagnostic technics to detect the sign of material degradation before cracking on the austenitic stainless steels used as structural material in nuclear power plants. In typical austenitic stainless steels such as type304, magnetic ferrite phase would exist in the alloy before irradiation, and it is concerned to disturb the magnetic measurement on irradiated material. Magnetic measurements were conducted in this work on type304 austenitic stainless steel specimens irradiated up to different doses. In addition, microstructure observation was conducted on the area including grain boundary to discuss the correlation of magnetization on irradiated austenitic stainless alloy and grain boundary cracking. Obtained magnetic data on irradiated type304 stainless steel were seen clearly different from that on un-irradiated specimen, and showed positive correlation with radiation dose, therefore it was thought that magnetic measurement technics can be applied for the material which contains certain quantity of ferrite phase before irradiation. In the microstructural observation, magnetic phase (FeNi
) formation along the grain boundary was revealed.
Terada, Takaya; Nishimura, Akihiko; Oka, Kiyoshi; Toyama, Ryoji*; Kitsunai, Daisuke*
no journal, ,
A prototype of maintenance machine for heat exchanger tubes was developed. It was based on the bench top size device which had been specially designed to have dual functions for fiber inspection and laser cladding in JAEA. A laser processing head was improved mechanically in 2 motions of axial translation and circumferential rotation to insert the heat exchanger tubes. Laser scattering monitoring was also discussed for the purpose of detecting small defects at the inner wall of the heat exchanger tubes.
Kishi, Yoshiyuki; Yasuda, Takeshi; Tokoro, Hayate; Yamanaka, Atsushi; Tsutagi, Koichi; Shirato, Yoji; Tanaka, Hitoshi
no journal, ,
High active liquid waste is stored in the facility equipped with cooling system and hydrogen scavenging system because of hydrogen generation due to radiolysis, and heat generation due to decay heat. The facilities related to these systems maintenance have been designed to be fed by emergency power generators in the conventional design. In addition to that, in light of the lessons learned from accident at the Fukushima Daiichi Nuclear Power Plant, we took safety measures for emergencies such as securement of the power supply system from power source vehicle to enable a quick recovery in the case of station black out in the Tokai Reprocessing Plant.
Kaito, Yasuaki; Ashida, Takashi; Imaizumi, Kazuyuki; Izawa, Osamu; Naito, Hiroyuki
no journal, ,
In the experimental fast reactor Joyo, it was confirmed that the top of the irradiation test sub-assembly(S/A) of MARICO-2 (material testing rig with temperature control) had been bent onto the in-vessel storage rack as an obstacle, and the MARICO-2 S/A had been stuck in the transfer pot owing to its deformation. Since the MARICO-2 S/A cannot be retrieved with the existing handling equipment, special handling equipment was developed while considering the anomalous conditions. Emphasis had been placed on (1) how to precisely control the gripper position in the reactor vessel, and (2) how to retrieve the MARICO-2 S/A together with transfer pot with high reliability in the design and development of equipment. The MARICO-2 S/A was retrieved successfully together with transfer pot in 2014. This paper describes the in-vessel repair techniques for MARICO-2 S/A retrieval.
Shimizu, Shunji; Okuda, Eiji; Kikuchi, Yuki; Kawasaki, Toru; Owada, Ryohei
no journal, ,
no abstracts in English
Ashida, Takashi; Takamatsu, Misao; Ito, Hideaki; Okawa, Toshikatsu; Yoshihara, Shizuya
no journal, ,
In the experimental fast reactor Joyo, it was confirmed that the top of the irradiation test sub-assembly (S/A) of MARICO-2 (material testing rig with temperature control) had been broken and bent onto the in-vessel storage rack as an obstacle and had damaged the upper core structure (UCS). This incident necessitated the replacement of the UCS and the retrieval of the MARICO-2 sub-assembly for Joyo restoration. This paper describes the summary of Joyo restoration works. In-vessel repair techniques for sodium cooled fast reactors (SFRs) are important in confirming safety and integrity of SFRs. However, the demonstrated techniques under the actual reactor environment with high temperature, high radiation dose and remaining sodium are not enough to secure the reliability of these techniques. The experience and knowledge accumulated in the UCS replacement provides valuable insights into further improvements for in-vessel repair techniques in SFRs.
Ouchi, Masayuki; Hoshi, Takahiro; Sasaki, Shunichi; Isobe, Hiroyasu; Nagaoka, Shinichi; Kurabayashi, Kazuaki; Obu, Tomoyuki
no journal, ,
Plutonium nitrate solution is stored in tanks that have cooling system to prevent boiling accident by decay heat and scavenging system to prevent explosion of hydrogen generated by radiolysis. These systems are designed to be provided power from emergency power generators at the Tokai Reprocessing Plant (TRP) when commercial power is lost. In addition, some apparatus (mobile power generators and nitrogen gas cylinder to scavenge hydrogen, etc.) are recently deployed at TRP on the basis of power loss in Fukushima Daiichi nuclear accident. In this report, these safety measures for emergency are shown.
Chikazawa, Yoshitaka; Asayama, Tai; Kamide, Hideki
no journal, ,
Fast reactors are vital to achieve nuclear fuel cycle sustainability. Extensive experiences of Sodium cooled fast reactors (SFRs) have been accumulated with past and existing prototype reactors in Japan and the world. Those technologies are matured from the view point of industrial development. In Japan, an advanced loop type SFR (Japan Sodium-cooled reactor (JSFR)) was selected in the Fast Reactor Cycle Development Project (FaCT) project showing that it could meet with requirements on GIF concepts. Monju is now the only prototype loop type reactor in the world. Operation of Monju is desired from the view point of development of key technologies and maintenance/repair program of loop type SFR.
Watahiki, Kenji; Ishii, Takahiro; Sukigara, Koji; Sanyoshi, Hirotaka; Inami, Shinichi
no journal, ,
Treated water of The Tokai Reprocessing Plant used for cooling water and fire fighting water, is supplied by the steel pipes. These pipes are laid down under the ground or in the trench. Treated water had leaked, because the part of the underground pipe flange corrosion. The cause of this breakage was presumed that the pipe flange has been loaded the stress continuously. The stress generated depend of that the underground pipes sank with the ground subsidence by influence of the 2011 earthquake of the Pacific coast of Tohoku. The repair of the breakage flange and the pipes was adopted to use clamp (flexibility joint) for pipe connection as one of the countermeasures for tremble of the earthquake. This report shows the technical evaluation for the clamp performance on the practical use.
Kawasumi, Hiroyuki; Takeuchi, Kenji; Domura, Kazuyuki; Sanyoshi, Hirotaka; Inami, Shinichi
no journal, ,
The Tokai Reprocessing Plant has recovered nuclear materials from spent nuclear fuel. Reprocessing process is consisted of shearing process, a dissolution process and an extraction process etc. Each facilities designed to confine radioactive materials in controlled area to keep negative pressure using ventilation system. This system has been used over 40 years. We analyzed influences to the trouble of aging equipment, and we considered necessary additional check out system to prevent troubles based on the results of analysis. This reports describe the present status of this system.
Ebine, Noriya; Yamaguchi, Yoshihito; Katsuyama, Jinya; Nishida, Akemi; Li, Y.
no journal, ,
A probabilistic analysis code PASCAL-EC (PFM Analysis of Structural Components in Aging LWR - Erosion and Corrosion) have been improved for evaluating structural integrity of wall thinned pipes on the basis of the Monte Carlo method. In order to perform the fragility evaluation for wall-thinned pipes, several functions are introduced to PASCAL-EC in this work, such as probabilistic evaluation model to consider the uncertainty of seismic response stress, a simple evaluation model to considered the increase of the seismic stress due to wall thinning and failure evaluation methods for wall-thinned pipes. This paper provides some details of these models and preliminary results of fragility evaluation using improved PASCAL-EC code for wall-thinned pipes in nuclear power plants.
Uno, Shumpei; Katsuyama, Jinya; Katsumata, Genshichiro*; Masaki, Koichi*; Osakabe, Kazuya*; Li, Y.
no journal, ,
Assuring the structural integrity of a reactor pressure vessel (RPV) is known as one of the critical issues to maintain the safe long-term operation of a nuclear power plant. Authors have been conducting extensive research on the improvement of structural integrity assessment methods of RPVs. In this paper, we describe some research results obtained from the recent activity including the review on the technical background of the methods and study on probabilistic methods for the applicability to the current code and standard.
Takaya, Shigeru; Machida, Hideo*; Kamishima, Yoshio*
no journal, ,
This paper describes the outline of the guidelines on structural reliability evaluation for the passive components of the fast breeder reactor (FBR). The guidelines are now being prepared by the working group for the system based code in the Japan society of mechanical engineers. They consist of five chapters, which are "General rules", "Reliability evaluation", "Failure scenario setting", "Modeling", and "Failure probability calculation", respectively. In the chapter of "Reliability evaluation", the general procedures of reliability evaluation are explained. Detailed procedures at each step are explained in the following chapters in the guidelines. Evaluation procedures for accumulation damage and crack propagation due to creep-fatigue interaction are also provided in the appendix of the guidelines because creep-fatigue interaction is a typical degradation phenomenon in FBR.
Shimomura, Kenta; Garcia Rodriguez, D.*
no journal, ,
There is a longstanding interest to improve the in-service inspection (ISI) of nuclear power plants in general, and sodium fast reactors (SFR) in particular. With the goal of improving both safety and economic performance, a risk-informed approach to in-service inspection (RI-ISI) has therefore been earmarked for future commercial-scale SFR. A reliable RI-ISI, however, can only be established based on a rich operational experience record. For long, JAEA has compiled information about SFR troubles, but up to now such effort had been limited to specific types of troubles or components, while lacking the reliability and completeness required for the RI-ISI approach. In this paper we introduce recent work aiming to develop a comprehensive SFR trouble database based on original information sources, which both expands and improves any previously available documentation.
Mihalache, O.; Garcia Rodriguez, D.*; Yamamoto, Toshihiro*; Cheng, W.*
no journal, ,