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Shibata, Akira; Kitagishi, Shigeru; Watashi, Katsumi; Matsui, Yoshinori; Omi, Masao; Sozawa, Shizuo; Naka, Michihiro
Nihon Hozen Gakkai Dai-13-Kai Gakujutsu Koenkai Yoshishu, p.290 - 297, 2016/07
The exhaust stack of Japan Materials Testing Reactor Hot laboratory is a part of gaseous waste treatment system. It was built in 1970 and is 40 m in height. In 2015, thinning was found at some anchor bolts on base of the stack. When thinning of anchor bolts were investigated, gaps between anchor bolt nuts and flange plate was found. JAEA removed steel cylinder of stack which is 33 m in height for safety. In the end of investigation, thinning was found in all anchor bolts of the stack. Cause investigation for the thinning and the gaps were performed. It is concluded that the thinning was caused by water infiltration over a long period of time and the gaps were caused by elongation of thinning part of anchor bolts by the 2011 earthquake off the Pacific coast of Tohoku.
Seguchi, Tadao*; Tamura, Kiyotoshi*; Watashi, Katsumi; Suzuki, Masahide; Shimada, Akihiko; Sugimoto, Masaki; Idesaki, Akira; Yoshikawa, Masahito; Oshima, Takeshi; Kudo, Hisaaki*
JAEA-Research 2012-029, 158 Pages, 2012/12
The degradation mechanisms of ethylene-propylene rubber (EPR), crosslinked polyethylene (XLPE), polyvinylchloride (PVC), and silicone rubber (SiR) as the cable insulation materials were investigated for the cable ageing research of the nuclear power plant. The materials as same insulations for the practical cable (practical formulation) and as the model formulation containing specific additive were selected. They were exposed to the accelerated radiation and thermal environments. The mechanical properties, the crosslinking and chain scission, and the distribution of antioxidant and of oxidative products were measured and analyzed.
Watashi, Katsumi
Nihon Hozen Gakkai Dai-6-Kai Gakujutsu Koenkai Yoshishu, p.337 - 341, 2009/08
Furnace tubes are widely used in heat transfer device in petroleum or chemical industries. This study proposes an alternative calculation method for the evaluation of creep rupture damage factor of furnace tubes. The method is consistency with an evaluation procedure described in American Petroleum Institute document API 581. Three types of creep parameters including Larson-Miller parameter, namely NRIM creep rupture data of furnace tube material, can be applied for evaluation. The method is rather simple compared with original method described in API 581 first edition.
Hattori, Shuji*; Inoue, Fumitaka*; Kurachi, Hiroaki*; Watashi, Katsumi; Tsukimori, Kazuyuki; Hashimoto, Takashi; Yada, Hiroki
JAEA-Research 2008-080, 45 Pages, 2009/02
Research on cavitation erosion in liquid metal is very important to confirm the safety of fast breeder reactor using sodium coolant. In this study, a cavitation erosion test apparatus was developed to carry out the erosion tests in low-temperature liquid metals. Cavitation erosion tests were carried out in liquid lead-bismuth alloy and in deionized water. We discuss the effect of liquid parameters and temperature effects on the erosion rate. We reach to the following conclusions. The erosion rate was evaluated in terms of a relative temperature which was defined as the percentage between freezing and boiling points. At 14 C relative temperature, the erosion rate is 10 times in lead-bismuth alloy, and 2 to 5 times in sodium, compared with that in deionized water. The erosion rate can be evaluated as a function of material density and sound velocity. Finally, the temperature dependence was discussed in term of liquid vapor pressure.
Sakakibara, Yasuhide; Isomura, Kazutoshi; Yamashita, Takuya; Watashi, Katsumi; Doi, Motoo; Okusa, Kyoichi; Tagawa, Akihiro; Hirahara, Kenji
Nihon Hozen Gakkai Dai-3-Kai Gakujutsu Koenkai Yoshishu, p.283 - 286, 2006/06
This study was performed to enrich the contents of measures for an ageing of the nuclear power plants at Fukui area, where infrastructures of research works, for example, institutes, universities etc. are intensively existed., according to the Road-Map established by the Atomic Energy Society of Japan at 2005.
Hattori, Shuji*; Ito, Takamoto*; Watashi, Katsumi; Hashimoto, Takashi
JNC TY4400 2005-003, 82 Pages, 2005/09
no abstracts in English
Meshii, Toshiyuki*; Watashi, Katsumi; Doi, Motoo; Hashimoto, Takashi
JNC TY4400 2004-001, 90 Pages, 2004/03
Based on the results of the Phase I of the titled research, we started Phase II scheduled to reach conclusion (develop basic methods necessary for FBR post-construction code) in three years. The following results were obtained. (1) Development of assessment procedure for multiple defects due to creep damage: We first confirmed that multiple surface cracks in axial and circumferential direction inside a cylinder can be evaluated independently by investigating multiple flaws found in a cylinder due to cyclic thermal shock. Then we developed a method to evaluate the stress intensity factors (SIFs) of the multiple circumferential cracks in a finite legth cylinder under axisymmetric loads. (2) Improving creep-fatigue crack grown (C-FCG) assessment pfodedure: We reviewed the French code A16 for flaw assessment and compared it with the JNC proposal. Since the fundamenfal philosophy for both was to evaluate the C-FCG by estimating elastic-plastic fracture mechanics parameter from the SIF, we developed a "3D crack finite element analysis system that can specify the target error in SIF." (3) Crack propagation assessment under thermal stresses (fatigue crack grown (FCG) resistace for small load): To improve accuracy of FCG assesment for components in FBR power plants (designed to minimize thermal stresses) under thermal cycles, we obtained near threshold FCG data for S55C, SUS304, HT60, SS400, 2.25Cr-1Mo, SUS316, SUS321, T91, Inconel718 by Kmax = constant test method. The results showed that FCG curves in the JSME post construction code (which is an extrapolation of the curves in ASME PVP code sec. XI) are valid in general. However, precise review of S55C, HT60's data suggested that the JSME FCG, evaluation curve may not have enough safety margin. In addition, we proposed a method to predict the decrease in the threshold SIF range DeltaKth due to high Kmax and showed its validity.
Meshii, Toshiyuki*; Watashi, Katsumi; Doi, Motoo; Hashimoto, Takashi
JNC TY4400 2003-002, 63 Pages, 2003/03
no abstracts in English
Watashi, Katsumi
JNC TN4410 2003-008, 62 Pages, 2001/04
Since May 2001, FBR basic lectures -fundamental and general contents on FBR- have been given to the personnel whose career is a few years after joining with JNC and concerned companies personnel at International Cooperation and Technology Development Center building up a close connection with Human Resources Development Section Personnel Division Head Quarter. The trend toward opening these lectures to the public and making use of them for other organizations has come to take place. This time one of these lectures 'Structural integrity of FBR' will be opened to the Fire and Disaster Management Agency Ministry of Public Management.
; ; Yoshida, Eiichi; ; ; Yaguchi, Katsumi*; Watashi, Katsumi
PNC TN9450 95-006, 175 Pages, 1995/04
In order to advancement in materials strength standard on elevated temperature design guide of the FBRs and evaluation method of materials strength behavior, this report are presented about the creep properties of FBR grade 316 (Abbreviation 316FR), based on the R&D results obtained through the activities of material tests. Contents of the data sheet are as follows; (1)Material : 316FR (Base Metal) Plate 8 heats (B7, B8, JA, MC,MD, ,ME, MG, MI heat) Tube 2 heats (S6F, B10 heat) (2)Test environment : In air , In sodium (3)Test temperature : 500C
800
C (4)Test method : According to JIS and FBR Metallic Materials Test Method (5)Number of deta : 211 points
; ; Yoshida, Eiichi; ; Yaguchi, Katsumi*; Watashi, Katsumi
PNC TN9450 95-003, 98 Pages, 1995/02
In order to advancement in materials strength standard on elevated temperature design guide of the FBRs and evaluation method of materials strength behavior, this report are presented about the tensile properties of FBR grade 316FR on air and sodium environment conditions. Contents of the data sheet are as follows; (1)Material : FBR grade 316FR (Base Metal) (a)B6 heat 25mm1,000㎜
1,000mm (Plate) (b)B7 heat 50mm
1,000mm
1,000mm (Plate) (c)B8 heat 40mm
1,000mm
1,000mm (Plate) (d)B9 heat 25mm
1,000mm
1,000mm (Plate) (d)B11 heat 50mm
1,000mm
1,000mm (Plate) (2)Pre-test treatment: (a)Argon aged for 5000hr at 500, 550, 600
C (B6, B7, B8 Heats) (b)Argon aged for 20000hr at 500, 550, 600
C (B7, B8 Heats) (c)Sodium exposed for 5000hr at 500, 550, 600
C (B6, B7, B8 Heats) (d)Sodium exposed for 20000hr at 500, 550, 600
C (B6, B7, B8 Heats) (e)As-recieved (B6, B7, B8, B9, B11 Heats) (3)Test temperature : R.T.
750
C (4)Test method : Accoding to JIS and FBR Metallic Materials Test Methods (5)Number of data : 153 points (Including 64 points on the last report)
Watashi, Katsumi
Nuclear Engineering and Design, 153(2-3), p.275 - 285, 1995/00
None
; Wakai, Takashi; ; ; Watashi, Katsumi;
PNC TN9410 93-220, 112 Pages, 1993/09
This report describes elastic thermal stress analysis and creep-fatigue damage estimation results of Mod.9Cr-1Mo piping specimen which consists of three different thickness portions (20, 15, 10mm), and contains six circumferential weldments. Thermal fatigue failure test on the specimen has been conducted in the sodium test loop named SPTT to clarify fatigue crack initiation behavior under thermal bending stress conditions, which is caused by temperature gradient arisen in the wall thickness of the specimen. The thermal transient test is now underway, with the condition where sodium of 550C and 300
C sodium alternately flow into the specimen for 5 minutes in one cycle. The test is scheduled to complete after loading 9,000 cycles of thermal transients. Crack inspection by using penetration test tequnique is planned after the test. For an analytical study, heat transfer analysis using measured temperature data and elastic thermal stress analysis were carried out with the finite element method, and the analysis results were utilized to develop a candidate creep-fatigue damage evaluation method for Mod.9Cr-1Mo steels based on elastic analysis. Some analysis and damage estimation results of the specimen are presented here, and crack initiation after 9,000 cycles of thermal transients are predicted based on the calculated creep-fatigue damage.
; Wakai, Takashi; ; ; Watashi, Katsumi;
PNC TN9410 93-209, 115 Pages, 1993/09
This report describes elastic thermal stress analysis and creep-fatiguedamage evaluation results of an SUS316FR piping specimen which contains circumferential weldment in the middle portion of it. Thermal creep-fatigue failure test on the specimen was conducted in a sodiumtest loop named STST to clarify creep-fatigue crack initiation behavior of SUS316FR base metal and weldment under thermal bending stress conditions, which is caused by temperature gradient arisen in the wall thickness or the specimen. Thermal transient test was conducted under the condition that 550C and 300
C sodium alternately flow into the specimen for 5 hours and 1 hour, respectively in one cycle. The test had been completed after loading 1,600 cycles or thermal transients. After the test, crack inspection by PT was performed, and cracks were observed successfully in both the base metal and weldment. For an analytical study, heat transfer analysis using measured temperature data and elastic thermal stress analysis were carried out with the finite element method, and the analysis results were utilized to develop a candidate creep-fatigue damage evaluation method for SUS316FR steel based on elasticanalysis. Some analysis results and damage evaluation results of the specimen arepresented here, and crack initiation after 1,600 cycles of thermal transients is predicted based on the calculated creep-fatigue damage, which demonstrates a good agreements with the crack distribution on the specimen inner surface.
Watashi, Katsumi; Aoto, Kazumi; Aoki, M; Komine, Ryuji; Ito, Takushi; Hasebe, Shinichi; Kato, Shoichi; Koi, Mamoru; Wada, Yusaku
PNC TN9410 93-142, 120 Pages, 1993/06
Much progress has been made in improving established creep properties of Type 316 stainless steel and to develop a new structural material named "FBR Grade Type 316 Stainless Steel", 316FR, with superior creep properties. This report includes a draft of Material Strength Standard of 316FR and its interpretation on the basis of the major result of research and development conducted so far. The draft includes identical items described in the "Standards for the Strength of Materials" for Monju, and was carefully prepared to have an identical style for convenience in design evaluation. Creep damage evaluation diagrams, which are depicted in the "Structural Design Guide for Class 1 Components of Prototype Fast Breeder Reactor for Elevated Temperature Service" (ETSDG) for individual materials, are also included in this report.
Umeda, Hisao; Tanaka, Nobuyuki; Watashi, Katsumi; Kikuchi, Masayuki; Iwata, Koji
Nuclear Engineering and Design, 140, p.349 - 372, 1993/06
None
Watashi, Katsumi; ; ; ; ; Asada, Yasuhide*
Nuclear Engineering and Design, 139(3), p.293 - 298, 1993/03
Times Cited Count:2 Percentile:29.55(Nuclear Science & Technology)None
Watashi, Katsumi; ; ; Asada, Yasuhide*
Nuclear Engineering and Design, 139, 283 Pages, 1993/00
Times Cited Count:10 Percentile:69.39(Nuclear Science & Technology)None