Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
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
JAEA-Research-2012-029.pdf:9.4MB
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
JAEA-Research-2008-080.pdf:2.86MB
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
JNC-TY4400-2004-001.pdf:10.67MB
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
JNC-TY4400-2003-002.pdf:3.95MB
no abstracts in English
Watashi, Katsumi
JNC TN4410 2003-008, 62 Pages, 2001/04
JNC-TN4410-2003-008.pdf:2.65MB
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.
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：29.79(Nuclear Science & Technology)None
Watashi, Katsumi; *; *; Asada, Yasuhide*
Nuclear Engineering and Design, 139, 283 Pages, 1993/00
Times Cited Count：9 Percentile：66.93(Nuclear Science & Technology)None
Kasahara, Naoto; Watashi, Katsumi; Iwata, Koji
PNC TN9410 92-150, 275 Pages, 1992/06
In order to clayify problems in elastic analyses of structures and to get fundamental data to develop evaluation methods of inelastic behaviors, we have carried out transient heat transfer analysis and thermo-elastic analysis on the Stress Mitigation Structure Model (2)tested by Thermal Transient Test Facility for Structures(TTS). Adopted Code is a universal non-linear structural analysis code FINAS. In the case of thermal transient structural analysis, points to which special attention should be paid are modeling of structures and thermal boundaries. Considering the former point, calculated results by part models were examined though comparing with a entire model. For the later one, influences of thermal boundaries on thermal stress were investigated. These analysis results were input to structural strength data base with the destructive inspection results to be compared with ones of other test models. Following results were obtained through this analytical study. (1)It was clarified that influences of boundary conditions were different between a entire analysis model and part models. (2)Heat convective conditions on the surfaces in open air spaces can not be ignored and should be considered. (3)Failure mode of an upper skirt was quite different from one of a plate to cylinder junction in spite of a small difference of elastical calculated damage. (4)It was found that there is such unexpected failure mode of a perforated plate as penetration clacks between holes. (5)Blastic analysis data of the Stress Mitigation Structure Model(2) was obtained for a structural strength database.
Furuhashi, Ichiro*; Watashi, Katsumi
PNC TN9410 91-035, 110 Pages, 1991/01
A probabilistic fracture mechanics analysis code CANIS-P was developped for reliability evaluation of FBR structural components. The CANIS-P code has following aspects. (1)Compute failure probability of a structure with a semi-elliptical surface crack. (2)Can treat plate, cylinder with circumferential or axial crack. (3)Statistical distributions of initial crack shape and material properties. (4)Many failure criterions such as geometrical codition, net-section collapse, brittle fracture and tearing instability. (5)Fatigue crack growth by 
K or 
J. (6)Many reference K-solutions for computation of K and elastic J. (7)Simplified reference stress method for computation of elasto-plastic J. (8)Creep crack growth based on creep J-integral range J
. (9)Simplified reference stress method for computation or J. (10)Can treat hydrostatic proof test, PSI, ISI, leak detection and earthquake. (11)Leak rate computation by simple formula at PWR condition, or by elasto-plastic crack opening area and Bernoulli's formula. We can use CANIS-P aiming following works. (1)Reliability evaluation or failure probability evaluation of FBR structural components. (2)Get key parameters that enough sensitive on failure probabilities. (3)Research of best structures and best operating schedules considering economical costs and reliabilities. This report describes physical/mathematical models, probabilistic/statistic models, a user's manual and analysis examples of CANIS-P.
Watashi, Katsumi; Furuhashi, Ichiro*; Sasaki, Toshihiko*
PNC TN9410 91-034, 125 Pages, 1991/01
This paper describes an experimental and analytical result of crack growth behavior from circumferential slitted-cylinders under cyclic cold transients. PNC is promoting R&D program aimed at assessment method of crack and/or defect at creep temperature for FBR application, the experiment and analysis is one item of them. The purpose of this study is to verify the applicability of the method developed recently in PNC. Test models made of 304 austenitic stainless steel are 1.5 m in hight, 70 mm in inner diameter and 30 mm in thickness, and have axisymmetrical circumferential initial machined-notches on inner surface. As a first step, five machined-notches with different depth and width were tested in Air Cooled Themal Transient Test Rig. One cycle of thermal loading is such that the model heated up tp 650 C by furnace, then air blow into the model for 5 min. This sequence caused cyclic temperature gradient in the wall of the model. The tests were continued till crack depth exceeded 20. DC potential method and precise ultrasonic examination were applied to measure the crack growth. After the test, the model were dismantled and laboratory fractured, then striation spacing was measured continuouasly in the direction of crack growth. The experimental result was summarized as crack growth rate and reliability of online monitoring measures. Thermal-inelastic finite element analysis facilitated evaluation of fracture mechanics parameters, J, for thermal fatigue. The crack growth behavior was well predicted by the analysis considering a scatter band in material crack growth character. A simple method for inelastic crack evaluation was developed in PNC. The method is based on a database of linear fracture mechanics, and includes plastic and creep/relaxation modification. Firstly the applicability of the database to present problem was damonstrated comparing with thermal-elastic finite element analysis. Then J, and the crack growth behavior ...
Kasahara, Naoto; Watashi, Katsumi; Iwata, Koji
Int Symp on Structral Mechanics in Reactor Technology, 0 Pages, 1991/00
None
Tanaka, Nobuyuki; Watashi, Katsumi; Umeda, Hisao; Kikuchi, Masayuki; Iwata, Koji
Int Symp on Structral Mechanics in Reactor Technology, ,
None
