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Imagawa, Yuya; Toyota, Kodai; Onizawa, Takashi; Kato, Shoichi
JAEA-Testing 2023-004, 76 Pages, 2024/03
JAEA-Testing-2023-004.pdf:2.08MB
This manual describes the methods for conducting material tests in air, argon gas, and sodium, and for organizing the data obtained, as a part of the development of high-temperature structural design technology for fast reactors. This manual reflects the revision of test methods in Japanese Industrial Standards (JIS) to the "FBR Metallic Materials Test Manual, PNC TN241 77-03" published in 1977 and the "FBR Metallic Materials Test Manual (Revised Edition), JNC TN9520 2001-001" published in 2001. Also, it was written with reference to the recommended room temperature / elevated temperature tensile test method by the Japan Society of Mechanical Engineers (JSME) and the test standard for the elevated-temperature low-cycle fatigue test method by the Society of Materials Science, Japan (JSMS), which are the standard for material test methods in the domestic academic society.
-ray irradiationAmbai, Hiromu; Nishizuka, Yusuke*; Sano, Yuichi; Uchida, Naoki; Iijima, Shizuka
QST-M-2; QST Takasaki Annual Report 2015, P. 90, 2017/03
The spent fuel stored in the storage pools at the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company Holdings, Inc. is exposed with the environment containing seawater components, owing to the injection of seawater into the storage pools. Therefore, during reprocessing, it is expected that the spent fuel will be contaminated with seawater components, and the influence of seawater on reprocessing needs to be investigated. We conducted the corrosion tests of the HAW storage tanks under -ray irradiation, and revealed that no significant effect of seawater components was emerged.
Kizaki, Minoru; Honda, Junichi; Usami, Koji; Ouchi, Asao*; Oeda, Etsuro; Matsumoto, Seiichiro
JAERI-Tech 2000-087, 50 Pages, 2001/02
JAERI-Tech-2000-087.pdf:2.78MB
no abstracts in English
Kitajima, Toshio; Abe, Shinichi; Takahashi, Kiyoshi; Onuma, Yuichi; Watanabe, Hiroyuki; Okada, Yuji; Oyake, Noriyuki*
UTNL-R-0404, p.6_1 - 6_9, 2001/00
no abstracts in English
Harada, Yuhei*; Maruyama, Yu; Maeda, Akio*; Chino, Eiichi; Shibazaki, Hiroaki*; Kudo, Tamotsu; Hidaka, Akihide; Hashimoto, Kazuichiro; Sugimoto, Jun
JAERI-Conf 2000-015, p.309 - 314, 2000/11
no abstracts in English
; ; Sakamoto, Naoki; *; Akasaka, Naoaki;
JNC TN9400 2000-095, 110 Pages, 2000/07
JNC-TN9400-2000-095.pdf:13.57MB
The effects of high fluence irradiation and swelling on the transient burst properties of austenitic steel fuel claddings; PNC316 and 15Cr-20Ni stcel, which were irradiated as the MONJU type fuel assemblies (MFA-1&MFA-2) in the FFTF reactor, were investigated. The temperature-transient-to-burst tests were conducted on a total of eight irradiation conditions. Fractographic examination and TEM observation were performed in order to evaluate the effect of high dose irradiation on the transient burst property and the relation between failure mechanism and microstructural change during rapid (ramp) heating. The results of the PIE showed that there was no significant effect of irradiation on the transient burst properties of these fuel claddings under the irradiation conditions examined. the results obtained in this study are as follows; (1)The rupture temperature of the irradiated PNC316 fuel cladding of MFA-1 was as same as that of our previous works for the fluence range up to 2.13
10
n/m
. There was no noticeable decrease in rupture temperature with increasing fluence in lower hoop stress region(
100MPa). (2)The rupture temperature of the irradiated 15Cr-20Ni fuel cladding of MFA-2 was almost as same as that of as-received cladding for the hoop stress range up to about 200MPa. The rupture temperature did not decrease significantly with fluence. (3)The rupture temperature of the irradiated PNC316 cladding tested at hoop stress 69MPa, which was the design hoop stress for MONJU fuel, was 1055.6
C. This suggested that the design cladding maximum temperature limit for MONJU (830C) was conservative. (4)There was no obvious relation between rupture temperature, swelling and microstructural change during transient heating under the irradiation conditions examined.
; *
JNC TN9450 2000-002, 335 Pages, 1999/10
JNC-TN9450-2000-002.pdf:21.65MB
This report summarizes the material test dala of SUS304 welded joints. Numbers of the data are as follows: [Tensile tests 71 (Post-irradiation: 39, others: 32) [Creep tests 77 (Post-irradiation: 20, others: 57) [Fatigue tests 50 (Post-irradiation: 0) [Creep-fatigue tests 14 (Post-irradiation: 0) This report consists of the printouts from "the structural material data processing system".
Tabata, Toshio; Komukai, Bunsaku; Nagao, Yoshiharu; Shimakawa, Satoshi; Koike, Sumio; Takeda, Takashi; Fujiki, Kazuo
JAERI-Tech 99-021, 68 Pages, 1999/03
no abstracts in English
; ; Fujisaku, Kazuhiko*; Ishibashi, Yuzo; Takeda, Seiichiro
PNC TN8410 98-060, 74 Pages, 1998/03
None
; Department of JMTR
JAERI-Conf 98-007, 197 Pages, 1998/03
no abstracts in English
Kurihara, Ryoichi
Nucl. Eng. Des., 172(3), p.317 - 325, 1997/00
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
Aoto, Kazumi; ;
PNC TN9410 97-037, 51 Pages, 1996/11
The basic material properties of a rolled steel for welded structure (present standard name is SM400B, old standard name SM41B) which is used as the liner plate in SHTS cells of "Monju plant". Based on the material testing data for evaluation of structural integity of the liner during sodium leakage are tentatively proposed. Main basic material properties are shown as follows. (1)The 0.2% offset yield stress (lower yield point). (2)The ultimate tensile strength. (3)The modulus of the longitudinal elasticity. (4)Static stress-strain relation. (Physical property in Ludwik equation). (5)The creep strain. (6)The linear thermal expansion coefficient. (7)The density. (8)A specific heat. (9)The thermal conductivity.
; ; ; ; ; ;
PNC TN9410 96-235, 258 Pages, 1996/03
The chemical decontamination technique has been developed in order to remove the crud adhering to the surface of the components constructing the primary coolant system, as a part of the measure to decrease the exposure in the annual inspection. The technique has been already applied to the prototype reactor "Fugen", in the core of which the fuel assemblies were not loaded. The chemical decontamination, for the core in which the fuel assemblies are loaded, has been planned for the purpose of improving the utilization factor. It is necessary to confirm, through the test before putting the plan into practice, that the decontamination reagent does not exert a bad influence upon the components constructing the fuel assembly. This report describes the test results which have been carried out so as to investigate the influence of the reagent on the components constructing the fuel assembly. The outline of the results is as follows: (1)The susceptibility to stress corrosion cracking of the chemical decontamination treatment and the residual decontamination reagent on the components constructing the fuel assembly is low enough. (2)The chemical decontamination treatment and the residual decontamination reagent do not exert a bad influence upon the integrity of the fuel assembly concerning the fuel rod holding function of the spacer and the characteristics of the fretting wear caused on the fuel claddings.
Tsuchida, Noboru; Ooka, Norikazu; ;
ASRR-V: Proc., 5th Asian Symp. on Research Reactors, 1, p.123 - 130, 1996/00
no abstracts in English
; Oku, Tatsuo*; Ishiyama, Shintaro; Eto, Motokuni
Nihon Kikai Gakkai Rombunshu, A, 62(593), p.10 - 17, 1996/00
no abstracts in English
JAERI-Research 95-026, 54 Pages, 1995/03
JAERI-Research-95-026.pdf:2.47MB
no abstracts in English
Iriya, Keishiro*; Fujiwara, Yasushi*; Motohashi, Kenichi*; Nakanishi, Masatoshi*
PNC TJ1449 93-001, 341 Pages, 1993/03
None
; ; ; ; ; Shibahara, Itaru
PNC TN9410 92-321, 30 Pages, 1992/10
Integrity evaluations have been performed for the 2nd Fugen pressure tube test (8 years irradiation, 5.6 10
n/cm (E
1Mev)). Test items mainly consist of tensile test, bending test, corrosion test and hydrogen analysis. It has become clear using these data that the pressure tube material has maintained its integrity during the irradiation by the integrity assessment on both tensile and fracture toughness properties. Besides, both thickness loss by corrosion and absorbed hydrogen content were lower than those of design values.
; ; *; *; *; *; Yoshida, Eiichi
PNC TN9450 91-008, 38 Pages, 1991/09
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 SUS316, based on the R&D results obtained through the activities of material tests. Contents of the data sheet are as follows; (1)Material : FBR grade SUS316 (Base Metal) B7 Heat 1,000mm1,000㎜50mm
(Plate) B8 Heat 1,000㎜1,000mm40mm(Plate) B9 Heat 1,000mm1,000㎜25㎜(Plate) (2)Test temperature : RT750C (3)Test method : According to JIS and FBR Metallic Materials Test Methods (4)Number of deta : 64 points
