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Journal Articles

Preliminary verification of water radiolysis and ECP calculation models by in-pile ECP measurements

Hanawa, Satoshi; Hata, Kuniki; Chimi, Yasuhiro; Kasahara, Shigeki

Proceedings of 21st International Conference on Water Chemistry in Nuclear Reactor Systems (Internet), 12 Pages, 2019/09

Journal Articles

Study of irradiation effect on ECP using in-pile loops in the JMTR

Hanawa, Satoshi; Uchida, Shunsuke; Hata, Kuniki; Chimi, Yasuhiro; Kasahara, Shigeki*; Nishiyama, Yutaka

Proceedings of 20th Nuclear Plant Chemistry International Conference (NPC 2016) (USB Flash Drive), 10 Pages, 2016/10

The authors proposed and ECP evaluation model introducing irradiation-induced diffusion in the oxide layer to simulate neutron irradiation effect, and predicted with this model that ECP is started to depress from the neutron flux of about ten to the fourteenth per square meter. As the JMTR has in-pile loops applicable to water chemistry experiments, degree of irradiation effect on ECP appears in the in-pile loop was estimated by the model. Under oxygen injected condition, ECP in a capsule becomes constant along the vertical direction due to the presence of high amount of oxygen and hydrogen peroxide in a capsule. However, if neutron irradiation depress ECP, ECP in a capsule along vertical direction wouldn't become constant, and the degree to the decrement is detectable by experiments.

Journal Articles

Present status of in-pile IASCC growth tests at JMTR

Kaji, Yoshiyuki; Ugachi, Hirokazu; Nakano, Junichi*; Matsui, Yoshinori; Kawamata, Kazuo; Tsukada, Takashi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*

HPR-364, Vol.1 (CD-ROM), 10 Pages, 2005/10

Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors (LWRs) for a long period. In the research field of IASCC, mainly PIEs for irradiated materials have been carried out, because there are many difficulties on SCC tests under neutron irradiation. Hence as a part of the key techniques for in-pile SCC tests, we have embarked on a development of the test technique to obtain information concerning effects of applied stress level, water chemistry, irradiation conditions, etc. In this paper, we describe the developed several techniques, especially control of loading on specimens, monitoring technique of crack propagation and so on, and the present status of in-pile IASCC growth tests using pre-irradiated materials at JMTR.

Journal Articles

Stress corrosion cracking growth behavior of in-core materials

Kaji, Yoshiyuki

Proceedings of KNS-AESJ Joint Summer School 2005 for Students and Young Researchers, 2, p.221 - 228, 2005/08

For core internals, the main research items are intergranular stress corrosion cracking (IGSCC) of low carbon stainless steel in core shrouds and primary loop recirculation pipes in boiling water reactor (BWR), and irradiation assisted stress corrosion cracking (IASCC) which is caused by the synergistic effects of neutron and gamma-ray radiation, corrosion by high temperature water, and the residual and/or applied stresses. This paper describes the current status and typical results of fundamental study for mechanistic understanding of IGSCC and IASCC, development of IASCC evaluation technology for BWR plants based on post-irradiation IASCC test data as a part of METI's national project, in-pile IASCC tests.

Journal Articles

Development of test techniques for in-pile SCC initiation and growth tests and the current status of in-pile testing at JMTR

Ugachi, Hirokazu; Kaji, Yoshiyuki; Nakano, Junichi*; Matsui, Yoshinori; Kawamata, Kazuo; Tsukada, Takashi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*

Proceedings of 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors (CD-ROM), p.319 - 325, 2005/00

Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors (LWRs) for a long period. In the research field of IASCC, mainly PIEs for irradiated materials have been carried out, because there are many difficulties on SCC tests under neutron irradiation. Hence as a part of the key techniques for in-pile SCC tests, we have embarked on a development of the test technique to obtain information concerning effects of applied stress level, water chemistry, irradiation conditions, etc. In this conference, we describe the developed several techniques, especially control of loading on specimens, monitoring technique of crack initiation, propagation and water chemistry, and the current status of in-pile SCC tests using thermally sensitized materials at JMTR.

Journal Articles

Assessment of irradiation temperature stability of the first irradiation testi rig in the HTTR

Shibata, Taiju; Kikuchi, Takayuki; Miyamoto, Satoshi*; Ogura, Kazutomo*

Nuclear Engineering and Design, 223(2), p.133 - 143, 2003/08

 Times Cited Count:1 Percentile:11.39(Nuclear Science & Technology)

The High Temperature Engineering Test Reactor (HTTR) can provide very large spaces at high temperatures for irradiation tests. The I-I type irradiation equipment was developed as the first irradiation rig. It will be served for an in-pile creep test on a stainless steel with large standard size specimens. It uses the ambient high temperature of the core for the irradiation temperature control. The target irradiation temperatures are 550 and 600$$^{circ}$$C with the target temperature deviation of $$pm$$3$$^{circ}$$C. In this study, the irradiation temperature changes at transient conditions were analyzed by an FEM code and the temperature controllability of the equipment was examined by a mockup test. The controllability was evaluated with the measured temperature transient data at the core graphite components in the Rise-to-Power tests of the HTTR. The result indicates that the temperature control method of the equipment is effective to keep the irradiation temperature stable in the irradiation test.

Journal Articles

Dosimetry plan at the first irradiation test in the HTTR

Shibata, Taiju; Kikuchi, Takayuki; Shimakawa, Satoshi

Reactor Dosimetry in the 21st Century, p.211 - 218, 2003/00

The High Temperature Engineering Test Reactor (HTTR) is the first HTGR in Japan with a maximum power of 30 MW. The construction of it was completed successfully in March 2002. The HTTR aims to perform irradiation studies at its very wide irradiation spaces at high temperatures. Although the creep behavior of materials is measured by the large standard size specimens at out-of-pile, small size ones are generally used for in-pile creep tests because of the irradiation capability of reactors. The I-I type irradiation equipment, the first rig for the HTTR, is to be used for the in-pile creep test on a stainless steel with the standard specimens. The rig can give big tensile loads of about 9.8 kN on them. The temperatures of 550 and 600$$^{circ}$$C and the fast neutron fluence of 1.2$$times$$10$$^{23}$$n/m$$^{2}$$ are the targets of the test. Prior to the in-pile creep test, the in-core irradiation properties at the irradiation region are to be obtained by the rig as the first irradiation test. This paper describes the dosimetry plan at the first irradiation test and the subsequent data assessment procedure.

JAEA Reports

Development of irradiation rig in HTTR and dosimetry method; I-I type irradiation equipment

Shibata, Taiju; Kikuchi, Takayuki; Miyamoto, Satoshi*; Ogura, Kazutomo*

JAERI-Tech 2002-097, 19 Pages, 2002/12

JAERI-Tech-2002-097.pdf:1.4MB

The HTTR aims to establish and upgrade the technological basis for the HTGRs and to perform the innovative basic research on high temperature engineering. The HTTR is planned to be used to perform various tests such as, the safety demonstration test, high temperature test operation and irradiation test with large irradiation fields at high temperatures. This paper describes the design of the I-I type irradiation equipment, developed as the first rig for the HTTR, and does the planned dosimetry method at the first irradiation test. It was developed to perform in-pile creep test on a stainless steel with large standard size specimens. It can give great loads on the specimens stably and can control the irradiation temperature precisely. The in-core data are measured by differential transformers, thermocouples, SPNDs and neutron fluence monitors. The obtained data at the first test can be contributed to upgrade the technological basis for the HTGRs, since it is the first direct measurement of the in-core irradiation environments.

Journal Articles

Evaluation of in-pile and out-of-pile stress relaxation in 316L stainless steel under uniaxial loading

Kaji, Yoshiyuki; Miwa, Yukio; Tsukada, Takashi; Kikuchi, Masahiko; Kita, Satoshi; Yonekawa, Minoru; Nakano, Junichi; Tsuji, Hirokazu; Nakajima, Hajime

Journal of Nuclear Materials, 307-311(Part1), p.331 - 334, 2002/12

 Times Cited Count:4 Percentile:30.46(Materials Science, Multidisciplinary)

Irradiation assisted stress corrosion cracking (IASCC) caused by simultaneous effects of neutron irradiation and high temperature water environments has been pointed out as one of the major concerns of in-core structural materials not only for the light water reactors (LWRs) but also for the water-cooled fusion reactor. It is necessary to evaluate precisely stress condition under irradiation environment, because stress is one of key factors on IASCC. Stress relaxation of tensile type specimens under fast neutron irradiation at 288$$^{circ}$$C has been studied for type 316L stainless steel in Japan Materials Testing Reactor (JMTR). This paper describes the in-pile and out-of-pile stress-relaxation test results of tensile type specimens for type 316L stainless steel as compared with the literature data by Foster, which were mainly obtained by bent beam specimens. Moreover these experimental results were compared with the analytical ones by using Nakagawa's model.

Journal Articles

New in-pile water loop facility for IASCC studies at JMTR

Tsukada, Takashi; Komori, Yoshihiro; Tsuji, Hirokazu; Nakajima, Hajime; Ito, Haruhiko

Proceedings of International Conference on Water Chemistry in Nuclear Reactor Systems 2002 (CD-ROM), 5 Pages, 2002/00

Irradiation assisted stress corrosion cracking (IASCC) is caused by the synergistic effects of neutron and gamma radiation, residual and applied stresses and high temperature water environment on the structural materials of vessel internals. IASCC has been studied since the beginning of the 1980s and the phenomenological knowledge on IASCC is accrued extensively. However, mainly due to the experimental difficulties, data for the mechanistic understanding and prediction of failures of the specific in-vessel components are still insufficient and further well-controlled experiments are needed [1]. In recent years, efforts to perform the in-pile materials test for IASCC study have been made at some research reactors [2-4]. At JAERI, a high temperature water loop facility was designed to install at the Japan Materials Testing Reactor (JMTR) to carry out the in-core IASCC testing. This report describes an overview of design and specification of the loop facility.

Journal Articles

Neutronic and thermal estimation of blanket in-pile mockup with Li$$_{2}$$TiO$$_{3}$$ pebbles

Nagao, Yoshiharu; Nakamichi, Masaru; Tsuchiya, Kunihiko; Kawamura, Hiroshi

Fusion Engineering and Design, 58-59, p.673 - 678, 2001/11

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

no abstracts in English

Journal Articles

In-pile and post-irradiation creep of type 304 stainless steel under different neutron spectra

Kurata, Yuji; Itabashi, Yukio; Mimura, Hideaki*; Kikuchi, Taiji; Amezawa, Hiroo; Shimakawa, Satoshi; Tsuji, Hirokazu; Shindo, Masami

Journal of Nuclear Materials, 283-287(Part.1), p.386 - 390, 2000/12

 Times Cited Count:5 Percentile:38.85(Materials Science, Multidisciplinary)

no abstracts in English

JAEA Reports

In-pile test of the crud separator system in the HBWR; Development of the crud separator system, II

; Iimura, Katsumichi; Yokouchi, Iichiro; Nakagawa, Tetsuya;

JAERI-M 90-231, 17 Pages, 1991/01

JAERI-M-90-231.pdf:0.8MB

no abstracts in English

JAEA Reports

In-pile loop OWL-2 and irradiation tests done with it

; ; ; Watanabe, Hiroyuki; ;

JAERI-M 90-196, 45 Pages, 1990/11

JAERI-M-90-196.pdf:1.67MB

no abstracts in English

Journal Articles

Safety aspects required to the in-pile irradiation facilities attached to the JMTR

;

The Safety,Status and Future of Non-Commercial Reactors and Irradiation Facilities,Vol. 1, p.509 - 515, 1990/09

no abstracts in English

Journal Articles

Diffusion coefficients of cesium in un-irradiated graphite and comparison with those obtained from in-pile experiments

Hayashi, Kimio; Fukuda, Kosaku

Journal of Nuclear Materials, 168, p.328 - 336, 1989/00

 Times Cited Count:9 Percentile:70.78(Materials Science, Multidisciplinary)

no abstracts in English

JAEA Reports

VHTR Fuel Irradiation Tests by the In-pile Gas Loop,OGL-1 at JMTR

; ; ; ;

JAERI-M 86-068, 17 Pages, 1986/04

JAERI-M-86-068.pdf:0.8MB

no abstracts in English

Journal Articles

In-pile eddy current test on PWR fuel rod failed by pellet-cladding interaction

; T.Johnsen*

Journal of Nuclear Science and Technology, 23(8), p.752 - 755, 1986/00

 Times Cited Count:1 Percentile:28.52(Nuclear Science & Technology)

no abstracts in English

52 (Records 1-20 displayed on this page)