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Journal Articles

Effect of dissolved oxygen concentration on dynamic strain aging and stress corrosion cracking of SUS304 stainless steel under high temperature pressurized water

Hirota, Noriaki; Nakano, Hiroko; Fujita, Yoshitaka; Takeuchi, Tomoaki; Tsuchiya, Kunihiko; Demura, Masahiko*; Kobayashi, Yoshinao*

The IV International Scientific Forum "Nuclear Science and Technologies"; AIP Conference Proceedings 3020, p.030007_1 - 030007_6, 2024/01

Dynamic strain aging (DSA) and intergranular stress corrosion cracking (intragranular SCC) occur in high temperature pressurized water simulating a boiling water reactor environment due to changes in dissolved oxygen (DO) content, respectively. In order to clearly understand the difference between these phenomena, the mechanism of their occurrence was summarized. As a result, it was found that DSA due to intragranular cracking occurred in SUS304 stainless steel at low DO $$<$$ 1 ppb, while DSA was suppressed at DO 100 to 8500 ppb due to the formation of oxide films on the surface. On the other hand, when DO was increased to 20000 ppb, the film was peeled from the matrix, O element diffused to the grain boundary of the matrix, resulting in intergranular SCC. These results are indicated that the optimum DO concentration must be adjusted to suppress crack initiation due to DSA and intergranular SCC.

Journal Articles

Modelling of intergranular corrosion using cellular automata, 1; Characteristics and corrosion rates of stainless steels in modified nuclear reprocessing solution

Yamamoto, Masahiro; Irisawa, Eriko; Igarashi, Takahiro; Komatsu, Atsushi; Kato, Chiaki; Ueno, Fumiyoshi

Proceedings of Annual Congress of the European Federation of Corrosion (EUROCORR 2019) (Internet), 5 Pages, 2019/09

Intergranular corrosion phenomena were analysed using modified reprocessing solution. The data indicated that corrosion rates increased with time at the initial stage, and these stayed at constant value. Intergranular corrosion propagated at grain boundary in the initial stage and then attacked whole grain boundary causing drop out of grains. Corrosion rates of steady state were sum of intergranular corrosion amounts and weight losses of dropped grains. Surface appearances and cross sections of corroded samples were analyzed. The results indicated that the initial stage of intergranular corrosion was characterized by the ratio of corrosion rates between grain boundary and matrix. These ratios differed from individual grain boundaries. Total corrosion rates were affected by the distribution of these ratios. These data were based on the numerical modelling of intergranular corrosion using cellular automata. And also, calculated results were compared with these analytical data.

Journal Articles

Intergranular strains of plastically deformed austenitic stainless steel

Suzuki, Kenji*; Shobu, Takahisa

E-Journal of Advanced Maintenance (Internet), 10(4), p.9 - 17, 2019/02

In materials with an elastic anisotropy, a stress difference is generated between crystals when plastic deformation occurs, and it is known that this is deeply involved in material fracture. In this study, the residual stress for load direction in the plastically deformed material was investigated for each crystal orientation using the high-energy synchrotron radiation diffraction method. As a result, it was found that the residual stress is a tensile residual stress at an index with a high X-ray elastic constant (Young's modulus obtained for each diffraction surface) and a compressive residual stress at an index with a low X-ray elastic constant. We believe that this result will be useful for the technique of controlling the crystal orientation like the texture as improving the material strength.

JAEA Reports

Report of Examination of the Samples from Core Shroud (2F3-H6a) at Fukushima Dai-ni Nuclear Power Station Unit-3 (Contract Research)

The Working Team for Examination Operation of Samples From Core Shroud at Fukushima Dai-ni Unit-3

JAERI-Tech 2004-044, 92 Pages, 2004/05

JAERI-Tech-2004-044.pdf:15.18MB

The present examination has been performed with the objective to ensure the transparency of the examination as the third-party organization by providing technical basis for identifying the causes of cracking through examination of the sample taken from the cracked region of outer H6a welding portion of the core shroud at Fukushima Dai-ni Nuclear Power Station Unit-3, which was a part of sample stored in the Nippon Nuclear Fuel Development Co., Ltd. in the examination of Tokyo Electric Power Company in 2001. The present examination of the sample was conducted at the post irradiation examination facilities of JAERI. The following findings were obtained from the result of the present examination. (1)Three cracks were observed at the portion 3 to 9mm apart from the weld metal and the maximum depth was about 8mm. (2)Intergranular cracking was observed in almost whole fracture surface. The transgranular cracking was partially observed within the depth of about 300$$mu$$m from the surface. (3)Hardening layer over Hv400 at its maximum was found from the surface to the depth of about 500$$mu$$m. Based on the examination results concerning presence of tensile residual stress by welding and relatively high dissolved oxygen contents in core coolant, it is concluded that the cracks were mainly initiated in the hardening layer by transgranular stress corrosion cracking and propagated along the grain boundaries.

JAEA Reports

Inspection of heat transfer tubes after mock-up tests of minituarized apparatus for the acid recovery evaporator (Contract research)

Hamada, Shozo; Fukaya, Kiyoshi*; Kato, Chiaki; Yanagihara, Takao; Doi, Masamitsu*; Kiuchi, Kiyoshi

JAERI-Tech 2001-063, 49 Pages, 2001/10

JAERI-Tech-2001-063.pdf:13.39MB

The demonstration test for the acid recovery evaporator and the dissolver used in the major equipment of Rokkasho Reprocessing Plant (RRP), has been carried out. The mock-up miniature equipment has been employed to it. This test had been performed from April in 1998. The total time of demonstration test using the mock-up equipment is about two and half years, which corresponds to about 20,000 hours. After that, four of the seven heat transfer tubes used in the evaporator were drawn out and the corrosion level and the mechanical properties were evaluated for one of them. As a result, intergranular corrosion was recognized in the inner surface of the heat transfer tube and the corrosion depth at the grain boundary was statistically shown to be about one grain from the inner surface. Further, no change in mechanical properties was observed and growth of intergranular cracks in the inner surface of the specimen was found after flattering test.

Journal Articles

Non-equilibrium intergranular segregation and embrittlement in neutron-irradiated ferritic alloys

Kameda, Jun*; Nishiyama, Yutaka; Bloomer, T. E.*

Surface and Interface Analysis, 31(7), p.522 - 531, 2001/07

 Times Cited Count:10 Percentile:28.89(Chemistry, Physical)

This study describes intergranular segregation and embrittlement in several model ferritic alloys doped with Mn, P, S and/or Cu subjected to neutron irradiation, irradiation-equivalent thermal ageing (ETA) and post-irradiation annealing (PIA). Neutron irradiation produced a larger amount of intergranular P segregation than S segregation. Intergranular C segregation remained small in all the as-irradiated alloys. A PIA study has shown that the P segregation in P-doped alloys subjected to lower temperature PIA proceeds via mobile P-interstitial complexes while the S segregation is controlled by vacancy-enhanced diffusion. The mechanisms of non-equilibrium intergranular segregation induced by neutron irradiation are discussed in light of coupled fluxes of point defects and impurities, and changes in the segregation capacity of grain boundaries. Small punch tests demonstrated how the impurity segregation or desegregation and hardening or softening induced by the irradiation, ETA and PIA influence intergranular embrittlement in the various ferritic alloys.

Journal Articles

Irradiation embrittlement of 2.25Cr-1Mo steel at 400$$^{circ}$$C and its electrochemical evaluation

Nishiyama, Yutaka; Fukaya, Kiyoshi; Suzuki, Masahide; Eto, Motokuni

Journal of Nuclear Materials, 258-263, p.1187 - 1192, 1998/00

 Times Cited Count:4 Percentile:38.68(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Electrochemical evaluation of thermal aging embrittlement of 21/4Cr-1Mo steel for a nuclear pressure vessel

Nishiyama, Yutaka; Fukaya, Kiyoshi; Suzuki, Masahide; Eto, Motokuni; *

Small Specimen Test Techniques; ASTM STP 1204, p.16 - 26, 1993/00

no abstracts in English

JAEA Reports

DBTT measurement by use of small punch(SP) test

Suzuki, Masahide; Fukaya, Kiyoshi; Nishiyama, Yutaka; Eto, Motokuni

JAERI-M 92-086, 44 Pages, 1992/06

JAERI-M-92-086.pdf:2.79MB

no abstracts in English

Journal Articles

Thermomechanical Treatment Aimed at The Protection of Type 316 Stainless Steel from IGC and IGSCC

; Kondo, Tatsuo

Boshoku Gijutsu, 32(9), p.503 - 511, 1983/00

no abstracts in English

Journal Articles

Journal Articles

Effect of thermal neutron irradiation on mechanical properties of alloys for HTR core application

Ogawa, Yutaka; Kondo, Tatsuo; ;

Proc.of 2nd Japan-US HTGR Safety Technology Seminar,Material Properties and Design Method Session, 9 Pages, 1978/00

no abstracts in English

Oral presentation

Evaluation of mechanical property in grain boundary character distribution-optimized Ni-based alloy

Yamashita, Shinichiro; Sekio, Yoshihiro; Sakaguchi, Norihito*; Shibayama, Tamaki*; Watanabe, Seiichi*; Kokawa, Hiroyuki*

no journal, , 

Oral presentation

Microstructural analysis on Japanese RPV steels irradiated in PWR, 2; Grain-boundary phosphorus segregation

Hata, Kuniki; Nagai, Yasuyoshi*; Nishiyama, Yutaka

no journal, , 

no abstracts in English

14 (Records 1-14 displayed on this page)
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