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Hata, Kuniki; Hanawa, Satoshi; Chimi, Yasuhiro; Uchida, Shunsuke; Lister, D. H.*
Journal of Nuclear Science and Technology, 60(8), p.867 - 880, 2023/08
Times Cited Count：2 Percentile：53.91(Nuclear Science & Technology)One of the major subjects for evaluating the corrosive conditions in the PWR primary coolant was to determine the optimal hydrogen concentration for mitigating PWSCC without any adverse effects on major structural materials. As suitable procedures for evaluating the corrosive conditions in PWR primary coolant, a couple of procedures, i.e., water radiolysis and ECP analyses, were proposed. The previous article showed the radiolysis calculation in the PWR primary coolant, which was followed by an ECP study here. The ECP analysis, a couple of a mixed potential model and an oxide layer growth model, was developed originally for BWR conditions, which was extended to PWR conditions with adding Li
(Na) and H effects on the anodic polarization curves. As a result of comparison of the calculated results with INCA in-pile-loop experiment data as well as other experimental data, it was confirmed that the ECPs calculated with the coupled analyses agreed with the measured within 
100mV discrepancies.
Kurata, Yuji; Kikuchi, Kenji; Saito, Shigeru; Kamata, Kinya*; Kitano, Teruaki*; Oigawa, Hiroyuki
Proceedings of 4th International Workshop on the Utilisation and Reliability of High Power Proton Accelerators, p.267 - 277, 2004/05
R&D on lead-bismuth technology have been conducted for accelerator driven system. From the test results of 316SS using JLBL-1 for 3000h without active oxygen control, mass transfer from high temperature to low temperature parts was observed. It was found that deposition of Pb/Bi and Fe-Cr grains in the annular channel of the electro-magnetic pump caused plugging and decrease in flow rate. The modification of the loop system brought about a good effect on operation. Significant erosion/corrosion was not observed in the experiment using MES loop for 1000h under 10
wt.% oxygen condition. The results of static corrosion tests showed the following: corrosion depth decreased at 450
C with increasing Cr content in steels while corrosion depth of JPCA and 316SS became larger due to ferritization caused by dissolution of Ni and Cr at 550C. Si-added steel exhibited good corrosion resistance at 550C.
Kurata, Yuji; Kikuchi, Kenji; Saito, Shigeru; Futakawa, Masatoshi; Sasa, Toshinobu
FZKA-6876, p.190 - 198, 2003/12
A report at MEGAPIE(Megawatt Pilot Experiment) Technical R & D Meeting is collected into an FZK report. According to the static corrosion tests, Al surface-treated layer produced by the gas diffusion method exhabited corrosion resistance to liquid Pb-Bi, while Al surface-treated layer produced by the melt dipping method suffered a severe corrosion attack. Furtheremore, it was found that a thick ferrite layer was formed in the surface of austenitic stainless steel at 550C. Dissolution at high temperature parts, precipitation of Fe-Cr alloy and deposition of PbO at low temperature parts occured in the first loop corrosion test. These caused plugging of the narrow passage in electro-magnetic pumu(EMP) system. Adoption of filters and a wide passage in an EMP system, and the use of an inner-polished tube specimen brought about a good effect.
Kurata, Yuji; Futakawa, Masatoshi; Kikuchi, Kenji; Saito, Shigeru; Osugi, Toshitaka
Journal of Nuclear Materials, 301(1), p.28 - 34, 2002/02
Times Cited Count：52 Percentile：93.86(Materials Science, Multidisciplinary)no abstracts in English
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.
; Shindo, Masami; Kondo, Tatsuo
JAERI-M 83-093, 16 Pages, 1983/07
no abstracts in English
