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材料と環境 Vol. 48 (1999), No. 12

ISIJ International
belloff
オンライン版ISSN: 1881-9664
冊子版ISSN: 0917-0480
発行機関: Japan Society of Corrosion Engineering

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  6. Vol. 68 (2019)

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  7. Vol. 67 (2018)

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  8. Vol. 66 (2017)

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  9. Vol. 65 (2016)

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  10. Vol. 64 (2015)

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  11. Vol. 63 (2014)

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  12. Vol. 62 (2013)

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  13. Vol. 61 (2012)

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  14. Vol. 60 (2011)

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  15. Vol. 59 (2010)

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  16. Vol. 58 (2009)

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  17. Vol. 57 (2008)

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  18. Vol. 56 (2007)

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  19. Vol. 55 (2006)

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  20. Vol. 54 (2005)

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  21. Vol. 53 (2004)

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  22. Vol. 52 (2003)

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  24. Vol. 50 (2001)

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  26. Vol. 48 (1999)

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  27. Vol. 47 (1998)

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  28. Vol. 46 (1997)

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  29. Vol. 45 (1996)

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  30. Vol. 44 (1995)

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  31. Vol. 43 (1994)

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  32. Vol. 42 (1993)

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  33. Vol. 41 (1992)

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  34. Vol. 40 (1991)

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材料と環境 Vol. 48 (1999), No. 12

Improvement of Reliability in Nuclear Power Plants by Water Chemistry Technology

Hideo Hirano

pp. 747-752

抄録

As of 1999, 28 BWR units and 23 PWR units are operated, and the generation of electric power by light water reactors accounts for 35% of the total electric power generation in Japan. However, light water reactors have experienced many corrosion problems since the first commercial operation started in 1970.
To cope with these corrosion problems, variour countermeasurements have been developed and employed. In the case of BWRs, nuclear grade 316LC stainless and new weld methods were developed in order to prevent pipe cracking in recirculation system. Hydrogen injection are also employed to prevent the SCC of in-reactor materials. On the other hand, in the case of PWRs, thermally treated alloy 690 (690TT) was developed and employed to prevent the IGA/SCC for Steam Generator tubing. Zinc injection has been investigated to control the radiation build-up in both BWRs and PWRs.
In this report, the status of advancement of reliability in nuclear power plant by the improvement of materials performance, water chemistry and design in Nuclear Power Plant is reviewed, and the subjects which will be dealed with in the future are discussed.

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Corrosion Behavior and Its Countermeasures of BWR Plant Components

Shunichi Suzuki

pp. 753-762

抄録

EAC (Environmental assisted cracking) such as SCC and corrosion fatigue is one of the most important phenomena to control the total life time of structural components of nuclear power plants. As has been well recognized, they are controlled by many parameters as sensitization of materials, water chemistry and stress/strain.
In this paper, EAC behavior of the materials used as BWR structural components such as Low alloy steel, carbon steel, stainless steel and Ni base alloy as well as IASCC (Irradiation Assisted SCC) of stainless steel are widely discussed. In addition, their mechanisms and various kinds of countermeasures are also summarized in terms of their controlling parameters mentioned above.

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Improvement of PWR Reliability by Corrosion Prevention

Hiroshi Takamatsu

pp. 763-770

抄録

Since first PWR in Japan started commercial operation in 1970, we have encountered the various modes of corrosion on primary and secondary side components. We have paied much efforts for resolving these corrosion problems, that is, investigating the causes of corrosion and establishing the countermeasures for these corrosion. We summarize these efforts in this article.

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Improvement of Reliability in Nuclear Fuel Reprocessing Plant

Fumihiro Wada

pp. 771-775

抄録

JNFL's Rokkasyo Reprocessing Plant (RRP) is under construction in Rokkasyo village Aomori prefecture, and it is required to select appropriate materials of equipment which possess high degree of corrosion resistance to process solutions based on nitric acid. In other existing reprocessing plants, some troubles caused by corrosion of stainless steel has been reported, and almost of these troubles had been occurred at the equipment used in boiling nitric acid condition. In consideration of these experiences and results, in RRP, distillation under reduced pressure technology adopted and zirconium was selected for boiling nitric acid condition under normal pressure.

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Short Crack Behavior on PWSCC of Mill Annealed Alloy 600

Masayuki Kamaya, Syunji Sakai, Nobuo Totsuka, Nobuo Nakajima

pp. 790-795

抄録

A short crack behavior is essential to construct a life time prediction model of a light water reactor component. The objective of this study is to evaluate a short crack behavior on primary water stress corrosion cracking (PWSCC) of mill annealed alloy 600. Constant load tests were conducted to investigate a short crack growth rate. Test time was fixed at 1000, 3000, 3880 and 6940 hr and no specimen was fractured during the test. Crack length was measured for all cracks and plotted to the Weibull probabilistic sheet. By calculating the maximum crack length using the Weibull parameter at each test time, crack growth rates were obtained. We found there were 2 kinds of crack growth stages. This means short cracks transform from slow speed stage to fast speed stage. Furthermore distribution of inclination of grain boundaries, which were cracked was investigated to consider the factors which influence to crack propagation. Consequently it is revealed that the crack propagation force is expressed as a function of not only stress intensity factor K but plastic strain and grain boundaries energy. Finally Monte Carlo simulation was performed to confirm the above considerations. As a result of simulation, we could clarify the reasons of existence of 2 kinds of short crack growth stages.

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Evaluation of Corrosivity of Atmosphere for Stainless Steels by Meteorological Data

Tadashi Shinohara, Shin-ichi Motoda, Koji Nabeshima, Yohnosuke Suzuki, Shigeo Tsujikawa

pp. 796-806

抄録

Makuhari-Messe Convention Center building is located in a seashore area, 500m from the sea, at Chiba city in Chiba prefecture. Its roof and eaves are made of 22Cr-0.8Mo stainless steel. To evaluate the corrosivity of atmosphere at the eaves in this building, ACM(Atmospheric Corrosion Monitor) sensors and 22Cr-0.8Mo steel specimens were exposed at five sites, where Sites 1 and 2 were located at the south-west side of the building, Side A, facing the sea, and Sites 3, 4 and 5 were located at the north-east side, Side B, behind the sea. They had been exposed since '94.11 and renewed every 1-4 months until '96.11. The outputs of ACM sensors were measured every 10 minutes together with temperature and relative humidity (RH). The amount of deposited sea salt, Ws, was also determined based on ACM sensor outputs combined with RH data. The 22Cr-0.8Mo steel specimens were rusted only at Sites 1 and 2 when Ws increased more than 1g/m2, during a single exposure period, '95.4.18-'95.6.22, in the whole exposure period of 2 years ('94.11-'96.11). The day, when Ws exceeded 1g/m2 and make the steel rusted, could be identified based on the effective coefficient of wind-force energy in a day, α*D.

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Hydrogen Absorption of Titanium for Nuclear Waste Container in Reducing Condition

Haruo Tomari, Tsuyoshi Masugata, Kazutoshi Shimogori, Tsutomu Nishimura, Ryutaro Wada, Akira Honda, Naoki Taniguchi

pp. 807-814

抄録

Effects of bentonite clay, applied potential, pH of solution and cathodic polarization time on hydrogen absorption into titanium, which is one of the candidate materials of over-pack for high-level radioactive waste container, have been investigated in artificial underground water. Considering the result at various test time, and assuming the hydrogen absorption is ruled by the parabolic law, the amount of hydrogen after 1000 year exposure calculated to about 17 ppm, which will be absorbed at the applied potential of -0.51V vs. SHE corresponds to equilibrium potential of hydrogen. It seems that the assumption of the parabolic law and the test period up to 1440 h are proper, because the linear relations were obtained between the amount of absorbed hydrogen and the logarithm of the averaged cathodic current and between the slopes of the lines and a square root of the test time. Titanium seems to have a life over 1000 years in deep underground depository according to assumption that about 500 ppm absorbed hydrogen is critical for hydrogen embrittlement of titanium.

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The Dissolution Behavior of Type 304, Type 316 and Type 430 Stainless Steels during the Process of Stress Corrosion Cracking

Rokuro Nishimura

pp. 815-821

抄録

The dissolution behavior during the SCC process of type 304, type 316 and type 430 stainless steels was investigated in 0.82 kmol/m3 HCl or 0.82 kmol/m3 chloride (pH 1.0) solutions by using ICP under constant applied stress or constant slow strain rate condition. In addition, the maximum stress (σmair) of the specimens interrupted at various elongations undera constant applied stress was measured at room temperature by using an Instron tensile maschine. The dissolution behavior of the specimens under both the conditions changed at tss, the transition time in corrosion elongation curve and at tm at which a stress becomes the maximum stress (σm) obtained under constant slow strain rate; that is, the dissolution rate before tss and tm was larger than that after those. The σmair decreased with time up to failure, but showed a rapid small change at tss. The results obtained were discussedin terms of the number of cracks, selective dissolution, active dissolution and reduction in cross sectional area.

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