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Of the twenty-three pressurized water reactor (PWR) plants currently operating in Japan, the older ones are over 30 years old, and damage to aged plants, such as stress corrosion cracking (SCC), has been observed. This paper describes countermeasures against SCC and introduces the latest maintenance techniques for ensuring the safe and steady operation of PWR plants, such as inspection techniques to confirm the integrity of components, mitigation (degradation reducing) techniques to prevent the occurrence of damage, and repair and replacement techniques for any damage detected.
It is important to prevent various problems arising from damage and/or leakage in nuclear power plants. This involves diagnosing the life remaining, reducing degradation and repairing components such as pressure vessels and piping. This is a major problems for aged plants especially.
One conspicuous problem seen in aged plants is the SCC (stress corrosion cracking) issue, which is consid- ered to be jointly caused by the environment, the materials and stress, as shown in Fig. 1. The damage caused by SCC to high nickel based alloy (Alloy 600 be- low) in pressurized water reactor (PWR) plants has become conspicuous both in Japan and abroad. Figure 2 shows the regions where Alloy 600 is used in conven- tional PWR plants. The stress corrosion cracking of Alloy 600 in PWR plants is called PWSCC (primary water SCC), and it is considered that PWSCC will occur when parts are used under high residual stress caused by machining, welding and so on.
In order to prevent damage and leakage and to en- sure the long-term steady operation of plants, we have developed the inspection and maintenance techniques for reducing the degradation shown in Fig. 1.
Maintenance Technologies for SCC which Support Stable Operations of Pressurized Water Reactor Power Plants
KOJI OKIMURA*1 MASAYUKI MUKAI*1 KAZUHIKO KAMO*2
NOBUYUKI HORI*1 KOICHIRO MASUMOTO*1 MASAAKI KUROKAWA*2
High-temperature environment in primary water
Reduction of environmental temperature
Tensile residual stress Improved to
Having susceptibility of SCC Material improvement
(replacement, cladding, etc.)
Reactor vessel head penetration nozzle Reactor vessel head penetration nozzle joint
Reactor vessel outlet nozzle safe end joint
Reactor vessel inlet nozzle safe end joint
Reactor vessel bottom mounted instrument Reactor vessel bottom mounted instrument joint
Pressurizer nozzle to safe end joints
Steam generator tube
Steam generator inlet nozzle safe end joint
Steam generator outlet nozzle safe end joint
Fig. 2 Region used Alloy 600 in primary system of PWR plants in Japan
Fig. 1 Three factors causing stress corrosion cracking and mitigation methods
*1 Kobe Shipyard & Machinery Works
*2 Takasago Research & Development Center, Technical Headquarters
Mitsubishi Heavy Industries, Ltd. Technical Review Vol. 43 No. 4 (Dec. 2006)
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