Maintenance Technologies for SCC which Support Stable Operations of Pressurized Water Reactor Power Plants

Transcription

1 Maintenance Technologies for 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 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 (), has been observed. This paper describes countermeasures against 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. 1. Introduction 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 (stress corrosion cracking) issue, which is considered to be jointly caused by the environment, the materials and stress, as shown in Fig The damage caused by to high nickel based alloy (Alloy 600 below) 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 conventional PWR plants. The stress corrosion cracking of Alloy 600 in PWR plants is called PW (primary water ), and it is considered that PW 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 ensure the long-term steady operation of plants, we have developed the inspection and maintenance techniques for reducing the degradation shown in Fig. 1. Having susceptibility of Material improvement (replacement, cladding, etc.) Environment High-temperature environment in primary water Reduction of environmental temperature Material Tensile residual stress Improved to compressive stress Fig. 1 Three factors causing stress corrosion cracking and mitigation methods Reactor vessel head penetration nozzle Reactor vessel head penetration nozzle joint Pressurizer nozzle to safe end joints Reactor vessel outlet nozzle safe end joint Steam generator tube Reactor vessel inlet nozzle safe end joint Steam generator inlet nozzle safe end joint Reactor vessel bottom mounted instrument Reactor vessel bottom mounted instrument joint Steam generator outlet nozzle safe end joint Fig. 2 Region used Alloy 600 in primary system of PWR plants in Japan *1 Kobe Shipyard & Machinery Works *2 Takasago Research & Development Center, Technical Headquarters 1

2 2. Inspection Technique The inspection techniques require higher defect detection and defect quantification accuracy, in order to ensure the reliability of components. In the steam generator tube (composed of about 4,000 small tube bundles with an inner diameter of approximately 20 mm) shown in Fig. 3, 3 the intelligent ECT (eddy current test), which are densely arranged coils using a multi-induction system for higher noise reduction, allow precise inspection. This technique was developed and applied in order to shorten the inspection time and improve the accuracy of the inspection of shallow defects. Further, the phased array UT (ultrasonic test) has been developed for pipe welding. In this method, the direction and focal position of an ultrasonic-wave beam can be freely changed by the UT probe, which is composed of multiple piezoelectric composites, which improves the accuracy of the quantification of shallow defects and defects in regions with complex shapes, which were conventionally considered difficult regions. Alloy 600 is used for the reactor vessel nozzle safe end joints (connecting nozzle and piping) and other joints with complex shapes and materials. Surface inspection is considered very effective for detecting PW where it is caused by the surface contacting water. Accordingly, we developed and verified the ECT techniques with remote control and automatic search. Figure 4 shows an example of the inspection technique for the reactor vessel inlet/outlet nozzle safe end joints (Alloy 600), where even extremely tiny flaws in welding can be detected by ECT. Further, by allowing the installation of multiple inspection heads, different inspection methods can be simultaneously applied, for example, when a defect is detected by ECT, defect quantification (confirmation of flaw depth) can also be carried out by using a UT probe capable of measuring defect depth. Further, improved detectability is also required for confirming the integrity of components before applying preventative maintenance techniques (for example, stress improvement techniques). The synergistic effect of securing integrity by inspections prior to stress improvement techniques, will suppress damage by reducing degradation, which is considered to further improve reliability. 2 coils for defect detection Axial direction Tube Thin-film array coil Inclined drive coil Realization of array coil for enhanced delectability and high-speed detection Circumferential direction Tube expansion transition Fig. 3 The steam generator tube inspection technique of Intelligent ECT Outer surface Joint (Alloy 600 weld metal) Inner surface (low-alloy steel) Reactor vessel side Stainless steel cladding Defect signal Alloy 600 weld metal Example of ECT detection Capable of mounting multiple inspection heads, enabling diverse inspections including ECT, UT, etc. Fig. 4 Inspection technique for Alloy 600 welding Example of the reactor vessel outlet/inlet nozzle safe end joints 2

3 3. Improvement Technique As a countermeasure against high residual stress, shot peening is used for components installed in air and water jet peening is used for components installed in water, in order to change the residual tensile stress into compressive stress of surfaces contacting the primary water. Figure 5 shows an outline of these peening techniques. It has been confirmed that the stress near the inner surface can be made into compressive stress by the plastic strain applied to the surface by using cavitation shock pressure during the water jet peening and by using the collision force of the shot material during the shot peening. For the shot peening, we apply a new method employing ultrasonic-wave vibration. Using a high power piezo-electric element as the drive source, we use shot material with a larger particle size. This technique will deeply relieve the tensile stress at the surface and allow easy control of the disposed shot material. On the other hand, when access to the inner surface of piping is difficult, it is necessary to relieve the tensile stress using an outer surface approach. One such technique, the outer surface irradiated laser stress improvement process, which we call L-SIP, is the use of the thermal expansion strain caused by the temperature difference between the surface walls generated by the high speed heating of the outer surface by a high power laser beam moving along the outer surface. Since high speed heating is possible through laser irradiation, the method can be used easily and simply by selecting the appropriate conditions without requiring water cooling of the inner surface. Figure 6 shows an example of the application of this method to the nozzle joints, where it is confirmed that the high tensile residual stress at and around the welded region on the inner surface has been relieved into compressive stress. Water jet peening Ultrasonic shot peening Principle and feature Water jet with cavitations bubbles Plastic strain applied to the surface by shock pressure of cavitations collapse Material surface relieved into compressive stress due to cavitation impact caused by water jet Example of the water jet peening (the reactor vessel bottom mounted instruments mock-up) Shot material drive source (piezo-electric element) Shot material with larger particle size Plastic strain applied to the surface by shot collision force Material surface relieved into compressive stress by shot material collision force Water jet nozzle The surface and the surrounding relieved into compressive stress at tensile residual stress field Water jet with cavitations bubbles Fig. 5 improvement technique by peening Principle and feature Laser oscillator During heating After cooling Temperature distribution Temp Temp Example of the outer surface irradiated laser stress improvement process + + distribution Temperature at inner and outer surfaces of piping during heating (analysis) When the outer surface is subjected to high-speed heating, temperature distribution occurs in the sheet metal, resulting in thermal expansion strain, causing the inner surface to undergo compressive stress during the process of cooling. A high energy laser is used as the heat source and as it is movable, high-speed heating as well as suppression of temperature rise on the inner surface can be obtained at the same time. Optical head The inner surface can undergo compressive stress even in a tensile residual stress field (welding residual stress). Fig. 6 Outer surface irradiated laser stress improvement technique 3

4 4. Improved Material and Cladding Techniques When a surface touches the primary system feed water a countermeasure employing high nickel based alloy TT690 (Alloy 690), which has excellent PW resistance, is used. Alloy 690 has double the chromium (Cr) content of Alloy 600. It is used for cladding, and prevents the Alloy 600 from coming in contact with the water by overlay welding its surface (with Alloy 690) or by replacement, which is described in the next section. As for an example of the application of the degradation technique to the reactor vessel outlet nozzle safe end joints (using welded Alloy 600), the cladding technique on the reactor vessel nozzle welded region is shown in Fig In order to ensure a flow passage and to allow inspection, a groove is machined at surface before cladding (overlay welding). An automatic 3-layer, temper-bead welding process, not requiring post-weld heat treatment, has been developed and applied to actual reactors. The cladding repair technique has also been applied to reactor vessel nozzle head penetration joints. 5. Replacement with improved material When components such as the reactor vessel head, the steam generator, and the reactor internals (core internals) are replaced, the new components are made with material with excellent resistance. As a recent example of component replacement, the replacement of the core internals is shown in Fig Since a PWR's core internals are designed to be removed, they can be removed as a unit, stored and preserved in a container, and new reactor internals installed. The replacement of the core internals has been carried out for the first time in the world, being done in three PWR plants in Japan. A partial replacement technique (replacement of nozzle safe end joint) is also used when total replacement is difficult and to repair damage and for preventative maintenance. An example of this is shown in Fig. 9, the replacement of a reactor vessel outlet nozzle safe end joint. In this method, the welding is done using Alloy 690 welding material with excellent resistance when Alloy 600 joints are removed and replaced with new ones. Outline Inner surface (Alloy 600) is clad (overlay welding) with Alloy 690 on preventing the occurrence of. In order to secure a flow passage equivalent to the conventional one and for inspectability, clad welding is performed at the channel region. In order to protect the low alloy steel from the heat, 3-layer temper-bead welding is performed. Cladding with Alloy 690 (low alloy steel) Safe end (stainless steel) Welding (Alloy 600) Image of on-site work View of automatic welding on the nozzle inner surface Fig. 7 Cladding technique: Example of the reactor vessel outlet nozzle safe end joint (3) Carrying out through the equipment hatch (1) Lifting and storage of the old core internals Ceiling (Polar) crane Temporary floor Temporary lifting equipment Outer shield Container for preserving the old core internals Containment vessel Reversing pedestal (2) Reversing and horizontal removal of the old core internals View of installing the new core internals to the reactor vessel Fig. 8 component replacement technique: Example of the reactor core internals 4

5 Outline After removing Alloy 600 joint, new member (spool piece) is welded using Alloy 690 with excellent resistance. Special welding process (temper-bead welding or local heat treatment of narrow-region) is used to prevent the low-alloy steel from the heat effects of welding. Welding step 1: Butte ring Alloy 600 joint (low-alloy steel) Main coolant pipe (stainless steel) Welding step 2: Joint welding Fig. 9 Partial replacement technique: Example of the reactor vessel outlet nozzle safe end joint Removal of defect (drilled repair) If size of the defect is within the permissible range of the component structure, the bottom mounted instrument is drilled for removing the defect by enlarging the inner diameter. Installation of cap for new boundary (cap repairing) The cap for new boundary is installed outside the bottom mounted instruments to prevent the leakage from component by retaining the leakage from the defect inside the cap. Enlargement of inner diameter Original inner diameter Initial (inner diameter 10mm) The bottom mounted instruments The reactor vessel lower head Assumed defect Cap The bottom mounted instruments After drilling (inner diameter 16mm) Fig. 10 Repair technique: Example of the reactor vessel bottom mounted instrument 6. Repair Technique When damage is detected and repair is required, it is possible to remove the defective part and install a new boundary component. Figure 10 shows an example of the repair of reactor vessel bottom mounted instruments. The removal of defects is a technique of partial removal when the component is structurally safe after removing the defect. The installation of a component for the new boundary is installed outside the damaged component to prevent leakages from the defect. 7. Conclusion In this paper, we have explained the latest maintenance techniques including inspection, mitigation (stress improvement and material improvement), replacement and repair. We consider that these techniques contribute to the safety and reliability of nuclear power plants for the continuous economic efficiency of plant operation over a long time period. In the future, we will continue to develop and verify these techniques. Koji Okimura Nobuyuki Hori Masayuki Mukai Koichiro Masumoto Kazuhiko Kamo Masaaki Kurokawa 5