Evaluation of Laser Peening for Mitigation of Primary Water Stress Corrosion Cracking in Pressurized Water Reactors

Author(s):  
Stephen Marlette ◽  
Stan Bovid

Abstract For several decades pressurized water reactors have experienced Primary Water Stress Corrosion Cracking (PWSCC) within Alloy 600 components and welds. The nuclear industry has developed several methods for mitigation of PWSCC to prevent costly repairs to pressurized water reactor (PWR) components including surface stress improvement by peening. Laser shock peening (LSP) is one method to effectively place the surface of a PWSCC susceptible component into compression and significantly reduce the potential for crack initiation during future operation. The Material Reliability Program (MRP) has issued MRP-335, which provides guidelines for effective mitigation of reactor vessel heads and nozzles constructed of Alloy 600 material. In addition, ASME Code Case N-729-6 provides performance requirements for peening processes applied to reactor vessel head penetrations in order to prevent degradation and take advantage of inspection relief, which will reduce operating costs for nuclear plants. LSP Technologies, Inc. (LSPT) has developed and utilized a proprietary LSP system called the Procudo® 200 Laser Peening System. System specifications are laser energy of 10 J, pulse width of 20 ns, and repetition rate of 20 Hz. Scalable processing intensity is provided through automated focusing optics control. For the presented work, power densities of 4 to 9.5 GW/cm2 and spot sizes of nominally 2 mm were selected. This system has been used effectively in many non-nuclear industries including aerospace, power generation, automotive, and oil and gas. The Procudo® 200 Laser Peening System will be used to process reactor vessel heads in the United States for mitigation of PWSCC. The Procudo® 200 Laser Peening System is a versatile and portable system that can be deployed in many variations. This paper presents test results used to evaluate the effectiveness of the Procudo® 200 Laser Peening System on Alloy 600 material and welds. As a part of the qualification process, testing was performed to demonstrate compliance with industry requirements. The test results include surface stress measurements on laser peened Alloy 600, and Alloy 182 coupons using x-ray diffraction (XRD) and crack compliance (slitting) stress measurement techniques. The test results are compared to stress criteria developed based on the performance requirements documented in MRP-335 and Code Case N-729-6. Other test results include surface roughness measurements and percent of cold work induced by the peening process. The test results demonstrate the ability of the LSP process to induce the level and depth of compression required for mitigation of PWSCC and that the process does not result in adverse conditions within the material.

Author(s):  
Donald C. Adamonis ◽  
Robert P. Vestovich ◽  
Fred G. Whytsell ◽  
Filippo D’Annucci ◽  
Eric Loehlein ◽  
...  

Several pressurized water reactors have experienced primary coolant leaks as a result of degradation in the tubes and J-groove welds of reactor vessel head penetrations. Leakage has been attributed to primary water stress corrosion cracking (PWSCC) of the Alloy 600 nozzle material and Alloy 182/82 weld materials. More recently, other Alloy 600 components including reactor vessel bottom mounted instrumentation nozzles, dissimilar metal pipe welds, hot leg instrument penetrations, and pressurizer heater sleeves have exhibited degradation. Westinghouse has been actively involved in the development of a comprehensive Alloy 600 degradation management program including investigation of root cause, establishing a safety position, and developing inspection and repair/replacement strategies to address Alloy 600 degradation issues. Personnel from Germany, Sweden and the United States have cooperatively developed equipment and nondestructive examination technologies for identification and characterization of degradation that might exist in these Alloy 600 components and attachment welds. These developments represent significant enhancements to technologies and equipment previously available in the industry and are driven by the need to meet new inspection applications and industry requirements which have continued to evolve over the last several years. This paper describes the inspection capabilities available to support Alloy 600 degradation management programs, field experience with those inspection technologies, and the status of ongoing NDE development efforts to enhance future Alloy 600 inspection programs.


2003 ◽  
Vol 40 (7) ◽  
pp. 509-516 ◽  
Author(s):  
Takumi TERACHI ◽  
Nobuo TOTSUKA ◽  
Takuyo YAMADA ◽  
Tomokazu NAKAGAWA ◽  
Hiroshi DEGUCHI ◽  
...  

Author(s):  
Hiroyuki Adachi ◽  
Akira Ito ◽  
Kazuto Imasaki ◽  
Masaki Yoda ◽  
Itaru Chida ◽  
...  

Primary water stress corrosion cracking (PWSCC) is a degradation process that has plagued nickel alloy components and welds in the nuclear industry for decades. Numerous mitigation techniques have been developed over the years that help reduce the potential for cracking in nickel alloy components exposed to the primary water environment. One such method is Laser Peening (LP), which improves the stress properties and helps to reduce the potential for crack initiation. The LP process has been applied in Japan to both boiling water reactors (BWR) and pressurized water reactors (PWR) for stress corrosion mitigation. The first application of LP in the US for the nuclear industry was applied in the fall of 2016 to the bottom mounted instrumentation (BMI) nozzles of a PWR. The bottom mounted nozzles are made from Alloy 600 tubing and attached with Alloy 82/182 welds, which are known to be susceptible to PWSCC. In order to prevent crack initiation, it is important for the peening mitigation process to induce sufficient compressive stress on the surface of the susceptible materials. However, it is not practical to take stress measurements directly on the reactor components in order to verify compression. Thus, the magnitude of compression induced by the LP process was verified prior to the application at the plant using mockups of the BMI nozzles. As a part of the qualification process, test coupons were peened and stress measurements were taken using X-ray diffraction (XRD). The results of the stress measurements demonstrate that sufficient surface compression was achieved by the LP process in order to provide PWSCC mitigation. This paper presents and discusses key stress measurement results taken during the qualification process for the first application of LP at a U.S. nuclear plant. Although not directly applicable in this case, the guidance in ASME Code Case N-729 Mandatory Appendix II and MRP-335 for PWR upper head nozzles was generally followed.


Author(s):  
Robert Ge´rard ◽  
Fre´de´ric Somville

The baffle to former bolts are used in Pressurized Water Reactors to attach the baffle plates to the former plates in the reactor vessel lower internals. The resulting structure forms a boundary for the flow of coolant and provides lateral support to the fuel assemblies. Some edge bolts are also present, assembling together the baffle plates. After an operating time of the order of 120 000 hours, some bolts exhibit cracking at the junction of the head and the shaft of the bolt. Examinations of failed bolts have made it possible to identify the cause of cracking as irradiation assisted stress corrosion cracking (IASCC). Up to now, baffle bolt cracking has been detected in units older than 15 years, where the baffle bolts are not cooled (no holes in the former to allow a water flow on the bolt shaft). In Belgium the concerned unit are Tihange 1 and Doel 1–2. The paper summarizes the experience with baffle bolts cracking in Belgian units and the strategy implemented to mitigate this problem, consisting of structural integrity analyses, baffle bolts inspections and replacement, and research programs in the field of IASCC, including examinations of highly irradiated replaced bolts.


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