Advanced Nondestructive Examination Technologies for Alloy 600 Components

2004 ◽  
Author(s):  
Donald C. Adamonis ◽  
Robert P. Vestovich ◽  
Fred G. Whytsell ◽  
Filippo D’Annucci ◽  
Eric Loehlein ◽  
...  
Keyword(s):  
The United States ◽  
Management Program ◽  
Reactor Vessel ◽  
Alloy 600 ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Nondestructive Examination ◽  
The Status ◽  
Primary Water ◽  
Water Reactors

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.

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

2019 ◽  
Author(s):  
Stephen Marlette ◽  
Stan Bovid
Keyword(s):  
Surface Stress ◽  
Reactor Vessel ◽  
Pressurized Water Reactors ◽  
Laser Peening ◽  
Alloy 600 ◽  
Test Results ◽  
Pressurized Water ◽  
Performance Requirements ◽  
Primary Water ◽  
Water Reactors

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.

Alloy 600 Cracking Prevention and Mitigation

2002 ◽  
Author(s):  
Edward A. Siegel ◽  
William M. Connor ◽  
David R. Forsyth ◽  
Manu Badlani ◽  
Paula A. Grendys
Keyword(s):  
Management Program ◽  
Nuclear Industry ◽  
Alloy 600 ◽  
Operating Life ◽  
Pressurized Water ◽  
Weld Joints ◽  
Zinc Addition ◽  
Primary Water ◽  
Water Reactors ◽  
Head Temperature

Primary Water Stress Corrosion Cracking (PWSCC) has created a concern for pressurized water reactors (PWRs) in recent years, causing cracking in Alloy 600 materials. The domestic nuclear industry is currently focusing on short-term plans directed towards the reactor vessel (RV) head, and in at least one case, the Alloy 600 weld joint between the RV and the Hot Leg piping. There are many additional locations within the reactor coolant pressure boundary (RCPB) that contain Alloy 600 base metal or weld metal that may be susceptible to PWSCC over time. The predictive models being used have a large uncertainty band on when cracking might occur at specific locations within the RCPB. An informal poll of metallurgists leads to a consensus that the question is WHEN the cracking will occur at these other locations, not IF cracking will occur. While the industry is reacting to the RV head issues, there is an opportunity to plan a preventive aging management program that will preclude, or at the very least, dramatically reduce the incidence of cracking at many of these other locations. The benefits of a preventative program are clear when compared to recent examples of unplanned outage extensions due to the unexpected discovery of Alloy 600 cracking. In this paper, these other locations are identified and ranked on a simple risk basis generically for PWRs. At each location, preventative or mitigative techniques are described along with some historical perspective on each. The mitigative techniques that are discussed in this paper include the following. Zinc addition, which inhibits crack initiation and slows crack growth throughout the RCPB, is the only mitigative technique that benefits all wetted Alloy 600 surfaces within the RCPB, and is not a location-specific technique. Zinc addition also provides dose reduction as a second important benefit. MSIP (Mechanical Stress Improvement Process) is a proven, permanent solution for piping weld joints and has been successfully applied to over 1300 joints in BWRs. The NRC has accepted MSIP after rigorous qualification testing and field experience that demonstrated the effectiveness of this technique. The upper head temperature reduction program is a method of lowering the RV head temperature and is already in effect at a number of plants. Alloy 600 PWSCC is very sensitive to temperature; and lower temperatures can be an effective way to extend the useful life of a head. Weld overlay or encapsulation is a technique for creating a nonstructural fluid barrier on the inner diameter (ID) of pipe weld joints and on the wetted Alloy 600 surfaces of a reactor head. The new barrier effectively stops the corrosion process from continuing by isolating the susceptible material from the corrosive environment. For reactor heads, this may be an alternative to head replacement or a repair/mitigation at a specific location where flaws are discovered. It is recommended that that all plants evaluate these mitigative techniques for inclusion in an overall Alloy 600 Program. Proactive implementation will diminish, and possibly preclude, the incidence of cracking at some specific locations during the subsequent operating life of the plant.

Comparison of Analytical Methods to Study the Impact of a Postulated Closure Head Drop Accident on the Structural Integrity of the Bottom-Mounted Instrumentation System

2010 ◽  
Author(s):  
William C. Castillo ◽  
Geoffrey M. Loy ◽  
Joseph M. Remic ◽  
David P. Molitoris ◽  
George J. Demetri ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
The United States ◽  
Reactor Vessel ◽  
Primary Coolant ◽  
Heavy Load ◽  
Pressurized Water ◽  
System Pressure ◽  
The Impact

During typical nuclear power plant refueling activities for a pressurized water reactor (PWR), the reactor vessel closure head assembly must be removed from the reactor vessel (RV), transported for storage, and returned to the RV after refueling. This is categorized as a critical heavy load lift in NUREG-0612 [1] because a drop accident could result in damage to the components required to cool the fuel in the RV core. In order to mitigate the potentially severe consequences of a closure head drop, the United States Nuclear Regulatory Commission (USNRC) has mandated that nuclear power plants upgrade to a single failure-proof crane, show single failure-proof crane equivalence, or perform a head drop analysis to demonstrate that the core remains covered with coolant and sufficient cooling is available after the head drop accident. The primary coolant-retaining components associated with the RV are the inlet and outlet nozzles and the hot and cold leg main loop piping. Typical head drop analyses have considered these components to ensure that their structural integrity is maintained. One coolant-retaining component that has not been included in head drop evaluations on a consistent basis is the bottom-mounted instrumentation (BMI) system. In a typical Westinghouse PWR, 50 to 60 BMI nozzles are connected through the bottom hemisphere of the RV to one-inch diameter guide tubes which run under the vessel to a seal table above. Failure of the BMI system has the potential to adversely affect core coolability, especially if multiple failures are postulated within the system. A study was performed to compare static and dynamic methods of analyzing the effects of a head drop accident on the structural integrity of the BMI system. This paper presents the results of that study and assesses the adequacy of each method. Acceptability of the BMI system pressure boundary is based on the Nuclear Energy Institute Initiative (NEI 08–05 [2]) criteria for coolant-retaining components, which are based on Section III, Appendix F of the ASME Code [3].

Development of Alternative Reductant Application in PWR Primary Systems

2009 ◽  
Author(s):  
Masafumi Domae ◽  
Hirotaka Kawamura ◽  
Taku Ohira
Keyword(s):  
Electrochemical Corrosion ◽  
Crack Propagation Rate ◽  
Propagation Rate ◽  
Aliphatic Alcohols ◽  
High Concentration ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Radiation Chemical ◽  
Primary Water ◽  
Water Reactors

In primary coolant of pressurized water reactors (PWRs), high concentration dissolved hydrogen (DH) has been added, to prevent generation of oxidizing species through radiolysis of coolant. Recently, number of ageing plants is increasing and utilities are concerned about primary water stress corrosion cracking (PWSCC). Although mechanism of PWSCC is not fully clarified, some researchers consider that occurrence of PWSCC and crack propagation rate are affected by DH concentration. The authors consider that one of possible mitigation methods toward PWSCC is use of alternative reductant for hydrogen. From the radiation chemical aspect, aliphatic alcohols are typical scavengers of oxidizing radical generated through the radiolysis of water. The aliphatic alcohols are promising candidates of the alternative reductant. In the present work, possible alternatives of hydrogen were screened, and methanol was selected as the best candidate. Corrosion tests of type 304 stainless steels were carried out at 320°C in two conditions: (1) DH 1.5 ppm (part per million) and (2) methanol 2.9 ppm. Under two conditions, electrochemical corrosion potential was measured during the immersion tests. After the immersion tests, surface morphology of the stainless steel specimens was observed by scanning probe microscope. Major component of oxide film was analyzed by X-ray diffraction. From comparison of the test results, it is concluded that addition of methanol 2.9 ppm has almost the same effect as addition of DH 1.5 ppm.

scholarly journals Influence of Dissolved Hydrogen on Structure of Oxide Film on Alloy 600 Formed in Primary Water of Pressurized Water Reactors

2003 ◽  
Vol 40 (7) ◽  
pp. 509-516 ◽  
Author(s):  
Takumi TERACHI ◽  
Nobuo TOTSUKA ◽  
Takuyo YAMADA ◽  
Tomokazu NAKAGAWA ◽  
Hiroshi DEGUCHI ◽  
...  
Keyword(s):  
Oxide Film ◽  
Pressurized Water Reactors ◽  
Alloy 600 ◽  
Pressurized Water ◽  
Primary Water ◽  
Dissolved Hydrogen ◽  
Water Reactors

Advanced First Core Design for the Westinghouse AP1000

2009 ◽  
Author(s):  
Robert J. Fetterman
Keyword(s):  
Fuel Cycle ◽  
The United States ◽  
Pressurized Water ◽  
Loading Pattern ◽  
The Core ◽  
Low Leakage ◽  
Fuel Assemblies ◽  
Cycle Loading ◽  
Water Reactors ◽  
Under Construction

As the nuclear renaissance is now upon us and new plants are either under construction or being ordered, a considerable amount of attention has also turned to the design of the first fuel cycle. Requirements for core designs originate in the Utilities Requirements Document (URD) for the United States and the European Utilities Requirements (EUR) for Europe. First core designs created during the development of these documents were based on core design technology dating back to the 1970’s, where the first cycle core loading pattern placed the highest enrichment fuel on the core periphery and two other lower enrichments in the core interior. While this sort of core design provided acceptable performance, it underutilized the higher enriched fuel assemblies and tended to make transition to the first reload cycle challenging, especially considering that reload core designs are now almost entirely of the Low Leakage Loading Pattern (LLLP) design. The demands placed on today’s existing fleet of pressurized water reactors for improved fuel performance and economy are also desired for the upcoming Generation III+ fleet of plants. As a result of these demands, Westinghouse has developed an Advanced First Core (AFCPP) design for the initial cycle loading pattern. This loading pattern design simulates the reactivity distribution of an 18 month low leakage reload cycle design by placing the higher enriched assemblies in the core interior which results in improved uranium utilization for those fuel assemblies carried through the first and second reload cycles. Another feature of the advanced first core design is radial zoning of the high enriched assemblies, which allows these assemblies to be located in the core interior while still maintaining margin to peaking factor limits throughout the cycle. Finally, the advanced first core loading pattern also employs a variety of burnable absorber designs and lengths to yield radial and axial power distributions very similar to those found in typical low leakage reload cycle designs. This paper will describe each of these key features and demonstrate the operating margins of the AFC design and the ability of the AFC design to allow easy transition into 18 month low leakage reload cycles. The fuel economics of the AFC design will also be compared to those of a more traditional first core loading pattern.

A Risk-Informed Methodology for ASME Section XI, Appendix G

2012 ◽  
Vol 134 (3) ◽  
Author(s):  
Ronald Gamble ◽  
William Server ◽  
Bruce Bishop ◽  
Nathan Palm ◽  
Carol Heinecke
Keyword(s):  
Nuclear Regulatory Commission ◽  
The United States ◽  
Service Level ◽  
Pressure Vessels ◽  
Operational Flexibility ◽  
Pressurized Water ◽  
Leak Testing ◽  
Alternative Methodology ◽  
Water Reactors ◽  
Regulatory Commission

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code [1], Section XI, Appendix G provides a deterministic procedure for defining Service Level A and B pressure–temperature limits for ferritic components in the reactor coolant pressure boundary. An alternative risk-informed methodology has been developed for ASME Section XI, Appendix G. This alternative methodology provides easy to use procedures to define risk-informed pressure–temperature limits for Service Level A and B events, including leak testing and reactor start-up and shut-down. Risk-informed pressure–temperature limits provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials. This work evaluated selected plants spanning the population of pressurized water reactors (PWRs) and boiling water reactors (BWRs). The evaluation included determining appropriate material properties, reviewing operating history and system operational constraints, and performing probabilistic fracture mechanics (PFM) analyses. The analysis results were used to define risk-informed pressure–temperature relationships that comply with safety goals defined by the United States (U.S.) Nuclear Regulatory Commission (NRC). This alternative methodology will provide greater operational flexibility, especially for Service Level A and B events that may adversely affect efficient and safe plant operation, such as low-temperature-over-pressurization for PWRs and system leak testing for BWRs. Overall, application of this methodology can result in increased plant efficiency and increased plant and personnel safety.

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