Structural Integrity Evaluation of a Reactor Vessel Outlet Nozzle Weld Inlay

Author(s):  
N. L. Glunt ◽  
A. Udyawar ◽  
C. K. Ng ◽  
S. E. Marlette

Nickel-base weldments such as Alloy 82/182 dissimilar metal (DM) butt welds used in Pressurized Water Reactor (PWR) nuclear power plant components have experienced Primary Water Stress Corrosion Cracking (PWSCC), resulting in the need to repair/replace these weldments. The nuclear industry has been actively engaged in inspecting and mitigating these susceptible DM butt welds for the past several years. Full and Optimized Structural Weld Overlay as well as Mechanical Stress Improvement Process (MSIP®) are some of the mitigation/repair processes that have been implemented successfully by the nuclear industry to mitigate PWSCC. Three conditions must exist simultaneously for PWSCC to occur: high tensile stresses, susceptible material and an environment that is conducive to stress corrosion cracking. These mitigation/repair processes are effective in minimizing the potential for future initiation and crack propagation resulting from PWSCC by generating compressive residual stress at the inner surface of the susceptible DM weld. Weld inlay is an alternative mitigation/repair process especially for large bore nozzles such as reactor vessel nozzles. The weld inlay process consists of excavating a small portion of the susceptible weld material at the inside surface of the component and then applying a PWSCC resistant Alloy 52/52M repair weld layer on the inside surface of the component to isolate the susceptible DM weld material from the primary water environment. The design and analysis requirements of the weld inlay are provided in ASME Code Case N-766. This paper provides the structural integrity evaluation results for a typical reactor vessel outlet nozzle weld inlay performed in accordance with the ASME Code Case N-766 design and analysis requirements. The evaluation results demonstrate that weld inlay is also a viable PWSCC mitigation and repair process especially for large bore reactor vessel nozzles.

Author(s):  
E. A. Ray ◽  
K. Weir ◽  
C. Rice ◽  
T. Damico

During the October 2000 refueling outage at the V.C. Summer Nuclear Station, a leak was discovered in one of the three reactor vessel hot leg nozzle to pipe weld connections. The root cause of this leak was determined to be extensive weld repairs causing high tensile stresses throughout the pipe weld; leading to primary water stress corrosion cracking (PWSCC) of the Alloy 82/182 (Inconel). This nozzle was repaired and V.C. Summer began investigating other mitigative or repair techniques on the other nozzles. During the next refueling outage V.C. Summer took mitigative actions by applying the patented Mechanical Stress Improvement Process (MSIP) to the other hot legs. MSIP contracts the pipe on one side of the weldment, placing the inner region of the weld into compression. This is an effective means to prevent and mitigate PWSCC. Analyses were performed to determine the redistribution of residual stresses, amount of strain in the region of application, reactor coolant piping loads and stresses, and effect on equipment supports. In May 2002, using a newly designed 34-inch clamp, MSIP was successfully applied to the two hot-leg nozzle weldments. The pre- and post-MSIP NDE results were highly favorable. MSIP has been used extensively on piping in boiling water reactor (BWR) plants to successfully prevent and mitigate SCC. This includes Reactor Vessel nozzle piping over 30-inch diameter with 2.3-inch wall thickness similar in both size and materials to piping in pressurized water reactor (PWR) plants such as V.C. Summer. The application of MSIP at V.C. Summer was successfully completed and showed the process to be predictable with no significant changes in the overall operation of the plant. The pre- and post-nondestructive examination of the reactor vessel nozzle weldment showed no detrimental effects on the weldment due to the MSIP.


2021 ◽  
Author(s):  
Akihiro Mano ◽  
Jinya Katsuyama ◽  
Yinsheng Li

Abstract Probabilistic fracture mechanics (PFM) is expected as a more rational methodology for the structural integrity assessments of nuclear power components because it can consider the inherent probabilistic distributions of various influencing factors and quantitatively evaluate the failure probabilities of the components. The Japan Atomic Energy Agency (JAEA) has developed a PFM analysis code, PASCAL-SP, to evaluate the failure probabilities of piping caused by aging degradation mechanisms, such as fatigue and stress corrosion cracking in the environments of both pressurized water and boiling water reactors. To improve confidence in the analysis results obtained from PASCAL-SP, a benchmarking study was conducted together with the PFM analysis code, xLPR, which was developed jointly by the U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute. The benchmarking study was composed of deterministic and probabilistic analyses related to primary water stress corrosion cracking in a dissimilar metal weld joint in a pressurized water reactor surge line. The analyses were conducted independently by NRC staff and JAEA using their own codes and under common analysis conditions. In the present paper, the analysis conditions for the deterministic and probabilistic analyses are described in detail, and the analysis results obtained from the xLPR and PASCAL-SP codes are presented. It was confirmed that the analysis results obtained from the two codes were in good agreement.


Author(s):  
Dennis P. Weakland ◽  
Glenn White ◽  
Paul Crooker

This paper will discuss the ASME Code Committee activities involved in the incorporation of surface stress improvement (SSI) into ASME Code Cases N-770-4 and N-729-5. ASME Code Cases N-770 [1] and N-770-1 introduced several mitigation approaches for dissimilar metal weld (DMW) locations in PWR primary system piping and provided inspection relief for locations that were mitigated. The initial approaches contained in N-770 and N-770-1 included mechanical stress improvement and weld overlay methods that have a global stress relief effect to achieve a very low tensile surface stress state or a compressive stress state at the weld inside surface to halt crack initiation, as well as growth of acceptably sized cracks. The weld overlay mitigation methods are also effective because they introduce PWSCC-resistant material, i.e., Alloys 52, 152, or their variants. (The initial approaches also included Alloy 52/152 weld inlay and weld onlay, methods that do not require stress improvement but do require access to the weld inside surface.) While the mechanical stress improvement and weld overlay methods address the majority of the DMW locations in the primary piping system, there are locations that cannot be treated by these approaches due to the weld geometry or access limitations for the needed equipment. Additionally the dissimilar metal J-groove welds in the reactor pressure vessel head penetration nozzles (RPVHPN) could not be addressed at all by the approaches developed for DMW locations. To address the industry need to mitigate the unfavorable DMW geometries and locations along with the RPVHPN locations, the use of surface stress improvement (SSI) was studied and documented in EPRI reports Materials Reliability Program (MRP)-267 [2], “Technical Basis for Primary Water Stress Corrosion Cracking by Surface Stress Improvement,” and MRP-335 [3], “Topical Report for Primary Water Stress Corrosion Cracking by Surface Stress Improvement.” These reports formed the technical basis for the SSI-related changes made in Code Cases N-770-4 and N-729-5. Along with the technical bases noted, support from the international community in terms of operational experience with SSI in their power plants was invaluable in providing the necessary understanding, context, and confidence to committee members. The ASME “Task Group High Strength Nickel Alloy Issues” (TGHSNAI) was assigned the task of revising the existing Code Cases, N-770 [1], “Alternate Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities” and N-729 [4], “Alternate Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds.” To incorporate the SSI approach into these Code Cases, the first action was to determine whether the SSI process was considered to be a peening process as defined by ASME Section III NB-4422 criteria. This required the submittal of an Interpretation of NB-4422 to determine if SSI techniques were considered a peening process under ASME Section III. The interpretation (Interpretation III-1-13-03), documented in ASME File 12-1192 [5], specified that SSI was not considered peening by Section III. This interpretation provided the framework by which SSI could be directly applied to ASME Section XI inspection criteria without the need to first revise ASME Section III NB-4422. SSI (peening) was first incorporated into Code Case N-770 [1] to provide a mitigation alternative for locations unable to be addressed by the methods addressed thus far. The revision to Code Case N-770 [1] does not provide guidance for the application of SSI activities but rather, it provides the process performance criteria and the inspection guidance following the application of SSI and establishes the pre-application inspection acceptance criteria. Following the approval of SSI in Code Case N-770 [1] addressing the DMW in the primary coolant piping system, the SSI approach was applied to the partial penetration dissimilar metal J-groove welds in RPVHPNs in Code Case N-729 [4]. The application to RPVHPNs provides the industry with a valuable asset preservation tool while significantly lowering the safety risks associated with primary water stress corrosion cracking (PWSCC) and degradation from borated water leakage for the RPVHPNs.


Author(s):  
Naoki Chigusa ◽  
Shinro Hirano ◽  
Takehiko Sera ◽  
Hitoshi Kaguchi ◽  
Masayuki Mukai ◽  
...  

Several Japanese PWR power plants have experienced Primary Water Stress Corrosion Cracking (PWSCC) on dissimilar weld joints since 2004. J weld of 3 Reactor Vessel Head Penetration in Ohi unit 3 is one of the PWSCC incidents occurred in 2004 and has been studied by sampling and opening the fracture surface after its repair. Including Ohi unit 3 Reactor Vessel Head Penetration repair, Japanese PWR utilities and MHI have been developing the preventive maintenance and repair technologies applicable to alloy 600 welds and base metal, following PWSCC events on the Bugey-3 and V.C. Summer. This paper describes recent Japanese PWSCC incidents and repair technologies developed in Japan.


Author(s):  
Doug Killian ◽  
Samer Mahmoud ◽  
Heqin Xu ◽  
Silvester Noronha ◽  
Ashok Nana

The potential for primary water stress corrosion cracking (PWSCC) of large diameter austenitic nickel alloy components and their associated welds presents a particular problem for the nuclear industry due to a limited number of available options for mitigating or repairing large bore pressure boundary components such as reactor vessel, reactor coolant pump, and steam generator inlet or outlet nozzles. While a full structural weld overlay (FSWOL), as governed by ASME Code Case N-740, is commonly used to mitigate and repair small (4″) to medium (10″) bore piping assemblies employing Alloy 82/182 dissimilar metal welds, the large amount of weld metal that would be have to be deposited on large components (and the associated impact on outage schedule) makes this an unattractive strategy for managing the degradation of Alloy 600 type materials. An alternative design option, specifically developed for the mitigation and repair of large bore (30″) components, utilizes a thinner weld overlay whose thickness has been optimized to achieve a specific level of stress on the inside surface of the PWSCC susceptible material. According to ASME Code Case N-754, inside surface stresses should be limited to 10 ksi during the design phase of an optimized weld overlay (OWOL) in order to minimize the initiation or consequences of primary water stress corrosion cracking. With the increased inspection requirements of Code Case N-754 and the corresponding smaller crack growth design flaw size, and along with the reduced weld volume of an OWOL, as compared to a FSWOL, an optimized weld overlay is often the preferred technique for mitigating or repairing large bore piping components. This paper investigates the influence of various parameters on the effectiveness of an optimized weld overlay in satisfying its principle design objective, to reduce the inside surface stresses in PWSCC susceptible materials to no more than 10 ksi. Inherent design parameters are the thickness of the underlying pipe or weld, and the depth of any recorded or postulated weld repairs in the pre-overlay configuration of the welded joint. Explicit design parameters include the thickness of the overlay, the number of weld layers used to form the overlay, and the length of the overlay. Finite element analysis is used to calculate residual and operating stresses in a representative large bore reactor vessel coolant nozzle dissimilar metal weld for various combinations of design parameters. The overall objective of this study is to identify the key parameters influencing inside surface stresses, and thereby provide screening criteria for use in determining the applicability of the optimized weld overlay as a viable PWSCC mitigation or repair option for large bore primary pressure boundary components.


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.


Author(s):  
Kenichi Takakura ◽  
Kiyotomo Nakata ◽  
Noboru Kubo ◽  
Koji Fujimoto ◽  
Kimihisa Sakima

Irradiation Assisted Stress Corrosion Cracking (IASCC) is a matter of great concern as degradation of core internal components in light water nuclear reactor. To clarify the IASCC initiation conditions of baffle former bolt (BFB), constant load stress corrosion cracking (SCC) tests were carried out in simulated PWR primary water (290, 320, 340°C) using C-ring type specimens. Based on the SCC test results, IASCC initiation time becomes shorter with increasing fluence and increasing applied stress, IASCC initiation threshold stress becomes lower with increasing fluence. A test temperature effect was observed in SCC initiation time, but it was not clear the effect of test temperature for SCC initiation threshold stress. These results suggest that IASCC initiation threshold criteria can be described with stress in specimen and fluence. This paper describes the whole evaluation procedure to secure structural integrity of irradiated baffle structure in PWR primary environments, including the threshold stress diagram of IASCC initiation and the irradiation creep formula.


2013 ◽  
Vol 135 (3) ◽  
Author(s):  
Chi Bum Bahn ◽  
Sasan Bakhtiari ◽  
Jangyul Park ◽  
Saurin Majumdar

To detect degradation in steam generator (SG) tubes, periodic inspection using nondestructive examination techniques, such as an eddy current testing, is a common practice. Therefore, it is critical to evaluate and validate the reliability of the eddy current technique for ensuring the structural integrity of the SG tubes. The eddy current technique could be evaluated by comparing the data estimated by the eddy current with the destructive examination data of field cracks, which would be both costly and labor intensive. A viable alternative to pulled tube data is to manufacture crack specimens that closely represent actual field cracks in laboratory environments. A crack manufacturing method that can be conducted at room temperature and atmospheric pressure conditions is proposed. The method was applied to manufacture different types of stress corrosion cracking (SCC) specimens: axial outer-diameter (OD) SCC for straight tubes, circumferential ODSCC and primary water SCC (PWSCC) at hydraulic expansion transition regions, and axial PWSCC at the apex and tangential regions of U-bend tubes. To help the growth of SCC into the tube, corrosive chemicals (sodium tetrathionate) and tensile stress were applied. Eddy current and destructive examination data for SCC specimens were compared with the available field crack data to determine whether those SCC specimens are representative. It was determined that the proposed method could manufacture the representative crack specimens.


Author(s):  
Akihiro Mano ◽  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Yinsheng Li

Structural integrity assessments for cracked nuclear components are currently performed on the basis of deterministic fracture mechanics in accordance with the Rules on Fitness-for-Service for Nuclear Power Plant of the Japan Society of Mechanical Engineers. On the other hand, probabilistic fracture mechanics (PFM) is expected as a rational method for a structural integrity assessment because it can account for the uncertainties and scatters of various influencing factors and can evaluate quantitative values such as the failure probabilities of the components as the solutions. In the Japan Atomic Energy Agency (JAEA), a PFM analysis code PASCAL-SP was developed in order to evaluate the failure probability of nuclear pipe by taking into account aging degradation mechanisms such as inter-granular stress corrosion cracking (IGSCC) and fatigue in the boiling water reactor (BWR) environment. Recently, a number of cracks due to primary water stress corrosion cracking (PWSCC) have been detected in nickel-based alloy welds in the primary piping of pressurized water reactors (PWRs). Therefore, the structural integrity assessment for primary piping taking PWSCC into consideration has become important. This paper details the improvement of PASCAL-SP to evaluate the failure probability of primary pipe taking PWSCC into consideration. We introduce several probabilistic evaluation models such as crack initiation, crack growth and crack detection models related to PWSCC into PASCAL-SP. As numerical examples, the failure probabilities for circumferential and axial cracks in nickel-based alloy weld in pipe in the PWR primary water environment are calculated. We also evaluate the influence of non-destructive inspection on failure probabilities. On the basis of the numerical results, we conclude that the improved PASCAL-SP is useful for evaluating the failure probability of primary pipe taking PWSCC into account.


Author(s):  
Terry Dickson ◽  
Eric Focht ◽  
Mark Kirk

The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned normal reactor startup (heat-up) and shut-down (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Code. The technical basis for these regulations are now recognized by the technical community as being conservative and some plants are finding it increasingly difficult to comply with the current regulations. Consequently, the nuclear industry has developed, and submitted to the ASME Code for approval, an alternative risk-informed methodology that reduces the conservatism and is consistent with the methods previously used to develop a risk-informed revision to the regulations for accidental transients such as pressurized thermal shock (PTS). The objective of the alternative methodology is to provide a relaxation to the current regulations which will provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials, while continuing to provide reasonable assurance of adequate protection to public health and safety. The NRC and its contractor at Oak Ridge National Laboratory (ORNL) have recently performed an independent review of the industry proposed methodology. The NRC / ORNL review consisted of performing probabilistic fracture mechanics (PFM) analyses for a matrix of cool-down and heat-up rates, permutated over various reactor geometries and characteristics, each at multiple levels of embrittlement, including 60 effective full power years (EFPY) and beyond, for various postulated flaw characterizations. The objective of this review is to quantify the risk of a reactor vessel experiencing non-ductile fracture, and possible subsequent failure, over a wide range of normal transient conditions, when the maximum allowable thermal-hydraulic boundary conditions, derived from both the current ASME code and the industry proposed methodology, are imposed on the inner surface of the reactor vessel. This paper discusses the results of the NRC/ORNL review of the industry proposal including the matrices of PFM analyses, results, insights, and conclusions derived from these analyses.


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