Deterministic Technical Basis for Re-Examination Interval of Every Second Refueling Outage for PWR Reactor Vessel Heads Operating at Tcold With Previously Detected PWSCC

2016 ◽  
Author(s):  
Glenn White ◽  
Kevin Fuhr ◽  
Markus Burkardt ◽  
Craig Harrington
Keyword(s):  
Operating Experience ◽  
Alloy 600 ◽  
Head Penetration ◽  
Inspection Interval ◽  
Asme Code ◽  
Surface Examination ◽  
Technical Basis ◽  
Growth Analyses ◽  
Reactor Pressure ◽  
The U.S

Plant operating experience with Alloy 600 reactor pressure vessel top head penetration nozzles in U.S. PWRs shows that the inspection intervals prescribed by ASME Code Case N-729-1 have been successful in managing the PWSCC concern. No through-wall cracking has been observed in the U.S. after the first in-service volumetric or surface examination was performed on all CRDM or CEDM nozzles in a given head. The current inspection intervals have facilitated identification of any PWSCC in its early stages, with small numbers of nozzles affected and substantial margins to leakage at the five affected heads operating at Tcold. MRP-395 demonstrated through both deterministic and probabilistic analyses that the inspection intervals of Code Case N-729-1 remain valid to conservatively address the PWSCC concern. This paper supplements MRP-395 with additional deterministic crack growth analyses coupled with assessments of the PWSCC indications detected in heads operating at Tcold. The supplemental deterministic assessments presented in this paper demonstrate the acceptability of a 36-month volumetric or surface inspection interval for heads with previously detected PWSCC and that operate at Tcold. Until Code Case N-729-5 is approved by U.S. NRC, use of the 36-month interval in the U.S. for such heads would require review and approval by U.S. NRC of a relief request submitted by the licensee.

Effects of Cladding on Reactor Pressure Vessel Integrity Based on the Revised PTS Rule (10 CFR 50.61)

2009 ◽  
Author(s):  
Vikram Marthandam ◽  
Timothy J. Griesbach ◽  
Jack Spanner
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Historical Perspective ◽  
Pressurized Thermal Shock ◽  
Asme Code ◽  
Potential Benefits ◽  
Vessel Integrity ◽  
Technical Basis ◽  
Reactor Pressure

This paper provides a historical perspective of the effects of cladding and the analyses techniques used to evaluate the integrity of an RPV subjected to pressurized thermal shock (PTS) transients. A summary of the specific requirements of the draft revised PTS rule (10 CFR 50.61) and the role of cladding in the evaluation of the RPV integrity under the revised PTS Rule are discussed in detail. The technical basis for the revision of the PTS Rule is based on two main criteria: (1) NDE requirements and (2) Calculation of RTMAX-X and ΔT30. NDE requirements of the Rule include performing volumetric inspections using procedures, equipment and personnel qualified under ASME Section XI, Appendix VIII. The flaw density limits specified in the new Rule are more restrictive than those stipulated by Section XI of the ASME Code. The licensee is required to demonstrate by performing analysis based on the flaw size and density inputs that the through wall cracking frequency does not exceed 1E−6 per reactor year. Based on the understanding of the requirements of the revised PTS Rule, there may be an increase in the effort needed by the utility to meet these requirements. The potential benefits of the Rule for extending vessel life may be very large, but there are also some risks in using the Rule if flaws are detected in or near the cladding. This paper summarizes the potential impacts on operating plants that choose to request relief from existing PTS Rules by implementing the new PTS Rule.

Do All RPV Head Penetration Leaks Have the Potential to Cause Head Wastage?

2004 ◽  
Author(s):  
Chandra M. Roy ◽  
John R. Fessler ◽  
Jude R. Foulds ◽  
Ronald M. Latanision ◽  
David E. Taylor
Keyword(s):  
Water Stress ◽  
Pressure Vessel ◽  
Recent Experience ◽  
Alloy 600 ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Head Penetration ◽  
Failure Risk ◽  
Primary Water ◽  
Reactor Pressure

The identification of the PWSCC (Primary Water Stress Corrosion Cracking) mechanism responsible for leakage from an Alloy 600 nozzle tube of a PWR RPV (pressurized water reactor reactor pressure vessel) head more than a decade ago led to a significant body of research into understanding the phenomenon and to development of bases for safely managing primary pressure boundary integrity. However, the relatively recent experience at Davis-Besse, wherein penetration leakage resulted in significant vessel head material wastage, led to the heretofore unconsidered issue of vessel failure risk due to head rupture. This paper addresses, in preliminary fashion, one key input to determining the risk associated with head material wastage and potential rupture — the local environmental and fluid conditions associated with a range of leak paths. The results indicate a need for rigorous prediction of fluid conditions for a range of leak situations to help establish criteria for addressing penetration leaks.

A Risk-Informed Assessment of ASME Section XI, Appendix E

2011 ◽  
Author(s):  
Ronald Gamble ◽  
William Server ◽  
Bruce Bishop ◽  
Nathan Palm ◽  
Carol Heinecke
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Evaluation Criteria ◽  
Nuclear Regulatory Commission ◽  
Pressurized Thermal Shock ◽  
Asme Code ◽  
American Society ◽  
Regulatory Commission ◽  
Reactor Pressure ◽  
The U.S

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code [1], Section XI, Non mandatory Appendix E, “Evaluation of Unanticipated Operating Events”, provides a deterministic procedure for evaluating reactor pressure vessel (RPV) integrity following an unanticipated event that exceeds the operational limits defined in plant operating procedures. The recently developed risk-informed procedure for Appendix G to Section XI of the ASME Code [2, 3], and the development by the U.S. Nuclear Regulatory Commission (NRC) of the alternate Pressurized Thermal Shock (PTS) rule [4, 5, 6] led to initiation of this study to determine if the Appendix E evaluation criteria are consistent with risk-informed acceptance criteria. The results of the work presented in this paper demonstrate that Appendix E is consistent with risk-informed criteria developed for PTS and Appendix G and ensures that evaluation of RPV integrity following an unanticipated event would not violate material or operational limits recently defined using risk-informed criteria. Currently, Appendix E does not have evaluation criteria for BWR vessels; however, as part of this study, risk-informed analyses were performed for unanticipated heat-up events and isothermal, overpressure events in BWR plant designs.

Cracking in Alloy 600 Penetration Nozzles: A Regulatory Perspective

2004 ◽  
Author(s):  
Allen L. Hiser
Keyword(s):  
Nuclear Regulatory Commission ◽  
Nuclear Industry ◽  
Alloy 600 ◽  
Pressurized Water ◽  
New Findings ◽  
Regulatory Perspective ◽  
Water Reactors ◽  
Regulatory Commission ◽  
Reactor Pressure ◽  
The U.S

Cracking in Alloy 600 penetration nozzles in the reactor pressure vessel upper heads of pressurized water reactors (PWRs) has been an issue in the U.S. nuclear industry for more than 10 years. This paper provides a regulatory perspective on how the U.S. industry and the U. S. Nuclear Regulatory Commission (NRC) have addressed new findings over the last 10 years and responded to ensure safety of U.S. PWRs. This paper provides a summary going back to the early 1990’s and emphasizes activities over the last three years since the discovery of circumferential cracking at Oconee, and provides a perspective regarding on-going NRC activities.

Technical Basis for Code Case N-853: A600 Branch Connection Weld Repair for SCC Mitigation

2016 ◽  
Author(s):  
David Waskey ◽  
Steven McCracken
Keyword(s):  
Alloy 600 ◽  
Weld Repair ◽  
Inspection Method ◽  
Regulatory Issue ◽  
Full Penetration ◽  
Asme Code ◽  
Bounding Analysis ◽  
Technical Basis ◽  
Service Inspection ◽  
Repair Techniques

The NRC Issued a Regulatory Issue Summary (RIS 15-10) which was an inquiry to the ASME Code Committee to review the in-service inspection requirements for Alloy 600 full penetration branch connections. This NRC request resulted in the initiation of two PWROG projects, PA-MSC-1283 and 1294. PA-MSC-1283 is a Fracture Analysis to evaluate the crack growth rate of the various existing A600 configurations, heat treat conditions and operating loads. The results from this study would provide the technical basis for in-service inspection method and frequency. PA-MSC-1294 is a project to consider contingency repair techniques and provide a bounding analysis for all nozzle configurations of the selected repair technique. The selected repair approach to be analyzed was a weld buildup referred as a Branch Connection Weld Metal Buildup (BCWMB) in the ASME Code Case (N-853) developed to provide the rules for design, implementation and inspection. Included in this project was a proof-of-principle phase to demonstrate that the repair methodology is implementable and assist in establishing the potential on component repair duration and anticipated dose.

Reactor Pressure Vessel Modification Approaches

2005 ◽  
Author(s):  
Ronald J. Payne ◽  
Stephen Levesque
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
State Of The Art ◽  
Electrical Power ◽  
Evaluation Criteria ◽  
Reactor Pressure Vessel ◽  
Alloy 600 ◽  
Electrical Power Research Institute ◽  
Reactor Pressure ◽  
Plant Components

Stress corrosion cracking of Alloy 600 has lead to the modification and replacement of many nuclear power plant components. Among these components are the Bottom Mounted Nozzles (BMN) of the Reactor Pressure Vessel (RPV). Modifications of these components have been performed on an emergent basis. Since that time, Framatome ANP has developed state-of-the-art modification methods for the repair of BMNs using the Electrical Power Research Institute (EPRI) managed Materials Reliability Program (MRP) attributes for an ideal repair as a basis for evaluation of modification concepts. These attributes were used to evaluate the optimal modification concepts and develop processes and tooling to support future modification activity. This paper details the BMN configurations, modification evaluation criteria, several modification concepts, and the development of the tooling to support the optimal modification scenarios.

Suggested Improvements to Appendix G of ASME Section XI Code

2008 ◽  
Author(s):  
Hardayal S. Mehta ◽  
Timothy J. Griesbach ◽  
Gary L. Stevens
Keyword(s):  
Thermal Gradient ◽  
Research Council ◽  
Ductile Failure ◽  
Calculation Methods ◽  
Technical Literature ◽  
Corner Crack ◽  
The Past ◽  
Asme Code ◽  
Reactor Pressure ◽  
Calculation Procedures

This paper reviews some of the original basis documents for ASME Section XI Nonmandatory Appendix G for calculating pressure-temperature (P-T) limits and recommends areas for improvement. The original Appendix G in Section XI of ASME Code was mainly based on Welding Research Council (WRC) Bulletin 175 (WRC-175). Changes have been made to Appendix G over the past 20 years such as the use of the KIC reference toughness curve instead of KIR. However, aspects of the Appendix G method still refer back to WRC Bulletin 175. The published technical literature since the development of WRC 175 could be used to enhance the Appendix in a number of areas. One such area is stress intensity factor (K) calculation procedures for thermal gradient loading at a nozzle corner. This paper will review and evaluate the available K calculation methods for a nozzle corner crack, and develop closed-form expressions for incorporation into Appendix G. Also, the following areas will be reviewed: (1) treatment of operating stresses exceeding the material yield stress, and (2) fracture toughness criteria typically used for other than reactor pressure vessel (RPV) and piping for protection against non-ductile failure. This paper will also identify areas for future improvements in Appendix G.

U.S. Operating Experience With Thermally Treated Alloy 600 Steam Generator Tubes

2002 ◽  
Author(s):  
April Smith ◽  
Kenneth J. Karwoski
Keyword(s):  
Steam Generator ◽  
Operating Experience ◽  
Alloy 600 ◽  
Steam Generators ◽  
Pressurized Water ◽  
Steam Generator Tube ◽  
Water Reactors ◽  
1960S And 1970S ◽  
The 1960S ◽  
Thermally Treated

Steam generators placed in service in the 1960s and 1970s were primarily fabricated from mill-annealed Alloy 600. Over time, this material proved to be susceptible to stress corrosion cracking in the highly pure primary and secondary water chemistry environments of pressurized-water reactors. The corrosion ultimately led to the replacement of steam generators at numerous facilities, the first U.S. replacement occurring in 1980. Many of the steam generators placed into service in the 1980s used tubes fabricated from thermally treated Alloy 600. This tube material was thought to be less susceptible to corrosion. Because of the safety significance of steam generator tube integrity, this paper evaluates the operating experience of thermally treated Alloy 600 by looking at the extent to which it is used and recent results from steam generator tube examinations.

Consideration of Environmental Fatigue in the ASME Code for Carbon and Low-Alloy Steel Components

2003 ◽  
Author(s):  
Stan T. Rosinski ◽  
Arthur F. Deardorff ◽  
Robert E. Nickell
Keyword(s):  
Fatigue Life ◽  
Alloy Steel ◽  
Water Environment ◽  
Operating Experience ◽  
Nuclear Industry ◽  
Operating Conditions ◽  
Low Alloy Steel ◽  
Reactor Water ◽  
Asme Code ◽  
The Impact

The potential impact of reactor water environment on reducing the fatigue life of light water reactor (LWR) piping components has been an area of extensive research. While available data suggest a reduction in fatigue life when laboratory samples are tested under simulated reactor water environments, reconciliation of this data with plant operating experience, plant-specific operating conditions, and established ASME Code design processes is necessary before a conclusion can be reached regarding the need for explicit consideration of reactor water environment in component integrity evaluations. U.S. nuclear industry efforts to better understand this issue and ascertain the impact, if any, on existing ASME Code guidance have been performed through the EPRI Materials Reliability Program (MRP). Based on the MRP activities completed to date there is no need for explicit incorporation of reactor water environmental effects for carbon and low-alloy steel components in the ASME Code. This paper summarizes ongoing MRP activities and presents the technical arguments for resolution of the environmental fatigue issue for carbon and low-alloy steel locations.

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