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.

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.

Technical Basis for Code Case N-847: Excavate and Weld Repair (EWR) for SCC Mitigation

2016 ◽  
Author(s):  
Steven McCracken ◽  
Jonathan Tatman ◽  
Pete Riccardella
Keyword(s):  
Nuclear Power ◽  
Outer Part ◽  
Nuclear Power Reactor ◽  
Preparation Time ◽  
Weld Repair ◽  
Butt Weld ◽  
Weld Overlay ◽  
New Strategy ◽  
Asme Code ◽  
Technical Basis

Stress corrosion cracking (SCC), though infrequent, is often detected in nuclear power reactor system piping and components. A number of approaches have been developed and successfully deployed for SCC repair and mitigation such as full structural weld overlay (FSWOL), optimized weld overlay (OWOL), and mechanical stress improvement process (MSIP). While these approaches are proven technologies and have served the industry well, a new strategy, excavate and weld repair (EWR), provides yet another option for repair or mitigation of SCC. The EWR approach excavates a portion of the outer part of the butt weld. The excavation is then filled with a weld metal with demonstrated SCC resistance. The EWR approach would require less welding compared to a weld overlay and may be the best option for large bore butt welds where restricted access may make FSWOL, OWOL, or MSIP impractical. For the situation where a flaw is detected and removal or reduction of the flaw to acceptable size is necessary for continued service, the approach would permit a local partial arc EWR where only a portion of the butt weld circumference is removed and repaired. While the partial arc EWR is not a full mitigation, it would provide the needed preparation time for a more permanent repair during a subsequent refueling or maintenance outage. ASME Code Case N-847 was developed to provide examination, design, installation, and preservice/inservice inspection requirements for the EWR repair and mitigation approach. This paper provides a background, description and the technical bases for the EWR case case.

3D Residual Stress Simulation of an Excavate and Weld Repair Mockup

2016 ◽  
Author(s):  
Francis H. Ku ◽  
Pete C. Riccardella ◽  
Steven L. McCracken
Keyword(s):  
Residual Stresses ◽  
Weld Metal ◽  
Nuclear Power ◽  
Three Dimensional ◽  
Weld Repair ◽  
Element Analysis ◽  
Pressurized Water ◽  
Piping Systems ◽  
Asme Code ◽  
Original Construction

This paper presents predictions of weld residual stresses in a mockup with a partial arc excavate and weld repair (EWR) utilizing finite element analysis (FEA). The partial arc EWR is a mitigation option to address stress corrosion cracking (SCC) in nuclear power plant piping systems. The mockup is a dissimilar metal weld (DMW) consisting of an SA-508 Class 3 low alloy steel forging buttered with Alloy 182 welded to a Type 316L stainless steel plate with Alloy 82/182 weld metal. This material configuration represents a typical DMW of original construction in a pressurized water reactor (PWR). After simulating the original construction piping joint, the outer half of the DMW is excavated and repaired with Alloy 52M weld metal to simulate a partial arc EWR. The FEA performed simulates the EWR weld bead sequence and applies three-dimensional (3D) modeling to evaluate the weld residual stresses. Bi-directional weld residual stresses are also assessed for impacts on the original construction DMW. The FEA predicted residual stresses follow expected trends and compare favorably to the results of experimental measurements performed on the mockup. The 3D FEA process presented herein represents a validated method to evaluate weld residual stresses as required by ASME Code Case N-847 for implementing a partial arc EWR, which is currently being considered via letter ballot at ASME BPV Standards Committee XI.

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.

Research on Acoustic Emission In-Service Inspection for Large Above-Ground Storage Tank Floors

2010 ◽  
Author(s):  
Mingchun Lin ◽  
Yewei Kang ◽  
Weibin Wang ◽  
Lei Zhang ◽  
Yi Sun
Keyword(s):  
Acoustic Emission ◽  
Storage Tank ◽  
Magnetic Flux Leakage ◽  
Analysis Method ◽  
Inspection Method ◽  
Oil Tank ◽  
Flux Leakage ◽  
Service Inspection ◽  
Service Field ◽  
Corrosion Data

Much manpower is needed and a lot of materials are wasted when the floor of large above-ground storage tank (AST) is inspected with conventional methods which need to shut down the tank, then to empty and clean it before inspection. Due to the disadvantages of that, an in-service inspection method using acoustic emission (AE) technology is presented. By this mean the rational inspection plan and integrity evaluation of tank floors can be constructed. First, specific inspection steps are established based on the acoustic emission principle for large AST’s floors and the practical condition of AST in order to acquire the AE corrosion data. Second, analysis method of acoustic emission dataset is studied. Finally, maintenance proposes are provided based on results of analysis for the corrosion status of the tank floors. In order to evaluate the performance of our method, an in-service field inspection is practiced on product oil tank with a volume of 5000 cubic meters. Then a traditional inspection procedure using magnetic flux leakage (MFL) technology is followed up. Comparative analysis of the results of the two inspection methods shows that there is consistency in localizing the position of corrosion between them. The feasibility of inservice inspection of AST’s floors with AE is demonstrated.

2004 Edition of Japanese Fitness-for-Service Code for Nuclear Power Plants

2006 ◽  
Author(s):  
Koichi Kashima ◽  
Tomonori Nomura ◽  
Koji Koyama
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Industrial Safety ◽  
National Rule ◽  
Class 1 ◽  
Asme Code ◽  
Service Inspection ◽  
Core Shroud

JSME (Japan Society of Mechanical Engineers) published the first edition of a FFS (Fitness-for-Service) Code for nuclear power plants in May 2000, which provided rules on flaw evaluation for Class 1 pressure vessels and piping, referring to the ASME Code Section XI. The second edition of the FFS Code was published in October 2002, to include rules on in-service inspection. Individual inspection rules were prescribed for specific structures, such as the Core Shroud and Shroud Support for BWR plants, in consideration of aging degradation by Stress Corrosion Cracking (SCC). Furthermore, JSME established the third edition of the FFS Code in December 2004, which was published in April 2005, and it included requirements on repair and replacement methods and extended the scope of specific inspection rules for structures other than the BWR Core Shroud and Shroud Support. Along with the efforts of the JSME on the development of the FFS Code, Nuclear and Industrial Safety Agency, the Japanese regulatory agency approved and endorsed the 2000 and 2002 editions of the FFS Code as the national rule, which has been in effect since October 2003. The endorsement for the 2004 edition of the FFS Code is now in the review process.

ASME Code-Case Proposal to Explicitly Quantify the Interaction Between the PWR Environment and Component Surface Finish

2017 ◽  
Author(s):  
Thomas Métais ◽  
Stéphan Courtin ◽  
Laurent De Baglion ◽  
Cédric Gourdin ◽  
Jean-Christophe Le Roux
Keyword(s):  
Fatigue Life ◽  
Surface Finish ◽  
Polished Surface ◽  
Design Curve ◽  
Fatigue Design ◽  
Wide Range ◽  
Asme Code ◽  
The Difference ◽  
The Uk ◽  
Technical Basis

Fatigue rules from ASME have undergone a significant change over the past decade, especially with the inclusion of the effects of BWR and PWR environments on the fatigue life of components. The incorporation of the environmental effects into the calculations is performed via an environmental factor, Fen, which is introduced in ASME BPV code-case N-792 [5], and depends on factors such as the temperature, dissolved oxygen and strain rate. Nevertheless, a wide range of factors, such as surface finish, have a deleterious impact on fatigue life, but their contribution to fatigue life is typically taken through the transition factors to build the fatigue design curve [2] and not in an explicit way, such as the Fen factor. The testing supporting the rules pertaining to Environmental Fatigue Correction Factor (Fen) Method in ASME BPV was performed on specimens with a polished surface finish and on the basis that the Fen factor was applicable without alteration of the historical practice of building the design curve through transition factors. The extensive amount of testing conducted and reported in References [2] and [7] (technical basis for ASME BPV current EAF rules) was used to propose a set of transition coefficients from the mean air curve to the design curve on one hand, and on the other hand to build a Fen factor expression, defined as the difference between the life in air and in PWR environments. The work initiated by AREVA in 2005 [9] [10] [11] demonstrated that there is a clear interaction between the two aggravating effects of surface finish and PWR environment for fatigue damage, which was not experimentally tested in the References [2] and [7]. These results have clearly been supported by testing carried out independently in the UK by Rolls-Royce and AMEC FW [12]. These results are all the more relevant as most NPP components do not have a polished surface finish. Most surfaces are either industrially polished or installed as-manufactured. It was concluded that this proposal could potentially be applicable to a wide range of components and could be of interest to a wider community. EDF/Areva/CEA have therefore authored a code-case introducing the Fen-threshold, a factor which explicitly quantifies the interaction between PWR environment and surface finish. This paper summarizes this proposal and provides the technical background and experimental work to justify this proposal.

Technical Basis for Flaw Acceptance Criteria for Cast Austenitic Stainless Steel Piping

2020 ◽  
Vol 142 (2) ◽  
Author(s):  
Do Jun Shim ◽  
Nathanial Cofie ◽  
Dilipkumar Dedhia ◽  
Tim Griesbach ◽  
Kyle Amberge
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Evaluation Procedure ◽  
Correction Factors ◽  
Acceptance Criteria ◽  
Regulatory Guidance ◽  
Asme Code ◽  
Tensile Data ◽  
Equivalent Factor ◽  
Technical Basis

Abstract According to the current ASME Code Section XI, IWB-3640 and Appendix C flaw evaluation procedure, cast austenitic stainless steel (CASS) piping with ferrite content (FC) less than 20% is treated as wrought stainless steel. For CASS piping with FC equal or greater than 20%, there was no flaw evaluation procedure in the ASME Code prior to the 2019 Edition. In this paper, the technical basis for the recently approved Code change containing flaw acceptance criteria for CASS piping is presented. The procedure utilizes the current rules in ASME Code Section XI, IWB 3640/Appendix C and the existing elastic-plastic correction factors (i.e., Z-factors) for other materials in the Code. The appropriate Z-factor to use for the CASS piping is determined based on the FC (using Hull's equivalent factor). Experimentally measured fully saturated fracture toughness and tensile data of the three most common grades of CASS material in the U.S. (CF3, CF8, and CF8M) were used to determine the flaw acceptance criteria in the Appendix C Code method. As described here, the method is conservative since it utilizes the fully saturated condition of CASS materials. In addition, it is simple and consistent with the current regulatory guidance on aging management of CASS piping.

Inspection and Repair Techniques and Strategies for Alloy 600 PWSCC in Reactor Vessel Head CRD Nozzles and Welds

2002 ◽  
Author(s):  
S. W. Glass ◽  
D. M. Schlader
Keyword(s):  
High Velocity ◽  
Decision Process ◽  
Safety Concern ◽  
Reactor Vessel ◽  
Alloy 600 ◽  
Seal Design ◽  
Weld Area ◽  
Repair Techniques

As a result of the Alloy 600 PWSCC CRD nozzle leaks discovered in the fall of 2000 and spring of 2001 in several US plants, the NRC has recommended a more pro-active effort by U.S. utilities to inspect similarly susceptible nozzles in all US plants. The primary safety concern is circumferential cracks that can permit the nozzle to separate from the head at high velocity and produce a large-break leak in the reactor vessel. A secondary concern is head leakage from any through-wall cracks in the nozzle or J-groove weld area. Although the fundamental weld and seal design are similar for all US PWR plants, the various surrounding geometry and repair probability considerations require multiple inspection and repair alternatives. Geometry issues include the head insulation design that influences the ability to perform visual examinations from above the head, and the presence or absence of thermal sleeves and funnels governing the type of NDE probes than can be used. Repair probability considerations primarily include the likelihood for repair of a small or large number of nozzles and the length of time the repair must last before a head replacement. This paper discusses the various inspection and repair alternatives offered by one service vendor and discusses a decision process for planning the inspection and repair effort.

Technical Basis for Structural Integrity Performance Criterion for Steam Generator Tubing

2006 ◽  
Author(s):  
Russell C. Cipolla ◽  
James A. Begley ◽  
Robert F. Keating
Keyword(s):  
Steam Generator ◽  
Nuclear Energy ◽  
Structural Integrity ◽  
Performance Criterion ◽  
Major Fraction ◽  
Reactor Coolant ◽  
Steam Generator Tubing ◽  
Accident Conditions ◽  
Technical Basis ◽  
Repair Techniques

General Design Criteria (GDC) 1, 2, 4, 14, 30, 31 and 32 of 10 CFR Part 50, Appendix A, define requirements for the reactor coolant pressure boundary (RCPB) with respect to structural and leakage integrity [1]. Steam generator tubing and tube repairs constitute a major fraction of the RCPB surface area. Steam generator tubing and associated repair techniques and components, such as sleeves, must be able to maintain reactor coolant inventory and pressure. The Structural Integrity Performance Criterion (SIPC) from Nuclear Energy Institute (NEI) 97-06 [2] was developed to provide reasonable assurance that a steam generator tube will not burst during normal or postulated accident conditions. This paper presents the SIPC and its technical basis.

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