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.

Effect of Size of Butt Weld Repairs on Weld Overlay Residual Stresses

2007 ◽  
Author(s):  
Carl R. Limpus ◽  
David G. Dijamco ◽  
Richard Bax ◽  
Nathaniel G. Cofie
Keyword(s):  
Residual Stresses ◽  
Nuclear Power ◽  
Analytical Modeling ◽  
Weld Repair ◽  
Butt Weld ◽  
Weld Overlay ◽  
Compressive Stresses ◽  
Element Simulation ◽  
Weld Overlays ◽  
Compressive Residual Stresses

Weld overlays have been used to provide repair and mitigation to stress corrosion cracking (SCC) susceptible butt welds in nuclear power plant piping. Among the several advantages associated with weld overlays are the beneficial compressive residual stresses that are developed in the inner portion of the component after application of the overlay. These compressive stresses can provide significant mitigation against SCC in these welds. To determine the residual stresses resulting from the weld overlay process in analytical modeling, a weld repair during original fabrication of the butt weld is typically assumed before application of the weld overlay. If the fabrication records are available, the details of the weld repair can be simulated in the analysis. However, in most cases, the weld records are not easily accessible and in instances where they are available, the quality and completeness of the information are questionable. As such, various conservative assumptions are made on the extent of the weld repair to be simulated in the analytical modeling. In this paper, the residual stress results of an axisymmetric finite element simulation of a bimetallic weld subjected to an inside surface weld repair followed by a weld overlay repair are presented. Three through-wall weld repair sizes (25%, 50% and 75% of the wall thickness without the overlay) assumed to be full 360° around the circumference were considered in the study. The results indicate that for all three weld repair cases, the inside of the configuration is very tensile after the weld repair indicating that regardless of the size of the weld repair, SCC is a possibility. The post weld repair stress distribution of the 50% and the 75% repair cases are similar indicating that an assumed 50% repair is fairly representative of repairs that can be assumed for analysis purposes. The application of the overlay resulted in favorable compressive stresses on the inside portion of the configuration for all the three weld repair cases indicating that regardless of the size of the initial weld repair, the application of the weld overlay provides mitigation against SCC.

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.

Technical Basis for Cases N-629 and N-631 as an Alternative for RTNDT Reference Temperature

2007 ◽  
Author(s):  
J. G. Merkle ◽  
K. K. Yoon ◽  
W. A. VanDerSluys ◽  
W. Server
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Threshold Level ◽  
Nuclear Regulatory Commission ◽  
New Approach ◽  
The Us ◽  
Asme Code ◽  
Regulatory Commission ◽  
Technical Basis

ASME Code Cases N-629/N-631, published in 1999, provided an important new approach to allow material specific, measured fracture toughness curves for ferritic steels in the code applications. This has enabled some of the nuclear power plants whose reactor pressure vessel materials reached a certain threshold level based on overly conservative rules to use an alternative RTNDT to justify continued operation of their plants. These code cases have been approved by the US Nuclear Regulatory Commission and these have been proposed to be codified in Appendix A and Appendix G of the ASME Boiler and Pressure Vessel Code. This paper summarizes the basis of this approach for the record.

Prediction of Plastic Collapse Moments for Circumferential Cracked Pipes With Weld Overlays

2009 ◽  
Author(s):  
Kunio Hasegawa ◽  
Yinsheng Li ◽  
Gery M. Wilkowski ◽  
Arthur F. Deardorff
Keyword(s):  
Experimental Data ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Method ◽  
Plastic Collapse ◽  
Analytical Evaluation ◽  
Weld Overlay ◽  
Asme Code ◽  
Weld Overlays

Weld overlay (WOL) is one of the useful repair methods for cracked piping that has been successfully applied for piping in many nuclear power plants. In addition, ASME Boiler and Pressure Vessel Code Section XI provides a WOL method in Non-mandatory Appendix Q and a number of Code Cases. Currently, an analytical evaluation method for predicting failure stresses for WOL piping is under discussion in a working group of ASME Code Section XI. This paper proposes an approach for predicting the plastic collapse moment for WOL piping using a net-section collapse stress approach. In addition, the predicted collapse moments are compared with experimental data.

Technical Justification for Increasing Temperbead Welding Area Limits on Low Alloy Steel

2007 ◽  
Author(s):  
Marcos L. Herrera ◽  
Shu S. Tang ◽  
Artie Peterson
Keyword(s):  
Ambient Temperature ◽  
Alloy Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Low Alloy Steel ◽  
Weld Overlay ◽  
Gas Tungsten Arc ◽  
Asme Code ◽  
Area Limit ◽  
Inspection Techniques

This paper presents the results of the analytical evaluation supporting the technical justification of increasing the amount of temperbead welding, currently limited to an area of 100 in2, that can be performed on low alloy steel (LAS) nuclear power plant components. The need to expand the application area limitations is increasing for ambient temperature Gas Tungsten Arc Weld (GTAW) temperbead weld overlay repairs on LAS components. As nuclear power plants age and as inspection techniques continue to improve increasing the area limit becomes increasingly important since more indications are being identified. Existing limitations of temperbead welding area of 100 in2 imposed in the ASME Code and in Code Cases 606 and 638 for ambient temperature temper bead welding are arbitrary and overly conservative. This paper presents the analyses supporting: 1) a weld overlay repair greater than 100 in2 on a Reactor Pressure Vessel (RPV) nozzle and 2) a weld cavity repair on an RPV of 500 in2 vertical shell weld. Based on the results of these cases, conclusions regarding temperbead welding in excess of the current 100 in2 limit are made.

Post Weld Overlay Residual Stresses in Dissimilar Metal Welds Considering Various Butt Weld and Weld Repair Fabrication Sequences

2009 ◽  
Author(s):  
Francis H. Ku ◽  
Christopher S. Lohse ◽  
David G. Dijamco ◽  
Charles J. Fourcade ◽  
Richard L. Bax ◽  
...  
Keyword(s):  
Residual Stresses ◽  
Welding Process ◽  
Reasonable Assumption ◽  
Weld Repair ◽  
Pressurized Water ◽  
Butt Weld ◽  
Weld Overlay ◽  
Water Reactor ◽  
Dissimilar Metal ◽  
Pipe Size

Weld overlays have been used to repair or mitigate stress corrosion cracking (SCC) in both boiling water reactor (BWR) and pressurized water reactor (PWR) nozzle-to-pipe dissimilar metal welds (DMW). One of the contributing factors to SCC is the high tensile residual stresses produced during the fabrication of the original butt weld, especially when local weld repairs were present during the welding process. In analytical simulations to determine the post weld overlay residual stresses, complete simulation of the original butt weld, weld repair and the overlay is desired. However, to reduce the computational effort, it is commonly assumed that the weld repair stresses overwhelm the original butt weld residual stresses such that the original butt weld need not be simulated and only the weld repair is simulated before the application of the overlay. Questions have also been raised as to why the butt weld and/or the weld repair need to be simulated since it is assumed that both of these fabrication processes would be overcome by the weld overlay process. This paper investigates three fabrication sequences in order to determine their effect on the post weld overlay residual stresses: (1) the butt weld is simulated followed by a weld repair and then the weld overlay is applied; (2) the butt weld is simulated followed by the weld overlay with no consideration of a weld repair; (3) the butt weld is not simulated but a weld repair is assumed and the weld overlay is applied. Five different nozzle-to-pipe size configurations were used in the study to determine the effect of pipe size on the three fabrication sequences described above. The investigation indicates that the post weld overlay residual stresses for Cases 1 and 3 are similar and hence simulation of the weld repair alone (without the butt weld simulation) prior to simulating the weld overlay is a reasonable assumption. However, not simulating the weld repair (corresponding to Case 2) may provide different residual stress distribution.

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.

A Technical Basis for Characterizing Flaws Detected by Preservice and Inservice Examinations of Nuclear Power Plant Components

1975 ◽  
Vol 97 (4) ◽  
pp. 322-326 ◽  
Author(s):  
R. R. Maccary
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Flaw Detection ◽  
Nondestructive Examination ◽  
Asme Code ◽  
Underlying Basis ◽  
Technical Basis ◽  
Plant Components

The nondestructive examination procedures specified by the rules of construction of the ASME Boiler and Pressure Vessel Code—Section III, “Nuclear Power Plant Components” require techniques whose flaw detection capabilities are well within the practical limits established for acceptable workmanship and quality of fabrication. The rules of the ASME Section XI, “Inservice Inspection of Nuclear Reactor Coolant Systems”, impose an additional series of examinations. Material or fabrication flaws detected during a preservice examination as well as flaws developed during service must be evaluated to establish the acceptability of the component for initial and continued service. These examination requirements have introduced the need to characterize the flaws detected by the examinations and to set “allowable flaw indication standards.” The principles of fracture mechanics provide an engineering tool which predicts the behavior of materials containing flaws under service loadings. These principles form the underlying basis upon which the allowable flaw indication standards of ASME Section XI were formulated. The development of new rules governing flaw indication characterization and allowable flaw indications standards, as specified in the ASME Code, Section XI, are reviewed.

SANICRO 71

Alloy Digest ◽  
1965 ◽  
Vol 14 (12) ◽  
Keyword(s):  
Nuclear Power ◽  
Elevated Temperatures ◽  
Power Reactor ◽  
Base Alloy ◽  
Heat Treating ◽  
Good Resistance ◽  
Nuclear Power Reactor ◽  
Temperature Performance ◽  
Resistance To Stress ◽  
Nickel Base

Abstract Sanicro 71 is a nickel-base alloy having good resistance to stress-corrosion, oxidation and creep at elevated temperatures. It is recommended for nuclear power reactor heat exchanger tubes, aircraft turbojet engines and for equipment in the textile, plastic, and chemical industries. This datasheet provides information on composition, physical properties, hardness, elasticity, and tensile properties. It also includes information on high temperature performance and corrosion resistance as well as forming, heat treating, machining, joining, and surface treatment. Filing Code: Ni-108. Producer or source: Sandvik.

G-30 FILLER METAL

Alloy Digest ◽  
1993 ◽  
Vol 42 (5) ◽  
Keyword(s):  
Fracture Toughness ◽  
Corrosion Resistance ◽  
Physical Properties ◽  
Tensile Properties ◽  
Filler Metal ◽  
Weld Repair ◽  
Weld Overlay ◽  
High Chromium ◽  
Cast Products

Abstract Hastelloy Alloy G-30 filler metal is used as matching composition filler metal for fabrication of Hastelloy G-30 wrought and cast products and as filler metal for fabrication of G/G-3 alloy wrought products. It is also used for weld repair of high chromium castings and for weld overlay cladding. This datasheet provides information on composition, physical properties, elasticity, and tensile properties as well as fracture toughness. It also includes information on corrosion resistance as well as joining. Filing Code: Ni-432. Producer or source: Haynes International Inc.

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