ML20246G707

Enclosure 4: Attachment 4 - WCAP-11164-NP, Rev. 2, Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for North Anna Units 1 and 2 Nuclear Power Plants for the Subsequent License Renewal

ML20246G707
Person / Time
Site:	North Anna
Issue date:	01/31/2020
From:	Johnson E Westinghouse
To:	Office of Nuclear Reactor Regulation
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ML20246G703	List: ML20246G697 ML20246G698 ML20246G700 ML20246G701 ML20246G706 ML20246G707 ML20246G709
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20-115
Download: ML20246G707 (92)

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Serial No.: 20-115 Docket Nos.: 50-338/339 Enclosure4 Attachment 4 WCAP-11164-NP, REVISION 2 (Redacted version of WCAP-11163-P)

Virginia Electric and Power Company

{Dominion Energy Virginia)

North Anna Power Station Units 1 and 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-11164-NP January 2020 Revision 2 Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for North Anna Units 1 and 2 Nuclear Power Plants for the Subsequent License Renewal Program (80 Years) Leak-Before-Break Evaluation

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• • • This record was final approved on 1/17/2020 11 :18:49 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 ii WCAP-11164-NP Revision 2 Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for North Anna Units 1 and 2 Nuclear Power Plants for the Subsequent License Renewal Program (80 Years) Leak-Before-Break Evaluation January 2020 Author: E. D. Johnson*

Reactor Vessel and Containment Vessel Design and Analysis Verifier: M. Wiratmo*

Piping Engineering Approved: L.A. Patterson, Manager*

Reactor Vessel and Containment Vessel Design and Analysis

• Electronically approved records are authenticated in the electronic document management system.

WCAP-11164-NP January 2020 Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii RECORD OF REVISIONS Rev Date Revision Description August 0 Original Issue (WCAP-11164) 1986 October 1 Non-Proprietary Issue of WCAP-11163-P, Revision 1.

2019 January 2 Non-Proprietary Issue of WCAP-11163-P, Revision 2.

2020 WCAP-11164-NP January 2020 Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 iv TABLE OF CONTENTS 1.0 Introduction .............................................................................................................................. 1-1 1.1 Purpose ........................................................................................................................ 1-1 1.2 Background Information ............................................................................................... 1-1 1.3 Scope and Objectives ................ :................................................................................. 1-2 1.4 References ................................................................................................................... 1-3 2.0 Operation .and Stability of the Reactor Coolant System ........................................................... 2-1 2.1 Stress Corrosion Cracking ........................................................................................... 2-1 2.2 Watei Hammer ............................................................................................................. 2-3 2.3 Low Cycle and High Cycle Fatigue ....................... :...................................................... 2-3 2.4 Wall Thinning, Creep, and Cleavage ............................................................................ 2-3 2.5 References ................................................................................................................... 2-4 3.0 Pipe Geometry and Loading .................................................................................................... 3-1 3.1 Introduction to Methodology ......................................................................................... 3-1 3.2 Calculation of Loads and Stresses ............................................................................... 3-2 3.3 Loads for Leak Rate Evaluation ................................................................................... 3-3 3.4 Load Combination for Crack Stability Analyses ............................................................ 3-3 3.5 References ................................................................................................................... 3-4 4.0 Material Characterization ......................................................................................................... 4-1 4.1 Primary Loop Pipe and Fittings Materials ..................................................................... 4-1 4.2 Tensile Properties ......................................................................................................... 4-1 4.3 Fracture Toughness Properties .................................................................................... 4-2 4.3.1 Fracture Toughness Properties of Centrifugal Cast Pipe (CF8A) ..................... 4-3 4.3.2 Fracture Toughness Properties of Static Cast Elbows (CF8M) ........................ 4-4 4.3.3 Consideration of Dissimilar Metal Weld Material Profiles ................................. 4-6 4.4 References ................................................................................................................... 4-7 5.0 Critical Location and Evaluation Criteria .................................................................................. 5-1 5.1 Critical Locations .......................................................................................................... 5-1 5.2 Evaluation Criteria ........................................................................................................ 5-2 5.3 References ................................................................................................................... 5-2 6.0 Leak Rate Predictions .............................................................................................................. 6-1 6.1 Introduction ................................................................................................................... 6-1 WCAP-11164-NP January 2020 Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 V 6.2 General Considerations ................................................................................................ 6-1 6.3 Calculation Method ....................................................................................................... 6-1

-, 6.4 Leak Rate Calculations ................................................................................................ 6-2 6.5 References ................................................................................................................... 6-2 7.0 Fracture Mechanics Evaluation ................................................................................................ 7-1 7.1 Local Failure Mechanism ............................................................................................. 7-1 7.2 Global Failure Mechanism ............................................................................................ 7-2 7.3 SGIN and SGON Alloy 82/182 Welds .................................:****--................................... 7-3 7.4 References ....................................... :........................................................................... 7-4 8.0 Fatigue Crack Growth Analysis ................................................................................................ 8-1 8.1 References ................................................................................................................... 8-3 9.0 Assessment of Margins ............................................................................................................ 9-1 10.0 Conclusions ............................................................................................................................ 10-1 Appendix A: Limit Moment ................................................................................................................A-1 Appendix 8: LBS History from Surry Power Station SLR ................................................................. 8-1 WCAP-11164-NP January 2020 Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 vi LIST OF TABLES Table ES-1 Summary of Analysis Locations, Materials, and Evaluation Types ............................................. x Table 3-1 Dimensions, Normal Loads and Stresses for North Anna Unit 1 ............................................ 3-5 Table 3-2 Dimensions, Normal Loads and Stresses for North Anna Unit 2 .......................................... 3-6 Table 3-3 Faulted Loads and Stresses for North Anna Unit 1 .................................................................. 3-7 Table 3-4 Faulted Loads and Stresses for North Anna Unit 2 ......................................................... 3-8 Table 4-1 Measured Tensile Properties for North Anna Unit 1 Primary Loop Pipes (A351-CF8A) ........... 4-8 Table 4-2 Measured Tensile Properties for North Anna Unit 2 Primary Loop Pipes (A351-CF8A) ........... 4-9 Table 4-3 Measured Tensile Properties for North Anna Unit 1 Primary Loop Elbows (A351-CF8M) ..... 4-1 O Table 4-4 Measured Tensile Properties for North Anna Unit 2 Primary Loop Elbows (A351-CF8M) ..... 4-11 Table 4-5 North Anna Units 1 and 2 Measured Tensile Properties for A351-CF8A and A351-CF8M ..... 4-12 Table 4-6 ASME Code Tensile Properties for Material A351-CF8A and A351-CF8M ............................ 4-13 Table 4-7 Material Properties for Operating Temperature Conditions on North Anna Units 1 and 2 RCL Lines ............................................................................................................................... 4-14 Table 4-8 North Anna Units 1 and 2 CF8A Chemical Composition ........................................................ 4-15 Table 4-9 North Anna Units 1 and 2 CF8A Fracture Toughness and Tearing Modulus Properties ......... 4-16 Table 4-10 North Anna Units 1 and 2 CF8A Pipes with Lowest Fracture Toughness Properties ............. 4-17 Table 4-11 North Anna Units 1 and 2 CF8M Chemical Composition ....................................................... 4-18 Table 4-12 North Anna Units 1 and 2 CF8M Fracture Toughness and Tearing Modulus Properties ........ 4-20 Table 4-13 North Anna Units 1 and 2 CF8M Elbows with Lowest Fracture Toughness Properties .......... 4-22 Table 5-1 Critical Analysis Locations for Leak-Before-Break of North Anna Unit 1 RCL Lines ................ 5-3 Table 5-2 Critical Analysis Locations for Leak-Before-Break of North Anna Unit 2 RCL Lines ................ 5-3 Table 6-1 Flaw Sizes for North Anna Unit 1 Yielding a Leak Rate of 1O gpm for the RCL Lines ............. 6-3 Table 6-2 Flaw Sizes for North Anna Unit 2 Yielding a Leak Rate of 1O gpm for the RCL Lines ............. 6-3 Table 7-1 Stability Results for North Anna Unit 1 Based on Elastic-Plastic J-lntegral Evaluations .......... 7-5 Table 7-2 Stability Results for North Anna Unit 2 Based on Elastic-Plastic J-lntegral Evaluations .......... 7-5 Table 7-3 Flaw Stability Results for North Anna Unit 1 Yielding a Leak Rate of 10 gpm for the RCL Lines Based on Limit Load ....................................................................................................... 7-6 Table 7-4 Flaw Stability Results for North Anna Unit 2 Yielding a Leak Rate of 10 gpm for the RCL Lines Based on Limit Load ....................................................................................................... 7-6

. Table 7-5 Leakage Flaw Sizes, Critical Flaw Size and Margins for North Anna Unit 1 SGON Alloy 82/182 Welds ........................................................................................................................... 7-7 Table 8-1 Summary of Transients (Representative 80-Year Design) ....................................................... 8-4 Table 8-2 Typical Fatigue Crack Growth at [ ]a,c,e (40, 60, and 80 years) ............. 8-5 Table 8-3 Summary of Transients for North Anna Units 1 and 2 (40, 60, 80 years) ................................. 8-5 Table 9-1 Leakage Flaw Sizes, Critical Flaw Sizes and Margins for North Anna Unit 1 ......................... 9-2 Table 9-2 Leakage Flaw Sizes, Critical Flaw Sizes and Margins for North Anna Unit 2 ......................... 9-3 WCAP-11164-NP January 2020 Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 vii LIST OF FIGURES Figure 3-1 Hot Leg Coolant Pipe ................................................................................................... 3-9 Figure 3-2 Schematic Diagram of North Anna Units 1 and 2 Primary Loop Showing Weld Locations ........................................................................................................... 3-1 0 Figure 4-1 Pre-Service J vs. ~a for SA351-CF8M Cast Stainless Steel at 600°F ....................... 4-23 Figure 5-1 Schematic Diagram of North Anna Units 1 and 2 Primary Loop Showing Critical Weld Locations ................................................................................................. 5-4 Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures ........................ 6-4 Figure 6-2 ]a,c,e Pressure Ratio as a Function of LID ......................................... 6-5 Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack ...................................... 6-6 Figure 7-1 [ ]a,c,e Stress Distribution ............................................................................ 7-8 Figure 8-1 Typical Cross-Section of [ ]a.c.e ......................................... 8-6 Figure 8-2 Reference Fatigue Crack Growth Curves for Carbon & Low Alloy Ferritic Steels ........ 8-7 Figure 8-3 Reference Fatigue Crack Growth Law for [ ]a.c,e in a Water Environment at 600°F ........................................................................................................................ 8-8 Figure A-1 Pipe with a Through-Wall Crack in Bending ................................................................. A-2 WCAP-11164-NP January 2020 Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 viii EXECUTIVE

SUMMARY

The original structural design basis of the reactor coolant system for the North Anna Units 1 and 2 Nuclear Power Plants required consideration of dynamic effects resulting from pipe breaks and that protective measures for such breaks are incorporated into the design. Subsequent to the original North Anna design, an additional concern of asymmetric blowdown loads was raised as described in Unresolved Safety Issue A-2 (Asymmetric Slowdown Loads on the Reactor Coolant System). North Anna Units 1 and 2 Nuclear Power Plants were not part of the utilities which sponsored Westinghouse to resolve the A-2 issue. Generic analyses by Westinghouse to resolve the A-2 issue was approved by the NRC and documented in Generic Letter 84-04 (Reference 1-1). Generic Letter 84-04 (Reference 1-1) serves as the generic safety evaluation report (SER) for leak-before-break.

Research by the NRC and industry coupled with operating experience determined that safety could be negatively impacted by placement of pipe whip restraints on certain systems. As a result, NRC and industry initiatives resulted in demonstrating that leak-before-break (LBB) criteria can be applied to reactor coolant system piping based on fracture mechanics technology and material toughness.

Subsequently, the NRC modified 10CFR50 General Design Criterion 4, and published in the Federal Register (Vol. 52, No. 207) on October 27, 1987 its final rule, "Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures,"

(Reference 1-2). This change to the rule allows use of leak-before-break (LBB) technology for excluding from the design basis the dynamic effects of postulated ruptures in primary coolant loop piping in pressurized water reactors (PWRs). With the LBB technology, North Anna Units 1 and 2 had shown that the dynamic effects of. postulated ruptures in primary coolant loop piping in pressurized water reactors (PWRs) can be excluded from the design basis, as demonstrated in WCAP-11163 and its supplement (References 1-7 and 1-8). The methodology of WCAP-11163 and its supplement (References 1-7 and 1-8) is consistent with the original, generic work performed as the basis of Generic Letter 84-04 (Reference 1-1).

References 1-7 and 1-8 had demonstrated compliance with LBB technology for the North Anna reactor coolant system piping based on a plant specific analysis. The LBB evaluation was performed based on loading, pipe geometry and fracture toughness considerations, enveloping critical locations were determined at which leak-before-break crack stability evaluations were made. Through-wall flaw sizes were found which would cause a leak at a rate of ten (10) times the leakage detection system capability of the plant. Large margins for such flaw sizes were demonstrated against flaw instability. Finally, fatigue crack growth was shown not to be an issue for the primary loops.

Over the years, subsequent LBB evaluations were performed, in order to maintain an updated and applicable analysis of record, for the Replacement Steam Generator (RSG) program, Reactor Coolant Pump {RCP) support modification, program, 60 year plant life First License Renewal (FLR) program, 2% power uprate program, Measurement Uncertainty Recapture {MUR) program, as well as Structural Weld Overlay (SWOL) and Weld Inlay programs. Since there are potential susceptibility to primary water stress corrosion cracking (PWSCC) phenomena that could occur due to 82/182 weld alloys at the Units 1 and 2 Steam Generator Inlet Nozzles (SGIN's) and Steam Generator Outlet Nozzles {SGON's), LBB evaluations also consider the mitigated 82/182 weld using 52/152 alloys and the unmitigated 82/182 weld.

For the current North Anna Units 1 and 2 Nuclear Power Plants Subsequent License Renewal (SLR) program, all the above input are reevaluated to ensure that the existing LBB evaluation conclusions remain applicable for the 80 year plant life.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 ix Along with the SLR program, the report also addresses North Anna primary coolant loop material toughness properties which are used in the fracture mechanics evaluation. The North Anna Units 1 and 2 primary coolant loop piping systems include cast austenitic stainless steel materials (A351-CF8A and A351-CF8M) that will thermally age. As of June 2018, North Anna Units 1 and 2 are operating at 32.1 *and 30.5 EFPY, respectively. For the SLR program that accounts for 80 year of plant operation, the materials will thermally age. Even though the aged flow stress are higher than the unaged flow stress because of thermally aging, material tearing modulus will be lower. To predict more accurately the impact of material aging to the material fracture toughness properties including the shearing modulus, NUREG/CR-4513, Revision 2, is used. Fully aged fracture toughness properties are used for the LBB evaluation.

Based on the LBB evaluations documented in this report (Revision 1), it is demonstrated that the conclusions of WCAP-11163 provided in References 1-7 and 1-8 remain applicable, and the primary loop integrity for the North Anna Units 1 and 2 plants remains consistent with the NRC position for exemption from consideration of dynamic effects. The dynamic effects of reactor coolant system primary loop pipe breaks need not be considered in the structural design basis of the North Anna Units 1 and 2 Nuclear Power Plants for the 80 year plant life (SLR Program).

The following table presents a brief summary of the analysis locations considered in this report. This table includes the material type(s) considered for each location, as well as the specific evaluation which are used to justify the applicability of Leak-Before-Break for the full Reactor Coolant Loop (RCL) piping system.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 X Table ES-1 Summary of Analysis Locations, Materials, and Evaluation Types Piping Location Unit Material Evaluation Type Results Summary Local stability<2l Table 7-1 and Table 7-2 1 Both CF8A<1>

Global failure<3> Table 7-3 and Table 7-4 Local stability(2> Table 7-1 and Table 7-2 Both CF8M Hot Global failure< 3> Table 7-3 and Table 7-4 Leg Alloy 82/182 weld 2 1 see note (4)

(with SWOL)

Alloy 82/182 weld 2 see note (5)

(with Alloy ~~/152 inlay)

Local stability<2> Table 7-1 and Table 7-2 Both CFBM Global failure<3> Table 7-3 and Table 7-4 3 1 Alloy 82/182 weld Global failure<6l Table 7-5 Alloy 82/182 weld 2 see note (5)

(with Alloy 52/152 inlay)

Crossover Local stability<2> Table 7-1 and Table 7-2 Leg CF8A< 1>

Global failure< 3> Table 7-3 and Table 7-4 4 Both Local stabi1ity<2> Table 7-1 and Table 7-2 CF8M<1l Global failure< 3l Table 7-3 and Table 7-4 Local stability<2l Table 7-1 and Table 7-2 11 Both CF8M< 1>

Global failure< 3l Table 7-3 and Table 7-4 Local stability(2> Table 7-1 and Table 7-2 12 Both CF8N 1l Cold Global failure(3 l Table 7-3 and Table 7-4 Leg Local stability<2l Table 7-1 and Table 7-2 5 Both CF8M<1>

Global failure< 3l Table 7-3 and Table 7-4 Notes: (1) Cast austenitic stainless steel .base metal is more limiting that stainless steel weld material, as noted in Section 4.3.

(2) Local stability evaluations accounts for the thermal aging effects on the cast austenitic stainless steel base metal.

{3) While the global failure evaluations use the base metal material properties, the analysis also accounts for the fracture strength reductions due to the welding process used during fabrication.

(4) The Unit 1 dissimilar metal weld with structural weld overlay is evaluated separately from this report, as noted in Section 7.3.

(5) Because the Alloy 52/152 inlay prevents primary water stress corrosion cracking, this evaluation is bounded by the CF8M base metal as justified in Section 4.3.3.

(6) Global failure evaluation includes crack morphology parameter to account for the effects of primary water stress corrosion cracking.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-1

1.0 INTRODUCTION

1.1 PURPOSE This report applies to the North Anna Units 1 and 2 Reactor Coolant System (RCS) primary loop piping. It is intended to demonstrate that for the specific parameters of the North Anna Units 1 and 2 Nuclear Power Plants, RCS primary loop pipe breaks need not be considered in the structural design basis for the 80 year plant life (Subsequent License Renewal Program). This report also includes the LBB evaluation results based on the Replacement Steam Generator (RSG) program, Reactor Coolant Pump (RCP) support modification program, 60 year plant life First License Renewal (FLR) program, 2% power uprate program as well as Measurement Uncertainty Recapture (MUR) program.

In addition, this report reviews the dissimilar metal weld (DMW) locations at the Units 1 and 2 SGIN's and SGON's using Alloy 82/182 nickel-base materials which are susceptible to primary water stress corrosion cracking (PWSCC) to confirm that those locations have been appropriately mitigated and evaluated for LBB evaluation. Except for the Unit 1 SGON nozzle welds, all DMW locations (Units 1 and 2 SG nozzles) have been mitigated from PWSCC, using weld overlay or weld inlay. All critical locations are evaluated, including the mitigated and unmitigated SGIN and SGON locations, to reconfirm that the LBB evaluation conclusions remain valid for 80 year plant life SLR program and for the limiting material profile at each analysis location.

1.2 BACKGROUND

INFORMATION Westinghouse has performed considerable testing and analysis to demonstrate that Reactor Coolant System (RCS) primary loop pipe breaks can be eliminated from the structural design basis of all Westinghouse plants. The concept of eliminating pipe breaks in the RCS primary loop was first presented to the NRC in 1978 in WCAP-9283 (Reference 1-3). That topical report employed a deterministic fracture mechanics evaluation and a probabilistic analysis to support the elimination of RCS primary loop pipe breaks. That approach was then used as a means of addressing Generic Issue A-2 and Asymmetric Loss of Coolant Accident (LOCA) Loads.

Westinghouse performed additional testing and analysis to justify the elimination . of RCS primary loop pipe breaks. This material was provided to the NRC along with Letter Report NS-EPR-2519 (Reference 1-4).

The NRC funded research through Lawrence Livermore National Laboratory (LLNL) to *address this same issue using a probabilistic approach. As part of the LLNL research effort, Westinghouse performed extensive evaluations of specific plant loads, material properties, transients, and system geometries to demonstrate that the analysis and testing previously performed by Westinghouse and the research performed by LLNL applied to all Westinghouse plants (References 1-5 and 1-6). The results from the LLNL study were released at a March 28, 1983, ACRS Subcommittee meeting. These studies, which are applicable to all Westinghouse plants east of the Rocky Mountains, determined the mean probability of a direct LOCA (RCS primary loop pipe break) to be 4.4 x 10-12 per reactor year and the mean probability of an indirect LOCA to be 10-7 per reactor year. Thus, the results previously obtained by Westinghouse (Reference 1-3) were confirmed by an independent NRC research study.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-2 Based on the studies by Westinghouse, LLNL, the ACRS, and the Atomic Industrial Forum (AIF), the NRC completed a safety review of the Westinghouse reports submitted to address asymmetric blowdown loads that result from a number of discrete break locations on the PWR primary systems. The NRC Staff evaluation (Reference 1-1) concludes that an acceptable technical basis has been provided so that asymmetric blowdown loads need not be considered for those plants that can demonstrate the applicability of the modeling and conclusions contained in the Westinghouse response or can provide an equivalent fracture mechanics demonstration of the primary coolant loop integrity. In a more formal recognition of Leak-Before-Break (LBS) methodology applicability for PWRs, the NRC appropriately modified 10 CFR 50, General Design Criterion 4, "Requirements for Protection Against Dynamic Effects for Postulated Pipe Rupture" (Reference 1-2).

For North Anna Nuclear Power Station Units 1 and 2, the postulated pipe breaks for the RCS primary loop piping have been evaluated using LBS evaluation methods. It is demonstrated that the dynamic effects of the pipe rupture resulting from postulated breaks in the primary loop piping need not be considered in the structural design basis of North Anna Units 1 and 2. The original LBB evaluation results for the RCS primary loop were documented in WCAP-11163 report (Reference 1-7) in 1986, and its supplement (Reference 1-8) in 1988.

For the Subsequent License Renewal (SLR) program, this report demonstrates that the conclusions reached in References 1-7 and 1-8 remain applicable in the structural design basis for the 80 year plant life for the specific parameters of the North Anna Units 1 and 2 Nuclear Power Stations.

1.3 SCOPE AND OBJECTIVES The general purpose of this investigation is to demonstrate leak-before-break for the primary loops in North Anna Units 1 and 2 on a plant specific basis for the 80 year plant life. The recommendations and criteria proposed in References 1-9 and 1-10 are used in this evaluation.

These criteria and resulting steps of the evaluation procedure can be briefly summarized as follows:

1. Calculate the applied loads. Identify the locations at which the highest stress occurs.

2. Identify the limiting material profiles and the associated material properties. 1

3. Postulate a surface. flaw at the governing locations. Determine fatigue crack growth.

Show that a through-wall crack will not result. '*

4. Postulate a through-wall flaw at the governing locations. The size of the flaw should be large enough so that the leakage is assured of detection with margin using the installed leak detection equipment when the pipe is subjected to normal operating loads. A margin of 10 is demonstrated between the calculated leak rate and the leak detection capability.

5. Using faulted loads, demonstrate that there is a margin of 2 between the leakage flaw size and the critical flaw size.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-3

6. Review the operating history to ascertain that operating experience has indicated no particular susceptibility to failure from the effects of corrosion, water hammer or low and high cycle fatigue.

7. For the materials actually used in the plant provide the properties including toughness and tensile test data. Evaluate long term effects such as thermal aging.

8. Demonstrate margin on the calculated applied load value; margin of 1.4 using algebraic summation of loads or margin of 1.0 using absolute summation of loads.

This report provides a fracture mechanics demonstration of primary loop integrity for the North Anna Units 1 and 2 plants consistent with the NRC position for exemption from consideration of dynamic effects.

The LBB evaluation summarized in this report consider the limiting weld locations. of the RCL piping. In general, the analyses consider the material properties of the piping base metal, which are more limiting that the weld materials. The evaluations also consider the Unit 1 SGON's base metal and the PWSCC susceptible 82/182 alloy and as well as for Unit 2 SGIN's and SGON's in the weld inlay program involving 52/152 alloy. Separate from this report, an updated LBB evaluation was performed for the full structural weld overlay (SWOL) program applied to the Unit 1 SGIN's (Reference 1-11). The re-evaluations were performed to ensure that the LBB evaluation conclusions remain valid for 80 year plant life in the SLR program.

It should be noted that the terms "flaw" and "crack" have the same meaning and are used interchangeably. "Governing location" and "critical location" are also used interchangeably throughout the report.

The computer codes used in this evaluation for leak rate and fracture mechanics calculations have been validated and used for all the LBB applications by Westinghouse.

1.4 REFERENCES

1-1 USNRC Generic Letter 84-04,

Subject:

"Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops," February 1, 1984.

1-2 Nuclear Regulatory Commission, 10 CFR 50, Modification of General Design Criteria 4 Requirement~ for Protection Against Dynamic Effects of Postulated Pipe Ruptures, Final Rule, Federal RegisterNol. 52, No. 207/Tuesday, October 27, 11987/Rules and Regulations, pp. 41288-41295.

1-3 WCAP-9283, "The Integrity of the Primary Piping Systems of Westinghouse Nuclear Power Plants During Postulated Seismic Events," March, 1978.

1-4 Westinghouse Proprietary Class 2 Letter Report NS-EPR-2519, Westinghouse (E. P.

Rahe) to NRC (D. G. Eisenhut), November 10, 1981.

1-5 Letter from Westinghouse (E. P. Rahe) to NRC (W. V. Johnston) dated April 25, 1983.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-4 1-6 Letter from Westinghouse (E. P. Rahe) to NRC (W. V. Johnston) dated July 25, 1983.

1-7 WCAP-11163, Rev. 0, "Technical Bases for Eliminating Large Primary Loop Pipe Rupture as a Structural Design Basis for North Anna Units 1 & 2," August 1986.

1-8 WCAP-11163, Supplement 1, Rev. 0, "Additional Information in Support of the Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for North Anna Units 1 and 2," January 1988.

1-9 Standard Review Plan: Public Comments Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal RegisterNol. 52, No. 167/Friday August 28, 1987/Notices, pp.

32626-32633.

1-10 NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.

1-11 Structural Integrity Associates, Inc. Letter, "Leak Before Break (LBB) Update for North Anna Unit 1 Steam Generator Inlet Nozzle Weld Overlay (WOL) for Subsequent License Renewal," July 10, 2019.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1 2.0 OPERATION AND STABILITY OF THE REACTOR COOLANT SYSTEM 2.1 STRESS CORROSION CRACKING The Westinghouse reactor coolant system primary loops have an operating history that demonstrates the inherent operating stability characteristics of the design. This includes a low susceptibility to cracking failure from the effects of corrosion (e.g., intergranular stress corrosion cracking (IGSCC)). This operating history totals over 1400 reactor-years, including 16 plants each having over 30 years of operation, 10 other plants each with over 25 years of operation, 11 plants each with over 20 years of operation, and 12 plants each with over 15 years of operation.

In 1978, the United States Nuclear Regulatory Commission (USNRC) formed the second Pipe Crack Study Group. (The first Pipe Crack Study Group (PCSG) established in 1975, addressed cracking in boiling water reactors only.) One of the objectives of the second PCSG was to include a review of the potential for stress corrosion cracking in Pressurized Water Reactors (PWR's). The results of the study performed by the PCSG were presented in NUREG-0531 (Reference 2-;1) entitled "Investigation and Evaluation of Stress Corrosion Cracking in Piping of Light Water Reactor Plants." In that report the PCSG stated:

"The PCSG has determined that the potential for stress-corrosion cracking in PWR primary system piping is extremely low because the ingredients that produce IGSCC are not all present. The use of hydrazine additives and a hydrogen overpressure limit the oxygen* in the coolant to very low levels. Other impurities that might cause stress-corrosion cracking, such as halides or caustic, are also rigidly controlled. Only for brief periods during reactor shutdown when the coolant is exposed to the air and during the subsequent startup are conditions even marginally capable of producing stress-corrosion cracking in the primary systems of PWRs. \

Operating experience in PWRs supports this determination. To date, no stress corrosion cracking has been reported in the primary piping or safe ends of any PWR."

During 1979, several instances of cracking in PWR feedwater p1pmg led to the establishment of the third PCSG. The investigations of the PCSG reported in NUREG-0691 (Reference 2-2) further confirmed that no occurrences of _IGSCC have been reported for PWR 'primary coolant systems.

• As stated above, for the Westinghouse plants there is no history of cracking failure in the reactor coolant system loop. The discussion below further qualifies the PCSG's findings.

For stress corrosion cracking (SCC) to occur in piping, the following three conditions must exist simultaneously: high tensile stresses, susceptible material, and a corrosive environment. Since some residual stresses and some degree of material susceptibility exist in any stainless steel piping, the potential for stress corrosion is minimized by properly selecting a material immune to sec as well as preventing the occurrence of a corrosive environment. The material specifications consider compatibility with the system's Operation and Stability of the Reactor Coolant System January 2020 WCAP-11164-NP Revision 2

• • • This record was final approved on 1/17/2020 11 :18:49 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 operating environment (both internal and external) as well as other material in the system, applicable ASME Code rules, fracture toughness, welding, fabrication, and processing.

The elements of a water environment known to increase the susceptibility of austenitic stainless steel to stress corrosion are: oxygen, fluorides, chlorides, hydroxides, hydrogen peroxide, and reduced forms of sulfur (e.g., sulfides, sulfites, and thionates). Strict pipe cleaning standards prior to operation and careful control of water chemistry during plant operation are used to prevent the occurrence of a corrosive environment. Prior to being put into service, the piping is cleaned internally and externally. During flushes and preoperational testing, water chemistry is controlled in accordance with written specifications. Requirements on chlorides, fluorides, conductivity, and pH are included in the acceptance criteria for the piping.

During plant operation, the reactor coolant water chemistry is monitored and maintained within very specific limits. Contaminant concentrations are kept below the thresholds known to be conducive to stress corrosion cracking with the major water chemistry control standards being included in the plant operating procedures as a condition for plant operation. For example, during normal power operation, oxygen concentration in the RCS is expected to be in the parts-per-billion (ppb) range by controlling charging flow chemistry and maintaining hydrogen in the reactor coolant at specified concentrations. Halogen concentrations are also stringently controlled by maintaining concentrations of chlorides and fluorides within the specified limits. Thus during plant operation, the likelihood of stress corrosion cracking is minimized.

The potential susceptibility to primary water stress corrosion cracking (PWSCC) in materials such as Alloy 82/182 in the dissimilar metal welds in the North Anna Units 1 and 2 RCS primary loop piping, especially in Steam generator Inlet and outlet nozzles, are addressed in this report. It should be noted that these locations, except for the Unit 1 SGON nozzle welds, all DMWs have been mitigated by implementing the weld overlay and weld inlay. The LBB assessments appropriately consider each material profile.

For the structural weld overlay repair applied for the Unit 1 SGIN's, the Leak Before Break evaluation for subsequent license renewal is documented in Reference 2-3. As stated in Reference 2-3, "Leak Before Break evaluation (for SG Inlet Nozzle as documented in calculation 1100226.303 Revision 3), can continue to be use~ as justification/demonstration of LBB behavior for the weld overlay, for the period of.. subsequent license renewal (80 years) at North Anna Unit 1."

The LBB evaluation for unmitigated weld locations at Unit 1 SG outlet nozzle includes an additional evaluation considering the Alloy 82/182 by assuming the PWSCC will actually occur and calculate the leakage flaw size based upon PWSCC crack morphology to determine the crack stability for 80 year plant life in the SLR program.

Operation and Stability of the Reactor Coolant System January 2020 WCAP-11164-NP Revision 2

• • • This record was final approved on 1/17/2020 11:18:49 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 2.2 WATER HAMMER Overall, there is a low potential for water hammer in the RCS since it is designed and operated to preclude the voiding condition in normally filled lines. The reactor coolant system, including piping and primary . components, is designed for normal, upset, emergency, and faulted condition transients. The design requirements are conservative relative to both the number of transients and their severity. Relief valve actuation and the associated hydraulic transients following valve opening are considered in the system design. *Other valve and pump actuations are relatively slow transients with no significant effect on the system dynamic loads. To ensure dynamic system stability, reactor coolant parameters are stringently controlled. Temperature during normal operation is maintained within a narrow range; pressure is controlled by pressurizer heaters and pressurizer spray also within a narrow range for steady-state conditions. The flow characteristics of the system remain constant during a fuel cycle because the only governing parameters, namely system resistance and the reactor coolant pump characteristics, are contro,lled in the design process. Additionally, Westinghouse has instrumented typical reactor coolant systems to verify the flow and vibration characteristics of the system. Preoperational testing and operating experience have verified the Westinghouse approach. The operating transients of the RCS primary piping are such that no significant water hammer can occur.

2.3 LOW CYCLE AND HIGH CYCLE FATIGUE An assessment of the low cycle fatigue loadings was carried out as part of this study in the form of a fatigue crack growth analysis, as discussed in Section 8.0.

High cycle fatigue loads in the system would result primarily from pump vibrations. These are minimized by restrictions placed on shaft vibrations during hot functional testing and operation. During operation, an alarm signals the exceedance of the vibration limits. Field vibration measurements have been made on the reactor coolant loop piping in a number of plants during hot functional testing, including plants similar to North Anna Units 1 and 2.

Stresses in the elbow below the reactor coolant pump resulting from system vibration have been found to be very small, between 2 and 3 ksi at the highest. These stresses are well below the fatigue endurance limit for the material and would also result in an applied stress intensity factor below the threshold for fatigue crack growth.

2.4 WALL THINNING, CREEP, AND CLEAVAGE; Wall thinning by erosion and erosion-corrosion effects ~hould not occur in the primary loop piping due to the low velocity, typically less than 1.0 ft/sec and the stainless steel material, which is highly resistant to these degradation mechanisms. The cause of wall thinning is'related to high water velocity and is therefore clearly not a mechanism that would affect the primary loop piping.

Creep is typical experienced for temperatures over 700°F for stainless steel material, and the maximum operating temperature of the primary loop piping is well below this temperature value; therefore, there would be no significant mechanical creep damage in stainless steel piping.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-4 Cleavage type failures are not a concern for the operating temperatures and the stainless steel material used in the primary loop piping.

2.5 REFERENCES

2-1 Investigation and Evaluation of Stress-Corrosion Cracking in Piping of Light Water Reactor Plants, NUREG-0531, U.S. Nuclear Regulatory Commission, February 1979.

2-2 Investigation and Evaluation of Cracking Incidents in Piping in Pressurized Water Reactors, NUREG-0691, U.S. Nuclear Regulatory Commission, September 1980.

2-3 Structural Integrity Associates, Inc. Letter, "Leak Before Break (LBB) Update for North Anna Unit 1 Steam Generator Inlet Nozzle Weld Overlay (WOL) for Subsequent License Renewal," July 10, 2019.

Operation and Stability of the Reactor Coolant System January 2020 WCAP-11164-NP Revision 2

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-7 I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 3.0 PIPE GEOMETRY AND LOADING

3.1 INTRODUCTION

TO METHODOLOGY The general approach is discussed first. As an example a segment of the primary coolant hot leg pipe is shown in Figure 3-1. The as-built outside diameter and minimum wall thickness of the pipe are 34.06 in. and 2.43 in., respectively, as shown in the figure. The normal stresses at the weld locations are from the load combination procedure discussed in Section 3.3 whereas the faulted loads are as described in Section 3.4. The components for normal loads are pressure, deadweight and thermal expansion. An additional component, Safe Shutdown Earthquake (SSE), is considered for faulted loads. Tables 3-1 through 3-4 show the enveloping loads for North Anna Units 1 and 2. As seen from Tables 3-3 and 3-4, the highest stressed location in the entire North Anna reactor coolant loop is at the Unit 1 rnactor vessel outlet nozzle to pipe weld (Location 1). This is one of the locations at which leak-before-break is to be established.

Essentially a circumferential flaw is postulated to exist at this location which is subjected to both the normal loads and faulted loads to assess leakage and stability, respectively. The loads (developed below) at this location are also given in Figure 3-1.

Since the primary loop piping are made of different materials (A351-CF8A and A351-CF8M),

locations other than the highest stressed pipe location were examined by taking into consideration both fracture toughness and stress. The seven most critical locations among the entire primary loop are identified after the full analysis is completed (see Section 5.0). Once loads (this section) and fracture toughnesses (Section 4.0) are obtained, the critical locations are determined (Section 5.0). At these locations, leak rate evaluations (Section 6.0) and fracture mechanics evaluations (Section 7.0) are performed per the guidance of References 3-1 and 3-2.

For local and global failure mechanisms, all locations are evaluated using the cast stainless steel material properties (A351-CF8A and A351-CF8M) which present a limiting condition not only due to their tensile properties in unaged condition but also the material fracture toughness and tearing modulus reductions due to the thermal aging effects for the entire 80 year plant life.

The cast stainless steel presents a more limiting tensile property condition when compared to the PWSCC susceptible Alloy 82/182 as well as non-susceptible Alloy 52/152 tensile properties found in Units 1 and 2 SG (inlet/outlet) nozzle welds. The A351-CF8A and A351-CF8M materials are even more limiting because they thermally age with time.

For completeness of the evaluation, the LBB analysis for the unmitigated locations at Unit 1 SG outlet nozzle considers both the A351 base metal and the Alloy 82/182 DMW by assuming the PWSCC will actually occur and calculate the leakage flaw size based upon PWSCC crack morphology. The LBB evaluation for the mitigated SG nozzle locations due to the applications of full weld overlay program for Unit 1 SGIN's and weld inlay for Unit 2 SGIN's and SGON's also considers the A351 base metal and both Alloys 52/152 and 82/182 as part of the 80 year plant life period in SLR program.

The underlying LBB evaluations based on the limiting cast austenitic stainless steel (CASS) base metal remain valid and satisfy the requirements to demonstrate the postulated crack Pipe Geometry and Loading January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-2 structural stability. Added assurance is obtained by reviewing and evaluating the presence of nickel alloy welds.

Fatigue crack growth (Section 8.0) assessment and stability margins are also evaluated (Section 9.0). All the weld locations considered for the LBB evaluation are those shown in Figure 3-2.

• 3.2 CALCULATION OF LOADS AND STRESSES The stresses due to axial loads and bending moments are calculated by the following equation:

F M (3-1) a=-+ -

A Z

where, a = stress, ksi F = axial load, kips M = bending moment, in-kips A = pipe cross-sectional area, in 2 Z = section modulus, in 3 The total moments for the desired loading combinations are calculated by the following equation:

M =.JM~ + M~ +Mi (3-2)

where, M = total moment for required loading Mx = X component of moment (torsion)

Mv = Y component of bending moment Mz = Z component of bendi~g moment NOTE: X-axis is along the center line of the pipe.

The axial load and bending moments for leak rate predictions and crack stability analyses are computed by the methods to be explained in Sections 3.3 and 3.4.

Pipe Geometry and Loading January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-3 3.3 LOADS FOR LEAK RATE EVALUATION The normal operating loads for leak rate predictions are calculated by the following equations:

Unit 1, Structural Weld Overlay (SWOL) loads are included:

F = Fow + FTH + Fp + FswOL (3-3)

Mx = (Mx)ow + (Mx)TH + (Mx)swoL (3-4)

Mv = (Mv)ow + (Mv)TH + (Mv)swOL (3-5)

Mz = (Mz)ow + (MzhH + (Mz)swoL (3-6)

Unit 2:

F = Fow+ Frn + Fp (3-7)

Mx = (Mx)ow + (MxhH (3-8)

Mv = (Mv)ow + (Mv)rn (3-9)

Mz = (Mz)ow + (Mz)TH (3-10)

The subscripts of the above equations represent the following loading cases:

OW = deadweight TH = normal thermal expansion p = load due to internal pressure SWOL = Structural Weld Overlay This method of combining loads is often referred to as the algebraic sum method (References 3-1 and 3-2).

The loads based on this method of combination are provided in Tables 3-1 and 3-2 at all the weld locations identified in Figure 3-2. These loads bound all three loops. The as-built dimensions are also given in Tables 3-1 and 3-2.

3.4 LOAD COMBINATION FOR CRACK STABILITY ANALYSES In accordance with Standard Review Plan 3.6.3 (References 3-1 and 3-2), the margin in terms of applied loads needs to be demonstrated by crack stability analysis. Margin on loads of 1.4 (2) can be demonstrated if normal plus Safe Shutdown Earthquake (SSE) are applied. The 1 ~4 (2) margin should be reduced to 1.0 if the deadweight, thermal expansion, internal pressure,'

pressure expansion, SWOL, SSEINERTIA and seismic anchor motion (SAM) loads are combined based on individual absolute values as shown below. It is noted that additional conservatism exists in this evaluation since the seismic loads are based on the original spectra model and do not account for advancements credited to ASME Code Case N-411. This Code Case could be used in future evaluation efforts if it is necessary to reduce conservatism and regain margin.

Pipe Geometry and Loading January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-4 The absolute sum of loading components is used for the LBB analysis which results in higher magnitude of combined loads and thus satisfies a margin on loads of 1.0. The absolute summation of loads is shown in the following equations:

Unit 1, Structural Weld Overlay (SWOL) loads are included:

.F = I Fow I+ I Fm I+ I Fp I+ I FssEINERTIA I+ I FssEAM I+ IFswod (3-11)

Mx = I (Mx)ow I+ I (Mx)rn I+ I (Mx)ssE1NERT1Al + I (Mx)ssEAM l + l(Mx)swod (3-12)

Mv = I (Mv)ow l + I(MvhH I+ I (Mv)ssEINERTIAl + l (Mv)ssEAM I+ l(Mv)swod (3-13)

Mz = l (Mz)ow l + I (Mz)TH l + I(Mz)ssE1NERT1AI + l (Mz)ssEAM I + l(Mz)swod (3-14)

Unit 2:

F = I Fowl + I Fm I+ I Fp I + I FssEJNERTIA I+ I FssEAM I (3-15)

Mx = I (Mx)ow I + I(Mx)rn I+ I(Mx)ssEINERTIA I+ I (Mx)ssEAM I (3-16)

Mv = I (Mv)ow I+ I (Mv)m I+ I (Mv)ssEINERT1AI + I (Mv)ssEAM I (3-17)

Mz = I (Mz)ow I + I(Mz)TH I + I (Mz)ssEINERTIA I+ I (Mz)ssEAMI (3-18) where subscript SSEINERTIA refers to safe shutdown earthquake inertia, SSEAM is safe shutdown earthquake anchor motion, respectively.

The loads so determined are used in the fracture mechanics evaluations (Section 7.0) to demonstrate the LBB margins at the locations established to be the governing locations. These loads at all the weld locations (see Figure 3-2) are given in Tables 3-3 and 3-4. The loads in these tabl.es bound all three loops.

3.5 REFERENCES

3-1 Standard Review Plan: Public Comments Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal RegisterNol. 52, No. 167/Friday, August 28, 1987/Notices, pp. 32626-32633.

3-2 NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before'"Break Evaluation Procedures.

Pipe Geometry and Loading January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-5 Table 3-1 Dimensions, Normal Loads and Stresses for North Anna Unit 1 Location 8 Outside Minimum Axial Loadb Moment Total Stress Node Weld Diameter (in) Thickness (in) (kips) (in-kips) (ksi)

Point Point 103 1 34.06 2.43 1556 10616 12.396 108 6 34.06 2.43 1574 1583 7.405 110 7 34.06 2.43 1574 5984 9.872 120 2 36.60 2.70 1710 9321 10.049 139 3 36.96 2.88 1650 8632 8.888 145 8 36.22 2.58 1547 6825 8.859 147 4 36.22 2.58 1542 7398 9.107

) 153 9 36.22 2.58 1751 2635 7.652 154 10 36.22 2.58 1751 4168 8.368 159 11 36:22 2.58 1859 12204 12.514 164 12 32.82 2.56 1297 5352 8.460 167 13 32.82 2.56 1297 4442 7.928 169 14 32.82 2.56 1298 2324 6.691 173 15 32.82 2.56 1287 5776 8.665 179 5 32.82 2.56 1274 7531 9.639 Notes:

a. See Figure 3-2

b. Included Pressure

c. All locations ~re evaluated based on the limiting properties of the CASS. base metal, except as iden~ified in notes (d) and (e).

d. Weld point 2 is a DMW using Alloy 82/182 with a full SWOL using Alloy 52/152. The CASS base metal, without SWOL is evaluated in this report. The SWOL condition is evaluated separately, as noted in Section 2.1 *

e. Weld point 3 is an unmodified DMW using Alloy 82/182. This location is evaluated considering both the CASS base metal and the nickel alloy material.

Pipe Geometry and Loading January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3

• 3-6 Table 3-2 Dimensions, Normal Loads and Stresses for North Anna Unit 2 Location a Outside Minimum Axial Loadb Moment Total Stress Node Weld Diameter (in) Thickness (in) (kips) (in-kips) (ksi)

Point Point 103 1 34.06 2.43 1487 21363 18.137 108 6 34.06 2.43 1505 1548 7.102 110 7 34.06 2.43 1506 11457 12.659 120 2 36.60 2.70 1599 17672 13.342 139 3 36.96 2.88 1642 8688 8.885 145 8 36.22 2.58 1550 7388 9.134 147 4 36.22 2.58 1545 7501 9.167 153 9 36.22 2.58 1768 2012 7.422 154 10 36.22 2.58 1764 3620 8.158 159 11 36.22 2.58 1856 11738 12.285 164 12 32.82 2.56 1361 4938 8.481 167 13 32.82 2.56 1361 4416 8.175 169 14 32.82 2.56 1362 2662 7.152 173 15 32.82 2.56 1351 5327 8.666 179 5 32.82 2.56 1315 6687 9.313 Notes:

a. See Figure 3-2

b. Included Pressure

c. All locations are evaluated based on the limiting properties of the GASS base metal, except as identified in note (d).

d. Weld points 2 and 3 are DMWs using Alloy 82/182 with an inlay of Alloy 52/152 .. This location is evaluated using the more limiting properties of the CASS base metal.

Pipe Geometry and Loading January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-7 Table 3-3 Faulted Loads and Stresses for North Anna Unit 1 Locationa Axial Loadb Moment Total Stress (kips) (in-kips) (ksi)

Node Point Weld Point 103 1 2039 39247 30.447 108 6 2019 5872 11.654 110 7 2062 23755 21.859 120 2 2285 35290 23.481 139 3 1972 25380 16.797 145 8 1897 19445 16.036 147 4 1893 14601 13.757 153 9 1836 9714 11.f-?O 154 10 1835 12071 -12.363 159 11 1935 23258 17.955 164 12 1841 20935 19.811 167 13 1842 16664 17.318 169 14 1854 10642 13.843 173 15 1796 13902 15.510 179 5 1680 18937 17.980 Notes:

a. See Figure 3-2

b. Included Pressure

\

Pipe Geometry and Loading January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-8 Table 3-4 Faulted Loads and Stresses for North Anna Unit 2 Locationa Axial Loadb Moment Total Stress (kips) (in-kips) (ksi)

Node Point Weld Point 103 1 1970 28705 24.251 108 6 1951 5561 11.196 110 7 1994 18309 18.523 120 2 2174 26989 19.444 139 3 1964 19605 14.404 145 8 1894 15378 14.126 147 4 1890 12283 12.664 153 9 1820 5765 9.366 154 10 1831 8305 10.592 159 11 1932 20801 16.797 164 12 1777 18463 18.102 167 13 1779 14613 15.856 169 14 1790 9450 12.882 173 15 1731 12780 14.590 179 5 1639 16978 16.666 Notes:

a. See Figure 3-2

b. Included Pressure Pipe Geometry and Loading January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-9 Unit 1 Location 1 ooa = 34.06 in ta = 2.43 in Normal Loadsa Faulted Loadsb Forcec: 1556 kips Force 0 : 2039 kips Bending Moment: 10616 in-kips Bending Moment: 39247 in-kips a See Table 3-1 b See Table 3-3 c Includes the force due to a pressure of 2259 psig Figure 3-1 Hot Leg Coolant Pipe Pipe Geometry and Loading January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-10 Renctor Pressure Vessel

?2 COLO LEG

'\__Reoctor Coolant Pump Steam Generator ~ Location 2

- Unit 1: Alloy 82/182 weld with SWOL

- Unit 2: Alloy 821182 weld with weld inlay Location 3

- Unit 1: Alloy 82/182 weld

- Unit 2: Alloy 82/182 weld with weld inlay t CROSSOVER LEG Hot leg: Temperature 622"F; Pressure 2259 psig Crossover Leg: Temperature 556°F; Pressure 2221 psig Cold Leg: Temperature 556°F; Pressure 2305 psig Figure 3-2 Schematic Diagram of North Anna Units 1 and 2 Primary Loop Showing Weld Locations Pipe Geometry and Loading January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-1 4.0 MATERIAL CHARACTERIZATION 4.1 PRIMARY LOOP PIPE AND FITTINGS MATERIALS The primary loop pipe is A351-CF8A and the elbow fittings are A351-CF8M for North Anna Units 1 and 2.

4.2 TENSILE PROPERTIES The Certified Materials Test Reports (CMTRs) for the North Anna Units 1 and 2 Reactor Coolant Loop Lines are used to establish the tensile properties for the Leak-Before-Break analyses. For the RCL Lines, Tables 4-1 and 4-2 provide the tensile properties for Units 1 and 2 pipes (A351-CF8A);

Tables 4-3 and 4-4 provide the tensile properties for Units 1 and 2 elbows (A351-CF8M).

For the A351-CF8A and A351-CF8M material, the representative properties at operating temperatures (622°F for Hot Leg and 556°F for Crossover Leg and Cold Leg) are established from the minimum and average tensile properties at temperature of 650°F, given in Table 4-5, by utilizing Section II of the 2007 ASME Boiler and Pressure Vessel Code (Reference 4-1). Code tensile properties at temperatures for the operating conditions considered in this LBB analysis are obtained by linear interpolation of tensile properties provided in _the Code. Ratios of the Code tensile properties at the operating temperatures to the corresponding properties at the CMTR temperature are then applied to the minimum and average tensile properties obtained from CMTRs (Table 4-5) to obtain the North Anna Units 1 and 2 line-specific properties at operating temperatures. For material heats* where CMTR data is not available

• for the 650°F test temperature, material interpolations are based on the CMTR properties taken at room temperature. It should be noted that there is no significant impact. by using the 2007 ASME Code Section II edition for material properties for the LBB analysis, as compared to the North Anna ASME Code of record.

Material modulus of elasticity is also interpolated from ASME Code values for the operating temperatures considered, and Poisson's ratio is taken as 0.3. The temperature-dependent material properties from the ASME Code are shown in Table 4-6. The average and lower bound yield strengths, ultimate strengths, and elastic moduli 'for the pipe material at applicable operating temperatures are tabulated in Table 4-7. The operating temperatures represent the applicable temperatures that include the design basis operating temperatures, uprating and Measurement Uncertainty Recapture (MUR) operating temperatures.

For the SLR program that accounts for 80 years of plant operation, the materials will thermally age.

The aged tensile (Sy and Su) properties, calculated per Reference 4-2, are higher than the unaged properties as demonstrated in the calculations for CF8A and CF8M. Since lower tensile properties are more conservative for the LBB evaluation by potentially reducing the critical flaw size/flaw size margins; therefore, for conservatism, the unaged tensile properties as shown in Table 4-7 are used in the LBB evaluation.

Material Characterization January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-2 4.3 FRACTURE TOUGHNESS PROPERTIES The pre-service fracture toughness (J) of cast stainless steels that are of interest are in terms of J,c (J at Crack Initiation) and have been found to be very high at 600°F. [

Ja,c,e However, cast stainless steel is susceptible to thermal aging at the reactor operating temperature, that is, about 290°C (550°F). Thermal aging of cast stainless steel results in embrittlemert, that is, a decrease in the ductility, impact strength, and fracture toughness of the material. Depending on the material composition, the Charpy impact energy of a cast stainless steel component could decrease to a small fraction of its original value after exposure to reactor temperatures during service.

In 1994, the Argonne National Laboratory (ANL) completed an extensive research program in assessing the extent of thermal aging of cast stainless steel materials. The ANL research program measured mechanical properties of cast stainless steel materials after they had been heated in controlled ovens for long periods of time. ANL compiled a data base, both from data within ANL and from international sources,

• of about 85 compositions of cast stainless steel exposed to a temperature range of 290-400°C (550-750°F) for up to 58,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (6.5 years). In 2015 the work done by ANL was augmented, and the fracture toughness database for CASS materials was aged to 100,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> at 290-350°C (554-633°F). The methodology for estimating fracture properties has been extended to cover CASS materials with a ferrite content of up to 40%. From this database (NUREG/CR-4513, Revision 2), ANL developed correlations for estimating the extent of thermal aging of cast stainless steel (Reference 4-2).

• ANL developed the fracture toughness estimation procedures by correlating data in the database conservatively. After developing the correlations, ANL validated the estimation procedures by comparing .the estimated fracture toughness with the measured value for several cast stainless steel plant components removed from actual plant service. The procedure developed by ANL was used to calculate the end of life fracture toughness values for this analysis. The ANL research program was sponsored and the procedure was accepted by the NRC.

The results from the ANL Research Program indicate that the lower-bound fracture toughness of thermally aged cast stainless steel is similar to that of submerged arc welds (SAWs). The applied value of the J-integral for a flaw in the weld regions will be lower than that in the base metal because the yield stress for the weld materials is much higher at the temperature. 1 Therefore, weld regions are less limiting than the cast material.

Based on Reference 4-2, the fracture toughness correlations used for the full aged condition is applicable for plants operating at ~30 EFPY (Effective Full Power Years) for the CF8A materials and

~15 EFPY for the CF8M materials. As of June 2018, North Anna Units 1 and 2 are operating at 32.1 and 30.5 EFPY, respectively. For the SLR program that accounts for 80 years of plant operation, the materials will thermally age. Therefore, the use of the fracture toughness correlations described in the following sections is applicable for the fully aged or saturated condition of the North Anna Units 1 and 2 materials made of CFBA for pipes and CFBM elbow fittings.

1 In the report all the applied J values were conservatively determined by using base metal strength properties.

Material Characterization January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-3 4.3.1 Fracture Toughness Properties of Centrifugal Cast Pipe (CF8A)

The chemical compositions of the North Anna Units 1 and 2 primary loop piping material are available from CMTRs. The following equations 4-1 to 4-3 for delta ferrite calculations are taken from Reference 4-2 and applicable for CF8A type materials.

Creq =Cr+ 1.21 (Mo)+ 0.48(Si) - 4.99 = (Chromium equivalent) (4-1)

Nieq = (Ni) + 0.11 (Mn) - 0.0086(Mn)2 + 18.4(N) + 24.S(C) + 2.77 = (Nickel equivalent) (4-2)

Note: N is not included in CMTR. Value of 0.04 is assumed per Reference 4-2.

Oc = 100.3 (Creq / Nieq)2-170.72(Creq / Nieq)+74.22 (4-3)

Note: for material heats 143325 and 147377, %Mo data is not available, but delta ferrite values are given in CMTR. However, even if %Mo of 0.5 is assumed, the calculated delta ferrite will be similar to delta ferrite values from CMTR. Therefore, 8c CMTR values are used in lieu of 8c (Eq. 4-3) for conservatism.

The elements are in percent weight and 8c is ferrite in percent volume.

The saturation room temperature (RT at 77°F) impact energies of the cast stainless steel materials are determined from the chemical compositions available from CMTRs and shown in Table 4-8.

For CF8A, the saturation value of RT impact energy Cvsat (J/cm 2) is the lower value determined from log10CVsat = 1.15 + 1.36 exp (-0.035<!>) (4-4) where the material parameter 4> is expressed as 4> = 8c (Cr + Si)(C + 0.4N) (4-5) and from log10CVsat = 5.64 - 0.0068c- 0.185Cr + 0.273Mo - 0.204Si+ 0.044Ni - 2.12(C + 0.4N) (4-6)

Note, that the calculated Cvsat value is conservatively used to represent Cv value in the following equations.

The J-R curve at RT, for centrifugal-cast CF8A steel is given by Jd = 57 (Cv) 0*52 (Lla)n (4-7) n = 0.18 + 0.10 10910 (Cv) (4-8) where Jd is the "deformation J" in kJ/m 2 and ,1.a is the crack extension in mm.

The J-R curve at 290-320°C (554-608°F), for centrifugal-cast CF8A steel is given by Jd = 134 (Cv) 0*28 (Lla)n (4-9) n = 0.17 + 0.0910910 (Cv) (4-10) where Jd is the "deformation J" in kJ/m 2 and ,1.a is the crack extension in mm. Note that the operating temperatures used in the LBB evaluations are 622°F for Hot Leg and 556°F (for Material Characterization January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-4 Crossover Leg and Cold Leg). Therefore, to obtain values at operating temperatures, data interpolation for 556°F and extrapolation for 622°F are performed using values calculated at RT condition (77°F) and hot condition (554-608°F). For hot condition, temperature upper bound (608°F) and temperature lower bound {554°F) are considered in the data interpolation/extrapolation processes, and then the minimum calculated value is conservatively selected.

J1c and Jmax Calculations:

T mat Calculations:

The material shearing modulus, Tmat, is calculated as follows:

T mat = dJ/da X E/(CTta) 2 Japp and Tapp Calculations:

The critical heats for CF8A with lowest fracture toughness property and lowest tearing modulus values from Table 4-9 on Cold Leg, Hot Leg, and Crossover Leg are summarized in Table 4-10.

Notes: U1 = Unit 1; U2 = Unit 2, CL= Cold Leg, HL = Hot Leg, XL= Crossover Leg.

The applied J Integral value, Japp, is calculated and compared to the J,c and Jmax values in Table 7-1 for Unit 1 and Table 7-2 for Unit 2.

• 4.3.2 Fracture Toughness Properties of Static Cast Elbows (CFBM)

The susceptibility of the material to thermal aging increases with increasing ferrite contents, and the molybdenum bearing CF8M shows increased susceptibility to thermal aging.

The chemical compositions of the North Anna Units 1 and 2 primary loop elbow fitting material are available from CMTRs. Note that the CMTRs for the cast elbow components contain two sets of chemistry composition data; labeled as Ladle and Check values. [

]a,c,e For the North Anna cast elbow heats, both the Ladle and Check properties meet the chemical composition requirements for ASTM A351 standards for grade CF8M. For conservatism in the LBB analysis, both the Ladle and Check properties are considered in the calculation of thermally aged fracture toughness properties. The following equations 4-11 to 4-13 for delta ferrite calculations are taken from Reference 4-2 and applicable for CF8M type materials.

Creq =Cr+ 1.21(Mo) + 0.48(Si)-4.99 = (Chromium equivalent) (4-11)

Nieq =(Ni)+ 0.11 (Mn) - 0.0086(Mn)2 + 18.4(N) + 24.5(C) + 2.77 = (Nickel equivalent) (4-12)

Note: N is not included in CMTR. Value of 0.04 is assumed per Reference 4-2.

Material Characterization January 2020 WCAP-11164-N P Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-5 6c =1OO.3(Creq / Nieq) 2 -17O.72(Creq / Nieq) + 74.22 (4-13) where the elements are in percent weight and 6c is ferrite in percent volume.

The saturation room temperature (RT at 77°F) impact energies of the cast stainless steel materials are determined from the chemical compositions available from CMTRs and shown in Table 4-11.

For CF8M steel with < 10% Ni, the saturation value of RT impact energy CVsat (J/cm2) is the lower value determined from log10CVsat = 0.27 + 2.81 exp (:-O.O22(jl) (4-14) where the material parameter $ is expressed as

~ = 6c (Ni + Si +Mn) 2(C + O.4N)/5.O (4-15) and from log10CVsat = 7.28 - O.0116c - O.185Cr - O.369Mo - 0.451 Si- O.OO7Ni - 4. 71 (C + OAN) (4-16)

For CF8M steel with ~ 10% Ni, the saturation value of RT impact energy Cvsat (J/cm2) is the lower value determined from log10CVsat = 0.84 + 2.54 exp (-0.047~) (4-17) where the material parameter ~ is expressed as

~ = 6c (Ni + Si +Mn}2(C + O.4N)/5.O (4-18) and from log10CVsat = 7.28 - O.O116c - O.185Cr - O.369Mo - O.451Si - O.OO7Ni -4.71 (C + O.4N) (4-19)

Note, that the calculated Cvsat value is conservatively used to represent Cv value in the following equations.

The J-R curve at RT, for static-cast CFBM steel is given by (4.:20)

Jd = 1.44 (Cv) 1*35 (Aa)n for Cv < 35 J/cm2 (4-21) n = 0.20 + 0.08 10910 (Cv) (4-22) where Jd is the "deformation J" in kJ/m 2 and Aa is the crack extension in mm.

The J-R curve at 29O-32O°C (554-6O8°F), for static-cast CF8M steel is given by Jd = 49 (Cv) 0*41 (Lla)n for Cv ~ 46-,J/cm2 (4-23)

Jd = 5.5 (Cv) 0*98 (Lla)n for Cv < 46 J/cm2 (4-24) n = 0.19 + 0.07 10910 (Cv) (4-25) where Jd is the "deformation J" in kJ/m 2 and .:la is the crack extension in mm. Note that the operating temperatures used in the LBB evaluations are 622°F for Hot Leg and 556°F (for Crossover Leg and Cold Leg). Therefore, to obtain values at operating temperatures, data interpolation for 556°F and extrapolation for 622°F are performed using values calculated at RT condition (77°F) and hot condition (554-6O8°F). For hot condition, temperature upper bound (6O8°F)

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-6 and temperature lower bound (554°F) are considered in the data interpolation/extrapolation processes, and then the minimum calculated value is conservatively selected.

J1c and Jmax Calculations:

]a,c,e T mat Calculations:

The material shearing modulus, Tmat, is calculated as follows:

T mat = dJ/da X E/(crta}2 Japp and Tapp Calculations:

The critical heats for CF8M with lowest fracture toughness property and lowest tearing modulus values from Table 4-12 on each Cold Leg, Hot Leg, and Crossover Leg are summarized in Table 4-

13. Notes: U1 = Unit 1; U2 = Unit 2, CL= Cold Leg, HL:::: Hot Leg, XL= Crossover Leg.

The applied J Integral value, Japp, is calculated and compared to the J1c and Jmax values in Table 7-1 for Unit 1 and Table 7-2 for Unit 2.

4.3.3 Consideration of Dissimilar Metal Weld Material Profiles For local and global failure mechanisms, all locations are evaluated using the cast stainless steel material properties (A351-CF8A and A351-CF8M) as shown in Tables 4-8 to 4-12 which present a limiting condition due to the thermal aging effects. As stated in Reference 4-3, "The fracture resistance of Alloy 82 and 52 welds have been investigated by conducting fracture toughness J-R curve tests at 24-338°C in deionized water [ ... ]. The results indicate that these welds exhibit high fracture toughness in air and high-temperature water

(>93°C)."

• Since nickel Alloys are known to have high toughness properties and because the CASS base metal of the RCL piping and elbows are susceptible to thermal aging degradation of the fracture toughness, it is determined that the CASS base metal presents the most limiting condition. The fracture mechanics evaluation of the Unit 2 DMW locations with 52/152 weld inlay (SGIN's and SGON's) and the Unit 1 DMW locations at the SGON's without PWSCC mitigation ar'9 bounded by the A351 base metal evaluations at these locations. The Unit 1 DMW locations at the SGIN's with SWOL are addressed separately from this report, in Reference 4-4.

As an additional defense; for the Unit 1 DMW locations at the SGON's without PWSCC mitigation, the evaluation additionally considers Alloy 82/182 material properties. For the mitigated locations due to the applications of full weld overlay program for Unit 1 SGIN's and weld inlay program for Unit 2 SGIN's and SGON's, the evaluations additionally considers both Alloys 52/152 and 82/182 properties (Table 4-7) as part of the 80 year plant life period in SLR program.

Material Characterization January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-7

4.4 REFERENCES

4-1 ASME Boiler and Pressure Vessel Code Section II, 2007 Edition through 2008 Addenda.

4-2 0. K. Chopra, "Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems," NUREG/CR-4513, Revision 2, U.S. Nuclear Regulatory Commission, Washington, DC, May 2016.

4-3 NUREG/CR-6721, "Effects of Alloy Chemistry, Cold Work, and Water Chemistry on Corrosion Fatigue and Stress Corrosion Cracking of Nickel Alloys and Welds," date published: April 2001.

4-4 Structural Integrity Associates, Inc. Letter, "Leak Before Break (LBB) Update for North Anna Unit 1 Steam Generator Inlet Nozzle Weld Overlay (WOL) for Subsequent License Renewal," July 10, 2019.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-8 Table 4-1 Measured Tensile Properties for North Anna Unit 1 Primary Loop Pipes (A351-CF8A)

At Room Temperature At 650°F Location Heat No.

(Leg Loop) /Serial No. Yield Ultimate Yield Ultimate Strength Strength Strength Strength (psi) (psi) (psi) (psi)

B-2356A 39450 81100 23400 60000 CL Loop 1 B-2620 40950 79900 24800 67000 143325NRA-1 50100 84250 24450 N/A B-2608 43950 84700 23700 64500 CL Loop 2 B-2659 39800 77611 20000 58500 143325NRA-2 50100 84250 24450 N/A B-2548 35000 78300 21500 63000 CL Loop 3 B-2715 43000 83283 21000 65000 143325NRA-3 50100 84250 24450 N/A HL Loop 1 B-2744 36960 79020 23300 64000 "HL Loop 2 B-2764 41958 82815 22200 64500 HL Loop 3 B-2641 39950 82900 20300 62500 C-1153A 39460 83310 23700 65750 XL Loop 1 C-1037C 43000 85700 27400 66000 C-1037A 43000 85700 27400 66000 XL Loop 2 C-10378 43000 85700 27400 66000 C-1632A 39960 83100 24200 66000 XL Loop 3 C-11538 39460 83310 23700 66000 Material Characterization January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-9 Table 4-2 Measured Tensile Properties for North Anna Unit 2 Primary Loop Pipes (A351-CF8A)

At Room Temperature At 650°F Location Heat No.

(Leg Loop) /Serial No. Yield Ultimate Yield Ultimate Strength Strength Strength Strength (psi) (psi) (psi) (psi)

C2205-A 44955 79320 22700 64000 CL Loop 1 147377-3 51380 83400 25650 N/A C2230 46200 82500 23700 66250 C2291-B 37460 80920 23100 65000 CL Loop 2 147377-2 51380 83400 25650 N/A C2246 39960 83920 26400 66250 C-2116 42460 81820 22100 63000 CL Loop 3 C 2152 35090 79200 19950. 62250 147377-1 51380 83400 25650 N/A HL Loop 1 C-1638 44455 84815 22300 64000 HL Loop 2

• C-1751 42600 83750 25000 67000 HL Loop 3 C-1804 41960 80920 25500 64500 C-1632-8 39960 83100 24200 66000 XL Loop 1 C-1153-C 39460 83310 23700 65750 C1767-A 43800 83750 23100 63750 XL Loop 2 C1767-B 43800 83750 23100 63750 C2103-A 40960 82320 23400 65000 XL Loop 3 C1767-C 43800 83750 23100 63750 Material Characterization January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-10 Table 4-3 Measured Tensile Properties for North Anna Unit 1 Primary Loop Elbows (A351-CF8M)

At Room Temperature At 650°F Location Heat No./

(Leg Loop) Serial No. Yield Ultimate Yield Ultimate Strength Strength Strength Strength (psi) (psi) (psi) (psi)

CL Loop 18 56818/1 41600 74700 24700 53900 CL Loop 28 56824/2 50500 76300 31400 66200 CL Loop 38 56844/3 46100 76300 25200 56100 HL Loop 1b 58071/51 50500 79700 NIA N/A HL Loop 2b 58090/52 52800 83100 NIA N/A HL Loop 3b 58118/53 47200 76900 N/A N/A XL Loop 1c 57528/9 42100 74100 24700 56100 XL Loop 2c

• 57528/10 42100 75800 24700 58400 XL Loop 3c 57604/11 39900 72400 23600 56100 XL Loop 1d 57412/9 46000 76300 25800 59500 XL Loop2d 57452/10 42700 75200 24700 58400 XL Loop 3c1 57604/11 39900 72400 23600 56100 XL Loop 18 K-0160/95 47200 80300 33700 64000 XL Loop 2e K-0169/96 48800 80300 30300 64000 XL Loop 3e K-0222/97 41100 74700 24700 57300 Note: a = 27.5-inch ID elbow - 35° b = 29-inch x 31-inch ID elbow- 50° c = 31-inch ID elbow-40° d = 31-inch ID elbow- 90° e = 31-inch ID elbow 90° wt splitter Material Characterization January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-11 Table 4-4 Measured Tensile Properties for North Anna Unit 2 Primary Loop Elbows {A351-CF8M}

At Room Temperature At 650°F Location Heat No./

(Leg Loop) Serial No. Yield Ultimate Yield Ultimate Strength Strength Strength Strength (psi) (psi) fpsi) (psi)

CL Loop 18 57663/57 54500 84800 N/A NIA CL Loop 28 57731/58 51600 81400 N/A NIA CL Loop 3° 57801/59 48300 79700 N/A N/A HL Loop 1b 58148/78 51600 85300 N/A N/A HL Loop 2b 58148/79 51600 85300 N/A N/A HL Loop 3b 58159180 46100 83600 NIA NIA XL Loop 1c 58090/71 52800 83100 N/A N/A XL Loop 2c 58118/72 47200 76900 N/A NIA XL Loop 3c 58118/73 47200 76900 N/A NIA XL Loop 1d 58037189 44900 76900 NIA N/A XL Loop2d 58071190 50500 79700 NIA N/A XL Loop 3d 58009188 47200 78000 N/A NIA XL Loop 18 K0134/107 50000 80300 31400 62900 XL Loop 2e K-0399/106 40400 72400 25800 55600 XL Loop 3e K-0876/105 41000 74700 21600 55000 Note: a = 27.5-inch ID elbow - 35° b 2g-inch x 31-inch ID elbow- 50° c = 31-inch ID elbow- 40° d = 31-inch ID elbow- go 0 0

e = 31-inch ID elbow- go w/ splitter Material Characterization January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-12 Table 4-5 North Anna Units 1 and 2 Measured Tensile Properties for A351-CF8A and A351-CF8M At 650°F Critical Location Material Heats#

Material Table 4-1 to 4-4 Yield Strength (psi) Ultimate Strength (psi) 1 CF8A B-2641 20300 62500 2 CF8M 58159/80 & 58118/53 46100 (70°F}* 76900 (70°F)*

3 CF8M 57604/11 23600 56100 4 CF8A C1767-A, -B, & -C 23100 63750 4 CF8M 57604/11 23600 56100 11 CF8M K-0876/105 21600 55000 12 CF8A C 2152 & B-2659 19950 58500 5 CF8M 56818/1 24700 53900 5 CF8M 57801/59 48300 (70°F}* 79700 (70°F}*

Note:

• CMTR data at 650°F is not available.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-13 Table 4-6 ASME Code Tensile Properties for Material A351-CF8A and A351-CF8M CF8A CF8M Temperature (OF) Yield Ultimate Elastic Yield Ultimate Elastic Strength Strength Modulus Strength Strength Modulus (ksi) (ksi) (ksi) (ksi) (ksi) (ksi) 70 35.00 77 28300 30.00 70 28300 100 35.00 30.00 150 31.20 27.30 200 29.20 72.9 27500 25.80 70 27500 250 27.50 24.50 300 68.0 27000 23.30 68 27000 400 24.10 65.7 26400 21.40 67.2 26400 500 22.60 65.1 25900 19.90 67.2 25900 556 8 21.98 65.1 25564 19.28 67.2 25564 600 21.50 65.1 25300 18.80 67.2 25300 21.28 65.1 25190 18.62 67.2 25190 650 21.00 65.1 18.40 67.2 700 20.50 65.1 24800 18.10 67.2 24800 Notes: a. XL, CL operating temperature, value is interpolated.

b. HL operating temperature, value is interpolated.

c. Material properties are from the 2007 Edition of the ASME Boiler and Pressure Vessel Code (Reference 4-1 ).

d. Shaded cells are based on linear interpolation of the values provided in the ASME Code Material Characterization January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-14 Table 4-7 Material Properties for Operating Temperature Conditions on North Anna Units 1 and 2 RCL Lines Lower Bound Modulus of Average Yield Temperature Elasticity Strength Ultimate Material (OF) Yield Stress (psi) (psi) Strength (psi)

(psi) 1 A351-CF8A 622 25190000 24263 22430 62500 2 A351-CF8M 622 25190000 31019 28619 73824 3, A351-CF8M 556 25564000 28521 24734 56100 4A A351-CF8A. 556 25564000 26529 24182 63750 4M A351-CF8M 556 25564000 26529 24734 56100 11 A351-CF8M 556 *25564000 29258 22638 55000 12 A351-CF8A 556 25564000 24581 20885 58500 5 (Unit 1) A351-.CF8M 556 25564000 28402 25887 53900 5 (Unit 2) A351-CF8M 556 25564000 33083 31047 76512 SGON 82/182 Alloy

• 556 28876000 49964 49964 84616 SGIN 82/182 Alloy 622 .' 28612000 49656 49656 83692 SGON 52/152 Alloy 556 ,28188000 49964 49964 84616

• SGIN 52/152 Alloy 622 28012000 49656 49656 83692 Notes:

(1) As identified in Section 4.2, the normal operating temperature of the hot leg piping is 622°F and the normal operating temperature of the Cold Leg and Crossover Leg piping is 556°F.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-15 a,c,e Material Characterization January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-16 a,c,e Material Characterization January 2020 WCAP-11164-NP Revision 2

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-)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-17 a,c,e Material Characterization January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-18 a,c,e Material Characterization January 2020 WCAP-11164-NP Revision 2 I

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-19 a,c,e Material Characterization January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-20 a,c,e Material Characterization January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-21 a,c,e Material Characterization January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-22 a,c,e Note: As discussed in Section 4.3.3, the cast stainless steel fracture toughness properties present a limiting condition when compared to the Alloys 82/182 and 52/152 fracture toughness properties found in Units 1 and 2 SG (inlet/outlet) nozzle welds.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-23 a,c,e Figure 4-1 Pre-Service J vs . .1\a for SA351-CF8M Cast Stainless Steel at 600°F Material Characterization January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-1 5.0 CRITICAL LOCATION AND EVALUATION CRITERIA 5.1 CRITICAL LOCATIONS The governing or critical locations for the LBB evaluation are established based on the fracture toughness properties of the metal base at the weld points and also on the basis of pipe geometry, welding process, operating temperature, operating pressure, and the highest faulted stresses at the welds.

The RCL weld points applicable for North Anna Unit 1 and 2 are shown in Figure 3-2. However, the critical locations for LBB analysis are shown i.n Figure 5-1. Note that for the LBS analysis, stresses at weld points 6 and 7 (hot leg) and weld points 13 and 14 (cold leg), where the loop isolation valves are connected, are enveloped by stresses at points 1 and 12, respectively. The by-pass line between the hot leg and cold leg isolation valves is not within the current LBB scope.

Since critical locations correspond with fracture toughness properties of the metal base at the weld points and also the maximum faulted stress locations, the same critical locations (weld points 1, 2, 3, 4, and 5) per References 5-1 and 5-2 are re-evaluated. In this report, additional critical locations (at the RCP inlet and outlet nozzles, which are weld points 11 and 12, respectively) are also evaluated to ensure that every high stress locations in the RCL lines are covered in the evaluation. It is noted, that weld points 1, 2, 3, 11, 12, and 5 are all at the equipment nozzle end weld points. Some of those points have low toughness properties (weld points 2, 3, 11, and 5 with base metal of CF8M). Weld point 4 also shows low toughness properties due to base metal with material type of CF8M at the weld connection. Weld point 1 at the hot leg and* point 12 at the cold leg have high stresses and have the lowest fracture toughness properties among CF8A piping component in the RCL lines.

For LBB evaluation, Table 5-1 and Table 5-2 show the critical locations for North Anna Unit 1 and Unit 2, respectively. Figure 5-1 *shows the locations of the critical welds for North Anna Units 1 and 2.

As noted above, SG nozzle weld points 2 and 3, with base metal of CF8M, have low toughness properties. Section 4.3.3 identifies that the CF8M base metal fracture toughness is more limiting than the Alloy 82/182 DMWs with Alloy 52/152 inlay for the Unit 2 SGIN and SGON locations (weld points 2 and 3). Since the Unit 1 SGON locations (weld point 3) are not mitigated for PWSCC effects, an additional evaluation is performed for this weld considering the Alloy 82/182 material properties and the PWSCC crack morphology.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-2 5.2 EVALUATION CRITERIA As will be discussed later, fracture mechanics analyses are made based on local failure mechanism as described in Section 7.1, and based on global failure mechanism as described in Section 7.2.

For local failure mechanism, stability analysis is performed using J-integral evaluation method with the criteria as follows:

(1) If Japp< J1c, then the crack will not initiate and the crack is stable; (2) If Japp ~ J1c; and Tapp < T mat and Japp < Jmax, then the crack is stable.

Where:

Japp = Applied J J1c = J at Crack Initiation Tapp = Applied Tearing Modulus Tmat = Material Tearing Modulus Jmax = Maximum J value of the material For global failure mechanism, the stability analysis is performed using limit load method based on loads and postulated flaw sizes related to leakage, with the criteria as follows:

• Margin of 10 on the Leak Rate

• Margin of 2.0 on Flaw Size

• Margin of 1.0 on Loads (using the absolute summation method for faulted load combination).

5.3 REFERENCES

5-1 WCAP-11163, Rev. 0, "Technical Bases for Eliminating Large Primary Loop Pipe Rupture as a Structural Design Basis for I North Anna Units 1 & 2," August 1986.

5-2 WCAP-11163, Supplement 1, Rev. 0, "Additional Information in Support of the Technical*,

Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for North Anna Units 1 and 2," January 1988.

Critical Location and Evaluation Criteria January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-3 Table 5-1 Critical Analysis Locations for Leak-Before-Break of North Anna Unit 1 RCL Lines Operating Operating Maximum Node Weld Do Pipe Thickness Welding Pressure Temperature Faulted Stress Point Point (in) (in) Process (OF)

(psi) (psi) 103 1 34.06 2.43 SMAW 2259 622 30.447 120 2 36.60 2.70 SMAW 2259 622 23.481 139 3 36.96 2.88 SMAW 2221 556 16.797 147 4 36.22 2.58 SAW 2221 556 13.757 159 11 36.22 2.58 SMAW 2221 556 17.955 164 12 32.82 2.56 SMAW 2305 556 19.811 179 5 32.82 2.56 SMAW 2305 556 17.980 Table 5-2 Critical Analysis Locations for Leak-Before-Break of North Anna Unit 2 RCL Lines Operating Operating Maximum Node Weld Do Pipe Thickness Welding Pressure Temperature Faulted Stress Point Point (in) (in) Process (OF)

(psi) (psi) 103 1 34.06 2.43 SMAW 2259 622 24.251 120 2 36.60 2.70 SMAW 2259 622 19.444 139 3 36.96 2.88 SMAW 2221 556 14.404 147 4 36.22 2.58 SAW 2221 556 12.664 159 11 36.22 2.58 SMAW 2221 556 16.797 164 12 32.82 2.56 SMAW 2305 556 18.102 179 5 32.82 2.56 SMAW 2305 556 16.666 Critical Location and Evaluation Criteria January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-4 HOT LEG

\__Reactor ~oolant Pump Steam Gencraio, Location! _

• Unit 1: Alloy 821182 weld with SWOL

, Unit 2: Alloy 821182 weld with weld Inlay l.ocalion3 *

~ Urdt 1: Al~y ,821182 WJlld

• Unit 2: Alloy 82/182 weld with weld il'!ll!Y CROSSOVER LEG Figure 5-1 Schematic Diagram of North Anna Units 1 and 2 Primary Loop Showing Critical Weld Locations Critical Location and Evaluation Criteria January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-1 6.0 LEAK RATE PREDICTIONS

6.1 INTRODUCTION

The purpose of this section is to discuss the method which is used to predict the floY" through postulated through-wall cracks and present the leak rate calculation results for through-wall circumferential cracks.

6.2 GENERAL CONSIDERATIONS The flow of hot pressurized water through an opening to a lower back pressure causes flashing which can result in choking. For long channels where the ratio of the channel length, ~. to hydraulic diameter, DH, (L/DH) is greater than [

6.3 CALCULATION METHOD The basic method used in the leak rate calculations is the method developed by [

The flow rate through a crack was calculated in the following manner. Figure 6-1 from Reference 6-2 was used to estimate the critical pressure, Pc, for the primary loop enthalpy condition and an assumed flow. Once Pc was found for a given mass flow, the [

]a,c,e was found from Figure 6-2 (taken from Reference 6-2). For all cases considered, since [ ]0 ,c,e Therefore, this method will yield the two-phase pressure drop due to momentum effects as illustrated in Figure 6-3, where Po is the operating pressure. Now using the assumed flow rate, G, the frictional pressure drop can be calculated using (6-1) where the friction factor f; is determined using the [ ]a,c,e The crack relative roughness, i::, was obtained from fatigue crack data on stainless steel samples. The relative roughness value used in these calculations was [ ]a,c,e The frictional pressure drop using equation 6-1 is then calculated for the assumed flow rate and added to the [ ]a,c,e to obtain the total pressure drop from the primary system to the atmosphere. That is, for the primary loop:

Absolute Pressure - 14. 7 = [ (6-2)

Leak Rate Predictions January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-2 for a given assumed flow rate G. If the right-hand side of equation 6-2 does not agree with the pressure difference between the primary loop and the atmosphere, then the procedure is repeated until equation 6-2 is satisfied to within an acceptable tolerance which in turn leads to flow rate value for a given crack size.

6.4 LEAK RATE CALCULATIONS Leak rate calculations were made as a function of crack length at the governing locations previously identified in Section 5.1. The normal operating loads of Tables 3-1 and 3-2 were applied in these calculations. The crack opening areas were estimated using the method of Reference 6-3, and the leak rates were calculated using the two-phase flow formulation described in the preceding section. The average material properties of Section 4.0 (see Table 4-7) were used for these calculations.

The flaw sizes to yield a leak rate of 10 gpm were calculated at the governing locations and are given in Table 6-1 for North Anna Unit 1 and in Table 6-2 for North Anna Unit 2. The flaw sizes so determined are called leakage flaw sizes.

The North Anna Units 1 and 2 RCS pressure boundary leak detection system meets the intent of Regulatory Guide 1.45, and the plant leak detection capability is 1 gpm. Thus, to satisfy the margin of 1O on the leak rate, the flaw sizes (leakage flaw sizes) are determined which yield a leak rate of 1O gpm.

6.5 REFERENCES

6-1 6-2 M. M, EI-Wakil, "Nuclear Heat Transport, International Textbook Company," New York, N.Y, 1971.

6-3 Tada, H., "The Effects of Shell Corrections on Stress Intensity Factors and the Crack Opening Area of Circumferential and a Longitudinal Through-Crack in a Pipe,"

Section 11-1, NUREG/CR-3464, September 1983.

Leak Rate Predictions January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3 Table 6-1 Flaw Sizes for North Anna Unit 1 Yielding a Leak Rate of 10 gpm for the RCL Lines Leakage Flaw Size Node Point Weld Point (in) 103 1 5.51 120 2 6.84 139 3 7.19 147 4 6.86 159 11 5.61 164 12 6.96 179 5 6.56 Table 6-2 Flaw Sizes for North Anna Unit 2 Yielding a Leak Rate of 1O gpm for the RCL Lines Leakage Flaw Size Node Point Weld Point (in) 103 1 3.96 120 2 5.69 139 3 7.19 147 4 6.83 159 11 5.68 164 12 . 6.94 179 5 6.81 Note: The flaw size in the Tables 6-1 and 6-2 refers to the flaw length of through-wall circumferential crack.

Leak Rate Predictions January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-4 a,c,e I l STAGNATION ENTHALPY {1()2 BtvJib)

Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures Leak Rate Predictions January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-5 a,c,e LENGTH/OIAMETIER RATIO fUDJ

/

Figure 6-2 [ ]a,c,e Pressure Ratio as a Function of LID Leak Rate Predictions January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-6 a.c.e

[

Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack Leak Rate Predictions January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-1 7.0 FRACTURE MECHANICS EVALUATION 7.1 LOCAL FAILURE MECHANISM The local mechanism of failure is primarily dominated by the crack tip behavior in terms of crack-tip blunting, initiation, extension, and final crack instability. The local stability will be assumed if the crack does not initiate at all. It has been accepted that the initiation toughness measured in terms of J1c from a J-integral resistance curve is a material parameter defining the crack initiation. If, for a given load, the calculated J-integral value is shown to be less than the J1c of the material, then the crack will not initiate. If the initiation criterion is not met, one can calculate the tearing modulus as defined by the following relation:

(7-1) where:

Tapp = applied tearing modulus E = modulus of elasticity O"f = 0.5 (cry+ cru) = flow stress a = crack length O"y, O"u = yield and ultimate strength of the material, respectively Stability is said to exist when ductile tearing does not occur if Tapp is less than T mat, the experimentally determined tearing modulus. Since a constant T mat is assumed a further restriction is placed in Japp- Japp must be less than Jmax where Jmax is the maximum value of J for which the experimental T mat is greater than or equal to the Tapp used.

As discussed in Section 5.2 the local crack stability criteria is a two-step process:

(1) If Japp < J1c, then the crack will not initiate and the crack is stable; (2) If Japp~ J1c; and Tapp < T mat and Japp < Jmax, then the crack is stable.

The calculations of Japp and Tapp for the critical locations are performed following the methodology developed in References 7-5 and 7-6. The stability results based on elastic-plastic J-integral evaluations for North Anna Units 1 and 2 are provided in Tables 7-1 and 7-2, respectively.

Fracture Mechanics Evaluation January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-2 7.2 GLOBAL FAILURE MECHANISM Determination of the conditions which lead to failure in stainless steel should be done with plastic fracture methodology because of the large amount of deformation accompanying fracture. One method for predicting the failure of ductile material is the plastic instability method, based on traditional plastic limit load concepts, but accounting for strain hardening and taking into account the presence of a flaw. The flawed pipe is predicted to fail when the remaining net section reaches a stress level at which a plastic hinge is formed. The stress level at which this occurs is termed as the flow stress. The flow stress is generally taken as the average of the yield and ultimate tensile strength of the material at the temperature of interest.

This methodology has been shown to be applicable to ductile piping through a large number of experiments and will be used here to predict the critical flaw size in the primary coolant piping.

The failure criterion has been obtained by requiring equilibrium of the section containing the flaw (Figure 7-1) when loads are applied. The detailed development is provided in Appendix A for a through-wall circumferential flaw in a pipe with internal pressure, axial force, and imposed bending moments. The limit moment for such a pipe is given by:

The analytical model described above accurately* accounts for the piping internal pressure as well as imposed axial force as they affect the limit moment. Good agreement was found between the analytical predictions and the experimental results (Reference 7-1). For application of the limit load methodology, the material, including consideration of the configuration, must have a sufficient ductility and ductile tearing resistance to sustain the limit load.

A stability analysis based on limit load was performed for all the critical locations (locations 1, 2, 3, 4, 11, 12, and 5). The field welds are made of GTAW and SMAW combination weld. The shop welds are made of GTAW, SMAW or SAW combination weld. Field welds are at the critical locations 1, 2, 3, 11, 12, and 5. Shop weld is at critical location 4. The "Z" factor correction for SMAW was applied (References 7-2 and 7-3) at the field weld critical locations (locations 1, 2, 3, Fracture Mechanics Evaluation January 2020 WCAP-11164-NP Revision 2

• • • This record was final approved on 1/17/2020 11 :18:49 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-3 11, 12, and 5) and the "Z" factor correction for SAW was applied (References 7-2 and 7-3) at the shop weld location (location 4) and the equations as follows:

Z = 1.15 [1.0 + 0.013 (OD-4)] for SMAW (7-4)

Z = 1.30 [1.0 + 0.01 (OD-4)] for SAW (7-5) where OD is the outer diameter of the pipe in inches.

The Z-factors were calculated for the critical locations, using the dimensions given in Tables 3-1 and 3-2. The applied loads were increased by the Z factors. Tables 7-3 and 7-4 summarize the results of the stability analyses based on limit load. The leakage flaw sizes are also presented on the same table.

7.3 SGIN AND SGON ALLOY 82/182 WELDS North Anna Units 1 and 2 reactor coolant system primary loop piping contains Alloy 82/182 dissimilar metal welds which are susceptible to PWSCC (Primary Water Stress Corrosion Cracking). The Alloy 82/182 welds are at Units 1 and 2 Steam Generator Inlet and Outlet Nozzles (SGIN's and SGON's). Structural weld overlay repair have been applied for the Unit 1 SGIN's. Leak Before Break evaluation for Unit 1 SGIN's weld overlay for subsequent license renewal is documented in Reference 7-4. As stated in Reference 7-4, "Leak Before Break evaluation (for SG Inlet Nozzle as documented in calculation 11_00226.303 Revision 3), can continue to be used as justification/demonstration of LBB behavior for the weld overlay, for the period of subsequent license renewal (80 years) at North Anna Unit 1."

The Unit 2 SGIN's and SGON's have included an inlay of Alloy 52/152 material to prevent PWSCC. In summary, for Unit 1 SGIN's, Unit 2 SGIN's and SGON's, the potential PWSCC have been mitigated either by weld inlay or weld overlay, therefore no further evaluations are required for those locations for the SLR program (80 year plant service). However, to add the degree of assurance, the presence of nickel base alloy welds are also evaluated by comparison of the material properties to demonstrate that the thermally aged CASS base metal is more limiting.

The Alloy 82/182 weld at the Unit 1 SGON's had been evaluated previously using LBB stability analysis methodology. The conclusion is that postulated cracks including postulated PWSCC are stable, therefore, acceptable. For the SLR program, the previous LBB evaluation is updated to ensure that the conclusions are applicable for 80 years of plant service.

In the LBB evaluation for SLR program, it is noted that Alloy 82/182 weld has high toughness and it does not degrade due to the thermal aging and therefore the LIMIT load method with a 'Z' factor of 1.0 is used to calculate the critical flaw sizes. A conservative factor of 1.69 to account for the PWSCC is applied to the leakage flaw size calculation. These approaches are consistent with the approach used in the previous LBB evaluation. The LBB evaluation results for Unit 1 SGON Alloy 82/182 welds are shown in Table 7-5, and found to be acceptable for SLR program.

In conclusion, the existence of Alloy 82/182 welds at North Anna Units 1 and 2 SGIN's and SGON's are acceptable for the SLR program (for 80 years of plant operation).

Fracture Mechanics Evaluation January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-4

7.4 REFERENCES

7-1 Kanninen, M. F., et. al., "Mechanical Fracture Predictions for Sensitized Stainless Steel Piping with Circumferential Cracks," EPRI NP-192, September 1976.

7-2 Standard Review Plan; Public Comment Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal RegisterNol. 52, No. 167/Friday, August 28, 1987/Notices, pp.

32626-32633.

7-3 NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.

7-4 Structural Integrity Associates, Inc. Letter, "Leak Before Break (LBB) Update for North Anna Unit 1 Steam Generator Inlet Nozzle Weld Overlay (WOL) for Subsequent License Renewal," July 10, 2019.

7-5

]a,c,e 7-6

] a,c,e Fracture Mechanics Evaluation January 2020 WCAP-11164-NP Revision 2

""* This record was final approved on 1/17/2020 11 :18:49 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-5

_a,c,e a,c,e Fracture Mechanics Evaluation January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-6 Table 7-3 Flaw Stability Results for North Anna Unit 1 Yielding a Leak Rate of 10 gpm for the RCL Lines Based on Limit Load Node Point Weld Point Leakage Flaw Size (in) Critical Flaw Size (in) 103 1 5.51 11.13 120 2 6.84 25.94 139 3 7.19 28.41 147 4 6.86 30.40 159 11 5.61 24.74 164 12 6.96 21.27 179 5 6.56 24.06 Note: results are based on the limiting material properties of the CASS base metal.

Table 7-4 Flaw Stability Results for North Anna Unit 2 Yielding a Leak Rate of 10 gpm for the RCL Lines Based on Limit Load Node Point Weld Point Leakage Flaw Size (in) Critical Flaw Size (in) 103 1 3.96 18.24 120 2 5.69 30.79 139 3 7.19 31.93 147 4 6.83 32.06 159 11 5.68 26.38 164 12 6.94 23.48 179 5 6.81 32.95 Note: results are based on the limiting material properties of the CASS base metal.

Fracture Mechanics Evaluation January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-7 Table 7-5 Leakage Flaw Sizes, Critical Flaw Size and Margins for North Anna Unit 1 SGON Alloy 82/182 Welds Flaw SGON Type Leakage Flaw Size (in) Critical Flaw Size (in) Margins Circumferential 19.74 64.46 >2.0 Axial 14.83 67.33 >2.0 Note: results are based on the material properties of the Alloy 82/182 weld metal.

Fracture Mechanics Evaluation January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-8 Neutral Axis Figure 7-1 [ ]a,c,e Stress Distribution Fracture Mechanics Evaluation January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-1 8.0 FATIGUE CRACK GROWTH ANALYSIS To determine the sensitivity of the primary coolant system to the presence of small cracks, a fatigue crack growth analysis was carried out for the [ ]a,c,e region of a typical system (see Location [ ]a,c,e of Figure 5-1). This region was selected because crack growth calculated here will be typical of that in the entire primary loop. Crack growths calculated at other locations can be expected to show minimal variation.

Even-though the North Anna plant does not have lnconel Alloy 600 weld at the vessel inlet nozzle, the lnconel 600 weld exists at the SG inlet and outlet nozzles. From the crack growth

. point of view, the crack growth at the vessel inlet nozzle remains typical. Therefore, the plant typical evaluation at the vessel inlet nozzle with lnconel Alloy 600 is used in this calculation note, to represent the evaluation of FCG at the primary equipment nozzles including at the SG inlet and outlet nozzles. Fatigue crack growth results for the lnconel 182 and lnconel 152 welds are expected to be about the same as lnconel 600 weld.

A[ ]a,c,e of a plant typical in geometry and operational characteristics to any Westinghouse PWR System. [

The normal, upset, and test conditions were considered. Circumferentially oriented surface flaws are postulated in the region, assuming the flaw was located in three different locations, as shown in Figure 8-1. Specifically, these are:

[

Fatigue crack growth rate laws were used [

The law for stairnless steel was derived from Reference 8-j, with a very conservative correction for the R ratio, which is the ratio of minimum to maximum stress during a transient.

For stainless steel, the fatigue crack growth formula is:

da dn

= (5 .4 X 10-12)Keff4.48 (8-1)

Where:

da ctn = crack growth rate (inches/cycles)

Fatigue Crack Growth Analysis January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-2 Keff = Kmax(l.O - R) 0*5 R = Kmin/Kmax

]a,c,e a,c,e (8-2)

The calculated fatigue crack growth for semi-elliptic surface flaws of circumferential orientation and various depths is summarized in Table 8-2, and shows that the crack growth is very small, regardless of which material is assumed.

The reactor vessel transients and cycles for North Anna Units 1 and 2 are shown in Table 8-3. By comparing the transients and cycles for the generic analysis shown in Table 8-1 and the North Anna plant specific transients and cycles shown in Table 8-3, it is concluded that the generic transients and cycles used for the fatigue crack growth analysis enveloped the North Anna transients and cycles. The transients and cycles (shown in Table 8-3) for the North Anna plants for 60 years are the same as those of 40 years, and remain applicable for 80 years of operation as well. Also any changes in the cycles for the 80 year design transients will not have a significant impact on the fatigue crack growth conclusions, since there is insignificant growth of small surface flaws as shown in Table 8-2.

The fatigue crack growth analysis is not a requirement for the LBB analysis (see References 8-4 and 8-5) since the LBB analysis is based on the postulation of through-wall flaws, whereas the FCG analysis is performed based on a surface flaw. In addition Reference 8-6 has indicated that, "the Commission deleted the fatigue crack growth analysis in the proposed rule. This requirement was found to be unnecessary because it was bounded by the crack stability analysis." 1 1 It is therefore, concluded that the generic fatigue crack growth analysis results shown in Table 8-2 is representative of the North Anna plants fatigue crack growth and also applicable for 80 years.

Fatigue Crack Growth Analysis January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-3

8.1 REFERENCES

8-1 Bamford, W. H., "Fatigue Crack Growth of Stainless Steel Piping in a Pressurized Water Reactor Environment," Trans. ASME Journal of Pressure Vessel Technology, Vol. 101, Feb. 1979.

8-2 James, L.A., "Fatigue Crack Propagation Behavior of lnconel 600," in Hanford Engineering Development Labs Report HEDL-TME-76-43, May 1976.

8-3 Hale, D. A., et. al., "Fatigue Crack Growth in Piping and RPV Steels in Simulated BWR Water Environment," Report GEAP 24098/NUREG CR-0390, Jan. 1978.

8-4 Standard Review Plan: Public Comments Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal RegisterNol. 52, No. 167/Friday August 28, 1987/Notices, pp.

32626-32633.

8-5 NUREG-0800, Revision 1, "Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures" March 2007.

8-6 Nuclear Regulatory Commission, 10 CFR 50, Modification of General Design Criteria 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures, Final Rule, Federal RegisterNol. 52, No. 207/Tuesday, October 27, 1987/Rules and Regulations, pp. 41288-41295.

Fatigue Crack Growth Analysis January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-4 Table 8-1 Summary of Transients (Representative 80-Year Design)

Number of Number Typical Transient Identification Cycles Normal Conditions Heatup and Cooldown at 100°F/hr.

1 200 (pressurizer cooldown 200°F/hr.)

Load Follow Cycles 2 18300 (Unit loading and unloading at 5% of full power/min.)

3 Step Load Increase and Decrease 2000 4 Large Step Load Decrease, with Steam Dump 200 5 Steady State Fluctuations 1000000 U~set Conditions 6 Loss of Load, without Immediate Turbine or Reactor Trip 80 Loss of Power 7 40 (blackout with natural circulation in the Reactor Coolant System)

Loss of Flow 8 80 (partial loss of flow, one pump only) 9 Reactor Trip from Full Power 400 Test Conditions 10 Turbine Roll Test 10 11 Hydrostatic Test Conditions Primary Side 5 Primary Side Leak Test 50 12 Cold Hydrostatic Test 10 Fatigue Crack Growth Analysis January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-5 Table 8-2 Typical Fatigue Crack Growth at [ ]a,c,e (40, 60, and 80 years)

Final Flaw Depth (in)

Initial Flaw Depth (in)

[ ]a,c,e [ ]a,c,e [ ]a,c,e 0.292 0.31097 0.30107 0.30698 0.300 0.31949 0.30953 0.31626 0.375 0.39940 0.38948 0.40763 0.425 0.45271 0.44350 0.47421 Table 8-3 Summary of Transients for North Anna Units 1 and 2 (40, 60, 80 years)

Number Transient Name Number of Cycles 1 Heatup/Cooldown 200 2 Pressurizer Cooldown at 200 deg./hr. 200 3 Loading and Unloading 18300 4 10% Step Load Decrease/Increase 2000 5 Large Step Decrease 200 6 Reactor Trip from Full Power 400 7 Loss of Load >15% 80 8 Loss of AC Power 40 9 Loss of Flow 80 10 Inadvertent Spray 10 11 Turbine Roll Test ' 10 12 Hydrostatic Test Primary Side 5 13 Hydrostatic Test Secondary Side 5 14 Primary Side Leak test 50 15 Steady State Fluctuations Infinite Fatigue Crack Growth Analysis January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-6 a,c,e Figure 8-1 Typical Cross-Section of [

Fatigue Crack Growth Analysis January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-7 1000 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

• Linear interpolation is recom*

700 mended to account for ratio dependence of water environment 500 cunies, for 0.25 <R< 0.65 for shallow slope:

da = 11.01 X 10*1 1 02*ll. K 1 -95 dN 02

• 3.75 R +0.06 R

• Kmin /Kmax 200 Subsurface flaws u 100 lair en11ironmend

~ da :: 10.0267 X 10-31 A dN 70

~

• g Determine the ll.K at whicti the 50

~ law changes by calculation of the inter,ection of the 1WO I curws,

"-a e

a:,

a: Surface flaws

.&. (water react(1r I

C, 20 enllironmentl applicable for

""- R-C 0.25 i!! "0.25 < R < 0.65 u

R ;a 0.65 10 R

• Kmin /Kmax 7 "'

0) it

<;) (

5 UJ-

• Linear interpolation is recommended

~

)(

I to account for R ratio dependence ot water environment curws, for

~ I "' )(

0.25 < R < 0.65 for steep slope:

~

I ~ de = (1.02 X 10-61 0 1 ll.K5 -9 5 dN 01 "'26.9R

• 5.725 R ~ Kmin /Kmax 2 5 7 10 20 50 70 100 Stress lntensilv Factor Range IAK1 ksi ..pi,>

Figure 8-2 Reference Fatigue Crack Growth Curves for Carbon & Low Alloy Ferritic Steels Fatigue Crack Growth Analysis January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-8 a,c,e Figure 8-3 Reference Fatigue Crack Growth Law for [ ]a,c,e in a Water Environment at 600°F Fatigue Crack Growth Analysis January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-1 9.0 ASSESSMENT OF MARGINS The results of the leak rates of Section 6.4 and the corresponding stability and fracture toughness evaluations of Sections 7.1 and 7.2 are used in performing the assessment of margins. Margins are shown in Table 9-1 for Unit 1 and Table 9-2 for Unit 2. All of the LBB recommended margins are satisfied. The LBB analyses results are acceptable for the subsequent license renewal program (80 years).

In summary, at all the critical locations relative to:

1. Flaw Size - Using faulted loads obtained by the absolute sum method, a margin of 2 or more exists between the critical flaw and the flaw having a leak rate of 10 gpm (the leakage flaw).

2. Leak Rate - A margin of 1O exists between the calculated leak rate from the leakage flaw and the plant leak detection capability of 1 gpm.

3. Loads - At the critical locations the leakage flaw was shown to be stable using the faulted loads obtained by the absolute sum method (i.e., a flaw twice the leakage flaw size is shown to be stable; hence the leakage flaw size is stable). A margin of 1 on loads using the absolute summation of faulted load combinations is satisfied.

Assessment of Margins January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-2 Table 9-1 Leakage Flaw Sizes, Critical Flaw Sizes and Margins for North Anna Unit 1 Leakage Flaw Size Critical Flaw Size Locationc Margin (in) (in) 5.51 11.13a 2.02 8 1 ,c 5.51 11.Q2b >2b 6.84 25.94a 3_79a 2d 6.84 13.68b >2b 7.19 28.41a 3.958 3

7.19 14.38b >2b 3 (Unit 1)

(Alloy 82/182 with 19.74 64.46 3.27 PWSCC) 6.86 30.40a 4.43a 4

6.86 13.72b >2b 5.61 24.748 4.41 8 11 5.61 11.22b >2b 6.96 21.278 3.06 8 12 6.96 13.92b >2b 6.56 24.068 3.678 5

6.56 13.12b - >2b abased on limit load bbased on J integral evaluation cAII results are based on the limiting material properties of the CASS base metal, except as noted for the location 3 Alloy 82/182 case with PWSCC effects dResults for location 2 are. applicable to the base metal material away from the SWOL geometry.

Complete evaluation of the applied SWOL is detailed in Reference 7-4 Assessment of Margins January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-3 Table 9-2 Leakage Flaw Sizes, Critical Flaw Sizes and Margins for North Anna Unit 2 Leakage Flaw Size Critical Flaw Size Locationc Margin (in) (in) 3.96 18.248 4.6P 1

3.96 7.92b >2b 5.69 30.798 5.41 8 2

5.69 11.38b >2b 7.19 31.938 4.44a 3

7.19 14.38b >2b 6.83 32.068 4.698 4

6.83 13.66b >2b 5.68 26.388 4.648 11 5.68 11.36b >2b 6.94 23.488 3.38 8 12 6.94 13.88b >2b 6.81 32.958 4.848 5

6.81 13.62b >2b abased on limit load bbased on J integral evaluation cAII results are based on the limiting material properties of the CASS base metal Assessment of Margins January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 10-1

10.0 CONCLUSION

S This report justifies the elimination of RCS primary loop pipe breaks from the structural design basis for the 80 year plant life of North Anna Units 1 and 2 as follows:

a. Stress corrosion cracking is precluded by use of fracture resistant materials in the piping system and controls on reactor coolant chemistry, temperature, pressure, and flow during normal operation. Alloy 82/182 welds are present at the North Anna Unit 1 and Unit 2 SGIN's and SGON's. The alloy 82/182 welds are susceptible to PWSCC (Primary Water Stress Corrosion Cracking).

To mitigate PWSCC due to the existence of Alloy 82/182, either weld inlay or weld overlay has been applied to the North Anna Unit 1 SGIN's, Unit 2 SGIN's and SGON's, therefore for those locations no further consideration of PWSCC effects is required for SLR program. Evaluation of the Unit 1 SGON's alloy 82/182 welds is updated for the SLR program, the results are documented in Section 7.3 and found to be acceptable.

b. As stated in Section 3.0, for local and global failure mechanisms, all locations are evaluated using the cast stainless steel material properties (A351-CF8A and A351-CF8M) which present a limiting condition due to the thermal aging effects. The cast stainless steel fracture toughness properties also present a limiting condition when compared to the fracture toughness properties of the Alloy 82/182 and 52/152 dissimilar metal weld materials found in Units 1 and 2 SG (inlet/outlet) nozzle welds.

For the 80 year plant life SLR program, for the unmitigated locations at Unit 1 SG outlet nozzles, the LBB has been reevaluated not only due to thermal aging effects but also by additionally considering alloy 82/182 material properties that includes appropriate PWSCC crack morphology parameter.

For the mitigated dissimilar metal weld locations, including the Unit 1 SGIN's with applications of full weld overlay program and the Unit 2 SGIN's and SGON's with weld inlay, LBB has been reevaluated by additionally considering both Alloys 52/152:and 82/182 and demonstrating that the material properties of the thermally aged CASS base metal are more limiting. Evaluation of the RCS piping considering the thermal aging effects for the 80 year plant life period of the SLR program and also the use of the most limiting fracture toughness properties ensures that each materials profile is appropriately bounded by the LBB results presented in this report.

c. Water hammer should not occur in the RCS piping because of system design, testing, and operational considerations.

d. The effects of low and high cycle fatigue on the integrity of the primary piping are negligible.

Conclusions January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 10-2

e. Ample margin exists between the leak rate of small stable flaws and the capability of the North Anna Units 1 and 2 reactor coolant system pressure boundary Leakage Detection System.

f. Ample margin exists between the small stable flaw sizes of item (e) and larger stable flaws.

g. Ample margin exists in the material properties used to demonstrate end-of-service life (fully aged) stability of the critical flaws.

For the critical locations, flaws are identified that will be stable because of the ample margins described in e, f, and g above.

Based on the above, the Leak-Before-Break conditions and margins are satisfied for the ,North Anna Units 1 and 2 primary loop piping. All the recommended margins are satisfied'. It is therefore concluded that dynamic effects of RCS primary loop pipe breaks need not be considered in the structural design basis for North Anna Units 1 and 2 Nuclear Power Plants for the 80 year plant life (subsequent license renewal program).

Conclusions January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-1 APPENDIX A: LIMIT MOMENT

]a,c,e Appendix A: Limit Moment January 2020 WCAP-11164-NP  : Revision 2 I *** This record was final approved on 1/17/2020 11 :18:49 AM. (This statement was added by the PRIME system upon its validation) l_

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-2 (I)

~.,---------------------,

Figure A-1 Pipe with a Through-Wall Crack in Bending Appendix A: Limit Moment January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-1 APPENDIX B: LBB HISTORY FROM SURRY POWER STATION SLR Recently, the Subsequent License Renewal (SLR) application for the Surry Power Station (SPS) was submitted to the US Nuclear Regulatory Commission (NRC) for review and approval. As part of the review process, NRC regulators reviewed the LBB analysis scope for Surry to ensure that the evaluation conclusions remain applicable for the period of extended operation.

Throughout that review process, the NRC regulators issued a number of Requests for Additional Information (RAls) related to the LBB analysis.

The Surry Units 1 and 2 pants are a similar design to the North Anna Units 1 and 2 plants, and both power stations are owned and operated by Dominion Energy. As such, the RAls related to the Surry LBB report for SLR are reviewed here for applicability to the North Anna LBB evaluation. The Surry RAls will be summarized in this appendix, and if found to be applicable to the North Anna LBB evaluation, the relevant information will be identified.

RAl-1 Summary (1) Discuss the fracture toughness data of plant-specific (or representative) primary loop stainless steel welds to confirm that the fracture toughness data of the welds are greater than the fracture toughness estimated for the CASS elbows. Alternatively, identify relevant references (e.g., references to topical reports) for the weld fracture toughness data.

(2) In addition, clarify how the limit load analysis determines the material properties of the welds (e.g., flow stresses). Alternatively, identify relevant references (e.g., references to topical reports) for the weld material properties considered in the limit load analysis.

(3) Clarify whether the fracture toughness values of the CASS elbows estimated in accordance with Revision 2 of NUREG/CR-4513 are more limiting than the saturated fracture toughness (fully aged) in accordance with Revision 1 of NUREG/CR-4513 for the cold leg, crossover leg and hot leg locations. If not, please discuss whether the use of the fracture toughness value in accordance with NUREG/CR-4513, Revision 1 affects the conclusion of the crack stability analysis.

Response for North Anna RCL Piping LBB (1) Historic testing done by Westinghouse on representative plants, documented in References B-1 and B-2, has shown that the wrought and cast stainless steel piping exhibits more limiting (unaged) fracture toughness properties than the weld metal. Since Reference B-3 indicates that CASS material's aged lower bound fracture toughness values are similar to that of Submerged Arc Welds (SAWs). And since SAWs are considered to be the most limiting of welding processes (with respect to GTAW and SMAW), it is concluded that the aged fracture toughness of the wrought and cast stainless steel base metal is more limiting than the aged fracture toughness of the weld Appendix 8: L88 History from Surry Power Station SLR January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-2 metal. The North Anna RCL piping system also includes dissimilar metal welds (DMWs),

fabricated using Alloy 82/182 material with various profiles including weld inlay or weld overlay of Alloy 52/152 material. This report presents additional information supporting the justification that the fracture toughness of the CASS base metal bounds that of the nickel alloy DMWs.

(2) The limit load analyses consider material properties (yield and ultimate strength) of the base metal, and not the material properties of the weld metal. The base metal (piping) is considered to have more limiting material properties than the weld metal. In addition to using the limiting yield and ultimate strength of the base metals, the limit load analyses at the critical locations consider a Z-factor penalty. The Z-factor is consistent with the methodology of Reference 8-4 and accounts for reduction of the material toughness due to the welding process_ used during construction. By combining the limiting yield and ultimate strength of the base metals with the Z-factor penalty of the weld, the limit load analysis ensures that it is bounding both the weld metal and base metal. For the North Anna DWM weld location without inlay/overlay, an additional limit load analysis is presented in Section 7.3 of this report which accounts for the crack morphology effects due to primary water stress corrosion cracking. This analysis considers the typical material properties of the Alloy 82/182 weld material.

(3) The LBS evaluations in this report uses Reference 8-3 for estimating the thermally aged properties of the CASS material since this is the most recent version of NUREG/CR-4513 approved by NRC in May 2016. Revision 2 of Reference 8-3 revises the procedure and correlations, used for predicting the change in fracture toughness properties of CASS components due to thermal aging, to account for the most current and larger fracture toughness database (than Revision 1) for CASS materials aged up to 100,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> at 290-350°C (554-633°F). The methodology for estimating fracture properties has been extended to cover a wider range of CASS materials with a ferrite content of up to 40% in Reference 8-3. A more detailed discussion on the updated methodology is presented in the Executive Summary and Section 1 Introduction of Reference 8-3.

Therefore, the fracture toughness correlations from Reference 8-3 are appropriate for use in the LBS evaluations.

RAl-2 Summary Explain why the applied J-integral for [Surry] location 3 is greater than that of [Surry] location 6 even though the axial force and moment of [Surry] location 3 are less than those of [Surry]

location 6, respectively. As part of the response, provide the K1 (stress intensity factor for axial tension) and Kb (stress intensity factor for bending) for each of [Surry] locations 3 and 6, as the plastic zone corrections are applied.

Appendix B: LBB History from Surry Power Station SLR January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-3 Response for North Anna RCL Piping LBB This RAI is not applicable for North Anna. The scenario identified in the RAI was due to using different fracture mechanics methodology for the two locations being discussed. For North Anna, the local failure mechanism uses a [ ]a,c,e analysis methodology for all analysis locations, as discussed in Section 7.1 of this report.

RAl-3 Summary Provide the following information: (1) the aspect ratio of the postulated initial crack sizes; and (2) the basis for the initial crack sizes for the fatigue analysis. As part of the response, clarify whether the initial crack depths are greater than those that are acceptable in accordance with the acceptance criteria of ASME Code,Section XI, inservice inspection requirements (e.g.,

Table IWB-3410-1). If not, explain why the analysis assumes initial cracks that are not large enough to be detected and repaired during the inservice inspection.

Response for North Anna RCL Piping LBB Note that the fatigue crack growth (FCG) analysis is not a requirement for the LBB analysis (per Reference B-4) since the LBB analysis is based on the postulation of a through-wall flaw, whereas the FCG analysis is performed based on the surface flaw. In addition, the NRC staff, in Reference B-5 had indicated that, "the Commission deleted the fatigue crack growth analysis in the proposed rule. This requirement was found to be unnecessary because it was bounded by the crack stability analysis."

Nevertheless, the FCG for North Anna RCL piping was documented in Section 8.0 of this report and retained to keep with historical precedence to demonstrate that small surface flaws will not result in a through-wall flaw over the design life of the plant. The aspect ratio for the postulated initial crack sizes are for a typical flaw shape of [ ]a,c,e (flaw length:flaw depth). Various initial flaw depths were considered in the FCG analysis to demonstrate that small, NOE-detectable flaw sizes on the order of [ ]a,c,e would be acceptable for the life of the plant (i.e. will not grow to the become complete through-wall).

The intent of FCG in the LBB analysis was not to use initial flaw depths that are larger than the Acceptance Tables of ASME Section XI IWB-3410-1, but rather to show a defense in-depth fatigue crack growth based on small flaw sizes that are detectable based on NOE examination techniques, which would not become through-wall flaws over the design life of the plant.

Appendix B: LBB History from Surry Power Station SLR January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-4 RAl-4 Summary This request had asked for a description of the basis for why the fatigue crack growth analysis does not include the "Inadvertent auxiliary pressurizer spray" transient that is listed in a table of the Surry SLR application.

Response for North Anna RCL Piping LBB This request was specific to the design transients considered for the Surry FCG evaluation and is not directly applicable to the scope of this report for North Anna. Section 8.0 of this report provides the details of the FCG evaluation for the North Anna RCL piping and Table 8-3 identifies the design transients which are applicable to North Anna for the 80 year plant life of the SLR program.

RAl-5 Summary (1) This request had asked to clarify whether the LBB TLAA (Time-Limited Aging Analysis) applies only to the RCS primary loop piping, as the NRG reviewers did not find this to be clearly communicated in the SLR application.

(2) The request asked for a revision to the UFSAR (Updated Final Safety Analysis Report) to include the specific revision number of the WCAP report applicable to the Surry LBB evaluation of the RCL piping.

Response for North Anna RCL Piping LBB (1) As for the Surry plants, the only LBB analysis completed for North Anna Units 1 and 2 is for the main loop reactor coolant piping described in this report. The only existing TLAA under 10CFR54.21 pertaining to LBB is for the main loop reactor coolant piping.

(2) Update of the North Anna UFSAR will occur if and when the SLR application for North Anna Units 1 and 2 is approved by the NRG. At this time, the UFSAR will be updated to reference the current revision of this WCAP report for the North Anna primary loop piping LBB evaluation.

Appendix B: LBB History from Surry Power Station SLR January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-5 RAl-6 Summary (1) Clarify whether Z factors are applied to both axial (including pressure) and moment loads. If not, provide the technical basis for why the Z factors are not applied to both axial (including pressure) and moment loads.

(2) Clarify why the applied Z factors are sufficiently high to confirm the structural integrity of the thermally aged CASS elbows. As part of the response, clarify whether the other conservatisms associated with the method and results of the limit load analysis (in addition to the Z factors) are sufficient to confirm the structural integrity of the CASS elbows.

Response for North Anna RCL Piping L88 (1) The North Anna L88 evaluation consist of evaluating two failure mechanisms as discussed in Section 7 of this report.

The first failure mechanism is based on a J-integral evaluation to assess the local crack stability. In thE:l J integral evaluation, Japp is calculated based on the faulted loads without any Z-factors to account for reduction in fracture toughness. This is because the calculation of J1c, as part of the J-integral evaluations, already considers reduction in fracture toughness due to thermal aging of the CASS materials at normal operating temperature over extended operating periods. This reduction in fracture toughness is based on Reference 8-3 correlations which have determined lower bound fracture toughness; as discussed in Section 4.3 of this report. The results from the Reference 8-3 ANL Research Program also states that "the lower-bound fracture toughness of thermally aged cast stainless steel is similar to that of submerged arc welds (SAWs)".

Additionally, as discussed in Section 4.3.3, the cast stainless steel fracture toughness properties present a limiting condition when compared to the Alloys 82/182 and 52/152 fracture toughness properties found in Units 1 and 2 SG (inlet/outlet) nozzle welds.

Therefore, no additional Z-factors are necessary, because the reduction in fracture toughness is already captured with the consideration of end-of-life fracture toughness values from Reference 8-3.

The second failure mechanism that is also evaluated is based on plastic instability (limit

• load). In this evaluation, the flawed pipe is predicted to fail when the remaining net section reaches a stress level at which a plastic hinge is formed; see Section 7.2 of this report. In this global failure mechanism, Z-factors are included based on the weld types and are applied on the faulted loads at the critical locations. The Z-factors, used in the limit load global failure evaluation, is applied to both the axial load (including pressure) and the moment load. The limit load evaluations with incorporation of Z-factors are equated to the material flow stress to determine the critical flaw sizes in the L88 evaluation. The incorporation of Z-factors for limit load, as performed in this report, is consistent with Reference 8-4.

Appendix B: LBB History from Surry Power Station SLR January 2020 WCAP-11164-NP Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-6 (2) Excluding the Alloy 82/182 weld location without inlay or overlay, the limit load analyses consider material properties (yield and ultimate strength) of the base metal, and not the material properties of the weld metal. The base metal (piping) is considered to have more limiting material properties than the weld metal. In addition to using the limiting yield and ultimate strength of the base metals, the limit load analyses at the critical locations consider a Z-factor penalty. The Z-factor is consistent with the methodology of Reference B-4 and accounts for reduction of the material toughness due to the welding process used during construction. By combining the limiting yield and ultimate strength of the base metals with the Z-factor penalty of the weld, the analysis is ensured to both the weld metal and base metal. Taken together, the use of more limiting material properties in combination with the Z-factor ensures that the LBB analysis is conservative relative to the structural integrity of the CASS materials.

For the SGON weld location on Unit 1, the Alloy 82/182 weld is known to be susceptible to the effects of PWSCC. The specific material properties of the Alloy 82/182 weld material are considered for the limit load analysis at this location. As noted in Section 7.3 of this report, the limit load analysis for this location considers a Z-factor of 1.0, since the effects of PWSCC have been addressed by applying a conservative crack morphology factor to the leakage flaw size calculation Reference B-4 does not explicitly discuss the consideration of the thermal aging of CASS material. However, to account for the thermal aging effect, a separate fracture mechanics analysis is performed which calculates the applied J-integral due to the faulted loading conditions (again, using the more-limiting base metal tensile properties) and compares the calculated Japp value to the thermally aged fracture toughness allowables (J1c, Jmax, Tmat) of the CASS material.

Therefore, the limit load analysis for CASS materials considers the reduced fracture toughness of the weld (Z-factor), the limit load analysis of the Alloy 82/182 weld considers a conservative crack morphology factor, and the J-applied analysis considers the reduced fracture toughness of the thermally aged CASS material per Reference B-3.

Appendix B: LBB History from Surry Power Station SLR January 2020 WCAP-11164-N P Revision 2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-7 Appendix B: References B-1 Westinghouse Report, WCAP-9787, "Tensile and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation," May 1981.

8-2 Westinghouse Report, WCAP-9558, Revision 2, "Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Through-Wall Crack,"

May 1981.

8-3 0. K. Chopra, "Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems," NUREG/CR-4513, Revision 2, U.S. Nuclear Regulatory Commission, Washington, DC, May 2016.

8-4 NUREG-0800, Revision 1, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures, March 2007.

8-5 Nuclear Regulatory Commission, 10 CFR 50, Modification of General Design Criteria 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures, Final Rule, Federal RegisterNol. 52, No. 207/Tuesday, October 27, 1987/Rules and Regulations, pp. 41288-41295.

Appendix B: LBB History from Surry Power Station SLR January 2020 WCAP-11164-NP Revision 2

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