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{{#Wiki_filter:Change Notice 2                                             Serial No.: 19-096 SPS SLRA                                              Docket Nos.: 50-280/281 Enclosure 5 TOPICAL REPORT WCAP-15550-NP, REV. 2 Virginia Electric and Power Company (Dominion Energy Virginia)
{{#Wiki_filter:Change Notice 2 SPS SLRA Serial No.: 19-096 Docket Nos.: 50-280/281 TOPICAL REPORT WCAP-15550-NP, REV. 2 Virginia Electric and Power Company (Dominion Energy Virginia)
Surry Power Station Units 1 and 2
Surry Power Station Units 1 and 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15550-N P                                      March 2019 Revision 2 Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Surry Units 1 and 2 Nuclear Power Plants for the Subsequent License Renewal Program (80 Years) Leak-Before-Break Evaluation
WCAP-15550-N P Revision 2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 March 2019 Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Surry Units 1 and 2 Nuclear Power Plants for the Subsequent License Renewal Program (80 Years) Leak-Before-Break Evaluation  


WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15550-NP Revision 2 Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Surry Units 1 and 2 Nuclear Power Plants for the Subsequent License Renewal Program (80 Years)
WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15550-NP Revision 2 Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Surry Units 1 and 2 Nuclear Power Plants for the Subsequent License Renewal Program (80 Years)
Leak-Before-Break Evaluation March 2019 Authors:   Mamo Wiratmo*
Leak-Before-Break Evaluation March 2019 Authors:
Structural Design & Analysis II Reviewer:   Eric D. Johnson*
Mamo Wiratmo*
Structural Design & Analysis II Approved: Benjamin A. Leber*, Manager Structural Design & Analysis II
Structural Design & Analysis II Reviewer: Eric D. Johnson*
*Electronically approved records are authenticated in the electronic document management system.
Structural Design & Analysis II Approved: Benjamin A. Leber*, Manager Structural Design & Analysis II  
Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA
*Electronically approved records are authenticated in the electronic document management system.
                            © 2019 Westinghouse Electric Company LLC All Rights Reserved
Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA  
© 2019 Westinghouse Electric Company LLC All Rights Reserved  


WESTINGHOU SE NON-PROPRIETARY CLASS 3                                 iii RECORD OF REVISIONS Rev        Date  Revision Description August 0                Original Issue (WCAP-15550) 2000 Revised to include the LBB results from the Measured Uncertainty 1-A                Recapture (MUR) Program.
Rev Date 0
(Draft for    March customer      2017 This WCAP revision also includes LBB evaluation for subsequent review) license renewal for 80 years of operation for Surry Units 1 and 2.
August 2000 1-A (Draft for March customer 2017 review) 1 June 2017 2
June 1                Finalized Revision 1-A and incorporated customer's comments.
March 2019 WCAP-15550-N P WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii RECORD OF REVISIONS Revision Description Original Issue (WCAP-15550)
2017 Revised to address three chemical content errors in Tables 4-6 and 4-7 and to address the updated CMTR data from Surry Unit 1 that have not been considered in Revision 1. The updated CMTR data is March  provided in Table 4-6.
Revised to include the LBB results from the Measured Uncertainty Recapture (MUR) Program.
2 2019  As shown in the revised Tables 4-6 and 4-7, the corrections of CMTR errors are shown in bold font and updates due to additional CMTRs are marked by grey-shaded color. Changes in the text of the document are shown with revision bars in the right margin.
This WCAP revision also includes LBB evaluation for subsequent license renewal for 80 years of operation for Surry Units 1 and 2.
WCAP-15550- N P                                                                      March 2019 Revision 2
Finalized Revision 1-A and incorporated customer's comments.
Revised to address three chemical content errors in Tables 4-6 and 4-7 and to address the updated CMTR data from Surry Unit 1 that have not been considered in Revision 1. The updated CMTR data is provided in Table 4-6.
As shown in the revised Tables 4-6 and 4-7, the corrections of CMTR errors are shown in bold font and updates due to additional CMTRs are marked by grey-shaded color. Changes in the text of the document are shown with revision bars in the right margin.
March 2019 Revision 2  


WESTINGH OUSE NON-PROPRIETARY CLASS 3                                                                         iv TABLE OF CONTENT S
WESTINGHOUSE NON-PROPRIETARY CLASS 3 iv TABLE OF CONTENTS
: 1. O 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
: 1. O 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     Water Hammer.............................................. : ..................................................... 2-2 2.3     Low Cycle and High Cycle Fatigue ..................................................................... 2-3 2.4     Wall Thinning, Creep, and Cleavage ................................................................... 2-3
: 2. 0 Operation and Stability of the Reactor Coolant System..................................................
: 2. 5     References ......................................................................................................... 2-3 3.0   Pipe Geometry and Loading ...................................................................... :.................... 3-1 3.1       Introduction to Methodology ................................................................................ 3-1 3.2     Calculation of Loads and Stresses .................................................................. :... 3-1 3.3     Loads for Leak Rate Evaluation .......................................................................... 3-2 3.4     Load Combination for Crack Stability Analyses ................................................... 3-3 3.5     References ......................................................................................................... 3-3 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 ........................................................................... 4-1 4.4     References ......................................................................................................... 4-4
* 2-1 2.1 Stress Corrosion Cracking.................................................................................. 2-1 2.2 Water Hammer.............................................. :..................................................... 2-2 2.3 Low Cycle and High Cycle Fatigue..................................................................... 2-3 2.4 Wall Thinning, Creep, and Cleavage................................................................... 2-3
: 5. 0   Critical Location and Evaluation Criteria ......................................................................... 5-1 5.1     Critical Locations ................................................................................................. 5-1 5.2     Fracture Criteria .................................................................................................. 5-1 6.0   Leak Rate Predictions .................................................................................................... 6-1 6.1       Introduction ......................................................................................................... 6-1
: 2. 5 References......................................................................................................... 2-3 3.0 Pipe Geometry and Loading...................................................................... :.................... 3-1 3.1 Introduction to Methodology................................................................................ 3-1 3.2 Calculation of Loads and Stresses.................................................................. :... 3-1 3.3 Loads for Leak Rate Evaluation.......................................................................... 3-2 3.4 Load Combination for Crack Stability Analyses................................................... 3-3 3.5 References......................................................................................................... 3-3 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........................................................................... 4-1 4.4 References......................................................................................................... 4-4
: 6. 2     General Considerations ...................................................................................... 6-1 6.3       Calculation Method ............................................................................................. 6-1 6.4       Leak Rate Calculations ....................................................................................... 6-2 6.5       References ......................................................................................................: .. 6-2
: 5. 0 Critical Location and Evaluation Criteria......................................................................... 5-1 5.1 Critical Locations................................................................................................. 5-1 5.2 Fracture Criteria.................................................................................................. 5-1 6.0 Leak Rate Predictions.................................................................................................... 6-1 6.1 Introduction......................................................................................................... 6-1
: 7. 0 Fracture Mechanics Evaluation ...................................................................................... 7-1 7.1       Local Failure Mechanism .................................................................................... 7-1 7.2       Global Failure Mechanism .................................................................................. 7-2 7.3       Crack Stability Evaluations .................................................................................. 7-3 7.4       References ......................................................................................................... 7-4 8.0   Fatigue Crack Growth Analysis ....................................................................................... 8-1 8.1       References ......................................................................................................... 8-2 9.0   Assessment of Margins .................................................................................................. 9-1 10.0 Conclusions .................................................................................................................. 10-1 Appendix A: Limit Moment. ......................................................................................................... A-1 WCAP-1555 0-N P                                                                                                             March 2019 Revision 2
: 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 Crack Stability Evaluations.................................................................................. 7-3 7.4 References......................................................................................................... 7-4 8.0 Fatigue Crack Growth Analysis....................................................................................... 8-1 8.1 References......................................................................................................... 8-2 9.0 Assessment of Margins.................................................................................................. 9-1 10.0 Conclusions.................................................................................................................. 10-1 Appendix A: Limit Moment.......................................................................................................... A-1 WCAP-15550-N P March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                                                           V LIST OF TABLES Table 3-1 Dimensions, Normal Loads and Stresses for Surry Units 1 and 2 ............................... 3-4 Table 3-2 Faulted Loads and Stresses for Surry Units 1 and 2 ................................................... 3-5 Table 4-1 Measured Tensile Properties (psi) for Surry Unit 1 Primary Loop Pipes ...................... 4-5 Table 4-2 Measured Tensile Properties (psi) for Surry Unit 2 Primary Loop Pipes ...................... 4-6 Table 4-3 Measured Tensile Properties (psi) for Surry Unit 1 Primary Loop Elbows ................... .4-7 Table 4-4 Measured Tensile Properties (psi) for Surry Unit 2 Primary Loop Elbows ................... .4-8 Table 4-5 Mechanical Properties for Surry Units 1 and 2 Materials at Operating Temperatures ............................................................................................. 4-9 Table 4-6 Chemistry and Fracture Toughness Elbow Properties of the Material Heats of Surry Unit 1 ............................................................................................................. 4-10 Table 4-7 Chemistry and Fracture Toughness Elbow Properties of the Material Heats of Surry Unit 2 ............................................................................................................. 4-12 Table 4-8 Fracture Toughness Properties for Surry Units 1 and 2 Primary Loops for Leak-Before-Break Evaluation at Critical Locations ................................. 4-13 Table 6-1 Flaw Sizes Yielding a Leak Rate of 10 gpm at the Governing Locations .................... 6-3 Table 7-1 Stability Results for Surry Units 1 and 2 Based on Elastic-Plastic J-lntegral Evaluations ............................................................................................................... 7-5 Table 7-2 Stability Results for Surry Units 1 and 2 Based on Limit Load .................................... 7-5 Table 8-1 Summary of Reactor Vessel Transients ..................................................................... 8-3 Table 8-2 Typical Fatigue Crack Growth at [                                                   rc,e (40, 60, and 80 years)****:****************************************************************************************** 8-4 Table 8-3 Summary of Reactor Vessel Transients for Surry Units 1 and 2 (40, 60, and 80 years) ............................................................................................... 8-5 Table 9-1 Leakage Flaw Size, Critical Flaw Sizes and Margins for Surry Units 1 and 2 .............................................................................................. 9-2 WCAP-15550-N P                                                                                                           March 2019 Revision 2
WESTINGHOUSE NON-PROPRIETARY CLASS 3 V
LIST OF TABLES Table 3-1 Dimensions, Normal Loads and Stresses for Surry Units 1 and 2............................... 3-4 Table 3-2 Faulted Loads and Stresses for Surry Units 1 and 2................................................... 3-5 Table 4-1 Measured Tensile Properties (psi) for Surry Unit 1 Primary Loop Pipes...................... 4-5 Table 4-2 Measured Tensile Properties (psi) for Surry Unit 2 Primary Loop Pipes...................... 4-6 Table 4-3 Measured Tensile Properties (psi) for Surry Unit 1 Primary Loop Elbows....................4-7 Table 4-4 Measured Tensile Properties (psi) for Surry Unit 2 Primary Loop Elbows....................4-8 Table 4-5 Mechanical Properties for Surry Units 1 and 2 Materials at Operating Temperatures............................................................................................. 4-9 Table 4-6 Chemistry and Fracture Toughness Elbow Properties of the Material Heats of Surry Unit 1............................................................................................................. 4-10 Table 4-7 Chemistry and Fracture Toughness Elbow Properties of the Material Heats of Surry Unit 2............................................................................................................. 4-12 Table 4-8 Fracture Toughness Properties for Surry Units 1 and 2 Primary Loops for Leak-Before-Break Evaluation at Critical Locations................................. 4-13 Table 6-1 Flaw Sizes Yielding a Leak Rate of 10 gpm at the Governing Locations.................... 6-3 Table 7-1 Stability Results for Surry Units 1 and 2 Based on Elastic-Plastic J-lntegral Evaluations............................................................................................................... 7-5 Table 7-2 Stability Results for Surry Units 1 and 2 Based on Limit Load.................................... 7-5 Table 8-1 Summary of Reactor Vessel Transients..................................................................... 8-3 Table 8-2 Typical Fatigue Crack Growth at [
rc,e (40, 60, and 80 years)****:****************************************************************************************** 8-4 Table 8-3 Summary of Reactor Vessel Transients for Surry Units 1 and 2 (40, 60, and 80 years)............................................................................................... 8-5 Table 9-1 Leakage Flaw Size, Critical Flaw Sizes and Margins for Surry Units 1 and 2.............................................................................................. 9-2 WCAP-15550-N P March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                                                     vi LIST OF FIGURES Figure 3-1 Hot Leg Coolant Pipe ............................................................................................. 3-6 Figure 3-2 Schematic Diagram of Surry Units 1 and 2 Primary Loop Showing Weld Locations ...................................................................................................... 3-7 Figure 4-1 Pre-Service J vs. ~a for SA351 CF8M Cast Stainless Steel at 600°F ................... 4-14 Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures ................... 6-4 Figure 6-2                     rc,e Pressure Ratio as a Function of UD ..................................... 6-5 Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack ................................. 6-6 Figure 7-1                 t,c,e Stress Distribution ................................................................... 7-6 Figure 7-2 Critical Flaw Size Prediction - Hot Leg at Location 1 ............................................. 7-7 Figure 7-3 Critical Flaw Size Prediction - Hot Leg at Location 3 ............................................. 7-8 Figure 7-4 Critical Flaw Size Prediction - Cross-over Leg at Location 6 ................................. 7-9 Figure 7-5 Critical Flaw Size Prediction - Cold Leg at Location 15 ....................................... 7-10 Figure 8-1 Typical Cross-Section of [                                                     ]a,c,e ............................ .... 8-6 Figure 8-2 Reference Fatigue Crack Growth Curves for Carbon and Low Alloy Ferritic Steels ................................................................................................ 8-7 FigureA-1 Pipe with a Through-Wall Crack in Bending ........................................................... A-2 WCAP-15550-N P                                                                                                         March 2019 Revision 2
WESTINGHOUSE NON-PROPRIETARY CLASS 3 vi LIST OF FIGURES Figure 3-1 Hot Leg Coolant Pipe............................................................................................. 3-6 Figure 3-2 Schematic Diagram of Surry Units 1 and 2 Primary Loop Showing Weld Locations...................................................................................................... 3-7 Figure 4-1 Pre-Service J vs. ~a for SA351 CF8M Cast Stainless Steel at 600°F................... 4-14 Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures................... 6-4 Figure 6-2 rc,e Pressure Ratio as a Function of UD..................................... 6-5 Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack................................. 6-6 Figure 7-1 t,c,e Stress Distribution................................................................... 7-6 Figure 7-2 Critical Flaw Size Prediction - Hot Leg at Location 1............................................. 7-7 Figure 7-3 Critical Flaw Size Prediction - Hot Leg at Location 3............................................. 7-8 Figure 7-4 Critical Flaw Size Prediction - Cross-over Leg at Location 6................................. 7-9 Figure 7-5 Critical Flaw Size Prediction - Cold Leg at Location 15....................................... 7-10 Figure 8-1 Typical Cross-Section of [  
]a,c,e................................ 8-6 Figure 8-2 Reference Fatigue Crack Growth Curves for Carbon and Low Alloy Ferritic Steels................................................................................................ 8-7 FigureA-1 Pipe with a Through-Wall Crack in Bending........................................................... A-2 WCAP-15550-N P March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                   vii EXECUTIVE  
WESTINGHOUSE NON-PROPRIETARY CLASS 3 vii EXECUTIVE  


==SUMMARY==
==SUMMARY==
The original structural design basis of the reactor coolant system for the Surry Units 1 and 2 Nuclear Power Plants required consideration of dynamic effects resulting from pipe break and that protective measures for such breaks be incorporated into the design. Subsequent to the original Surry 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).
The original structural design basis of the reactor coolant system for the Surry Units 1 and 2 Nuclear Power Plants required consideration of dynamic effects resulting from pipe break and that protective measures for such breaks be incorporated into the design. Subsequent to the original Surry 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).
Surry Units 1 and 2 Nuclear Power Plants were 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 served as the original basis for the elimination of large primary loop pipe rupture from the structural design basis for Surry Units 1 and 2. As identified in Generic Letter 84-04, the primary technical references supporting the NRC's safety evaluation of eliminating postulated pipe breaks are documented in WCAP-9558 (Reference 1-9) and WCAP-9787 (Reference 1-10).
Surry Units 1 and 2 Nuclear Power Plants were 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 served as the original basis for the elimination of large primary loop pipe rupture from the structural design basis for Surry Units 1 and 2. As identified in Generic Letter 84-04, the primary technical references supporting the NRC's safety evaluation of eliminating postulated pipe breaks are documented in WCAP-9558 (Reference 1-9) and WCAP-9787 (Reference 1-10).
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.
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 1OCFR50 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 technology for excluding from the design basis the dynamic effects of postulated ruptures in primary coolant loop piping in pressurized water reactors (PWRs).
Subsequently, the NRC modified 1 OCFR50 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 technology for excluding from the design basis the dynamic effects of postulated ruptures in primary coolant loop piping in pressurized water reactors (PWRs).
The LBB evaluation is 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.
The LBB evaluation is 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.
Revision O of this report had demonstrated compliance with LBB technology for the Surry reactor coolant system piping for the 60 year plant life based on a plant specific analysis. Subsequently, an LBB evaluation was performed for the MUR (Measurement Uncertainty Program), the results of that particular analysis are also incorporated in Revision 1 of this report. Lastly, based on the LBB evaluation in Revision 1 of this report herein, it also demonstrated that dynamic effects of reactor coolant system primary loop pipe breaks need not be considered in the structural design basis of the Surry Units 1 and 2 Nuclear Power Plants for the 80 year plant life (Subsequent License Renewal Program). The technical evaluations utilized in Revision O through Revision 2 of this report are consistent with the methodology and principles of WCAP-9558 and WCAP-9787.
Revision O of this report had demonstrated compliance with LBB technology for the Surry reactor coolant system piping for the 60 year plant life based on a plant specific analysis. Subsequently, an LBB evaluation was performed for the MUR (Measurement Uncertainty Program), the results of that particular analysis are also incorporated in Revision 1 of this report. Lastly, based on the LBB evaluation in Revision 1 of this report herein, it also demonstrated that dynamic effects of reactor coolant system primary loop pipe breaks need not be considered in the structural design basis of the Surry Units 1 and 2 Nuclear Power Plants for the 80 year plant life (Subsequent License Renewal Program). The technical evaluations utilized in Revision O through Revision 2 of this report are consistent with the methodology and principles of WCAP-9558 and WCAP-9787.
Therefore, the justifications demonstrated herein are compliant with the original conclusions of Generic Letter 84-04.
Therefore, the justifications demonstrated herein are compliant with the original conclusions of Generic Letter 84-04.
The report documents the plant specific geometry, loading, and material properties used in the fracture mechanics evaluation. Mechanical properties were determined at operating temperatures.
The report documents the plant specific geometry, loading, and material properties used in the fracture mechanics evaluation. Mechanical properties were determined at operating temperatures.
WCAP-15550-NP                                                                             March 2019 Revision 2
WCAP-15550-NP March 2019 Revision 2  
 
WESTINGHOUSE NON-PROPRIETARY CLASS 3                                viii Since the piping systems include cast austenitic stainless steel, fracture toughness considering thermal aging was determined for each heat of material. Fully aged fracture toughness properties were used for the LBB evaluation. The full aged condition is applicable for plants operating at beyond 15 EFPY (Effective Full Power Years) for the CF8M materials (elbows for Surry Units 1 and 2). As of January 2017, Surry Units 1 and 2 are operating at 33.78 and 33.69 EFPY, respectively. Thus, the LBB evaluation in this report has been demonstrated for the primary loops at Surry Units 1 and 2 for 80 years of plant operation.
WCAP-15550-N P                                                                        March 2019 Revision 2


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                 1-1
WESTINGHOUSE NON-PROPRIETARY CLASS 3 viii Since the piping systems include cast austenitic stainless steel, fracture toughness considering thermal aging was determined for each heat of material. Fully aged fracture toughness properties were used for the LBB evaluation. The full aged condition is applicable for plants operating at beyond 15 EFPY (Effective Full Power Years) for the CF8M materials (elbows for Surry Units 1 and 2).
As of January 2017, Surry Units 1 and 2 are operating at 33.78 and 33.69 EFPY, respectively. Thus, the LBB evaluation in this report has been demonstrated for the primary loops at Surry Units 1 and 2 for 80 years of plant operation.
WCAP-15550-N P March 2019 Revision 2


==1.0      INTRODUCTION==
WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-1  


1.1     PURPOSE This report applies to the Surry Units 1 and 2 Reactor Coolant System (RCS) primary loop piping. It is intended to demonstrate that for the specific parameters of the Surry 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 LBS evaluation results based on the Measurement Uncertainty Recapture (MUR)
==1.0 INTRODUCTION==
Program.
1.1 PURPOSE This report applies to the Surry Units 1 and 2 Reactor Coolant System (RCS) primary loop piping. It is intended to demonstrate that for the specific parameters of the Surry 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 LBS evaluation results based on the Measurement Uncertainty Recapture (MUR)
Program.  


==1.2     BACKGROUND==
==1.2 BACKGROUND==
INFORMATION Westinghouse has performed considerable testing and analysis to demonstrate that 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 LOCA Loads.
INFORMATION Westinghouse has performed considerable testing and analysis to demonstrate that 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 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).
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 12 primary loop pipe break) to be 4.4 x 10- per reactor year and the mean probability of an 7
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.
indirect LOCA to be 10- per reactor year. Thus, the results previously obtained by Westinghouse (Reference 1-3) were confirmed by an independent NRC research study.
Based on the studies by Westinghouse, LLNL, the ACRS, and the 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.
Based on the studies by Westinghouse, LLNL, the ACRS, and the 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).
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).
Introduction                                                                         March 2019 WCAP-15550-NP                                                                           Revision 2
Introduction WCAP-15550-NP March 2019 Revision 2  


WESTING HOUSE NON-PROPRIETARY CLASS 3                                 1-2 1.3       SCOPE AND OBJECTIVES The general purpose of this investigation is to demonstrate leak-before-break for the primary loops in Surry Units 1 and 2 on a plant specific basis for the 80 year plant life. The recommendations and criteria proposed in References 1-7 and 1-8 are used in this evaluation.
WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-2 1.3 SCOPE AND OBJECTIVES The general purpose of this investigation is to demonstrate leak-before-break for the primary loops in Surry Units 1 and 2 on a plant specific basis for the 80 year plant life. The recommendations and criteria proposed in References 1-7 and 1-8 are used in this evaluation.
These criteria and resulting steps of the evaluation procedure can be briefly summarized as follows:
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.
: 1.
: 2.       Identify the materials and the associated material properties.
Calculate the applied loads. Identify the locations at which the highest stress occurs.
: 3.       Postulate a surface flaw at the governing locations. Determine fatigue crack growth.
: 2.
Identify the materials and the associated material properties.
: 3.
Postulate a surface flaw at the governing locations. Determine fatigue crack growth.
Show that a through-wall crack will not result.
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.
: 4.
: 5.       Using faulted loads, demonstrate that there is a margin of 2 between the leakage flaw size and the critical flaw size.
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.
: 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.
: 5.
: 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.
Using faulted loads, demonstrate that there is a margin of 2 between the leakage flaw size and the critical flaw size.
: 8.       Demonstrate margin on applied load.
: 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 applied load.
This report provides a fracture mechanics demonstration of primary loop integrity for the Surry Units 1 and 2 plants consistent with the NRC position for exemption from consideration of dynamic effects.
This report provides a fracture mechanics demonstration of primary loop integrity for the Surry Units 1 and 2 plants consistent with the NRC position for exemption from consideration of dynamic effects.
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.
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.
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.
March 2019 Introduction Revision 2 WCAP-15 550-NP
Introduction WCAP-15550-NP March 2019 Revision 2
 
WESTINGHOU SE NON-PROPRIETARY CLASS 3                              1-3


==1.4    REFERENCES==
WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-3


1-1     USNRC Generic Letter 84-04,  
==1.4 REFERENCES==
1-1 USNRC Generic Letter 84-04,  


==Subject:==
==Subject:==
  "Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops," February 1, 1984.
  "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 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures, Final Rule, Federal RegisterNol. 52, No. 207fruesday, October 27, 1987/Rules and Regulations, pp. 41288-41295.
1-2 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. 207fruesday, October 27, 1987/Rules and Regulations, pp. 41288-41295.
1-3     WCAP-9283, "The Integrity of Primary Piping Systems of Westinghouse Nuclear Power Plants During Postulated Seismic Events," March, 1978.
1-3 WCAP-9283, "The Integrity of Primary Piping Systems of Westinghouse Nuclear Power Plants During Postulated Seismic Events," March, 1978.
1-4     Letter Report NS-EPR-2519, Westinghouse (E. P. Rahe) to NRC (D. G. Eisenhut),
1-4 Letter Report NS-EPR-2519, Westinghouse (E. P. Rahe) to NRC (D. G. Eisenhut),
Westinghouse Proprietary Class 2, November 10, 1981.
Westinghouse Proprietary Class 2, November 10, 1981.
1-5     Letter from Westinghouse (E. P. Rahe) to NRC (W. V. Johnston) dated April 25, 1983.
1-5 Letter from Westinghouse (E. P. Rahe) to NRC (W. V. Johnston) dated April 25, 1983.
1-6     Letter from Westinghouse (E. P. Rahe) to NRC (W. V. Johnston) dated July 25, 1983.
1-6 Letter from Westinghouse (E. P. Rahe) to NRC (W. V. Johnston) dated July 25, 1983.
1-7     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.
1-7 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.
32626-32633.
1-8     NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.
1-8 NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.
1-9     WCAP-9558, Revision 2, "Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Throughwall Crack," May, 1981.
1-9 WCAP-9558, Revision 2, "Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Throughwall Crack," May, 1981.
1-10   WCAP-9787, "Tensile and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation," May, 1981.
1-10 WCAP-9787, "Tensile and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation," May, 1981.
Introduction                                                                         March 2019 WCAP-15550-N P                                                                       Revision 2
Introduction WCAP-15550-N P March 2019 Revision 2  


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.
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:
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."
"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.
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.
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 SCC as well as preventing the occurrence of a Operation and Stability of the Reactor Coolant System                               March 2019 WCAP-15550-NP                 .                                                      Revision 2
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 SCC as well as preventing the occurrence of a Operation and Stability of the Reactor Coolant System WCAP-15550-NP March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                             2-2 corrosive environment. The material specifications consider compatibility with the system's operating environment (both internal and external) as well as other material in the system, applicable ASME Code rules, fracture toughness, welding, fabrication, and processing.
WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 corrosive environment. The material specifications consider compatibility with the system's 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.
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 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.
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 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.
It should be noted that there are no primary water stress corrosion cracking material such as Alloy 82/182 in the dissimilar metal welds in the Surry Units 1 and 2 Reactor Coolant System (RCS) primary loop piping.
It should be noted that there are no primary water stress corrosion cracking material such as Alloy 82/182 in the dissimilar metal welds in the Surry Units 1 and 2 Reactor Coolant System (RCS) primary loop piping.
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 controlled in the design process. Additionally, Westinghouse has instrumented typical reactor coolant systems to verify the flow and vibration characteristics of the system. Preoperational Operation and Stability of the Reactor Coolant System                               March 2019 WCAP-15550-N P                                                                       Revision 2
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 controlled in the design process. Additionally, Westinghouse has instrumented typical reactor coolant systems to verify the flow and vibration characteristics of the system.
Preoperational Operation and Stability of the Reactor Coolant System WCAP-15550-N P March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                 2-3 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.
WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 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.
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 measurements have been made on a number of plants during hot functional testing, including plants similar to Surry 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.
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 measurements have been made on a number of plants during hot functional testing, including plants similar to Surry 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 should 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.
2.4 WALL THINNING, CREEP, AND CLEAVAGE Wall thinning by erosion and erosion-corrosion effects should 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.
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.
Cleavage type failures are not a concern for the operating temperatures and the stainless steel material used in the primary loop piping.
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.
==2.5 REFERENCES==
2-2     Investigation and Evaluation of Cracking Incidents in Piping in Pressurized Water Reactors, NUREG-0691, U.S. Nuclear Regulatory Commission, September 1980.
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.
Operation and Stability of the Reactor Coolant System                                     March 2019 WCAP-15550-NP                                                                               Revision 2
2-2 Investigation and Evaluation of Cracking Incidents in Piping in Pressurized Water Reactors, NUREG-0691, U.S. Nuclear Regulatory Commission, September 1980.
Operation and Stability of the Reactor Coolant System WCAP-15550-NP March 2019 Revision 2  


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


==3.1     INTRODUCTION==
==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 34.00 in. and 2.395 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, dead weight and thermal expansion. An additional component, Safe Shutdown Earthquake (SSE), is considered for faulted loads. Tables 3-1 and 3-2 show the enveloping loads for Surry Units 1 and 2; these loads were determined as part of the MUR project. As seen from Table 3-2, the highest stressed location in the entire loop is at Location 1 at the reactor vessel outlet nozzle to pipe weld. This is one of the locations at which, as an enveloping location, 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.
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 34.00 in. and 2.395 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, dead weight and thermal expansion. An additional component, Safe Shutdown Earthquake (SSE), is considered for faulted loads. Tables 3-1 and 3-2 show the enveloping loads for Surry Units 1 and 2; these loads were determined as part of the MUR project. As seen from Table 3-2, the highest stressed location in the entire loop is at Location 1 at the reactor vessel outlet nozzle to pipe weld. This is one of the locations at which, as an enveloping location, 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 elbows are made of different materials than the pipe, locations other than the highest stressed pipe location were examined taking into consideration both fracture toughness and stress. The four most critical locations among the entire primary loop are identified after the full analysis is completed. 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. 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.
Since the elbows are made of different materials than the pipe, locations other than the highest stressed pipe location were examined taking into consideration both fracture toughness and stress. The four most critical locations among the entire primary loop are identified after the full analysis is completed. 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. 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.
Please note that the piping loads and stresses based on the MUR Program were considered in the LBB evaluation as part of Revision 1 of this WCAP report.
Please note that the piping loads and stresses based on the MUR Program were considered in the LBB evaluation as part of Revision 1 of this WCAP report.
3.2     CALCULATION OF LOADS AND STRESSES The stresses due to axial loads and bending moments are calculated by the following equation:
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=-+ -
F M
A     Z
a=-+ -
: where, cr =     stress, ksi F   =     axial load, kips M   =     bending moment, in-kips 2
A Z
A    =     pipe cross-sectional area, in z   =     section modulus, in 3 Pipe Geometry and Loading                                                                 March 2019 WCAP-15550-NP                                                                               Revision 2
: where, cr  
=
stress, ksi F  
=
axial load, kips M  
=
bending moment, in-kips A
=
pipe cross-sectional area, in2 z  
=
section modulus, in3 Pipe Geometry and Loading WCAP-15550-NP (3-1)
March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                 3-2 The total moments for the desired loading combinations are calculated by the following equation:
WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-2 The total moments for the desired loading combinations are calculated by the following equation:
(3-2)
(3-2)
: where, M     =     total moment for required loading Mx   =     X component of moment (torsion)
: where, M  
Mv   =     Y component of bending moment Mz   =     Z component of bending moment NOTE:       X-axis is along the center line of the pipe.
=
total moment for required loading Mx  
=
X component of moment (torsion)
Mv  
=
Y component of bending moment Mz  
=
Z component of bending moment NOTE:
X-axis is along the center line of the pipe.
I 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.
I 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.
3.3     LOADS FOR LEAK RATE EVALUATION The normal operating loads for leak rate predictions are calculated by the following equations:
3.3 LOADS FOR LEAK RATE EVALUATION The normal operating loads for leak rate predictions are calculated by the following equations:
F     =     Fow + FTH + Fp                                                         (3-3)
F  
Mx    =     (Mx)ow + (Mx)TH                                                         (3-4)
=
Mv    =     (Mv)ow + (MvhH                                                         (3-5)
Fow + FTH + Fp Mx
Mz    =     (Mz)ow + (MzhH                                                         (3-6)
=
The subscripts of the above equations represent the following loading cases:
(Mx)ow + (Mx)TH Mv
ow           =     deadweight TH          =      normal thermal expansion p            =      load due to internal pressure This method of combining loads is often referred to as the algebraic sum method (References 3-1 and 3-2).
=
(Mv)ow + (MvhH Mz
=
(Mz)ow + (MzhH The subscripts of the above equations represent the following loading cases:
ow TH p
=  
=
=
deadweight normal thermal expansion load due to internal pressure (3-3)
(3-4)
(3-5)
(3-6)
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 Table 3-1 at all the weld locations identified in Figure 3-2. The as-built dimensions are also given in Table 3-1.
The loads based on this method of combination are provided in Table 3-1 at all the weld locations identified in Figure 3-2. The as-built dimensions are also given in Table 3-1.
Pipe Geometry and Loading                                                               March 2019 WCAP-15550-N P                                                                           Revision 2
Pipe Geometry and Loading WCAP-15550-N P March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                         3-3 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
WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-3 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) 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, SSE INERTIA and seismic anchor motion (SAM) loads are combined based on individual absolute values as shown below.
(../2) margin should be reduced to 1.0 if the deadweight, thermal expansion, internal pressure, pressure expansion, SSE INERTIA and seismic anchor motion (SAM) loads are combined based on individual absolute values as shown below.
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:
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.
F   = I Fow I + I FTH I + I Fp I + I FssEINERTIA I + I FssEAM I                 (3-7)
The absolute summation of loads is shown in the following equations:
Mx = I (Mx)ow I + I (MxhH I + I (Mx)ssEINERTIA I + I (Mx)ssEAM I               (3-8)
F = I Fow I + I FTH I + I Fp I + I FssEINERTIA I + I FssEAM I Mx = I (Mx)ow I + I (MxhH I + I (Mx)ssEINERTIA I + I (Mx)ssEAM I Mv = I (Mv)ow I+ I (MvhH I+ I (Mv)ssEINERTIAI + I (Mv)ssEAM I Mz = I (Mz)ow I + I (Mz)TH I + I (Mz)ssEINERTIA I + I (Mz)ssEAM I (3-7)
Mv = I (Mv)ow I+ I (MvhH I+ I (Mv)ssEINERTIAI + I (Mv)ssEAM I                   (3-9)
(3-8)
Mz = I (Mz)ow I + I (Mz)TH I + I (Mz)ssEINERTIA I + I (Mz)ssEAM I             (3-10) where subscript SSEINERTIA refers to safe shutdown earthquake inertia, SSEAM is safe shutdown earthquake anchor motion, respectively.
(3-9)
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 Table 3-2.
(3-10) 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 Table 3-2.  
==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.
==3.5 REFERENCES==
3-2       NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.
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.
Pipe Geometry and Loading                                                           March 2019 WCAP-15550-N P                                                                       Revision 2
3-2 NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.
Pipe Geometry and Loading WCAP-15550-N P March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                               3-4 Table 3-1 Dimensions, Normal Loads and Stresses for Surry Units 1 and 2 3       Outside       Minimum      Axial Loadb      Moment        Total Stress Location Diameter (in) Thickness (in)    (kips)        (in-kips)          (ksi) 1             34.00         2.395          1482            19815          17.51 2              34.00         2.395          1482              837            6.71 3             34.00         2.395          1482            9661            11.73 4             37.75           3.270        1614            14956            9.87 5             37.625         3.208        1628              7852            7.55 6             36.32         2.555          1605            6815            9.11 7             36.32         2.555          1599            6786            9.07 8               36.32         2.555          1709              1071            6.81 9             37.625         3.208        1709              2666            5.90 10             37.625         3.208        1844              9466            8.76 11             32.26         2.270        1365              2588            8.11 12             32.26         2.270        1365            2927            8.33 13             32.26         2.270        1366            2373            7.97 14             33.60         2.940        1366              3636            6.64 15             33.60         2.940        1363              4955            7.29 Notes:
WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-4 Table 3-1 Dimensions, Normal Loads and Stresses for Surry Units 1 and 2 Location 3
: a.     See Figure 3-2
Outside Diameter (in) 1 34.00 2
: b.     Included Pressure
34.00 3
_Pipe Geometry and Loading                                                           March 2019 WCAP-15550-N P                                                                         Revision 2
34.00 4
37.75 5
37.625 6
36.32 7
36.32 8
36.32 9
37.625 10 37.625 11 32.26 12 32.26 13 32.26 14 33.60 15 33.60 Notes:
: a.
See Figure 3-2
: b.
Included Pressure
_Pipe Geometry and Loading WCAP-15550-N P Minimum Thickness (in) 2.395 2.395 2.395 3.270 3.208 2.555 2.555 2.555 3.208 3.208 2.270 2.270 2.270 2.940 2.940 Axial Loadb Moment (kips)
(in-kips) 1482 19815 1482 837 1482 9661 1614 14956 1628 7852 1605 6815 1599 6786 1709 1071 1709 2666 1844 9466 1365 2588 1365 2927 1366 2373 1366 3636 1363 4955 Total Stress (ksi) 17.51 6.71 11.73 9.87 7.55 9.11 9.07 6.81 5.90 8.76 8.11 8.33 7.97 6.64 7.29 March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                               3-5 Table 3-2 Faulted Loads and Stresses for Surry Units 1 and 2 Locationa,b       Axial Loadc         Moment           Total Stress (kips)          (in-kips)           (ksi) 1                1640            24646             20.93 2                  1640            2652               8.41 3                1639            12918             14.25 4                  1941            20101             12.62 5                1927            22956             13.89 6                1870            15673             14.23 7                1864            9928             11.52 8                1839            8475             10.75 9                1839            11375             9.43 10                1885            14032             10.53 11                1413            7850             11.84 12                1413            7390             11.54 13                1420            6032             10.66 14                1422            7923               8.99 15                1418            10829             10.42 Notes:
WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 3-2 Faulted Loads and Stresses for Surry Units 1 and 2 Locationa,b Axial Loadc (kips) 1 1640 2
: a. See Figure 3-2
1640 3
: b. See Table 3-1 for dimensions
1639 4
: c. Included Pressure Pipe Geometry and Loading                                                              March 2019 WCAP-15550-N P                                                                          Revision 2
1941 5
1927 6
1870 7
1864 8
1839 9
1839 10 1885 11 1413 12 1413 13 1420 14 1422 15 1418 Notes:
: a.
See Figure 3-2
: b.
See Table 3-1 for dimensions
: c.
Included Pressure Pipe Geometry and Loading WCAP-15550-N P Moment Total Stress (in-kips)
(ksi) 24646 20.93 2652 8.41 12918 14.25 20101 12.62 22956 13.89 15673 14.23 9928 11.52 8475 10.75 11375 9.43 14032 10.53 7850 11.84 7390 11.54 6032 10.66 7923 8.99 10829 10.42 3-5 March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                             3-6 Crack
WESTINGHOUSE NON-PROPRIETARY CLASS 3  
                                      '
------------------------------------------------- *)
(-
(-
            -------------------------------------------------
Location 1 Normal Loadsa Forcec:
                                                                  *)    M
1482 kips Bending Moment:
: 1. OD
19815 in-kips a See Table 3-1 b See Table 3-2 Faulted Loadsb Forcec:
                                                                                        .1 ooa = 34.00 in Location 1                                                                 ta = 2.395 in Normal Loadsa                             Faulted Loadsb Forcec:               1482 kips          Forcec:        1640 kips Bending Moment:       19815 in-kips      Bending Moment: 24646 in-kips a See Table 3-1 b See Table 3-2 c Includes the force due to a pressure of 2250 psia Figure 3-1 Hot Leg Coolant Pipe Pipe Geometry and Loading                                                            March 2019 WCAP-15550-N P                                                                        Revision 2
1640 kips Bending Moment:
24646 in-kips c Includes the force due to a pressure of 2250 psia Pipe Geometry and Loading WCAP-15550-N P Figure 3-1 Hot Leg Coolant Pipe M
3-6 Crack
: 1.
OD
.1 ooa = 34.00 in ta
= 2.395 in March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                           3-7
WESTINGHOUSE NON-PROPRIETARY CLASS 3
* Reactor Pressure Vessel
* Reactor Pressure Vessel  
                                    .....-coLDLEG
.....-coLDLEG  
                                        ...--@
\\_
                                          ...--@
Reactor Coolant Pump  
                                          \ _ Reactor Coolant Pump
\\...._ __ Steam Generator HOT LEG CROSS-OVER LEG COLD LEG CROSSOVER LEG Temperature 609.1 °F Pressure: 2250 psia Temperature 542.6°F Pressure: 2250 psia Temperature 542.9°F Pressure: 2250 psia Figure 3-2 Schematic Diagram of Surry Units 1 and 2 Primary Loop Showing Weld Locations Pipe Geometry and Loading WCAP-15550-NP 3-7 March 2019 Revision 2  
                    \....__ _ Steam Generator CROSSOVER LEG HOT LEG                            Temperature 609.1 °F Pressure: 2250 psia CROSS-OVER LEG                    Temperature 542.6°F Pressure: 2250 psia COLD LEG                          Temperature 542.9°F Pressure: 2250 psia Figure 3-2 Schematic Diagram of Surry Units 1 and 2 Primary Loop Showing Weld Locations Pipe Geometry and Loading                                                         March 2019 WCAP-15550-NP                                                                     Revision 2


WESTINGHOUSE NON-PROPRIETARY CLASS 3                               4-1 4.0       MATERIAL CHARACTERIZATION 4.1       PRIMARY LOOP PIPE AND FITTINGS MATERIALS The primary loop pipe is A376-TP316 and the elbow fittings are A351-CF8M for Surry Units 1 and 2.
WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-1 4.0 MATERIAL CHARACTERIZATION 4.1 PRIMARY LOOP PIPE AND FITTINGS MATERIALS The primary loop pipe is A376-TP316 and the elbow fittings are A351-CF8M for Surry Units 1 and 2.
4.2       TENSILE PROPERTIES The Pipe Certified Materials Test Reports (CMTRs) for Surry Units 1 and 2 were used to establish the tensile properties for the leak-before-break analyses. The CMTRs include tensile properties at room temperature and/or at 650°F for each of the heats of material. These properties are given in Table 4-1 for the Surry Unit 1 pipe, Table 4-2 for the Surry Unit 2 pipe, Table 4-3 for Unit 1 elbows and in Table 4-4 for Unit 2 elbows.
4.2 TENSILE PROPERTIES The Pipe Certified Materials Test Reports (CMTRs) for Surry Units 1 and 2 were used to establish the tensile properties for the leak-before-break analyses. The CMTRs include tensile properties at room temperature and/or at 650°F for each of the heats of material.
These properties are given in Table 4-1 for the Surry Unit 1 pipe, Table 4-2 for the Surry Unit 2 pipe, Table 4-3 for Unit 1 elbows and in Table 4-4 for Unit 2 elbows.
The representative properties at 609°F (represents actual 609.1 °F for Hot Leg) for the pipe were established from the tensile properties at 650°F given in Tables 4-1 and 4-2 by utilizing Section II of the 1999ASME Boiler and Pressure Vessel Code (Reference 4-1). Code tensile properties at 609°F were obtained by interpolating between the 600°F and 650°F tensile properties. Ratios of the code tensile properties at 609°F to the corresponding tensile properties at 650°F were then applied to the 650°F tensile properties given in Tables 4-1 and 4-2 to obtain the plant specific properties for A376-TP316 at 609°F. It should be noted that there is no significant impact by using the 1999 ASME Code Section II edition for material properties for the LBB analysis, as compared to the Surry ASME code of record.
The representative properties at 609°F (represents actual 609.1 °F for Hot Leg) for the pipe were established from the tensile properties at 650°F given in Tables 4-1 and 4-2 by utilizing Section II of the 1999ASME Boiler and Pressure Vessel Code (Reference 4-1). Code tensile properties at 609°F were obtained by interpolating between the 600°F and 650°F tensile properties. Ratios of the code tensile properties at 609°F to the corresponding tensile properties at 650°F were then applied to the 650°F tensile properties given in Tables 4-1 and 4-2 to obtain the plant specific properties for A376-TP316 at 609°F. It should be noted that there is no significant impact by using the 1999 ASME Code Section II edition for material properties for the LBB analysis, as compared to the Surry ASME code of record.
The representative properties at 609°F (represents actual 609.1 °F for Hot Leg) and 543°F (represents actual 542.9°F for Cold Leg and 542.6°F for Crossover Leg) for the elbows were established from the tensile properties at room temperature properties given in Tables 4-3 and 4-4 by utilizing Section II of the 1999 ASME Boiler and Pressure Vessel Code (Reference 4-1).
The representative properties at 609°F (represents actual 609.1 °F for Hot Leg) and 543°F (represents actual 542.9°F for Cold Leg and 542.6°F for Crossover Leg) for the elbows were established from the tensile properties at room temperature properties given in Tables 4-3 and 4-4 by utilizing Section II of the 1999 ASME Boiler and Pressure Vessel Code (Reference 4-1).
Line 223: Line 302:
The average and lower bound yield strengths and ultimate strengths are given in Table 4-5. The ASME Code moduli of elasticity values are also given, and Poisson's ratio was taken as 0.3.
The average and lower bound yield strengths and ultimate strengths are given in Table 4-5. The ASME Code moduli of elasticity values are also given, and Poisson's ratio was taken as 0.3.
Updated CMTRs from Replacement Steam Generator (RSG) replacement elbows on Surry Unit 1 are also considered. The added tensile properties are shown in Table 4-3 (marked by grey-shaded color). Of these added properties, the minimum Yield Strength is 38350 psi and the minimum Ultimate Strength is 81200 psi. These values are bounded by lower bound values in Table 4-5. It has also been reviewed that the updated tensile data has negligible impact to the average Yield Strength in Table 4-5.
Updated CMTRs from Replacement Steam Generator (RSG) replacement elbows on Surry Unit 1 are also considered. The added tensile properties are shown in Table 4-3 (marked by grey-shaded color). Of these added properties, the minimum Yield Strength is 38350 psi and the minimum Ultimate Strength is 81200 psi. These values are bounded by lower bound values in Table 4-5. It has also been reviewed that the updated tensile data has negligible impact to the average Yield Strength in Table 4-5.
4.3       FRACTURE TOUGHNESS PROPERTIES The pre-service fracture toughness (J) of cast stainless steels that are of interest are in terms of J 1c (J at Crack Initiation) and have been found to be very high at 600°F. [
4.3 FRACTURE TOUGHNESS PROPERTIES The pre-service fracture toughness (J) of cast stainless steels that are of interest are in terms of J1c (J at Crack Initiation) and have been found to be very high at 600°F. [
                                                                                ]a,c,e However, cast Material Characterization                                                                March 2019 WCAP-15550-N P                                                                            Revision 2
Material Characterization WCAP-15550-N P
]a,c,e However, cast March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                   4-2 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 embrittlement, 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.
WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-2 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 embrittlement, 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.
The susceptibility of the material to thermal aging increases with increasing ferrite contents. The molybdenum bearing CF8M shows increased susceptibility to thermal aging.
The susceptibility of the material to thermal aging increases with increasing ferrite contents. The molybdenum bearing CF8M shows increased susceptibility to thermal aging.
The method described below was used to calculate the end of life toughness properties for the cast material of the Surry Units 1 and 2 primary coolant loop piping and elbows.
The method described below was used to calculate the end of life toughness properties for the cast material of the Surry Units 1 and 2 primary coolant loop piping and elbows.
In 1994, the Argonne National Laboratory (ANL) completed an extensive research program in assessing the extent of thermal aging of cast stainless steel materials (Reference 4-2). 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 hours (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 hours 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 4-3).
In 1994, the Argonne National Laboratory (ANL) completed an extensive research program in assessing the extent of thermal aging of cast stainless steel materials (Reference 4-2). 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 hours (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 hours 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 4-3).
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 in Reference 4-3 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.
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 in Reference 4-3 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.
Based on NUREG/CR-4513, Revision 2, the fracture toughness correlations used for the full aged condition is applicable for plants operating at and beyond 15 EFPY (Effective Full Power Years) for the CF8M materials (elbows for Surry Units 1 and 2). As of January 2017, Surry Units 1 and 2 are operating at 33. 78 and 33.69 EFPY, respectively. Therefore, the use of the fracture toughness correlations described below is applicable for the fully aged or saturated condition of the Surry Units 1 and 2 elbow materials made of CF8M.
Based on NUREG/CR-4513, Revision 2, the fracture toughness correlations used for the full aged condition is applicable for plants operating at and beyond 15 EFPY (Effective Full Power Years) for the CF8M materials (elbows for Surry Units 1 and 2). As of January 2017, Surry Units 1 and 2 are operating at 33. 78 and 33.69 EFPY, respectively. Therefore, the use of the fracture toughness correlations described below is applicable for the fully aged or saturated condition of the Surry Units 1 and 2 elbow materials made of CF8M.
The chemical compositions of the Surry Units 1 and 2 primary loop elbow fitting material are available from CMTRs and are provided in Table 4-6 and Table 4-7 of this report. The following equations are taken from Reference 4-3 and applicable for CF8M type material:
The chemical compositions of the Surry Units 1 and 2 primary loop elbow fitting material are available from CMTRs and are provided in Table 4-6 and Table 4-7 of this report. The following equations are taken from Reference 4-3 and applicable for CF8M type material:
Creq = Cr+ 1.21 (Mo) + 0.48(Si) - 4.99 = (Chromium equivalent)                             (4-1) 2 Nieq =(Ni)+ 0.11(Mn)- 0.0086(Mn) + 18.4(N) + 24.5(C) + 2.77 = (Nickel equivalent)           (4-2)
Creq = Cr+ 1.21 (Mo) + 0.48(Si) - 4.99 = (Chromium equivalent)
Oc =100.3(Creq I Nieq )2-170.72(Creq I Nieq )+74.22 = (Ferrite Content)                     (4-3)
(4-1)
Material Characterization                                                               March 2019 WCAP-15550-NP                                                                             Revision 2
Nieq =(Ni)+ 0.11(Mn)- 0.0086(Mn)2 + 18.4(N) + 24.5(C) + 2.77 = (Nickel equivalent)
(4-2)
Oc =100.3(Creq I Nieq )2-170.72(Creq I Nieq )+74.22 = (Ferrite Content)
(4-3)
Material Characterization WCAP-15550-NP March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                     4-3 where the elements are in percent weight and oc is ferrite in percent volume.
WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-3 where the elements are in percent weight and oc is ferrite in percent volume.
The saturation room temperature (RT) impact energies of the cast stainless steel materials were determined from the chemical compositions available from CMTRs and provided in Tables 4-6 and 4-7.
The saturation room temperature (RT) impact energies of the cast stainless steel materials were determined from the chemical compositions available from CMTRs and provided in Tables 4-6 and 4-7.
For CF8M steel with < 10% Ni, the saturation value of RT impact energy CVsat (J/cm 2 ) is the lower value determined from log10CVsat = 0.27 + 2.81 exp (-0.022$)                                                         (4-4) where the material parameter $ is expressed as
For CF8M steel with < 10% Ni, the saturation value of RT impact energy CV sat (J/cm2) is the lower value determined from log10CVsat = 0.27 + 2.81 exp (-0.022$)
  $ = oc (Ni + Si + Mn)2(C + 0.4N)/5.0                                                             (4-5) and from log 10CVsat = 7.28 - 0.011oc - 0.185Cr - 0.369Mo - 0.451 Si - 0.007Ni - 4.71 (C + 0.4N)         (4-6)
where the material parameter $ is expressed as  
For CF8M steel with ~ 10% Ni, the saturation value of RT impact energy Cvsat (J/cm 2
$ = oc (Ni + Si + Mn)2(C + 0.4N)/5.0 and from (4-4)
                                                                                                ) is the lower value determined from log10CVsat = 0.84 + 2.54 exp (-0.047$)                                                         (4-7) where the material parameter $ is expressed as
(4-5) log10CVsat = 7.28 - 0.011oc - 0.185Cr - 0.369Mo - 0.451 Si - 0.007Ni - 4.71 (C + 0.4N)
  $ = oc (Ni + Si + Mn)2(C + 0.4N)/5.0                                                             (4-8) and from log10CVsat = 7 .28 - 0.011 oc - 0.185Cr - 0.369Mo - 0.451 Si - 0.007Ni - 4. 71 (C + 0.4N)      (4-9)
(4-6)
The saturation J-R curve at RT, for static-cast CF8M steel is given by 135                            2 (4-10)
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$)
Jd = 1.44 (CVsat) ' (.!iat     for CVsat < 35 J/cm for CVsat ~ 35 J/cm 2
where the material parameter $ is expressed as  
(4-11) n = 0.20 + 0.08 log10 (CVsat)                                                                (4-12) where Jct is the "deformation J" in kJ/m 2 and Lia is the crack extension in mm.
$ = oc (Ni + Si + Mn)2(C + 0.4N)/5.0 and from log10CVsat = 7.28 - 0.011 oc - 0.185Cr - 0.369Mo - 0.451 Si - 0.007Ni - 4. 71 (C + 0.4N)
The saturation J-R curve at 290-320&deg;C (554-608&deg;F), for static-cast CF8M steel is given by 98                              2 Jd = 5.5 (CVsal' (Liat       for CVsat < 46 J/cm                                              (4-13) 41 for CVsat ~ 46 J/cm 2
The saturation J-R curve at RT, for static-cast CF8M steel is given by Jd = 1.44 (CV sat) 1'35(.!iat for CVsat < 35 J/cm2 for CVsat ~ 35 J/cm 2
Jd = 49 (CVsal' (Lia)"                                                                         (4-14) n = 0.19 + 0.07 log10 (CVsat)                                                                (4-15) where Jct is the "deformation J" in kJ/m 2 and Lia is the crack extension in mm.
n = 0.20 + 0.08 log10 (CVsat) where Jct is the "deformation J" in kJ/m2 and Lia is the crack extension in mm.
Material Characterization                                                                 March 2019 WCAP-15550-N P                                                                               Revision 2
The saturation J-R curve at 290-320&deg;C (554-608&deg;F), for static-cast CF8M steel is given by Jd = 5.5 (CVsal'98(Liat for CVsat < 46 J/cm2 Jd = 49 (CVsal'41 (Lia)"
for CVsat ~ 46 J/cm2 n = 0.19 + 0.07 log10 (CVsat) where Jct is the "deformation J" in kJ/m2 and Lia is the crack extension in mm.
(4-7)
(4-8)
(4-9)
(4-10)
(4-11)
(4-12)
(4-13)
(4-14)
(4-15)
Material Characterization WCAP-15550-N P March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                               4-4 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 1
WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-4 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.
metal because the yield stress for the weld materials is much higher at the temperature.
Therefore, weld regions are less limiting than the cast material.
In the fracture mechanics analyses that follow, the fracture toughness properties given in Table 4-8 will be used as the criteria against which the applied fracture toughness values will be compared.
In the fracture mechanics analyses that follow, the fracture toughness properties given in Table 4-8 will be used as the criteria against which the applied fracture toughness values will be compared.
As indicated in the record of revisions table, this stress report is revised to address CMTR errors and to address updated CMTRs from RSG replacement elbows on Unit 1.
As indicated in the record of revisions table, this stress report is revised to address CMTR errors and to address updated CMTRs from RSG replacement elbows on Unit 1.
As shown in the revised Tables 4-6 and 4-7, the corrections of CMTR errors (shown in bold font) and updates due to additional CMTR's (marked by grey-shaded color) have been included in this stress report. It has been reviewed that the revisions do not impact the evaluation results in this Section 4 material characterization. The summary of limiting fracture toughness properties provided in Table 4-8 remains valid and applicable. The revisions also do not affect the conclusions of this stress report.
As shown in the revised Tables 4-6 and 4-7, the corrections of CMTR errors (shown in bold font) and updates due to additional CMTR's (marked by grey-shaded color) have been included in this stress report. It has been reviewed that the revisions do not impact the evaluation results in this Section 4 material characterization. The summary of limiting fracture toughness properties provided in Table 4-8 remains valid and applicable. The revisions also do not affect the conclusions of this stress report.  
 
==4.4      REFERENCES==


4-1     ASME Boiler and Pressure Vessel Code, An International Code, Section II, Materials, Part D-Properties, 1999 Addenda, July 1, 1999.
==4.4 REFERENCES==
4-2     0. K. Chopra and W. J. Shack, "Assessment of Thermal Embrittlement of Cast Stainless Steels," NUREG/CR-6177, U.S. Nuclear Regulatory Commission, Washington, DC, May 1994.
4-1 ASME Boiler and Pressure Vessel Code, An International Code, Section II, Materials, Part D-Properties, 1999 Addenda, July 1, 1999.
4-3     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.
4-2
: 0. K. Chopra and W. J. Shack, "Assessment of Thermal Embrittlement of Cast Stainless Steels," NUREG/CR-6177, U.S. Nuclear Regulatory Commission, Washington, DC, May 1994.
4-3
: 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.
1 In the report all the applied J values were conservatively determined by using base metal strength properties.
1 In the report all the applied J values were conservatively determined by using base metal strength properties.
Material Characterization                                                               March 2019 WCAP-15550-N P                                                                           Revision 2
Material Characterization WCAP-15550-N P March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                               4-5 Table 4-1 Measured Tensile Properties (psi) for Surry Unit 1 Primary Loop Pipes At Room Temp.                          At 650&deg;F Heat No./Serial                     Yield            Ultimate          Yield          Ultimate Location                       Strength        Strength        Strength No.                         Strength F021212867Z         X-over Leg     43500            84200            26400          67000 F021212867Z         X-over Leg     43900            82500              NIA              NIA F021512894Y         X-over Leg     40000            84800            21700          66800 F021512894Y         X-over Leg     41900            87700              NIA              NIA F022212900Y         X-over Leg     44000            86500            21500          66200 F022212900Y         X-over Leg     41500            83200              NIA              NIA 877712883X         X-over Leg     36100            78200            21000          62000 877712883X         X-over Leg     38500            77800              NIA              NIA 878512881X         X-over Leg     36100            74200            20400          57200 878512881X         X-over Leg     39700            79800              NIA              NIA 891512885X         X-over Leg     38500            77400            24200          62300 891512885X         X-over Leg     38600            77200              NIA              NIA 877712874Y           Cold Leg     35300            79200            24100          65600 877712874Y           Cold Leg     34900            78200              NIA              NIA 891312875Y           Cold Leg     35100            78400            24300          68500 891312875Y           Cold Leg     41100            84800              NIA              NIA F022812996           Cold Leg     45100            88600            24700          69800 F022812996           Cold Leg     43900            87400              NIA              NIA F022912952           Cold Leg     42200              85300            24500          68200 F022912952           Cold Leg     39700              83400            NIA              NIA F016212859           Cold Leg     46000              83500            27600            69400 F016212859           Cold Leg     52900              87900            NIA              NIA F016212860           Cold Leg     43700              83900            NIA              NIA F016212860           Cold Leg     47400              90400            NIA              NIA F021612861           Cold Leg     40600              83400            21300            66600 F021612861           Cold Leg     40000              80500            NIA              NIA F022712947           Cold Leg     42500              83700            20900            66600 F022712947           Cold Leg     43200              87100            NIA              NIA F022912951           Cold Leg     42000              84400            24500            68200 F022912951           Cold Leg     44500              86800              NIA            NIA 878512963X             Hot Leg     33400              75700            22000            61000 878512963X             Hot Leg     33600              75300              NIA            NIA FO 19012847Y            Hot Leg     42000              88800            21300            58200 F019012847Y             Hot Leg     43000              86000              NIA            NIA F005812629             Hot Leg     41200              85200            27000            74000 F005812629           Hot Leg     44300              89500              NIA            NIA F018812845             Hot Leg     40900              83000            26100            67000 F018812845             Hot Leg     43000              84000              NIA            NIA E148213353             Hot Leg     41800              84400            24500            66200 E148213353             Hot Leg     41700              84600              NIA            NIA Note:   NIA= Not Applicable Material Characterization                                                               March 2019 WCAP-15550-NP                                                                           Revision 2
WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-5 Table 4-1 Measured Tensile Properties (psi) for Surry Unit 1 Primary Loop Pipes Heat No./Serial Location No.
F021212867Z X-over Leg F021212867Z X-over Leg F021512894Y X-over Leg F021512894Y X-over Leg F022212900Y X-over Leg F022212900Y X-over Leg 877712883X X-over Leg 877712883X X-over Leg 878512881X X-over Leg 878512881X X-over Leg 891512885X X-over Leg 891512885X X-over Leg 877712874Y Cold Leg 877712874Y Cold Leg 891312875Y Cold Leg 891312875Y Cold Leg F022812996 Cold Leg F022812996 Cold Leg F022912952 Cold Leg F022912952 Cold Leg F016212859 Cold Leg F016212859 Cold Leg F016212860 Cold Leg F016212860 Cold Leg F021612861 Cold Leg F021612861 Cold Leg F022712947 Cold Leg F022712947 Cold Leg F022912951 Cold Leg F022912951 Cold Leg 878512963X Hot Leg 878512963X Hot Leg FO 1901284 7Y Hot Leg F019012847Y Hot Leg F005812629 Hot Leg F005812629 Hot Leg F018812845 Hot Leg F018812845 Hot Leg E148213353 Hot Leg E148213353 Hot Leg Note:
NIA= Not Applicable Material Characterization WCAP-15550-NP At Room Temp.
Yield Ultimate Strength Strength 43500 84200 43900 82500 40000 84800 41900 87700 44000 86500 41500 83200 36100 78200 38500 77800 36100 74200 39700 79800 38500 77400 38600 77200 35300 79200 34900 78200 35100 78400 41100 84800 45100 88600 43900 87400 42200 85300 39700 83400 46000 83500 52900 87900 43700 83900 47400 90400 40600 83400 40000 80500 42500 83700 43200 87100 42000 84400 44500 86800 33400 75700 33600 75300 42000 88800 43000 86000 41200 85200 44300 89500 40900 83000 43000 84000 41800 84400 41700 84600 At 650&deg;F Yield Strength 26400 NIA 21700 NIA 21500 NIA 21000 NIA 20400 NIA 24200 NIA 24100 NIA 24300 NIA 24700 NIA 24500 NIA 27600 NIA NIA NIA 21300 NIA 20900 NIA 24500 NIA 22000 NIA 21300 NIA 27000 NIA 26100 NIA 24500 NIA Ultimate Strength 67000 NIA 66800 NIA 66200 NIA 62000 NIA 57200 NIA 62300 NIA 65600 NIA 68500 NIA 69800 NIA 68200 NIA 69400 NIA NIA NIA 66600 NIA 66600 NIA 68200 NIA 61000 NIA 58200 NIA 74000 NIA 67000 NIA 66200 NIA March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                               4-6 Table 4-2 Measured Tensile Properties (psi) for Surry Unit 2 Primary Loop Pipes At Room Temperature                      At 650&deg;F Heat No./Serial                       Yield          Ultimate          Yield          Ultimate Location                                       Strength        Strength No.                         Strength          Strength F0213/2889         Hot Leg     43200            84800            23700          69800 F0213/2889         Hot Leg     44000            88400              N/A              N/A F0213/2890X         Hot Leg     41900            82500              N/A              N/A F0213/2890X         Hot Leg     46500            84500              N/A              N/A F0213/2891X         Hot Leg     44000            86000              N/A              N/A F0213/2891X         Hot Leg     41000            83000              N/A              N/A F0227/2897X         Hot Leg     41800            86800              N/A              N/A F0227/2897X         Hot Leg     42500            85500              N/A              N/A F0373/3169         Cold Leg     41400            85800            25600          71900 F0373/3169         Cold Leg     44800            89700              N/A              N/A F0221 /2866X     X-over Leg     44000            83800              N/A              N/A F0221/2866Y       X-over Leg     42700            86000            21600          65200 F0222/2900X       X-over Leg     44000            86500            21500          66200 F0222/2900X       X-over Leg     41500            83200              N/A              N/A 52154/2843         Cold Leg     43500            86500            23100          57400 52154/2843         Cold Leg     33300            75600              N/A              N/A F0228/2948         Cold Leg     42500            87400            24700          69800 F0228/2948         Cold Leg     45000            87500              N/A              N/A F0229/2994         Cold Leg     41000            82800            24500          68200 F0229/2994         Cold Leg     45000            87900              N/A              N/A E1490/3347Y         Hot Leg     43100            86000            23700          68000 E1490/3347Y         Hot Leg     43000            82700              N/A              N/A F0189/2869Y       X-over Leg     37700            80600              N/A              N/A F0189/2869Z       X-over Leg     44100            91000              N/A              N/A F0189/2868X       X-over Leg     37700            80600            25200          70000 F0189/2868X       X-over Leg     44100            91000              N/A              N/A F0226/2946         Cold Leg     43000              85000            21500          66400 F0226/2946         Cold Leg     42100              86000            N/A              N/A K2011/3683X         Cold Leg     32100            75900            20600          56200 1<2011 /3683X       Cold Leg     38400            81900            N/A              N/A E1478/3257         Cold Leg     38000            82900            25300          72400 E1478/3257         Cold Leg     42400              84900            N/A              N/A V0629/3262         Cold Leg     48400              88800            21100          67400 V0629/3262         Cold Leg     41300            82200            N/A              N/A F0229/2953         Cold Leg     41000              84900            24500          68200 F0229/2953         Cold Leg     45000            86200            N/A              N/A l
l WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-6 Table 4-2 Measured Tensile Properties (psi) for Surry Unit 2 Primary Loop Pipes Heat No./Serial Location No.
F0215/2892Y         Hot Leg     43000            83000            21700          66800 F0215/2892Y         Hot Leg     42000            84600            N/A              N/A Note:   N/A = Not Applicable Material Characterization                                                             March 2019 WCAP-15550-NP                                                                           Revision 2
F0213/2889 Hot Leg F0213/2889 Hot Leg F0213/2890X Hot Leg F0213/2890X Hot Leg F0213/2891X Hot Leg F0213/2891X Hot Leg F0227/2897X Hot Leg F0227/2897X Hot Leg F0373/3169 Cold Leg F0373/3169 Cold Leg F0221 /2866X X-over Leg F0221/2866Y X-over Leg F0222/2900X X-over Leg F0222/2900X X-over Leg 52154/2843 Cold Leg 52154/2843 Cold Leg F0228/2948 Cold Leg F0228/2948 Cold Leg F0229/2994 Cold Leg F0229/2994 Cold Leg E1490/3347Y Hot Leg E1490/3347Y Hot Leg F0189/2869Y X-over Leg F0189/2869Z X-over Leg F0189/2868X X-over Leg F0189/2868X X-over Leg F0226/2946 Cold Leg F0226/2946 Cold Leg K2011/3683X Cold Leg 1<2011 /3683X Cold Leg E1478/3257 Cold Leg E1478/3257 Cold Leg V0629/3262 Cold Leg V0629/3262 Cold Leg F0229/2953 Cold Leg F0229/2953 Cold Leg F0215/2892Y Hot Leg F0215/2892Y Hot Leg Note:
N/A = Not Applicable Material Characterization WCAP-15550-NP At Room Temperature Yield Ultimate Strength Strength 43200 84800 44000 88400 41900 82500 46500 84500 44000 86000 41000 83000 41800 86800 42500 85500 41400 85800 44800 89700 44000 83800 42700 86000 44000 86500 41500 83200 43500 86500 33300 75600 42500 87400 45000 87500 41000 82800 45000 87900 43100 86000 43000 82700 37700 80600 44100 91000 37700 80600 44100 91000 43000 85000 42100 86000 32100 75900 38400 81900 38000 82900 42400 84900 48400 88800 41300 82200 41000 84900 45000 86200 43000 83000 42000 84600 Yield Strength 23700 N/A N/A N/A N/A N/A N/A N/A 25600 N/A N/A 21600 21500 N/A 23100 N/A 24700 N/A 24500 N/A 23700 N/A N/A N/A 25200 N/A 21500 N/A 20600 N/A 25300 N/A 21100 N/A 24500 N/A 21700 N/A At 650&deg;F Ultimate Strength 69800 N/A N/A N/A N/A N/A N/A N/A 71900 N/A N/A 65200 66200 N/A 57400 N/A 69800 N/A 68200 N/A 68000 N/A N/A N/A 70000 N/A 66400 N/A 56200 N/A 72400 N/A 67400 N/A 68200 N/A 66800 N/A March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                   4-7 Table 4-3 Measured Tensile Properties (psi) for Surry Unit 1 Primary Loop Elbows At Room Temperature Heat No.                 Location Yield Strength         Ultimate Strength 10360-1               X-over Leg               46500                   88000 10844-1               X-over Leg                 39000                 78000 11046-1               X-over Leg               43500                   87000 11246-1               X-over Leg               42000                   82500 11441-1               X-over Leg               48000                   88500 11937-1               X-over Leg               45000                   88500 10442-1               X-over Leg                 48000                   88500 11168-1               X-over Leg                 46500                   88000 12198-1               X-over Leg                 45000                   85000 12623-1               X-over Leg                 42000                   80000 10482-1               X-over Leg                 45750                   80250 10723-1               X-over Leg                 46200                   88600 29943-2                 Cold Leg                 39300                   77300 30690-1                 Cold Leg                 39700                   80200 29943-1                 Hot Leg                 38400                   74900 31011-5                 Hot Leg                 43000                   84800 28387-2                 Hot Leg                 42300                   85800 28387-1                 Cold Leg                 42300                   85800 30597-2               X-over Leg                 42300                   84800 10128-2               X-over Leg                 46500                   89000 10243-2               X-over Leg                 48000                   90000 Note:
WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 4-3 Measured Tensile Properties (psi) for Surry Unit 1 Primary Loop Elbows Heat No.
* tensile properties from the added heats are not included in the average tensile property calculations in Table 4-5, but the impact of the added tensile data are negligible.
Location At Room Temperature Yield Strength Ultimate Strength 10360-1 X-over Leg 46500 88000 10844-1 X-over Leg 39000 78000 11046-1 X-over Leg 43500 87000 11246-1 X-over Leg 42000 82500 11441-1 X-over Leg 48000 88500 11937-1 X-over Leg 45000 88500 10442-1 X-over Leg 48000 88500 11168-1 X-over Leg 46500 88000 12198-1 X-over Leg 45000 85000 12623-1 X-over Leg 42000 80000 10482-1 X-over Leg 45750 80250 10723-1 X-over Leg 46200 88600 29943-2 Cold Leg 39300 77300 30690-1 Cold Leg 39700 80200 29943-1 Hot Leg 38400 74900 31011-5 Hot Leg 43000 84800 28387-2 Hot Leg 42300 85800 28387-1 Cold Leg 42300 85800 30597-2 X-over Leg 42300 84800 10128-2 X-over Leg 46500 89000 10243-2 X-over Leg 48000 90000 Note:
Material Characterization                                                               March 2019 WCAP-15550-N P                                                                             Revision 2
tensile properties from the added heats are not included in the average tensile property calculations in Table 4-5, but the impact of the added tensile data are negligible.
4-7 Material Characterization WCAP-15550-N P March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                   4-8 Table 4-4 Measured Tensile Properties (psi) for Surry Unit 2 Primary Loop Elbows At Room Temperature Heat No.             Location Yield Strength        Ultimate Strength 30080-2               Hot Leg                  42400                    82400 32535-2               Cold Leg                  44100                    87500 31571-6               Cold Leg                  42800                    85600 14008-1             X-over Leg                 48000                    87000 14085-1              X-over Leg                 45000                    86000 14165-1              X-over Leg                   48000                  89000 12709-1              X-over Leg                   48000                  85500 14507-1              X-over Leg                   40500                  82000 13579-1              X-over Leg                   48000                  85500 13781-5              X-over Leg                   4500                  89500 13826-2              X-over Leg                   48000                  87500 30690-2                Hot Leg                   38100                  82200 31427-2                Hot Leg                   41250                  86250 14786-3              X-over Leg                   42000                  84500 14990-1              X-over Leg                   43500                  84500 15384-1              X-over Leg                   45000                  88500 15769-1              X-over Leg                   46500                  85500 30080-1              Cold Leg                   45100                  84200 12087-3              X-over Leg                   42000                  80500 12547-2              X-over Leg                   40500                  83500 13051-4              X-over Leg                 40500                   83500 Material Characterization                                                                March 2019 WCAP-15550-N P                                                                            Revision 2
WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-8 Table 4-4 Measured Tensile Properties (psi) for Surry Unit 2 Primary Loop Elbows Heat No.
30080-2 32535-2 31571-6 14008-1 14085-1 14165-1 12709-1 14507-1 13579-1 13781-5 13826-2 30690-2 31427-2 14786-3 14990-1 15384-1 15769-1 30080-1 12087-3 12547-2 13051-4 Material Characterization WCAP-15550-N P Location Hot Leg Cold Leg Cold Leg X-over Leg X-over Leg X-over Leg X-over Leg X-over Leg X-over Leg X-over Leg X-over Leg Hot Leg Hot Leg X-over Leg X-over Leg X-over Leg X-over Leg Cold Leg X-over Leg X-over Leg X-over Leg At Room Temperature Yield Strength 42400 44100 42800 48000 45000 48000 48000 40500 48000 4500 48000 38100 41250 42000 43500 45000 46500 45100 42000 40500 40500 Ultimate Strength 82400 87500 85600 87000 86000 89000 85500 82000 85500 89500 87500 82200 86250 84500 84500 88500 85500 84200 80500 83500 83500 March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                   4-9 Table 4-5 Mechanical Properties for Surry Units 1 and 2 Materials at Operating Temperatures Lower Bound Average Yield     Modulus of Temperature*                                            Yield      Ultimate Material                           Strength           Elasticity (OF)                                             Strength    Strength (psi)             (psi)
WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-9 Table 4-5 Mechanical Properties for Surry Units 1 and 2 Materials at Operating Temperatures Lower Bound Temperature*
                                                        -                    (psi)         (psi) 6 A376 TP316               609           23835         25.255 X 10          20,762       56,200 6
Average Yield Modulus of Material Strength Elasticity Yield Ultimate (OF)
609           27468         25.255 X 10          23,785       71,904 A351 CF8M                                                           6 543            28493          25.585 X 10          24,672       71,904 Poisson's ratio: 0.3 Note: Representative temperature. The actual temperatures are provided in Figure 3-2.
(psi)
Material Characterization                                                                 March 2019 WCAP-15550-NP                                                                               Revision 2
(psi)
Strength Strength (psi)
(psi)
A376 TP316 609 23835 25.255 X 106 20,762 56,200 609 27468 25.255 X 106 23,785 71,904 A351 CF8M 25.585 X 106 543 28493 24,672 71,904 Poisson's ratio: 0.3 Note: Representative temperature. The actual temperatures are provided in Figure 3-2.
Material Characterization WCAP-15550-NP March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                         4-10 Table 4-6 Chemistry and Fracture Toughness Elbow Properties of the Material Heats of Surry Unit 1 a,c,e Material Characterization                                                                                           March 2019 WCAP-15550-N P                                                                                                       Revision 2
WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 4-6 Chemistry and Fracture Toughness Elbow Properties of the Material Heats of Surry Unit 1 Material Characterization WCAP-15550-N P 4-10 March 2019 Revision 2 a,c,e


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                           4-11 Table 4-6 Chemistry and Fracture Toughness Elbow Properties of the Material Heats of Surry Unit 1 a,c,e Material Characterization                                                                                           March 2019 WCAP-15550-NP                                                                                                       Revision 2
WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 4-6 Chemistry and Fracture Toughness Elbow Properties of the Material Heats of Surry Unit 1 Material Characterization WCAP-15550-NP 4-11 March 2019 Revision 2 a,c,e


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                           4-12 Table 4-7 Chemistry and Fracture Toughness Elbow Properties of the Material l:feats of Surry Unit 2 a,c,e Material Characterization                                                                                           March 2019 WCAP-15550-N P                                                                                                       Revision 2
WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 4-7 Chemistry and Fracture Toughness Elbow Properties of the Material l:feats of Surry Unit 2 Material Characterization WCAP-15550-N P 4-12 March 2019 Revision 2 a,c,e


WESTINGHOUSE NON-PROPRIETARY CLASS 3                             4-13 Table 4-8 Fracture Toughness Properties for Surry Units 1 and 2 Primary Loops for Leak-Before-Break Evaluation at Critical Locations a,c,e Material Characterization                                                           March 2019 WCAP-15550-N P                                                                       Revision 2
WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 4-8 Fracture Toughness Properties for Surry Units 1 and 2 Primary Loops for Leak-Before-Break Evaluation at Critical Locations 4-13 a,c,e Material Characterization WCAP-15550-N P March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                         4-14 a .. c .. e Figure 4-1 Pre-Service J vs. ~a for SA351-CF8M Cast Stainless Steel at 600&deg;F Material Characterization                                                       March 2019 WCAP-15550-NP                                                                     Revision 2
WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-14 a.. c.. e Figure 4-1 Pre-Service J vs. ~a for SA351-CF8M Cast Stainless Steel at 600&deg;F Material Characterization WCAP-15550-NP March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                             5-1 5.0       CRITICAL LOCATION AND EVALUATION CRITERIA 5.1       CRITICAL LOCATIONS The leak-before-break (LBB) evaluation margins are to be demonstrated for the limiting locations (governing locations). Such locations are established based on the loads (Section 3.0) and the material properties established in Section 4.0. These locations are defined below for Surry Units 1 and 2. Table 3-2 as well as Figure 3-2 are used for this evaluation.
WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-1 5.0 CRITICAL LOCATION AND EVALUATION CRITERIA 5.1 CRITICAL LOCATIONS The leak-before-break (LBB) evaluation margins are to be demonstrated for the limiting locations (governing locations). Such locations are established based on the loads (Section 3.0) and the material properties established in Section 4.0. These locations are defined below for Surry Units 1 and 2. Table 3-2 as well as Figure 3-2 are used for this evaluation.
Critical Locations The highest stressed location for the entire primary loop is at Location 1 (in the Hot Leg) (See Figure 3-2) at the reactor vessel outlet nozzle to pipe weld. Location 1 is critical for all the weld locations of pipe.
Critical Locations The highest stressed location for the entire primary loop is at Location 1 (in the Hot Leg) (See Figure 3-2) at the reactor vessel outlet nozzle to pipe weld. Location 1 is critical for all the weld locations of pipe.
Since the elbows are made of cast materials, the critical locations for the elbows are: for the hot leg, the highest stressed location is at weld location 3; for the cross-over leg, the highest stressed location is at weld location 6; and for the cold leg; the highest stressed location is at weld location 15. It is thus concluded that the enveloping locations in Surry Units 1 and 2 for which LBB methodology is to be applied are locations 1, 3, 6 and 15. The tensile properties and the allowable toughness for the critical locations are shown in Tables 4-5 and 4-8.
Since the elbows are made of cast materials, the critical locations for the elbows are: for the hot leg, the highest stressed location is at weld location 3; for the cross-over leg, the highest stressed location is at weld location 6; and for the cold leg; the highest stressed location is at weld location 15. It is thus concluded that the enveloping locations in Surry Units 1 and 2 for which LBB methodology is to be applied are locations 1, 3, 6 and 15. The tensile properties and the allowable toughness for the critical locations are shown in Tables 4-5 and 4-8.
5.2     FRACTURE CRITERIA As will be discussed later, fracture mechanics analyses are made based on loads and postulated flaw sizes related to leakage. The stability criteria against which the calculated J and tearing modulus are compared are:
5.2 FRACTURE CRITERIA As will be discussed later, fracture mechanics analyses are made based on loads and postulated flaw sizes related to leakage. The stability criteria against which the calculated J and tearing modulus are compared are:
(1)   If Japp < J 1c, 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.
(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:
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 critical locations, the limit load method discussed in Section 7.0 was also used.
Japp  
Critical Location and Evaluation Criteria                                                 March 2019 WCAP-15550-NP                                                                               Revision 2
=
 
Applied J J1c  
WESTINGHOUSE NON-PROPRIETARY CLASS 3                                      6-1 6.0      LEAK RATE PREDICTIONS
=
J at Crack Initiation Tapp  
=
Applied Tearing Modulus Tmat  
=
Material Tearing Modulus Jmax =
Maximum J value of the material For critical locations, the limit load method discussed in Section 7.0 was also used.
Critical Location and Evaluation Criteria WCAP-15550-NP March 2019 Revision 2  


==6.1      INTRODUCTION==
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 flow through postulated through-wall cracks and present the leak rate calculation results for through-wall circumferential cracks.
The purpose of this section is to discuss the method which is used to predict the flow 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, L, to hydraulic diameter, DH, (UDH) is greater than [
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, L, to hydraulic diameter, DH, (UDH) is greater than [
6.3     CALCULATION METHOD The basic method used in the leak rate calculations is the method developed by [
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 [
The flow rate through a crack was calculated in the following manner.
                                                ]8-c,e was found from Figure 6-2 (taken from Reference 6-2). For all cases considered, since [                                                   ]8-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, c, 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:
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 [  
Ab~olute Pressure - 14. 7 = [                                                       (6-2)
]8-c,e was found from Figure 6-2 (taken from Reference 6-2). For all cases considered, since [  
Leak Rate Predictions                                                                       March 2019 WCAP-15550-N P                                                                               Revision 2
]8-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, c, 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:
Ab~olute Pressure - 14. 7 = [
Leak Rate Predictions WCAP-15550-N P (6-2)
March 2019 Revision 2  


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.
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 Table 3-1 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 above. The average material properties of Section 4.0 (see Table 4-5) were used for these calculations.
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 Table 3-1 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 above. The average material properties of Section 4.0 (see Table 4-5) 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 Surry Units 1 and 2. The flaw sizes so determined are called leakage flaw sizes.
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 Surry Units 1 and 2. The flaw sizes so determined are called leakage flaw sizes.
The Surry 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 10 on the leak rate, the flaw sizes (leakage flaw sizes) are determined which yield a leak rate of 10 gpm.
The Surry 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 10 on the leak rate, the flaw sizes (leakage flaw sizes) are determined which yield a leak rate of 10 gpm.  
 
==6.5      REFERENCES==


6-1 6-2     M. M, EI-Wakil, "Nuclear Heat Transport, International Textbook Company," New York, N.Y, 1971.
==6.5 REFERENCES==
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,"
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.
Section 11-1, NUREG/CR-3464, September 1983.
Leak Rate Predictions                                                                 March 2019 WCAP-15550-NP                                                                         Revision 2
Leak Rate Predictions WCAP-15550-NP March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                         6-3 Table 6-1   Flaw Sizes Yielding a Leak Rate of 10 gpm at the Governing Locations Location                           Leakage Flaw Size (in) 1                                         4.02 3                                         5.85 6                                         6.84 15                                         7.96 Note: The flaw size in the above table refers to the flaw length of through-wall circumferential crack.
WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3 Table 6-1 Flaw Sizes Yielding a Leak Rate of 10 gpm at the Governing Locations Location Leakage Flaw Size (in) 1 4.02 3
Leak Rate Predictions                                                                             March 2019 WCAP-15550-N P                                                                                     Revision 2
5.85 6
6.84 15 7.96 Note: The flaw size in the above table refers to the flaw length of through-wall circumferential crack.
Leak Rate Predictions WCAP-15550-N P March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                            6-4 a,c,e
; = -i -& ->... u 0...
    ;   -
      -=i
    -&
      ...-
      >
u
    ...0
    "'>
ti C
ti C
2 STAGNATION ENTHALPY       no2 Btu/lb)
2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-4 a,c,e STAGNATION ENTHALPY no2 Btu/lb)
Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures Leak Rate Predictions                                                           March 2019 WCAP-15550-N P                                                                   Revision 2
Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures Leak Rate Predictions WCAP-15550-N P March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                         6-5 a,c,e LENGTH/OIAMET'EA RATIO (L/D)
Figure 6-2 [
Figure 6-2 [            ia,c,e Pressure Ratio as a Function of UD Leak Rate Predictions                                                          March 2019 WCAP-15550-NP                                                                  Revision 2
Leak Rate Predictions WCAP-15550-NP WESTINGHOUSE NON-PROPRIETARY CLASS 3 LENGTH/OIAMET'EA RATIO (L/D) ia,c,e Pressure Ratio as a Function of UD 6-5 a,c,e March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                           6-6
WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-6
[
[
Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack Leak Rate Predictions                                                         March 2019 WCAP-15550-N P                                                                 Revision 2
Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack Leak Rate Predictions WCAP-15550-N P March 2019 Revision 2  


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 J 1c 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:
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.
dJ E Tapp= -d Xz a at where:
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:
Tapp            =        applied tearing modulus E                =        modulus of elasticity
where:
                          =        0.5 (cry+ cru) = flow stress a                =        crack length cry, cru        =        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.
Tapp E
a cry, cru
=
=
=
=
=
dJ E
Tapp= -d Xz a at applied tearing modulus modulus of elasticity 0.5 (cry+ cru) = flow stress crack length 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:
As discussed in Section 5.2 the local crack stability criteria is a two-step process:
( 1)     If Japp < J 1c, 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.,
( 1)
Fracture Mechanics Evaluation                                                                 March 2019 WCAP-15550-NP                                                                                   Revision 2
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.,
Fracture Mechanics Evaluation WCAP-15550-NP March 2019 Revision 2  


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.
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.
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 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:  
r,c,e
=
                =    0.5 (cry+ cru) = flow stress, psi 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.
r,c,e 0.5 (cry+ cru) = flow stress, psi The analytical model described above accurately accounts for the piping internal pressure as well as imposed axial force as they affect the limit moment.
Fracture Mechanics Evaluation                                                         March 2019 WCAP-15550-N P                                                                           Revision 2
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.
Fracture Mechanics Evaluation WCAP-15550-N P March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3        7-3 7.3     CRACK STABILITY EVALUATIONS Local Failure Mechanism:
WESTINGHOUSE NON-PROPRIETARY CLASS 3 7.3 CRACK STABILITY EVALUATIONS Local Failure Mechanism:
Fracture Mechanics Evaluation                                 March 2019 WCAP-15550-N P                                                 Revision 2
Fracture Mechanics Evaluation WCAP-15550-N P 7-3 March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                   7-4 Global failure mechanism:
WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-4 Global failure mechanism:
A stability analysis based on limit load was performed for all the critical locations (locations 1, 3, 6 and 15) as described in Section 7.2. The field welds are made of GTAW and SMAW combination weld. The shop welds are made of GTAW, SMAW or SAW combination weld.
A stability analysis based on limit load was performed for all the critical locations (locations 1, 3, 6 and 15) as described in Section 7.2.
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, 3 and 15. Shop weld is at critical location 6. The "Z" factor correction for SMAW was applied (References 7-5 and 7-6) at the field weld critical locations (locations 1, 3 and 15) and the "Z" factor correction for SAW was applied (References 7-5 and 7-6) at the shop weld location (location 6) and the equations as follows:
Field welds are at the Critical locations 1, 3 and 15. Shop weld is at critical location 6. The "Z" factor correction for SMAW was applied (References 7-5 and 7-6) at the field weld critical locations (locations 1, 3 and 15) and the "Z" factor correction for SAW was applied (References 7-5 and 7-6) at the shop weld location (location 6) and the equations as follows:
Z = 1.15 [1.0 + 0.013 (OD-4)]                     for SMAW Z = 1.30 [1.0 + 0.01 (OD-4)]                       for SAW where OD is the outer diameter of the pipe in inches.
Z = 1.15 [1.0 + 0.013 (OD-4)]
The Z-factors were calculated for the critical locations, using the dimensions given in Table 3-1.
Z = 1.30 [1.0 + 0.01 (OD-4)]
where OD is the outer diameter of the pipe in inches.
for SMAW for SAW The Z-factors were calculated for the critical locations, using the dimensions given in Table 3-1.
The Z factor was 1.599 for locations 1 and 3, 1. 72 for location 6 and 1.592 for location 15. The applied loads were increased by the Z factors and plots of limit load versus crack length were generated as shown in Figures 7-2, 7-3, 7-4 and 7-5. Table 7-2 summarizes the results of the stability analyses based on limit load. The leakage flaw sizes are also presented on the same table.
The Z factor was 1.599 for locations 1 and 3, 1. 72 for location 6 and 1.592 for location 15. The applied loads were increased by the Z factors and plots of limit load versus crack length were generated as shown in Figures 7-2, 7-3, 7-4 and 7-5. Table 7-2 summarizes the results of the stability analyses based on limit load. The leakage flaw sizes are also presented on the same table.
 
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.
==7.4     REFERENCES==
7-2 Johnson, W. and Mellor, P. B., Engineering Plasticity, Van Nostrand Relmhold Company, New York, (1973), pp. 83-86.
 
7-3 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.
7-1     Kanninen, M. F., et. al., "Mechanical Fracture Predictions for Sensitized Stainless Steel Piping with Circumferential Cracks," EPRI NP-192, September 1976.
7-4 Irwin, G. R., "Plastic Zone near a Crack and Fracture Toughness," Proc.ih Sagamore Conference, P. IV-63 (1960).
7-2     Johnson, W. and Mellor, P. B., Engineering Plasticity, Van Nostrand Relmhold Company, New York, (1973), pp. 83-86.
7-5 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.
7-3     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.
7-4     Irwin, G. R., "Plastic Zone near a Crack and Fracture Toughness," Proc.ih Sagamore Conference, P. IV-63 (1960).
7-5     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.
32626-32633.
7-6     NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.
7-6 NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.
Fracture Mechanics Evaluation                                                             March 2019 WCAP-15550-NP                                                                             Revision 2
Fracture Mechanics Evaluation WCAP-15550-NP March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                     7-5 Table 7-1   Stability Results for Surry Units 1 and 2 Based on Elastic-Plastic J-lntegral Evaluations                                   a,c,e Table 7-2     Stability Results for Surry Units 1 and 2 Based on Limit Load Critical Location               Critical Flaw Size (in)       Leakage Flaw Size (in) 1                               19.87                         4.02 3                                 34.08                         5.85 6                                 34.81                         6.84 15                               40.07                         7.96 Fracture Mechanics Evaluation                                                             March 2019 WCAP-15550-N P                                                                               Revision 2
WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 7-1 Stability Results for Surry Units 1 and 2 Based on Elastic-Plastic J-lntegral Evaluations Table 7-2 Stability Results for Surry Units 1 and 2 Based on Limit Load Critical Location Critical Flaw Size (in)
Leakage Flaw Size (in) 1 19.87 4.02 3
34.08 5.85 6
34.81 6.84 15 40.07 7.96 7-5 Fracture Mechanics Evaluation WCAP-15550-N P March 2019 Revision 2 a,c,e


WESTINGHOUSE NON-PROPRIETARY CLASS 3               7-6 Neutral Axis Figure 7-1 [     ]a,c,e Stress Distribution Fracture Mechanics Evaluation                                        March 2019 WCAP-15550-N P                                                        Revision 2
WESTINGHOUSE NON-PROPRIETARY CLASS 3 Figure 7-1 [
Fracture Mechanics Evaluation WCAP-15550-N P Neutral Axis
]a,c,e Stress Distribution 7-6 March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                        7-7 80000 70000 60000
80000 70000 60000 II)
- I I) 0.
: 0.
:SC:I 50000
50000
--C C
:SC:
Q) 40000 E
I C --
0    30000
40000 C
:1E
Q)
.:==
E 0
E   20000
30000
:1E  
.:==
E 20000
:J 10000 0
:J 10000 0
0               10               20           30           40          50 Flaw Length (inches)
WESTINGHOUSE NON-PROPRIETARY CLASS 3 0
OD = 34.00 in. cry  = 20. 76 ksi F = 1640 kips t = 2.395 in. cru   = 56.20 ksi M = 24646 in-kips A376-TP316 Material with SMAW Weld Figure 7-2 Critical Flaw Size Prediction - Hot Leg at Location 1 Fracture Mechanics Evaluation                                                   March 2019 WCAP-15550-N P                                                                   Revision 2
10 20 30 Flaw Length (inches)
OD = 34.00 in.
t = 2.395 in.
cry
= 20. 76 ksi cru  
= 56.20 ksi A376-TP316 Material with SMAW Weld F = 1640 kips M = 24646 in-kips 40 Figure 7-2 Critical Flaw Size Prediction - Hot Leg at Location 1 Fracture Mechanics Evaluation WCAP-15550-N P 7-7 50 March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                                       7-8.
WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-8.
100000 90000 80000
100000 90000 80000 70000 0 Q.
  -0 Q.
:i 60000 C:
:i..
50000 C:
70000 60000
Q)
  -
E 40000 0  
C:
!2E
:,:..
:t:
C:
30000 E
50000 Q)
:J 20000 (34_080 in_, 20650 in-kips) 10000 0
E     40000 0
0 10 20 30 40 Flaw Length (inches)
  !2E
OD= 34.00 in.
:t:   30000 E
cry  
:J     20000 ------------****---------------------- -----------------*----------------- (34_080 in_, 20650 in-kips) 10000 0
= 23. 78 ksi F = 1639 kips t= 2.395 in.
0                   10                   20                   30                 40                 50 Flaw Length (inches)
cru  
OD= 34.00 in.               cry = 23. 78 ksi           F = 1639 kips t= 2.395 in.               cru = 71.90 ksi           M = 12918 in-kips A351-CF8M Material with SMAW Weld Figure 7-3 Critical Flaw Size Prediction - Hot Leg at Location 3 Fracture Mechanics Evaluation                                                                                 March 2019 WCAP-15550-N P                                                                                                   Revision 2
= 71.90 ksi M = 12918 in-kips A351-CF8M Material with SMAW Weld Figure 7-3 Critical Flaw Size Prediction - Hot Leg at Location 3 Fracture Mechanics Evaluation WCAP-15550-N P 50 March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                       7-9 140000 120000
WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-9 140000 120000  
  - II)
- 100000 II)
Q.
Q.
:ii:I 100000
:ii:
  -....
I.5 80000 C
  .5 C
Cl)
Cl) 80000 E
E 60000 0
0   60000
:E
:E
:t::
:t:: E 40000
E
:::i (34.814 in., 26960 in-kips) 20000 0
:::i 40000                                                          (34.814 in., 26960 in-kips) 20000 0
0 10 20 30 40 Flaw Length (inches)
0             10             20           30               40                   50 Flaw Length (inches)
OD = 36.32 in.
OD = 36.32 in.         cry   = 24.67 ksi F = 1870 kips t = 2.555 in.           cru = 71.90 ksi   M = 15673 in-kips A351-CF8M material with SAW Weld Figure 7-4 Critical Flaw Size Prediction - Crossover Leg at Location 6 Fracture Mechanics Evaluation                                                               March 2019 WCAP-15550-NP                                                                                 Revision 2
cry  
= 24.67 ksi F = 1870 kips t = 2.555 in.
cru  
= 71.90 ksi M = 15673 in-kips A351-CF8M material with SAW Weld Figure 7-4 Critical Flaw Size Prediction - Crossover Leg at Location 6 Fracture Mechanics Evaluation WCAP-15550-NP 50 March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                     7-10 100000
WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-10 100000 Cl)
  - Cl)
Q.
Q.
:.i:I 80000
80000
  -....
:.i:
  .5 C:
I.5 -
G)   60000 E
C:
0
60000 G)
:IS
E 0
:!::
:IS 40000 E
E   40000
:J 20000 0...... -
:J 20000 0 ......- .............-+__.______+-_.............-+__.______+-_..............-+.....____--...
............. -+__._ _____ +-_............. -+__._ _____ +-_.............. -+..... ____ --...
0                     10         20             30         40             50           60 Flaw Length (inches)
0 10 20 30 40 50 Flaw Length (inches)
OD = 33.60 in.             cry   =24.67ksi       F=1418kips t = 2.940 in.               cru   = 71.90 ksi     M = 10829 in-kips A351-CF8M material with SMAW Weld Figure 7-5 Critical Flaw Size Prediction - Cold Leg at Location 15 Fracture Mechanics Evaluation                                                                     March 2019 WCAP-15550-NP                                                                                       Revision 2
OD = 33.60 in.
cry  
=24.67ksi F=1418kips t = 2.940 in.
cru  
= 71.90 ksi M = 10829 in-kips A351-CF8M material with SMAW Weld Figure 7-5 Critical Flaw Size Prediction - Cold Leg at Location 15 Fracture Mechanics Evaluation WCAP-15550-NP 60 March 2019 Revision 2  


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 [                               rc,e region of a typical system (see Location     [ ]8,c,e of Figure   3-2). 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 less than 10% variation.
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[                                                                                           ]a,c,e of a plant typical in geometry and operational characteristics to any Westinghouse PWR System.
rc,e region of a typical system (see Location [ ]8,c,e of Figure 3-2). 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 less than 10% variation.
[
A [  
                                                ]a,c,e The normal, upset, and test conditions were considered. A summary of generic applied transients is provided in Table 8-1. Circumferentially oriented surface flaws were postulated in the region, assuming the flaw was located in two different locations, as shown in Figure 8-1. Specifically, these were:
]a,c,e of a plant typical in geometry and operational characteristics to any Westinghouse PWR System.
Cross Section A: Stainless Steel Cross Section B: SA 508 Cl. 2 or 3 Low Alloy Steel Fatigue crack growth rate laws were used [
[  
                    ]8,c,e The law for stainless steel was derived from Reference 8-1, a compilation of data for austenitic stainless steel in a PWR water environment was presented in Reference 8-2, and it was found that the effect of the environment on the crack growth rate was very small. From this information it was estimated that the environmental factor should be conservatively set at [ rc,e in the crack growth rate equation from Reference 8-1.
]a,c,e The normal, upset, and test conditions were considered. A summary of generic applied transients is provided in Table 8-1. Circumferentially oriented surface flaws were postulated in the region, assuming the flaw was located in two different locations, as shown in Figure 8-1. Specifically, these were:
Cross Section A: Stainless Steel Cross Section B: SA 508 Cl. 2 or 3 Low Alloy Steel Fatigue crack growth rate laws were used [  
]8,c,e The law for stainless steel was derived from Reference 8-1, a compilation of data for austenitic stainless steel in a PWR water environment was presented in Reference 8-2, and it was found that the effect of the environment on the crack growth rate was very small.
From this information it was estimated that the environmental factor should be conservatively set at [ rc,e in the crack growth rate equation from Reference 8-1.
For stainless steel, the fatigue crack growth formula is:
For stainless steel, the fatigue crack growth formula is:
Fatigue Crack Growth Analysis                                                             March 2019 WCAP-15550-NP                                                                               Revision 2
Fatigue Crack Growth Analysis WCAP-15550-NP March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                 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.
WESTINGHOUSE NON-PROPRIETARY CLASS 3 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 Surry 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 Surry 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 Surry transients and cycles. The transients and cycles (shown in Table 8-3) for the Surry 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 reactor vessel transients and cycles for Surry 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 Surry 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 Surry transients and cycles. The transients and cycles (shown in Table 8-3) for the Surry 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.
It is therefore, concluded that the generic fatigue crack growth analysis shown in Table 8-2 is representative of the Surry plants fatigue crack growth and also applicable for 80 years.
It is therefore, concluded that the generic fatigue crack growth analysis shown in Table 8-2 is representative of the Surry plants fatigue crack growth and also applicable for 80 years.  
 
==8.1      REFERENCES==


8-1     James, L.A. and Jones, D.P., "Fatigue Crack Growth Correlations for Austenitic Stainless Steel in Air, Predictive Capabilities in Environmentally Assisted Cracking," ASME Publication PVP-99, December 1985.
==8.1 REFERENCES==
8-2     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. 1.979.
8-1 James, L.A. and Jones, D.P., "Fatigue Crack Growth Correlations for Austenitic Stainless Steel in Air, Predictive Capabilities in Environmentally Assisted Cracking," ASME Publication PVP-99, December 1985.
Fatigue Crack Growth Analysis                                                         March 2019 WCAP-15550-N P                                                                           Revision 2
8-2 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. 1.979.
Fatigue Crack Growth Analysis WCAP-15550-N P March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                        8-3 Table 8-1 Summary of Reactor Vessel Transients Number of Number                    Typical Transient Identification Cycles 1             Turbine Roll                                       10 2             Cold Hydro                                         10 3             Heatup/Cooldown                                   200 4             Loading and Unloading                           18300 5             10% Step Load Decrease/Increase                 2000 6             Large Step Decrease                               200 7           Steady State Fluctuation                       1000000 8           Loss of Load from Full Power                       80 9           Loss of Power                                     40 10           Partial Loss of Flow                               80 11           Reactor Trip from Full Power                     400 12           Hot Hydro Test                                     50 Fatigue Crack Growth Analysis                                               March 2019 WCAP-15550-N P                                                               Revision 2
WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 8-1 Summary of Reactor Vessel Transients Number Typical Transient Identification 1
Turbine Roll 2
Cold Hydro 3
Heatup/Cooldown 4
Loading and Unloading 5
10% Step Load Decrease/Increase 6
Large Step Decrease 7
Steady State Fluctuation 8
Loss of Load from Full Power 9
Loss of Power 10 Partial Loss of Flow 11 Reactor Trip from Full Power 12 Hot Hydro Test Fatigue Crack Growth Analysis WCAP-15550-N P 8-3 Number of Cycles 10 10 200 18300 2000 200 1000000 80 40 80 400 50 March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3               8-4 a,c,e Table 8-2 Typical Fatigue Crack Growth at [     1 (40, 60, and 80 years)
WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 8-2 Typical Fatigue Crack Growth at [
FINAL FLAW ( in.)                       a,c,e Fatigue Crack Growth Analysis                                        March 2019 WCAP-15550-NP                                                        Revision 2
Fatigue Crack Growth Analysis WCAP-15550-NP (40, 60, and 80 years)
FINAL FLAW ( in.)
1 a,c,e 8-4 March 2019 Revision 2 a,c,e


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                 8-5 Table 8-3 Summary of Reactor Vessel Transients For Surry Units 1 and 2 (40, 60, 80 years)
WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-5 Table 8-3 Summary of Reactor Vessel Transients For Surry Units 1 and 2 (40, 60, 80 years)
Number                   Typical Transient Identification       Number of Cycles 1           Heatu p/Cooldown                                           200 2           Loading and Unloading                                     18300 3           10% Step Load Decrease/Increase                             2000 4 '        Large Step Decrease                                         200 5           Reactor Trip from Full Power                               400 6           Hot Hydro Test                                               45 Fatigue Crack Growth Analysis                                                         March 2019 WCAP-15550-N P                                                                         Revision 2
Number Typical Transient Identification 1
Heatu p/Cooldown 2
Loading and Unloading 3
10% Step Load Decrease/Increase 4
Large Step Decrease 5
Reactor Trip from Full Power 6
Hot Hydro Test Fatigue Crack Growth Analysis WCAP-15550-N P Number of Cycles 200 18300 2000 200 400 45 March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3               8-6
WESTINGHOUSE NON-PROPRIETARY CLASS 3 Figure 8-1 Typical Cross-Section of [
                                                              ]a,c,e Figure 8-1 Typical Cross-Section of [
Fatigue Crack Growth Analysis WCAP-15550-N P  
Fatigue Crack Growth Analysis                                       March 2019 WCAP-15550-N P                                                       Revision 2
]a,c,e 8-6 March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                                                                 8-7 1000 - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~
.!! u > u....
* Linear interpolation is recom-700                   mended to account for ratio dependence of water environment 500                   curves, for 0.25 <R< 0.65 for shallow slope:
J::
                                      ~ = (1.01 X 10*11 o ti. K1 .95 dN                    2 02 = 3.75 R +0.06 R = /(min /Kmax 200 Subsurface fla\NS
u C *e u i
          .!!
&sect;
u    100                     lair environment)
~
            .
e a:
            >
J:: j e t.7 u
u J::
I! u WESTINGHOUSE NON-PROPRIETARY CLASS 3 1000 ----------------------------~
u    70 C
700 500 200 100 70 50 20 10 7
          *eu              Determine the ti.K at which the 50 i                  law changes by calculation of the intersection of the two
          &sect;                  curves.
          ""~
e a:'"                    Surface flaws J::                      (water reactor j                      environment) 20 e
t.7                      applicable for
          "'I!
u                              R < 0.25 u                      "0.25 < R < 0.65 R;, 0.65 10             R = Kmin /Krnax 7
5
5
* Linear interpolation is recom-mended to account for ratio dependence of water environment curves, for 0.25 <R< 0.65 for shallow slope:
~ = (1.01 X 10*11 o2 ti. K1.95 dN 02 = 3.75 R +0.06 R = /(min /Kmax Subsurface fla\\NS lair environment)
Determine the ti.K at which the law changes by calculation of the intersection of the two curves.
2 Surface flaws (water reactor environment) applicable for R < 0.25 "0.25 < R < 0.65 R;, 0.65 R = Kmin /Krnax 5
7 10 20 Stress lntensitv Factor Range (ti.K1 ksi ~.I
* Linear interpolation is recommended to account for R ratio dependence of water environment curves, for 0.25 < R < 0.65 fer steep s!ope:
* Linear interpolation is recommended to account for R ratio dependence of water environment curves, for 0.25 < R < 0.65 fer steep s!ope:
da = (1.02 X 10-61 01 ti.K5 *95 dN 01 = 26.9R
da = (1.02 X 10-61 01 ti.K5*95 dN 01 = 26.9R
* 5.725 R = Kmin /Kmax 2                  5      7      10            20              50     70     100 Stress lntensitv Factor Range (ti.K1 ksi ~.I Figure 8-2 Reference Fatigue Crack Growth Curves for Carbon & Low Alloy Ferritic Steels Fatigue Crack Growth Analysis                                                                                             March 2019 WCAP-15550-NP                                                                                                               Revision 2
* 5.725 R = Kmin /Kmax 50 70 100 8-7 Figure 8-2 Reference Fatigue Crack Growth Curves for Carbon & Low Alloy Ferritic Steels Fatigue Crack Growth Analysis WCAP-15550-NP March 2019 Revision 2  
 
WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-1


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, 7.2 and 7.3 are used in performing the assessment of margins. Margins are shown in Table 9-1. All of the LBB recommended margins are satisfied.
==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, 7.2 and 7.3 are used in performing the assessment of margins. Margins are shown in Table 9-1. All of the LBB recommended margins are satisfied.
In summary, at all the critical locations relative to:
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).
: 1.
: 2. Leak Rate - A margin of 10 exists between the calculated leak rate from the leakage flaw and the plant leak detection capability of 1 gpm.
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).
: 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.
: 2.
Assesment of Margins                                                                 March 2019 WCAP-15550-N P                                                                         Revision 2
Leak Rate - A margin of 10 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.
Assesment of Margins WCAP-15550-N P March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                             9-2 Table 9-1 Leakage Flaw Sizes, Critical Flaw Sizes and Margins for Surry Units 1 and 2 Location               Leakage Flaw Size     Critical Flaw Size              Margin 8                        8 1                       4.02 in.               19.87 in.                 4.9 8                        8 3                       5.85 in.               34.08 in.                 5.8 3                      5.85 in.               11.70bin.                 >2.0b 8                       8 6                      6.84 in.               34.81 in.                 5.1 6                      6.84 in.               13.68b in.               >2.0b 8                        8 15                     7.96 in.               40.07 in.                  5.0 15                      7.96 in.               15.92b in.               >2.0b "based on limit load bbased on J integral evaluation Assesment of Margins                                                                    March 2019 WCAP-15550-NP                                                                            Revision 2
WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-2 Table 9-1 Leakage Flaw Sizes, Critical Flaw Sizes and Margins for Surry Units 1 and 2 Location Leakage Flaw Size 1
4.02 in.
3 5.85 in.
3 5.85 in.
6 6.84 in.
6 6.84 in.
15 7.96 in.
15 7.96 in.  
"based on limit load bbased on J integral evaluation Assesment of Margins WCAP-15550-NP Critical Flaw Size 19.87 8 in.
34.08 8 in.
11.70bin.
34.81 8 in.
13.68b in.
40.07 8 in.
15.92b in.
Margin 4.9 8
5.8 8
>2.0b 5.1 8
>2.0b 5.0 8
>2.0b March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                             10-1
WESTINGHOUSE NON-PROPRIETARY CLASS 3 10-1  


==10.0 CONCLUSION==
==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 Surry Units 1 and 2 as follows:
S This report justifies the elimination of RCS primary loop pipe breaks from the structural design basis for the 80 year plant life of Surry 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. There is no Alloy 82/182 material present in the welds for the Surry Units 1 and 2 Reactor Coolant System (RCS) primary loop piping.
: a.
* b. Water hammer should not occur in the RCS piping because of system design, testing, and operational considerations.
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. There is no Alloy 82/182 material present in the welds for the Surry Units 1 and 2 Reactor Coolant System (RCS) primary loop piping.
: c. The effects of low and high cycle fatigue on the integrity of the primary piping are negligible.
* b.
: d. Ample margin exists between the leak rate of small stable flaws and the capability of the Surry Units 1 and 2 reactor coolant system pressure boundary Leakage Detection System.
Water hammer should not occur in the RCS piping because of system design, testing, and operational considerations.
: e. Ample margin exists between the small stable flaw sizes of item (d) and larger stable flaws.
: c.
: f. Ample margin exists in the material properties used to demonstrate end-of-service life (fully aged) stability of the critical flaws.
The effects of low and high cycle fatigue on the integrity of the primary piping are negligible.
: d.
Ample margin exists between the leak rate of small stable flaws and the capability of the Surry Units 1 and 2 reactor coolant system pressure boundary Leakage Detection System.
: e.
Ample margin exists between the small stable flaw sizes of item (d) and larger stable flaws.
: f.
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 d, e, and f above.
For the critical locations, flaws are identified that will be stable because of the ample margins described in d, e, and f above.
Based on the above, the Leak-Before-Break conditions and margins are satisfied for the Surry 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 Surry Units 1 and 2 Nuclear Power Plants for the 80 year plant life (subsequent license renewal program).
Based on the above, the Leak-Before-Break conditions and margins are satisfied for the Surry 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 Surry Units 1 and 2 Nuclear Power Plants for the 80 year plant life (subsequent license renewal program).
Conclusions                                                                           March 2019 WCAP-15550-NP                                                                           Revision 2
Conclusions WCAP-15550-NP March 2019 Revision 2  


WESTINGHOU SE NON-PROPRIETARY CLASS 3       A-1 APPENDIX A: LIMIT MOMENT
Appendix A: Limit Moment WCAP-15550-N P WESTINGHOUSE NON-PROPRIETARY CLASS 3 APPENDIX A: LIMIT MOMENT  
                                                  ]a,c,e Appendix A: Limit Moment                                      March 2019 WCAP-15550- N P                                                Revision 2
]a,c,e A-1 March 2019 Revision 2  


WESTINGHOUSE NON-PROPRIETARY CLASS 3                   A-2 a,
WESTINGHOUSE NON-PROPRIETARY CLASS 3 a, 0.-----------------------
0.-------------------
ca Figure A-1 Pipe with a Through-Wall Crack in Bending Appendix A: Limit Moment WCAP-15550-NP A-2 March 2019 Revision 2  
ca                                              ----
Figure A-1 Pipe with a Through-Wall Crack in Bending Appendix A: Limit Moment                                                 March 2019 WCAP-15550-NP                                                             Revision 2


WCAP-15550-NP Revision 2                                                                                                                                 Proprietary Class 3
WCAP-15550-NP Revision 2 Proprietary Class 3  
        **This page was added to the quality record by the PRIME system upon its validation and shall not be considered in the page numbering of this document.**
**This page was added to the quality record by the PRIME system upon its validation and shall not be considered in the page numbering of this document.**
Author Approval Wiratmo Mamo Mar-14-2019 12:06:23 Reviewer Approval Johnson Eric D Mar-14-2019 12:16:52 Manager Approval Leber Benjamin A Mar-14-2019 12:27:28 Files approved on Mar-14-2019}}
Author Approval Wiratmo Mamo Mar-14-2019 12:06:23 Reviewer Approval Johnson Eric D Mar-14-2019 12:16:52 Manager Approval Leber Benjamin A Mar-14-2019 12:27:28 Files approved on Mar-14-2019}}

Latest revision as of 03:42, 5 January 2025

Enclosure 5 - Topical Report WCAP-15550-NP, Rev 2, Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Surry Units 1 and 2 Nuclear Power Plants for the Subsequent License Renewal Progra
ML19095A605
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Issue date: 03/31/2019
From:
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To:
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19-096 WCAP-15550-NP, Rev 2
Download: ML19095A605 (66)
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Text

Change Notice 2 SPS SLRA Serial No.: 19-096 Docket Nos.: 50-280/281 TOPICAL REPORT WCAP-15550-NP, REV. 2 Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

WCAP-15550-N P Revision 2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 March 2019 Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Surry Units 1 and 2 Nuclear Power Plants for the Subsequent License Renewal Program (80 Years) Leak-Before-Break Evaluation

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15550-NP Revision 2 Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Surry Units 1 and 2 Nuclear Power Plants for the Subsequent License Renewal Program (80 Years)

Leak-Before-Break Evaluation March 2019 Authors:

Mamo Wiratmo*

Structural Design & Analysis II Reviewer: Eric D. Johnson*

Structural Design & Analysis II Approved: Benjamin A. Leber*, Manager Structural Design & Analysis II

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

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

© 2019 Westinghouse Electric Company LLC All Rights Reserved

Rev Date 0

August 2000 1-A (Draft for March customer 2017 review) 1 June 2017 2

March 2019 WCAP-15550-N P WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii RECORD OF REVISIONS Revision Description Original Issue (WCAP-15550)

Revised to include the LBB results from the Measured Uncertainty Recapture (MUR) Program.

This WCAP revision also includes LBB evaluation for subsequent license renewal for 80 years of operation for Surry Units 1 and 2.

Finalized Revision 1-A and incorporated customer's comments.

Revised to address three chemical content errors in Tables 4-6 and 4-7 and to address the updated CMTR data from Surry Unit 1 that have not been considered in Revision 1. The updated CMTR data is provided in Table 4-6.

As shown in the revised Tables 4-6 and 4-7, the corrections of CMTR errors are shown in bold font and updates due to additional CMTRs are marked by grey-shaded color. Changes in the text of the document are shown with revision bars in the right margin.

March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iv TABLE OF CONTENTS

1. O 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 Water Hammer.............................................. :..................................................... 2-2 2.3 Low Cycle and High Cycle Fatigue..................................................................... 2-3 2.4 Wall Thinning, Creep, and Cleavage................................................................... 2-3
2. 5 References......................................................................................................... 2-3 3.0 Pipe Geometry and Loading...................................................................... :.................... 3-1 3.1 Introduction to Methodology................................................................................ 3-1 3.2 Calculation of Loads and Stresses.................................................................. :... 3-1 3.3 Loads for Leak Rate Evaluation.......................................................................... 3-2 3.4 Load Combination for Crack Stability Analyses................................................... 3-3 3.5 References......................................................................................................... 3-3 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........................................................................... 4-1 4.4 References......................................................................................................... 4-4
5. 0 Critical Location and Evaluation Criteria......................................................................... 5-1 5.1 Critical Locations................................................................................................. 5-1 5.2 Fracture Criteria.................................................................................................. 5-1 6.0 Leak Rate Predictions.................................................................................................... 6-1 6.1 Introduction......................................................................................................... 6-1
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 Crack Stability Evaluations.................................................................................. 7-3 7.4 References......................................................................................................... 7-4 8.0 Fatigue Crack Growth Analysis....................................................................................... 8-1 8.1 References......................................................................................................... 8-2 9.0 Assessment of Margins.................................................................................................. 9-1 10.0 Conclusions.................................................................................................................. 10-1 Appendix A: Limit Moment.......................................................................................................... A-1 WCAP-15550-N P March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 V

LIST OF TABLES Table 3-1 Dimensions, Normal Loads and Stresses for Surry Units 1 and 2............................... 3-4 Table 3-2 Faulted Loads and Stresses for Surry Units 1 and 2................................................... 3-5 Table 4-1 Measured Tensile Properties (psi) for Surry Unit 1 Primary Loop Pipes...................... 4-5 Table 4-2 Measured Tensile Properties (psi) for Surry Unit 2 Primary Loop Pipes...................... 4-6 Table 4-3 Measured Tensile Properties (psi) for Surry Unit 1 Primary Loop Elbows....................4-7 Table 4-4 Measured Tensile Properties (psi) for Surry Unit 2 Primary Loop Elbows....................4-8 Table 4-5 Mechanical Properties for Surry Units 1 and 2 Materials at Operating Temperatures............................................................................................. 4-9 Table 4-6 Chemistry and Fracture Toughness Elbow Properties of the Material Heats of Surry Unit 1............................................................................................................. 4-10 Table 4-7 Chemistry and Fracture Toughness Elbow Properties of the Material Heats of Surry Unit 2............................................................................................................. 4-12 Table 4-8 Fracture Toughness Properties for Surry Units 1 and 2 Primary Loops for Leak-Before-Break Evaluation at Critical Locations................................. 4-13 Table 6-1 Flaw Sizes Yielding a Leak Rate of 10 gpm at the Governing Locations.................... 6-3 Table 7-1 Stability Results for Surry Units 1 and 2 Based on Elastic-Plastic J-lntegral Evaluations............................................................................................................... 7-5 Table 7-2 Stability Results for Surry Units 1 and 2 Based on Limit Load.................................... 7-5 Table 8-1 Summary of Reactor Vessel Transients..................................................................... 8-3 Table 8-2 Typical Fatigue Crack Growth at [

rc,e (40, 60, and 80 years)****:****************************************************************************************** 8-4 Table 8-3 Summary of Reactor Vessel Transients for Surry Units 1 and 2 (40, 60, and 80 years)............................................................................................... 8-5 Table 9-1 Leakage Flaw Size, Critical Flaw Sizes and Margins for Surry Units 1 and 2.............................................................................................. 9-2 WCAP-15550-N P March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vi LIST OF FIGURES Figure 3-1 Hot Leg Coolant Pipe............................................................................................. 3-6 Figure 3-2 Schematic Diagram of Surry Units 1 and 2 Primary Loop Showing Weld Locations...................................................................................................... 3-7 Figure 4-1 Pre-Service J vs. ~a for SA351 CF8M Cast Stainless Steel at 600°F................... 4-14 Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures................... 6-4 Figure 6-2 rc,e Pressure Ratio as a Function of UD..................................... 6-5 Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack................................. 6-6 Figure 7-1 t,c,e Stress Distribution................................................................... 7-6 Figure 7-2 Critical Flaw Size Prediction - Hot Leg at Location 1............................................. 7-7 Figure 7-3 Critical Flaw Size Prediction - Hot Leg at Location 3............................................. 7-8 Figure 7-4 Critical Flaw Size Prediction - Cross-over Leg at Location 6................................. 7-9 Figure 7-5 Critical Flaw Size Prediction - Cold Leg at Location 15....................................... 7-10 Figure 8-1 Typical Cross-Section of [

]a,c,e................................ 8-6 Figure 8-2 Reference Fatigue Crack Growth Curves for Carbon and Low Alloy Ferritic Steels................................................................................................ 8-7 FigureA-1 Pipe with a Through-Wall Crack in Bending........................................................... A-2 WCAP-15550-N P March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vii EXECUTIVE

SUMMARY

The original structural design basis of the reactor coolant system for the Surry Units 1 and 2 Nuclear Power Plants required consideration of dynamic effects resulting from pipe break and that protective measures for such breaks be incorporated into the design. Subsequent to the original Surry 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).

Surry Units 1 and 2 Nuclear Power Plants were 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 served as the original basis for the elimination of large primary loop pipe rupture from the structural design basis for Surry Units 1 and 2. As identified in Generic Letter 84-04, the primary technical references supporting the NRC's safety evaluation of eliminating postulated pipe breaks are documented in WCAP-9558 (Reference 1-9) and WCAP-9787 (Reference 1-10).

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 1 OCFR50 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 technology for excluding from the design basis the dynamic effects of postulated ruptures in primary coolant loop piping in pressurized water reactors (PWRs).

The LBB evaluation is 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.

Revision O of this report had demonstrated compliance with LBB technology for the Surry reactor coolant system piping for the 60 year plant life based on a plant specific analysis. Subsequently, an LBB evaluation was performed for the MUR (Measurement Uncertainty Program), the results of that particular analysis are also incorporated in Revision 1 of this report. Lastly, based on the LBB evaluation in Revision 1 of this report herein, it also demonstrated that dynamic effects of reactor coolant system primary loop pipe breaks need not be considered in the structural design basis of the Surry Units 1 and 2 Nuclear Power Plants for the 80 year plant life (Subsequent License Renewal Program). The technical evaluations utilized in Revision O through Revision 2 of this report are consistent with the methodology and principles of WCAP-9558 and WCAP-9787.

Therefore, the justifications demonstrated herein are compliant with the original conclusions of Generic Letter 84-04.

The report documents the plant specific geometry, loading, and material properties used in the fracture mechanics evaluation. Mechanical properties were determined at operating temperatures.

WCAP-15550-NP March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 viii Since the piping systems include cast austenitic stainless steel, fracture toughness considering thermal aging was determined for each heat of material. Fully aged fracture toughness properties were used for the LBB evaluation. The full aged condition is applicable for plants operating at beyond 15 EFPY (Effective Full Power Years) for the CF8M materials (elbows for Surry Units 1 and 2).

As of January 2017, Surry Units 1 and 2 are operating at 33.78 and 33.69 EFPY, respectively. Thus, the LBB evaluation in this report has been demonstrated for the primary loops at Surry Units 1 and 2 for 80 years of plant operation.

WCAP-15550-N P March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-1

1.0 INTRODUCTION

1.1 PURPOSE This report applies to the Surry Units 1 and 2 Reactor Coolant System (RCS) primary loop piping. It is intended to demonstrate that for the specific parameters of the Surry 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 LBS evaluation results based on the Measurement Uncertainty Recapture (MUR)

Program.

1.2 BACKGROUND

INFORMATION Westinghouse has performed considerable testing and analysis to demonstrate that 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 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.

Based on the studies by Westinghouse, LLNL, the ACRS, and the 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).

Introduction WCAP-15550-NP March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-2 1.3 SCOPE AND OBJECTIVES The general purpose of this investigation is to demonstrate leak-before-break for the primary loops in Surry Units 1 and 2 on a plant specific basis for the 80 year plant life. The recommendations and criteria proposed in References 1-7 and 1-8 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 materials and the associated material properties.

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.

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 applied load.

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

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.

Introduction WCAP-15550-NP March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-3

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 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures, Final Rule, Federal RegisterNol. 52, No. 207fruesday, October 27, 1987/Rules and Regulations, pp. 41288-41295.

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

1-4 Letter Report NS-EPR-2519, Westinghouse (E. P. Rahe) to NRC (D. G. Eisenhut),

Westinghouse Proprietary Class 2, November 10, 1981.

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

1-6 Letter from Westinghouse (E. P. Rahe) to NRC (W. V. Johnston) dated July 25, 1983.

1-7 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.

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

1-9 WCAP-9558, Revision 2, "Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Throughwall Crack," May, 1981.

1-10 WCAP-9787, "Tensile and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation," May, 1981.

Introduction WCAP-15550-N P March 2019 Revision 2

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 SCC as well as preventing the occurrence of a Operation and Stability of the Reactor Coolant System WCAP-15550-NP March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 corrosive environment. The material specifications consider compatibility with the system's 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 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.

It should be noted that there are no primary water stress corrosion cracking material such as Alloy 82/182 in the dissimilar metal welds in the Surry Units 1 and 2 Reactor Coolant System (RCS) primary loop piping.

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 controlled in the design process. Additionally, Westinghouse has instrumented typical reactor coolant systems to verify the flow and vibration characteristics of the system.

Preoperational Operation and Stability of the Reactor Coolant System WCAP-15550-N P March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 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 measurements have been made on a number of plants during hot functional testing, including plants similar to Surry 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 should 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.

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.

Operation and Stability of the Reactor Coolant System WCAP-15550-NP March 2019 Revision 2

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 34.00 in. and 2.395 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, dead weight and thermal expansion. An additional component, Safe Shutdown Earthquake (SSE), is considered for faulted loads. Tables 3-1 and 3-2 show the enveloping loads for Surry Units 1 and 2; these loads were determined as part of the MUR project. As seen from Table 3-2, the highest stressed location in the entire loop is at Location 1 at the reactor vessel outlet nozzle to pipe weld. This is one of the locations at which, as an enveloping location, 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 elbows are made of different materials than the pipe, locations other than the highest stressed pipe location were examined taking into consideration both fracture toughness and stress. The four most critical locations among the entire primary loop are identified after the full analysis is completed. 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. 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.

Please note that the piping loads and stresses based on the MUR Program were considered in the LBB evaluation as part of Revision 1 of this WCAP report.

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

F M

a=-+ -

A Z

where, cr

=

stress, ksi F

=

axial load, kips M

=

bending moment, in-kips A

=

pipe cross-sectional area, in2 z

=

section modulus, in3 Pipe Geometry and Loading WCAP-15550-NP (3-1)

March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-2 The total moments for the desired loading combinations are calculated by the following equation:

(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 bending moment NOTE:

X-axis is along the center line of the pipe.

I 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.

3.3 LOADS FOR LEAK RATE EVALUATION The normal operating loads for leak rate predictions are calculated by the following equations:

F

=

Fow + FTH + Fp Mx

=

(Mx)ow + (Mx)TH Mv

=

(Mv)ow + (MvhH Mz

=

(Mz)ow + (MzhH The subscripts of the above equations represent the following loading cases:

ow TH p

=

=

=

deadweight normal thermal expansion load due to internal pressure (3-3)

(3-4)

(3-5)

(3-6)

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 Table 3-1 at all the weld locations identified in Figure 3-2. The as-built dimensions are also given in Table 3-1.

Pipe Geometry and Loading WCAP-15550-N P March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-3 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, SSE INERTIA and seismic anchor motion (SAM) loads are combined based on individual absolute values as shown below.

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:

F = I Fow I + I FTH I + I Fp I + I FssEINERTIA I + I FssEAM I Mx = I (Mx)ow I + I (MxhH I + I (Mx)ssEINERTIA I + I (Mx)ssEAM I Mv = I (Mv)ow I+ I (MvhH I+ I (Mv)ssEINERTIAI + I (Mv)ssEAM I Mz = I (Mz)ow I + I (Mz)TH I + I (Mz)ssEINERTIA I + I (Mz)ssEAM I (3-7)

(3-8)

(3-9)

(3-10) 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 Table 3-2.

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 WCAP-15550-N P March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-4 Table 3-1 Dimensions, Normal Loads and Stresses for Surry Units 1 and 2 Location 3

Outside Diameter (in) 1 34.00 2

34.00 3

34.00 4

37.75 5

37.625 6

36.32 7

36.32 8

36.32 9

37.625 10 37.625 11 32.26 12 32.26 13 32.26 14 33.60 15 33.60 Notes:

a.

See Figure 3-2

b.

Included Pressure

_Pipe Geometry and Loading WCAP-15550-N P Minimum Thickness (in) 2.395 2.395 2.395 3.270 3.208 2.555 2.555 2.555 3.208 3.208 2.270 2.270 2.270 2.940 2.940 Axial Loadb Moment (kips)

(in-kips) 1482 19815 1482 837 1482 9661 1614 14956 1628 7852 1605 6815 1599 6786 1709 1071 1709 2666 1844 9466 1365 2588 1365 2927 1366 2373 1366 3636 1363 4955 Total Stress (ksi) 17.51 6.71 11.73 9.87 7.55 9.11 9.07 6.81 5.90 8.76 8.11 8.33 7.97 6.64 7.29 March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 3-2 Faulted Loads and Stresses for Surry Units 1 and 2 Locationa,b Axial Loadc (kips) 1 1640 2

1640 3

1639 4

1941 5

1927 6

1870 7

1864 8

1839 9

1839 10 1885 11 1413 12 1413 13 1420 14 1422 15 1418 Notes:

a.

See Figure 3-2

b.

See Table 3-1 for dimensions

c.

Included Pressure Pipe Geometry and Loading WCAP-15550-N P Moment Total Stress (in-kips)

(ksi) 24646 20.93 2652 8.41 12918 14.25 20101 12.62 22956 13.89 15673 14.23 9928 11.52 8475 10.75 11375 9.43 14032 10.53 7850 11.84 7390 11.54 6032 10.66 7923 8.99 10829 10.42 3-5 March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3

*)

(-

Location 1 Normal Loadsa Forcec:

1482 kips Bending Moment:

19815 in-kips a See Table 3-1 b See Table 3-2 Faulted Loadsb Forcec:

1640 kips Bending Moment:

24646 in-kips c Includes the force due to a pressure of 2250 psia Pipe Geometry and Loading WCAP-15550-N P Figure 3-1 Hot Leg Coolant Pipe M

3-6 Crack

1.

OD

.1 ooa = 34.00 in ta

= 2.395 in March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3

.....-coLDLEG

\\_

Reactor Coolant Pump

\\...._ __ Steam Generator HOT LEG CROSS-OVER LEG COLD LEG CROSSOVER LEG Temperature 609.1 °F Pressure: 2250 psia Temperature 542.6°F Pressure: 2250 psia Temperature 542.9°F Pressure: 2250 psia Figure 3-2 Schematic Diagram of Surry Units 1 and 2 Primary Loop Showing Weld Locations Pipe Geometry and Loading WCAP-15550-NP 3-7 March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-1 4.0 MATERIAL CHARACTERIZATION 4.1 PRIMARY LOOP PIPE AND FITTINGS MATERIALS The primary loop pipe is A376-TP316 and the elbow fittings are A351-CF8M for Surry Units 1 and 2.

4.2 TENSILE PROPERTIES The Pipe Certified Materials Test Reports (CMTRs) for Surry Units 1 and 2 were used to establish the tensile properties for the leak-before-break analyses. The CMTRs include tensile properties at room temperature and/or at 650°F for each of the heats of material.

These properties are given in Table 4-1 for the Surry Unit 1 pipe, Table 4-2 for the Surry Unit 2 pipe, Table 4-3 for Unit 1 elbows and in Table 4-4 for Unit 2 elbows.

The representative properties at 609°F (represents actual 609.1 °F for Hot Leg) for the pipe were established from the tensile properties at 650°F given in Tables 4-1 and 4-2 by utilizing Section II of the 1999ASME Boiler and Pressure Vessel Code (Reference 4-1). Code tensile properties at 609°F were obtained by interpolating between the 600°F and 650°F tensile properties. Ratios of the code tensile properties at 609°F to the corresponding tensile properties at 650°F were then applied to the 650°F tensile properties given in Tables 4-1 and 4-2 to obtain the plant specific properties for A376-TP316 at 609°F. It should be noted that there is no significant impact by using the 1999 ASME Code Section II edition for material properties for the LBB analysis, as compared to the Surry ASME code of record.

The representative properties at 609°F (represents actual 609.1 °F for Hot Leg) and 543°F (represents actual 542.9°F for Cold Leg and 542.6°F for Crossover Leg) for the elbows were established from the tensile properties at room temperature properties given in Tables 4-3 and 4-4 by utilizing Section II of the 1999 ASME Boiler and Pressure Vessel Code (Reference 4-1).

Code tensile properties at 609°F and 543°F were obtained by interpolating between the 500°F, 600°F and 650°F tensile properties. Ratios of the code tensile properties at 609°F and 543°F to the corresponding tensile properties at room temperature were then applied to the room temperature tensile properties given in Tables 4-3 and 4-4 to obtain the plant specific properties for A351-CF8M at 609°F and 543°F.

The average and lower bound yield strengths and ultimate strengths are given in Table 4-5. The ASME Code moduli of elasticity values are also given, and Poisson's ratio was taken as 0.3.

Updated CMTRs from Replacement Steam Generator (RSG) replacement elbows on Surry Unit 1 are also considered. The added tensile properties are shown in Table 4-3 (marked by grey-shaded color). Of these added properties, the minimum Yield Strength is 38350 psi and the minimum Ultimate Strength is 81200 psi. These values are bounded by lower bound values in Table 4-5. It has also been reviewed that the updated tensile data has negligible impact to the average Yield Strength in Table 4-5.

4.3 FRACTURE TOUGHNESS PROPERTIES The pre-service fracture toughness (J) of cast stainless steels that are of interest are in terms of J1c (J at Crack Initiation) and have been found to be very high at 600°F. [

Material Characterization WCAP-15550-N P

]a,c,e However, cast March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-2 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 embrittlement, 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.

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

The method described below was used to calculate the end of life toughness properties for the cast material of the Surry Units 1 and 2 primary coolant loop piping and elbows.

In 1994, the Argonne National Laboratory (ANL) completed an extensive research program in assessing the extent of thermal aging of cast stainless steel materials (Reference 4-2). 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 4-3).

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 in Reference 4-3 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.

Based on NUREG/CR-4513, Revision 2, the fracture toughness correlations used for the full aged condition is applicable for plants operating at and beyond 15 EFPY (Effective Full Power Years) for the CF8M materials (elbows for Surry Units 1 and 2). As of January 2017, Surry Units 1 and 2 are operating at 33. 78 and 33.69 EFPY, respectively. Therefore, the use of the fracture toughness correlations described below is applicable for the fully aged or saturated condition of the Surry Units 1 and 2 elbow materials made of CF8M.

The chemical compositions of the Surry Units 1 and 2 primary loop elbow fitting material are available from CMTRs and are provided in Table 4-6 and Table 4-7 of this report. The following equations are taken from Reference 4-3 and applicable for CF8M type material:

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) + 24.5(C) + 2.77 = (Nickel equivalent)

(4-2)

Oc =100.3(Creq I Nieq )2-170.72(Creq I Nieq )+74.22 = (Ferrite Content)

(4-3)

Material Characterization WCAP-15550-NP March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-3 where the elements are in percent weight and oc is ferrite in percent volume.

The saturation room temperature (RT) impact energies of the cast stainless steel materials were determined from the chemical compositions available from CMTRs and provided in Tables 4-6 and 4-7.

For CF8M steel with < 10% Ni, the saturation value of RT impact energy CV sat (J/cm2) is the lower value determined from log10CVsat = 0.27 + 2.81 exp (-0.022$)

where the material parameter $ is expressed as

$ = oc (Ni + Si + Mn)2(C + 0.4N)/5.0 and from (4-4)

(4-5) log10CVsat = 7.28 - 0.011oc - 0.185Cr - 0.369Mo - 0.451 Si - 0.007Ni - 4.71 (C + 0.4N)

(4-6)

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$)

where the material parameter $ is expressed as

$ = oc (Ni + Si + Mn)2(C + 0.4N)/5.0 and from log10CVsat = 7.28 - 0.011 oc - 0.185Cr - 0.369Mo - 0.451 Si - 0.007Ni - 4. 71 (C + 0.4N)

The saturation J-R curve at RT, for static-cast CF8M steel is given by Jd = 1.44 (CV sat) 1'35(.!iat for CVsat < 35 J/cm2 for CVsat ~ 35 J/cm 2

n = 0.20 + 0.08 log10 (CVsat) where Jct is the "deformation J" in kJ/m2 and Lia is the crack extension in mm.

The saturation J-R curve at 290-320°C (554-608°F), for static-cast CF8M steel is given by Jd = 5.5 (CVsal'98(Liat for CVsat < 46 J/cm2 Jd = 49 (CVsal'41 (Lia)"

for CVsat ~ 46 J/cm2 n = 0.19 + 0.07 log10 (CVsat) where Jct is the "deformation J" in kJ/m2 and Lia is the crack extension in mm.

(4-7)

(4-8)

(4-9)

(4-10)

(4-11)

(4-12)

(4-13)

(4-14)

(4-15)

Material Characterization WCAP-15550-N P March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-4 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.

In the fracture mechanics analyses that follow, the fracture toughness properties given in Table 4-8 will be used as the criteria against which the applied fracture toughness values will be compared.

As indicated in the record of revisions table, this stress report is revised to address CMTR errors and to address updated CMTRs from RSG replacement elbows on Unit 1.

As shown in the revised Tables 4-6 and 4-7, the corrections of CMTR errors (shown in bold font) and updates due to additional CMTR's (marked by grey-shaded color) have been included in this stress report. It has been reviewed that the revisions do not impact the evaluation results in this Section 4 material characterization. The summary of limiting fracture toughness properties provided in Table 4-8 remains valid and applicable. The revisions also do not affect the conclusions of this stress report.

4.4 REFERENCES

4-1 ASME Boiler and Pressure Vessel Code, An International Code,Section II, Materials, Part D-Properties, 1999 Addenda, July 1, 1999.

4-2

0. K. Chopra and W. J. Shack, "Assessment of Thermal Embrittlement of Cast Stainless Steels," NUREG/CR-6177, U.S. Nuclear Regulatory Commission, Washington, DC, May 1994.

4-3

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.

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

Material Characterization WCAP-15550-N P March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-5 Table 4-1 Measured Tensile Properties (psi) for Surry Unit 1 Primary Loop Pipes Heat No./Serial Location No.

F021212867Z X-over Leg F021212867Z X-over Leg F021512894Y X-over Leg F021512894Y X-over Leg F022212900Y X-over Leg F022212900Y X-over Leg 877712883X X-over Leg 877712883X X-over Leg 878512881X X-over Leg 878512881X X-over Leg 891512885X X-over Leg 891512885X X-over Leg 877712874Y Cold Leg 877712874Y Cold Leg 891312875Y Cold Leg 891312875Y Cold Leg F022812996 Cold Leg F022812996 Cold Leg F022912952 Cold Leg F022912952 Cold Leg F016212859 Cold Leg F016212859 Cold Leg F016212860 Cold Leg F016212860 Cold Leg F021612861 Cold Leg F021612861 Cold Leg F022712947 Cold Leg F022712947 Cold Leg F022912951 Cold Leg F022912951 Cold Leg 878512963X Hot Leg 878512963X Hot Leg FO 1901284 7Y Hot Leg F019012847Y Hot Leg F005812629 Hot Leg F005812629 Hot Leg F018812845 Hot Leg F018812845 Hot Leg E148213353 Hot Leg E148213353 Hot Leg Note:

NIA= Not Applicable Material Characterization WCAP-15550-NP At Room Temp.

Yield Ultimate Strength Strength 43500 84200 43900 82500 40000 84800 41900 87700 44000 86500 41500 83200 36100 78200 38500 77800 36100 74200 39700 79800 38500 77400 38600 77200 35300 79200 34900 78200 35100 78400 41100 84800 45100 88600 43900 87400 42200 85300 39700 83400 46000 83500 52900 87900 43700 83900 47400 90400 40600 83400 40000 80500 42500 83700 43200 87100 42000 84400 44500 86800 33400 75700 33600 75300 42000 88800 43000 86000 41200 85200 44300 89500 40900 83000 43000 84000 41800 84400 41700 84600 At 650°F Yield Strength 26400 NIA 21700 NIA 21500 NIA 21000 NIA 20400 NIA 24200 NIA 24100 NIA 24300 NIA 24700 NIA 24500 NIA 27600 NIA NIA NIA 21300 NIA 20900 NIA 24500 NIA 22000 NIA 21300 NIA 27000 NIA 26100 NIA 24500 NIA Ultimate Strength 67000 NIA 66800 NIA 66200 NIA 62000 NIA 57200 NIA 62300 NIA 65600 NIA 68500 NIA 69800 NIA 68200 NIA 69400 NIA NIA NIA 66600 NIA 66600 NIA 68200 NIA 61000 NIA 58200 NIA 74000 NIA 67000 NIA 66200 NIA March 2019 Revision 2

l WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-6 Table 4-2 Measured Tensile Properties (psi) for Surry Unit 2 Primary Loop Pipes Heat No./Serial Location No.

F0213/2889 Hot Leg F0213/2889 Hot Leg F0213/2890X Hot Leg F0213/2890X Hot Leg F0213/2891X Hot Leg F0213/2891X Hot Leg F0227/2897X Hot Leg F0227/2897X Hot Leg F0373/3169 Cold Leg F0373/3169 Cold Leg F0221 /2866X X-over Leg F0221/2866Y X-over Leg F0222/2900X X-over Leg F0222/2900X X-over Leg 52154/2843 Cold Leg 52154/2843 Cold Leg F0228/2948 Cold Leg F0228/2948 Cold Leg F0229/2994 Cold Leg F0229/2994 Cold Leg E1490/3347Y Hot Leg E1490/3347Y Hot Leg F0189/2869Y X-over Leg F0189/2869Z X-over Leg F0189/2868X X-over Leg F0189/2868X X-over Leg F0226/2946 Cold Leg F0226/2946 Cold Leg K2011/3683X Cold Leg 1<2011 /3683X Cold Leg E1478/3257 Cold Leg E1478/3257 Cold Leg V0629/3262 Cold Leg V0629/3262 Cold Leg F0229/2953 Cold Leg F0229/2953 Cold Leg F0215/2892Y Hot Leg F0215/2892Y Hot Leg Note:

N/A = Not Applicable Material Characterization WCAP-15550-NP At Room Temperature Yield Ultimate Strength Strength 43200 84800 44000 88400 41900 82500 46500 84500 44000 86000 41000 83000 41800 86800 42500 85500 41400 85800 44800 89700 44000 83800 42700 86000 44000 86500 41500 83200 43500 86500 33300 75600 42500 87400 45000 87500 41000 82800 45000 87900 43100 86000 43000 82700 37700 80600 44100 91000 37700 80600 44100 91000 43000 85000 42100 86000 32100 75900 38400 81900 38000 82900 42400 84900 48400 88800 41300 82200 41000 84900 45000 86200 43000 83000 42000 84600 Yield Strength 23700 N/A N/A N/A N/A N/A N/A N/A 25600 N/A N/A 21600 21500 N/A 23100 N/A 24700 N/A 24500 N/A 23700 N/A N/A N/A 25200 N/A 21500 N/A 20600 N/A 25300 N/A 21100 N/A 24500 N/A 21700 N/A At 650°F Ultimate Strength 69800 N/A N/A N/A N/A N/A N/A N/A 71900 N/A N/A 65200 66200 N/A 57400 N/A 69800 N/A 68200 N/A 68000 N/A N/A N/A 70000 N/A 66400 N/A 56200 N/A 72400 N/A 67400 N/A 68200 N/A 66800 N/A March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 4-3 Measured Tensile Properties (psi) for Surry Unit 1 Primary Loop Elbows Heat No.

Location At Room Temperature Yield Strength Ultimate Strength 10360-1 X-over Leg 46500 88000 10844-1 X-over Leg 39000 78000 11046-1 X-over Leg 43500 87000 11246-1 X-over Leg 42000 82500 11441-1 X-over Leg 48000 88500 11937-1 X-over Leg 45000 88500 10442-1 X-over Leg 48000 88500 11168-1 X-over Leg 46500 88000 12198-1 X-over Leg 45000 85000 12623-1 X-over Leg 42000 80000 10482-1 X-over Leg 45750 80250 10723-1 X-over Leg 46200 88600 29943-2 Cold Leg 39300 77300 30690-1 Cold Leg 39700 80200 29943-1 Hot Leg 38400 74900 31011-5 Hot Leg 43000 84800 28387-2 Hot Leg 42300 85800 28387-1 Cold Leg 42300 85800 30597-2 X-over Leg 42300 84800 10128-2 X-over Leg 46500 89000 10243-2 X-over Leg 48000 90000 Note:

tensile properties from the added heats are not included in the average tensile property calculations in Table 4-5, but the impact of the added tensile data are negligible.

4-7 Material Characterization WCAP-15550-N P March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-8 Table 4-4 Measured Tensile Properties (psi) for Surry Unit 2 Primary Loop Elbows Heat No.

30080-2 32535-2 31571-6 14008-1 14085-1 14165-1 12709-1 14507-1 13579-1 13781-5 13826-2 30690-2 31427-2 14786-3 14990-1 15384-1 15769-1 30080-1 12087-3 12547-2 13051-4 Material Characterization WCAP-15550-N P Location Hot Leg Cold Leg Cold Leg X-over Leg X-over Leg X-over Leg X-over Leg X-over Leg X-over Leg X-over Leg X-over Leg Hot Leg Hot Leg X-over Leg X-over Leg X-over Leg X-over Leg Cold Leg X-over Leg X-over Leg X-over Leg At Room Temperature Yield Strength 42400 44100 42800 48000 45000 48000 48000 40500 48000 4500 48000 38100 41250 42000 43500 45000 46500 45100 42000 40500 40500 Ultimate Strength 82400 87500 85600 87000 86000 89000 85500 82000 85500 89500 87500 82200 86250 84500 84500 88500 85500 84200 80500 83500 83500 March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-9 Table 4-5 Mechanical Properties for Surry Units 1 and 2 Materials at Operating Temperatures Lower Bound Temperature*

Average Yield Modulus of Material Strength Elasticity Yield Ultimate (OF)

(psi)

(psi)

Strength Strength (psi)

(psi)

A376 TP316 609 23835 25.255 X 106 20,762 56,200 609 27468 25.255 X 106 23,785 71,904 A351 CF8M 25.585 X 106 543 28493 24,672 71,904 Poisson's ratio: 0.3 Note: Representative temperature. The actual temperatures are provided in Figure 3-2.

Material Characterization WCAP-15550-NP March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 4-6 Chemistry and Fracture Toughness Elbow Properties of the Material Heats of Surry Unit 1 Material Characterization WCAP-15550-N P 4-10 March 2019 Revision 2 a,c,e

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 4-6 Chemistry and Fracture Toughness Elbow Properties of the Material Heats of Surry Unit 1 Material Characterization WCAP-15550-NP 4-11 March 2019 Revision 2 a,c,e

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 4-7 Chemistry and Fracture Toughness Elbow Properties of the Material l:feats of Surry Unit 2 Material Characterization WCAP-15550-N P 4-12 March 2019 Revision 2 a,c,e

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 4-8 Fracture Toughness Properties for Surry Units 1 and 2 Primary Loops for Leak-Before-Break Evaluation at Critical Locations 4-13 a,c,e Material Characterization WCAP-15550-N P March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-14 a.. c.. e Figure 4-1 Pre-Service J vs. ~a for SA351-CF8M Cast Stainless Steel at 600°F Material Characterization WCAP-15550-NP March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-1 5.0 CRITICAL LOCATION AND EVALUATION CRITERIA 5.1 CRITICAL LOCATIONS The leak-before-break (LBB) evaluation margins are to be demonstrated for the limiting locations (governing locations). Such locations are established based on the loads (Section 3.0) and the material properties established in Section 4.0. These locations are defined below for Surry Units 1 and 2. Table 3-2 as well as Figure 3-2 are used for this evaluation.

Critical Locations The highest stressed location for the entire primary loop is at Location 1 (in the Hot Leg) (See Figure 3-2) at the reactor vessel outlet nozzle to pipe weld. Location 1 is critical for all the weld locations of pipe.

Since the elbows are made of cast materials, the critical locations for the elbows are: for the hot leg, the highest stressed location is at weld location 3; for the cross-over leg, the highest stressed location is at weld location 6; and for the cold leg; the highest stressed location is at weld location 15. It is thus concluded that the enveloping locations in Surry Units 1 and 2 for which LBB methodology is to be applied are locations 1, 3, 6 and 15. The tensile properties and the allowable toughness for the critical locations are shown in Tables 4-5 and 4-8.

5.2 FRACTURE CRITERIA As will be discussed later, fracture mechanics analyses are made based on loads and postulated flaw sizes related to leakage. The stability criteria against which the calculated J and tearing modulus are compared are:

(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 critical locations, the limit load method discussed in Section 7.0 was also used.

Critical Location and Evaluation Criteria WCAP-15550-NP March 2019 Revision 2

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 flow 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, L, to hydraulic diameter, DH, (UDH) 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 [

]8-c,e was found from Figure 6-2 (taken from Reference 6-2). For all cases considered, since [

]8-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, c, 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:

Ab~olute Pressure - 14. 7 = [

Leak Rate Predictions WCAP-15550-N P (6-2)

March 2019 Revision 2

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 Table 3-1 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 above. The average material properties of Section 4.0 (see Table 4-5) 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 Surry Units 1 and 2. The flaw sizes so determined are called leakage flaw sizes.

The Surry 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 10 on the leak rate, the flaw sizes (leakage flaw sizes) are determined which yield a leak rate of 10 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 WCAP-15550-NP March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3 Table 6-1 Flaw Sizes Yielding a Leak Rate of 10 gpm at the Governing Locations Location Leakage Flaw Size (in) 1 4.02 3

5.85 6

6.84 15 7.96 Note: The flaw size in the above table refers to the flaw length of through-wall circumferential crack.

Leak Rate Predictions WCAP-15550-N P March 2019 Revision 2

= -i -& ->... u 0...

ti C

2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-4 a,c,e STAGNATION ENTHALPY no2 Btu/lb)

Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures Leak Rate Predictions WCAP-15550-N P March 2019 Revision 2

Figure 6-2 [

Leak Rate Predictions WCAP-15550-NP WESTINGHOUSE NON-PROPRIETARY CLASS 3 LENGTH/OIAMET'EA RATIO (L/D) ia,c,e Pressure Ratio as a Function of UD 6-5 a,c,e March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-6

[

Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack Leak Rate Predictions WCAP-15550-N P March 2019 Revision 2

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:

where:

Tapp E

a cry, cru

=

=

=

=

=

dJ E

Tapp= -d Xz a at applied tearing modulus modulus of elasticity 0.5 (cry+ cru) = flow stress crack length 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.,

Fracture Mechanics Evaluation WCAP-15550-NP March 2019 Revision 2

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:

=

r,c,e 0.5 (cry+ cru) = flow stress, psi 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.

Fracture Mechanics Evaluation WCAP-15550-N P March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7.3 CRACK STABILITY EVALUATIONS Local Failure Mechanism:

Fracture Mechanics Evaluation WCAP-15550-N P 7-3 March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-4 Global failure mechanism:

A stability analysis based on limit load was performed for all the critical locations (locations 1, 3, 6 and 15) as described in Section 7.2.

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, 3 and 15. Shop weld is at critical location 6. The "Z" factor correction for SMAW was applied (References 7-5 and 7-6) at the field weld critical locations (locations 1, 3 and 15) and the "Z" factor correction for SAW was applied (References 7-5 and 7-6) at the shop weld location (location 6) and the equations as follows:

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

Z = 1.30 [1.0 + 0.01 (OD-4)]

where OD is the outer diameter of the pipe in inches.

for SMAW for SAW The Z-factors were calculated for the critical locations, using the dimensions given in Table 3-1.

The Z factor was 1.599 for locations 1 and 3, 1. 72 for location 6 and 1.592 for location 15. The applied loads were increased by the Z factors and plots of limit load versus crack length were generated as shown in Figures 7-2, 7-3, 7-4 and 7-5. Table 7-2 summarizes the results of the stability analyses based on limit load. The leakage flaw sizes are also presented on the same table.

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 Johnson, W. and Mellor, P. B., Engineering Plasticity, Van Nostrand Relmhold Company, New York, (1973), pp. 83-86.

7-3 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.

7-4 Irwin, G. R., "Plastic Zone near a Crack and Fracture Toughness," Proc.ih Sagamore Conference, P. IV-63 (1960).

7-5 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.

7-6 NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.

Fracture Mechanics Evaluation WCAP-15550-NP March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 7-1 Stability Results for Surry Units 1 and 2 Based on Elastic-Plastic J-lntegral Evaluations Table 7-2 Stability Results for Surry Units 1 and 2 Based on Limit Load Critical Location Critical Flaw Size (in)

Leakage Flaw Size (in) 1 19.87 4.02 3

34.08 5.85 6

34.81 6.84 15 40.07 7.96 7-5 Fracture Mechanics Evaluation WCAP-15550-N P March 2019 Revision 2 a,c,e

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Figure 7-1 [

Fracture Mechanics Evaluation WCAP-15550-N P Neutral Axis

]a,c,e Stress Distribution 7-6 March 2019 Revision 2

80000 70000 60000 II)

0.

50000

SC:

I C --

40000 C

Q)

E 0

30000

1E

.:==

E 20000

J 10000 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 0

10 20 30 Flaw Length (inches)

OD = 34.00 in.

t = 2.395 in.

cry

= 20. 76 ksi cru

= 56.20 ksi A376-TP316 Material with SMAW Weld F = 1640 kips M = 24646 in-kips 40 Figure 7-2 Critical Flaw Size Prediction - Hot Leg at Location 1 Fracture Mechanics Evaluation WCAP-15550-N P 7-7 50 March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-8.

100000 90000 80000 70000 0 Q.

i 60000 C:

50000 C:

Q)

E 40000 0

!2E

t:

30000 E

J 20000 (34_080 in_, 20650 in-kips) 10000 0

0 10 20 30 40 Flaw Length (inches)

OD= 34.00 in.

cry

= 23. 78 ksi F = 1639 kips t= 2.395 in.

cru

= 71.90 ksi M = 12918 in-kips A351-CF8M Material with SMAW Weld Figure 7-3 Critical Flaw Size Prediction - Hot Leg at Location 3 Fracture Mechanics Evaluation WCAP-15550-N P 50 March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-9 140000 120000

- 100000 II)

Q.

ii:

I.5 80000 C

Cl)

E 60000 0

E
t:: E 40000
i (34.814 in., 26960 in-kips) 20000 0

0 10 20 30 40 Flaw Length (inches)

OD = 36.32 in.

cry

= 24.67 ksi F = 1870 kips t = 2.555 in.

cru

= 71.90 ksi M = 15673 in-kips A351-CF8M material with SAW Weld Figure 7-4 Critical Flaw Size Prediction - Crossover Leg at Location 6 Fracture Mechanics Evaluation WCAP-15550-NP 50 March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-10 100000 Cl)

Q.

80000

.i:

I.5 -

C:

60000 G)

E 0

IS 40000 E
J 20000 0...... -

............. -+__._ _____ +-_............. -+__._ _____ +-_.............. -+..... ____ --...

0 10 20 30 40 50 Flaw Length (inches)

OD = 33.60 in.

cry

=24.67ksi F=1418kips t = 2.940 in.

cru

= 71.90 ksi M = 10829 in-kips A351-CF8M material with SMAW Weld Figure 7-5 Critical Flaw Size Prediction - Cold Leg at Location 15 Fracture Mechanics Evaluation WCAP-15550-NP 60 March 2019 Revision 2

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 [

rc,e region of a typical system (see Location [ ]8,c,e of Figure 3-2). 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 less than 10% variation.

A [

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

[

]a,c,e The normal, upset, and test conditions were considered. A summary of generic applied transients is provided in Table 8-1. Circumferentially oriented surface flaws were postulated in the region, assuming the flaw was located in two different locations, as shown in Figure 8-1. Specifically, these were:

Cross Section A: Stainless Steel Cross Section B: SA 508 Cl. 2 or 3 Low Alloy Steel Fatigue crack growth rate laws were used [

]8,c,e The law for stainless steel was derived from Reference 8-1, a compilation of data for austenitic stainless steel in a PWR water environment was presented in Reference 8-2, and it was found that the effect of the environment on the crack growth rate was very small.

From this information it was estimated that the environmental factor should be conservatively set at [ rc,e in the crack growth rate equation from Reference 8-1.

For stainless steel, the fatigue crack growth formula is:

Fatigue Crack Growth Analysis WCAP-15550-NP March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 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 Surry 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 Surry 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 Surry transients and cycles. The transients and cycles (shown in Table 8-3) for the Surry 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.

It is therefore, concluded that the generic fatigue crack growth analysis shown in Table 8-2 is representative of the Surry plants fatigue crack growth and also applicable for 80 years.

8.1 REFERENCES

8-1 James, L.A. and Jones, D.P., "Fatigue Crack Growth Correlations for Austenitic Stainless Steel in Air, Predictive Capabilities in Environmentally Assisted Cracking," ASME Publication PVP-99, December 1985.

8-2 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. 1.979.

Fatigue Crack Growth Analysis WCAP-15550-N P March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 8-1 Summary of Reactor Vessel Transients Number Typical Transient Identification 1

Turbine Roll 2

Cold Hydro 3

Heatup/Cooldown 4

Loading and Unloading 5

10% Step Load Decrease/Increase 6

Large Step Decrease 7

Steady State Fluctuation 8

Loss of Load from Full Power 9

Loss of Power 10 Partial Loss of Flow 11 Reactor Trip from Full Power 12 Hot Hydro Test Fatigue Crack Growth Analysis WCAP-15550-N P 8-3 Number of Cycles 10 10 200 18300 2000 200 1000000 80 40 80 400 50 March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 8-2 Typical Fatigue Crack Growth at [

Fatigue Crack Growth Analysis WCAP-15550-NP (40, 60, and 80 years)

FINAL FLAW ( in.)

1 a,c,e 8-4 March 2019 Revision 2 a,c,e

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-5 Table 8-3 Summary of Reactor Vessel Transients For Surry Units 1 and 2 (40, 60, 80 years)

Number Typical Transient Identification 1

Heatu p/Cooldown 2

Loading and Unloading 3

10% Step Load Decrease/Increase 4

Large Step Decrease 5

Reactor Trip from Full Power 6

Hot Hydro Test Fatigue Crack Growth Analysis WCAP-15550-N P Number of Cycles 200 18300 2000 200 400 45 March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Figure 8-1 Typical Cross-Section of [

Fatigue Crack Growth Analysis WCAP-15550-N P

]a,c,e 8-6 March 2019 Revision 2

.!! u > u....

J::

u C *e u i

§

~

e a:

J:: j e t.7 u

I! u WESTINGHOUSE NON-PROPRIETARY CLASS 3 1000 ----------------------------~

700 500 200 100 70 50 20 10 7

5

  • Linear interpolation is recom-mended to account for ratio dependence of water environment curves, for 0.25 <R< 0.65 for shallow slope:

~ = (1.01 X 10*11 o2 ti. K1.95 dN 02 = 3.75 R +0.06 R = /(min /Kmax Subsurface fla\\NS lair environment)

Determine the ti.K at which the law changes by calculation of the intersection of the two curves.

2 Surface flaws (water reactor environment) applicable for R < 0.25 "0.25 < R < 0.65 R;, 0.65 R = Kmin /Krnax 5

7 10 20 Stress lntensitv Factor Range (ti.K1 ksi ~.I

  • Linear interpolation is recommended to account for R ratio dependence of water environment curves, for 0.25 < R < 0.65 fer steep s!ope:

da = (1.02 X 10-61 01 ti.K5*95 dN 01 = 26.9R

  • 5.725 R = Kmin /Kmax 50 70 100 8-7 Figure 8-2 Reference Fatigue Crack Growth Curves for Carbon & Low Alloy Ferritic Steels Fatigue Crack Growth Analysis WCAP-15550-NP March 2019 Revision 2

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, 7.2 and 7.3 are used in performing the assessment of margins. Margins are shown in Table 9-1. All of the LBB recommended margins are satisfied.

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 10 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.

Assesment of Margins WCAP-15550-N P March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-2 Table 9-1 Leakage Flaw Sizes, Critical Flaw Sizes and Margins for Surry Units 1 and 2 Location Leakage Flaw Size 1

4.02 in.

3 5.85 in.

3 5.85 in.

6 6.84 in.

6 6.84 in.

15 7.96 in.

15 7.96 in.

"based on limit load bbased on J integral evaluation Assesment of Margins WCAP-15550-NP Critical Flaw Size 19.87 8 in.

34.08 8 in.

11.70bin.

34.81 8 in.

13.68b in.

40.07 8 in.

15.92b in.

Margin 4.9 8

5.8 8

>2.0b 5.1 8

>2.0b 5.0 8

>2.0b March 2019 Revision 2

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 Surry 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. There is no Alloy 82/182 material present in the welds for the Surry Units 1 and 2 Reactor Coolant System (RCS) primary loop piping.

  • b.

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

c.

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

d.

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

e.

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

f.

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 d, e, and f above.

Based on the above, the Leak-Before-Break conditions and margins are satisfied for the Surry 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 Surry Units 1 and 2 Nuclear Power Plants for the 80 year plant life (subsequent license renewal program).

Conclusions WCAP-15550-NP March 2019 Revision 2

Appendix A: Limit Moment WCAP-15550-N P WESTINGHOUSE NON-PROPRIETARY CLASS 3 APPENDIX A: LIMIT MOMENT

]a,c,e A-1 March 2019 Revision 2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 a, 0.-----------------------

ca Figure A-1 Pipe with a Through-Wall Crack in Bending Appendix A: Limit Moment WCAP-15550-NP A-2 March 2019 Revision 2

WCAP-15550-NP Revision 2 Proprietary Class 3

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Author Approval Wiratmo Mamo Mar-14-2019 12:06:23 Reviewer Approval Johnson Eric D Mar-14-2019 12:16:52 Manager Approval Leber Benjamin A Mar-14-2019 12:27:28 Files approved on Mar-14-2019

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