RA-22-0290, License Amendment Request to Exclude the Dynamic Effects of Specific Postulated Pipe Ruptures from the Design and Licensing Basis Based on Leak-Before-Break Methodology

License Amendment Request to Exclude the Dynamic Effects of Specific Postulated Pipe Ruptures from the Design and Licensing Basis Based on Leak-Before-Break Methodology

ML23242A086
Person / Time
Site:	Robinson
Issue date:	08/30/2023
From:	Basta L Duke Energy Progress
To:	Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML23242A084	List: RA-22-0290, License Amendment Request to Exclude the Dynamic Effects of Specific Postulated Pipe Ruptures from the Design and Licensing Basis Based on Leak-Before-Break Methodology
References
RA-22-0290
Download: ML23242A086 (1)

Text

Laura A. Basta H. B. Robinson Steam Electric Plant Unit 2 Site Vice President Duke Energy 3581 West Entrance Road Hartsville, SC 29550 O: 843 951 1701 Laura.Basta@duke-energy.com PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENTS 6, 7, AND 8 THIS LETTER IS UNCONTROLLED Serial: RA-22-0290 August 30, 2023 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 H. B. Robinson Steam Electric Plant, Unit No. 2 Docket No. 50-261 / Renewed License Number DPR-23

SUBJECT:

License Amendment Request to Exclude the Dynamic Effects of Specific Postulated Pipe Ruptures from the Design and Licensing Basis Based on Leak-Before-Break Methodology Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy) is submitting a request for an amendment to the Facility Operating License for H. B. Robinson Steam Electric Plant, Unit No. 2 (RNP). The proposed amendment uses the Leak-Before-Break (LBB) methodology to eliminate the dynamic effects of postulated pipe ruptures to auxiliary piping systems attached to the Reactor Coolant System (RCS) from the RNP design and licensing basis. There are no proposed changes to the Technical Specifications associated with this License Amendment Request (LAR).

This LAR is submitted in accordance with 10 CFR 50, Appendix A, General Design Criterion (GDC) 4, "Environmental and dynamic effects design bases," following the guidance of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition", Section 3.6.3, "Leak-Before-Break Evaluation Procedures".

As noted in GDC 4, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping. An evaluation of the proposed change is provided in Attachment 1.

The supporting technical bases for applying LBB methodology to the RCS auxiliary piping for RNP is provided by the following documents:

• WCAP-17776-P/NP, Revision 1, "Technical Justification for Eliminating Pressurizer Surge Line Rupture as the Structural Design Basis for H. B. Robinson, Unit 2, March 2023,

• WCAP-17778-P/NP, Revision 1, "Technical Justification for Eliminating Residual Heat Removal (RHR) Line Rupture as the Structural Design Basis for H. B. Robinson Unit 2, March 2023,

• WCAP-17779-P/NP, Revision 1, "Technical Justification for Eliminating Accumulator Line Rupture as the Structural Design Basis for H. B. Robinson Unit 2, March 2023 PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENTS 6, 7, AND 8 THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENTS 6, 7, AND 8 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-22-0290 Page 2 WCAP-17776-P, Revision 1, WCAP-17778-P, Revision 1, and WCAP-17779-P, Revision 1 (Attachments 6, 7 and 8 respectively) include information proprietary to Westinghouse Electric Company, LLC (Westinghouse) , and the Westinghouse documents CAW-23-008 , CAW-23-009, and CAW-23-010 in Attachment 5 provide the affidavits signed by the owner of the information .

Accordingly, it is respectfully requested that the proprietary information be withheld from public disclosure in accordance with 10 CFR 2.390. The redacted , non-proprietary versions are provided in Attachments 2, 3, and 4. Correspondence with respect to the proprietary aspects of the affidavits should be addressed to the Westinghouse representative identified in the respective affidavits.

Duke Energy has evaluated the proposed amendment and has determined it does not involve a significant hazards consideration as defined in 10 CFR 50.92. The basis for this determination is included in Attachment 1.

Duke Energy requests approval of the proposed license amendment within one year of completion of the NRC's acceptance review to support the Subsequent License Renewal Application for RNP . Following NRC approval , Duke Energy will implement the amendment within 120 days.

In accordance with 10 CFR 50.91, Duke Energy is notifying the state of South Carolina of this license amendment request by transmitting a copy of this letter to the designated state officials.

Should you have any questions concerning this letter, or require additional information , please contact Ryan Treadway, Director - Nuclear Fleet Licensing at (980) 373-5873 .

This submittal contains no new regulatory commitments.

I declare under penalty of perjury that the foregoing is true and correct. Executed on August 30, 2023.

Sincerely, Laura A. Basta Site Vice President PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENTS 6, 7, AND 8 THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENTS 6, 7, AND 8 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-22-0290 Page 3 Attachments:

1. Evaluation of the Proposed Change

2. WCAP-17776-NP, Revision 1, "Technical Justification for Eliminating Pressurizer Surge Line Rupture as the Structural Design Basis for H. B. Robinson Unit 2, March 2023 (Redacted)

3. WCAP-17778-NP, Revision 1, "Technical Justification for Eliminating Residual Heat Removal (RHR) Line Rupture as the Structural Design Basis for H. B. Robinson Unit 2, March 2023 (Redacted)

4. WCAP-17779-NP, Revision 1, "Technical Justification for Eliminating Accumulator Line Rupture as the Structural Design Basis for H. B. Robinson Unit 2, March 2023 (Redacted)

5. Westinghouse Affidavits

6. WCAP-17776-P, Revision 1, "Technical Justification for Eliminating Pressurizer Surge Line Rupture as the Structural Design Basis for H. B. Robinson Unit 2, March 2023 (Proprietary)

7. WCAP-17778-P, Revision 1, "Technical Justification for Eliminating Residual Heat Removal (RHR) Line Rupture as the Structural Design Basis for H. B. Robinson Unit 2, March 2023 (Proprietary)

8. WCAP-17779-P, Revision 1, "Technical Justification for Eliminating Accumulator Line Rupture as the Structural Design Basis for H. B. Robinson Unit 2, March 2023 (Proprietary)

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENTS 6, 7, AND 8 THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENTS 6, 7, AND 8 THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission RA-22-0290 Page 4 cc: (Without Attachments)

L. Dudes, Regional Administrator USNRC Region II J. Zeiler, NRC Senior Resident Inspector L. Haeg, NRR Project Manager A. Wilson, Attorney General (SC)

R. S. Mack, Assistant Bureau Chief, Bureau of Environmental Health Services (SC)

L. Garner, Manager, Radioactive and Infectious Waste Management Section (SC)

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENTS 6, 7, AND 8 THIS LETTER IS UNCONTROLLED

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 ATTACHMENT 1 EVALUATION OF THE PROPOSED CHANGE

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 1 of 21 EVALUATION OF THE PROPOSED CHANGE TABLE OF CONTENTS 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specifications Requirements 2.3 Reason for the Proposed Change 2.4 Description of the Proposed Change

3.0 TECHNICAL EVALUATION

3.1 Comparison of WCAP Topical Reports to NUREG-0800, Section 3.6.3.III, Review Procedures 3.2 Margin Assessment 3.3 Reactor Coolant System Leakage Detection Systems 3.4 Conclusion

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination 4.4 Conclusions

6.0 REFERENCES

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 2 of 21 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy) requests an amendment to the Facility Operating License for H. B. Robinson Steam Electric Plant, Unit No. 2 (RNP).

NRC approval is requested for application of the leak-before-break (LBB) methodology to auxiliary piping systems attached to the Reactor Coolant System (RCS) for RNP to eliminate the dynamic effects of postulated pipe ruptures.

This license amendment request (LAR) is submitted in accordance with 10 CFR 50, Appendix A, General Design Criterion (GDC) 4, "Environmental and dynamic effects design bases,"

following the guidance of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Section 3.6.3, "Leak-Before-Break Evaluation Procedures." The LAR applies LBB methodology to demonstrate the risk of pipe rupture is extremely low for portions of auxiliary lines attached to the Reactor Coolant Loops (RCLs).

No changes to the Technical Specifications (TS) are required by this LAR.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation RNP is a three-loop Westinghouse pressurized water reactor. As described in the Updated Final Safety Analysis Report (UFSAR) Section 3.6.1, the current design basis includes application of LBB to the RCS primary loop piping. This LAR would expand the scope of the LBB methodology to include specific portions of piping systems attached to the RCS. The auxiliary lines attached to the RCLs included in the scope of this proposed change include:

• The Pressurizer Surge Line from the primary loop nozzle junction (i.e., weld that connects the nozzle to the surge line piping) to the pressurizer nozzle junction (i.e., weld that connects the pressure surge nozzle safe end to the pressurizer surge nozzle)

• The Residual Heat Removal (RHR) Lines, limited to the high energy Class 1 portion of the RHR lines (primary loop junction to the second isolation valve)

• The 10-inch Accumulator Lines (from the cold legs Loop A, Loop B and Loop C) and attached 8-inch line connected to 10-inch accumulator lines except for the piping upstream of Valves SI-875D, SI-875E, and SI-875F Pressurizer Surge Line - Pressurizer pressure is transmitted to the remainder of the RCS via the surge line that connects the bottom of the pressurizer with the RCS hot leg piping. The pressurizer surge line connects the bottom of the pressurizer to the hot leg of RCL 3.

Residual Heat Removal Lines - The RHR system is a low-pressure, low-temperature fluid system that is not used during power operation. The system is designed to operate at pressures less than 375 pounds per square inch gauge (psig) and at temperatures less than 350 degrees Fahrenheit (°F). The system is operated during plant cooldown after RCS pressure and temperature are within RHR system limitations. The primary purpose of the RHR system is to remove decay heat energy generated in the reactor core during plant cooldown and refueling operations. During plant shutdown and refueling, reactor coolant is drawn from the hot leg of RCS Loop 2 by the RHR pumps, discharged through the tube side of the RHR heat exchangers, and returned to the RCS via all three cold legs.

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 3 of 21 Accumulator Lines - An accumulator filled with borated water and pressurized with nitrogen is connected to each RCS cold leg. When RCS pressure drops below the nitrogen pressure setpoint, the accumulators discharge their borated water into the RCS. This action provides rapid refilling of the lower core plenum in the event of a large break in the RCS.

Materials that are susceptible to Primary Water Stress Corrosion Cracking (PWSCC), such as Alloy 600 and Alloy 82/182 weld metal, are not found at the RNP RHR line, Accumulator line, or Pressurizer Surge Line. In addition, the RNP RHR line, Accumulator line, and Pressurizer Surge Line do not contain Cast Austenitic Stainless Steel (CASS) material.

Figure 1 Reactor Coolant Loop Piping

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 4 of 21 2.2 Current Technical Specifications Requirements RNP Technical Specifications Limiting Condition for Operation (LCO) 3.4.13, specifies that RCS operational leakage shall be limited to.

a) No pressure boundary leakage, b) 1 gpm unidentified leakage, c) 10 gpm identified leakage, and d) 75 gallons per day primary to secondary leakage through any one Steam Generator Applicability includes Modes 1-4. Actions for specified conditions are as described in LCO 3.4.13. Surveillance Requirements (SR) for RCS operational leakage are provided in SR 3.4.13.1 and 13.4.13.2 RNP Technical Specifications LCO 3.4.15, specifies that the following RCS leakage detection instrumentation shall be operable:

a) One containment sump level monitor, b) One containment atmosphere radioactivity monitor (gaseous or particulate), and c) One containment fan cooler condensate flow rate monitor Applicability includes Modes 1-4. Actions for specified conditions are as described in LCO 3.14.15. SR for RCS leakage detection instrumentation are provided in SR 3.4.15.1 through 13.4.15.5 2.3 Reason for Proposed Change Duke Energy is requesting the proposed amendment to apply LBB analyses to the RCS branch piping to facilitate potential future plant changes and Subsequent License Renewal Application (SLRA) for RNP. As stated in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Section 3.6.3, "Leak-Before-Break Evaluation Procedures," NRC staff approval of an LBB analysis permits an operating plant licensee to "remove protective hardware such as pipe whip restraints and jet impingement barriers, redesign pipe connected components, their supports and their internals, and other related changes."

During preparation of the SLRA for RNP it was determined that there were selected issues that would benefit significantly from a reduction in RCS loads through extending the LBB methodology to the branch lines in the scope of this LAR.

Reduced loads will be used to regain margin for the following NRC focus areas for subsequent license renewal:

1. Equivalent Margins Analyses of the Reactor Pressure Vessel (RPV) upper shell assembly and RPV nozzles,

2. Fracture Mechanics Evaluation of the RPV supports located beneath each of the three RPV inlet nozzles, and

3. Assessment of potential flaws in the core barrel for American Society of Mechanical EngineersSection XI or MRP-227 inspections The RPV upper shell assembly, RPV nozzles, and RPV support assemblies are all susceptible to reduction of fracture toughness owing to additional neutron exposure associated with operation to 80-years. As such, structural integrity of these items with reduced fracture toughness will be evaluated as part of the SLRA as specified in NUREG-2192, "Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants."

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 5 of 21 2.4 Description of Proposed Change The proposed change would revise the RNP design and licensing basis to expand the scope of NRC staff's approval of LBB to auxiliary piping connected to the RCLs. The use of LBB for RNP is currently limited to the large, primary loop RCS piping, as discussed in UFSAR Section 3.6.1.

The expanded scope LBB would eliminate the dynamic effects of postulated ruptures of specific portions of piping for the Pressurizer Surge Line, the RHR lines, and the Accumulator Lines.

Upon implementation of this LAR, relevant sections of RNP UFSAR will be updated to reflect this expanded scope of LBB methodology.

3.0 TECHNICAL EVALUATION

The LBB concept is based on calculations and experimental data demonstrating that certain pipe material has sufficient fracture toughness (ductility) to prevent a small through-wall flaw from propagating rapidly and uncontrollably to catastrophic pipe rupture and to ensure that the probability of a pipe rupture is extremely low. The small leaking flaw is demonstrated to grow slowly, and the limited leakage would be detected by the RCS leakage detection systems early on such that licensees can shut down the plant to repair the degraded pipe long before the potential catastrophic pipe rupture.

While the dynamic effects of pipe breaks have been eliminated for the RNP RCL piping, additional breaks remain applicable for the auxiliary piping systems connected to the RCLs. The auxiliary piping systems attached to the RCLs within the scope of this LAR include the following:

• The Pressurizer Surge Line from the primary loop nozzle junction (i.e., weld that connects the nozzle to the surge line piping) to the pressurizer nozzle junction (i.e., weld that connects the pressure surge nozzle safe end to the pressurizer surge nozzle)

• The Residual Heat Removal (RHR) Lines, limited to the high energy Class 1 portion of the RHR lines (primary loop junction to the second isolation valve)

• The 10-inch Accumulator Lines (from the cold legs Loop A, Loop B and Loop C) and attached 8-inch line connected to 10-inch accumulator lines except for the piping upstream of Valves SI-875D, SI-875E, and SI-875F to this LAR provides WCAP-17776-P, Revision 1, "Technical Justification for Eliminating Pressurizer Surge Line Rupture as the Structural Design Basis for H. B. Robinson, Unit 2. WCAP-17776-P, Revision 1 provides a description of a mechanistic pipe break evaluation method and the analytical results that can be used for establishing that a circumferential type of break will not occur within the pressurizer surge line. The evaluations consider that circumferentially oriented flaws cover longitudinal cases. Additionally, a fracture mechanics analysis that demonstrates the pressurizer surge line integrity for RNP consistent with the NRCs position for exemption from consideration of postulated pipe rupture dynamic effects is presented. The pressurizer surge line is known to be subjected to thermal stratification and the effects of thermal stratification for the RNP pressurizer surge line have been used in the LBB evaluation presented in WCAP-17776-P, Revision 1. Attachment 2 to this LAR provides a non-proprietary version of WCAP-17776-P, Revision 1. Figure 2 below (from WCAP-17776-P Figure 3-1) shows the Pressurizer Surge line layout.

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 6 of 21 Figure 2 Pressurizer Surge Line Layout to this LAR provides WCAP-17778-P, Revision 1, "Technical Justification for Eliminating RHR Line Rupture as the Structural Design Basis for H. B. Robinson, Unit 2.

WCAP-17778-P, Revision 1 provides a description of a mechanistic pipe break evaluation method and the analytical results that can be used for establishing that a circumferential type of break will not occur within the RHR lines. Consistent with the pressurizer surge line reported in WCAP-17776-P, Revision 1, the evaluations consider that circumferentially oriented flaws cover longitudinal cases. Additionally, a fracture mechanics demonstration of the RHR integrity for RNP consistent with the NRC position for exemption from consideration of postulated pipe rupture dynamic effects is presented. Attachment 3 to this LAR provides a non-proprietary version of WCAP-17778-P, Revision 1. Figure 3 below (from WCAP-17778-P Figure 3-1) shows the RHR line layout.

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 7 of 21 Figure 3 RHR Line Layout

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 8 of 21 to this LAR provides WCAP-17779-P, Revision 1, "Technical Justification for Eliminating Accumulator Line Rupture as the Structural Design Basis for H. B. Robinson, Unit

2. WCAP-17779-P, Revision 1 provides a description of a mechanistic pipe break evaluation method and the analytical results that can be used for establishing that a circumferential type of break will not occur within the accumulator lines that includes the 10-inch accumulator lines (from the cold legs Loop 1, Loop 2, and Loop 3) and attached 8-inch line connected to the 10-inch accumulator lines. NRC Standard Review Plan (SRP) Section 3.6.3, "Leak-Before-Break Evaluation Procedures," Revision 1 requires that LBB should only be applied to high energy, ASME Code Class 1 or 2 piping. As noted in WCAP-17779-P, Revision 1, the scope includes the 10-inch Class 2 and Class 1 Safety Injection System piping from the accumulators to each of the cold legs. Although WCAP-17779-P, Revision 1 includes the Class 2 piping, the scope requested in this LAR does not include the Class 2 piping (piping upstream of Check Valves SI-875-D, SI-875-E, and SI-875-F) since this piping is not a high energy line (temperature less than 200 OF). Figures 4, 5, and 6 below (from WCAP-17779-P Figures 3-1, 3-2, and 3-3) show the Accumulator line layouts. Consistent with the pressurizer surge line reported in WCAP-17776-P, Revision 1, the evaluations consider that circumferentially oriented flaws cover longitudinal cases. Additionally, a fracture mechanics demonstration of accumulator line piping integrity for RNP consistent with the NRC position for exemption from consideration of postulated pipe rupture dynamic effects is presented. Attachment 4 to this LAR provides a non-proprietary version of WCAP-17779-P, Revision 1.

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 9 of 21 Figure 4 Loop A Accumulator Line Layout SI-875-D

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 10 of 21 Figure 5 Loop B Accumulator Line Layout SI-875-E

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 11 of 21 Figure 6 Loop C Accumulator Line Layout SI-875-F 3.1 Comparison of WCAP Topical Reports To NUREG-0800, Section 3.6.3.III, Review Procedures The branch line Technical Documents (WCAP-17776, Revision 1; WCAP-17778, Revision 1; and WCAP-17779, Revision 1) have not been reviewed and approved by the NRC. Therefore, the following table presents SRP Section 3.6.3.III requirements and sections of the WCAP that address the specific requirements.

Table 1 Compliance with SRP 3.6.3 SRP 3.6.3, III, Subparagraphs WCAP Sections That Address Requirement SRP Requirement

1. The reviewer should verify that the licensees or Sections 3.1 through 3.6 for applicants LBB evaluation uses design basis loads WCAP-17776-P/NP, Revision 1, and is based on the as-built piping configuration, as WCAP-17778-P/NP, Revision 1, opposed to the design configuration. and WCAP-17779-P/NP, Revision 1

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 12 of 21 Table 1 Compliance with SRP 3.6.3 SRP 3.6.3, III, Subparagraphs WCAP Sections That Address Requirement SRP Requirement

2. The reviewer should evaluate the potential for Sections 2.1 through 2.4 WCAP-degradation by erosion, erosion/corrosion, and 17776-P/NP, Revision 1, WCAP-erosion/cavitation due to unfavorable flow conditions 17778-P/NP, Revision 1, and and water chemistry. Industry experience for specific WCAP-17779-P/NP, Revision 1.

piping systems plays an important role in the evaluation of these degradation mechanisms.

Additionally, an evaluation of wall thinning of elbows and other fittings is undertaken to ensure that American Society of Mechanical Engineers Code minimum wall requirements are met

3. The review should evaluate the material Section 2.1 for WCAP-17776-susceptibility to corrosion, the potential for high P/NP, Revision 1, WCAP-17778-residual stresses, and environmental conditions that P/NP, Revision 1, and WCAP-could lead to degradation by stress corrosion 17779-P/NP, Revision 1.

cracking. Primary water stress corrosion cracking (PWSCC) is considered to be an active degradation mechanism in Alloy 600/82/182 materials in pressurized water reactor plants

4. The reviewer should evaluate the adequacy of the Section 6.4 for WCAP-17776-leakage detection systems associated with the P/NP, Revision 1, WCAP-17778-reactor coolant system P/NP, Revision 1, and WCAP-17779-P/NP, Revision 1 and section 2.2 of this Evaluation.

5. The reviewer should verify that the potential for Section 2.2 for WCAP-17776-water hammer in the candidate piping systems is P/NP, Revision 1, WCAP-17778-very low P/NP, Revision 1, and WCAP-17779-P/NP, Revision 1.

6. The reviewer should verify that the candidate piping Section 2.4 for WCAP-17776-is not susceptible to creep and creep-fatigue. P/NP, Revision 1, WCAP-17778-Operation below 700°F in ferritic steel piping and P/NP, Revision 1, and WCAP-below 800°F in austenitic steel piping can alleviate 17779-P/NP, Revision 1.

concerns of creep.

7. The reviewer should evaluate the corrosion Section 2.1 for WCAP-17776-resistance of piping, which can be demonstrated by P/NP, Revision 1, WCAP-17778-the frequency and degree of corrosion in the specific P/NP, Revision 1, and WCAP-piping systems 17779-P/NP, Revision 1.

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 13 of 21 Table 1 Compliance with SRP 3.6.3 SRP 3.6.3, III, Subparagraphs WCAP Sections That Address Requirement SRP Requirement

8. The reviewer should assess the potential for indirect All of the piping segments within sources of pipe ruptures to ensure that indirect the scope of this LAR are inside failure mechanisms defined in the plant SAR are containment and protected negligible causes of pipe rupture. Compliance with relative to missiles and dynamic the snubber surveillance requirements of the effects. Missiles and dynamic technical specifications ensures that snubber failure effects are discussed in UFSAR, rates are acceptably low. Sections 3.5.1 and 3.6.1.

9. The reviewer should determine that the piping There are no damage material will not become susceptible to brittle mechanisms that can lead to cleavage-type failures over the full range of system reduction of fracture toughness of operating temperatures (that is, the material is on the the piping materials, radiation upper shelf of the Charpy Impact energy versus test levels are low and there are no temperature curve). CASS product forms.

10. The reviewer should determine that the candidate Sections 2.3 and 8.0 for WCAP-piping does not have a history of fatigue cracking or 17776-P/NP, Revision 1, WCAP-failure. An evaluation to ensure that the potential for 17778-P/NP, Revision 1, and pipe rupture due to thermal and mechanical induced WCAP-17779-P/NP, Revision 1.

fatigue is unlikely should be performed.

11. The following steps constitute an acceptable Sections 4, 5, 6, and 7 for deterministic LBB evaluation procedure: WCAP-17776-P/NP, Revision 1, WCAP-17778-P/NP, Revision 1, and WCAP-17779-P/NP, Revision 1.

3.2 Margin Assessment The results of the leak rates of Section 6.4 and the corresponding stability evaluations of Section 7.2 for each of the three attached WCAPs are used in performing the assessment of margins. All 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.

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 14 of 21 3.3 Reactor Coolant System Leakage Detection Systems As discussed in GL 80-04 and SRP 3.6.3, the licensee leakage detection systems should be sufficient to provide adequate margin to detect the leakage from a postulated circumferential throughwall flaw. The leak detection systems associated with the RCS are described in the UFSAR, Section 5.2.5, and consist of the following: (1) two radiation sensitive instruments, (2) humidity detector, (3) a condensate monitoring system which determines leakage losses from all water and steam systems within the containment, including that from the RCS, and (4) an increase in the amount of coolant makeup water which is required to maintain normal level in the pressurizer, or an increase in containment sump level are also used as leakage detection systems.

The following requirement is provided in Section 5.2.5 of the UFSAR: To support the application of Leak Before Break methodology, at least one leakage detection system must be operable with a sensitivity capable of detecting a 1 gallon per minute leak within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This requirement was implemented when RNP incorporated WCAP-9558 and WCAP-9787 (References 1 and 2, respectively) into the Current Licensing Basis (CLB) for LBB of main coolant piping. This requirement is an exception to the guidance in Regulatory Guide 1.45, which requires detection of 1 gallon(s) per minute (gpm) leakage within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The capability of the RNP leak detection systems to detect 1 gpm leakage is within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, consistent with the conditions of Generic Letter 84-04 (Reference 3).

3.4 Conclusion The elimination of pressurizer surge line breaks, the RHR line breaks, and the accumulator line breaks from the structural design basis for RNP is justified 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.

b. Water hammer should not occur in the pressurizer surge piping, the RHR piping, and the accumulator piping because of system design, testing, and operational considerations.

c. The effects of low and high cycle fatigue on the integrity of the pressurizer surge line piping, the RHR line piping, and the accumulator line piping are negligible.

d. Ample margin exists between the leak rate of small stable flaws and the capability of the RNP, Unit 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 stability of the critical flaws.

g. Fatigue crack growth results using the 40-year design transients and cycles (shown to be applicable for 60 years) show that there will be insignificant growth through the wall for the license renewal period (60-year plant life).

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 loading, pipe geometry and pipe material properties considerations, enveloping critical (governing) locations were determined at which leak-before-break crack stability evaluations were made. Through-wall flaw sizes were postulated 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 pressurizer surge line piping, RHR line piping, and accumulator line piping.

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 15 of 21 Therefore, the LBB conditions and margins are satisfied for the RNP pressurizer surge line piping, RHR line piping, and accumulator line piping. It is demonstrated that the dynamic effects of the pipe rupture resulting from postulated breaks in the pressurizer surge piping, the RHR piping, and the accumulator piping need not be considered in the structural design basis of RNP for the license renewal period (60-year plant life).

4.0 REGULATORY EVALUATION

The proposed change would revise RNP licensing and design bases to expand the LBB scope to eliminate the dynamic effects of postulated ruptures of specific portions of piping for the pressurizer surge line, RHR lines, and accumulator lines. The following regulatory requirements have been reviewed and a No Significant Hazards Consideration Determination has been performed as discussed below.

4.1. Applicable Regulatory Requirements/Criteria Current General Design Criteria (GDC), Criterion 4 - Environmental and dynamic effects design bases states that structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.

During the initial plant licensing of RNP, it was demonstrated that the design of the reactor coolant pressure boundary met the regulatory requirements in place at that time. The GDC included in Appendix A to 10 CFR Part 50 did not become effective until May 21, 1971. The construction permit for RNP was issued prior to May 21, 1971; consequently, RNP is not subject to current GDC requirements (SECY-92-223, dated September 18, 1992, ADAMS Accession Number ML18100B279). RNP's conformance with the GDC in existence at the time RNP was licensed (contained in Proposed Appendix A to 10 CFR 50, "General Design Criteria for Nuclear Power Plants," published in the Federal Register on July 11, 1967) is described in UFSAR Section 3.1.2. As defined in UFSAR Section 3.6.2, high energy piping systems are those whose service temperature exceeds 200°F and whose design pressure exceeds 275 psig.

This LAR is based on evaluations to demonstrate that the piping in the scope of the request has an extremely low probability of rupture, consistent with the contemporary version of GDC 4.

The following information demonstrates compliance with GDC 14, 30, and 31 of 10 CFR 50, Appendix A. GDC 14, 30, and 31 of 10 CFR 50, Appendix A states:

Criterion 14 - Reactor coolant pressure boundary. The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

Criterion 30 - Quality of reactor coolant pressure boundary. Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 16 of 21 Criterion 31 - Fracture prevention of reactor coolant pressure boundary. The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.

NRC GDC 14 and 30 are similar to Atomic Energy Commission (AEC) Criterion 9 (RNP UFSAR Section 3.1.2.9), Reactor Coolant Pressure Boundary, and AEC Criterion 16, Monitoring Reactor Coolant Leakage, (RNP UFSAR Section 3.1.2.16). NRC GDC 31 is similar to AEC Criterion 34, Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention, (RNP UFSAR Section 3.1.2.34).

The piping in the scope of this LAR is designated Class 1 (Class 1 portion is defined herein as piping that is within the American Society of Mechanical EngineersSection XI Subsection, IWB inspection boundary) reactor coolant pressure boundary piping whose materials, design and as-built configuration, analysis, fabrication, and testing preclude the possibility of gross rupture or significant leakage, as supported by the enclosed LBB evaluations based on as-built configuration, material properties, and design transients. This LAR also addresses the capability to detect and respond to piping system leakage prior to a potential flaw reaching a critical size. Therefore, the request is consistent with GDC 14, 30, and 31.

The reactor coolant pressure boundary is designed, fabricated, and constructed to have an exceedingly low probability of gross rupture or significant uncontrolled leakage throughout its design lifetime. Reactor coolant pressure boundary piping and components have provisions for inspection, testing and surveillance of critical areas by appropriate means to assess the structural and leak tight integrity of the boundary components during their service lifetime. The TS reactor coolant system leakage limits ensure the reactor coolant pressure boundary will retain adequate structural and leakage integrity during normal operating, transient, and postulated accident conditions.

NRC Standard Review Plan (SRP) Section 3.6.3, "Leak-Before-Break Evaluation Procedures,"

Revision 1, provides guidance for review of the LBB application, including guidance for determining an acceptable leakage crack and the reactor coolant system leakage detection sensitivity based on the fracture mechanics analysis. The guidance states that determination of leakage from a crack in a piping system under pressure involves uncertainties and, therefore, margins are needed.Section III.4 of SRP 3.6.3 states that the NRC staff evaluates the proposed leakage detection systems to determine whether they are sufficiently reliable, redundant, and sensitive so that a margin on the detection of unidentified leakage exists for through-wall flaws to support the deterministic fracture mechanics evaluation. The guidance specifies that the predicted leakage rate from the postulated leakage crack should be a factor of 10 times greater than the minimum leakage the detection system is capable of sensing unless the licensee provides justification accounting for the effects of uncertainties in the leakage measurement.

The guidance of SRP Section 3.6.3 also states that specifications for plant specific leakage detection systems inside the containment should be equivalent to those in Regulatory Guide (RG) 1.45, Revision 0, "Reactor Coolant Pressure Boundary Leakage Detection Systems." The RNP reactor coolant system pressure boundary leak detection system, consistent with the conditions of Generic Letter 84-04, meets the intent of Regulatory Guide 1.45 and meets a leak

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 17 of 21 detection capability of 1 gpm. The conditions provided in Generic Letter 84-04 were implemented when RNP incorporated WCAP-9558 and WCAP-9787 (References 1 and 2, respectively) into the CLB for LBB of main coolant piping. Generic Letter 84-04 provides an exception to the guidance in Regulatory Guide 1.45, which requires detection of 1 gpm leakage within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The capability of the RNP leak detection systems to detect 1 gpm leakage is within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, consistent with the conditions of Generic Letter 84-04.

Licensees are required to submit, for NRC review and approval, a fracture mechanics evaluation of specific piping configurations to meet the requirements of GDC 4. A candidate pipe should satisfy the screening criteria of SRP, Section 3.6.3, by demonstrating that it experiences no active degradation. The candidate pipe should be demonstrated by the fracture mechanics analysis to satisfy the safety margins in SRP, Section 3.6.3. Finally, the licensee must demonstrate that the reactor coolant system leakage detection systems have the capability to detect a certain leak rate, with margins, when compared to the leak rate from the leakage flaw size of the candidate pipe.

The implementation of LBB requires a license amendment under 10 CFR 50.90 because one or more of the criteria of 10 CFR 50.59(c)(2) applies to LBB. When the proposed LBB LAR is approved by the NRC, the licensee is required to amend its final safety analysis report to document that the LBB methodology has become a part of the licensing basis for the candidate piping.

The requirements related to the content of the TS are contained in 10 CFR 50.36, which requires that the TS include LCOs. The criteria defined by 10 CFR 50.36(c)(2)(ii) relevant to determining whether capabilities related to reactor coolant pressure boundary (RCPB) leakage detection should be included in the TS LCOs, are as follows:

a) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

b) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The lowest flow rate calculated for the LBB leakage flaws is 10 gpm as stated in WCAP-17776, WCAP-17778 and WCAP-17779. The proposed change maintains the minimum required unidentified leakage detection capability of 1 gpm after applying the margin factor of 10, in accordance with SRP 3.6.3 criteria. The 1 gpm limit assures timely identification of reactor coolant pressure boundary degradation, and the measurement capability is sufficient to ensure reactor coolant system leakage can be detected well in advance of a through wall flaw propagating to a pipe rupture. The adequacy of the current TS is supported by the margins used in the LBB evaluations, i.e., a margin factor of 10 between leakage crack flow rate and leakage detection capability, and a factor of two between leakage crack size and critical crack size.

These margins offset uncertainties associated with leakage detection and prediction.

In summary, the request is consistent with the GDC 4 provision that dynamic effects associated with postulated pipe ruptures may be removed from the design basis if NRC-approved analyses demonstrate an extremely low probability of pipe rupture occurring under design basis conditions. The proposed change maintains consistency with GDC 14 and GDC 30 criteria for maintaining the integrity of the reactor coolant pressure boundary and being able to detect reactor coolant system leakage. The existing TS for leakage detection and leakage limits are consistent with 10 CFR 50.36 and do not require revision to support this request.

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 18 of 21 4.2. Precedents Several other licensees have requested and received approval to use the LBB methodology to eliminate the dynamic effects of pipe rupture for auxiliary piping systems attached to the reactor coolant system main piping including the following:

• Letter from US NRC to Entergy, "Waterford Steam Electric Station, Unit 3 - Issuance of Amendment Re: Approval of Leak-Before-Break of the Pressurizer Surge Line", dated February 28, 2011 (ADAMS Accession Number ML110410119)

• Letter from US NRC to Northern States Power Company, Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendments Re: Request to Exclude the Dynamic Effects Associated with Certain Postulated Pipe Ruptures from the Licensing Basis based upon Application of Leak-Before-Break Methodology", dated October 27, 2011 (ADAMS Accession Number ML112200856)

• Letter from US NRC to Indiana Michigan Power Company, "Donald C. Cook Nuclear Plant, Unit No. 1 - Issuance of Amendment Number 346 Re: Approval of Application of Proprietary Leak-Before-Break Methodology for Reactor Coolant System Small Diameter Piping", dated August 1, 2019 (ADAMS Accession Number ML19170A362)

• Letter from US NRC to PSE&G LLC "Salem Nuclear Generating Station, Unit Nos. 1 and 2 - Issuance of Amendment Numbers 336 and 317 Re: Leak-Before-Break for Accumulator, Residual Heat Removal, Safety Injection, and Pressurizer Surge Lines,"

dated February 23, 2021 (ADAMS Accession No. ML20338A038)

• Letter from US NRC to Dominion Energy Virginia "Surry Power Station, Units 1 and 2 -

Issuance of Amendment Numbers 304 and 304 Re: Leak-Before-Break for Pressurizer Surge, Residual Heat Removal, Safety Injection Accumulator, Reactor Coolant System Bypass and Safety Injection Lines," dated August 20, 2021 (ADAMS Accession Number ML21175A185) 4.3 No Significant Hazards Consideration Duke Energy Progress, LLC (Duke Energy) requests an amendment to the H. B. Robinson Steam Electric Plant, Unit No. 2 (RNP) Facility Operating License. The proposed amendment would change the RNP design and licensing basis as described in the RNP Updated Final Safety Analysis Report (UFSAR) to eliminate the dynamic effects of postulated pipe ruptures in specific portions of systems attached to the Reactor Coolant System (RCS) in accordance with 10 CFR 50, Appendix A, General Design Criterion 4, "Environmental and dynamic effects design bases." This License Amendment Request (LAR) uses Leak-Before-Break (LBB) methodology to demonstrate the risk of pipe rupture is extremely low for portions of the following systems piping connected to the RCS loop piping:

Duke Energy has evaluated the proposed changes using the criteria in 10 CFR 50.92 and determined that the proposed changes do not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change requests plant-specific approval of a previously approved Leak-Before-Break (LBB) evaluation methodology, in accordance with 10 CFR 50, Appendix A, General Design Criterion (GDC) 4. The LBB evaluations demonstrate that the probability of a rupture of the piping in the scope of the request is extremely low under

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 19 of 21 design basis conditions, such that the dynamic effects of postulated pipe ruptures may be removed from the design basis of RNP. The proposed change does not adversely affect accident initiators or precursors. Overall protection system performance will remain within the bounds of the previously performed accident analyses. The design of the protection systems will be unaffected. The Reactor Protection System (RPS) and Emergency Core Cooling System (ECCS) will continue to function in a manner consistent with the plant design basis. All design, material, and construction standards that were applicable prior to the request will remain applicable.

There will be no change to normal plant operating parameters or accident mitigation performance. The proposed amendment will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the RNP Updated Final Safety Analysis Report (UFSAR).

Therefore, these proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed change requests NRC approval of LBB methodology to demonstrate an extremely low probability of pipe rupture. It does not introduce any new accident scenarios, failure mechanisms, or single failures. All systems, structures, and components previously required for the mitigation of an event remain capable of fulfilling their intended design function. The proposed change has no adverse effects on any safety related systems or components and does not challenge the performance or integrity of any safety related system. Further, there are no changes in the method by which any safety-related plant system performs its safety function. This amendment will not affect the normal method of power operation or change any operating parameters.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No The proposed change does not adversely affect the ability of the fuel cladding, reactor coolant pressure boundary, or containment to perform their design basis functions as fission product barriers. The proposed change uses previously accepted analytical methods to demonstrate that the probability of a fluid system rupture is extremely low. It has no effect on the manner in which safety limits or limiting safety system settings are determined and it does not adversely affect any plant systems necessary to assure the accomplishment of protection functions.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based upon the above, Duke Energy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 20 of 21 4.4 Conclusion Based on the considerations discussed herein, Duke Energy concludes that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

The proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9) as follows:

(i) The proposed change involves no significant hazards consideration.

As described in Section 4.3 above, the proposed change involves no significant hazards consideration.

(ii) There are no significant changes in the types or significant increase in the amounts of any effluents that may be released off-site.

The proposed change would change the RNP design and licensing basis as described in the RNP UFSAR to eliminate the dynamic effects of postulated pipe ruptures in specific portions of systems attached to the RCS. The proposed change does not alter the design function or operation of any plant structure, system, or component. The reactor coolant pressure boundary will continue to meet its specific structural and leakage integrity performance criteria. The proposed change does not involve the installation of any new equipment or the modification of any equipment that may affect the types or amounts of effluents that may be released off-site. The proposed change will have no impact on normal plant releases and will not increase the predicted radiological consequences of accidents postulated in the UFSAR. Therefore, there are no significant changes in the types or significant increase in the amounts of any effluents that may be released off-site.

ii) There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed change would change the RNP design and licensing basis as described in the RNP UFSAR to eliminate the dynamic effects of postulated pipe ruptures in specific portions of systems attached to the RCS. The proposed change does not implement plant physical changes or result in plant operation in a configuration outside the plant safety analyses or design basis. Furthermore, reactor coolant pressure boundary will continue to meet specific structural and leakage integrity performance criteria. Therefore, there is no significant increase in individual or cumulative occupational radiation exposure associated with the proposed change.

Based on the above, Duke Energy concludes that, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

RA-22-0290, Attachment 1 H. B. Robinson Steam Electric Plant, Unit No. 2 Page 21 of 21

6.0 REFERENCES

1. WCAP-9558, "Mechanistic Fracture Evaluation of Reactor Coolant Piping Containing a Postulated Circumferential Through-Wall Crack," dated May 1981.

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

3. Generic Letter 84-04, "Safety Evaluation of Westinghouse Topical Report Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops," dated February 1, 1984.

RA-22-0290, Attachment 2 H. B. Robinson Steam Electric Plant, Unit No. 2 ATTACHMENT 2 WCAP-17776-NP, REVISION 1, "TECHNICAL JUSTIFICATION FOR ELIMINATING PRESSURIZER SURGE LINE RUPTURE AS THE STRUCTURAL DESIGN BASIS FOR H. B.

ROBINSON UNIT 2, MARCH 2023 (REDACTED)

Westinghouse Non-Proprietary Class 3 WCAP-17776-NP March 2023 Revision 1 Technical Justification for Eliminating Pressurizer Surge Line Rupture as the Structural Design Basis for H. B. Robinson Unit 2

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17776-NP Revision 1 Technical Justification for Eliminating Pressurizer Surge Line Rupture as the Structural Design Basis for H. B. Robinson Unit 2 March 2023 Author: Nadia B. Petkova

• Operating Plants Piping and Supports Reviewer: Momo Wiratmo*

Operating Plants Piping and Supports Approved: Lynn A. Patterson, Manager*

Reactor Vessel and Containment Vessel Design and Analysis

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

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

© 2023 Westinghouse Electric Company LLC All Rights Reserved

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii RECORD OF REVISIONS Revision Date Revision Description Original Issue (WCAP-17776-NP). This is the non-proprietary class 3 0 August 2013 version of WCAP-17776-P, Revision 0.

This is the non-proprietary class 3 version of WCAP-17776-P, Revision 1 March 2023 1.

WCAP-17776-NP March 2023 Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iv TABLE OF CONTENTS 1.0 Introduction ................................................................................................................................... 1-1 1.1 Background ...................................................................................................................... 1-1 1.2 Scope and Objectives ....................................................................................................... 1-1 1.3 References ........................................................................................................................ 1-2 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 ................................................................................. 2-2 2.4 Summary Evaluation of Surge Line for Potential Degradation During Service .............. 2-3 2.5 References ........................................................................................................................ 2-4 3.0 Pipe Geometry and Loading ......................................................................................................... 3-1 3.1 Calculation of Loads and Stresses ................................................................................... 3-1 3.2 Loads for Leak Rate Evaluation ...................................................................................... 3-1 3.3 Load Combination for Crack Stability Analyses ............................................................. 3-2 3.4 Loading Conditions.......................................................................................................... 3-2 3.5 Summary of Loads ........................................................................................................... 3-4 3.6 Pipe Geometry ................................................................................................................. 3-4 3.7 References ........................................................................................................................ 3-4 4.0 Material Characterization.............................................................................................................. 4-1 4.1 Pressurizer Surge Line Piping, Fittings, and Weld Materials .......................................... 4-1 4.2 Tensile Properties ............................................................................................................. 4-1 4.3 References ........................................................................................................................ 4-1 5.0 Critical Location and Evaluation Criteria ..................................................................................... 5-1 5.1 Critical Locations ............................................................................................................. 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 Global Failure Mechanism............................................................................................... 7-1 7.2 Results of Crack Stability Evaluation .............................................................................. 7-2 7.3 References ........................................................................................................................ 7-2 8.0 Assessment of Fatigue Crack Growth ........................................................................................... 8-1 8.1 Introduction...................................................................................................................... 8-1 8.2 Results.............................................................................................................................. 8-2 WCAP-17776-NP March 2023 Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 v 8.3 References ........................................................................................................................ 8-3 9.0 Assessment of Margins ................................................................................................................. 9-1 10.0 Conclusions ................................................................................................................................. 10-1 Appendix A: Limit Moment...................................................................................................................... A-1 WCAP-17776-NP March 2023 Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vi LIST OF TABLES Table 3-1 Types of Loadings ............................................................................................................ 3-5 Table 3-2 Normal and Faulted Loading Cases for Leak-Before-Break Evaluations ....................... 3-5 Table 3-3 Associated Load Cases for Analyses ............................................................................... 3-6 Table 3-4 Summary of Leak-Before-Break Loads and Stresses at the Three Governing Stressed Weld Locations (Material A 376 TP316, Weld Process GTAW/SMAW Combination)... 3-7 Table 4-1 Measure Tensile Properties for Pressurizer Surge Line Material A376 TP316................ 4-2 Table 4-2 Mechanical Properties for the Pressurizer Surge Line Material at Operating Temperatures

......................................................................................................................................... 4-2 Table 6-1 Leakage Flaw Sizes ......................................................................................................... 6-3 Table 7-1 Summary of Critical Flaw Sizes for the Pressurizer Surge Line ..................................... 7-3 Table 8-1 Pressurizer Surge Line Fatigue Crack Growth Results .................................................... 8-4 Table 9-1 Leakage Flaw Sizes, Critical Flaw Sizes, and Margins for the Pressurizer Surge Line .. 9-2 WCAP-17776-NP March 2023 Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vii LIST OF FIGURES Figure 3-1 H. B. Robinson Unit 2 Pressurizer Surge Line Layout .................................................... 3-9 Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures ........................ 6-4 Figure 6-2 [ ]a,c,e Pressure Ratio as a Function of L/D .................................................. 6-5 Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack .......................................... 6-6 Figure 7-1 [ ]a,c,e Stress Distribution ................................................................................. 7-4 Figure 7-2 Critical Flaw Size Prediction - Node 130 Case D ........................................................... 7-5 Figure 7-3 Critical Flaw Size Prediction - Node 130 Case E ........................................................... 7-6 Figure 7-4 Critical Flaw Size Prediction - Node 130 Case F ............................................................ 7-7 Figure 7-5 Critical Flaw Size Prediction - Node 380 Case D ........................................................... 7-8 Figure 7-6 Critical Flaw Size Prediction - Node 380 Case E ........................................................... 7-9 Figure 7-7 Critical Flaw Size Prediction - Node 380 Case F .......................................................... 7-10 Figure 7-8 Critical Flaw Size Prediction - Node 600 Case D ......................................................... 7-11 Figure 7-9 Critical Flaw Size Prediction - Node 600 Case E ......................................................... 7-12 Figure 7-10 Critical Flaw Size Prediction - Node 600 Case F .......................................................... 7-13 Figure 8-1 Orientation of Stress Cuts for the Fatigue Crack Growth Analysis ................................. 8-5 Figure 8-2 Reference Crack Growth Curves for Stainless Steel in Air Environments ...................... 8-6 Figure A-1 Pipe with a Through-Wall Crack in Bending .................................................................. A-2 WCAP-17776-NP March 2023 Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-1

1.1 BACKGROUND

The current structural design basis for the pressurizer surge line requires postulating non-mechanistic circumferential and longitudinal pipe breaks. This results in additional plant hardware (e.g. pipe whip restraints and jet shields) that would mitigate the dynamic consequences of the pipe breaks. It is therefore highly desirable to be realistic in the postulation of pipe breaks for the surge line. Presented in this report are the descriptions of a mechanistic pipe break evaluation method and the analytical results that can be used for establishing that a circumferential type break will not occur within the pressurizer surge line.

The evaluations consider that circumferentially oriented flaws cover longitudinal cases. The pressurizer surge line is known to be subjected to thermal stratification and the effects of thermal stratification for the H. B. Robinson presurizer surge line have been evaluated and documented in WCAP-12962 (Reference 1-1). The results of the stratification evaluation as described in WCAP-12962 have been used in the Leak-Before-Break evaluation presented in this report.

1.2 SCOPE AND OBJECTIVES The purpose of this investigation is to demonstrate Leak-Before-Break (LBB) for the H. B. Robinson Unit 2 pressurizer surge line. The scope of this work covers the entire pressurizer surge line from the primary loop nozzle junction to the pressurizer nozzle junction. A schematic drawing of the piping system is shown in Section 3.0. The recommendations and criteria proposed in SRP 3.6.3 (References 1-2 and 1-3) are used in this evaluation. The criteria and the resulting steps of the evaluation procedure can be briefly summarized as follows:

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

2. Identify the materials and the material properties.

3. Postulate a through-wall flaw at the governing location. 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. Demonstrate that there is a margin of 10 between the calculated leak rate and the leak detection capability.

4. Using maximum faulted loads in the stability analysis, demonstrate that there is a margin of 2 between the leakage size flaw and the critical size flaw.

5. 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.

6. For the materials types used in the plant, provide representative material properties.

7. Demonstrate margin on applied load.

Introduction March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-2

8. Perform an assessment of fatigue crack growth. Show that a through-wall crack will not result.

This report provides a fracture mechanics analysis that demonstrates the pressurizer surge line integrity for H. B. Robinson Unit 2 consistent with the NRCs position for exemption from consideration of postulated pipe rupture dynamic effects (Reference 1-4).

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.

Note that there are several locations in this report where proprietary information has been identified and bracketed. For each of the bracketed locations, the reason for the proprietary classification is given using a standardized system. The proprietary brackets are labeled with three different letters, to provide this information, and the explanation for each letter is given below:

a. The information reveals the distinguishing aspects of a process or component, structure, tool, method, etc., and the prevention of its use by Westinghouses competitors, without license from Westinghouse, gives Westinghouse a competitive economic advantage.

b. The information, if used by a competitor, would reduce the competitors expenditure of resources or improve the competitors advantage in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.

c. The information reveals aspects of past, present, or future Westinghouse or customer-funded development plans and programs of potential commercial value to Westinghouse.

The proprietary information in the brackets which has been deleted in this version of this report are provided in the proprietary class 2 document (WCAP-17776-P, Revision 1).

1.3 REFERENCES

1-1 WCAP-12962, Revision 0, Structural Evaluation of the H. B. Robinson Unit 2 and Shearon Harris Pressurizer Surge lines, Considering the Effects of Thermal Stratification, September 1991 including WCAP-12962 Supplement 1, Revision 0, October 1995 (Westinghouse Proprietary).

1-2 Standard Review Plan: Public Comments Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 52, No. 167/Friday August 28, 1987/Notices, pp. 32626-32633.

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

1-4 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 Register/Vol. 52, No. 207/Tuesday, October 27, 1987/Rules and Regulations, pp. 41288 41295.

Introduction March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

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 pressurizer surge lines 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 (PWRs). 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.

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 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.

Operation and Stability of the Reactor Coolant System March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 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 1979, several instances of cracking in PWR feedwater piping 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.

Primary Water Stress Corrosion Cracking (PWSCC) occurred in V. C. Summer reactor vessel hot leg nozzle, Alloy 82/182 weld. It should be noted that this susceptible material is not found at the H. B.

Robinson Unit 2 pressurizer surge line.

2.2 WATER HAMMER Overall, there is a low potential for water hammer in the RCS and the connecting surge line since they are designed and operated to preclude the voiding condition in the normally filled surge line. The RCS and connecting surge line including piping and components, are designed for normal, upset, emergency, and faulted condition transients. The design requirements are conservative relative to both the number of transients and their severity. Pressurizer safety and relief valve actuation and the associated hydraulic transients following valve opening are considered in the system design. Only relatively slow transients are applicable to the surge line and there is 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 by the control rod positions. Pressure is also controlled within a narrow range for steady-state conditions by the pressurizer heaters and the pressurizer spray. 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 and the connecting auxiliary lines.

Preoperational testing and operating experience have verified the Westinghouse approach. The operating transients of the RCS primary piping and the connected surge line are such that no significant water hammer can occur.

2.3 LOW CYCLE AND HIGH CYCLE FATIGUE Fatigue considerations are accounted for in the surge line piping through the fatigue usage factor evaluation for the stratification analyses (Reference 2-3) to show compliance with the rules of Section III Operation and Stability of the Reactor Coolant System March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 of the ASME Code. A further assessment of the low cycle fatigue loading is discussed in Section 8.0 as part of this study in the form of a fatigue crack growth assessment.

Pump vibrations during operation would result in high cycle fatigue loads in the piping system. During operation, an alarm signals the exceeding of the RC pump vibration limits. Field measurements have been made on the reactor coolant loop piping in a number of plants during hot functional testing. Stresses in the elbow below the RC pump have been found to be very small, between 2 and 3 ksi at the highest.

Field measurements on a typical PWR plant indicate vibration amplitudes less than 1 ksi. When translated to the connecting surge line, these stresses would be even lower, well below the fatigue endurance limit for the surge line material and would result in an applied stress intensity factor below the threshold for fatigue crack growth. H. B. Robinson Unit 2 configurations are similar and the results are expected to be the similar.

2.4

SUMMARY

EVALUATION OF SURGE LINE FOR POTENTIAL DEGRADATION DURING SERVICE There has never been any service cracking or wall thinning identified in the pressurizer surge line of Westinghouse PWR design. The design, construction, inspection, and operation of the pressurizer surge line piping mitigate sources of such degradation.

There is no known mechanism for water hammer in the pressurizer/surge system. The pressurizer safety and relief piping system that is connected to the top of the pressurizer could have loading from water hammer events. However, these loads are effectively mitigated by the pressurizer and have a negligible effect on the surge line.

Wall thinning by erosion and erosion-corrosion effects should not occur in the surge line due to the low velocity, typically less than 1.0 ft/sec and the material, austenitic stainless steel, which is highly resistant to these degradation mechanisms. Per NUREG-0691 (Reference 2-1), a study on pipe cracking in PWR piping reported only two incidents of wall thinning in stainless steel pipe and these were not in the surge line. The cause of wall thinning is related to the high water velocity and is therefore clearly not a mechanism that would affect the surge line.

It is well known that the pressurizer surge line is subjected to thermal stratification and the effects of stratification are particularly significant during certain modes of heatup and cooldown operation. The effects of stratification have been evaluated for the H. B. Robinson Unit 2 surge line and the loads, accounting for the stratification effects, have been derived in WCAP-12962 (Reference 2-3). These loads are used in the Leak-Before-Break evaluation described in this report.

The H. B. Robinson Unit 2 surge line piping system is fabricated from forged products (see Section 4) which are not susceptible to toughness degradation due to thermal aging.

Finally, the maximum operating temperature of the pressurizer surge line piping, which is about 650°F, is well below the temperature that would cause any creep damage in stainless steel piping. Cleavage type failures are not a concern for the operating temperatures and the material used in the stainless steel piping of the pressurizer surge line.

Operation and Stability of the Reactor Coolant System March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-4

2.5 REFERENCES

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

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

2-3 WCAP-12962, Revision 0, Structural Evaluation of the H. B. Robinson Unit 2 and Shearon Harris Pressurizer Surge lines, Considering the Effects of Thermal Stratification, September 1991 including WCAP-12962 Supplement 1, Revision 0, October 1995 (Westinghouse Proprietary).

Operation and Stability of the Reactor Coolant System March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 3.0 PIPE GEOMETRY AND LOADING 3.1 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 Z

where,

= stress F = axial load M = moment A = pipe cross-sectional area Z = section modulus The moments for the desired loading combinations are calculated by the following equation:

M = M2x + M2y + M2z (3-2)

where, Mx = X component of moment, Torsion My = Y component of bending moment Mz = Z component of bending moment The axial load and moments for leak rate predictions and crack stability analyses are computed by the methods to be explained in Sections 3.2 and 3.3.

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

F = FDW + FTH + FP (3-3)

MX = (MX)DW + (MX)TH (3-4)

MY = (MY)DW + (MY)TH (3-5)

MZ = (MZ)DW + (MZ)TH (3-6)

Pipe Geometry and Loading March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-2 The subscripts of the above equations represent the following loading cases:

DW = deadweight TH = normal thermal expansion or thermal stratification 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).

The loads based on this method of combination are provided in Table 3-4 at the weld locations identified in Figure 3-1.

3.3 LOAD COMBINATION FOR CRACK STABILITY ANALYSES In accordance with Standard Review Plan 3.6.3 (References 3-1 and 3-2), the absolute sum of loading components can be applied which results in higher magnitude of combined loads. If crack stability is demonstrated using these loads, the LBB margin on loads can be reduced from 2 to 1.0. The absolute summation of loads is shown in the following equations:

F = FDW + FTH + FP + FSSEINERTIA + FSSEAM (3-7)

MX = (MX)DW + (MX)TH + (MX)SSEINERTIA+ (MX)SSESAM (3-8)

MY = (MY)DW + (MY)TH + (MY)SSEINERTIA+ (MY)SSEAM (3-9)

MZ = (MZ)DW + (MZ)TH + (MZ)SSEINERTIA+ (MZ)SSEAM (3-10) where subscript SSEINERTIA refers to safe shutdown earthquake inertia and SSEAM is safe shutdown earthquake anchor motion.

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 the three governing the weld locations (see Figure 3-1) are given in Table 3-4.

3.4 LOADING CONDITIONS Because thermal stratification can cause large stresses during heatup and cooldown, a review of the stratification stresses was performed to identify the upper bound loadings. The identified types of loading are given in Table 3-1.

Seven loading cases were identified and are shown in Table 3-2. Cases A, B and C are the normal operating load cases and Cases D, E, F and G are the faulted load cases.

The cases postulated for Leak-Before-Break evaluation are summarized in Table 3-3. The cases of primary interest are the postulation of a detectable leak at normal 100% power [

Pipe Geometry and Loading March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-3 a,c,e

]

Case Combination [

a,c,e

]

The case combination [

a,c,e

]

The realistic cases A/D or B/E or [

a,c,e

]

[

a,c,e

]

The logic for this system T of [ ] a,c,e is based on the following:

Actual practice, based on experience from other plants with this type of situation, indicates that the plant operators complete the cool down as quickly as possible once a leak in the primary system is detected.

Technical Specifications may require cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, but actual practice is that the plant operators depressurize the system as soon as possible once a primary system leak is detected. Therefore, the hot leg is generally on the warmer side of the limits (~200°F) when the pressurizer bubble is quenched. Once the bubble is quenched, the pressurizer is cooled down fairly quickly reducing the T in the system.

Pipe Geometry and Loading March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-4 3.5

SUMMARY

OF LOADS The combined loads were evaluated at the various weld locations. Normal loads were determined using the algebraic sum method whereas faulted loads were combined using the absolute sum method. Table 3-4 shows loads and stresses at the three governing stressed weld locations for A376 TP316 material with GTAW/SMAW combination. Loads and stresses for Case C and Case G in Table 3-4 are shown for information only and they are not used in the LBB analysis.

3.6 PIPE GEOMETRY The H. B. Robinson pressurizer surge line is 12-inch schedule 140; pipe outer diameter is 12.75 inch and a minimum pipe wall thickness, based on the maximum allowed counterbore at a butt weld (Reference 3-3), at the weld counterbore of 1.005 inches was used in the analysis.

3.7 REFERENCES

3-1 Standard Review Plan: Public Comments Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 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.

3-3 American National Standards, Butt Welding Ends, ANSI B16.25-1979.

Pipe Geometry and Loading March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-5 Table 3-1 Types of Loadings Pressure (P)

Dead Weight (DW)

Normal Operating Thermal Expansion (TH)

Safe Shutdown Earthquake including Seismic Anchor Motion (SSE) a,c,e

[ ]

a,c,e

[ ]

a,c,e

[ ]

Table 3-2 Normal and Faulted Loading Cases for Leak-Before-Break Evaluations CASE A This is the normal operating case at 653°F consisting of the algebraic sum of the loading components due to P, DW and TH.

CASE B [

a,c,e

]

CASE C1 [

a,c,e

]

CASE D This is the faulted operating case at 653°F consisting of the absolute sum (every component load is taken as positive) of P, DW, TH and SSE.

CASE E [

a,c,e

]

CASE F This is a forced cooldown case [

a,c,e

] with stratification [

a,c,e

]

CASE G1 [ ]

a,c,e 1

Case C and Case G are shown for information only.

Pipe Geometry and Loading March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-6 Table 3-3 Associated Load Cases for Analyses A/D This is the standard Leak-Before-Break evaluation.

A/F This depicts a postulated forced cooldown event resulting from experiencing a detectable leak [

a,c,e

]

B/E [

a,c,e

]

B/F This depicts a postulated forced cooldown event resulting from experiencing a detectable leak [

a,c,e

]

Pipe Geometry and Loading March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-7 Table 3-4 Summary of Leak-Before-Break Loads and Stresses at the Three Governing Stressed Weld Locations (Material A 376 TP316, Weld Process GTAW/SMAW Combination)

Nodea Case Axial Axial Stress Moment Moment Total stress Forceb (ksi) (in-kips) Stress (ksi) (ksi)

(kips) 130 A 210.562 5.678 436.785 4.32 10.00 130 B 210.562 5.678 405.433 4.01 9.69 130 Cc 44.846 1.209 3285.262 32.52 33.73 130 D 214.142 5.775 865.628 8.57 14.34 130 E 214.142 5.775 1046.293 10.36 16.13 130 F 45.076 1.216 2834.232 28.05 29.27 130 Gc 48.426 1.306 3893.478 38.54 39.84 380 A 210.562 5.678 751.769 7.44 13.12 380 B 210.562 5.678 1042.106 10.31 15.99 380 Cc 44.846 1.209 2411.412 23.87 25.08 380 D 212.011 5.717 805.097 7.97 13.69 380 E 212.011 5.717 1111.973 11.01 16.72 380 F 45.076 1.216 1920.339 19.01 20.22 380 Gc 46.295 1.248 2522.435 24.97 26.21 600 A 203.837 5.497 1182.233 11.70 17.20 600 B 203.737 5.494 1083.418 10.72 16.22 600 Cc 39.521 1.066 1424.814 14.10 15.17 Pipe Geometry and Loading March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-8 Table 3-4 Summary of Leak-Before-Break Loads and Stresses at the Three Governing Stressed Weld Locations (Material A 376 TP316, Weld Process GTAW/SMAW Combination)

Nodea Case Axial Axial Stress Moment Moment Total stress Forceb (ksi) (in-kips) Stress (ksi) (ksi)

(kips) 600 D 206.959 5.581 1513.316 14.98 20.56 600 E 206.859 5.578 1377.651 13.64 19.21 600 F 40.701 1.098 1158.077 11.46 12.56 600 Gc 42.643 1.150 1802.016 17.84 18.99 Notes:

a. See Figure 3-1

b. Included Pressure

c. For information only Pipe Geometry and Loading March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-9 Figure 3-1 H. B. Robinson Unit 2 Pressurizer Surge Line Layout Pipe Geometry and Loading March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-1 4.0 MATERIAL CHARACTERIZATION 4.1 PRESSURIZER SURGE LINE PIPING, FITTINGS, AND WELD MATERIALS The pipe material of the pressurizer surge line for H. B. Robinson Unit 2 is A376 TP316. This is a wrought product of the type used for the piping of several PWR Plants. The surge line is connected to the primary loop at one end and at the other end to the pressurizer nozzle. The surge line does not include any cast pipes or cast fittings. The welding processes used are Gas Tungsten Arc Weld (GTAW)/Shielded Metal Arc Weld (SMAW) combination. Figure 3-1 shows the schematic layout of the surge line and identifies the weld locations by node points.

In the following sections the tensile properties of the materials are presented for use in the Leak-Before-Break analyses.

4.2 TENSILE PROPERTIES The Certified Materials Test Reports (CMTRs) for the H. B. Robinson Unit 2 pressurizer surge line were used to establish the tensile properties for the leak-before-break analyses. The tensile properties for the pipe material are provided in Table 4-1.

For H. B. Robinson Unit 2, specific data was used as a basis for determining tensile properties. The room temperature mechanical properties of the surge line material were obtained from the Certified Materials Test Reports (CMTRs) and are given in Table 4-1. The representative minimum and average tensile properties were established. The material properties at temperatures (653°F, 605°F, 455°F and 205°F) are required for the leak rate and stability analyses. The minimum and average tensile properties were calculated by using the ratio of ASME Boiler and Pressure Vessel Code Section II of the 2007 Edition with the 2008 Addenda (Reference 4-1) properties at the temperatures of interest stated above. Table 4-2 shows the tensile properties at various temperatures. The moduli of elasticity values were established at various temperatures from the ASME Code Section III (see Table 4-2). In the Leak-Before-Break evaluation, the representative minimum yield strength and minimum ultimate strength at temperature were used for the flaw stability evaluations and the representative average yield strength was used for the leak rate predictions.

The average and lower bound yield strengths and ultimate strengths for the pipe material are tabulated in Table 4-2. The ASME Code modulus of elasticity values are also given, and Poisson's ratio was taken as 0.3.

4.3 REFERENCES

4-1 ASME Boiler and Pressure Vessel Code, 2007 Edition with the 2008 Addenda,Section II, Part D

- Properties (Customary) Materials.

Material Characterization March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-2 Table 4-1 Measure Tensile Properties for Pressurizer Surge Line Material A376 TP316 Heat Number Yield Strength (psi) Ultimate Strength (psi)

Room Temp. Room Temp.

8935(SER 2161) 39500 81600 8935(SER 2161) 40760 81600 8935(SER 2162) 37700 79400 8935(SER 2162) 40160 88200 8935(SER 2163) 35900 81000 8935(SER 2163) 38500 81000 8935(SER 2164) 41500 82600 8935(SER 2164) 39500 81800 Table 4-2 Mechanical Properties for the Pressurizer Surge Line Material at Operating Temperatures Lower Bound Average Yield Yield Stress Ultimate Material Temperature (°F) Strength (psi) (psi) Strength (psi)

A376 TP316 653 24144 22117 76012 A376 TP316 605 24637 22569 76012 A376 TP316 455 26950 24687 76060 A376 TP316 205 33671 30844 79289 Modulus of E = 25.035 x 106 psi at 653°F ; E = 25.275 x 106 psi at 605°F; E = 26.125 x 106 psi Elasticity: at 455°F; E = 27.475 x 106 psi at 205°F Poissons ratio: 0.3 Material Characterization March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

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 critical 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 the H. B. Robinson Unit 2 pressurizer surge line piping.

Critical Locations Node 130 (hot leg nozzle to pipe weld location), Node 380 (intermediate elbow weld location) and Node 600 (pressurizer nozzle location, reducer to pipe weld location) are the governing weld locations identified for the LBB analysis. Node 130 is determined to be the critical location at the HL surge nozzle due to the limiting cross-sectional properties of the pipe, rather than the thicker reinforcement area at the nozzle to reactor coolant loop branch weld.

Critical Location and Evaluation Criteria March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

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, (L/DH) is greater than [

]a,c,e 6.3 CALCULATION METHOD The basic method used in the leak rate calculations is the method developed by [

]a,c,e 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 pressurizer surge line enthalpy condition and an assumed flow. Once Pc was found for a given mass flow, the [

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

]a,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 Pf = [ ] a,c,e (6-1) where the friction factor f is determined using the [ ]a,c,e The crack relative roughness, , 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.

Leak Rate Predictions March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-2 That is, for the pressurizer surge line:

Absolute Pressure - 14.7 = [ ]a,c,e (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 pressurizer surge line 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 3-4 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 4-2) 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 H. B. Robinson Unit 2. The flaw sizes so determined are called leakage flaw sizes.

The H. B. Robinson Unit 2 RCS pressure boundary leak detection system meets the intent of Regulatory Guide 1.45. 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 [

] a,c,e 6-2 M. M, El-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 II-1, NUREG/CR-3464, September 1983.

Leak Rate Predictions March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3 Table 6-1 Leakage Flaw Sizes Node Point Load Case Temperature Leakage Flaw Size (in.)

(°F) (for 10 gpm leakage) 130 A 653 4.91 a,c,e 130 B [ ] 4.84 380 A 653 4.08 a,c,e 380 B [ ] 3.48 600 A 653 3.26 a,c,e 600 B [ ] 3.44 Leak Rate Predictions March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-4 Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures Leak Rate Predictions March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-5 Figure 6-2 [ ]a,c,e Pressure Ratio as a Function of L/D Leak Rate Predictions March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-6 Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack Leak Rate Predictions March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-1 7.0 FRACTURE MECHANICS EVALUATION 7.1 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:

[ ]a,c,e where:

[

]a,c,e f = 0.5 (y + u) = flow stress, psi

[

]a,c,e 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 March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-2 7.2 RESULTS OF CRACK STABILITY EVALUATION A stability analysis based on limit load was performed for governing locations as described in Section 7.1.

The weld process types, at these locations Node 130, Node 380 and Node 600 are used as GTAW and SMAW combination. The Z correction factor for SMAW (References 7-2 and 7-3) are as follows:

Z = 1.15 [1.0 + 0.013 (OD-4)] for SMAW where OD is the outer diameter of the pipe in inches.

The Z-factor for the GTAW weld is 1.0. The Z-factor for the SMAW was calculated for the governing locations, using the outer diameter of 12.75 inches. The applied faulted loads (Table 3-4) were increased by the Z factor and plots of limit load versus crack length were generated as shown in Figures 7-2 to 7-10.

Lower bound material properties were used from Table 4-2. Table 7-1 summarizes the results of the stability analyses based on limit load. The leakage flaw sizes are also presented in the same table.

7.3 REFERENCES

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

7-2 Standard Review Plan; Public Comment Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 52, No. 167/Friday, August 28, 1987/Notices, pp. 32626-32633.

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

Fracture Mechanics Evaluation March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-3 Table 7-1 Summary of Critical Flaw Sizes for the Pressurizer Surge Line Node Point Load Case Temperature Critical

(°F) Flaw Size (in) 130 D 653 14.98 130 E 605 14.27 130 F 205 10.85 380 D 653 15.28 380 E 653 14.01 380 F 455 13.58 600 D 653 12.54 600 E 653 13.06 600 F 455 17.32 Fracture Mechanics Evaluation March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-4 Figure 7-1 [ ]a,c,e Stress Distribution Fracture Mechanics Evaluation March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-5 a,c,e OD = 12.75 in. y-min = 22.117 ksi F = 214.142 kips t = 1.005 in. u-min = 76.012 ksi M = 865.628 in-kips A376 TP316 with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-2 Critical Flaw Size Prediction - Node 130 Case D Fracture Mechanics Evaluation March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-6 a,c,e OD = 12.75 in. y-min = 22.569 ksi F = 214.142 kips t = 1.005 in. u-min = 76.012 ksi M = 1046.293 in-kips A376 TP316 with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-3 Critical Flaw Size Prediction - Node 130 Case E Fracture Mechanics Evaluation March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-7 a,c,e OD = 12.75 in. y-min = 30.844 ksi F = 45.076 kips t = 1.005 in. u-min = 79.289 ksi M = 2834.232 in-kips A376 TP316 with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-4 Critical Flaw Size Prediction - Node 130 Case F Fracture Mechanics Evaluation March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-8 a,c,e OD = 12.75 in. y-min = 22.117 ksi F = 212.011 kips t = 1.005 in. u-min = 76.012 ksi M = 805.097 in-kips A376 TP316 with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-5 Critical Flaw Size Prediction - Node 380 Case D Fracture Mechanics Evaluation March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-9 a,c,e OD = 12.75 in. y-min = 22.117 ksi F = 212.011 kips t = 1.005 in. u-min = 76.012 ksi M = 1111.973 in-kips A376 TP316 with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-6 Critical Flaw Size Prediction - Node 380 Case E Fracture Mechanics Evaluation March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-10 a,c,e OD = 12.75 in. y-min = 24.687 ksi F = 45.076 kips t = 1.005 in. u-min = 76.060 ksi M = 1920.339 in-kips A376 TP316 with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-7 Critical Flaw Size Prediction - Node 380 Case F Fracture Mechanics Evaluation March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-11 a,c,e OD = 12.75 in. y-min = 22.117 ksi F = 206.959 kips t = 1.005 in. u-min = 76.012 ksi M = 1513.316 in-kips A376 TP316 with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-8 Critical Flaw Size Prediction - Node 600 Case D Fracture Mechanics Evaluation March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-12 a,c,e OD = 12.75 in. y-min = 22.117 ksi F = 206.859 kips t = 1.005 in. u-min = 76.012 ksi M = 1377.651 in-kips A376 TP316 with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-9 Critical Flaw Size Prediction - Node 600 Case E Fracture Mechanics Evaluation March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-13 a,c,e OD = 12.75 in. y-min = 24.687 ksi F = 40.701 kips t = 1.005 in. u-min = 76.060 ksi M = 1158.077 in-kips A376 TP316 with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-10 Critical Flaw Size Prediction - Node 600 Case F Fracture Mechanics Evaluation March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-1 8.0 ASSESSMENT OF FATIGUE CRACK GROWTH

8.1 INTRODUCTION

The fatigue crack growth (FCG) analysis is not a requirement for the LBB analysis (see References 8-1 and 8-2) since the LBB analysis is based on the postulation of through-wall flaw, whereas the FCG analysis is performed based on the surface flaw. However, a fatigue crack growth (FCG) assessment of the H. B. Robinson Unit 2 pressurizer surge line was performed. The fatigue crack growth (FCG) of the H. B. Robinson Unit 2 pressurizer surge line was determined by comparison with a fatigue crack growth analysis of a similar pressurizer surge line. The details of the fatigue crack growth analysis are presented below. By comparing the parameters critical to the fatigue crack growth analysis between H. B. Robinson Unit 2 and the similar pressurizer surge line analysis, it was concluded that the similar analysis would adequately cover the fatigue crack growth assessment of the H. B. Robinson Unit 2 pressurizer surge line.

Due to similarities in Westinghouse PWR designs, it was possible to perform a representative fatigue crack growth assessment which would be applicable to H. B. Robinson Unit 2.

The methodology consists of first obtaining the local and structural transient stress analyses results and then superimposing the local and structural transient stresses. The design transients and cycles used in the FCG analyses were the similar ones used in Reference 8-3. An initial flaw size was postulated and the calculation of crack growth for the design plant life using the austenitic stainless steel crack growth law was performed. This fatigue crack growth analysis was performed at the hot leg nozzle location. At this location five through wall stress cuts were analyzed and their orientations are shown in Figure 8-1.

An extensive study was performed by the Materials Property Council Working Group on Reference Fatigue Crack Growth concerning the crack growth behavior of the austenitic stainless steels in an air environment, published in Reference 8-4. A reference fatigue crack growth curve for stainless steels in an air environment, is from Appendix C of the ASME Section XI Code, 2007 Edition (Reference 8-5). This curve is shown in Figure 8-2.

A compilation of data for austenitic stainless steels in a PWR water environment was made by Bamford (Reference 8-6), and it was found that the effect of the environment on the crack growth rate was small.

For this reason it was conservatively estimated that the environmental factor should be set at [ ]a,c,e in the crack growth rate equation from Reference 8-4. Based on these works (References 8-4 and 8-6) the stainless steel fatigue crack growth law used in the analyses is:

[

Assessment of Fatigue Crack Growth March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-2

] a,c,e 8.2 RESULTS Fatigue crack growth analyses were carried out along five stress cuts (Figure 8-1). The analyses were completed for postulated initial flaws oriented circumferentially. The flaws were assumed to be semi-elliptical with an aspect ratio of six to one. The initial flaw sizes were assumed to be 10% of the nominal wall thickness. The results of the fatigue crack growth analyses are presented in Table 8-1. For an initial flaw size of 0.14 inch, the result projects that the maximum final flaw size after 40 /60 years is about 14.8% of the nominal wall thickness. Therefore, flaw growth through the wall is not expected to occur during the 40/60 year design life of the plant and it is concluded that fatigue crack growth should not be a concern for the pressurizer surge line. Transients and cycles for the H. B. Robinson Unit 2 plant for a 40-year transient set will remain bounding for 60 years (Reference 8-7), the FCG results shown in Table 8-1 is also applicable for the 60 years.

Assessment of Fatigue Crack Growth March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-3

8.3 REFERENCES

8-1 Standard Review Plan; Public Comment Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 52, No. 167/Friday, August 28, 1987/Notices, pp. 32626-32633.

8-2 NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.

8-3 WCAP-12962, Revision 0, Structural Evaluation of the H. B. Robinson Unit 2 and Shearon Harris Pressurizer Surge lines, Considering the Effects of Thermal Stratification, September 1991 including WCAP-12962 Supplement 1, Revision 0, October 1995 (Westinghouse Proprietary).

8-4 James, L. A. and Jones, D. P., Fatigue Crack Growth Correlations for Austenitic Stainless Steel in Air, in Predictive Capabilities in Environmentally Assisted Cracking, ASME publication PVP-99, December 1985.

8-5 ASME Boiler and Pressure Vessel Code Section XI, 2007 Edition with the 2008 Addenda, Rules for Inservice Inspection of Nuclear Power Plant Components.

8-6 Bamford, W. H., Fatigue Crack Growth of Stainless Steel Reactor Coolant Piping in a Pressurized Water Reactor Environment, ASME Trans. Journal of Pressure Vessel Technology February 1979.

8-7 NUREG-1785, Safety Evaluation Report Related to the License Renewal of H. B. Robinson Steam Electric Plant, Unit 2.

Assessment of Fatigue Crack Growth March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-4 Table 8-1 Pressurizer Surge Line Fatigue Crack Growth Results Orientation Crack Parameters a,c,e (Figure 8-1) Postulated Initial Flaw (% Final Flaw Size Final Flaw Initial Flaw size of wall (in) (% of wall (in) thickness) thickness) 40/60 Years*

Note:

• Transients and cycles for the H. B. Robinson Unit 2 plant for a 40-year transient set will remain bounding for 60 years, the FCG results shown in Table 8-1 are also applicable for the 60 years.

Assessment of Fatigue Crack Growth March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-5 Figure 8-1 Orientation of Stress Cuts for the Fatigue Crack Growth Analysis Assessment of Fatigue Crack Growth March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-6 Figure 8-2 Reference Crack Growth Curves for Stainless Steel in Air Environments Assessment of Fatigue Crack Growth March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

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 evaluations of Section 7.2 are used in performing the assessment of margins. Margins are shown in Table 9-1. All 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.

Assessment of Margins March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-2 Table 9-1 Leakage Flaw Sizes, Critical Flaw Sizes, and Margins for the Pressurizer Surge Line Node Load Case Critical Flaw Size Leakage Flaw Size (in) (in) Margin 130 A/D 14.98 4.91 3.0 130 A/F 10.84 4.91 2.2 130 B/E 14.27 4.84 2.9 130 B/F 10.84 4.84 2.2 380 A/D 15.28 4.08 3.7 380 A/F 13.58 4.08 3.3 380 B/E 14.01 3.48 4.0 380 B/F 13.58 3.48 3.9 600 A/D 12.54 3.26 3.8 600 A/F 17.32 3.26 5.3 600 B/E 13.06 3.44 3.8 600 B/F 17.32 3.44 5.0 Assessment of Margins March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 10-1

10.0 CONCLUSION

S This report justifies the elimination of pressurizer surge line breaks from the structural design basis for the H. B. Robinson Unit 2 license renewal period 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.

b. Water hammer should not occur in the pressurizer surge line piping because of system design, testing, and operational considerations.

c. The effects of low and high cycle fatigue on the integrity of the pressurizer surge line are negligible.

d. Ample margin exists between the leak rate of small stable flaws and the capability of the H. B. Robinson Unit 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.

g. Fatigue crack growth results using the 40-year design transients and cycles (shown to be applicable for 60 years) show that there will be insignificant growth through the wall for the license renewal period (60-year plant life).

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 loading, pipe geometry and pipe material properties considerations, enveloping critical (governing) locations were determined at which leak-before-break crack stability evaluations were made.

Through-wall flaw sizes were postulated 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 pressurizer surge line piping. Therefore, the Leak-Before-Break conditions and margins are satisfied for the H. B. Robinson Unit 2 pressurizer surge line piping. It is demonstrated that the dynamic effects of the pipe rupture resulting from postulated breaks in the pressurizer surge line piping need not be considered in the structural design basis of H. B. Robinson Unit 2 for the license renewal period (60-year plant life).

Conclusions March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

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

[

] a,c,e Appendix A: Limit Moment March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-2 Figure A-1 Pipe with a Through-Wall Crack in Bending Appendix A: Limit Moment March 2023 WCAP-17776-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:06:10 AM. (This statement was added by the PRIME system upon its validation)

RA-22-0290, Attachment 3 H. B. Robinson Steam Electric Plant, Unit No. 2 ATTACHMENT 3 WCAP-17778-NP, REVISION 1, "TECHNICAL JUSTIFICATION FOR ELIMINATING RESIDUAL HEAT REMOVAL (RHR) LINE RUPTURE AS THE STRUCTURAL DESIGN BASIS FOR H. B.

ROBINSON UNIT 2, MARCH 2023 (REDACTED)

Westinghouse Non-Proprietary Class 3 WCAP-17778-NP March 2023 Revision 1 Technical Justification for Eliminating Residual Heat Removal (RHR) Line Rupture as the Structural Design Basis for H. B. Robinson Unit 2

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17778-NP Revision 1 Technical Justification for Eliminating Residual Heat Removal (RHR) Line Rupture as the Structural Design Basis for H. B. Robinson Unit 2 March 2023 Author: Nadia B. Petkova

• Operating Plants Piping and Supports Reviewer: Momo Wiratmo*

Operating Plants Piping and Supports Approved: Lynn A. Patterson, Manager*

Reactor Vessel and Containment Vessel Design and Analysis

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

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

© 2023 Westinghouse Electric Company LLC All Rights Reserved

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii RECORD OF REVISIONS Revision Date Revision Description Original Issue (WCAP-17778-NP). This is the non-proprietary class 3 0 August 2013 version of WCAP-17778-P, Revision 0.

This is the non-proprietary class 3 version of WCAP-17778-P, Revision 1 March 2023 1.

WCAP-17778-NP March 2023 Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iv TABLE OF CONTENTS

1.0 INTRODUCTION

........................................................................................................................ 1-1 1.1 PURPOSE ........................................................................................................................ 1-1 1.2 SCOPE AND OBJECTIVES ........................................................................................... 1-1

1.3 REFERENCES

................................................................................................................ 1-2 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............................................................... 2-2 2.4 OTHER POSSIBLE DEGRADATION DURING SERVICE OF THE RHR LINES ..... 2-3

2.5 REFERENCES

................................................................................................................ 2-3 3.0 PIPE GEOMETRY AND LOADING ........................................................................................... 3-1 3.1 CALCULATION OF LOADS AND STRESSES ............................................................ 3-1 3.2 LOADS FOR LEAK RATE EVALUATION ................................................................... 3-1 3.3 LOAD COMBINATION FOR CRACK STABILITY ANALYSES................................ 3-2 3.4

SUMMARY

OF LOADS AND GEOMETRY FOR THE RHR LINES .......................... 3-2 3.5 GOVERNING LOCATIONS FOR THE RHR LINES.................................................... 3-3

3.6 REFERENCES

................................................................................................................ 3-3 4.0 MATERIAL CHARACTERIZATION.......................................................................................... 4-1 4.1 RHR LINE PIPE, FITTINGS AND WELD MATERIALS ............................................. 4-1 4.2 TENSILE PROPERTIES ................................................................................................. 4-1

4.3 REFERENCES

................................................................................................................ 4-1 5.0 CRITICAL LOCATION AND EVALUATION CRITERIA......................................................... 5-1 5.1 CRITICAL LOCATIONS ................................................................................................ 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

................................................................................................................ 6-3 7.0 FRACTURE MECHANICS EVALUATION ............................................................................... 7-1 7.1 GLOBAL FAILURE MECHANISM .............................................................................. 7-1 7.2 RESULTS OF CRACK STABILITY EVALUATION..................................................... 7-2

7.3 REFERENCES

................................................................................................................ 7-3 8.0 ASSESSMENT OF FATIGUE CRACK GROWTH .................................................................... 8-1

8.1 INTRODUCTION

........................................................................................................... 8-1 8.2 CRITICAL LOCATION FOR FATIGUE CRACK GROWTH ANALYSIS ................... 8-1 8.3 DESIGN TRANSIENTS ................................................................................................. 8-1 WCAP-17778-NP March 2023 Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 v 8.4 STRESS ANALYSIS ....................................................................................................... 8-1 8.5 OBE LOADS ................................................................................................................... 8-2 8.6 TOTAL STRESS FOR FATIGUE CRACK GROWTH .................................................. 8-2 8.7 FATIGUE CRACK GROWTH ANALYSIS .................................................................... 8-2 8.8 ANALYSIS PROCEDURE ............................................................................................. 8-2 8.9 RESULTS ........................................................................................................................ 8-4 8.10 REFERENCES ................................................................................................................ 8-5 9.0 ASSESSMENT OF MARGINS.................................................................................................... 9-1

10.0 CONCLUSION

S ........................................................................................................................ 10-1 APPENDIX A: LIMIT MOMENT ........................................................................................................... A-1 WCAP-17778-NP March 2023 Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vi LIST OF TABLES Table 3-1 Summary of H. B. Robinson Nuclear Power Plant Piping Geometry and Normal Operating Condition for the RHR Line..................................................................................................... 3-4 Table 3-2 Summary of H. B. Robinson Unit 2 Normal Loads and Stresses for Residual Heat Removal (RHR) Line .............................................................................................................................. 3-5 Table 3-3 Summary of H. B. Robinson Unit 2 Faulted Loads and Stresses for Residual Heat Removal (RHR) Line .............................................................................................................................. 3-6 Table 4-1 Measured Tensile Properties for RHR Line Material A376 TP316 ......................................... 4-2 Table 4-2 Mechanical Properties for RHR Line Material at Operating Temperatures............................. 4-2 Table 6-1 Flaw Sizes Yielding a Leak Rate of 10 gpm for the RHR Line ............................................... 6-4 Table 7-1 Stability Results for the RHR line Based on Limit Load ......................................................... 7-4 Table 8-1 Design Transients Considered for Fatigue Crack Growth Evaluation ..................................... 8-6 Table 8-2 RHR Lines Fatigue Crack Growth Results .............................................................................. 8-7 Table 9-1 Leakage Flaw Sizes, Critical Flaw Sizes and Margins for RHR Line ..................................... 9-2 WCAP-17778-NP March 2023 Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vii LIST OF FIGURES Figure 3-1 H. B. Robinson Unit 2 RHR Line Layout ........................................................................ 3-7 Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures ........................ 6-5 Figure 6-2 [ ]a,c,e Pressure Ratio as a Function of L/D ............................................... 6-6 Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack .......................................... 6-7 Figure 7-1 [ ]a,c,e Stress Distribution ............................................................................. 7-5 Figure 7-2 Critical Flaw Size Prediction - Node 323........................................................................ 7-6 Figure 7-3 Critical Flaw Size Prediction - Node 320........................................................................ 7-7 Figure 8-1 Schematic of RHR Line at RCL Hot Leg Nozzle Weld Location.................................... 8-8 Figure 8-2 Reference Crack Growth Curves for Stainless Steel in Air Environments ..................... 8-9 Figure A-1 Pipe with a Through-Wall Crack in Bending .................................................................. A-2 WCAP-17778-NP March 2023 Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-1

1.0 INTRODUCTION

1.1 PURPOSE The current structural design basis for the Residual Heat Removal (RHR) lines requires postulating non-mechanistic circumferential and longitudinal pipe breaks. This results in additional plant hardware (e.g., pipe whip restraints and jet shields) which would mitigate the dynamic consequences of the pipe breaks. It is, therefore, highly desirable to be realistic in the postulation of pipe breaks for the RHR lines.

Presented in this report are the descriptions of a mechanistic pipe break evaluation method and the analytical results that can be used for establishing that a circumferential type of break will not occur within the RHR lines.

1.2 SCOPE AND OBJECTIVES The scope of this report is limited to the high energy Class 1 portion of the RHR lines (primary loop junction to the second isolation valve). A schematic drawing of the piping system is shown in Section 3.

The recommendations and criteria proposed in SRP 3.6.3 (References 1-1 and 1-2) are used in this evaluation. The criteria and the resulting steps of the evaluation procedure can be briefly summarized as follows:

1. Calculate the applied loads. Identify the location(s) at which the highest faulted stress occurs.

2. Identify the materials and the material properties.

3. Postulate a surface flaw governing location. Determine fatigue crack growth. Show that a through-wall crack will not result.

4. Postulate a through-wall flaw at the governing location(s). 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. Demonstrate that there is a margin of 10 between the calculated leak rate and the leak detection capability.

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

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 material types used in the Plant, provide representative material properties.

This report provides a fracture mechanics demonstration of RHR integrity for H. B. Robinson Unit 2 consistent with the NRC position for exemption from consideration of postulated pipe rupture dynamic effects (Reference 1-3).

Introduction March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-2 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.

Note that there are several locations in this report where proprietary information has been identified and bracketed. For each of the bracketed locations, the reason for the proprietary classification is given using a standardized system. The proprietary brackets are labeled with three different letters, to provide this information, and the explanation for each letter is given below:

a. The information reveals the distinguishing aspects of a process or component, structure, tool, method, etc., and the prevention of its use by Westinghouses competitors, without license from Westinghouse, gives Westinghouse a competitive economic advantage.

b. The information, if used by a competitor, would reduce the competitors expenditure of resources or improve the competitors advantage in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.

c. The information reveals aspects of past, present, or future Westinghouse or customer-funded development plans and programs of potential commercial value to Westinghouse.

The proprietary information in the brackets which has been deleted in this version of this report are provided in the proprietary class 2 document (WCAP-17778-P, Revision 1).

1.3 REFERENCES

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

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

1-3 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 Register/Vol. 52, No. 207/Tuesday, October 27, 1987/Rules and Regulations, pp. 41288 41295.

Introduction March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

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 loop piping and connected Class 1 piping 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 (PWRs). 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.

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 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.

Operation and Stability of the Reactor Coolant System March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 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 1979, several instances of cracking in PWR feedwater piping 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.

Primary Water Stress Corrosion Cracking (PWSCC) occurred in V. C. Summer reactor vessel hot leg nozzle, Alloy 82/182 weld. It should be noted that this susceptible material is not found at the H. B.

Robinson Unit 2 RHR line.

2.2 WATER HAMMER Overall, there is a low potential for water hammer in the RCS and connecting RHR lines since they are designed and operated to preclude the voiding condition in normally filled lines. The RCS and connecting RHR lines including piping and components are 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 by the control rod positions; pressure is controlled also within a narrow range for steady-state conditions by the pressurizer heaters and pressurizer spray. 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 and the connecting auxiliary lines. Preoperational testing and operating experience have verified the Westinghouse approach. The operating transients of the RCS primary piping and connected RHR lines 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 is discussed in the form of a fatigue crack growth assessment, in Section 8.0.

Operation and Stability of the Reactor Coolant System March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 Pump vibrations during operation would result in high cycle fatigue loads in the piping system. During operation, an alarm signals the exceedance of the RC pump shaft vibration limits. Field measurements have been made on the reactor coolant loop piping in a number of plants during hot functional testing.

Stresses in the elbow below the RC pump have been found to be very small, between 2 and 3 ksi at the highest. Field measurements on typical PWR plants indicate vibration amplitudes less than 1 ksi. When translated to the connecting RHR lines, these stresses would be even lower, well below the fatigue endurance limit for the RHR line materials and would result in an applied stress intensity factor below the threshold for fatigue crack growth.

2.4 OTHER POSSIBLE DEGRADATION DURING SERVICE OF THE RHR LINES Thermal stratification occurs when conditions permit hot and cold layers of water to exist simultaneously in a horizontal pipe. This can result in significant thermal loadings due to the high fluid temperature differentials. Changes in the stratification state result in thermal cycling, which can cause fatigue damage. This was an important issue in PWR feedwater line and pressurizer surge line piping, where temperature differentials of 300°F were not uncommon.

For the RHR piping in the H. B. Robinson Unit 2, thermal stratification due to NRC Bulletin 88-08 is not a concern (Reference 2-3).

The RHR Lines and the associated fittings for H. B. Robinson Nuclear Power Plant are forged product forms, which are not susceptible to toughness degradation due to thermal aging.

The maximum normal operating temperature of the RHR piping is about 605°F. This is well below the temperature that would cause any creep damage in stainless steel piping.

2.5 REFERENCES

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

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

2-3 NRC letter, NRC Bulletin 88-08, Thermal Stresses in Piping Connected to Reactor Coolant Systems- H. B. Robinson Steam Electric Plant Unit no. 2 (TAC No. 69679) Docket No. 50-261, October 1, 1991.

Operation and Stability of the Reactor Coolant System March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 3.0 PIPE GEOMETRY AND LOADING 3.1 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 Z

where,

= stress F = axial load M = moment A = pipe cross-sectional area Z = section modulus The moments for the desired loading combinations are calculated by the following equation:

M = M2x + M2y + M2z (3-2)

where, Mx = X component of moment, Torsion My = Y component of bending moment Mz = Z component of bending moment The axial load and moments for leak rate predictions and crack stability analyses are computed by the methods to be explained in Sections 3.3 and 3.4.

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

F = FDW + FTH + FP (3-3)

MX = (MX)DW + (MX)TH (3-4)

MY = (MY)DW + (MY)TH (3-5)

MZ = (MZ)DW + (MZ)TH (3-6)

Pipe Geometry and Loading March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-2 The subscripts of the above equations represent the following loading cases:

DW = 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).

The as-built dimensions and normal operating conditions are given in Table 3-1. The loads based on this method of combination are provided in Table 3-2 at all the weld locations identified in Figure 3-1.

3.3 LOAD COMBINATION FOR CRACK STABILITY ANALYSES In accordance with Standard Review Plan 3.6.3 (References 3-1 and 3-2), the absolute sum of loading components can be applied which results in higher magnitude of combined loads. If crack stability is demonstrated using these loads, the LBB margin on loads can be reduced from 2 to 1.0. The absolute summation of loads is shown in the following equations:

F = FDW + FTH + FP + FSSEINERTIA + FSSEAM (3-7)

MX = (MX)DW + (MX)TH + (MX)SSEINERTIA+ (MX)SSESAM (3-8)

MY = (MY)DW + (MY)TH + (MY)SSEINERTIA+ (MY)SSEAM (3-9)

MZ = (MZ)DW + (MZ)TH + (MZ)SSEINERTIA+ (MZ)SSEAM (3-10) where subscript SSEINERTIA refers to safe shutdown earthquake inertia, SSEAM is safe shutdown earthquake anchor motion.

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 3-1) are given in Table 3-3.

3.4

SUMMARY

OF LOADS AND GEOMETRY FOR THE RHR LINES The load combinations were evaluated at the various weld locations. Normal loads were determined using the algebraic sum method whereas the faulted loads were combined using the absolute sum method.

The normal operating loadings for the RHR lines are Pressure (P), Deadweight (DW) and Normal Operating Thermal Expansion (TH) loads. The faulted loadings consist of Normal Operating loads plus Safe Shutdown Earthquake (SSE) loads including the Seismic Anchor Motion.

Table 3-1 shows the piping geometry and normal operating condition for the RHR line at the weld locations. The minimum pipe wall thickness at the weld counterbore is used in the analysis, which is Pipe Geometry and Loading March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-3 based on the maximum allowed counterbore at a butt weld (Reference 3-3). The normal and faulted loads are tabulated in Tables 3-2 and 3-3 respectively at the weld locations for RHR Line.

3.5 GOVERNING LOCATIONS FOR THE RHR LINES All the welds at the RHR line are fabricated using the GTAW/SMAW combination or GTAW weld process procedures. The governing locations were established on the basis of the pipe schedules, material type, operating temperature, operating pressure, and the highest faulted stresses at the welds. Figure 3-1 shows the schematic layout of the RHR line and also identifies the governing weld locations.

The governing locations enveloping the RHR line are found to be: Node 323 and Node 320.

3.6 REFERENCES

3-1 Standard Review Plan: Public Comments Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 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.

3-3 American National Standards, Butt Welding Ends, ANSI B16.25-1979.

Pipe Geometry and Loading March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-4 Table 3-1 Summary of H. B. Robinson Nuclear Power Plant Piping Geometry and Normal Operating Condition for the RHR Line Weld Outer Minimum Wall Normal Operating Material Location Diameter Thickness Pressure Temperature Type Node (in) (in) (psig) (oF) 323 A376 TP316 14.00 1.114 2235 605 322 A376 TP316 14.00 1.114 2235 605 320 A376 TP316 14.00 1.114 2235 350 319 A376 TP316 14.00 1.114 2235 350 317 A376 TP316 14.00 1.114 2235 350 316 A376 TP316 14.00 1.114 2235 350 315 A376 TP316 14.00 1.114 2235 350 314 A376 TP316 14.00 1.114 2235 350 313 A376 TP316 14.00 1.114 2235 350 312 A376 TP316 14.00 1.114 2235 350 1311 A376 TP316 14.00 1.114 2235 350 Pipe Geometry and Loading March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-5 Table 3-2 Summary of H. B. Robinson Unit 2 Normal Loads and Stresses for Residual Heat Removal (RHR) Line Weld Moment Axial Forceb Moment Axial Stress Total Stress Locationa Stress (lbs) (in-lbs) (psi) (psi)

Node (psi) 323 245667 244817 5446 1817 7263 322 245667 296448 5446 2200 7646 320 245643 299346 5446 2222 7667 319 243621 250051 5401 1856 7257 317 244046 64144 5410 476 5886 316 245471 44861 5442 333 5775 315 245464 36213 5442 269 5711 314 243585 43166 5400 320 5721 313 244024 109089 5410 810 6220 312 243958 130700 5408 970 6378 1311 243933 112207 5408 833 6241 Notes:

a. See Figure 3-1

b. Included Pressure Pipe Geometry and Loading March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-6 Table 3-3 Summary of H. B. Robinson Unit 2 Faulted Loads and Stresses for Residual Heat Removal (RHR) Line Weld Moment Axial Forcec Moment Axial Stress Total Stress Location Stress (lbs) (in-lbs) (psi) (psi)

Nodea,b (psi) 323 252725 590504 5603 4382 9985 322 252725 559261 5603 4151 9753 320 252466 602162 5597 4469 10066 319 244997 448750 5431 3330 8762 317 249680 579326 5535 4299 9835 316 250893 581339 5562 4314 9877 315 250812 364615 5560 2706 8266 314 244609 419216 5423 3111 8534 313 249430 469278 5530 3483 9013 312 248953 348625 5519 2587 8106 1311 248680 380022 5513 2820 8333 Notes:

a. See Figure 3-1

b. See Table 3-1 for dimensions

c. Included Pressure Pipe Geometry and Loading March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-7 Figure 3-1 H. B. Robinson Unit 2 RHR Line Layout Pipe Geometry and Loading March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-1 4.0 MATERIAL CHARACTERIZATION 4.1 RHR LINE PIPE, FITTINGS AND WELD MATERIALS The material type of the RHR line for the H. B. Robinson Unit 2 Nuclear Power Plant is A376 TP316.

This is a wrought product of the type used for the piping in several PWR plants. The RHR line system does not include any cast pipes or cast fittings. The welding processes used are Gas Tungsten Arc Weld (GTAW) and Shielded Metal Arc Weld (SMAW) combination. Figure 3-1 show the schematic layout of the RHR line and also identifies the weld locations by node points.

In the following sections the tensile properties of the materials are presented for use in the Leak-Before-Break analyses.

4.2 TENSILE PROPERTIES The Certified Materials Test Reports (CMTRs) for the H. B. Robinson Unit 2 RHR line were used to establish the tensile properties for the leak-before-break analyses. The tensile properties for the pipe material are provided in Table 4-1 for H. B. Robinson Unit 2.

For the A376 TP316 pipe material, the representative properties at operating temperatures of 605°F and 350°F are established from the tensile properties at room temperature given in Table 4-1 by utilizing Section II of the 2007 with the 2008 Addenda of ASME Boiler and Pressure Vessel Code (Reference 4-1).

Code tensile properties at the operating temperatures were obtained by interpolating between 300°F, 400°F, 600°F and 700°F tensile Code properties. Ratios of the Code tensile properties at the operating temperatures to the corresponding properties were then applied to the room temperature tensile properties obtained from CMTRs (Table 4-1) to obtain the H. B. Robinson Unit 2 RHR line specific properties at operating temperatures of 605°F and 350°F.

The average and lower bound yield strengths and ultimate strengths for the pipe material are tabulated in Table 4-2. The ASME Code modulus of elasticity values are also given, and Poisson's ratio was taken as 0.3.

4.3 REFERENCES

4-1 ASME Boiler and Pressure Vessel Code, 2007 Edition with the 2008 Addenda,Section II, Part D

- Properties (Customary) Materials.

Material Characterization March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-2 Table 4-1 Measured Tensile Properties for RHR Line Material A376 TP316 Heat Number Yield Strength (psi) Ultimate Strength (psi)

At Room Temp. At Room Temp.

J1434 (ser 3939) 40900 83800 J1434 (ser 3939) 46900 84800 J1434 (ser 3940) 44400 87400 J1434 (ser 3940) 41400 85500 J1434 (ser 3941) 43400 86400 J1434 (ser 3941) 43600 83400 J1434 (ser 3942) 48600 93200 J1434 (ser 3942) 43400 84900 DYEF(27009) 37200 78000 DYEF (27009) 37200 78000 DYEA (139788) 38600 83000 Table 4-2 Mechanical Properties for RHR Line Material at Operating Temperatures Lower Bound Average Yield Yield Stress Ultimate Strength Material Temperature (F) Strength (psi) (psi) (psi)

A376 TP316 605 26610 23386 74672 A376 TP316 350 31604 27776 75296 Modulus of E = 25.275 x 106 psi at 605F; E = 26.700 x 106 psi at 350F Elasticity:

Poissons ratio: 0.3 Material Characterization March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

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 critical 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 the H. B. Robinson RHR line piping. Tables 3-3 as well as Figure 3-1 are used for this evaluation.

Critical Locations The highest stressed location for the RHR line from hot leg to first valve is at Node 323 (See Table 3-3 and Figure 3-1). The highest stressed location for the RHR line from first valve to second valve is at Node 320 (See Table 3-3 and Figure 3-1). Node 323 is determined to be the critical location at the HL RHR branch nozzle due to the limiting cross-sectional properties of the pipe, rather than the thicker reinforcement area at the nozzle to reactor coolant loop branch weld. Node 323 and Node 320 are the load critical locations for all the weld locations in the RHR line piping.

Critical Location and Evaluation Criteria March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

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, (L/DH) is greater than [

]a,c,e 6.3 CALCULATION METHOD The basic method used in the leak rate calculations is the method developed by [

]a,c,e 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 RHR line enthalpy condition and an assumed flow.

Once Pc was found for a given mass flow, the [ ]a,c,e was found from Figure 6-2 (taken from Reference 6-2). For all cases considered, since [

]a,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 Pf =[ ]a,c,e (6-1) where the friction factor f is determined using the [ ]a,c,e The crack relative roughness, , 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.

Leak Rate Predictions March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-2 That is, for the RHR line:

Absolute Pressure - 14.7 = [ ]a,c,e (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 RHR line 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.

For the single phase cases with lower temperature (350°F at Node 320), leakage rate is calculated by the following equation (Reference 6-4) with crack opening area obtained by the method from Reference 6-3.

Q A(2 gP / k ) 0.5 ft3/sec; (6-3)

Where, P = pressure difference between stagnation and back pressure (lb/ft2), g = acceleration of gravity (ft/sec2), = fluid density at atmospheric pressure (lb/ft3), k = friction loss including passage loss, inlet and outlet of the through-wall crack, A = crack opening area (ft2).

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 3-2 were applied in these calculations. The crack opening areas were estimated using the method of Reference 6-3 and the leak rates were calculated using the formulation described above. The average material properties of Section 4.0 (see Table 4-2) 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 H. B. Robinson Unit 2. The flaw sizes so determined are called leakage flaw sizes.

The H. B. Robinson Unit 2 RCS pressure boundary leak detection system meets the intent of Regulatory Guide 1.45. 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.

Leak Rate Predictions March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3

6.5 REFERENCES

6-1 [

]a,c,e 6-2 M. M, El-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 II-1, NUREG/CR-3464, September 1983.

6-4 Crane, D. P., Handbook of Hydraulic Resistance Coefficient.

Leak Rate Predictions March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-4 Table 6-1 Flaw Sizes Yielding a Leak Rate of 10 gpm for the RHR Line Location Leakage Flaw Size (in)

Node 323 6.00 Node 320 5.25 Leak Rate Predictions March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-5 Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures Leak Rate Predictions March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-6 Figure 6-2 [ ]a,c,e Pressure Ratio as a Function of L/D Leak Rate Predictions March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-7 Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack Leak Rate Predictions March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-1 7.0 FRACTURE MECHANICS EVALUATION 7.1 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:

[ ]a,c,e where:

[

]a,c,e f = 0.5 (y + u) = flow stress, psi

[

]a,c,e 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 March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-2 7.2 RESULTS OF CRACK STABILITY EVALUATION A stability analysis based on limit load was performed for these locations as described in Section 7.1. The weld process types, at the critical locations at Node 323 and Node 320 are used as GTAW and SMAW combination. The Z correction factor for SMAW (References 7-2 and 7-3) are as follows:

Z = 1.15 [1.0 + 0.013 (OD-4)] for SMAW where OD is the outer diameter of the pipe in inches.

The Z-factor for the GTAW weld is 1.0. The Z-factor for the SMAW was calculated for the critical locations, using the dimensions given in Table 3-1. The applied faulted loads (Table 3-3) were increased by the Z factor and plots of limit load versus crack length were generated as shown in Figures 7-2 and 7-3. Lower bound material properties were used from Table 4-2. Table 7-1 summarizes the results of the stability analyses based on limit load. The leakage flaw sizes are also presented in the same table.

Fracture Mechanics Evaluation March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-3

7.3 REFERENCES

7-1 Kumar, V., German, M. D. and Shih, C. P., An Engineering Approach for Elastic-Plastic Fracture Analysis, EPRI Report NP-1931, Project 1237-1, Electric Power Research Institute, July 1981.

7-2 Standard Review Plan; Public Comment Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 52, No. 167/Friday, August 28, 1987/Notices, pp. 32626-32633.

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

Fracture Mechanics Evaluation March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-4 Table 7-1 Stability Results for the RHR line Based on Limit Load Location Critical Flaw Size (in) Leakage Flaw Size (in)

Node 323 18.52 6.00 Node 320 19.13 5.25 Fracture Mechanics Evaluation March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-5 Figure 7-1 [ ]a,c,e Stress Distribution Fracture Mechanics Evaluation March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-6 a,c,e OD = 14.00 in. y-min = 23.386 ksi F = 252.725 kips t = 1.114 in. u-min = 74.672 ksi M = 590.504 in-kips A376 TP316 with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-2 Critical Flaw Size Prediction - Node 323 Fracture Mechanics Evaluation March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-7 a,c,e OD = 14.00 in. y-min = 27.776 ksi F = 252.466 kips t = 1.114 in. u-min = 75.296 ksi M = 602.162 in-kips A376 TP316 with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-3 Critical Flaw Size Prediction - Node 320 Fracture Mechanics Evaluation March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-1 8.0 ASSESSMENT OF FATIGUE CRACK GROWTH

8.1 INTRODUCTION

The fatigue crack growth (FCG) analysis is not a requirement for the LBB analysis (see References 8-1 and 8-2) since the LBB analysis is based on the postulation of a through-wall flaw, whereas the FCG analysis is performed based on the surface flaw. However, a fatigue crack growth (FCG) assessment of the H. B. Robinson Unit 2 RHR line was determined by comparison with a generic fatigue crack growth analysis of a similar piping system. The details of the generic fatigue crack growth analysis are presented below. By comparing the parameters critical to the fatigue crack growth analysis between H. B. Robinson and the generic analysis, it was concluded that the generic analysis would adequately cover the fatigue crack growth of the H. B. Robinson Unit 2 RHR lines.

Due to similarities in Westinghouse PWR designs, it was possible to perform a representative fatigue crack growth calculation which would be applicable to the H. B. Robinson Plant.

8.2 CRITICAL LOCATION FOR FATIGUE CRACK GROWTH ANALYSIS The weld location at the RCL hot leg nozzle to RHR line (see Figure 8-1) was determined to be the most critical location for the fatigue crack growth evaluation. The nozzle configuration and weld location is shown in Figure 8-1. The geometry of the pipe was identical between the H. B. Robinson Unit 2 and the generic model (14 Schedule 140). Both analyses used austenitic stainless steel at the critical location.

8.3 DESIGN TRANSIENTS The transient conditions selected for this evaluation are based on conservative estimates of the magnitude and the frequency of the temperature fluctuations resulting from various operating conditions in the plant.

These are representative of the conditions which are considered to occur during plant operation. The fatigue evaluation based on these transients provides confidence that the component is appropriate for its application over the design life of the plant. The normal operating and upset thermal transients were considered for this evaluation. Out of these 20 transients were used in the final fatigue crack growth analysis as listed in Table 8-1.

8.4 STRESS ANALYSIS A thermal transient stress analysis was performed for a typical plant similar to the H. B. Robinson Unit 2 to obtain the through-wall stress profiles for use in the fatigue crack growth analysis. The generic RHR line design transients described in Section 8.3 were used.

A simplified analysis method was used to develop conservative maximum and minimum linear through wall stress distributions due to thermal transients. In this method, a 1-D computer program was used to perform the thermal analysis to determine the through wall temperature gradients as a function of time.

The inside surface stress was calculated by using an equation, which is similar to the transient portion of ASME Section III NB 3600, Equation (11). The effect of discontinuity was included in the analysis by performing a separate 1-D thermal analysis for the pipe and nozzle. The maximum and minimum inside surface stresses were then obtained by searching the inside surface stress values calculated for each time Assessment of Fatigue Crack Growth March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-2 step of the transient solution. The outside surface stresses corresponding to the maximum and minimum inside surface stresses were then calculated by a similar method. The maximum and minimum linear through wall stress distribution for each thermal transient was obtained by joining the corresponding inside and outside surface stresses by a straight line. These two stress profiles are called the maximum and minimum through wall stress distributions respectively, for convenience. The stresses due to the generic pressure and the generic moment loading were then superimposed on the through wall cyclical stresses to obtain the total maximum and minimum stress profile for each transient.

8.5 OBE LOADS The stresses due to OBE loads were neglected in the fatigue crack growth analysis since these loads are not expected to contribute significantly to crack growth due to the small number of cycles.

8.6 TOTAL STRESS FOR FATIGUE CRACK GROWTH The total through wall stress at a section was obtained by superimposing the generic pressure stress and the generic moment stresses on the thermal transient stresses. Thus, the total stress for fatigue crack growth at any point is given by the following equation:

Stress due to Total Stress Stress due to Moment (DW + Thermal Transient For Fatigue = + +

Internal Pressure Thermal Stress Crack Growth Expansion) 8.7 FATIGUE CRACK GROWTH ANALYSIS The fatigue crack growth analysis was performed to determine the effect of the design thermal transients tabulated in Table 8-1. The analysis was performed for the critical cross-section identified in Figure 8-1.

A range of crack depths was postulated, and each was subjected to the transients in Table 8-1, which included pressure and moment loads.

8.8 ANALYSIS PROCEDURE The fatigue crack growth analyses presented herein were conducted in the same manner as suggested by Section XI, Appendix A of the ASME Boiler and Pressure Vessel Code (Reference 8-3). The analysis procedure involves assuming an initial flaw exists at some point and predicting the growth of that flaw due to an imposed series of transient stresses. The growth of a crack per loading cycle is dependent on the range of applied stress intensity factor, KI, by the following:

da C o K In (8-1) dN where "Co" and the exponent "n" are material properties, and KI is defined later. For inert environments these material properties are constants, but for some water environments they are dependent on the level Assessment of Fatigue Crack Growth March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-3 of mean stress present during the cycle. This can be accounted for by adjusting the value of "Co" by a function of the ratio of minimum to maximum stress for any given transient, as will be discussed later.

Fatigue crack growth properties of stainless steel in a pressurized water environment have been used in the analysis.

The input required for a fatigue crack growth analysis is basically the information necessary to calculate the parameter KI, which depends on crack and structure geometry and the range of applied stresses in the area where the crack exists. Once KI is calculated, the growth due to that particular cycle can be calculated by Equation (8-1). This increment of growth is then added to the original crack size, the KI adjusted, and the analysis proceeds to the next transient. The procedure is continued in this manner until all the transients have been analyzed.

The applied stresses at the flaw locations are resolved into membrane and bending stresses with respect to the wall thickness. Pressure, thermal, and discontinuity stresses are considered in the determination of the KI factors.

The stress intensity factor at the point of maximum depth is calculated from the membrane and bending stresses using the following equation taken from the ASME Code (Reference 8-3):

a KI [ mMm bMb ]

Q where : m, b = Membrane and Bending Stress, respectively a = Minor Semi-Axis (flaw depth)

Q = Flaw Shape Parameter Including A Plastic Zone Correction Factor for Plane Strain Condition Q = [ 12 - 0.212 ( / ys)2]

1/ 2

/2 b2 a2 1 = 0 1 (

b 2

) cos 2 d

ys = Yield Strength of the Material

= m + b b = Major Semi-Axis (Flaw Length/2)

= Parametric Angle of the Ellipse Mm = Correction Factor for Membrane Stress Assessment of Fatigue Crack Growth March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-4 Mb = Correction Factor for Bending Stress The appropriate values of Mm and Mb as a function of crack geometry can be found in Reference 8-3. The range of stress intensity factor (KI) for fluctuation of applied stress is determined by first finding the maximum and minimum stress intensity factor (KI max, KI min) during a given transient and then calculating the range of stress intensity factor (KI = KI max - KI min). At times KI min may go below zero, in these cases, KI min is set equal to zero before KI is determined.

Calculation of the fatigue crack growth for each cycle was then carried out using the reference fatigue crack growth rate law determined from consideration of the available data for stainless steel in a pressurized water environment. This law allows for the effect of mean stress or R ratio (KI min/KI max) on the growth rates.

The reference crack growth law used for the stainless steel RHR pipe system was taken from that developed by the Metal Properties Council - Pressure Vessel Research Committee Task Force In Crack Propagation Technology. The reference curve has the equation:

[ (8-2)

]a,c,e This equation appears in Appendix C of ASME Section XI for air environments and its basis is provided in Reference 8-4 in Figure 8-2. For water environments, an environmental factor of [ ]a,c,e was used, based on the crack growth tests in PWR environments reported in Reference 8-5.

8.9 RESULTS Fatigue crack growth analyses were carried out at the critical cross-section. Analysis was completed for a range of postulated flaw sizes oriented circumferentially, and the results are presented in Table 8-2. The postulated flaws are assumed to have an aspect ratio of six to one. Even for the largest postulated flaw of 0.35 inch, which is about 35 percent of the wall thickness, the results project that the flaw growth through the wall will not occur during the 40/60 year design life of the plant. Transients and cycles for the H. B.

Robinson Unit 2 plant for 40-year transient set will remain bounding for 60 years (Reference 8-6), the FCG results shown in Table 8-2 is also applicable for the 60 years.

Therefore, fatigue crack growth should not be a concern for the H. B. Robinson RHR Line.

Assessment of Fatigue Crack Growth March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-5 8.10 REFERENCES 8-1 Standard Review Plan; Public Comment Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 52, No. 167/Friday, August 28, 1987/Notices, pp. 32626-32633.

8-2 NUREG-0800 Revision 1, March 2007, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures.

8-3 ASME Boiler and Pressure Vessel Code Section XI, 2007 Edition with the 2008 Addenda, Rules for Inservice Inspection of Nuclear Power Plant Components.

8-4 James, L. A., and Jones, D. P., Fatigue Crack Growth Correlations for Austenitic Stainless Steel in Air, in Predictive Capabilities in Environmentally Assisted Cracking, ASME publication PVP-99, Dec. 1985.

8-5 Bamford, W. H., Fatigue Crack Growth of Stainless Steel Piping in a Pressurized Water Reactor Environment, Trans ASME, Journal of Pressure Vessel Technology, Feb. 1979. Engineering Development Labs Report HEDL-TME-76-43, May 1976.

8-6 NUREG-1785, Safety Evaluation Report Related to the License Renewal of H. B. Robinson Steam Electric Plant, Unit 2.

Assessment of Fatigue Crack Growth March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-6 Table 8-1 Design Transients Considered for Fatigue Crack Growth Evaluation Trans. No. Description No. of Occurrences 1 Unit Loading 13,200 2 Unit Unloading 13,200 3 Step Load Increase 2,000 4 Step Load Decrease 2,000 5 Large Step Load Decrease with Steam Dump 200 6 Feedwater Cycling 2000 7 Unit Loading Between 0 and 15% Power 500 8 Unit Unloading Between 0 and 15% Power 500 9 Loss of Load 80 10 Loss of Power 40 11 Partial Loss of Flow-Dead Loop 80 12 Partial Loss of Flow-Active Loop 80 13 Reactor Trip with no Inadvertent Cooldown 230 14 Reactor Trip with Cooldown; No Safety Injection 160 15 Reactor Trip with Cooldown Actuating Safety Injection 10 16 Inadvertent RCS Depressurization 20 17 Control Rod Drop 80 18 Inadvertent Safety Injection 60 19 Turbine Roll Test 20 20 Steady-State and Random Fluctuations 3.2 x 106 Assessment of Fatigue Crack Growth March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-7 Table 8-2 RHR Lines Fatigue Crack Growth Results Initial Crack Depth (in) After a,c,e Crack Depth (in) 10 Years 20 Years 30 Years 40/60 Years*

Note: *Because transients and cycles for the H. B. Robinson Unit 2 plant for 40-year transient set will remain bounding for 60 years, the FCG results shown in Table 8-2 are also applicable for 60 years.

Assessment of Fatigue Crack Growth March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-8 Figure 8-1 Schematic of RHR Line at RCL Hot Leg Nozzle Weld Location Assessment of Fatigue Crack Growth March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-9 Figure 8-2 Reference Crack Growth Curves for Stainless Steel in Air Environments Assessment of Fatigue Crack Growth March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

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 evaluations of Section 7.2 are used in performing the assessment of margins. Margins are shown in Table 9-1. All 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.

Assessment of Margins March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-2 Table 9-1 Leakage Flaw Sizes, Critical Flaw Sizes and Margins for RHR Line Location Critical Flaw Size Leakage Flaw Size Margin (in) (in)

Node 323 18.52 6.00 3.1 Node 320 19.13 5.25 3.6 Assessment of Margins March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 10-1

10.0 CONCLUSION

S This report justifies the elimination of RHR line pipe break from the structural design basis for the H. B.

Robinson Unit 2 during the 60 years plant life 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.

b. Water hammer should not occur in the primary loop piping and connected Class 1 piping because of system design, testing, and operational considerations.

c. The effects of low and high cycle fatigue on the integrity of the RHR line are negligible.

d. Ample margin exists between the leak rate of small stable flaws and the capability of the H. B. Robinson Unit 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.

g. Fatigue crack growth results using the 40-year design transients and cycles (shown to be applicable for 60 years) show that there will be insignificant growth through the wall for the license renewal period (60-year plant life).

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 loading, pipe geometry and material properties considerations, enveloping critical (governing) locations were determined at which leak-before-break crack stability evaluations were made.

Through-wall flaw sizes were postulated 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 RHR line piping.

Therefore, the Leak-Before-Break conditions and margins are satisfied for the H. B. Robinson Unit 2 RHR line piping. It is demonstrated that the dynamic effects of the pipe rupture resulting from postulated breaks in the RHR line piping need not be considered in the structural design basis of H. B. Robinson Unit 2 for the 60-years.

Conclusions March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

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

[

] a,c,e Appendix A: Limit Moment March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-2 Figure A-1 Pipe with a Through-Wall Crack in Bending Appendix A: Limit Moment March 2023 WCAP-17778-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:07:50 AM. (This statement was added by the PRIME system upon its validation)

RA-22-0290, Attachment 4 H. B. Robinson Steam Electric Plant, Unit No. 2 ATTACHMENT 4 WCAP-17779-NP, REVISION 1, "TECHNICAL JUSTIFICATION FOR ELIMINATING ACCUMULATOR LINE RUPTURE AS THE STRUCTURAL DESIGN BASIS FOR H. B.

ROBINSON UNIT 2, MARCH 2023 (REDACTED)

Westinghouse Non-Proprietary Class 3 WCAP-17779-NP March 2023 Revision 1 Technical Justification for Eliminating Accumulator Line Rupture as the Structural Design Basis for H. B. Robinson Unit 2

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17779-NP Revision 1 Technical Justification for Eliminating Accumulator Line Rupture as the Structural Design Basis for H. B. Robinson Unit 2 March 2023 Author: Nadia B. Petkova

• Operating Plants Piping and Supports Reviewer: Momo Wiratmo*

Operating Plants Piping and Supports Approved: Lynn A. Patterson, Manager*

Reactor Vessel and Containment Vessel Design and Analysis

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

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

© 2023 Westinghouse Electric Company LLC All Rights Reserved

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii RECORD OF REVISIONS Revision Date Revision Description Original Issue (WCAP-17779-NP). This is the non-proprietary class 3 0 August 2013 version of WCAP-17779-P, Revision 0.

This is the non-proprietary class 3 version of WCAP-17779-P, Revision 1 March 2023 1.

WCAP-17779-NP March 2023 Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iv TABLE OF CONTENTS

1.0 INTRODUCTION

........................................................................................................................ 1-1 1.1 PURPOSE ........................................................................................................................ 1-1 1.2 SCOPE AND OBJECTIVES ........................................................................................... 1-1

1.3 REFERENCES

................................................................................................................ 1-2 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............................................................... 2-2

2.4 REFERENCES

................................................................................................................ 2-3 3.0 PIPE GEOMETRY AND LOADING ........................................................................................... 3-1 3.1 CALCULATION OF LOADS AND STRESSES ............................................................ 3-1 3.2 LOADS FOR LEAK RATE EVALUATION ................................................................... 3-1 3.3 LOAD COMBINATION FOR CRACK STABILITY ANALYSES ................................ 3-2

3.4 REFERENCES

................................................................................................................ 3-2 4.0 MATERIAL CHARACTERIZATION.......................................................................................... 4-1 4.1 ACCUMULATOR LINE PIPING AND WELD MATERIALS ...................................... 4-1 4.2 TENSILE PROPERTIES ................................................................................................. 4-1 4.3 REFERENCE................................................................................................................... 4-1 5.0 CRITICAL LOCATION AND EVALUATION CRITERIA......................................................... 5-1 5.1 CRITICAL LOCATIONS ................................................................................................ 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

................................................................................................................ 6-3 7.0 FRACTURE MECHANICS EVALUATION ............................................................................... 7-1 7.1 GLOBAL FAILURE MECHANISM .............................................................................. 7-1 7.2 RESULTS OF CRACK STABILITY EVALUATION..................................................... 7-2

7.3 REFERENCES

................................................................................................................ 7-2 8.0 ASSESSMENT OF FATIGUE CRACK GROWTH .................................................................... 8-1

8.1 INTRODUCTION

........................................................................................................... 8-1 8.2 CRITICAL LOCATION FOR FATIGUE CRACK GROWTH ANALYSIS ................... 8-1 8.3 DESIGN TRANSIENTS ................................................................................................. 8-1 8.4 STRESS ANALYSIS ....................................................................................................... 8-1 8.5 OBE LOADS ................................................................................................................... 8-2 8.6 TOTAL STRESS FOR FATIGUE CRACK GROWTH .................................................. 8-2 WCAP-17779-NP March 2023 Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 v 8.7 FATIGUE CRACK GROWTH ANALYSIS .................................................................... 8-2 8.7.1 ANALYSIS PROCEDURE ................................................................................ 8-2 8.8 RESULTS ........................................................................................................................ 8-3

8.9 REFERENCES

................................................................................................................ 8-4 9.0 ASSESSMENT OF MARGINS.................................................................................................... 9-1

10.0 CONCLUSION

S ........................................................................................................................ 10-1 APPENDIX A: LIMIT MOMENT .......................................................................................................... A-1 WCAP-17779-NP March 2023 Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vi LIST OF TABLES Table 3-1 Summary of H. B. Robinson Nuclear Power Plant Piping Geometry and Normal Operating Condition for Accumulator Lines Loop A, Loop B and Loop C.............................................. 3-3 Table 3-2 Summary of Robinson Unit 2 Normal Loads and Stresses for Accumulator Line Loop A ..... 3-4 Table 3-3 Summary of Robinson Unit 2 Normal Loads and Stresses for Accumulator Line Loop B ...... 3-5 Table 3-4 Summary of Robinson Unit 2 Normal Loads and Stresses for Accumulator Line Loop C ..... 3-6 Table 3-5 Summary of Robinson Unit 2 Faulted Loads and Stresses for Accumulator Line Loop A ...... 3-7 Table 3-6 Summary of Robinson Unit 2 Faulted Loads and Stresses for Accumulator Line Loop B ...... 3-8 Table 3-7 Summary of Robinson Unit 2 Faulted Loads and Stresses for Accumulator Line Loop C ..... 3-9 Table 4-1 Measured Tensile Properties for the Accumulator Line Material A376 TP316 ........................ 4-2 Table 4-2 Mechanical Properties for Accumulator Line Material at Operating Temperatures ................. 4-2 Table 6-1 Flaw Sizes Yielding a Leak Rate of 10 gpm for the Accumulator Line ................................... 6-4 Table 7-1 Stability Results for the Accumulator Lines Based on Limit Load .......................................... 7-3 Table 8-1 Design Transients Considered for Fatigue Crack Growth Evaluation ...................................... 8-5 Table 8-2 Accumulator Line Fatigue Crack Growth Results .................................................................... 8-6 Table 9-1 Leakage Flaw Sizes, Critical Flaw Sizes and Margins for Accumulator Line ......................... 9-2 WCAP-17779-NP March 2023 Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vii LIST OF FIGURES Figure 3-1 H. B. Robinson Unit 2 Loop A Accumulator Line Layout ........................................... 3-10 Figure 3-2 H. B. Robinson Unit 2 Loop B Accumulator Line Layout...3-11 Figure 3-3 H. B. Robinson Unit 2 Loop C Accumulator Line Layout ........................................... 3-12 Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures ...................... 6-5 Figure 6-2 [ ]a,c,e Pressure Ratio as a Function of L/D ................................................ 6-6 Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack ......................................... 6-7 Figure 7-1 [ ]a,c,e Stress Distribution ............................................................................. 7-4 Figure 7-2 Critical Flaw Size Prediction - Node 409 Loop C......................................................... 7-5 Figure 7-3 Critical Flaw Size Prediction - Node 360 Loop B.....7-5 Figure 7-4 Critical Flaw Size Prediction - Node 364 Loop B......................................................... 7-7 Figure 7-5 Critical Flaw Size Prediction - Node 3811 Loop C ....................................................... 7-8 Figure 8-1 Schematic of 10 Accumulator Line at RCL Cold Leg Nozzle Weld Location ............. 8-7 Figure 8-2 Reference Crack Growth Curves for Stainless Steel in Air Enviroments ...................... 8-8 Figure A-1 Pipe with a Through-Wall Crack in Bending ................................................................ A-2 WCAP-17779-NP March 2023 Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-1

1.0 INTRODUCTION

1.1 PURPOSE The current structural design basis for the H. B. Robinson Unit 2, 10" accumulator lines (from the cold legs Loop A, Loop B and Loop C) and attached 8" line connected to 10" accumulator lines require postulating non-mechanistic circumferential and longitudinal pipe breaks. This results in additional plant hardware (e.g., pipe whip restraints and jet shields) which would mitigate the dynamic consequences of the pipe breaks. It is, therefore, highly desirable to be realistic in the postulation of pipe breaks for the accumulator lines. Presented in this report are the descriptions of a mechanistic pipe break evaluation method and the analytical results that can be used for establishing that a circumferential type of break will not occur within the accumulator lines. This report includes the 10" accumulator lines (from the cold legs Loop A, Loop B and Loop C) and attached 8" line connected to the 10" accumulator lines (see Figures in Section 3), for convenient purpose throughout the report it is called as accumulator line.

1.2 SCOPE AND OBJECTIVES The scope of this report is limited to the accumulator line. Schematic drawings of the piping system are shown in Section 3. The recommendations and criteria proposed in SRP 3.6.3 (References 1-1 and 1-2) are used in this evaluation. The criteria and the resulting steps of the evaluation procedure can be briefly summarized as follows:

1. Calculate the applied loads. Identify the location(s) at which the highest faulted stress occurs.

2. Identify the materials and the material properties.

3. Postulate a surface flaw governing location. Determine fatigue crack growth. Show that a through-wall crack will not result.

4. Postulate a through-wall flaw at the governing location(s). 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. Demonstrate that there is a margin of 10 between the calculated leak rate and the leak detection capability.

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

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 types used in the Plant, provide representative material properties.

This report provides a fracture mechanics demonstration of accumulator line piping integrity for H. B.

Robinson Unit 2 consistent with the NRC position for exemption from consideration of postulated pipe rupture dynamic effects (Reference 1-3).

INTRODUCTION March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-2 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.

Note that there are several locations in this report where proprietary information has been identified and bracketed. For each of the bracketed locations, the reason for the proprietary classification is given using a standardized system. The proprietary brackets are labeled with three different letters, to provide this information, and the explanation for each letter is given below:

a. The information reveals the distinguishing aspects of a process or component, structure, tool, method, etc., and the prevention of its use by Westinghouses competitors, without license from Westinghouse, gives Westinghouse a competitive economic advantage.

b. The information, if used by a competitor, would reduce the competitors expenditure of resources or improve the competitors advantage in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.

c. The information reveals aspects of past, present, or future Westinghouse or customer-funded development plans and programs of potential commercial value to Westinghouse.

The proprietary information in the brackets which has been deleted in this version of this report are provided in the proprietary class 2 document (WCAP-17779-P, Revision 1).

1.3 REFERENCES

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

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

1-3 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 Register/Vol. 52, No. 207/Tuesday, October 27, 1987/Rules and Regulations, pp. 41288-41295.

INTRODUCTION March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

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 and attached class 1 piping 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 (PWRs). 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.

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 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.

OPERATION AND STABILITY OF THE REACTOR COOLANT SYSTEM March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 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 1979, several instances of cracking in PWR feedwater piping 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.

Primary Water Stress Corrosion Cracking (PWSCC) occurred in V. C. Summer reactor vessel hot leg nozzle, Alloy 82/182 weld. It should be noted that this susceptible material is not found at the H. B.

Robinson Unit 2 accumulator line.

2.2 WATER HAMMER Overall, there is a low potential for water hammer in the RCS and connecting accumulator lines since they are designed and operated to preclude the voiding condition in normally filled lines. The RCS and connecting accumulator lines including piping and components are 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 by the control rod positions; pressure is controlled also within a narrow range for steady-state conditions by the pressurizer heaters and pressurizer spray. 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 and the connecting auxiliary lines. Preoperational testing and operating experience have verified the Westinghouse approach. The operating transients of the RCS primary piping and connected accumulator lines 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 is discussed in the form of a fatigue crack growth assessment, in Section 8.0.

OPERATION AND STABILITY OF THE REACTOR COOLANT SYSTEM March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 Pump vibrations during operation would result in high cycle fatigue loads in the piping system. During operation, an alarm signals the exceedance of the RC pump shaft vibration limits. Field measurements have been made on the reactor coolant loop piping in a number of plants during hot functional testing.

Stresses in the elbow below the RC pump have been found to be very small, between 2 and 3 ksi at the highest. Field measurements on typical PWR plants indicate vibration amplitudes less than 1 ksi. When translated to the connecting accumulator lines, these stresses would be even lower, well below the fatigue endurance limit for the accumulator line materials and would result in an applied stress intensity factor below the threshold for fatigue crack growth.

2.4 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 March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 3.0 PIPE GEOMETRY AND LOADING 3.1 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 Z

where,

= stress F = axial load M = moment A = pipe cross-sectional area Z = section modulus The moments for the desired loading combinations are calculated by the following equation:

M = M2x + M2y + M2z (3-2)

where, Mx = X component of moment, Torsion My = Y component of bending moment Mz = Z component of bending moment The axial load and moments for leak rate predictions and crack stability analyses are computed by the methods to be explained in Sections 3.3 and 3.4.

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

F = FDW + FTH + FP (3-3)

MX = (MX)DW + (MX)TH (3-4)

MY = (MY)DW + (MY)TH (3-5)

MZ = (MZ)DW + (MZ)TH (3-6)

PIPE GEOMETRY AND LOADING March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-2 The subscripts of the above equations represent the following loading cases:

DW = 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).

The as-built dimensions and normal operating conditions are given in Table 3-1. The minimum pipe wall thickness at the weld counterbore is used in the analysis, which is based on the maximum allowed counterbore at a butt weld (Reference 3-3). The loads based on this method of combination are provided in Tables 3-2 to 3-4 at all the weld locations identified in Figures 3-1 to 3-3.

3.3 LOAD COMBINATION FOR CRACK STABILITY ANALYSES In accordance with Standard Review Plan 3.6.3 (References 3-1 and 3-2), the absolute sum of loading components can be applied which results in higher magnitude of combined loads. If crack stability is demonstrated using these loads, the LBB margin on loads can be reduced from 2 to 1.0. The absolute summation of loads is shown in the following equations:

F = FDW + FTH + FP + FSSEINERTIA + FSSEAM (3-7)

MX = (MX)DW + (MX)TH + (MX)SSEINERTIA+ (MX)SSESAM (3-8)

MY = (MY)DW + (MY)TH + (MY)SSEINERTIA+ (MY)SSEAM (3-9)

MZ = (MZ)DW + (MZ)TH + (MZ)SSEINERTIA+ (MZ)SSEAM (3-10) where subscript SSEINERTIA refers to safe shutdown earthquake inertia, SSEAM is safe shutdown earthquake anchor motion.

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 Figures 3-1 to 3-3) are given in Tables 3-5 to 3-7.

3.4 REFERENCES

3-1 Standard Review Plan: Public Comments Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 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.

3-3 American National Standards, Butt Welding Ends, ANSI B16.25-1979.

PIPE GEOMETRY AND LOADING March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-3 Table 3-1 Summary of H. B. Robinson Nuclear Power Plant Piping Geometry and Normal Operating Condition for Accumulator Lines Loop A, Loop B and Loop C Minimum Normal Operating Weld Location Nodes Outer Wall Loop (See Figures 3-1, 3-2 Diameter (in) Thickness Pressure Temperature and 3-3) (psig) (oF)

(in)

A 449 to 4470 10.750 0.896 2235 547 A 446 to 433 10.750 0.896 1500 280 A 43 to 4321 8.625 0.650 1500 280 A 4500 to 4550 10.750 0.896 1500 280 A 4551 to 4570 10.750 0.896 660 140 B 3421 to 3440 8.625 0.650 1500 280 B 354 to 3620 10.750 0.896 1500 280 B 3621 to 3640 10.750 0.896 660 140 B 3450 to 3500 10.750 0.896 1500 280 B 3510 to 353 10.750 0.896 2235 547 C 3782 to 3811 8.625 0.650 1500 280 C 38 to 4080 10.750 0.896 1500 280 C 4081 to 410 10.750 0.896 2235 547 C 383 to 384 10.750 0.896 1500 280 C 3851 to 3870 10.750 0.896 660 140 PIPE GEOMETRY AND LOADING March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-4 Table 3-2 Summary of Robinson Unit 2 Normal Loads and Stresses for Accumulator Line Loop A Weld Axial Axial Force Moment Moment Stress Total Stress Location Stress Node (lbs) (in-lbs) (psi) (psi)

(psi) 449 138403 142003 4992 2249 7242 4480 138496 101836 4996 1613 6609 448 138895 89770 5010 1422 6432 4470 138895 83139 5010 1317 6327 446 92561 103376 3339 1638 4976 445 92561 179516 3339 2844 6182 444 92696 173839 3344 2754 6097 441 92696 355914 3344 5638 8982 440 92561 357607 3339 5665 9004 439 96537 327912 3482 5194 8677 437 95882 119449 3459 1892 5351 4341 95653 237310 3450 3759 7209 4331 94085 152532 3394 2416 5810 433 94078 149182 3393 2363 5757 43 62728 146149 3854 4838 8692 4321 62728 139785 3854 4627 8481 4500 94470 119442 3408 1892 5300 450 94634 130785 3413 2072 5485 451 97073 141437 3501 2240 5742 452 95800 136998 3456 2170 5626 453 96110 138587 3467 2195 5662 4550 96110 132529 3467 2099 5566 4551 43158 126048 1557 1997 3553 456 43158 125275 1557 1984 3541 457 42119 120578 1519 1910 3429 4570 42177 148761 1521 2356 3878 PIPE GEOMETRY AND LOADING March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-5 Table 3-3 Summary of Robinson Unit 2 Normal Loads and Stresses for Accumulator Line Loop B Weld Axial Axial Force Moment Moment Stress Total Stress Location Stress Node (lbs) (in-lbs) (psi) (psi)

(psi) 3421 61438 47848 3775 1584 5359 343 50855 81499 3124 2698 5822 344 62368 91412 3832 3026 6858 3440 62410 93381 3834 3091 6926 354 90007 332774 3247 5271 8518 355 104571 448633 3772 7107 10879 357 80751 579267 2913 9176 12089 358 80853 598731 2916 9484 12401 359 79845 805253 2880 12756 15636 360 90007 900522 3247 14265 17512 361 80751 651617 2913 10322 13235 3620 80751 573146 2913 9079 11992 3621 27799 169356 1003 2683 3685 363 27799 320980 1003 5085 6087 364 58797 749985 2121 11880 14001 3640 58855 688251 2123 10902 13025 3450 91206 280065 3290 4436 7726 348 97910 123191 3532 1951 5483 349 91172 121410 3289 1923 5212 3500 91172 68726 3289 1089 4377 3510 137506 135966 4960 2154 7114 352 137506 161774 4960 2563 7523 353 137503 170760 4960 2705 7665 PIPE GEOMETRY AND LOADING March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-6 Table 3-4 Summary of Robinson Unit 2 Normal Loads and Stresses for Accumulator Line Loop C Weld Axial Axial Force Moment Moment Stress Total Stress Location Stress Node (lbs) (in-lbs) (psi) (psi)

(psi) 3782 57397 337881 3526 11185 14712 379 69042 344978 4242 11420 15662 380 62720 275322 3853 9114 12968 3811 62250 437154 3825 14472 18296 38 93519 505161 3373 8002 11375 390 93386 201953 3368 3199 6568 391 95608 258517 3449 4095 7544 392 95608 267108 3449 4231 7680 393 97139 282209 3504 4470 7974 395 92013 149457 3319 2368 5686 396 88256 189171 3183 2997 6180 398 88832 150942 3204 2391 5595 399 88256 370601 3183 5871 9054 400 91541 365139 3302 5784 9086 401 89045 254825 3212 4037 7249 403 89045 376076 3212 5957 9169 404 89695 244909 3235 3880 7115 406 90679 331474 3271 5251 8522 407 101032 450150 3644 7131 10775 4080 101032 473827 3644 7506 11150 4081 147365 539677 5316 8549 13864 409 147366 558488 5316 8847 14162 4100 142698 495409 5147 7848 12995 410 142617 454623 5144 7202 12346 383 100765 469596 3635 7439 11073 384 89084 527800 3213 8361 11574 3851 36133 413573 1303 6551 7855 386 36131 398282 1303 6309 7612 387 43822 298125 1581 4723 6303 3870 43880 265969 1583 4213 5796 PIPE GEOMETRY AND LOADING March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-7 Table 3-5 Summary of Robinson Unit 2 Faulted Loads and Stresses for Accumulator Line Loop A Weld Axial Axial Force Moment Moment Stress Total Stress Location Stress Node (lbs) (in-lbs) (psi) (psi)

(psi) 449 144964 410234 5229 6498 11727 4480 144871 386831 5226 6128 11353 448 144025 378465 5195 5995 11190 4470 143969 347372 5193 5503 10696 446 97268 313285 3509 4963 8471 445 97214 379897 3507 6018 9524 444 99502 397398 3589 6295 9884 441 97299 396402 3510 6279 9789 440 97268 398964 3509 6320 9828 439 97247 368204 3508 5833 9340 437 96901 253120 3495 4010 7505 4341 99013 472207 3571 7480 11052 4331 97719 546337 3525 8654 12179 433 97717 517809 3525 8203 11727 43 66394 491441 4079 16269 20348 4321 66394 421965 4079 13969 18048 4500 96324 457613 3474 7249 10723 450 96290 459802 3473 7284 10757 451 106295 428914 3834 6794 10628 452 105022 210080 3788 3328 7116 453 98958 225244 3569 3568 7138 4550 98960 277698 3570 4399 7969 4551 46077 293946 1662 4656 6318 456 46093 296695 1663 4700 6362 457 43540 272316 1570 4314 5884 4570 43599 258522 1573 4095 5668 PIPE GEOMETRY AND LOADING March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-8 Table 3-6 Summary of Robinson Unit 2 Faulted Loads and Stresses for Accumulator Line Loop B Weld Axial Axial Force Moment Moment Stress Total Stress Location Stress Node (lbs) (in-lbs) (psi) (psi)

(psi) 3421 67378 228446 4140 7562 11702 343 77663 294457 4771 9748 14519 344 65497 248602 4024 8230 12254 3440 65463 335167 4022 11095 15117 354 100313 511033 3618 8095 11714 355 109422 601889 3947 9534 13481 357 109282 722323 3942 11442 15384 358 111280 706172 4014 11186 15200 359 112303 965649 4051 15297 19347 360 100411 1077485 3622 17068 20690 361 109285 788847 3942 12496 16438 3620 109288 697501 3942 11049 14991 3621 56596 284563 2041 4508 6549 363 56670 435221 2044 6894 8938 364 61918 903679 2233 14315 16548 3640 61978 829002 2236 13132 15368 3450 99554 421100 3591 6671 10262 348 100896 350121 3639 5546 9186 349 100385 340525 3621 5394 9015 3500 100466 293526 3624 4650 8274 3510 147033 409511 5304 6487 11791 352 147069 468711 5305 7425 12730 353 145972 571906 5265 9059 14325 PIPE GEOMETRY AND LOADING March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-9 Table 3-7 Summary of Robinson Unit 2 Faulted Loads and Stresses for Accumulator Line Loop C Weld Axial Axial Force Moment Moment Stress Total Stress Location Stress Node (lbs) (in-lbs) (psi) (psi)

(psi) 3782 71908 584336 4418 19344 23762 379 71899 595281 4417 19706 24124 380 74965 497991 4606 16486 21091 3811 74483 662826 4576 21942 26518 38 105746 757192 3814 11995 15809 390 96980 278708 3498 4415 7913 391 97597 329686 3520 5222 8743 392 97618 340846 3521 5399 8920 393 98446 364469 3551 5773 9324 395 98532 249644 3554 3955 7509 396 103096 305489 3719 4839 8558 398 101887 257901 3675 4085 7760 399 101887 535666 3675 8485 12160 400 98309 500908 3546 7935 11481 401 100892 409096 3639 6480 10120 403 101512 450033 3662 7129 10790 404 101169 480368 3649 7609 11259 406 100491 530129 3625 8398 12022 407 102086 608055 3682 9632 13314 4080 102090 626090 3682 9918 13600 4081 148633 710349 5361 11253 16614 409 148676 754046 5363 11945 17308 4100 145202 710407 5237 11253 16491 410 145283 665487 5240 10542 15782 383 101925 821895 3676 13019 16696 384 103802 712679 3744 11289 15034 3851 51072 569239 1842 9017 10859 386 51117 520793 1844 8250 10094 387 48608 438125 1753 6940 8694 3870 48551 389013 1751 6162 7914 Notes (for Table 3-2 to 3-7):

a. See Figures 3-1 to 3-3 for weld locations

b. Axial force included pressure PIPE GEOMETRY AND LOADING March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-10 Figure 3-1 H. B. Robinson Unit 2 Loop A Accumulator Line Layout PIPE GEOMETRY AND LOADING March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-11 Figure 3-2 H. B. Robinson Unit 2 Loop B Accumulator Line Layout PIPE GEOMETRY AND LOADING March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-12 Figure 3-3 H. B. Robinson Unit 2 Loop C Accumulator Line Layout PIPE GEOMETRY AND LOADING March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-1 4.0 MATERIAL CHARACTERIZATION 4.1 ACCUMULATOR LINE PIPING AND WELD MATERIALS The material type of the accumulator line for H. B. Robinson Unit 2 is A376 TP316. This is a wrought product of the type used for the piping in several PWR plants. The accumulator line system does not include any cast pipes or cast fittings. The welding processes used are Gas Tungsten Arc Weld (GTAW) and Shielded Metal Arc Weld (SMAW) combination. Figures 3-1 to 3-3 show the schematic layout of the accumulator lines and also identify the weld location by node points.

In the following sections the tensile properties of the materials are presented for use in the Leak-Before-Break analyses.

4.2 TENSILE PROPERTIES The Certified Materials Test Reports (CMTRs) for the H. B. Robinson Unit 2 accumulator lines were used to establish the tensile properties for the leak-before-break analyses. The tensile properties for the pipe material are provided in Table 4-1 for H. B. Robinson Unit 2.

For the A376 TP316 pipe material, the representative properties at operating temperatures of 140°F, 280°F and 547°F are established from the tensile properties at room temperature given in Table 4-1 by utilizing Section II of the 2007 Edition with the 2008 Addenda of the ASME Boiler and Pressure Vessel Code (Reference 4-1). Code tensile properties at the operating temperatures were obtained by interpolating between 70°F, 200°F, 300°F, 500°F and 600°F tensile Code properties. Ratios of the Code tensile properties at the operating temperatures to the corresponding properties were then applied to the room temperature tensile properties obtained from CMTRs (Table 4-1) to obtain the H. B. Robinson Unit 2 accumulator lines specific properties at operating temperatures of 140°F, 280°F and 547°F.

The average and lower bound yield strengths and ultimate strengths for the pipe material are tabulated in Table 4-2. The ASME Code modulus of elasticity values are also given, and Poisson's ratio was taken as 0.3.

4.3 REFERENCE 4-1 ASME Boiler and Pressure Vessel Code, 2007 Edition with the 2008 Addenda,Section II, Part D

- Properties (Customary) Materials.

MATERIAL CHARACTERIZATION March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-2 Table 4-1 Measured Tensile Properties for the Accumulator Line Material A376 TP316 Heat Number Yield Strength (psi) Ultimate Strength (psi)

Room Temp. Room Temp.

DXNS(139284) 36000 78800 DXNS(139284) 36000 78800 80275 35270 76950 48899 33710 81360 48977 31860 78940 48993 33850 79650 49096 32570 79790 80275 32860 78370 49065* 32570 76520 49083* 33990 76810 Note: *Applicable to the 8 inch line Table 4-2 Mechanical Properties for Accumulator Line Material at Operating Temperatures Lower Bound Average Yield Yield Stress Ultimate Strength Material Temperature (°F) Strength (psi) (psi) (psi)

A376 TP316 547 22090 20691 73667 A376 TP316 280 27099 25382 75226 A376 TP316 140 31512 29515 76950 A376 TP316 280* 26513 25947 74806 Modulus of E = 25.618 x 106 psi at 547°F ; E = 27.100 x 106 psi at 280°F; E = 27.869 x 106 psi Elasticity: at 140°F Poissons ratio: 0.3 Note: *Applicable to the 8 inch line MATERIAL CHARACTERIZATION March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

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 critical 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 the H. B. Robinson accumulator lines. Tables 3-1 and 3-5 to 3-7 as well as Figures 3-1 to 3-3 are used for this evaluation.

Critical Locations All the welds in the accumulator line are fabricated using the GTAW/SMAW combination. The pipe material type is A376 TP 316. The governing locations were established on the basis of the pipe geometry, material type, operating temperature, operating pressure, and the highest faulted stresses at the welds. Figures 3-1 to 3-3 show the schematic layout of the accumulator lines. The nozzle to reactor coolant loop branch welds is not shown in these figures, but based on the increased reinforcement thickness at these branch welds, the resulting faulted stress would not be limiting compared to the pipe weld thickness of the critical locations shown.

Critical Locations for the 10 inch Accumulator lines:

The highest faulted stress location is at Node 409 Loop C with temperature 547oF and pressure 2235 psig.

The highest faulted stress location is at Node 360 Loop B with temperature 280oF and pressure 1500 psig.

The highest faulted stress location is at Node 364 Loop B with temperature 140oF and pressure 660 psig.

Therefore, Node 409 Loop C, Node 360 Loop B and Node 364 Loop B are the critical locations.

Critical Location for the 8 inch lines:

Highest faulted stress location is Node 3811 Loop C with temperature 280oF and pressure 1500 psig.

Therefore, Node 3811 Loop C is the critical location.

CRITICAL LOCATION AND EVALUATION CRITERIA March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

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, (L/DH) is greater than [

]a,c,e 6.3 CALCULATION METHOD The basic method used in the leak rate calculations is the method developed by [

]a,c,e 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 accumulator line enthalpy condition and an assumed flow. Once Pc was found for a given mass flow, the [

]a,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 Pf = [ ] a,c,e (6-1) where the friction factor f is determined using the [ ]a,c,e The crack relative roughness, ,

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.

LEAK RATE PREDICTIONS March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-2 That is, for the accumulator line:

Absolute Pressure - 14.7 = [ ]a,c,e (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 accumulator line and the atmosphere, then the procedure is repeated until equation 6-2 is satisfied to within an acceptable tolerance which in turn leads to flow rate value for a given crack size.

6.4 LEAK RATE CALCULATIONS Leak rate calculations were made as a function of crack length at the governing locations previously identified in Section 5.1. The normal operating loads of Tables 3-2 through Table 3-4 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 formulation described above. The average material properties of Section 4.0 (see Table 4-2) were used for these calculations.

For the single phase cases with lower temperature, leakage rate is calculated by the following equation (Reference 6-4) with crack opening area obtained by the method from Reference 6-3.

Q = A(2 gP / k ) 0.5 ft3/sec; (6-3)

Where, P = pressure difference between stagnation and back pressure (lb/ft2), g = acceleration of gravity (ft/sec2), = fluid density at atmospheric pressure (lb/ft3), k = friction loss including passage loss, inlet and outlet of the through-wall crack, A = crack opening area (ft2).

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 H. B. Robinson Unit 2. The flaw sizes so determined are called leakage flaw sizes.

The H. B. Robinson Unit 2 RCS pressure boundary leak detection system meets the intent of Regulatory Guide 1.45. 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.

LEAK RATE PREDICTIONS March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3

6.5 REFERENCES

6-1 [

] a,c,e 6-2 M. M, El-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 II-1, NUREG/CR-3464, September 1983.

6-4 Crane, D. P., Handbook of Hydraulic Resistance Coefficient.

LEAK RATE PREDICTIONS March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-4 Table 6-1 Flaw Sizes Yielding a Leak Rate of 10 gpm for the Accumulator Line Location Leakage Flaw Size (in)

Node 409 Loop C 3.20 Node 360 Loop B 2.91 Node 364 Loop B 4.21 Node 3811 Loop C* 2.52 Note: *Applicable to the 8 inch line LEAK RATE PREDICTIONS March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-5 a,c,e Figure 6-1 Analytical Predictions of Critical Flow Rates of Steam-Water Mixtures LEAK RATE PREDICTIONS March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-6 Figure 6-2 [ ]a,c,e Pressure Ratio as a Function of L/D LEAK RATE PREDICTIONS March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-7 Figure 6-3 Idealized Pressure Drop Profile Through a Postulated Crack LEAK RATE PREDICTIONS March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-1 7.0 FRACTURE MECHANICS EVALUATION 7.1 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:

[ ]a,c,e where:

[

]a,c,e f = 0.5 (y + u) = flow stress, psi

[

]a,c,e 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 March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-2 7.2 RESULTS OF CRACK STABILITY EVALUATION A stability analysis based on limit load was performed for these locations as described in Section 7.1. The weld process types at the critical locations are used as GTAW and SMAW combination. The Z correction factor for SMAW (References 7-2 and 7-3) are as follows:

Z = 1.15 [1.0 + 0.013 (OD-4)] for SMAW where OD is the outer diameter of the pipe in inches.

The Z-factor for the GTAW weld is 1.0. The Z-factor for the SMAW was calculated for the critical locations, using the dimensions given in Table 3-1. The applied faulted loads (Table 3-5 through Table 3-

7) were increased by the Z factor and plots of limit load versus crack length were generated as shown in Figure 7-2 through Figure 7-5. Lower bound material properties were used from Table 4-2. Table 7-1 summarizes the results of the stability analyses based on limit load. The leakage flaw sizes are also presented in the same table.

7.3 REFERENCES

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

7-2 Standard Review Plan; Public Comment Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 52, No. 167/Friday, August 28, 1987/Notices, pp. 32626-32633.

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

FRACTURE MECHANICS EVALUATION March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-3 Table 7-1 Stability Results for the Accumulator Lines Based on Limit Load Location Critical Flaw Size (in) Leakage Flaw Size (in)

Node 409 Loop C 11.53 3.20 Node 360 Loop B 11.17 2.91 Node 364 Loop B 13.19 4.21 Node 3811 Loop C* 7.62 2.52 Note: *Applicable to the 8 inch line FRACTURE MECHANICS EVALUATION March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-4 Figure 7-1 [ ]a,c,e Stress Distribution FRACTURE MECHANICS EVALUATION March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-5 a,c,e OD = 10.75 in. y-min = 20.691 ksi F = 148.676 kips t = 0.896 in. u-min = 73.667 ksi M = 754.046 in-kips A376 TP316 with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-2 Critical Flaw Size Prediction - Node 409 Loop C FRACTURE MECHANICS EVALUATION March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-6 a,c,e OD = 10.75 in. y-min = 25.382 ksi F = 100.411 kips t = 0.896 in. u-min = 75.226 ksi M = 1077.485 in-kips A376 TP316 with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-3 Critical Flaw Size Prediction - Node 360 Loop B FRACTURE MECHANICS EVALUATION March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-7 a,c,e OD = 10.75 in. y-min = 29.515 ksi F = 61.918 kips t = 0.896 in. u-min = 76.950 ksi M = 903.679 in-kips A376 TP316 with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-4 Critical Flaw Size Prediction - Node 364 Loop B FRACTURE MECHANICS EVALUATION March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-8 a,c,e OD = 8.625 in. y-min = 25.947ksi F = 74.483 kips t = 0.650 in. u-min = 74.806 ksi M = 662.826 in-kips A376 TP316 with SMAW Weld Note: OD = outer diameter, t = thickness Figure 7-5 Critical Flaw Size Prediction - Node 3811 Loop C FRACTURE MECHANICS EVALUATION March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-1 8.0 ASSESSMENT OF FATIGUE CRACK GROWTH

8.1 INTRODUCTION

The fatigue crack growth (FCG) analysis is not a requirement for the LBB analysis. The LBB analysis is based on the postulation of through-wall flaw, whereas the FCG analysis is performed based on the surface flaw. However, a fatigue crack growth (FCG) assessment of the H. B. Robinson Unit 2 accumulator lines was determined by comparison with a generic fatigue crack growth analysis of a similar piping system. The details of the generic fatigue crack growth analysis are presented below. By comparing the parameters critical to the fatigue crack growth analysis between H. B. Robinson and the generic analysis, it was concluded that the generic analysis would adequately cover the fatigue crack growth of the H. B. Robinson Unit 2 accumulator lines.

Due to similarities in Westinghouse PWR designs, it was possible to perform a representative fatigue crack growth calculation which would be applicable to H. B. Robinson Unit 2.

8.2 CRITICAL LOCATION FOR FATIGUE CRACK GROWTH ANALYSIS The weld location at the RCL cold leg nozzle to accumulator pipe was determined to be the most critical location for the fatigue crack growth evaluation. The nozzle configuration and weld location are shown in Figure 8-1. The geometry of the accumulator pipe was identical between the H. B. Robinson Unit 2 and the generic model (10). Both analyses used austenitic stainless steel at the critical location.

8.3 DESIGN TRANSIENTS The transient conditions selected for this evaluation are based on conservative estimates of the magnitude and the frequency of the temperature fluctuations documented in various operating plant reports. These are representative of the conditions which are considered to occur during plant operation. The normal operating and upset thermal transients, in accordance with the design specification and the applicable system design criteria document, were considered for this evaluation. Out of these, 15 transients were used in the fatigue crack growth analysis and are listed in Table 8-1. There are some differences between the generic transients used in the fatigue crack growth evaluation and the H. B. Robinson Unit 2 transients but these differences will have insignificant impact on the fatigue crack growth results.

8.4 STRESS ANALYSIS A thermal transient stress analysis was performed for a typical plant similar to H. B. Robinson Unit 2 to obtain the through-wall stress profiles for use in the fatigue crack growth analysis. The generic accumulator line design transients described in Section 8.3 were used.

A simplified analysis method was used to develop conservative maximum and minimum linear through wall stress distributions due to minor thermal transients. In this method, a 1-D computer program was used to perform the thermal analysis to determine the through wall temperature gradients as a function of time. The inside surface stress was calculated by using an equation, which is similar to the transient portion of ASME Section III NB 3600, Equation (11). The effect of discontinuity was included in the analysis by performing a separate 1-D thermal analysis for the pipe and nozzle. The maximum and ASSESSMENT OF FATIGUE CRACK GROWTH March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-2 minimum inside surface stresses were then obtained by searching the inside surface stress values calculated for each time step of the transient solution. The outside surface stresses corresponding to the maximum and minimum inside surface stresses were then calculated by a similar method. The maximum and minimum linear through wall stress distribution for each thermal transient was obtained by joining the corresponding inside and outside surface stresses by a straight line. These two stress profiles are called the maximum and minimum through wall stress distributions respectively, for convenience.

The above methodology was used for minor thermal transients. For severe thermal transients, a 1-D axisymmetric finite element model of the accumulator piping was used to determine the nonlinear stress distributions. The effects of discontinuity at the critical location were included by increasing the magnitude of 1-D nonlinear through-wall stress by 20 percent at the inside one-third thickness of the pipe wall.

The stresses due to the generic pressure and the generic moment loading were superimposed on the through-wall cyclical stresses to obtain the total maximum and minimum stress profile for each transient.

8.5 OBE LOADS The stresses due to OBE loads were neglected in the fatigue crack growth analysis since these loads are not expected to contribute significantly to crack growth due to the small number of cycles.

8.6 TOTAL STRESS FOR FATIGUE CRACK GROWTH The total through-wall stress at a section was obtained by superimposing the generic pressure stress and the generic moment stresses on the thermal transient stresses. Thus, the total stress for fatigue crack growth at any point is given by the following equation:

Stress due to Total Stress Stress due to Moment (DW + Thermal Transient For Fatigue = + +

Internal Pressure Thermal Stress Crack Growth Expansion) 8.7 FATIGUE CRACK GROWTH ANALYSIS The fatigue crack growth analysis was performed to determine the effect of the design thermal transients tabulated in Table 8-1. The analysis was performed for the critical cross-section identified in Figure 8-1.

A range of crack depths was postulated, and each was subjected to the transients in Table 8-1, which included pressure and moment loads.

8.7.1 Analysis Procedure The fatigue crack growth analyses presented herein were conducted in the same manner as suggested by Section XI, Appendix A of the ASME Boiler and Pressure Vessel Code (Reference 8-1). The analysis procedure involves assuming an initial flaw exists at some point and predicting the growth of that flaw due to an imposed series of fluctuating stresses. The growth of a crack per loading cycle is dependent on the range of applied stress intensity factor KI, by the following:

ASSESSMENT OF FATIGUE CRACK GROWTH March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-3 da

= C o K In (8-1) dN where "Co" and the exponent "n" are material properties, and KI is defined as (KI = Kmax - Kmin). For inert environments these material properties are constants, but for some water environments they are dependent on the level of mean stress present during the cycle. This can be accounted for by adjusting the value of "Co" by a function of the ratio of minimum to maximum stress for any given transient. Fatigue crack growth properties of stainless steel in a pressurized water environment have been used in the analysis.

The input required for a fatigue crack growth analysis is basically the information necessary to calculate the parameter KI, which depends on crack size and structure geometry and the range of applied stresses in the area where the crack exists. Once KI is calculated, the growth due to that particular cycle can be calculated by Equation (8-1). This increment of growth is then added to the original crack size, the KI adjusted, and the analysis proceeds to the next transient. The procedure is continued in this manner until all the transients have been analyzed.

The reference crack growth law used for the stainless steel accumulator pipe system was taken from that developed by the Metal Properties Council - Pressure Vessel Research Committee Task Force In Crack Propagation Technology. The reference curve has the equation:

[ (8-2)

]a,c,e This equation appears in Appendix C of ASME Section XI for air environments and its basis is provided in Reference 8-2, and shown in Figure 8-2. For water environments, an environmental factor of [ ]a,c,e was used, based on the crack growth tests in PWR environments reported in Reference 8-3.

8.8 RESULTS Fatigue crack growth analyses were carried out at the critical cross section. Analysis was completed for a range of postulated flaw sizes oriented circumferentially, and the results are presented in Table 8-2. The postulated flaws are assumed to have an aspect ratio of six to one. Even for the largest postulated flaw of 0.25 inch, which is about 28 percent of the wall thickness, the result projects that flaw growth through the wall will not occur during the 40/60 year design life of the plant. Transients and cycles for the H. B.

Robinson Unit 2 plant for 40-year transient set will remain bounding for 60 years (Reference 8-4), the FCG results shown in Table 8-2 are also applicable for the 60 years. Therefore, fatigue crack growth should not be a concern for the H. B. Robinson Unit 2 accumulator line.

ASSESSMENT OF FATIGUE CRACK GROWTH March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-4

8.9 REFERENCES

8-1 ASME Boiler and Pressure Vessel Code Section XI, 2007 Edition with the 2008 Addenda, Rules for Inservice Inspection of Nuclear Power Plant Components.

8-2 James, L. A., and Jones, D. P., Fatigue Crack Growth Correlations for Austenitic Stainless Steel in Air, in Predictive Capabilities in Environmentally Assisted Cracking, ASME publication PVP-99, Dec. 1985.

8-3 Bamford, W. H., Fatigue Crack Growth of Stainless Steel Piping in a Pressurized Water Reactor Environment, Trans ASME, Journal of Pressure Vessel Technology, Feb. 1979. Engineering Development Labs Report HEDL-TME-76-43, May 1976.

8-4 NUREG-1785, Safety Evaluation Report Related to the License Renewal of H. B. Robinson Steam Electric Plant, Unit 2.

ASSESSMENT OF FATIGUE CRACK GROWTH March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-5 Table 8-1 Design Transients Considered for Fatigue Crack Growth Evaluation No. of Trans. No. Description Occurrences 1 Unit Loading 13200 2 Unit Unloading 13200 3 Step Load Increase 2,000 4 Step Load Decrease 2,000 5 Feedwater Cycling 2,000 6 Reactor Trip with Cooldown No Safety Injection 160 7 Inadvertent RCS Depressurization 20 8 Control Rod Drop 80 9 Turbine Roll Test 20 10 Accumulator Actuation, Accident Operation 21 11 Accumulator Actuation, Inadvertent During Cooldown 4 12 High Head Safety Injection 110 13 Steady-State and Random Fluctuations 3.2 x 106 14 RHR Operations During Plant Cooldown 200 15 RHR Operations During Refueling 80 ASSESSMENT OF FATIGUE CRACK GROWTH March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-6 Table 8-2 Accumulator Line Fatigue Crack Growth Results Initial Crack Depth (in) After a,c,e Crack Depth (in) 10 Years 20 Years 30 Years 40/60 Years*

Note:

• Transients and cycles for the H. B. Robinson Unit 2 plant for the 40-year transient set will remain bounding for 60 years, the FCG results shown in Table 8-2 is also applicable for the 60 years.

ASSESSMENT OF FATIGUE CRACK GROWTH March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-7 Figure 8-1 Schematic of 10 Accumulator Line at RCL Cold Leg Nozzle Weld Location ASSESSMENT OF FATIGUE CRACK GROWTH March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-8 Figure 8-2 Reference Crack Growth Curves for Stainless Steel in Air Environments ASSESSMENT OF FATIGUE CRACK GROWTH March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

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 evaluations of Section 7.2 are used in performing the assessment of margins. Margins are shown in Table 9-1. All 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.

ASSESSMENT OF MARGINS March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-2 Table 9-1 Leakage Flaw Sizes, Critical Flaw Sizes and Margins for Accumulator Line Location Critical Flaw Size Leakage Flaw Size Margin (in) (in)

Node 409 Loop C 11.53 3.20 3.6 Node 360 Loop B 11.17 2.91 3.8 Node 364 Loop B 13.19 4.21 3.1 Node 3811 Loop C* 7.62 2.52 3.0 Note: *Applicable to the 8 inch line ASSESSMENT OF MARGINS March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 10-1

10.0 CONCLUSION

S This report justifies the elimination of accumulator line break from the structural design basis for H. B.

Robinson Unit 2 during the 60 years plant life 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.

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

c. The effects of low and high cycle fatigue on the integrity of the accumulator line piping are negligible.

d. Ample margin exists between the leak rate of small stable flaws and the capability of the H. B. Robinson Unit 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.

g. Fatigue crack growth results using the 40 year design transients and cycles (shown to be applicable for 60 years) show that there will be insignificant growth through the wall for the license renewal period (60 year plant life).

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 loading, pipe geometry and material properties considerations, enveloping critical (governing) locations were determined at which leak-before-break crack stability evaluations were made.

Through-wall flaw sizes were postulated 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 accumulator line piping. Therefore, the Leak-Before-Break conditions and margins are satisfied for H. B. Robinson Unit 2 accumulator line piping. It is demonstrated that the dynamic effects of the pipe rupture resulting from postulated breaks in the accumulator line piping need not be considered in the structural design basis of H. B. Robinson Unit 2 for the 60 years.

CONCLUSIONS March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

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

[

] a,c,e APPENDIX A: LIMIT MOMENT March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-2 Figure A-1 Pipe with a Through-Wall Crack in Bending APPENDIX A: LIMIT MOMENT March 2023 WCAP-17779-NP Revision 1

• • • This record was final approved on 3/8/2023, 7:08:28 AM. (This statement was added by the PRIME system upon its validation)

RA-22-0290, Attachment 5 H. B. Robinson Steam Electric Plant, Unit No. 2 ATTACHMENT 5 WESTINGHOUSE AFFIDAVITS AFFIDAVIT CAW-23-008 for WCAP-17776-P, Revision 1 AFFIDAVIT CAW-23-009 for WCAP-17778-P, Revision 1 AFFIDAVIT CAW-23-010 for WCAP-17779-P, Revision 1

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-23-008 Page 1 of 3 Commonwealth of Pennsylvania:

County of Butler:

(1) I, Camille Zozula, Manager/Interim Director, Management Systems & Regulatory Compliance, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).

(2) I am requesting the proprietary portions of WCAP-17776-P Revision 1 be withheld from public disclosure under 10 CFR 2.390.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.

(4) Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.

(ii) The information sought to be withheld is being transmitted to the Commission in confidence and, to Westinghouses knowledge, is not available in public sources.

(iii) Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a document should be withheld from public disclosure. Nevertheless, public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

• • • This record was final approved on 3/13/2023, 8:59:50 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-23-008 Page 2 of 3 (5) Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

(6) The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means of lower-case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower-case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5)(a) through (f) of this Affidavit.

• • • This record was final approved on 3/13/2023, 8:59:50 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-23-008 Page 3 of 3 I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief. I declare under penalty of perjury that the foregoing is true and correct.

Executed on: 3/13/2023 _____________________________

Signed electronically by Camille Zozula

• • • This record was final approved on 3/13/2023, 8:59:50 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-23-009 Page 1 of 3 Commonwealth of Pennsylvania:

County of Butler:

(1) I, Camille Zozula, Manager/Interim Director, Management Systems & Regulatory Compliance, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).

(2) I am requesting the proprietary portions of WCAP-17778-P Revision 1 be withheld from public disclosure under 10 CFR 2.390.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.

(4) Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.

(ii) The information sought to be withheld is being transmitted to the Commission in confidence and, to Westinghouses knowledge, is not available in public sources.

(iii) Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a document should be withheld from public disclosure. Nevertheless, public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

• • • This record was final approved on 3/13/2023, 9:10:00 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-23-009 Page 2 of 3 (5) Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

(6) The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means of lower-case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower-case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5)(a) through (f) of this Affidavit.

• • • This record was final approved on 3/13/2023, 9:10:00 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-23-009 Page 3 of 3 I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief. I declare under penalty of perjury that the foregoing is true and correct.

Executed on: 3/13/2023 _____________________________

Signed electronically by Camille Zozula

• • • This record was final approved on 3/13/2023, 9:10:00 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-23-010 Page 1 of 3 Commonwealth of Pennsylvania:

County of Butler:

(1) I, Camille Zozula, Manager/Interim Director, Management Systems & Regulatory Compliance, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).

(2) I am requesting the proprietary portions of WCAP-17779-P Revision 1 be withheld from public disclosure under 10 CFR 2.390.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.

(4) Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.

(ii) The information sought to be withheld is being transmitted to the Commission in confidence and, to Westinghouses knowledge, is not available in public sources.

(iii) Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a document should be withheld from public disclosure. Nevertheless, public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

• • • This record was final approved on 3/13/2023, 9:22:04 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-23-010 Page 2 of 3 (5) Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

(6) The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means of lower-case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower-case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5)(a) through (f) of this Affidavit.

• • • This record was final approved on 3/13/2023, 9:22:04 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-23-010 Page 3 of 3 I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief. I declare under penalty of perjury that the foregoing is true and correct.

Executed on: 3/13/2023 _____________________________

Signed electronically by Camille Zozula

• • • This record was final approved on 3/13/2023, 9:22:04 AM. (This statement was added by the PRIME system upon its validation)