ML24114A015

Unit 2 – Issuance of Amendment No. 279 Regarding Application of Leak-Before-Break Methodology for Auxiliary Reactor Coolant System Piping

ML24114A015
Person / Time
Site:	Robinson
Issue date:	06/03/2024
From:	Luke Haeg Plant Licensing Branch II
To:	Basta L Duke Energy Progress
Haeg, T
References
EPID L-2023-LLA-0122
Download: ML24114A015 (1)

Text

June 3, 2024

Laura Basta Site Vice President H. B. Robinson Steam Electric Plant Duke Energy Progress, LLC 3581 West Entrance Road, RNPA11 Hartsville, SC 29550

SUBJECT:

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 - ISSUANCE OF AMENDMENT NO. 279 REGARDING APPLICATION OF LEAK-BEFORE-BREAK METHODOLOGY FOR AUXILIARY REACTOR COOLANT SYSTEM PIPING (EPID L-2023-LLA-0122)

Dear Laura Basta:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 279 to Renewed Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant, Unit No. 2 (Robinson). This amendment is in response to your application dated August 30, 2023, as supplemented by letter dated February 2, 2024.

This amendment allows for the use of leak-before-break methodology to eliminate the dynamic effects of postulated pipe ruptures to auxiliary piping systems attached to the reactor coolant system from the Robinson design and licensing bases.

A copy of the safety evaluation is also enclosed. A Notice of issuance will be included in the Commissions monthly Federal Register notice.

L. Basta

If you have any questions, please contact me at (301) 415-0272 or by e-mail at Lucas.Haeg@nrc.gov.

Sincerely,

/RA/

Lucas Haeg, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Docket No. 50-261

Enclosure:

1. Amendment No. 279 to DPR-23

2. Safety Evaluation

cc: Listserv DUKE ENERGY PROGRESS, LLC

DOCKET NO. 50-261

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2

AMENDMENT TO RENEWED FA CILITY OPERATING LICENSE

Amendment No. 279 Renewed License No. DPR-23

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by Duke Energy Progress, LLC (the licensee), dated August 30, 2023, as supplemented by letter dated February 2, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I;

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations;

D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, by Amendment No. 279, Renewed Facility Operating License No. DPR-23 is hereby amended to authorize the change to the Updated Final Safety Analysis Report (UFSAR) as requested by letter dated August 30, 2023, as supplemented by letter dated February 2, 2024, and evaluated in the NRC staffs safety evaluation enclosed within this amendment.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of the date of issuance. The licensee shall submit the update of the UFSAR authorized by this amendment in accordance with 10 CFR 50.71(e).

FOR THE NUCLEAR REGULATORY COMMISSION

David Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Date of Issuance: June 3, 2024

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

RELATED TO AMENDMENT NO. 279

RENEWED FACILITY OPERATING LICENSE NO. DPR-23

DUKE ENERGY PROGRESS, LLC

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2

DOCKET NO. 50-261

1.0 INTRODUCTION

By letter dated August 30, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23242A086), as supplemented by letter dated February 2, 2024 (ML24033A059), Duke Energy Progress, LLC (Duke Energy, the licensee) submitted a license amendment request (LAR) for an amendment for the H.B. Robinson Steam Electric Plant, Unit No. 2 (Robinson). The proposed amendment allow s for use of 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 Robinson design and licensing bases. Specifically, the amendment allows application of LBB methodology to specific sections of the pressurizer surge line, residual heat removal (RHR) piping, and accumulator piping and connected piping.

The supplemental letter dated February 2, 2024, provided additional information that clarified the submittal, did not expand the scope of the submittal as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register (FR) on October 31, 2023 (88 FR 74530).

The subject LAR proposed the elimination of the dynamic effects of postulated pipe ruptures in the subject piping from the design basis of Robinson. The licensee submitted the LAR in accordance with General Design Criterion (GDC) 4, Environmental and dynamic effects design bases, of Appendix A, General Design Criteria for Nuclear Power Plants, to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic licensing of production and utilization facilities. Attachments 6, 7, 8 of the LAR contained proprietary information. As the licensee requested, the NRC has withheld the proprietary information from public disclosure pursuant to 10 CFR 2.390, Public inspections, exemptions, requests for withholding.

Enclosure 2

2.0 REGULATORY EVALUATION

2.1 Proposed Change and Related Systems

The LBB concept is based on calculations and experimental data demonstrating that certain pipe materials have sufficient fracture toughness to prevent a small through-wall crack from propagating rapidly and unstably to catastrophic pipe rupture and to ensure that the probability of pipe rupture is extremely low. In the LBB concept, the RCS leakage detection system will detect small through-wall cracks and the associated leakage promptly so that operators can shut down the reactor and take corrective actions before pipe rupture.

Robinson is a three-loop Westinghouse pressurized water reactor. As noted in the Updated Final Safety Analysis Report (UFSAR) (ML23145A162), Section 3.6.1, Postulated Piping Failures in Fluid Systems Inside Containment, the current design basis includes the application of LBB to the RCS primary loop piping. The LAR would expand the scope of the LBB methodology to include specific portions of piping system attached to the RCS. The auxiliary lines attached to the reactor coolant loops (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 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 an attached 8-inch line connected to the 10-inch accumulator lines, except for the piping upstream of Valves SI-875D, SI-875E, and SI-875F.

The expanded scope of the LBB methodology described in the LAR would eliminate analysis of the dynamic effects of postulated rupture of these specific portions of RCS branch piping. The licensee also stated that materials that are susceptible to primary water stress-corrosion cracking (PWSCC), such as Alloy 600 and Alloy 82/182 weld metal, are not found in the Robinson RHR lines, the 10-inch accumulator lines and the attached 8-inch line connected to 10-inch accumulator lines, or the pressurizer surge line. In addition, the licensee stated there is no cast austenitic stainl ess steel (CASS) material in these lines.

2.2 Current Technical Specifications Requirements

The requirements related to the content of the technical specifications (TSs) are contained in 10 CFR 50.36, Technical Specifications, which requires, in part, that the TS include limiting conditions for operation (LCOs). The following criteria defined by 10 CFR 50.36(c)(2)(ii) are relevant to determining whether capabilities related to reactor coolant pressure boundary (RCPB) leakage detection should be included in the TS LCOs:

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.

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.

Robinson TS LCO 3.4.13 specifies that RCS operation leakage shall be limited to:

a. No pressure boundary leakage;

b. 1 [gallon per minute] gpm unidentified leakage;

c. 10 gpm identified leakage; and

d. 75 gallons per day primary to secondary leakage through any one steam generator.

LCO 3.4.13 provides actions for specified conditions. Surveillance Requirement (SR) 3.4.13.1 and SR 3.4.13.2 provide requirements for RCS operational leakage.

Robinson TS 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.

The Robinson SR for RCS leakage detection instrumentation is provided in SR 3.4.15.1 through SR 13.4.15.5.

The licensee has proposed no changes to the TSs in this LAR.

2.3 Applicable Regulatory Requirements and Guidance

The NRC issued a construction permit to Robinson prior to May 21, 1971; consequently, Robinson is not subject to current GDC requirements in 10 CFR Part 50, Appendix A (SECY-92-223, Resolution of Deviations Identif ied during the Systematic Evaluation Program, dated September 18, 1992, [ML18100B279]). Robinsons conformance with the proposed GDC in existence at the time Robinson 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 Robinsons UFSAR, Section 3.1.2, Evaluation per General Design Criteria. As defined in UFSAR 3.6.2, Postulated Piping Failures in Fluid Systems Outside of Containment, high energy piping s ystem are those whose service temperature exceeds 200 degrees Fahrenheit (°F) and whose design pressure exceeds 275 pounds per square inch gauge (psig). The licensee stated that the 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 current version of GDC 4.

The regulations in GDC 4 state, in part, t hat structures, systems, and components (SSCs) important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with postulated accidents, including loss-of-coolant accidents (LOCA). These SSCs 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 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.

The licensee provided information that demonstrated compliance with the intent of 10 CFR Part 50, Appendix A, GDC 14, 30, and 31, provided below, which are similar to Atomic Energy Commission (AEC) Criteria 9, 16, and 34 as described in Robinsons UFSAR Sections 3.1.2.9, 3.1.2.16, and 3.1.2.34:

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.

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.

The licensee stated that the piping in the scope of this LAR is designated as Class 1 (Class 1 is defined herein as piping that is within the American Society of Mechanical Engineers (ASME)

Section XI, Subsection, IWB inspection boundary) RCPB piping whose materials, design, 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. The LAR also addressed the capability to detect and respond to piping system leakage prior to a potential flaw reaching critical size.

Therefore, the licensee stated that the request is consistent with GDC 14, 30, and 31.

The licensee stated that the RCPB is designed, fabricated, and constructed to have an exceedingly low probability of gross rupture or significant uncontrolled leakage throughout its design lifetime. RCPB 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 RCS leakage limits ensure the RCPB will retain adequate structural and leakage integrity during normal operating, transient, and postulated accident conditions.

The licensee stated that the guidance provided in NRC Standard Review Plan (SRP)

(NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear

Power Plants: LWR [light-water reactor] Edition), Section 3.6.3, Leak-Before-Break Evaluation Procedures, Revision 1 for review of the LBB application was followed, including guidance for determining an acceptable leakage crack and the RCS leakage detection sensitivity based on the fracture mechanics analysis.

In addition, the RCS leakage detection system should be able to detect a certain leak rate with margins, compared to the leak rate from the leakage crack size of the subject piping in accordance with SRP 3.6.3, Revision 1.

SRP 3.6.3 states that the leakage detection systems 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 also specifies that the predicted leakage from the postulated leakage crack should be a factor of 10 times greater than the minimum leakage the detection system is c apable of sensing unless the licensee provides justification accounting for the effects of uncertainties in the leakage measurement. The 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, and meet a leak detection capability 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 />.

3.0 TECHNICAL EVALUATION

The NRC staff followed the guidance in SRP, Section 3.6.3, Revision 1 to review the licensees LBB analysis. The NRC staff reviewed the segments of piping lines for the LBB analysis (i.e.,

scope of LBB analysis), as discussed in Section 3.1 of this safety evaluation (SE). The NRC staff also evaluated whether the subject piping lines satisfy the screening criteria for various degradation mechanisms, as documented in Section 3.2 of this SE. The NRC staff further reviewed the fracture mechanics analysis of the s ubject piping, as discussed in Section 3.3 of this SE. In addition, the NRC staff evaluated the capability of the RCS leakage detection system, as documented in Section 3.4 of this SE.

As part of the LAR submittal, the licensee provided the following non-proprietary reports describing the LBB analysis:

WCAP-17776, Revision 1, Technical Justificat ion for Eliminating Pressurizer Surge Line Rupture as the Structural Design Basis for H. B. Robinson, Unit 2. (Attachment 2)

WCAP-17778, Revision 1, Technical Justific ation for Eliminating RHR Line Rupture as the Structural Design Basis for H. B. Robinson, Unit 2. (Attachment 3)

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

3.1 Scope of Leak-Before-Break Analysis

As described in Attachment 1 of the LAR, the licensees analysis is applied to specific portions of the accumulator, RHR, and pressurizer surge lines that are connected to the RCS. The licensee described the subject piping and related systems as follows:

Accumulator Lines

The 10-inch accumulator lines (from the cold legs Loop A, Loop B, and Loop C) and the attached 8-inch line connected to the 10-inch accumulator lines except for the piping upstream of Valves SI-875D, SI-875E, and SI-875F), 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.

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 psig and at temperatures less than 350°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. The licensee stated that 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.

Pressurizer Surge Lines

The 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.

The NRC staff finds that the LAR, as supplemented, has clearly identified the specific portions of the accumulator, RHR and pressurizer surge line piping that are subject to the licensees LBB analysis and, therefore, the scope of the LBB analysis is identified appropriately.

3.2 Screening Based on Applicable Degradation Mechanisms

The SRP, Section 3.6.3, Subsection III, Review Procedures, specifies that active degradation should not be present in system being evaluated for LBB (e.g., degradation due to stress-corrosion cracking (SCC), fatigue, water hammer, corrosion, wall thinning, creep, or brittle cleavage-type failure).

In the following sections, the NRC staff evaluates the LBB analysis in accordance with the degradation screening criteria of SRP Section 3.6.3, Subsection III.

3.2.1 Stress-Corrosion Cracking

As discussed in Section 2.1, Stress Corrosion Cracking, of WCAP-17776, Revision 1, WCAP-17778, Revision 1, and WCAP-17779, Revision 1, the following elements or contaminants in a reactor coolant environment are known to increase the susceptibility of austenitic stainless steel to SCC: oxygen, fluor ides, chlorides, and reduced forms of sulfur (e.g.,

sulfides and sulfites). The licensee stated that strict pipe cleaning standards were implemented prior to operation. Prior to being put into service, the piping was cleaned both internally and externally. During flushes and preoperational testing, water chemistry was controlled in

accordance with written specifications. During plant operation, the reactor coolant water chemistry is monitored and maintained within very specific limits, and the contaminant concentrations are kept below the thresholds known to be conducive to SCC with the major water chemistry control standards being included in the plant operating procedures as a condition for plant operation. Therefore, the likelihood of SCC is minimized during plant operation.

The Westinghouse RCS primary loop piping and connected Class 1 piping have an operating history that demonstrates the inherent operating stability characteristics of the design. The operating experience also confirms that the RCS piping is resistant to intergranular and/or PWSCC. Operating experience has shown that PWSCC has occurred in nickel-based Alloy 82/182 dissimilar metal butt welds in pressurized-water reactor (PWR) coolant environment. The licensee confirmed that these materials, which are susceptible to PWSCC, are not used in the accumulator, RHR and pressurizer surge lines of Robinson. Therefore, the NRC staff finds that SCC is not an active degradation mechanism for the subject piping based on the discussion above.

3.2.2 Fatigue

The licensee evaluated the piping susceptibility to low-cycle and high-cycle fatigue and the potential impact of fatigue on the piping integrity. In addition, the licensee provided the evaluation of thermal stratification that could cause fatigue. The sections below describe the NRC staffs evaluation on these matters.

3.2.2.1 Low-Cycle and High-Cycle Fatigue

As discussed in Section 2.3, Low Cycle and High Cycle Fatigue, of WCAP-17776-P/NP, Revision 1, WCAP-17778-P/NP, Revision 1, and WCAP-17779-P/NP, Revision 1, the licensee stated that pump vibrations during operation would result in high cycle fatigue loads in the piping system. During operation, the licensee stated that an alarm signals the exceedance of the reactor coolant pump shaft vibration limits. The licensee stated that field vibration measurements have been made on the reactor cool ant loop piping in a number of plants during hot functional testing. Stresses in the elbow below the reactor coolant pump have been found to be very small, between 2 and 3 thousand pounds per square inch (ksi) at the highest. The licensee stated that field measurements on typi cal 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.

Section 8.0, Assessment of Fatigue Crack Growth, of WCAP-17776-P/NP, Revision 1, WCAP-17778-P/NP, Revision 1, and WCAP-17779-P/NP, Revision 1 addresses the fatigue crack growth (FCG) analyses of postulated circumferential inner-surface cracks for the subject piping. The FCG assessment of the Robinson subject piping systems was determined by comparison with a generic FCG analysis of a similar piping system. The licensee stated that comparing the parameters critical to the FCG analysis between Robinson and the similar pressure line analysis, the licensee concluded that the similar analysis would adequately cover the FCG assessment of the Robinson Unit 2 pressure lines. The licensee also stated that due to the similarities in the Westinghouse PWR designs, it was possible to perform a representative FCG assessment which would be applicable to Robinson.

The licensee stated that the methodology consists of first obtaining the local and structural transient stress analyses results and then superimposing the local and structural transient stresses. 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 FCG analysis was performed at the hot leg nozzle location. The licensee stated that five through-wall stress cuts were analyzed at these locations.

The results from WCAP-17776, Revision 1 showed the FCG stress cuts with the initial flaws oriented circumferentially. The licensee stated the flaws were assumed to be semi-elliptical with an aspect ratio of six to one. The initial flaw size s were assumed to be 10 percent of the nominal wall thickness. For the initial flaw size of 0.14 inch, the result projects that the maximum final flaw size after 40/60 years is about 14.8 percent 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. The licensee concluded that the transients and cycles for Robinson plant for a 40-year transient set will remain bounding for 60 years.

As stated in WCAP-17778, Revision 1, the weld location at the RCL hot leg nozzle to RHR line was determined to be the most critical location for the FCG evaluation. The licensee stated that the FCG assessment of the RHR line was det ermined by comparison with a generic FCG analysis of a similar piping system. The geometry of the pipe was identical between the Robinson and the generic model.

The licensee stated that the FCG were carried out at the critical cross-section. 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. The licensee concluded that the transients and cycles for the Robinson Unit 2 plan for the 40-year transient set will remain bounding for 60 years. Based on the results, the licensee concluded that the FCG should not be a concern for the Robinson RHR line.

As stated in WCAP-17779, Revision 1, the weld location at the RCL cold leg nozzle to accumulator pipe was determined to be the most critical location for the FCG evaluation. The licensee stated that the FCG assessment of the accumulator lines was determined by comparison with a generic FCG analysis of a similar piping system. The geometry of the pipe was identical between the Robinson and the generic model.

The licensee stated that the FCG analysis were carried out at the critical cross section. 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. The licensee stated that the transients and cycles for the Robinson for the 40-year transient set will remain bounding for 60 years. Based on the results, the licensee concluded the FCG should not be a concern for the Robinson accumulator line.

The staff reviewed the licensees FCG analyses as described above. Based on its review, the staff finds that the FCG analyses for the subject piping support the licensees conclusions that the potential FCGs are insignificant and do not affect crack stability or LBB applicability, and that the high-cycle fatigue due to pump vibration is in significant. Therefore, the NRC staff finds that low-cycle and high-cycle fatigue is not a potential source of pipe rupture for the subject piping.

3.2.2.2 Thermal Stratification

Section 2.4, Summary Evaluation of Surge Line for Potential Degradation During Service, of WCAP-17776, Revision 1, and Section 2.4, Other Possible Degradation During Service of the RHR Lines, of WCAP-17778, Revision 1, states that 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. The licensee stated that therefore, changes in the stratification state result in thermal cycling, which can cause fatigue degradation. The licensee stated that the effects of thermal stratification have been evaluated for the RHR piping due to 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), dated October 1, 1991, and is not a concern. For the pressurizer surge lines and the accumulator lines, the licensee stated that the thermal stratification loads were used for the LBB analysis of the LAR.

The NRC staff finds that potential cracking due to thermal stratification is not an active degradation mechanism because the licensees inservice inspections have confirmed the absence of cracks due to thermal stratification.

3.2.2.3 Brittle Fracture and Cleavage-Type Failure

The licensee stated that there are no damage mechanisms that can lead to reduction of fracture toughness of the piping materials. The operating temperatures of the subject lines are higher than 70°F and therefore, brittle fracture is not a concern. The licensee further stated that brittle cleavage type failures are not a concern based on the operating temperatures, there is no CASS product forms, and the radiation levels are low.

The NRC staff finds that brittle fracture or cl eavage-type failure is not an active degradation mechanism for the subject piping because the operating temperature of the subject piping are not within the temperature range that would cause brittle fracture.

3.2.2.4 Creep

The licensee stated that the maximum operating temperature of the subject piping lines is about 650°F, which is below the temperature which creep damage would occur in stainless steel. The NRC staff recognizes that the operating temperatur e of the subject piping is well below 800°F, the temperature that would cause creep degradation in stainless steel material. Therefore, the NRC staff finds that creep is not an active degradation mechanism for the subject piping.

3.2.2.5 Wall Thinning

The licensee stated that the wall thinning by erosion and erosion-corrosion should not occur in the accumulator, the RHR, and the pressurizer surge line because the austenitic stainless steel material used for the piping is highly resistant to these degradation mechanisms. Additionally the licensee stated that the piping systems are typically low velocity and therefore wall thinning should not occur or be a concern. The licensee explained that as discussed in NUREG-0691, Investigation and Evaluation of Cracking Incidents in Piping in Pressurized Water Reactors, September 1980 (ML070040195; not publicly available), a study on pipe cracking in PWR piping reported only two incidents of wall thinning in stainless steel pipe, and those were not in the subject piping.

The NRC staff finds that wall thinning is not an active degradation mechanism for the subject piping because of the low velocity flow in the subject piping and the piping fabrication material that is resistant to wall thinning.

3.2.2.6 Water Hammer

The licensee indicated that there is a low potential for water hammer in the RCS and connecting auxiliary piping systems since they are designed and operated to preclude the voiding condition in the water-filled lines. The licensee explained that the design requirements for the subject piping lines are conservative relative to both the number of transients and their severity. In addition, the system design also considered relief valve actuation and the associated hydraulic transients following valve opening. 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, and pressure is also controlled within a narrow range for steady-state conditions by the pressurizer heaters and pressurizer spray. Accordingly, the flow characteristics of the RCS remain constant during a fuel cycle, which minimizes the potential for water hammer.

Additionally, the licensee explained that Westinghouse performed instrumentation and monitoring activities to verify the flow and vibration characteristics of the RCS and the connected auxiliary lines. The licensees preoperational testing and operating experience have verified that the Westinghouse approach is effective.

The NRC staff finds that based on the activi ties performed by the licensee on the instrumentation monitoring to verify flow and vibration characteristics of the system, no significant water hammer can occur.

3.2.3 Conclusion on the Screening Based on Applicable Degradation Mechanisms

On the basis of the above evaluation, the NRC staff finds that the accumulator, the RHR, and the pressurizer surge lines applicable to this LAR are not subject to any active degradation that can be a potential source of pipe rupture, consistent with SRP Section 3.6.3. The licensee also reviewed inservice inspection results and confirmed that there are no known relevant indications in the subject piping. Based on the absence of active degradation that can cause pipe rupture in the subject piping, the NRC staff concludes that the LBB analysis meets the acceptance criteria for a fracture mechanics analysis to be further used for the determination of crack stability in accordance with SRP Section 3.6.3.

3.3 Fracture Mechanics Analyses

3.3.1 Materials Properties

SRP Section 3.6.3, Subsection III.11 specifies that the LBB analysis should identify the types of materials and materials specifications used for, in part, base metal and weldments. This section also specifies that the licensee should provide the material properties, including toughness, tensile data, and long-term effects such as thermal aging.

The licensee reported that the subject piping is fabricated with stainless steel materials.

Specifically, the materials for fabrication of the subject piping as stated in WCAP-17776, Revision 1, WCAP-17778, Revision 1, and WCAP-17779, Revision 1 is a wrought product fabricated from A376 TP316. The subject piping does not include any cast pipes or cast fittings.

The licensee confirmed that the subject piping lines are not susceptible to fracture toughness degradation due to thermal aging. The licensee further indicated that the weld processes used in the subject piping are gas tungsten arc welding (GTAW) and shielded metal arc welding (SMAW) combination.

The licensees LBB analysis used ASME Code mechanical properties to establish the tensile properties of the piping materials, consistent with ASME Code,Section II, 2007 Edition with the 2008 Addenda. The material modulus of elasticity was also interpolated from ASME Code values for the operating temperatures for the subject piping, and Poissons ratio was taken as 0.3. Section 4.0, Material Characterization, of WCAP-17776, Revision 1, WCAP-17778, Revision 1, and WCAP-17779, Revision 1 provides the yield strengths, ultimate strengths, and elastic moduli for the materials of each subject piping system.

The NRC staff finds the licensees approach to be acceptable because (1) the licensee adequately used the code property data to estimate the operating temperature properties of the subject piping, and (2) the licensee accounted for the temperature effects on material properties by interpolating material property data at different temperatures.

3.3.2 Load Combinations

SRP Section 3.6.3, Subsection III.1 specifies that the LBB analysis should use design-basis loads that are based on the as-built piping configuration as opposed to the design configuration.

As described in WCAP-17776, Revision 1, WCAP-17778, Revision 1, and WCAP-17779, Revision 1 for the subject piping lines, the licensee stated that the LBB analysis used the as-built piping configurations and the associated piping loads.

SRP Section 3.6.3, Subsection III.11.C.v addresses the level of conservatism that needs to be applied to the load calculation in the crack stability analysis. The SRP indicates that if the deadweight, thermal expansion, pressure, safe-shutdown earthquake, and seismic anchor motion loads are combined based on the individual absolute values of the loads (i.e., absolute sum load combination method) no additional margin may be applied to the limiting load calculation.

In the crack stability analysis, the licensee used the absolute sum load combination method in accordance with the guidance in the SRP. The licensee also considered the bending and torsional moments to obtain the limiting to tal applied moment. In addition, the licensee calculated the applied moment based on the square root of the sum of squares of the bending and torsional moments, which is consistent with SRP Section 3.6.3, Subsection III.11.C.v.

The NRC staff finds that the licensees load combi nations in the crack stability analysis of the subject piping are acceptable because (1) the licensee used the absolute sum load combination method, (2) the licensee appropriately considered the upper bound loads under the faulted conditions in the limiting load combinations (including deadweight, thermal expansion, safe shutdown earthquake, and seismic anchor motion loads), (3) the calculations of the total moment considered the bending and torsional moments, and (4) these methods are consistent with the guidance in SRP Section 3.6.3, Subsections III.1 and III.11.C.v.

3.3.3 Leakage Crack Size Calculation

SRP Section 3.6.3, Subsection III.11.C.iii specifies that the estimated leak rate from the leakage crack during normal operation should be 10 times greater than the minimum leak rate that the RCS leakage detection system can detect.

The licensee stated that Sections 6.4 and 7.2 of WCAP-17776, Revision 1, WCAP-17778, Revision 1, and WCAP-17779, Revision 1 provide the leak rates and corresponding stability evaluations that were used in performing the assessment of margins. The licensee stated that the RCS pressure boundary leakage detection system can detect a leak rate of 1 gpm.

Accordingly, the LBB analysis for the subject piping lines uses a leakage rate of 10 gpm, which is 10 times the leak rate detection capability. In response to RAI-1 that requested clarifications for the methodology for the calculation of the leakage flaw size, the licensee clarified the approach for the calculation of the leakage flaw size and stated that the approach was consistent with WCAP-9558, Revision 2, Mechanistic Fracture Evaluation of Reactor Coolant Piping Containing a Postulated Circumferential Through-Wall Crack, May 1981, which has been used in previous applications and accepted by the NRC staff since 1981.

Based on its review of the licensees leakage crack size calculation method, the NRC staff finds that the estimated size (length) of the leakage crack is large enough that leakage from the flaw during normal operation would be 10 times greater than the minimum leakage that the RCS detection system is capable of detecting. Theref ore, the NRC staff finds that the licensees approach is consistent with the guidance in SRP Section 3.6.3, Subsection III.11.C.iii. The NRC staff also noted that the licensees methods used to estimate the leakage rates and leakage crack sizes for the given leakage detection limit are consistent with those used in existing LBB analysis for the primary coolant loops, which cons titute the current licensing basis of Robinson (see WCAP-17776, Revision 1, Technical Justification for Eliminating Pressurizer Surge Line Rupture as the Structural Design Basis for H. B. Robinson, Units 2, dated March 2023, WCAP-17778, Revision 1, Technical Justification for Eliminating RHR Line Rupture as the Structural Design Basis for H. B. Robinson, Unit 2, dated March 2023, WCAP-17779, Revision 1, Technical Justification for Eliminating Accumulator Line Rupture as the Structural Design Basis for H. B. Robinson, Unit 2, dated March 2023).

3.3.4 Crack Stability Analysis

SRP 3.6.3, Subsection III.11.C describes how the critical crack sizes should be calculated. SRP 3.6.3, Subsection III.11.C.iv specifies that a crack stability analysis should be performed to demonstrate that the leakage crack size will not become unstable by comparing the leakage crack size to the critical crack size. Specifically, a margin of 2 should exist between the leakage crack size and the critical crack size.

The licensee derived the critical crack sizes at t he critical locations in accordance with the guidance in SRP Section 3.6.3. The limiting locations were established on the basis of the pipe geometry, welding process, material type, oper ating temperature, operating pressure, and the highest faulted stresses at the welds. Therefore, the critical locations are established based on the bounding stresses and piping material properties.

As part of the crack stability analysis, the licensee used the limit load method to predict the critical crack size for the critical locations in the subject piping. The failure criteria was obtained by requiring equilibrium of the section containing a through-wall circumferential crack. The applied loads are calculated in consideration of internal pressure, axial force, and imposed

moments. The limiting moment for the analyzed pipe is calculated based on the flow stress, axial force (including internal pressure), pipe dimensions, and crack size and configuration. For the limit load method, the licensee also multiplied the pipe loads by the Z-factor, considering the welding processes (i.e., SMAW and GTAW processes). The licensee stated that it derived the Z factors for SMAW and GTAW in accordance with SRP Section 3.6.3.

For the piping segments identified in WCAP-17776, Revision 1, WCAP-17778, Revision 1, and WCAP-17779, Revision 1, the licensee performed a J-integral fracture mechanics analysis and demonstrated that the applied J-integral value is less than the fracture toughness of the piping base material. Accordingly, the postulated through-wall crack, which has a crack size twice the leakage crack size, would not result in crack tip extension.

The NRC staff finds the crack stability analysis to be acceptable because (1) a safety margin of 2 is demonstrated between the leakage crack size and the critical crack size, and (2) the licensees methodology is consistent with the guidance in SRP Section 3.6.3.

3.3.5 NRC Staffs Confirmatory Analysis

The NRC staff performed a confirmatory analysis to check the adequacy of the licensees analysis results. The confirmatory analysis used t he PICEP computer code described in Electric Power Research Institute NP-3596-SR, PICEP: Pipe Crack Evaluation Program, Revision 1, issued December 1987. The analysis evaluated the critical location Node 323 identified in Table 7.1, Stability Results for the RHR line Based on Limit Load of WCAP-17778.

The PICEP code conducted a limit load analysis with Z factors applied to the loads, which are used in the licensees evaluation. In the analysis, the NRC staff confirmed that the critical crack sizes determined by the licensee were in agreement with those calculated by the PICEP code.

The NRC staff also used the PICEP code to estima te the leakage crack sizes and to check the margin between the leakage crack size and the cr itical crack size. The confirmatory analysis used the same leakage rates as those used in the licensees analysis (10 gpm). In the analysis, the NRC staff confirmed that a margin of 2 is met between the leakage crack size and the critical crack size.

3.4 Reactor Coolant Pressure Boundary Leakage Detection System Capability

The licensee stated in the LAR:

As discussed in [GL 84-04, Safety Evaluation of Westinghouse Topical Report Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops, February 1, 1984] 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 through wall flaw.

As stated in Robinson UFSAR, Section 5.2.5, Detection of Leakage Through the Reactor Coolant Pressure Boundary, the leak detection systems associated with the RCS consist of the following: (a) two radiation sensitive instruments provide the capability for detection of leakage from the RCS, (b) a humidity detector that provides a means of measuring overall leakage from all water and steam systems within the containment, which is less sensitive than the radiation

monitors, and therefore, used as a backup to the radiation monitoring methods, and (c) 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.

Robinson UFSAR, Section 5.2.5 states, in part, that 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 />... The licensee states in the LAR that this was implemented when Robinson incorporated WCAP-9558 [Mechanistic Fracture Evaluation of Reactor Coolant Piping Containing a Postulated Circumferential Through-Wall Crack, May 1981] and WCAP-9787 [Tensile and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation, May 1981] into the Current Licensing Basis (CLB) for LBB of main coolant piping. The licensee noted that it is an exception to the guidance in RG 1.45, which provides that detection of 1 gpm leakage should be within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and that the capability of the Robinson leak detection systems to detect 1 gpm leakage should be 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.

The Robinson license renewal application, dated June 14, 2002, stated the following:

WCAP-15628 [Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the H.B. Robinson Unit 2 Nuclear Power Plant for the License Renewal Program, July 2001 (ML031320381)] is a new leak-before-break (LBB) calculation applicable to ROBINSON large bore Reactor Coolant System (RCS) piping and components... The new analysis meets the requirements for LBB required by 10 CFR 50, Appendix A, General Design Criterion 4, and uses the recommendations and criteria from the NRC Standard Review Plan for LBB evaluations.

In its Safety Evaluation Report for Robinson license renewal application (NUREG-1785 (ML041810563)) the NRC staff stated that the LBB application for the primary loop piping and components is acceptable for the period of extended operation.

As such, the staff has previously reviewed and accepted the licensees LBB application for primary loop piping and components, and the capability of the Robinson leak detection systems to detect 1 gpm leakage 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.

The licensee proposes to amend application of the LBB methodology to auxiliary piping systems attached to the RCS from the Robinson design and licensing basis, which does not involve changes to the Robinson leak detection syste m. Therefore, the NRC staff finds that the Robinson leak detection system is acceptable with the proposed changes.

3.5 Technical Evaluation Summary

On the basis of its review of the LAR, as supplemented, the NRC staff finds that, for the subject 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 nozzle surge nozzle safe end to the pressurizer surge nozzle), the RHR lines, limited to the high energy Class 1 portions 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, SI875E, and SI-875F, the licensee has demonstrated that (1) screening criteria of SRP Section 3.6.3 are satisfied in the evaluation of applicable degradation mechanisms, (2) a margin

of 10 exists between the calculated leak rate from the postulated leakage crack sizes and the RCS leakage detection system capability, (3) a margin of 2 exists between the leakage crack sizes and the critical crack sizes, (4) the critical cracks were calculated conservatively in consideration of the bounding material properties and load conditions in the limit load and fracture mechanics analyses, (5) the licensees LBB methods are consistent with the guidance in SRP Section 3.6.3, and (6) the potential FCG in the subject piping is insignificant and does not affect the crack stability and the validity of the LBB analysis.

Accordingly, the NRC staff finds that the licen sees analysis has demonstrated that the subject piping has an extremely low probability of rupture. As described in Section 3.1 of this SE, the

LBB application pertains to specific RCS piping segments associated with the accumulator, the RHR and the pressurizer surge lines.

Based on the evaluation above, the NRC staff has concluded that the Robinson LBB leakage detection capability remains consistent with (1) the requirement of 10 CFR 50.36(c)(2)(i), (2) the intent of 10 CFR Part 50, Appendix A, GDC 30, and (3) the intent of RG 1.45, Revision 0. The NRC staff also concludes that the use of the current TS leakage limit for the proposed LBB application is acceptable because (1) the TS requires that, if the leakage rate exceeds 1 gpm and the source of leakage cannot be identified in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the reactor must be brought to shutdown, and (2) the leakage detection system is capable of detecting 1 gpm, consistent with the leakage rate used in the LBB analysis.

Pursuant to 10 CFR Part 50, Appendix A, GDC 4, the NRC staff concludes that the licensee is permitted to exclude consideration of the dynamic effects associated with the postulated rupture of the subject accumulator, the RHR and the pressurizer surge lines piping from current licensing basis at Robinson.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the State of South Carolina official was notified of the proposed issuance of the amendment on April 22, 2024. The State of South Carolina official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes the requirements with respect to installation or use of a facilitys components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration in the Federal Register on October 31, 2023 (88 FR 74530),

and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need to be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, 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) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: E. Reichelt, NRR H. Wagage, NRR

Date: June 3, 2024

ML24114A015 OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DNRL/NPHP/BC NRR/DSS/SCPB/BC NAME LHaeg ABaxter MMitchell MValentin DATE 4/23/2024 4/25/2024 3/28/2024 5/1/2024 OFFICE OGC - NLO NRR/DORL/LPL2-NRR/DORL/LPL2-2/BC 2/PM NAME AGhosh DWrona LHaeg DATE 5/24/2024 5/31/2024 6/3/2024