ML20153A014

Relief from the Requirements of the American Society of Mechanical Engineers Code Regarding Request VRR4 (EPID L-2020-LLR-0052 (COVID-19))

ML20153A014
Person / Time
Site:	Beaver Valley
Issue date:	07/16/2020
From:	James Danna Plant Licensing Branch 1
To:	Penfield R Energy Harbor Nuclear Corp
Tobin J
References
EPID L-2020-LLR-0052 [COVID 19]
Download: ML20153A014 (12)

Text

July 16, 2020 Mr. Rod Penfield Site Vice President Energy Harbor Nuclear Corp.

Beaver Valley Power Station Mail Stop A-BV-SEB1 P.O. Box 4, Route 168 Shippingport, PA 15077

SUBJECT:

BEAVER VALLEY POWER STATION, UNITS 1 AND 2 - RELIEF FROM THE REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS CODE REGARDING RELIEF REQUEST VRR4 (EPID L-2020-LLR-0052 [COVID-19])

Dear Mr. Penfield:

By letter dated April 2, 2020, as supplemented by letter dated April 6, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML20093C288 and ML20097C373, respectively), Energy Harbor Nuclear Corp. (the licensee) proposed an alternative to certain inservice testing (IST) program requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code),Section XI, 2004 Edition through 2006 Addenda, for Beaver Valley Power Station (Beaver Valley), Units 1 and 2, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 55a.

Specifically, pursuant to 10 CFR 50.55a(z)(2), the licensee requested to extend the performance of IST program leakage testing for 72 specific isolation valves at Beaver Valley, Unit 1, during the fifth 10-year IST program interval, and at Beaver Valley, Unit 2, during the fourth 10-year IST program interval. In its submittal, the licensee requested the use of proposed alternative VRR4 for the 72 specified isolation valves at Beaver Valley, Units 1 and 2, on the basis that compliance with the ASME OM Code requirements for leakage testing would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

On April 8, 2020, the U.S. Nuclear Regulatory Commission (NRC) provided a verbal authorization (ADAMS Accession No. ML20099K215) of proposed alternative VRR4 for the one-time extension of the leakage testing interval for certain isolation valves at Beaver Valley, Unit 2, needed for the 2020 spring outage.

The NRC staff has concluded that the proposed alternative in VRR4 provides an acceptable level of quality and safety. The NRC staff finds that complying with the specified requirements in the ASME OM Code for testing of the specified isolation valves at Beaver Valley, Unit 1 and Unit 2, would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed the regulatory requirements set forth in 10 CFR 50.55a(z)(2).

R. Penfield All other ASME OM Code,Section XI requirements for which relief was not specifically requested and approved remain applicable.

If you have any questions, please contact the Beaver Valley Project Manager, Jennifer Tobin, at 301-415-2328 or Jennifer.Tobin@nrc.gov.

Sincerely, Digitally signed by James James G. G. Danna Date: 2020.07.16 Danna 08:04:10 -04'00' James G. Danna, Chief Plant Licensing Branch 1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-334 and 50-412

Enclosure:

Safety Evaluation cc: Listserv

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REGARDING RELIEF REQUEST VRR4 FOR THE TESTING OF CERTAIN VALVES ENERGY HARBOR NUCLEAR CORP.

ENERGY HARBOR NUCLEAR GENERATION LLC BEAVER VALLEY POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-334 AND 50-412

1.0 INTRODUCTION

By letter dated April 2, 2020, as supplemented on April 6, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML20093C288 and ML20097C373, respectively, Energy Harbor Nuclear Corp. (the licensee) proposed an alternative to specific inservice testing (IST) program requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code),Section XI, 2004 Edition through 2006 Addenda, for Beaver Valley Power Station, Unit 1 (BVPS-1), and Unit 2 (BVPS-2), pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 55a (10 CFR 50.55a).

Specifically, the licensee submitted Relief Request VRR4, Revision 0, Containment Isolation Valve Test Frequency, on April 2, 2020, requesting authorization by the U.S. Nuclear Regulatory Commission (NRC) to extend the performance of IST program leakage testing for 72 specific isolation valves at BVPS-1 during the fifth 10-year IST program interval and BVPS-2 during the fourth 10-year IST program interval. In its submittal, the licensee requested use of proposed alternative VRR4 for the 72 specified isolation valves at BVPS-1 and BVPS-2 on the basis that compliance with the ASME OM Code requirements for leakage testing would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety pursuant to 10 CFR 50.55a(z)(2).

On April 8, 2020, the NRC provided a verbal authorization (ADAMS Accession No. ML20099K215) of proposed alternative VRR4 for the one-time extension of the leakage testing interval for the isolation valves included in the April 6, 2020, supplement for the remainder of the fourth IST interval at BVPS-2, as needed for the 2020 spring outage.

Enclosure

2.0 REGULATORY EVALUATION

The NRC regulations in 10 CFR 50.55a(f)(4), Inservice testing standards requirement for operating plants, state, in part, that throughout the service life of a boiling or pressurized water-cooled nuclear power facility, pumps and valves that are within the scope of the ASME OM Code must meet the IST requirements (except design and access provisions) set forth in the ASME OM Code and addenda that become effective subsequent to editions and addenda specified in 10 CFR 50.55a(f)(2) and (3) and that are incorporated by reference in 10 CFR 50.55a(a)(1)(iv) to the extent practical within the limitations of design, geometry, and materials of construction of the components.

The NRC regulations in 10 CFR 50.55a(z) state, in part, that alternatives to the requirements of 10 CFR 50.55a(f) may be used, when authorized by the NRC, if the licensee demonstrates (1) the proposed alternatives would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The applicable ASME OM Code of record for the IST program at BVPS-1 for the fifth 10-year IST program interval and BVPS-2 for the fourth 10-year IST program interval, which began on September 20, 2017, and are currently scheduled to end on September 19, 2027, is the 2004 Edition through 2006 Addenda (OMb-2006) of the ASME OM Code, as incorporated by reference in 10 CFR 50.55a.

3.0 TECHNICAL EVALUATION

3.1 Licensees Alternative Request VRR4 The IST requirements of the ASME OM Code, as incorporated by reference in 10 CFR 50.55a, related to this alternative request are as follows:

ASME OM Code 2004 Edition through 2006 Addenda, Subsection ISTC, Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants, paragraph ISTC-3630, Leakage Rate for Other Than Containment Isolation Valves, subparagraph (a), Frequency, states that Tests shall be conducted at least once every 2 years.

In its submittal, the licensee requests an extension of the leakage testing interval for the following 72 isolation valves at BVPS-1 and BVPS-2:

Table 1 Valve ID Function Code OM Class Category 1SI-83 Hot Legs HHSI Supply Check 1 A/C MOV-1SI-869A HHSI to RCS Hot Legs Isolation 1 A 1CH-31 Regen H/X Inlet Check 2 A/C MOV-1CH-289 Chg PP Disch Hdr to Regen H/X 2 A 1SI-84 Hot Legs HHSI Supply Check 1 A/C MOV-1SI-869B HHSI to RCS Hot Legs Isolation 1 A 1CH-181 RCP 1A Seal Supply Check 2 A/C MOV-1CH-308A RCP 1A Seal Injection Isolation 2 A

Table 1 Valve ID Function Code OM Class Category 1CH-182 RCP 1B Seal Supply Check 2 A/C MOV-1CH-308B RCP 1B Seal Injection Isolation 2 A 1CH-183 RCP 1C Seal Supply Check 2 A/C MOV-1CH-308C RCP 1C Seal Injection Isolation 2 A 1CH-170 Fill Header Check 1 A/C FCV-1CH-160 Fill Header Flow Control Valve 1 A MOV-1SI-890A A LHSI to RCS Hot Legs Isolation 1 A MOV-1SI-890C LHSI to RCS Cold Legs Isolation 1 A MOV-1SI-890B B LHSI to RCS Hot Legs Isolation 1 A MOV-1SI-860A A LHSI PP CNMT Sump Suct Isol 2 A MOV-1SI-860B B LHSI PP CNMT Sump Suct Isol 2 A 1SI-95 Cold Legs HHSI Supply Check 1 A/C MOV-1SI-836 HHSI to RCS Cold Legs Isolation 1 A 1SI-94 Cold Legs HHSI Supply Check 1 A/C MOV-1SI-867C Boron Injection Tank Outlet Isol 1 A MOV-1SI-867D Boron Injection Tank Outlet Isol 1 A 2SIS*83 HHSI Check Valve to RCS Hot Legs 2 A/C 2SIS*MOV869A HHSI Hot Legs Injection Isolation 2 A 2CHS*31 Charging Header Isolation Check 2 A/C 2CHS*MOV289 Normal Charging Header Isolation 2 A 2SIS*84 HHSI Check Valve to RCS Hot Legs 2 A/C 2SIS*MOV869B HHSI Hot Legs Injection Isolation 2 A 2SIS*94 HHSI Check Valve to RCS Cold Legs 2 A/C 2SIS*MOV836 High Head to Cold Legs Injection Isolation 2 A 2SIS*MOV840 High Head to Cold Legs Injection Isolation 2 A 2CHS*474 Reactor Coolant Pump (RCP) 21A Seal Supply 2 A/C Containment Check 2CHS*MOV308A RCP 21A Seal Water Injection Isolation 2 A 2CHS*476 RCP 21B Seal Supply Containment Check 2 A/C 2CHS*MOV308B RCP 21B Seal Water Injection Isolation 2 A 2CHS*475 RCP 21C Seal Supply Containment Check 2 A/C 2CHS*MOV308C RCP 21C Seal Water Injection Isolation 2 A 2CHS*472 Loop Fill Containment Isolation Check 2 A/C 2CHS*FCV160 RCS Loop Fill Header Flow Control Valve 2 A 2SIS*95 HHSI Check Valve to RCS Cold Legs 2 A/C 2SIS*MOV867C HHSI Pump Isolation to Cold Leg Injection 2 A 2SIS*MOV867D HHSI Pump Isolation to Cold Leg Injection 2 A Table 2 Valve ID Valve Description Code OM Class Category MOV-1SI-885A LHSI PP 1A Min Flow Line (Train A) 2 A MOV-1SI-885B LHSI PP 1B Min Flow Line (Train A) 2 A MOV-1SI-885C LHSI PP 1B Min Flow Line (Train B) 2 A MOV-1SI-885D LHSI PP 1A Min Flow Line (Train B) 2 A 1SI-27 Charging Pump RWST Supply Check 2 A/C

Table 2 Valve ID Valve Description Code OM Class Category MOV-1CH-115B RWST Outlet to Chg PP Suction Hdr Isol 2 A MOV-1CH-115D RWST Outlet to Chg PP Suction Hdr Isol 2 A 2SIS*6 LHSI Pump 21A Discharge Check 2 A/C 2SIS*MOV8809A LHSI Pump 21A Suction Isolation 2 A 2SIS*MOV8890A LHSI Pump 21A Min Flow Recirc Isolation 2 A 2SIS*7 LHSI Pump 21B Discharge Check 2 A/C 2SIS*MOV8809B LHSI Pump 21B Suction Isolation 2 A 2SIS*MOV8890B LHSI Pump 21B Min Flow Recirc Isolation 2 A 2SIS*27 Check Valve to HHSI Pumps from RWST 2 A/C 2CHS*LCV115B Charging Pump Suction from RSWT 2 A 2CHS*LCV115D Charging Pump Suction from RSWT 2 A Table 3 Valve ID Valve Description Code OM Class Category 1CCR-289 RCP 1A Thermal Barrier Cooler Inlet Chk 3 A/C TV-1CC-107A RCP 1A Thermal Barrier Cooler Outlet Isol 3 A 1CCR-290 RCP 1B Thermal Barrier Cooler Inlet Chk 3 A/C TV-1CC-107B RCP 1B Thermal Barrier Cooler Outlet Isol 3 A 1CCR-291 RCP 1C Thermal Barrier Cooler Inlet Chk 3 A/C TV-1CC-107C RCP 1C Thermal Barrier Cooler Outlet Isol 3 A 2CCP*289 RCP 21A Thermal Barrier Clr Supply Chk 3 A/C 2CCP*MOV107A RCP 21A Thermal Barrier Clr Disch Isol 3 A 2CCP*290 RCP 21B Thermal Barrier Clr Supply Chk 3 A/C 2CCP*MOV107B RCP 21B Thermal Barrier Clr Disch Isol 3 A 2CCP*291 RCP 21C Thermal Barrier Clr Supply Chk 3 A/C 2CCP*MOV107C RCP 21C Thermal Barrier Clr Disch Isol 3 A

Reason for Request

The licensee states, in part:

The valves in Table 1 above are containment isolation valves (CIVs) that are expected to remain water filled following a loss of coolant accident (LOCA).

Therefore, they are not subject to Type-C leakage tests as noted in the Beaver Valley Power Station, Unit No. 1 (BVPS-1) and Unit No. 2 (BVPS-2) Licensing Requirements Manuals, Table 3.6.1-1, Containment Penetrations, but instead are leakage tested once every two years in accordance with Paragraph ISTC-3630(a) of the ASME OM Code using water.

The valves in Table 2 above are valves that are being leakage tested in response to Information Notice 91-56, Potential Radioactive Leakage to Tank Vented to Atmosphere. These valves have been determined to have a safety function with respect to control room and offsite dose limits and will prevent leakage of containment sump water from the emergency core cooling system (ECCS) recirculation lines to the refueling water storage tank (RWST) which is vented to

atmosphere. These non-CIVs are currently leakage tested once every two years in accordance with Paragraph ISTC-3630(a) of the ASME OM Code using water.

The valves in Table 3 above are reactor coolant pump thermal barrier cooler isolation valves that isolate the lower pressure component cooling water system from the higher-pressure reactor coolant system in the event of a thermal barrier cooler rupture. These non-CIVs are currently leakage tested once every two years in accordance with Paragraph ISTC-3630(a) of the ASME OM Code using water.

The valves identified in Tables 1, 2 and 3 above are all non-Type-C tested valves in water applications that are leakage tested using water at least once every two years. Although they are not specifically included in the scope for performance based testing as described in NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50 Appendix J, and are not leakage tested in accordance with 10 CFR 50 Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Option B, Performance Based Requirements, Energy Harbor Nuclear Corp. intends to implement an approach similar to the performance based testing frequency of Option B instead of performing leakage testing of these valves once every two years at BVPS-1 and BVPS-2.

The reason for requesting this relief is dose reduction to comport with NRC and industry as low as reasonably achievable (ALARA) radiation dose principles.

Additionally, the extended test frequency will reduce the amount of out of service time for certain high head and low head safety injection valves in the boration flow paths during a refueling outage.

Proposed Alternative In its submittal, the licensee stated, in part:

The proposed alternative is to perform leakage testing of the valves identified in Tables 1, 2 and 3 at intervals based on valve performance ranging from every refueling outage to every fourth refueling outage. The specific interval for each valve would be a function of its performance and would be established in a manner consistent with the CIV process under 10 CFR 50 Appendix J, Option B.

Valves that have demonstrated good performance for two consecutive cycles may have their test interval extended to every third refueling, not to exceed 60 months. Valves that have demonstrated good performance for three consecutive cycles may have their test interval extended to every fourth refueling, not to exceed 75 months, if additional considerations as specified in NEI 94-01, Revision 3-A, Section 11.3.2, Programmatic Controls, are met. Any leakage test failure would require the valve to be returned to the initial test frequency of every refueling outage until good performance can again be reestablished.

The valves listed in Table 1 and Table 2 are arranged with two valves in series.

This valve arrangement ensures that leakage outside of containment or leakage from the containment sump to the refueling water storage tank that is vented to atmosphere will be kept to a minimum. One exception is for BVPS-1

containment penetration numbers 68 and 69 that are located between the water filled containment sump (that is, post LOCA) and one valve (that is, MOV-ISI-860A, and MOV-ISI-860B, respectively) just outside the containment penetrations. A second exception is the RCP thermal barrier cooler valves because only one Category A valve is located on either side of each thermal barrier cooler.

As stated on page 22 of NEI 94-01, Revision 3-A, NUREG-1493, Performance-Based Containment Leak-Test Program, provided the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessments of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate testing intervals.

NUREG-1493 found the effect of Type C testing on overall accident risk is small and concluded that performance-based alternatives to local leakage rate testing requirements are feasible without significant risk impacts; and although extended testing intervals led to minor increases in potential off-site dose consequences, the actual decrease in on-site (worker) doses exceeded (by at least an order of magnitude) the potential off-site dose increases.

Option B of Appendix J to 10 CFR 50 provides performance-based primary reactor containment leakage-rate test requirements. NEI 94-01, Revision 3-A, allows for an extended leak test interval of up to 75 months. Although NEI 94-01 does not address seat leakage testing with water, it has been determined that extending the frequency for leak rate testing provides a low level of risk, therefore, the proposed alternative method would provide an acceptable level of quality and safety.

A review of recent historical data identified that leak testing each of the valves listed in Section 3.1 every refueling outage results in a total personnel dose of approximately 75 millirem at BVPS-1 and 96 millirem at BVPS-2. The proposed extended test interval (assuming all valves are on an extended frequency) would provide a total dose savings at the site of approximately 0.5 Rem over the course of four refueling outages at each Unit.

The extended test frequency will also reduce the amount of out of service time during refueling outages for certain high head and low head safety injection valves located in the boration flow paths at each Unit by approximately 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> per Unit per refueling outage.

A review of the leakage test results for the valves identified in Tables 1, 2 and 3 shows that these valves have maintained a history of excellent performance with a few exceptions noted below:

Fill Header Check Valve [1CH-170] failed its leak test in 2000 during the thirteenth refueling outage and was repaired (seats were lapped). Leakage rates since then have been acceptable.

Reactor Coolant Pump 1A Thermal Barrier Cooler Inlet Check Valve

[1CCR-289] had a history of leaking more than its sister valves [1CCR-290 and 291] even though its seats were passing a 360-degree blue check. The

valve was replaced during the seventeenth refueling outage in 2006, and the subsequent leak rates have been acceptable.

Reactor Coolant System Loop Fill Header Flow Control Valve

[2CHS*FCV160] failed its leak test in 2012 during the sixteenth refueling outage and was adjusted (stem valve travel adjusted to make tighter).

Subsequent leak rates have been acceptable.

This excellent valve performance supports extending the leakage test frequency for valves in Tables 1, 2 and 3 to four refueling outages based on achieving acceptable results for a least two consecutive outages prior to the submittal of this relief request. Therefore, the proposed alternative to perform leakage testing of the valves identified in Tables 1, 2 and 3 at intervals based on valve performance in lieu of the current ASME Code paragraph ISTC-3630(a) requirement to conduct tests at least once every two years, provides reasonable assurance that the valves are operationally ready.

3.2 NRC Staff Evaluation As incorporated by reference in 10 CFR 50.55a, ASME OM Code (2004 Edition through 2006 Addenda), paragraph ISTC-3630(a), requires that valves within the scope of the ASME OM Code that have specific leakage criteria (other than containment isolation valves that are tested in accordance with Appendix J to 10 CFR Part 50) are required to be leak-rate tested at least once every 2 years. The licensee has proposed an alternative test in lieu of this requirement for the 72 specific isolation valves listed in Tables 1, 2, and 3 of this SE.

Specifically, the licensee proposes to functionally test and verify the leakage rate of 72 valves listed in Tables 1, 2, and 3 of this SE using an Option B to Appendix J of 10 CFR Part 50 performance-based schedule. Valves would initially be tested at the required interval schedule, which is currently every refueling outage, or 2 years, as specified by ASME OM Code, subsection ISTC, paragraph ISTC-3630(a). Valves that have demonstrated good performance for two consecutive cycles may have their test interval extended to 75 months with a permissible extension for nonroutine emergent conditions of 9 months (84 months total). Any valve leakage test failure would require the component to return to the initial interval of every 30 months until good performance can again be established.

Option B of Appendix J to 10 CFR Part 50 is a performance-based containment leakage test program. Guidance for implementation of acceptable leakage rate test methods, procedures, and analyses is provided in Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program (ADAMS Accession No. ML003740058). Regulatory Guide 1.163 endorses Nuclear Energy Institute (NEI) Topical Report 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated July 21, 1995 (ADAMS Accession No. ML11327A025), with the limitation that Type C component test intervals cannot extend greater than 60 months. The current version of NEI 94-01 is Revision 3-A, which allows Type C containment isolation valve test intervals to be extended to 75 months with a permissible extension for nonroutine emergent conditions of 9 months (84 months total). By

letter dated December 6, 2012, the NRC staff found the guidance in NEI 94-01, Revision 3-A, to be acceptable (see ADAMS Accession Nos. ML121030286 and ML12226A546) with the following conditions:

1. Extended interval for Type C local leakage-rate tests (LLRTs) may be increased to 75 months with the requirement that a licensees post-outage report include the margin between Type B and Type C leakage rate summation and its regulatory limit. In addition, a corrective action plan shall be developed to restore the margin to an acceptable level.

Extensions of up to 9 months (total maximum interval of 84 months for Type C tests) are permissible only for nonroutine emergent conditions. This provision (9-month extension) does not apply to valves that are restricted and/or limited to 30-month intervals in Section 10.2 (such as boiling-water reactor main steam isolation valves) or to valves held to the base interval (30 months) due to unsatisfactory LLRT performance.

2. When routinely scheduling any LLRT valve interval beyond 60-months and up to 75 months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B and Type C total and must be included in a licensees post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

In its request, the licensee provided justification that compliance with the provisions in the ASME OM Code, subsection ISTC, paragraph ISTC-3630(a), as incorporated by reference in 10 CFR 50.55a, to conduct leakage testing of those isolation valves at BVPS-1 and BVPS-2 would result in a hardship without a compensating increase in the level of quality and safety in accordance with 10 CFR 50.55a(z)(2).

The licensee also detailed the total dose to personnel at each BVPS unit for leak testing each of the valves and calculated the savings of 0.5 rem at each BVPS unit over the course of four refueling outages in conforming with 10 CFR 20.1101.

In its submittal, the licensee reported that its review of the leakage test results for the specific isolation valves at BVPS-1 and BVPS-2 showed that all but three of those isolation valves maintained a history of excellent performance. Fill Header Check Valve 1CH-170 failed its leakage test in 2000, but after repair, has subsequently demonstrated acceptable performance.

Reactor Coolant Pump 1A Thermal Barrier Cooler Inlet Check Valve 1CCR-289 had a history of leakage, but was replaced in 2006, and has subsequently demonstrated acceptable performance. Reactor Coolant System Loop Fill Header Flow Control Valve 2CHS*FCV160 failed its leakage test in 2012 but, after adjustments, has subsequently demonstrated acceptable performance. The licensee considered that the performance of these isolation valves with acceptable results for at least two consecutive outages prior to submittal of its request supports an extension of the leakage test interval for these valves.

In response to the licensees request, the NRC staff reviewed the historical performance data of the isolation valves listed in Tables 1, 2, and 3 of this SE and performed a review of isolation valve operating experience at BVPS-1 and BVPS-2 using the Industry Reporting Information System database established by the Institute of Nuclear Power Operations. The NRC staff did not identify any events for the isolation valves listed in Tables 1, 2, and 3 of this SE for the last 20 years. The staff considered that the isolation valves listed in Tables 1, 2, and 3 of this SE have had excellent performance history with minor maintenance issues, justifying the extension of the leakage testing interval for these valves.

Based on the information provided by the licensee for the 72 specific isolation valves at BVPS-1 and BVPS-2 identified in its submittal, the NRC staff found that (1) previous leakage testing of these isolation valves indicates their acceptable historical performance, (2) no current concerns with the performance of these isolation valves have been identified, (3) periodic maintenance activities are not modified by this request, and (4) a hardship exists for the performance of leakage testing of these isolation valves. Therefore, the staff determined that the licensees proposed alternative for an extension of the leakage testing interval for the specified 72 isolation valves at BVPS-1 and BVPS-2 is acceptable in accordance with 10 CFR 50.55a(z)(2).

The proposed alternative will provide reasonable assurance that these isolation valves will be operationally ready to perform their safety functions during the remainder of the fifth 10-year IST program interval at BVPS-1 and during the remainder of the fourth 10-year IST program interval at BVPS-2, which began on September 20, 2017, and are currently scheduled to end on September 19, 2027.

4.0 CONCLUSION

As described in this SE, the NRC staff concluded that proposed alternative VRR4 will provide reasonable assurance that the valves at BVPS-1 and BVPS-2 specified in the licensees submittal dated April 2, 2020, as supplemented April 6 ,2020, are operationally ready to perform their safety functions until the end of the IST intervals highlighted above.1 The NRC staff finds that complying with the provisions of the ASME OM Code and its Code cases for testing of the specified isolation valves at BVPS-1 and BVPS-2 would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed the regulatory requirements set forth in 10 CFR 50.55a(z)(2).

All other ASME OM Code requirements, as incorporated by reference in 10 CFR 50.55a, for which relief or an alternative was not specifically requested and approved as part of these subject requests remain applicable.

Principal Contributors: J. Huang M. Farnan Date: July 16, 2020 1

On April 8, 2020, the NRC provided verbal authorization for the use of proposed alternative VRR4 at BVPS-2 for the isolation valves included in the April 6, 2020, supplement for the remainder of the fourth IST interval.

ML20153A014 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DEX/EMIB/BC(A)

NAME JTobin LRonewicz ABuford DATE 6/04/2020 6/03/2020 5/28/2020 OFFICE NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME JDanna JTobin DATE 7/16/2020 7/16/2020