ML18345A060

Relief Request Regarding Use of a Performance Based Testing Frequency for Pressure Isolation Valves (RR-4-14)

ML18345A060
Person / Time
Site:	Summer
Issue date:	12/17/2018
From:	Markley M Plant Licensing Branch II
To:	Lippard G South Carolina Electric & Gas Co
Williams S, 415-1009
References
EPID L-2018-LLR-0129
Download: ML18345A060 (9)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 17, 2018 Mr. George A. Lippard, Ill Vice President, Nuclear Operations South Carolina Electric & Gas Company Virgil C. Summer Nuclear Station P.O. Box 88, Mail Code 800 Jenkinsville, SC 29065

SUBJECT:

VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. 1 - RELIEF REQUEST REGARDING USE OF A PERFORMANCE BASED TESTING FREQUENCY FOR PRESSURE ISOLATION VALVES (RR-4-14) (EPID NO. L-2018-LLR-0129)

Dear Mr. Lippard:

By letter dated October 8, 2018, South Carolina Electric & Gas (SCE&G), the licensee, submitted alternative request RR-4-14 to the U.S. Nuclear Regulatory Commission (NRC). The licensee requested an alternative test plan in lieu of certain inservice testing (1ST) requirements of the 2004 Edition through 2006 Addenda of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) for the 1ST program at Virgil C. Summer Nuclear Station (Summer), Unit No. 1 for the remainder of the fourth 10-year 1ST program interval, which began on January 1, 2014 and is scheduled to end on December 31, 2023.

Specifically, pursuant to Title 10 of the Code of Federal Regulations 10 CFR 50.55a(z)( 1), the licensee requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and safety.

The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation that the proposed alternative described in Relief Request No. R-4-14 provides an acceptable level of quality and safety for components listed on page 2 of the licensee's proposed alternative in its letter dated October 8, 2018. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)( 1).

Therefore, the NRC staff authorizes the proposed alternative in Relief Request No. R-4-14 for the remainder of the fourth 10-year inservice testing interval at Summer, Unit No. 1, which is both currently scheduled to end on December 31, 2023. The performance-based program interval shall not exceed three refuelings outages (RFOs) or 60 months.

All other ASME OM Code requirements for which relief was not specifically requested and approved in the subject request for relief remain applicable.

G. Lippard, Ill If you have any questions, please contact the Vogtle project manager, Shawn Williams, at 301-415-1009 or by e-mail at Shawn.Williams@nrc.gov.

Sincerely, Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-395

Enclosure:

Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST RR-4-14 REGARDING USE OF A PERFORMANCE BASED TESTING FREQUENCY FOR PRESSURE ISOLATION VALVES SOUTH CAROLINA ELECTRIC & GAS SOUTH CAROLINA PUBLIC SERVICE AUTHORITY VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. 1 DOCKET NO. 50-395

1.0 INTRODUCTION

By letter dated October 8, 2018, (Agencywide Documents and Access Management System (ADAMS) Accession No. ML18282A046), South Carolina Electric & Gas (SCE&G), the licensee, submitted alternative request RR-4-14 to the U.S. Nuclear Regulatory Commission (NRC of the Commission). The licensee requested an alternative test plan in lieu of certain inservice testing (1ST) requirements of the 2004 Edition through 2006 Addenda of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) for the 1ST program at Virgil C. Summer Nuclear Station (Summer), Unit No. 1 during the remainder of the fourth 10-year 1ST program interval, which began on January 1, 2014 and is scheduled to end on December 31, 2023.

Specifically, pursuant to Title 10 of the Code of Federal Regulations 10 CFR 50.55a(z)( 1), the licensee requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and safety.

2.0 REGULATORY REQUIREMENTS Paragraph 10 CFR 50.55a(f), "lnservice Testing Requirements," requires, in part, that 1ST of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME OM Code and applicable addenda incorporated by reference in the regulations. Exceptions are allowed where alternatives have been authorized by the NRC pursuant to paragraphs 10 CFR 50.55a(z}(1) and 10 CFR 50.55a(z)(2).

In proposing alternatives, the licensee must demonstrate that ( 1) the proposed alternatives provide an acceptable level of quality and safety or (2) compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Section 50.55a allows the NRC to authorize alternatives from the ASME OM Code requirements upon making necessary findings.

Enclosure

Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request and the Commission to authorize the alternative requested by the licensee.

3.0 TECHNICAL EVALUATION

3.1 Licensee's Relief Request No. RR-4-14 3.1.1 Applicable ASME OM Code The licensee requested an alternative test plan in lieu of certain 1ST requirements of the 2004 Edition through 2006 Addenda of the ASME OM Code for the 1ST program at VCSNS for the remainder of the fourth 10-year 1ST interval which began on January 1, 2014 and is currently scheduled to end on December 31, 2023.

The ISTC-3630 "Leakage Rate for Other Than Containment Isolation Valve", states that "Category A valves with a leakage requirement not based on an Owner's 10 CFR 50, Appendix J program, shall be tested to verify their seat leakages are within acceptable limits.

Valve closure before seat leakage testing shall be by using the valve operator with no additional closing force applied."

The ISTC-3630(a) "Frequency", states that "Tests shall be conducted at least once every 2 years."

3.1.2 Components for Which Relief is Requested In its submittal, the licensee requested alternative testing for the following valves:

Valve ID Description Code Code Category Class XVG08701A-RH Residual Heat Removal (RH) Header A Isolation A 1 Valve (IRC)

XVG08701 B-RH RH Header B Isolation Valve (IRC) A 1 XVG08702A-RH RH Inlet Header A Isolation Valve A 1 XVG08702B-RH RH Inlet Header B Isolation Valve A 1 XVC08948A-SI Safety Injection (SI) Loop A Outlet Header Check A/C 1 Valve XVC08948B-SI SI Loop B Outlet Header Check Valve A/C 1 XVC08948C-SI SI Loop C Outlet Header Check Valve A/C 1 XVC08956A-SI SI Accum A Discharge Header Check Valve A/C 1 XVC08956B-SI SI Accum B Discharqe Header Check Valve A/C 1 XVC08956C-SI SI Accum C Discharge Header Check Valve A/C 1 XVC08973A-SI Reactor Coolant System (RCS) Loop A Cold Leg A/C 1 Inlet Header Check Valve XVC089738-SI RCS Loop B Cold Leg Inlet Header Check Valve A/C 1 XVC08973C-SI RCS Loop C Cold Leg Inlet Header Check Valve A/C 1 XVC0897 4A-SI SI Header A Check Valve (IRC) A/C 2 XVC089748-SI SI Header B Check Valve (IRC) A/C 2 XVC08988A-SI RHR Supply Header Check Valve A/C 1 XVC089888-SI RHR Supply Header Check Valve A/C 1

Valve ID Description Code Code Category Class XVC08990A-S1 Loop A Low Head Hot Leg Check Valve A/C 1 XVC08990B-S1 Loop B Low Head Hot Leg Check Valve A/C 1 XVC08990C-S1 Loop C Low Head Hot Leg Check Valve AIC 1 XVC08992A-S1 Loop A Hioh Head Hot Leo Check Valve A/C 1 XVC08992B-S1 Loop B High Head Hot Leg Check Valve A/C 1 XVC08992C-S1 Loop C High Head Hot Leg Check Valve AIC 1 XVC08993A-S1 Loop A High Head Hot Leg Header Check Valve A/C 1 XVC08993B-S1 Loop B High Head Hot Leg Header Check Valve AIC 1 XVC08993C-SI Loop C Hioh Head Hot Leo Header Check Valve A/C 1 XVC08995A-SI Loop A High Head Cold Leg Check Valve A/C 1 XVC08995B-S1 Loop B Hiqh Head Cold Leq Check Valve A/C 1 XVC08995C-SI Loop C High Head Cold Leg Check Valve A/C 1 XVC08997A-SI Loop A Low Head Cold Leq Check Valve A/C 1 XVC08997B-SI Loop B Low Head Cold Leo Check Valve A/C 1 XVC08997C-SI Loop C Low Head Cold Leg Check Valve AIC 1 XVC08998A-SI Loop A Low Head Cold Leg Check Valve A/C 1 XVC08998B-SI Loop B Low Head Cold Leq Check Valve AIC 1 XVC08998C-SI Loop C Low Head Cold Leg Check Valve A/C 1 3.1.3 Reason for Request In its submittal, the licensee stated, in part:

ASME OM Code Subsection ISTC-3630 paragraph (a) requires that leakage rate testing for pressure isolation valves (PIVs) be performed at least once every two years. PIVs are not specifically included in the scope of performance-based testing as provided for in 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Option B, Performance-Based Requirements. These motor-operated and check valve PIVs are, in some cases, containment isolation valves (CIVs).

The Virgil C. Summer Nuclear Station (VCSNS), Unit 1, Technical Specification (TS) 6.8.4.g, Containment Leakage Rate Testing Program, currently contains a requirement to establish the containment leakage rate testing program in accordance with the guidelines contained in [Nuclear Energy Institute] NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J.

NEI 94-01, Paragraph 10.2.3.2, Extended Test Interval, states, in part:

Test intervals for Type C valves may be increased based upon completion of two consecutive periodic as-found Type C tests where the result of each test is within a licensee's allowable administrative limits. Elapsed time between the first and last tests in a series of consecutive passing tests used to determine performance shall be 24 months or the nominal test interval (e.g., refueling cycle) for the valve prior to implementing Option B to Appendix J. Test intervals for Type C valves should be determined by a licensee in accordance with Section 11.0.

The licensee further stated, in part:

The concept behind the Option B alternative for CIVs is that licensees should be allowed to adopt cost effective methods for complying with regulatory requirements. Additionally, NEI 94-01 describes the risk-informed basis for the extended test intervals under Option B. The discussion concludes that CIVs, which have demonstrated good performance by the successful completion of two consecutive leakage rate tests over two consecutive cycles, may increase their test frequencies. NEI 94-01 also presents the results of a comprehensive risk analysis, including the conclusion that "the risk impact associated with increasing

[leak rate] test intervals is negligible (i.e., less than 0.1 percent of total risk).

This proposed alternative is intended to provide for a performance-based scheduling of PIV tests at VCSNS. The primary reason for requesting this alternative is to eliminate unnecessary thermal cycles in the RCS cold leg safety injection piping. A periodic thermal transient was identified in the RCS Cold Leg Safety Injection (SI) piping after every post-refueling heat-up, since approximately 1999. These transients coincide with the testing of the RCS PIVs, which causes the inlet check valves (XVC08998A-SI, XVC08998B-SI, and XVC08998C-SI) to open during this portion of testing, allowing cooler Volume Control Tank (VCT) temperature water into the Safety Injection (SI) piping.

These thermal transients, identified by the plant thermal cycle counting software, are counted against allowable fatigue usage totals for the affected piping system.

For the RCS Cold Leg SI lines, the approximate fatigue usage is at 70% of the allowable. As a result of the high cumulative usage factor, additional ultrasonic inspections of the welds and elbows of the RCS Cold Leg SI lines A, B, and C in areas susceptible to thermal stratification were performed during refueling outage RF-21 in April 2014, with acceptable exam results. The proposed extended test intervals would reduce the frequency and, therefore, the impact of injecting ECCS water into the RCS during testing.

An additional reason for requesting this alternative is dose reduction to conform with Nuclear Regulatory Commission (NRC) and industry As Low As Reasonably Achievable (ALARA) radiation dose principles. The nominal fuel cycle lengths at VCSNS are 18 months. However, since refueling outages (RFOs) may be scheduled slightly beyond 18 months, a 60-month period is used to provide a bounding timeframe to encompass three RFOs. The review of recent historical data identified that PIV testing results in a total personnel dose of approximately 300 milli-Roentgen Equivalent Man (mREM) each RFO.

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

SCE&G proposes to perform PIV testing at VCSNS at intervals ranging from every refueling outage to every third 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 extended test eligibility process guidance under 10 CFR 50, Appendix J, Option B. These valves have been historically tested at the required interval schedule, which is currently every refueling outage, or two

years, as specified in ASME OM Code Subsection ISTC-3630 paragraph (a).

Leakage rates less than the leakage limits found in TS and VCSNS procedure STP-215.008, "SI and RFI System Valve Leakage Test", shall be considered acceptable. Valves that have demonstrated good performance for two consecutive cycles may have their test interval extended to every third refueling outage, not to exceed 60-months. Any PIV leakage test failure would require the component to return to the initial interval of every RFO or two years until good performance is re-established. This request, upon approval, will be applied to the remainder of the station's fourth 10-year interval, which commenced January 1, 2014, and is currently scheduled to end on December 31, 2023.

3.2 NRC Staff Evaluation The licensee has proposed an alternative test in lieu of the requirements found in 2004 Edition through the 2006 Addenda of the ASME OM Code, Section ISTC-3630(a) for 35 PIVs listed in section 3.1.2 of this safety evaluation. Specifically, the licensee proposes to functionally test and verify the leakage rate of 35 PIVs using 10 CFR 50, Appendix J, Option B performance based schedule. Valves would initially be tested at the required interval schedule which is currently every RFO or two years as specified by ASME OM Code, Section ISTC-3630(a).

Valves that have demonstrated good performance for two consecutive cycles may have their test interval extended to every three RFOs, not to exceed five years. Any PIV leakage test failure would require the component to return to the initial interval of every RFO or two years until good performance can again be established.

PIVs are defined as two valves in series within the reactor coolant pressure boundary which separate the high pressure reactor coolant system from an attached lower pressure system.

Failure of a PIV could result in an over-pressurization event which could lead to a system rupture and possible release of fission products to the environment. This type of failure event was analyzed under NUREG/CR-5928, "[Inter System Loss of Coolant Accident] ISLOCA Research Program" (ADAMS Accession No. ML072430731). The purpose of NUREG/CR-5928 was to quantify the risk associated with an ISLOCA event. NUREG/CR-5928 analyzed Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) designs.

Paragraph 10 CFR 50, Appendix J, Option B is a performance based leakage test program.

Guidance for implementation of acceptable leakage rate test methods, procedures, and analyses is provided in Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program" (ADAMS Accession No. ML003740058). RG 1.163 endorses Nuclear Energy Institute (NEI) Topical Report (TR} 94-01, Revision 0, "Industry Guideline For Implementing Performance-Based Option of 10 CFR 50, Appendix J" dated July 26, 1995, with the limitation that Type-C components test interval cannot extend greater than 60 months The 35 PIVs are currently being leak tested every RFO or 2 years. Performance of the leakage test of the 35 PIVs could cause thermal system transients and additional occupational exposure to radiation. The licensee estimates that completion of leak test requirements averages a dose of 300 mRem. The valves have a maintained a history of overall good performance. Extending the leakage test interval based on good performance, low risk factor as noted in NUREG/CR-5928, and NEI 94-01 report conclusion that the risk impact associated with increasing leak rate test intervals being negligible supports the basis for a performance-based program. Therefore, the NRC staff concludes that the licensee's proposed alternative provides an acceptable level of quality and safety.

The licensee is authorized to implement a performance based program for the 35 PIVs at Summer, Unit No. 1. The performance based program interval shall not exceed three RFOs or 60 months.

4.0 CONCLUSION

As set forth above, the NRC staff finds that the proposed alternative described in alternative request RR-4-14 provides an acceptable level of quality and safety for components listed in Table 1. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1 ).

Therefore, the NRC staff authorizes the proposed alternative in Relief Request No. R-4-14 for the remainder of the fourth 10-year inservice testing interval at Summer, Unit No. 1, which is currently scheduled to end on December 31, 2023. The performance-based program interval shall not exceed three RFOs or 60 months.

All other ASME OM Code requirements for which relief was not specifically requested and approved in the subject requests for relief remain applicable.

Principle Contributor: M. Farnan, NRR Date: December 17, 2018

G. Lippard, Ill

SUBJECT:

VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. 1 - RELIEF REQUEST REGARDING USE OF A PERFORMANCE BASED TESTING FREQUENCY FOR PRESSURE ISOLATION VALVES (RR-4-14) (EPID NO. L-2018-LLR-0129)

DATED DECEMBER 17, 2018 DISTRIBUTION:

PUBLIC RidsNrrKGoldstein Resource RidsNrrPMSummer Resource RidsRgn2MailCenter Resource RidsACRS MailCTR Resource RidsNrrDeEMIB Resource RidsNrrDorlLpl2-1 Resource MFarnan, NRR ADAMS Accession No.: ML18345A060 *B Memo Dated OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DE/EMIB/BC NAME MMahoney KGoldstein SBailey DATE 12/17/18 12/11/18 11/21/18 OFFICE NRR/DORL/LPL2-1/BC NAME MMarkley DATE 12/17/18 OFFICIAL RECORD COPY