ML17136A112

Alternative to the Requirements of the American Society of Mechanical Engineers Code for Operation and Maintenance of Nuclear Power Plants

ML17136A112
Person / Time
Site:	Nine Mile Point
Issue date:	05/30/2017
From:	James Danna Plant Licensing Branch 1
To:	Bryan Hanson Exelon Generation Co
Marshall M
References
CAC MF9073, CAC MF9074
Download: ML17136A112 (10)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001

.May 30, 201 7 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

NINE MILE POINT NUCLEAR STATION, UNITS 1 AND 2 - RE: ALTERNATIVE TO THE REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS CODE FOR OPERATION AND MAINTENANCE OF NUCLEAR POWER PLANTS (CAC NOS. MF9073 AND MF9074)

Dear Mr. Hanson:

By letter dated December 27, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17003A096), as supplemented by letter dated April 6, 2017 (ADAMS Accession No. ML17096A553), Exelon Generation Company, LLC (the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to certain American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) requirements at Nine Mile Point Nuclear Station (Nine Mile Point), Units 1 and 2.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(a)(3)(i),

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. GVRR-3 provides an acceptable level of quality and safety for components listed on pages 1 and 2 of the proposed alternative attached to the licensee's letter dated December 27, 2016. 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 ).

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

Therefore, the NRC staff authorizes the proposed alternative in Relief Request No. GVRR-3 for the remainder of the fourth 10-year inservice testing interval at Nine Mile Point, Unit 1, and the third 10-year inservice testing interval at Nine Mile Point, Unit 2, which are both currently scheduled to end on December 31, 2018. The performance-based program interval shall not exceed 75 months, with the exception that non-routine emergent conditions may extend the program interval 9 months.

B. Hanson If you have any questions, please contact the project manager, Michael L. Marshall, Jr., at (301) 415-2871 or Michael.Marshall@nrc.gov.

Sincerely, James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-220 and 50-410

Enclosure:

Safety Evaluation cc w/encls: Distribution via Listserv

B. Hanson

SUBJECT:

NINE MILE POINT NUCLEAR STATION, UNITS 1 AND 2 - RE: ALTERNATIVE TO THE REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS CODE FOR OPERATION AND MAINTENANCE OF NUCLEAR POWER PLANTS (CAC NOS. MF9073 AND MF9074) DATED MAY 30, 2017 DISTRIBUTION:

Public LPL 1 R/F RidsNrrDorlLpl 1 RidsRgn1 MailCenter RidsNrrLALRonewicz RidsNrrPMNineMilePoint RidsNrrDeEpnb Resource MFarnan, NRR JBowen, OEDO ADAMS Accession Number: ML17136A112 *by safety evaluation OFFICE DORL/LPL 1/PM DORL/LPL 1/LA DE/EPNB/BC* DORL/LPL 1/BC DORL/LPL 1/PM NAME MMarshall LRonewicz DAiiey JDanna MMarshall DATE 05/22/2017 05/19/2017 04/14/2017 05/26/2017 05/30/2017 OFFICIAL RECORD COPY

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST NO. GVRR-3 RELATED TO THE INSERVICE TESTING PROGRAM FOURTH AND THIRD 10-YEAR INTERVALS NINE MILE POINT NUCLEAR STATION, LLC EXELON GENERATION COMPANY. LLC NINE MILE POINT NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-220 AND 50-410

1.0 INTRODUCTION

By letter dated December 27, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17003A096), as supplement by letter dated April 6, 2017 (ADAMS Accession No. ML17096A553), Exelon Generation Company, LLC (the licensee) submitted Relief Request No. GVRR-3 to the U.S. Nuclear Regulatory Commission (NRC or the Commission). This request is applicable to the fourth 10-year inservice testing (IST) program interval at Nine Mile Point Nuclear Station (Nine Mile Point), Unit 1, and the third 10-year IST program interval at Nine Mile Point, Unit 2. The licensee requested an alternative test plan in lieu of certain IST requirements of the 2004 Edition of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) for the IST program at Nine Mile Point, Units 1 and 2.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1 ), the licensee requested to use proposed alternative Relief Request No. GVRR-3 on the basis that the alternative provides an acceptable level of quality and safety.

2.0 REGULATORY EVALUATION

Section 50.55a(f), "lnservice Testing Requirements," of 10 CFR requires, in part, that IST 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 or relief has been granted by the NRC pursuant to 10 CFR 50.55a(z)(1) or 10 CFR 50.55a(z)(2).

In proposing alternatives or requesting relief, 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 of 10 CFR allows the NRC to authorize alternatives and to grant relief from 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. GVRR-3 3.1.1 Applicable ASME OM Code The request proposes an alternative test plan in lieu of certain IST requirements of the 2004 Edition of the ASME OM Code for the IST program at Nine Mile Point, Unit 1, for the fourth 10-year IST interval and Nine Mile Point, Unit 2, for the third 10-year IST interval, which are both currently scheduled to end on December 31, 2018.

ASME OM Code Section ISTC-3630, "Leakage Rate for Other Than Containment Isolation Valves", states:

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." ASME OM Code Section 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 submittals, the licensee requested alternative testing for the following valves:

Table 1 Valve ID Function Category Class CKV-40-03 Core Spray (CS) A/C 1 CKV-40-13 cs A/C 1 CKV-40-20 cs A/C 2 CKV-40-21 cs A/C 1 CKV-40-22 cs A/C 1 CKV-40-23 cs A/C 2 CKV-38-165 Shut Down Cooling (SOC) A/C 2 CKV-38-166 soc A/C 2 CKV-38-167 soc A/C 2 CKV-38-168 soc A/C 2 CKV-38-169 soc A/C 1 CKV-38-170 soc A/C 1 CKV-38-171 soc A/C 1 CKV-38-172 soc A/C 1 2CSH*V108 High Pressure Core Spray (CSH) A/C 1 2CSH*MOV107 CSH A 1 2CSL*V101 Low Pressure Core Spray (CSL) A/C 1 2CSL*MOV104 CSL A 1 21CS*V156 Reactor Core Isolation Cooling (ICS) A/C 1

Table 1 Valve ID Function Category Class 21CS*V157 ICS A/C 1 2RHS*V16A Residual Heat Removal (RHS) A/C 1 2RHS*V16B RHS A/C 1 2RHS*V16C RHS A/C 1 2RHS*V39A RHS A/C 1 2RHS*V39B RHS A/C 1 2RHS*MOV104 RHS A 1 2RHS*MOV112 RHS A 1 2RHS*MOV113 RHS A 1 2RHS*MOV24A RHS A 1 2RHS*MOV24B RHS A 1 2RHS*MOV24C RHS A 1 2RHS*MOV40A RHS A 1 2RHS*MOV40B RHS A 1 2RHS*MOV67A RHS A 1 2RHS*MOV67B RHS A 1 3.1.3 Reason for Request In its submittals, the licensee stated, in part:

ISTC-3630 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 for 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), but are not within the Appendix J scope since the Reactor Shutdown Cooling System valves are considered water-sealed.

The Nine Mile Point, Unit 1 (NMP1) leakage rate testing program is in accordance with NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 0, dated July21, 1995.

The Nine Mile Point, Unit 2 (NMP2) Technical Specifications contain a requirement to establish the leakage rate testing program in accordance with the guidelines contained in NEI 94-01, Revision 2-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated October 2008.

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. That justification shows that for 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. Further, it states that if the component does not fail within two

operating cycles, further failures appear to be governed by the random failure rate of the component. 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)."

The valves identified in this relief request are all in water applications. Testing is performed with water pressurized to pressures lower than function maximum pressure differential; however, the observed leakage is adjusted to the function maximum pressure differential value in accordance with ISTC 3630(b)(4). This relief request is intended to provide for a performance-based scheduling of PIV tests at NMP1 and NMP2. The reason for requesting this relief is dose reduction to conform with NRC and industry As Low As Reasonably Achievable (ALARA) radiation dose principles. The nominal fuel cycle lengths at NMP1 and NMP2 are 24 months. However, since refueling outages may be scheduled slightly beyond 24 months, a 4-1/2 year period is used to provide a bounding timeframe to encompass two refueling outages. The review of recent historical data identified that PIV testing each refueling outage results in a total personnel dose of approximately 1 Rem [roentgen equivalent man], assuming all of the PIVs remain classified as good performers. The proposed extended test intervals would provide for a savings of approximately 1 Rem over an approximate 4 year period (two refuel outages).

NUREG-0933, "Resolution of Generic Safety Issues," Issue 105, "Interfacing Systems LOCA at LWRs," discussed the need for PIV leak rate testing based primarily on three pre-1985 historical failures of applicable valves industry-wide.

These failures all involved human errors in either operations or maintenance.

None of these failures involved inservice equipment degradation. The performance of PIV leak rate testing provides assurance of acceptable seat leakage with the valve in a closed condition. Typical PIV testing does not identify functional problems which may inhibit the valves ability to reposition from open to closed. For check valves, functional testing is accomplished in accordance with ASME OM Code Section ISTC-3520, "Exercising Requirements," and Section ISTC-3522, "Category C Check Valves." For power-operated valves, testing is full stroke testing in accordance with the ASME OM Code to ensure their functional capabilities. Performance of the separate two-year PIV leak rate testing does not contribute any additional assurance of functional capability; it only determines the seat tightness of the closed valves.

3.1.4 Proposed Alternative In its submittals, the licensee stated:

NMP1 and NMP2 propose to perform PIV testing 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 process under 10 CFR 50 Appendix J, Option B.

A conservative control will be established such that if any valve fails either PIV test, the test interval for both tests will be reduced consistent with Appendix J, Option B requirements until good performance is reestablished.

The primary basis for this relief request is the historically good performance of the PIVs.

The functional capability of the check valves is demonstrated by the open and close exercising. This testing is separate and distinct from PIV testing and is performed at a refuel outage frequency in accordance with ASME OM Code, Section ISTC-3522.

The extension of test frequencies will be consistent with the guidance provided for Appendix J, Type C leak rate tests as detailed in NEI 94-01, Revision 3-A, Paragraph 10.2.3.2, "Extended Test Interval," (as approved by letter dated June 8, 2012 (ADAMS Accession No. ML121030286)) which states:

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. Intervals for Type C testing may be increased to a specific value in a range of frequencies from 30 months up to a maximum of 75 months. Test intervals for Type C valves should be determined by a licensee in accordance with Section 11.0 Additional basis for this relief request is provided below:

• Separate functional testing of motor-operated valve (MOV) PIVs and Check Valve PIVs per ASME OM Code will continue.

• The low likelihood of valve mis-positioning during power operations (e.g.,

procedures, interlocks).

• Relief valves in the low pressure (LP) piping - these relief valves may not provide Inter-System Loss of Coolant Accident (ISLOCA) mitigation for inadvertent PIV mis-positioning but their relief capacity can accommodate conservative PIV seat leakage rates.

• Alarms that identify high pressure (HP) to LP leakage - Operators are highly trained to recognize symptoms of a present ISLOCA and to take appropriate actions.

The proposed alternative will be utilized for the remainder of the third and fourth 120 month interval which is currently scheduled to end on December 31, 2018 for NMP1 and NMP2.

3.2 NRC Staff Evaluation The licensee has proposed an alternative test in lieu of the requirements found in the 2004 Edition of the ASME OM Code Section ISTC-3630(a) for 35 PIVs noted in Table 1 above.

Specifically, the licensee proposes to functionally test and verify the leakage rate of these PIVs using the 10 CFR Part 50, Appendix J, Option, B 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 Section ISTC-3630(a). Valves that have demonstrated good performance for two consecutive cycles may have their test interval extended to 75 months. Any PIV leakage test failure would require the component to return to the initial interval of every refueling outage, or 2 years, until good performance can again be established.

Pressure isolation valves 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, "ISLOCA [Inter System Loss of Coolant Accident] Research Program," July 1993. 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 designs. The conclusion of the analysis resulted in ISLOCA not being a risk concern for BWR design. Nine Mile Point, Units 1 and 2, are a BWR design.

Option B in 10 CFR Part 50, Appendix J, is a performance-based 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,"

September 1995 (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 50, Appendix J," dated July 26, 1995, 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 (ADAMS Package Accession No. ML122210254),

which allows Type C containment isolation valves test intervals to be extended to 75 months, with a permissible extension for non-routine emergent conditions of 9 months (i.e., 84 months total).

The NRC staff finds the guidance in NEI 94-01, Revision 3-A, acceptable (NRC letter to NEI dated December 6, 2012; ADAMS Accession No. ML12226A546), with the following conditions:

1) Extended interval for Type C local leak rate tests (LLRTs) may be increased to 75 months with the requirement that a licensee's 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 non-routine 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 BWR 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 licensee's 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.

The 35 PIVs are currently being leak-tested every refueling outage, or 2 years. Performance of the leakage test of the 35 PIVs is inconsistent with ALARA based on radiation exposure.

Overall completion of leak test requirements averages a dose of 1 Rem over a 4-year period.

As noted in the licensee's relief request proposal, the valves have maintained a history of good performance. Extending the leakage test interval based on good performance and the low risk factor as noted in NUREG/CR-5928 is a logical progression to a performance-based program.

Therefore, the NRC staff concludes that the licensee's proposed alternative provides an acceptable level of quality and safety.

4.0 CONCLUSION

As set forth above, the NRC staff finds that the proposed alternative described in Alternative Request No. GVRR-3 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 ).

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

Therefore, the NRC staff authorizes the proposed alternative in Relief Request No. GVRR-3 for the remainder of the fourth 10-year IST interval at Nine Mile Point, Unit 1, and the third 10-year IST interval at Nine Mile Point, Unit 2, which are both currently scheduled to end on December 31, 2018. The performance-based program interval shall not exceed 75 months, with the exception that non-routine emergent conditions may extend the program interval 9 months.

Principal Contributor: M. Farnan Date: May 30, 2017