ML17275A061

Relief Requests Associated with Fifth Ten-Year Inservice Testing Interval

ML17275A061
Person / Time
Site:	Peach Bottom
Issue date:	09/29/2017
From:	David Helker Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML17275A061 (19)

Text

200 Exelon Way Exelon Generation Kennett Squa1e, PA www.exeloncorp.com 193~8 10 CFR 50.55a September 29, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278

Subject:

Relief Requests Associated with Fifth Ten-Year lnservice Testing Interval Attached for your review are relief requests associated with the fifth ten-year lnservice Testing (IST) Interval for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3. The fifth interval of the PBAPS IST program is being prepared to comply with the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2012 Edition with no Addenda. The fifth ten-year IST Interval for PBAPS, Units 2 and 3, begins on August 15, 2018.

We request your approval by August 15, 2018.

There are no regulatory commitments contained within this letter.

If you have any questions concerning this letter, please contact Mr. David Neff at (610) 765-5631.

Sincerely, David P. Helker Manager - Licensing & Regulatory Affairs Exelon Generation Company, LLC

Attachment:

Relief Requests Associated with the Fifth Ten-Year Interval for Peach Bottom Atomic Power Station cc: USNRC Region I, Regional Administrator USNRC Senior Resident Inspector, PBAPS USNRC Project Manager, PBAPS R. R. Janati, Commonwealth of Pennsylvania S. T. Gray, State of Maryland

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUESTS ASSOCIATED WITH FIFTH TEN YEAR INSERVICE TESTING INTERVAL Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Relief Request No. Description GVRR-2 Changes to Pressure Isolation Valve Test Intervals 01 A-VRR-2 Main Steam Safety Relief Valves (SRVs) with ADS Function 01A-VRR-3 Use of ASME Code Case OMN-17 for Safety Relief Valve Testing 01A-VRR-4 Revise Main Steam Isolation Valve Partial Stroke Testing Frequency

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

GVRR-2 Changes to Pressure Isolation Valve Test Intervals

1. ASME Code Component(s) Affected Table GVRR-2 Valve System Code Class Category M0-2(3)-10-17 RHR 1 A M0-2(3)-10-18 RHR 1 A M0-2(3)-10-25A RHR 1 A M0-2(3)-10-258 RHR 1 A A0-2(3)-10-46A RHR 1 A/C A0-2(3)-10-468 RHR 1 A/C HV-2-10-23451 A/B RHR 1 A HV-3-10-33451 A/B RHR 1 A M0-2(3)-14-12A cs 1 A M0-2(3)-14-128 cs 1 A A0-2(3)-14-13A cs 1 A/C A0-2(3)-14-138 cs 1 A/C HV-2-14-29046A/B cs 1 A HV-3-14-39046A/B cs 1 A

2. Applicable Code Edition and Addenda

American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2012 Edition with no Addenda.

3. Applicable Code Requirement

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

ISTC-3630{a), Frequency, states, "Tests shall be conducted at least once every 2 yr."

Page 1 of 5

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

GVRR-2 Changes to Pressure Isolation Valve Test Intervals

4. Reason for Request

Pursuant to 10 CFR 50.55a, "Codes and standards," paragraph (z)(1 ), an alternative is proposed to the requirement of ASME OM Code ISTC-3630(a). The basis of the proposed request is that the proposed alternative would provide an acceptable level of quality and safety.

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 Residual Heat Removal (AHR) valves are considered water-sealed and the Core Spray (CS) system is not exposed to containment atmosphere.

Peach Bottom Atomic Power Station (PBAPS) Technical Specification (TS) 5.5.12, "Primary Containment Leakage Rate Testing Program," states, in part:

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in NE/ 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 3-A, dated July 2012 ...

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

Guidance for implementation of acceptable leakage rate test methods, procedures, and analyses is provided in NRC Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995 (ML003740058). RG 1.163 endorses Nuclear Energy Institute (NEI) 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision O, dated July 26, 1995, as an acceptable method for complying with the provisions of Option B to 10 CFR 50, Appendix J, with certain limitations. The current approved version of NEI 94-01 is Revision 3-A (ML12221A202), which allows Type CCIV test intervals to be extended to 75 months, with permissible extension for non-routine emergent condition of 9 months (84 months total) with conditions.

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 Cl Vs, which have demonstrated good performance by passing their leak rate tests for two consecutive cycles, further failures would 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)."

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EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

GVRR-2 Changes to Pressure Isolation Valve Test Intervals 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 proposed alternative is intended to provide for a performance-based scheduling of PIV tests at PBAPS. The reason for this alternative is dose reduction to comport with NRC and industry As Low As Reasonably Achievable (ALARA) radiation dose principles. The nominal fuel cycle length at PBAPS Units 2 and 3 is 24 months. However, since refueling outages (RFOs) may be scheduled slightly beyond 24 months, a 4-year period plus 6 months grace allowance is used to provide a bounding timeframe to encompass two RFOs. The review of recent historical data identified that PIV testing each RFO results in a total personnel dose of approximately 550 millirem assuming the PIVs remain classified as good performers. The proposed extended test intervals would provide for a savings of approximately 1.1 Roentgen equivalent man (rem) over a two-RFO (4-1/2 year) period.

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-1980 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.

5. Proposed Alternative and Basis for Use PBAPS proposes to perform PIV testing at intervals ranging from every RFO to every third RFO. 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. If any valve fails its PIV test, the test interval for the test will be reduced consistent with Appendix J, Option B requirements until good performance is re-established.

The primary basis for this proposed alternative is the historically good performance of the PIVs. The historical test data that demonstrates acceptable PIV performance for the RHR and CS systems was provided in the previous proposed alternative for the PBAPS Fourth IST Interval for PIV Leakage Testing Frequency, dated April 27, 2016 (Reference 6). The data showed that there were no seat leakage failures of the PIVs. Following NRC approval of the previous interval relief requests, the leakage test intervals were established based on performance and the leakage test intervals remain consistent with the CIV process under 10 CFR 50 Appendix J, Option B.

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EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

GVRR-2 Changes to Pressure Isolation Valve Test Intervals The functional capability of the check valves is demonstrated by the opening and closing of the valves each refueling outage using an external air actuator, which is directly coupled to the valve shaft and disc. This test is separate and distinct from the PIV testing and is performed at a Cold Shutdown frequency in accordance with ASME OM Code, Section ISTC-3522.

Note that NEI 94-01 is not the sole basis for this request, given that NEI 94-01 does not address seat leakage testing with water. This document was cited as an approach similar to the requested alternative method. If the proposed alternative is authorized and the valves exhibit good performance, the PIV test frequency will be controlled similar to the method used in NEI 94-01 so that the test would not be required each RFO.

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, Paragraph 10.2.3.2, "Extended Test Interval," 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 are determined in accordance with NE/ 94-01, Section 11.0, "Basis for Performance and Risk-Based Testing Frequencies for Type A, Type B, and Type C Tests."

Additional basis for this request is provided below:

• Separate functional testing of motor-operated valve (MOV) PIVs and check valve PIVs are performed per the ASME OM Code.

• There is a 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.

Extending the leakage test interval based on good performance and low risk factor, as noted in NUREG/CR-5928, ISLOCA Research Program Final Report, is a logical progression to a performance-based program, and provides an acceptable level of quality and safety. Therefore, this proposed alternative is being submitted pursuant to 10 CFR 50.55a(z)(1 ).

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EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

GVRR-2 Changes to Pressure Isolation Valve Test Intervals

6. Duration of Proposed Alternative This request, upon approval, will be applied to the PBAPS, Unit Nos. 2 and 3 Fifth 10-year Intervals, which begin on August 15, 2018, and are scheduled to end on August 14, 2028.

7. Precedence

1. Fermi Power Station, Fermi 2 - Evaluation of In-Service Testing Program Relief Requests VRR-011, VRR-012, and VRR-013, dated September 28, 201 O (ML102360570)

2. Quad Cities Nuclear Power Stations, Units 1 and 2 - Safety Evaluation in Support of Request for Relief Associated with the Fifth 1O Year Interval lnservice Testing Program, dated February 14, 2013 (ML13042A348)

3. Dresden Nuclear Power Station, Units 2 and 3- Relief Request To Use An Alternative from the American Society of Mechanical Engineers Code Requirements, dated October 27, 2015 (ML15174A303)

4. Peach Bottom Atomic Power Station, Units 2 and 3 - Safety Evaluation of Relief Request GVRR-2 Regarding the Fourth 10-year Interval of the lnservice Testing Program, dated September 21, 2016 (ML16235A340)

8. References

1. NRC Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program"

2. NUREG-0933, "Resolution of Generic Safety Issues," Issue 105, "Interfacing Systems LOCA at LWRs"

3. NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision O, dated September 1995 (ML003740058)

4. NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, dated July 2012(ML12221A202)

5. NUREG/CR-5928, "ISLOCA Research Program Final Report," dated July 1993 (ML072430731) (non-publicly available)

6. Letter from Exelon Generation Company (D. P. Helker) to the NRC (Document Control Desk), "Submittal of Relief Request Associated with the Fourth lnservice Testing Interval - Pressure Isolation Valve Leakage Testing Frequency," dated April 27, 2016 (ML16118A414)

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EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) 01 A-VRR-2 Main Steam Safety Relief Valves (SRVs) with ADS Function

1. ASME Code Component(s) Affected Valve Description Class Category Unit RV-2-02-071 A ADS/Safety Relief Valve A 1 B/C 2 RV-2-02-071 B ADS/Safety Relief Valve B 1 B/C 2 RV-2-02-071 C ADS/Safety Relief Valve C 1 B/C 2 RV-2-02-071 G ADS/Safety Relief Valve G 1 B/C 2 RV-2-02-071 K ADS/Safety Relief Valve K 1 B/C 2 RV-3-02-071 A ADS/Safety Relief Valve A 1 B/C 3 RV-3-02-071 B ADS/Safety Relief Valve B 1 B/C 3 RV-3-02-071 C ADS/Safety Relief Valve C 1 B/C 3 RV-3-02-071 G ADS/Safety Relief Valve G 1 B/C 3 RV-3-02-071 K ADS/Safety Relief Valve K 1 B/C 3

2. Applicable Code Edition and Addenda

American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2012 Edition with no Addenda.

3. Applicable Code Requirement

Mandatory Appendix I, lnservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants, Section 1-341 O(d), Class 1 Main Steam Pressure Relief Valves With Auxiliary Actuating Devices, states, "Each valve with an auxiliary actuation device that has been removed for maintenance or testing and reinstalled after meeting the requirements of para. 1-331 O, shall have the electrical and pneumatic connections verified either through mechanical/electrical inspection or test prior to the resumption of electric power generation. Main disk movement and set-pressure verification are not required."

4. Reason for Request

Pursuant to 10 CFR 50.55a, "Codes and standards," paragraph (z)(2), an alternative is proposed to the requirement of ASME OM Code Mandatory Appendix I, Section l-3410(d).

This section requires in-situ exercising testing of the Automatic Depressurization System (ADS) Safety Relief Valves (SRVs) at reduced power operation. This in-situ test imposes an unnecessary challenge on the valves and has been linked to valve degradation (e.g.,

pilot and/or valve leakage). Pilot degradation, while not a concern with respect to the ADS safety function could, if severe enough, lead to SRV set-point drift, spurious actuation, and/or failure to properly re-seat. Such events have occurred at other boiling water reactors (BWRs) with similar SRVs. If any of these valves fail to re-close after testing, the plant would be placed in a loss-of-coolant accident (LOCA) condition requiring plant shutdown in accordance with Technical Specification 3.6.2.1, "Suppression Pool Average Temperature."

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EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) 01A-VRR-2 Main Steam Safety Relief Valves (SRVs) with ADS Function As originally stated in NUREG-1482, Rev. 0, Guidelines for lnservice Testing at Nuclear Power Plants, and NUREG-0626, Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in GE Designed Operating Plants and Near-Term Operating License Applications, the NRG staff recommends reducing the number of challenges to the ADS valves. NUREG-1482, Revision 2, Section 4.3.2.1, describes how several licensees have determined that in-situ testing of the SRV/SVs can contribute to undesirable seat leakage of the valve during subsequent plant operation.

ASME Code development has recognized that unnecessary challenges to ADS valves should be avoided. Paragraph ISTC-1200 of the ASME OM 2012 Code exempts safety and relief valves from the requirements of ISTC-3700, Valve Position Verification, and ISTC-3500, Valve Testing Requirements.

In-situ testing of ADS SRVs is not necessary because the remaining ADS and SRV tests provide an acceptable level of quality and safety. These remaining tests and the associated ADS SRV performance requirements provide adequate demonstration of ADS SRV operability as described below:

A. ASME OM Code Appendix I Setpoint/Leakage Testing These functional tests and inspections, performed on at least 50% of the SRVs each refueling outage, verify that the valves self-actuate to open and close at the required set pressure and that leakage is within specified limits. After as-found testing is completed, disassembly and inspection is performed and the valves are refurbished. Manually exercising of the valves via the solenoid mode (i.e., ADS mode) is performed after refurbishment and as-left set pressure and seat leakage testing is performed. Specifically, Procedure ST-M-016-220-2(3) (for PBAPS, Units 2 and 3) performs a leak check of the air accumulators and piping used for actuation of the ADS SRVs.

B. ADS Logic System Functional Test This test, performed once per 24 months, verifies the ability of the ADS system logic to initiate and sustain automatic operation of the ADS system during design accident conditions. The surveillance tests the logic by simulating Reactor Low Water Level and High Drywall Pressure conditions, times and verifies proper operation of the ADS Bypass time delay relay, and verifies ADS SRV solenoid valve circuit operability. These tests verify proper installation of the electrical connections following installation of the valve. Specifically, Procedure ST-1-016-220-2(3) (for PBAPS, Units 2 and 3) tests:

• The ADS logic/function from the Main Control Room

• The ADS logic/function from the ADS Alternative Control Panels

• The ADS logic/function from the HPCI Alternative Shutdown Panel

• The logic/function of the Backup Nitrogen Supply Valves Additionally, procedure M-001-006 is performed to verify the polarity of the SRV solenoid cables prior to the SRV solenoid being installed.

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EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) 01 A-VRR-2 Main Steam Safety Relief Valves (SRVs) with ADS Function C. ADS Leak Check This test, performed each refueling outage, verifies that the ADS instrument nitrogen accumulator leakage is low enough to ensure that there will be sufficient pneumatic pressure for design basis ADS SRVs operation. This leak check confirms the integrity of the air line connections following reinstallation of the valve.

D. SRV Cyclic Test This test, performed each refueling outage, verifies proper operation of the ADS solenoid valves and air operator.

These combined tests described above verify the required ADS critical components' performance requirements. This proposed alternative will only eliminate the post-installation stroke test. This ADS SRV function is considered to be extremely reliable based on the simplicity of this aspect of the SRV design and is supported by PBAPS and industry performance history.

5. Proposed Alternative and Basis for Use PBAPS proposes to continue performance of the ASME OM Code Appendix I setpoint/leakage testing, ADS logic system functional tests, ADS leak check, and an SRV cyclic test (as indicated in Section 4, items A through D above).

Station procedures are in place to ensure that proper installation of the electrical and air-line connection to each SRV is verified following install of the valves following maintenance. Testing at reduced or normal system pressure would be a hardship without a compensating increase in the level of quality and safety because cycling the valves with steam as the medium can cause seat leakage. Stroke testing at the test facility coupled with the testing performed following maintenance is an acceptable alternative to the remote testing requirement in the ASME OM Code. Therefore, the proposed alternative provides reasonable assurance that the valves are operationally ready and is thereby requested pursuant to 10 CFR 50.55a(z)(2).

6. Duration of Proposed Alternative This request, upon approval, will be applied to the PBAPS, Unit Nos. 2 and 3 Fifth 10-year Interval, which begins on August 15, 2018, and is scheduled to end on August 14, 2028.

7. Precedent Peach Bottom Atomic Power Station, Units 2 and 3 - Requests for Relief Associated with the Fourth lnservice Testing Interval [Request No. 01 A-VRR-2], dated September 3, 2008 (ML081790539)

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EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) 01 A-VRR-2 Main Steam Safety Relief Valves (SRVs) with ADS Function

8. References

1. NUREG-1482, Rev. O, Section 4.3.4, Guidelines for lnseNice Testing at Nuclear Power Plants

2. NUREG-1482, Revision 2, Section 4.3.2.1, Boiling Water Reactor Safety/Alternative Valve Stroke Testing

3. NUREG-0626, Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in GE Designed Operating Plants and Near-Term Operating License Applications Page 4 of 4

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) 01A-VRR-3, Use of ASME Code Case OMN-17 for Safety Relief Valve Testing

1. ASME Code Component(s) Affected Table 1 01 A-VRR-3 Valve Description Class Category Unit RV-2-02-070A Main Steam Safety Valve 1 c 2 RV-2-02-0708 Main Steam Safety Valve 1 c 2 RV-2-02-070C Main Steam Safety Valve 1 c 2 RV-2-02-071 A Main Steam Safety Relief Valve 1 c 2 RV-2-02-071 B Main Steam Safety Relief Valve 1 c 2 RV-2-02-071 C Main Steam Safety Relief Valve 1 c 2 RV-2-02-071 D Main Steam Safety Relief Valve 1 c 2 RV-2-02-071 E Main Steam Safety Relief Valve 1 c 2 RV-2-02-071 F Main Steam Safety Relief Valve 1 c 2 RV-2-02-071 G Main Steam Safety Relief Valve 1 c 2 RV-2-02-071 H Main Steam Safety Relief Valve 1 c 2 RV-2-02-071J Main Steam Safety Relief Valve 1 c 2 RV-2-02-071 K Main Steam Safety Relief Valve 1 c 2 RV-2-02-071 L Main Steam Safety Relief Valve 1 c 2 RV-3-02-070A Main Steam Safety Valve 1 c 3 RV-3-02-0708 Main Steam Safety Valve 1 c 3 RV-3-02-070C Main Steam Safety Valve 1 c 3 RV-3-02-071 A Main Steam Safety Relief Valve 1 c 3 RV-3-02-071 B Main Steam Safety Relief Valve 1 c 3 RV-3-02-071 C Main Steam Safety Relief Valve 1 c 3 RV-3-02-071 D Main Steam Safety Relief Valve 1 c 3 RV-3-02-071 E Main Steam Safety Relief Valve 1 c 3 RV-3-02-071 F Main Steam Safety Relief Valve 1 c 3 RV-3-02-071 G Main Steam Safety Relief Valve 1 c 3 RV-3-02-071 H Main Steam Safety Relief Valve 1 c 3 RV-3-02-071 J Main Steam Safety Relief Valve 1 c 3 RV-3-02-071 K Main Steam Safety Relief Valve 1 c 3 RV-3-02-071 L Main Steam Safety Relief Valve 1 c 3

2. Applicable Code Edition and Addenda

American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2012 Edition with no Addenda.

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EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a(z)( 1) 01A-VRR-3, Use of ASME Code Case OMN-17 for Safety Relief Valve Testing

3. Applicable Code Requirement

Mandatory Appendix I, lnservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants, Section 1-1320, Test Frequencies, Class 1 Pressure Relief Valves, paragraph (a), 5-Year Test Interval, states, in part, that "Class 1 pressure relief valves shall be tested at least once every 5 year, ... "

4. Reason for Request

Pursuant to 10 CFR 50.55a, "Codes and standards," paragraph (z)(1 ), an alternative is proposed to the requirement of ASME OM Code ASME OM Code, Appendix I, Section l-1320(a) Section ISTC-3200, lnservice Testing, states, in part, that "lnservice testing ...

shall commence when the valves are required to be operable to fulfill their required function(s) ... Subsection ISTC-5240, Safety and Relief Valves, states, in part, that "Safety and relief valves shall meet the inservice test requirements of Mandatory Appendix I of this division.

According to Appendix I, Subsection 1-1320, paragraph (a) of the ASME OM Code, Class 1 pressure relief valves shall be tested at least once every 5 years, starting with initial electric power generation. l-1320(a) states, in part, that "a minimum of 20% of the valves from each valve group shall be tested within any 24-mo interval ... The test interval for any installed valve shall not exceed 5 yr." The required tests ensure that the SRVs/SVs, which are located on each of the main steam (MS) lines between the reactor vessel and the first isolation valve within the drywall, will open at the pressures assumed in the safety analysis.

The ASME Code requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, the size and number of Safety Relief Valves (SRVs) and Safety Valves (SVs) are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the Reactor Coolant Pressure Boundary (RCPB).

The SRVs are Target Rock three-stage pilot operated safety/relief valves (11 SRVs per unit). The SVs are Dresser spring-loaded safety valves (3 SVs per unit). One new SV for each unit (RV-2-02-0?0C and RV-3-02-0?0C) were recently installed as part of the power up-rate modifications. These valves are the same make and model as the existing valves identified in the original relief request from the previous IST 10-year interval. The SRVs and SVs are located on the MS lines between the reactor vessel and the first isolation valve within the drywell. The SRVs can actuate by either of two modes: the safety mode or the depressurization mode. In the safety mode, the pilot disc opens when steam pressure at the valve inlet expands the bellows to the extent that the hydraulic seating force on the pilot disc is reduced to zero. Opening of the pilot stage allows a pressure differential to develop across the second stage disc, which opens the second stage disc, thus, venting the chamber over the main valve piston. This causes a pressure differential across the main valve piston, which opens the main valve. The SVs are spring-loaded valves that actuate when steam pressure at the inlet overcomes the spring force holding the valve disc closed. This satisfies the ASME Code requirement. The proposed changes do not impact the depressurization mode function of the SRVs.

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EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) 01A-VRR-3, Use of ASME Code Case OMN-17 for Safety Relief Valve Testing The SRVs/SVs have shown acceptable test history at PBAPS, Units 2 and 3. However, given the current 24-month operating cycle for each unit, PBAPS is required to remove and test approximately half of the SRVs/SVs every refueling outage to ensure that all valves are removed and tested in accordance with the ASME OM Code requirements.

This ensures compliance with the ASME OM Code requirements for testing Class 1 pressure relief valves within a five-year interval.

With the current five-year interval, PBAPS is required to remove all 14 SRVs/SVs over 2 refuel cycles (i.e., 4 years). Approval of extending the test interval to 6 years (plus 6 months grace, as needed) would reduce the number of SRVs/SVs removed during an individual outage, such that the full scope of 14 SRVs/SVs are replaced over 3 refuel cycles (i.e., 6 years). Without Code relief, the incremental outage work due to the inclusion of the additional 2 - 3 SRVs/SVs per outage would be contrary to the principle of maintaining radiation dose As Low As Reasonably Achievable (ALARA). The removal and replacement of the additional 2 - 3 SRVs/SVs per outage without Code relief results in an additional exposure of approximately 21/2 - 4 Roentgen equivalent man (rem) each outage.

The ASME OM Committee developed Code Case OMN-17, Alternative Rules for Testing ASME Class 1 Pressure Relief/Safety Valves, and published it in the 2009 and 2012 Editions of the OM Code. However, Code Case OMN-17 has not been added to Regulatory Guide 1.192, Operation and Maintenance Code Case Acceptability, ASME OM Code; nor is it included in 10 CFR 50.55a by reference. However, the NRC has allowed use of Code Case OMN-17, provided all requirements in the Code Case are met.

In accordance with 10 CFR 50.55a(z)(1 ), PBAPS requests approval of an alternative to the five-year test interval requirements of ASME OM Code, Appendix I, Section l-1320(a) for the SRVs/SVs for Units 2 and 3. PBAPS requests that the test interval be increased from five years to six years (plus 6 months grace, as needed) as described in ASME Code Case OMN-17. All other related requirements in the 2012 Edition of the ASME OM Code will be met.

5. Proposed Alternative and Basis for Use PBAPS proposes that the subject Class 1 pressure relief valves be tested at least once every three (3) refueling cycles (approximately 6 years/72 months) with a minimum of 20%

of the valves tested within any 24-month interval as an alternative to the Code-required 5-year test interval per Appendix I, paragraph 1-1320(a). This 20% would consist of valves that have not been tested during the current 72-month interval, if they exist. The test interval for any individual valve would not exceed 72 months except that a 6-month grace period is allowed to coincide with refueling outages to accommodate extended shutdown periods and certification of the valve prior to installation.

After as-found set-pressure testing, the valves shall be disassembled and inspected to verify that parts are free of defects resulting from time-related degradation or service induced wear. As-left set-pressure testing shall be performed following maintenance and prior to returning the valve to service. Each valve shall have been disassembled and Page 3 of 5

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) 01A-VRR-3, Use of ASME Code Case OMN-17 for Safety Relief Valve Testing inspected prior to the start of the 72-month interval. Disassembly and inspection performed prior to the implementation of this alternative based on Code Case OMN-17 may be used.

The relief valve testing and maintenance cycle at PBAPS consists of removal of the SRV/SV complement requiring testing and transportation to an off-site test facility. Upon receipt at the off-site facility, the valves are subject to an as-found inspection and set pressure testing. Prior to the return of the complement of SRVs/SVs for installation in the plant, the valves are disassembled and inspected to verify that internal surfaces and parts are free from defects or service-induced wear prior to the start of the next test interval.

During this process, anomalies or damage are identified and dispositioned for resolution.

Damaged or worn parts, springs, gaskets and seals are replaced as necessary. Following reassembly, the valve's set pressure is recertified. This existing process is in accordance with ASME OM Code Case OMN-17, paragraphs (d) and (e). PBAPS previously reviewed the as-found set point test results for all of the SRV/SVs tested since 2000 as detailed in Table 1 of the original alternative request associated with the PBAPS fourth IST interval dated July 29, 2013. The test results from the original request are summarized in the following paragraph.

Since 2000, the PBAPS, Units 2 and 3 SRVs/SVs have a history of 96 as-found lift tests.

Of these 96 tests, 99% were found within a +/-3% tolerance. During the PBAPS Fall 2012 Unit 2 outage, one of the MSSVs (S/N 1095) as-found lift pressure was identified to be 3.4% above the setpoint. An expanded scope removal of the other SV was performed with a satisfactory lift (the S/N 1093 as-found test (November 2012) was exactly at set pressure). An investigation was performed regarding the one valve that lifted outside of a

+/-3% tolerance. It was identified that the previous certification practices were not using the current best-known practices and procedure enhancements have been implemented to prevent recurrence. Additionally, PBAPS implemented enhanced subcomponent replacement and testing criteria to provide further assurance of repeatable as-found lift results.

Accordingly, the proposed alternative of increasing the test interval for the subject Class I pressure relief valves from 5 years to 3 fuel cycles (approximately 6 years/72 months) would continue to provide an acceptable level of quality and safety while restoring the operational and maintenance flexibility that was lost when the 24-month fuel cycle created the unintended consequences of more frequent testing. This proposed alternative will continue to provide assurance of the valves' operational readiness and provides an acceptable level of quality and safety pursuant to 10 CFR 50.55a(z)(1 ).

6. Duration of Proposed Alternative This request, upon approval, will be applied to the PBAPS, Unit Nos. 2 and 3 Fifth 10-year Interval, which begins on August 15, 2018, and is scheduled to end on August 14, 2028.

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EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) 01A-VRR-3, Use of ASME Code Case OMN-17 for Safety Relief Valve Testing

7. Precedence

1. Monticello Nuclear Generating Plant - Relief from the Requirements of the ASME Code for Operation and Maintenance of Nuclear Power Plants for the Fifth 10-year lnservice Testing Program Interval, dated September 26, 2012 (ML12244A272)

2. Quad Cities Nuclear Power Stations, Units 1 and 2 - Safety Evaluation in Support of Request for Relief Associated with the Fifth 1O Year Interval lnservice Testing Program," dated February 14, 2013 (ML13042A348)

3. Peach Bottom Atomic Power Station, Units 2 and 3 - Safety Evaluation of Relief Request 01A-VRR-3 Regarding the Fourth 10-Year Interval of the lnservice Testing Program," dated April 30, 2014 (ML14094A051)

8. References

1. ASME OM Code, Code Case OMN-17, Alternative Rules for Testing ASME Class 1 Pressure Relief/Safety Valves

2. Request for Relief from ASME OM Code 5-Year Test Interval for Class 1 Safety Relief Valves/Safety Valves (SRVs/SVs) (Relief Request 01 A-VRR-3), Table 1, Relief Valve

[Test] History, dated July 29, 2013(ML13211A054).

3. Regulatory Guide 1.192, Operation and Maintenance Code Case Acceptability Page 5 of 5

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a{z)(1) 01 A-VRR-4 Revise Main Steam Isolation Valve Partial Stroke Testing Frequency

1. ASME Code Component(s) Affected Valve Svstem Code Class CateQorv A0-2(3)-01 A-080A MS 1 A A0-2(3)-01 A-0808 MS 1 A A0-2(3)-01 A-080C MS 1 A A0-2(3)-01 A-0800 MS 1 A A0-2(3)-01 A-086A MS 1 A A0-2(3)-01 A-0868 MS 1 A A0-2(3)-01 A-086C MS 1 A A0-2(3)-01 A-0860 MS 1 A

2. Applicable Code Edition and Addenda

American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2012 Edition with no Addenda.

3. Applicable Code Requirement

ISTC-3520, "Exercising Requirements," Section ISTC-3521, "Category A and Category B Valves," states in part, that "Category A and Category B valves shall be tested as follows:

(b) if full-stroke exercising during operation at power is not practicable, it may be limited to part-stroke during operation at power and full-stroke during cold shutdowns.

4. Reason for Request

Pursuant to 10 CFR 50.55a, "Codes and standards," paragraph (z)(1 ), an alternative is proposed to the requirements of ASME OM Code ISTC-3521 (b). The basis of the relief request is the proposed alternative would provide an acceptable level of quality and safety.

An existing PBAPS lnservice Testing (IST) Program Cold Shutdown Justification (CSJ) 01 A-VCS-2, for full-stroke testing, under ISTC-3521 (c), will be modified to remove the existing quarterly partial stroke exercise testing of the Main Steam Isolation Valves (MSIVs), under ISTC-3521 (b). This will be done to address the potential for the valves to fully close inadvertently during the quarterly exercise testing. Full closure of the valves, at power, will cause a reactivity event and potential loss of power production of the affected unit. Challenges like these, and their potential consequence(s), have also been recognized in NUREG-1482, Revision 2. NUREG-1482, Section 2.4.5, Deferring Valve Testing to Cold Shutdown or Refueling Outages, discusses activities generating these challenges and states they should be considered impracticable, thereby supporting the CSJ principal arguments.

In PBAPS Technical Specification (TS) 3.3.1.1 - Reactor Protection System (RPS)

Instrumentation, Surveillance Requirement (SR) 3.3.1.1.9 - Channel Functional Test (CFT), the frequency of testing is stated as, "In accordance with the Surveillance Page 1 of 3

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) 01 A-VRR-4 Revise Main Steam Isolation Valve Partial Stroke Testing Frequency Frequency Control Program (SFCP)." The only practical method to perform the RPS CFT for the MSIV position switch input into the RPS logic is to actually stroke the MSIV. There are no other TS-compliant methods available without reducing reactor power and entering the normally inerted primary containment. This would result in unwarranted power reductions and personnel radiation exposures.

PBAPS has elected, due to recent documentation describing MSIV industry test failures, to utilize the SFCP for the CFT to extend the test frequency of the CFT in increments, over a period of time, up to two years. A two-year test frequency would coincide with refueling outages and eliminate stroking of MSIVs during power operation of the units.

In order to utilize the SFCP for this MSIV testing, the valves will have to be partial stroke exercised at power, for a number of years, to achieve the final goal of stroking at a two-year frequency. This methodology will allow for a progressively longer test interval until the final biennial testing interval is achieved. This test frequency change cannot be done with the CSJ in the IST Program, as the stroking of the valves in accordance with the SFCP would be in contradiction with the CSJ, which would not permit stroking of the valves during normal power operation (except for emergent issues such as post maintenance testing).

5. Proposed Alternative and Basis for Use PBAPS proposes to continue partial stroke exercising the MSIVs for the sole purpose of supporting the requirements of the SFCP testing intervals that would require progressively longer surveillance intervals until the final biennial testing frequency is achieved. The CSJ would restrict any other stroking of the MSIVs, except for emergent issues such as post maintenance testing. Both the CSJ and the SFCP are needed together to address the removal of the challenges of partial stroke exercising, as defined in the CSJ, to support safer and more reliable continued operation of the units. Therefore, the proposed alternative would provide an acceptable level of quality and safety pursuant to 10 CFR 50.55a(z)(1 ).

6. Duration of Proposed Alternative This request, upon approval, will be applied to the PBAPS, Unit Nos. 2 and 3 Fifth 10-year Interval, which begins on August 15, 2018, and is scheduled to end on August 14, 2028.

7. Precedent Peach Bottom Atomic Power Station, Units 2 and 3 - Safety Evaluation of Relief Request 01 A-VRR-4 Regarding the Fourth 10-year Interval of the lnservice Testing Program, dated April 28, 2017(ML17108A762).

Page 2 of 3

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) 01 A-VRR-4 Revise Main Steam Isolation Valve Partial Stroke Testing Frequency

8. References

1. NUREG-1482, Revision 2, "Guidelines for lnservice Testing at Nuclear Power Plants,"

published October 2013 (ML13295A020)

2. PBAPS Technical Specifications, Surveillance Requirement (SR) 3.3.1.1.9 - Channel Functional Test (CFT)

3. PBAPS Cold Shutdown Justification (CSJ) 01 A-VCS-2, for full and partial stroke testing MSIVs, under ISTC-3521 (b)

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