ML17354A837

Relief Requests Associated with Fifth and Fourth Ten-Year Inservice Testing Intervals

ML17354A837
Person / Time
Site:	Nine Mile Point
Issue date:	12/20/2017
From:	David Helker Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML17354A837 (44)

Text

Exelon Generation December 20, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Nine Mile Point Nuclear Station, Units 1 and 2 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.55a Renewed Facility Operating License Nos. DPR-63 and NPF-69 NRC Docket Nos. 50-220 and 50-41 O

Subject:

Relief Requests Associated with Fifth and Fourth Ten-Year lnservice Testing Intervals Attached for your review are relief requests for Nine Mile Point Nuclear Station (NMPNS),

Units 1 and 2, associated with the fifth and fourth ten-year lnservice Testing (IST) intervals, respectively. The fifth and fourth intervals of the NMPNS, Units 1 and 2 IST program comply with the ASME OM Code, 2012 Edition. The fifth and fourth ten-year intervals for NMPNS, Units 1 and 2, both begin on January 1, 2019, and will conclude on December 31, 2028. We request your approval by December 20, 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 Nine Mile Point Nuclear Station, Unit 1 and the Fourth Ten-Year Interval for Nine Mile Point Nuclear Station, Unit 2 cc:

USNRC Region I, Regional Administrator NRC Senior Resident Inspector - NMP NRC Project Manager, NRR - NMP A. L. Peterson, NYSERDA

ATTACHMENT Relief Requests Associated with the Fifth Ten-Year Interval for Nine Mile Point Nuclear Station, Unit 1 and the Fourth Ten-Year Interval for Nine Mile Point Nuclear Station, Unit 2 Relief Request No.

Description Unit 1 and Unit 2 GVRR-3 Elapsed Time Between Successive Openings of PIVs Unit 1 ADS-VR-01 ERV Power Operated Valves Tested Each Cycle CRD-VR-01 Stroke Time Testing of SCRAM Discharge Volume Valves CTNH202-VR-01 H2/O2 Sample and Return Valve Stroke Time Testing Group 1 CTNH202-VR-02 H2/O2 Sample and Return Valve Stroke Time Testing Group 2 MS-VR-01 Reactor Pressure Vessel Safety Valve Testing RBCLC-PR-01 Alternate Frequency for Pump Testing Unit 2 MSS-VR-01 Reactor Pressure Vessel Safety Relief Valve Testing

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Units 1 and 2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

GVRR Elapsed Time Between Successive Openings of PIVs Page 1 of 6

1.

ASME Code Component(s) Affected Refer to GVRR-3 Tables 1 and 2, Affected Components

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 [are] 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.

4.

Reason for Request

Pursuant to 10 CFR 50.55a, Codes and standards, paragraph (z)(1), an alternative to the requirement of ASME OM Code ISTC-3630(a) is requested. The basis of the 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 Reactor Shutdown Cooling System valves are considered water-sealed.

The Nine Mile Point, Unit 1 (NMP1) leakage rate testing program is in accordance with the Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Revision 0, dated July 21, 1995.

The Nine Mile Point, Unit 2 (NMP2) Technical Specifications (TS) 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.

NEI 94-01, Paragraph 10.2.3.2, "Extended Test Interval," [as approved in the final safety evaluation for NEI 94-01, Revision 3, via letter dated June 8, 2012 (ADAMS Accession No. ML121030286)], 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 licensees allowable administrative limits. Elapsed time between the first and last tests

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Units 1 and 2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

GVRR Elapsed Time Between Successive Openings of PIVs Page 2 of 6 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.

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 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 NMP1 and NMP2. The reason for requesting this alternative 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, assuming all of the PlVs 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 paragraph ISTC-3522, "Category C Check Valves." For power-operated valves, full stroke functional testing is accomplished in accordance with the ASME OM Code paragraph ISTC-3521 Category A and Category B Valves. 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.

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Units 1 and 2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

GVRR Elapsed Time Between Successive Openings of PIVs Page 3 of 6

5.

Proposed Alternative and Basis for Use 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 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, paragraph ISTC-3522.

Note that NEI 94-01 is not the sole basis for this relief request, given 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 this proposed alternative is authorized and the PIVs continue to exhibit good performance, the PIV test frequency could be extended such that the leak test would not be required each refueling outage. 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."

Additional basis for this relief request is provided below:

• Separate functional testing of motor-operated valve (MOV) PIVs and Check Valve PIVs per the 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 primary basis for this alternative request is the historically good performance of the PIVs.

Since approval of the previous request, (as described in precedent 1 in Section 7), the leakage test intervals are based on performance; therefore, additional leakage test data is not included in this interval alternative request.

Based on valve performance history, there is continued assurance of valve operational readiness, as required by ASME OM-2012 Code, paragraph ISTC-3630. Therefore, this proposed alternative to extend the testing frequency consistent with the testing frequencies discussed within NEI 94-01, Revision 3-A, Paragraph 10.2.3.2, "Extended Test Interval," 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).

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Units 1 and 2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

GVRR Elapsed Time Between Successive Openings of PIVs Page 4 of 6

6.

Duration of Proposed Alternative This request, upon approval, will be applied to the NMP1 Fifth 10-Year and NMP2 Fourth10-year intervals, which are scheduled to begin January 1, 2019, and conclude on December 31, 2028.

7.

Precedence

1. This relief request was previously approved for the fourth and third 10-year intervals for NMP1 and NMP2, respectively, in letter from NRC (J. G. Danna) to Exelon Generation Company, LLC (B. C. Hanson), Nine Mile Point Nuclear Station, Units 1 and 2 - Re:

Alternative [GVRR-3] 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 (ML17136A112).

2. A similar relief request was approved for Fermi Power Station for the third IST Interval in a letter from NRC (R. J. Pascarelli) to Detroit Edison (J. M. Davis), "Fermi 2 - Evaluation of In-Service Testing Program Relief Requests VRR-011, VRR-012, and VRR-013 (TAC Nos. ME2558, ME2557, and ME2556), dated September 28, 2010 (ML102360570).

3. A similar relief request was approved for Quad Cities Nuclear Power Station, Units 1 and 2 for the fifth IST interval in a letter from NRC (J. Wiebe) to Exelon (M. J. Pacilio), "Quad Cities Nuclear Power Station, Units 1 and 2 - Safety Evaluation in Support of Request for Relief Associated with the Fifth 10-Year Interval Inservice Testing Program (TAC Nos.

ME7981, ME7982, ME7983, ME7984, ME7985, ME7986, ME7987, ME7988, ME7990, ME7991, ME7992, ME7993, ME7994, and ME7995), dated February 14, 2013 (ML13042A348).

4. A similar relief request was approved for Dresden Nuclear Power Station, Units 2 and 3 for the fifth IST interval in a letter from NRC (T. L. Tate) to Exelon (B. C. Hanson), "Dresden Nuclear Power Station, Units 2 and 3 - Relief Request to Use an Alternative from the American Society of Mechanical Engineers Code Requirements (CAC Nos. MF5089 andMF5090), dated October 27, 2015 (ML15174A303).

5. A similar relief request was approved for Peach Bottom Atomic Power Station, Units 2 and 3 for the fourth interval in a letter from NRC (D. A. Broaddus) to Exelon (B. C. Hanson),

"Peach Bottom Atomic Power Station, Units 2 and 3 - Safety Evaluation of Relief Request GVRR-2 Regarding the Fourth 10-Year Interval of the Inservice Testing Program (CAC Nos. MF7630 and MF7631)," dated September 21, 2016 (ML16235A340).

8.

References

1. NEI 94-01,Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Revision 3-A

2. NUREG 0933, Resolution of Generic Safety Issues, Section 3, Issue 105: "Interfacing Systems LOCA at LWRs (Rev. 4)

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Units 1 and 2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

GVRR Elapsed Time Between Successive Openings of PIVs Page 5 of 6

3. Letter from NRC (S. Bahadur) to NEI (B. Bradley), Final Safety Evaluation of Nuclear Energy Institute (NEI) Report, 94-01, Revision 3, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J (TAC No. ME2164), dated June 8, 2012 (ADAMS Accession No. ML121030286)

GVRR-3 Table 1 - Affected Components-Unit 1 Unit 1 Components System Code Class Category CKV-40-03 CS 1

A/C CKV-40-13 CS 1

A/C CKV-40-20 CS 2

A/C CKV-40-21 CS 1

A/C CKV-40-22 CS 1

A/C CKV-40-23 CS 2

A/C CKV-38-165 SDC 2

A/C CKV-38-166 SDC 2

A/C CKV-38-167 SDC 2

A/C CKV-38-168 SDC 2

A/C CKV-38-169 SDC 1

A/C CKV-38-170 SDC 1

A/C CKV-38-171 SDC 1

A/C CKV-38-172 SDC 1

A/C

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Units 1 and 2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

GVRR Elapsed Time Between Successive Openings of PIVs Page 6 of 6 GVRR-3 Table 2 - Affected Components - Unit 2 Components System Code Class Category 2CSH*V108 CSH 1

A/C 2CSH*MOV107 CSH 1

A 2CSL*V101 CSL 1

A/C 2CSL*MOV104 CSL 1

A 2ICS*V156 ICS 1

A/C 2ICS*V157 ICS 1

A/C 2RHS*V16A RHS 1

A/C 2RHS*V16B RHS 1

A/C 2RHS*V16C RHS 1

A/C 2RHS*V39A RHS 1

A/C 2RHS*V39B RHS 1

A/C 2RHS*MOV104 RHS 1

A 2RHS*MOV112 RHS 1

A 2RHS*MOV113 RHS 1

A 2RHS*MOV24A RHS 1

A 2RHS*MOV24B RHS 1

A 2RHS*MOV24C RHS 1

A 2RHS*MOV40A RHS 1

A 2RHS*MOV40B RHS 1

A 2RHS*MOV67A RHS 1

A 2RHS*MOV67B RHS 1

A

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

ADS-VR ERV Power Operated Valves Tested Each Cycle Page 1 of 12

1.

ASME Code Component(s) Affected The following Main Steam Electromatic Relief Valves (ERVs) are affected:

Component Description Class Category PSV-01-102A Main Steam ERV 1

B PSV-01-102B Main Steam ERV 1

B PSV-01-102C Main Steam ERV 1

B PSV-01-102D Main Steam ERV 1

B PSV-01-102E Main Steam ERV 1

B PSV-01-102F Main Steam ERV 1

B

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-3510, Exercising Test Frequency, states, in part: Power-operated relief valves shall be exercise tested once per fuel cycle.

ISTC-3700, Position Verification Testing, states, in part: Valves with remote position indicators shall be observed locally at least once every 2 yr to verify that valve operation is accurately indicated.

ISTC-5111, Valve Testing Requirements, states, in part: (a) Testing shall be performed in the following sequence or concurrently. If testing in the following sequence is impractical, it may be performed out of sequence, and a justification shall be documented in the record of tests for each test or in the test plan:

(1) leakage testing (2) stroke testing (3) position indication testing ISTC-5113, Valve Stroke Testing, states, in part: (a) Active valves shall have their stroke times measured when exercised in accordance with para. ISTC-3500.

ISTC-5114, Stroke Test Acceptance Criteria, states, in part: Test results shall be compared to the reference values established in accordance with para. ISTC-3300, ISTC-3310, or ISTC-3320.

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

ADS-VR ERV Power Operated Valves Tested Each Cycle Page 2 of 12

4.

Reason for Request

Pursuant to 10 CFR 50.55a, Codes and standards, paragraph (z)(1), an alternative is proposed to the power-operated relief valve test requirements of the ASME OM-2012 Code.

The basis of the request is that the proposed alternative would provide an acceptable level of quality and safety.

There are six ERVs installed on the main steam (MS) lines inside the drywell. Each ERV consists of a main valve, a pilot valve assembly, and a solenoid actuator (see Figure 1). The ERVs are opened by either signals from automatic actuation instrumentation or manually and, thus, do not rely on spring setpoints for valve actuation.

The ASME OM-2012 Code-required testing for the six ERVs would be satisfied by manually stroking open each ERV with the reactor at pressure once every operating cycle. It would be performed during plant startup following a refueling outage (RFO). Experience in the industry and at NMPNS, Unit No. 1 (NMP1), indicates that manually opening the ERVs during plant operation can increase the potential for main disc seat leakage and pilot valve seat leakage.

NMP1 experienced main disc seat leakage in March 2001 and pilot valve seat leakage in December 2002, both of which were attributed to debris on the seats caused by testing the valves using steam. Leakage from the main valve disc can cause increases in suppression pool (torus) temperature and level, requiring more frequent suppression pool cooling and pump-down operations, and diverts steam from the power generation steam cycle. Excessive leakage from the pilot valve can cause inadvertent opening of the main valve and impair its ability to re-close.

The proposed alternative will allow testing of the ERVs that is appropriate to demonstrate functionality without cycling the valves in place using reactor steam pressure. This is consistent with NUREG-0737, "Clarification of TMI Action Item Requirements," Item II.K.3.16, "Reduction of Challenges and Failures of Relief Valves," which recommended that the number of relief valve openings be reduced as much as possible and that unnecessary challenges should be avoided.

5.

Proposed Alternative and Basis for Use This relief request proposes an alternative to performing in-situ ERV steam pressure testing every RFO. The proposed alternative consists of a combination of offsite steam testing of the main valves, actuator cycling, and other inspections and maintenance activities. The proposed alternative would provide an acceptable level of quality and safety, as further discussed below.

=

System Description===

There are six Dresser model 1525VX solenoid-actuated, pilot-operated ERVs installed at NMP1. The ERVs are connected to the MS lines between the MS line flow restrictor and the inboard MS isolation valve. Each ERV has its own discharge pipe that is equipped with an acoustic monitor to detect flow noise and a thermocouple to sense discharge fluid temperature to monitor for valve actuation and/or leakage.

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

ADS-VR ERV Power Operated Valves Tested Each Cycle Page 3 of 12 The ERVs have two functional modes of operation: the automatic depressurization system (ADS) mode and the overpressurization relief mode. In the ADS mode, the ERVs depressurize the reactor vessel in the event of a small break loss of coolant accident (SBLOCA) by relieving steam to the torus, allowing the core spray system to inject (spray) cooling water into the reactor vessel. The ADS mode actuates on concurrent lo-lo-lo reactor water level and high drywell pressure signals. The six ERVs, three primary valves, and three backup valves are required to be operable for the ADS mode. Operation of three ERVs is sufficient to depressurize the reactor coolant system (RCS) and permit full core spray system flow.

The ERVs also provide overpressure protection (relief mode) for the reactor and MS piping by limiting reactor pressure during transients that result in a pressure increase. In the overpressurization relief mode, pressure switches that monitor reactor vessel pressure actuate six ERVs at staggered setpoints to ensure sufficient margin between the analyzed peak transient pressure and the lowest setpoint for the reactor head safety valves to prevent safety valve actuation during anticipated transients.

Valve Operation Steam under pressure from the reactor enters the main valve and passes upward around the disc guide. Steam enters the chamber below the main disc through a small orifice located in the disc retainer plate. Inlet steam pressure holds the main valve disc closed. A main disc spring is provided to keep the main valve disc in the closed position at low pressures or while depressurized. The pilot valve disc is held in the closed position by a pilot valve spring and steam pressure in the chamber below the pilot disc. When the solenoid actuator is energized, the actuator plunger depresses the pilot valve operating lever, thereby opening the pilot valve.

When the pilot valve is opened, steam is released through the outlet port at a faster rate than supplied through the inlet orifice. This causes the chamber below the main disc to depressurize, causing the valve to open. To close the valve, the solenoid actuator is de-energized, thereby closing the pilot valve and allowing steam pressure to reseat the main valve.

1. Exercise Test Frequency Alternative to ISTC-3510 The six ERVs are currently tested in accordance with approved relief request ADS-VR-01.

This includes the exercising of all six ERV solenoid actuators and the replacement of three (3) of the six (6) main valves each RFO. Inspections and precision preventative maintenance are performed each RFO for all six of the solenoid actuators and pilot valve assemblies.

For the proposed alternative, all six of the ERV solenoid actuators will be exercised each RFO, and two (2) of the six (6) main valves will be replaced with pretested spare valves each RFO such that all six valves will be replaced with pretested spare valves over a 6-year period with a six-month grace period. Inspections and precision preventive maintenance (described below) will be performed each RFO for all six of the solenoid actuators and pilot valve assemblies, with the IST requirements incorporated as part of the preventive maintenance activities. This combination of testing, inspections, and maintenance activities provides an acceptable level of quality and safety without requiring the six ERVs to be stroked with reactor steam during plant startup.

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

ADS-VR ERV Power Operated Valves Tested Each Cycle Page 4 of 12 Solenoid Actuator Each ERV solenoid actuator will be exercised each RFO. The closing stroke de-energizes the solenoid and allows the actuator to return to its fail-safe (closed) position. This test will be performed with the pilot valve and solenoid actuator mounted in their normal installed positions inside the drywell, which allows the solenoid actuator to be actuated electrically from the control room by placing the control switch in the Open position. The pilot valve operating lever and pilot valve stem will be secured in the Open position during this test to prevent damage to the pilot valve assembly, which could result from dry-stroking with no backpressure. The maintenance activities include detailed inspections of the electrical and mechanical components of the solenoid actuator.

NMP1 licensee event report (LER)03-001, Technical Specification Cooldown Rate Exceeded During Required Cooldown for a Failed Solenoid Actuated Pressure Relief Valve, reported an event involving an ERV that failed to open due to high resistance in the solenoid actuator cutout switch contacts. The high resistance contacts limited the current through the solenoid operating coil, which reduced the force that the plunger applied to the pilot valve operating lever. Further investigation and examination showed that the high contact resistance was due to the tin coating having been worn off the cutout switch contacts, allowing excessive contact oxidation to occur. Preventive maintenance activities now include inspection and cleaning of the cutout switch contacts, as necessary, to assure that the contact surfaces are clean and free of oxidation, corrosion, and discoloration. The contact tin plating is verified to be intact and not worn off exposing the copper base material. Associated springs and mechanisms are inspected, and the as-left contact resistances are verified. Resistance checks are performed on both actuator coils, and actuator operating currents during electrical actuation are verified to be within acceptance limits. These steps provide substantial indication that the solenoid actuator is capable of functioning as designed and producing its full output force.

Stroke timing of the solenoid actuator is not performed since the actuator is a sub-component of the total ERV. Degradation is monitored through the preventive maintenance inspections in lieu of trending millisecond stroke time variations.

Pilot Valve Each ERV pilot valve will be exercised each RFO when the new/refurbished pilot valve assembly is installed in the pilot housing. Note that the pilot valve housing is permanently welded to the outside of the ERV enclosure located in the drywell (see Figure 1). Removal and reinstallation of the pilot valve assembly does not affect the ERV main valve. The maintenance activities include inspections of the pilot valve assembly parts and the pilot valve housing interior to identify any damage or wear that could impair free movement of the stem or proper valve seating. Parts are refurbished or replaced as necessary. Cleanliness of parts and components and absence of foreign material are verified prior to reassembly.

NMP1 has experienced a stuck-open ERV event caused by improper maintenance. NMP1 LER 04-001, "Manual Reactor Scram and Cooldown Rate Exceeding Technical Specification Limits Due to Electromatic Relief Valve Failure to Close," reported an event involving an ERV that stuck open due to a maintenance error in which an extraneous gasket was installed in the pilot valve housing. This condition allowed steam to bypass the pilot valve seat, thereby preventing steam pressure from building up under the main valve disc to close the valve when

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

ADS-VR ERV Power Operated Valves Tested Each Cycle Page 5 of 12 given the closure signal. Appropriate precautions and instructions have been incorporated into the ERV maintenance procedure to ensure that the correct gasket is used and sufficient torque is applied to prevent steam from bypassing the pilot valve seat.

Prior to re-installing the pilot valve assembly inside the pilot housing, pilot stem/disc leak testing and freedom of movement and reseat functionality are demonstrated. A complete cleanliness inspection must be performed prior to installing the pilot valve assembly back into the housing. The housing is thoroughly cleaned and vacuumed to remove moisture and debris to minimize the potential for debris blocking or hindering pilot valve performance.

Following installation of the pilot valve assembly inside the housing, the pilot valve operating lever and pilot valve assembly freedom of movement and clearance adjustments are confirmed, followed by stroking the solenoid actuator plunger by hand to the full extent of travel. This ensures that the solenoid actuator plunger, pilot valve operating lever, and pilot valve assembly function as a unit, while eliminating the risk of damage resulting from electrically stroking the pilot valve in the absence of steam pressure (referred to as dry-stroking). The pilot valve freedom of movement check allows the pilot valve disc to return to its fail-safe (closed) position. NMP1 LER 00-04-01, "Manual Reactor Scram and Unusual Event Declaration Due to Stuck Open Electromatic Relief Valve and Failed Vacuum Breaker on Electromatic Relief Valve Discharge Line," reported an event involving an ERV that unexpectedly opened and would not reclose. The cause was attributed to a bent stem in the pilot valve assembly and partial disengagement of the pilot valve disc from the stem. It was determined that the pilot valve stem-disc separation had occurred as a result of dry-stroking the ERV pilot valve using the solenoid actuator. (Reference NRC Inspection Report 2000-008.)

Stroke timing of the pilot valve is not practical since the test is performed by hand and the pilot valve is a sub-component of the total ERV. Degradation of the pilot valve assemblies is monitored through the preventive maintenance inspections.

Main Valve A sampling program is proposed that will remove and replace two of the six ERV main valves with pre-tested spare main valves during each RFO, such that all six ERV main valves are replaced every three RFOs (approximately 6 years). Each ERV main valve will be stroke tested at an offsite steam test facility once every 6 years (three RFOs). A 6-month grace period would be allowed to accommodate variations in fuel cycle length and extended shutdown periods. The main valve testing will capture the exercise and stroke time test data required by the ASME OM-2012 Code.

Main valve testing will be performed at an offsite steam test facility. As shown in Figure 1, the main valve is housed in a heavy steel enclosure that is attached to the main steam line inlet flange. The pilot valve assembly is installed inside the pilot valve housing, and the housing is welded onto the outside of the enclosure and physically separated from the ERV main valve body. Thus, only the main valve of the ERV can be sent to the test facility. A spare pilot valve assembly and a spare solenoid actuator, both representative of the components used at the plant, will be installed at the test facility to allow testing the main valve. The valve will be installed on a test steam header in the same orientation as the plant installation. The test conditions at the test facility will be similar to those in the plant, including ambient temperature and steam conditions. The main valve will receive an initial seat leak test, a functional test to

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

ADS-VR ERV Power Operated Valves Tested Each Cycle Page 6 of 12 ensure it is capable of operating and closing, and a final seat leak test. Valve stroke time will be obtained during the exercise test. Valve seat tightness will be verified by a cold bar test, and if not free of fog, leakage will be measured and verified to be below specified acceptance criteria. This initial testing will be completed prior to plant startup from the RFO.

After initial testing, the main valves will be completely disassembled, inspected and refurbished, and then retested. The refurbished main valves will be stored at the offsite test facility and returned to the plant prior to the next scheduled use. The offsite test facility's storage requirements will ensure the valves are protected from physical damage. The valves will be stored in an area meeting ANSI/ ASME N45.2.2 Class B storage requirements, with the storage area temperature maintained between 50°F and 90°F. Maintaining the ERVs in a controlled environment during storage minimizes the potential for any valve degradation.

Prior to installation at the plant, the spare main valves will be inspected for foreign material and damage. The steam line and ERV discharge line openings will also be inspected to verify cleanliness and absence of foreign material. Procedural requirements ensure that the proper ERV inlet flange gasket separating ring thickness is provided so proper crush of the flexitallic gasket is achieved when the valve is installed. The valves are then installed and necessary connections completed, including connecting the vent tube and installing the enclosure cover and bellows assembly. Proper connections will be verified per procedure.

The four main valve discs that are not exercised during each RFO will have inspections and maintenance performed on their solenoid actuators and pilot valve assemblies as described above. Review of past surveillance testing and preventive maintenance history indicates that the ERV main valves are highly reliable. During the second 10-year IST interval (1986 to 1999), the ERVs were inspected and refurbished at 48-month intervals (every two RFOs).

From 1999 to 2004, the preventive maintenance interval for the ERV main valves was extended to 6 years, and since 2004, the preventive maintenance interval for the ERV main valves has been 10 years. These preventive maintenance activities have found the ERV main valves in excellent condition with no significant degradation noted. The table below lists the most recent preventive maintenance performed for each ERV main valve:

Table 1 ADS-VR-01: ERV Surveillance Test Valve Number Date of Last Preventive Maintenance PSV-01-102A (ERV-111) 2/17/2017 PSV-01-102B (ERV-112) 2/19/2015 PSV-01-102C (ERV-121) 9/10/2015 PSV-01-102D (ERV-122) 2/25/2015 PSV-01-102E (ERV-113) 2/20/2015 PSV-01-102F (ERV-123) 2/10/2017 Prior to 2011 the plant Technical Specifications (TS) required each ERV to be manually stroked open, once per operating cycle, until the downstream acoustic monitors or

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

ADS-VR ERV Power Operated Valves Tested Each Cycle Page 7 of 12 thermocouples indicated the valve was open and passing steam. However, following approval of the license amendment in a Safety Evaluation Report (SER) dated September 28, 2011, Nine Mile Point Nuclear Station, Unit No. 1 - Issuance of Amendment Regarding Changes to Modify Surveillance Requirements for Testing of the Main Steam Electromatic Relief Valves (TAC No. ME4849), the TS only requires manual operation of each ERV solenoid actuator each operating cycle. Therefore, no TS surveillance test data exists to show how the main valve will perform if not stroked, fully open, for a period beyond approximately 48 months, which is the present test interval. The only failure of an ERV to open during the last 20 years is the event reported in NMP1 LER 03-001, which was caused by a problem with the solenoid actuator for ERV-111, not with the main valve.

The proposed sampling program whereby two of the six ERV main valves will be removed and replaced with pre-tested spare main valves during each RFO is consistent with the requirements of ASME OM Code Case OMN-17, Alternative Rules for Testing ASME Class 1 Pressure Relief/Safety Valves. OMN-17 specifies that Class 1 pressure relief valves shall be tested at least once every 72 months (6 years) with a 6-month grace period and that a minimum of 20% of the valves from each valve group shall be tested within any 24-month interval. It also allows that the testing requirements may be satisfied by installing pretested valves to replace valves that have been in service provided the valves removed from service are tested prior to resumption of electric power generation and have been subjected to the specified maintenance.

Additionally, the NRC has previously authorized extensions of the Mandatory Appendix I, 5-year test interval, for testing ASME Class 1 pressure relief valves. For Nine Mile Point Unit 2 (NMP2), the NRC authorized the alternative described in Relief Request MSS-VR-01, Revision 1, to extend the test interval for Class 1 MS safety relief valves to 3 refueling cycles (approximately 6 years, plus 6 months grace), as documented in SER dated January 29, 2016. Also, for the James A. Fitzpatrick plant, the NRC authorized the alternative described in Relief Request VRR-06 Revision 1, to extend the test interval for Class 1 MS safety relief valves to 72 months (6 years) with a 6-month grace period, as documented in SER dated October 1, 2009, and associated license amendment dated July 21, 2010.

Based on the above discussion, extending the main valve exercising interval from every RFO (approximately 2 years) to every 6 years, plus a 6-month grace period, is reasonable and will not adversely impact the ability of the valves to perform their safety functions or result in additional valve failures. The testing and refurbishment activities performed at the off-site test facility on the partial compliment sample (two valves each RFO) will ensure that main valve degradation mechanisms are detected in a timely manner. Monitoring of the ERV discharge line temperatures during plant operation also provides an indication of degradation of the installed main valves.

2. Position Indication Verification Alternative to ISTC-3700 and ISTC-5111 This proposed alternative performs position indication verification for the six ERVs by observing the control room position indicating lights during the solenoid actuator test. Each ERV is equipped with red and green indicating lights, which provide control room open and closed indication, respectively, by monitoring the solenoid actuator plunger position. A blue indicating light is also provided in the control room, which monitors power to the solenoid actuator. The blue light is "On" when the solenoid is deenergized (valve closed) and "Off

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ADS-VR ERV Power Operated Valves Tested Each Cycle Page 8 of 12 when the solenoid is energized (valve open). As previously noted, the pilot valve operating lever and pilot valve stem will be secured in the open position during this test to prevent damage to the pilot valve assembly, which could result from dry-stroking with no backpressure. Solenoid actuator plunger movement will be observed locally in the drywell and compared to the control room indication to verify that solenoid actuator operation is accurately indicated. The proposed position indication verification alternative provides indirect pilot valve position, which ultimately represents the position of the main valve disc when steam is present, without cycling the ERVs in place with reactor steam pressure. This test is performed every RFO for each of the six ERVs.

The proposed position indication verification alternative provides an acceptable level of quality and safety without requiring indication of main valve obturator movement.

3. Stroke Time Testing Alternative to ISTC-5113 and ISTC-5114 Since the ERVs are not being in-situ tested, and since only the main valve is being tested at the offsite test facility (as previously noted), ERV full stroke time from initiating signal to indication of the end of the operating stroke cannot be obtained. Instead, main valve stroke times will be measured at the test facility. Stroke time acceptance criteria will use a pre-established reference value that represents good performance for this valve type. Since the whole valve assembly is not being tested and the test facility cannot duplicate the control circuitry, a simplified valve actuation circuit will be used. Although these differences may result in minor differences in measured stroke time compared to previous test data for in-situ testing of the complete ERV, the stroke times measured at the test facility will be comparable to each other and, thus, can be used to detect abnormalities in valve performance.

The proposed alternatives described above will maintain acceptable power-operated relief valve test accuracy and continue to provide an acceptable level of quality and safety; therefore, this proposed alternative is being requested pursuant to 10 CFR 50.55a(z)(1).

6.

Duration of Proposed Alternative This request, upon approval, will be applied to the NMP1 fifth 10-year interval, which begins on January 1, 2019, and is scheduled to end on December 31, 2028.

7.

Precedence

1. A similar relief request was previously approved for the fourth 10-year interval at NMP1, and all ERVs were removed and tested every four years, as documented in NRC safety evaluation, Nine Mile Point Nuclear Station, Unit No. 1 - Request for Alternative:

Automatic Depressurization System (ADS)-VR-01 for the Testing of Main Steam Electromatic Relief Valves Associated with the Fourth 10-Year Inservice Testing Interval (TAC No. ME4848), dated September 28, 2011 (ML112660001).

2. The NRC has authorized similar alternatives to the current method of ERV in-situ steam pressure testing, via the following NRC safety evaluations:

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

ADS-VR ERV Power Operated Valves Tested Each Cycle Page 9 of 12 Oyster Creek Nuclear Generating Station - Relief Request [RR-RV-53] Re: Reactor Inservice Testing of Main Steam Electromatic Relief Valves (TAC No. MC8672), dated August 31, 2006 - ADAMS Accession No. ML062220410.

Dresden Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2 - Issuance of Amendments for Main Steam Line Relief Valves and Associated Relief Requests [RV-021 and RV-030E, respectively] (TAC Nos. MC1792 through MC1795), dated October 19, 2004 - ADAMS Accession No. ML042600571 (ML042600563).

NMP1, Dresden, Quad Cities, and Oyster Creek all use the same Dresser model 1525VX solenoid actuated, pilot-operated ERVs. Similarities and differences between the proposed NMP1, alternative and the authorized alternatives for Dresden, Quad Cities, and Oyster Creek are summarized below:

Similarities A partial compliment of the ERV main valves will be removed and replaced with pre-tested spare main valves during each RFO.

Preventive maintenance is performed on all of the ERV solenoid actuators and their associated cutout switches during each RFO.

Differences NMP1 classifies the ERVs as OM Category B power-operated valves and tests the ERVs in accordance with the OM Code-2012, Subsection ISTC. The Dresden and Quad Cities relief requests cite both Subsection ISTC and Appendix I of the OM Code for ERV testing requirements, and the Oyster Creek relief request cites only Appendix I of the OM Code.

The proposed change for NMP1, includes a 6-month grace period on the 6-year interval for steam testing the ERVs at the offsite test facility. For Dresden, Quad Cities, and Oyster Creek, the ERV main valve testing is in accordance with Appendix I of the OM Code, which specifies a test interval of five years.

For Dresden, Quad Cities, and Oyster Creek, the offsite testing includes both the main valve and the pilot valve. For NMP1, the pilot valve cannot be tested with the main valve because of the unique heavy steel enclosure that houses the main valve. The pilot valve housing is welded onto the outside of the enclosure and is physically separated from the ERV main valve body.

For Dresden, Quad Cities, and Oyster Creek, the pilot valve assemblies for the ERVs that are not scheduled for removal and offsite testing are replaced with new or refurbished assemblies each RFO. For NMP1, all six of the ERV pilot valve assemblies will be replaced with new or refurbished assemblies each RFO.

For Dresden, Quad Cities, and Oyster Creek, the pilot valve is dry stroked using the solenoid actuator. For NMP1, dry stroking will not be performed due to a past event where dry stroking caused pilot valve damage. Instead, separate testing, inspections, and maintenance will be performed for the solenoid actuators and the pilot valve assemblies.

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

ADS-VR ERV Power Operated Valves Tested Each Cycle Page 10 of 12

8.

References

1. Letter from L.A. Hopkins (Niagara Mohawk Power Corporation) to Document Control Desk (NRC), LER 03-001, Technical Specification Cooldown Rate Exceeded During Required Cooldown for a Failed Solenoid Actuated Pressure Relief Valve, dated June 23, 2003 (ML031880016)

2. Letter from L.A. Hopkins (NMPC) to Document Control Desk (NRC), LER 04-001, "Manual Reactor Scram and Cooldown Rate Exceeding Technical Specification Limits Due to Electromatic Relief Valve Failure to Close," dated July 1, 2004 (ML041950181)

3. Letter from L. A. Hopkins (NMPC) to Document Control Desk (NRC), LER 00-[0]04, Supplement 1, "Manual Reactor Scram and Unusual Event Declaration Due to Stuck Open Electromatic Relief Valve and Failed Vacuum Breaker on Electromatic Relief Valve Discharge Line," dated December 6, 2000 (ML003777259)

4. Letter from M. G. Evans (NRC) to J. H. Mueller (NMPC), NRC's Nine Mile Point Inspection Report 05000220/2000-008, 05000410/2000-008, dated December 22, 2000 (ML003780274)

5. Letter from T. L. Tate (NRC) to B. Hanson (Exelon Generation Company, LLC), Nine Mile Point Nuclear Station, Unit 2 - Safety Evaluation by the Office of Nuclear Reactor Regulation for Relief Request MSS-VR-01, Revision 1 (CAC No. MF5773), dated January 29, 2016 (ML15345A006)

6. Letter from N. L. Salgado (NRC) to Vice President, Operations (Entergy Nuclear Operations, Inc.), James A. Fitzpatrick Nuclear Power Plant - Relief Request VRR-06, Revision 1 from the Requirements of the OM Code Re: Inservice Testing of Safety Relief Valves (TAC No. ME1818), dated October 1, 2009 (ML092730032)

7. Letter from B. K. Vaidya (NRC) to Vice President, Operations (Entergy Nuclear Operations, Inc.), James A. Fitzpatrick Nuclear Power Plant - Issuance of Amendment Regarding Testing of Safety/Relief Valves (TAC No. ME2810), dated July 21, 2010 (ML101750325)

8. NUREG-0737, "Clarification of TMI Action Item Requirements," Item II.K.3.16, "Reduction of Challenges and Failures of Relief Valves," November 1980 (ML051400209)

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

10. ANSI/ ASME N45.2.2, Packaging, Receiving, Storage, and Handling of Items for Nuclear Power Plants; QA Cases - December 1978

11. Letter from NRC (R. V. Guzman) to NMPNS (T. A. Lynch), Nine Mile Point Nuclear Station, Unit No. 1 - Issuance of Amendment [No. 210] Regarding Changes to Modify Surveillance Requirements for Testing of the Main Steam Electromatic Relief Valves (TAC No. ME4849), dated September 28, 2011 (ML112500067)

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

ADS-VR ERV Power Operated Valves Tested Each Cycle Page 11 of 12 ADS-VR-01 Figure 1 ERV with Enclosure as Installed at NMPNS, Unit 1 Expansion Bellows Assembly Small Bore Piping Threaded Union Enclosure Solenoid Actuator

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ADS-VR ERV Power Operated Valves Tested Each Cycle Page 12 of 12 ADS-VR-01 Figure 1 (Continued)

Standard ERV similar to other Plants (Cutaway view of internals is applicable to NMP1)

Below Disc Chamber Vent Tube (extended and routed to outside of enclosure for NMP application)

Pilot Valve Outlet Piping Pilot Valve Actuating Lever Disc Guide Retainer Plate Below Disc Chamber Inlet Orifice Main Disc Solenoid Actuator Plunger

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

CRD-VR Stroke Time Testing of SCRAM Discharge Volume Valves Page 1 of 2

1.

ASME Code Component(s) Affected The following Scram Discharge Volume (SDV) containment isolation valves (CIVs) are affected:

Component Description Class Category IV-44.2-15 SDV VENT INBOARD IV 2

A IV-44.2-16 SDV VENT OUTBOARD IV 2

A IV-44.2-17 SDV DRAIN OUTBOARD IV 2

A IV-44.2-18 SDV DRAIN INBOARD IV 2

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-5131, Valve Stroke Testing, states, in part, (a) Active valves shall have their stroke times measured when exercised in accordance with para. ISTC-3500.

4.

Reason for Request

Pursuant to 10 CFR 50.55a, Codes and standards, paragraph (z)(1), an alternative is proposed to the power-operated valve test requirements of the ASME OM-2012 Code.

The basis of the request is that the proposed alternative would provide an acceptable level of quality and safety.

The SDV CIVs are normally open valves. These valves close on the loss of air or the de-energizing of the solenoid valves (SOV-113-275 and SOV-113-276 for IV-44.2-16 and IV-44.2-17; and SOV-113-273 and SOV-113-274 for IV-44.2-15 and IV-44.2-18). The SDV air header and valve arrangement are single failure proof. The solenoid valves are powered from either reactor trip bus (RTB) 131 or 141 through fuses. Removing the fuses to fail-safe test these valves causes a scram in approximately six (6) seconds due to the de-energizing of SOV-113-271 and SOV-113-272. Venting the scram air header due to exercising of the valves by pulling fuses subjects the control rod drives to higher differential pressures than observed during a scram at normal operating conditions. The high differential pressure applied to control rods fully inserted has resulted in equipment damage.

Testing via the safety-related scram exhaust path cannot be performed during power operation since this could result in a plant trip. The safety-related exhaust path (scram path) is through SOV-113-275 and SOV-113-276 or SOV-113-273 and SOV-113-274 exhaust ports. A test solenoid valve (SOV-113-277) was installed as a result of Information Bulletin (IEB) No. 80-17 dated July 3, 1980, to permit fail-safe and stroke time

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

CRD-VR Stroke Time Testing of SCRAM Discharge Volume Valves Page 2 of 2 testing without causing a scram. The test solenoid exhaust path (test path) adds a restriction that is not present in the scram path. When the test solenoid is energized, the SDV air header and valve actuators are vented through SOV-113-277.

The restriction is due to exhausting air through the SOV-113-274 and SOV-113-276 air inlet supply port, since the solenoids are energized. The solenoid valve employs an internal pilot in the inlet port. Air can exhaust through the inlet port; however, the flow path is not a fixed resistance path. The variable resistance can cause variations in the quarterly stroke time measurements of the valves. These variations can result in inaccurate stroke times and mask the true valve performance. This limits the ability to accurately monitor for and detect degradation. Additionally, the test path is not the safety-related exhaust path (scram path) for the CIVs.

Stroke time testing through the scram path can be performed during refueling outages.

Stroke times obtained during refueling outage tests (using the scram vent path) have provided consistent accurate results. This testing method provides an accurate indication of valve performance and provides the ability to monitor for and detect degradation.

5.

Proposed Alternative and Basis for Use The SDV CIVs will be full stroke exercised and fail safe tested quarterly using the test solenoid valve. These valves will be stroke-time tested through the scram path during refueling outages.

The basis of the request is that the proposed alternative would provide an acceptable level of quality and safety; therefore, the alternative testing is proposed pursuant to 10 CFR 50.55a(z)(1).

6.

Duration of Proposed Alternative This request, upon approval, will be applied to the NMPNS, Unit No. 1, fifth 10-year interval, which begins on January 1, 2019, and is scheduled to end on December 31, 2028.

7.

Precedent This relief request was previously approved for the fourth 10-year interval at NMPNS, Unit No. 1, as documented in NRC safety evaluation, Nine Mile Point Nuclear Station - Safety Evaluation of Relief Requests [CRD-VR-01] for the Unit No. 1 Fourth 10-Year and Unit No.

2 Third 10-Year Pump and Valve Inservice Testing Program (TAC Nos. MD9202 and MD9203), dated December 29, 2008 (ML083500039).

8.

Reference NRC Information Bulletin (IEB) No. 80-17, Failure of 76 of 185 Control Rods to Fully Insert During Scram at a BWR, dated July 3, 1980 (ML8005050076)

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

CTNH202-VR H2/O2 Sample and Return Valve Stroke Time Testing Group 1 Page 1 of 3

1.

ASME Code Component(s) Affected The following Hydrogen and Oxygen sample and return containment isolation valves (CIVs) are affected:

Component Description Class Category IV-201.2-109

1. 11 TORUS RETURN INBOARD IV 2

A IV-201.2-110

1. 11 TORUS SAMPLE INBOARD IV 2

A IV-201.2-111

1. 11 TORUS SAMPLE OUTBOARD IV 2

A IV-201.2-112

1. 11 TORUS RETURN OUTBOARD IV 2

A IV-201.7-01

1. 11 SAMPLE STREAM B INBOARD IV 2

A IV-201.7-02

1. 11 SAMPLE STREAM B OUTBOARD IV 2

A IV-201.7-08 DW CAM SAMPLE INBOARD IV 2

A IV-201.7-09 DW CAM SAMPLE OUTBOARD IV 2

A IV-201.7-10

1. 11 DW RETURN INBOARD IV 2

A IV-201.7-11

1. 11 DW RETURN OUTBOARD IV 2

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-5131, Valve Stroke Testing, states, in part (a) Active valves shall have their stroke times measured when exercised in accordance with para. ISTC-3500.

ISTC-5132, Stroke Test Acceptance Criteria, states, Test results shall be compared to the reference values established in accordance with para. ISTC-3300, ISTC-3310, or ISTC-3320.

4.

Reason for Request

Pursuant to 10 CFR 50.55a, Codes and standards, paragraph (z)(1), an alternative is proposed to the power-operated valve test requirements of the ASME OM-2012 Code.

The basis of the request is that the proposed alternative would provide an acceptable level of quality and safety.

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

CTNH202-VR H2/O2 Sample and Return Valve Stroke Time Testing Group 1 Page 2 of 3 These pneumatically operated valves are grouped together on common control switches.

The groups are:

• IV-201.7-08, IV-201.7-09, IV-201.7-10, & IV-201.7-11

• IV-201.2-109, IV-201.2-112, IV-201.2-110, IV-201.2-111, IV-201.7-01, & IV-201.7-02 These arrangements have a common closed light (green) for a group of valves and individual open lights (red) for each valve. Reference values are established for each group by timing the valves for at least three exercises. The exercising is conducted over a sufficient interval to prevent erroneous data due to pre-conditioning. An individual reference value is developed for each valve in a group. A composite (group) reference value is developed by averaging the individual reference values. Typically, the individual valves reference values are within +/- 0.5 second of the group reference value.

As needed, primarily after rework or repair, the individual reference values and the group reference value are re-established. This group reference value is used as a common reference value for each valve in the group. The valve stroke-time test uses switch-actuation-to-red-light-out (closed indication) for open-to-close stroke time. The stroke-time of the slowest valve is observed and recorded. Typically, the slowest valve is not always the same component within the group. If the slowest valve exceeds the acceptance criterion (i.e., +/- 50% of the group reference value), the group is declared inoperable.

Corrective action is then taken, per ISTC-5133, Stroke Test Corrective Action.

The group reference values are less than 10 seconds, significantly below the Updated Final Safety Analysis Report (UFSAR), Table VI-3b, Primary Containment Isolation Valves Lines Entering Free Space of the Containment, maximum operating time of 60 seconds.

While some performance degradation is masked by this testing methodology, nuclear safety will not be compromised. Prior to any valve degrading and exceeding the UFSAR maximum operating time of 60 seconds, the acceptance criterion would be significantly exceeded, and corrective action would be taken. The proposed alternate testing method provides an adequate capability to monitor and detect individual valve degradation prior to exceeding the UFSAR maximum operating time. This method provides an equivalent level of quality and safety compared to the Code required individual valve stroke-timing.

5.

Proposed Alternative and Basis for Use NMPNS proposes to establish individual reference values, group reference values, and group acceptance criteria. Stroke-timing of the valve groups will record the slowest operating valves corresponding stroke-time. NMPNS will then compare the slowest valve stroke-time to the acceptance criterion to determine the valve group operability status.

Corrective actions will be taken, as required, for exceeding the acceptance criterion.

Therefore, relief is requested pursuant to 10 CFR 50.55a(z)(1) based on the proposed alternative providing an acceptable level of quality and safety.

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

CTNH202-VR H2/O2 Sample and Return Valve Stroke Time Testing Group 1 Page 3 of 3

6.

Duration of Proposed Alternative This request, upon approval, will be applied to the NMPNS, Unit No. 1, fifth 10-year interval, which begins on January 1, 2019, and is scheduled to end on December 31, 2028.

7.

Precedent This relief request was previously approved for the fourth 10-year interval at NMPNS, Unit No. 1, as documented in NRC safety evaluation, Nine Mile Point Nuclear Station - Safety Evaluation of Relief Requests [CTNH202-VR-01] for the Unit No. 1 Fourth 10-Year and Unit No. 2 Third 10-year Pump and Valve Inservice Testing Program (TAC Nos. MD9202 and MD9203), dated December 29, 2008 (ML083500039).

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

CTNH202-VR H2/O2 Sample and Return Valve Stroke Time Testing Group 2 Page 1 of 2

1.

ASME Code Component(s) Affected The following Hydrogen and Oxygen sample and return containment isolation valves (CIVs) are affected:

Component Description Class Category IV-201.2-23

1. 12 TORUS SAMPLE INBOARD IV 2

A IV-201.2-24

1. 12 TORUS SAMPLE OUTBOARD IV 2

A IV-201.2-29

1. 12 DRYWELL SAMPLE INBOARD IV 2

A IV-201.2-30

1. 12 DRYWELL SAMPLE OUTBOARD IV 2

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-5151, Valve Stroke Testing, states, in part, (a) Active valves shall have their stroke times measured when exercised in accordance with para. ISTC-3500.

ISTC-5152, Stroke Test Acceptance Criteria, states, Test results shall be compared to the reference values established in accordance with para. ISTC-3300, ISTC-3310, or ISTC-3320.

4.

Reason for Request

Pursuant to 10 CFR 50.55a, Codes and standards, paragraph (z)(1), an alternative is proposed to the power-operated valve test requirements of the ASME OM-2012 Code.

The basis of the request is that the proposed alternative would provide an acceptable level of quality and safety.

These solenoid operated valves are grouped together on a common control switch. The group is:

IV-201.2-23, IV-201.2-24, IV-201.2-29, & IV-201.2-30 This arrangement has a common closed light (green) for each pair of valves and individual open lights (red) for each valve. A reference value is established for each pair by timing the valves for at least three exercises. The exercising is conducted over a sufficient interval to prevent erroneous data due to pre-conditioning. A composite (group) reference value is developed by averaging the valve pair reference values. Individual reference values are not established. These valves stroke in less than 2 seconds and are all

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

CTNH202-VR H2/O2 Sample and Return Valve Stroke Time Testing Group 2 Page 2 of 2 designated as rapid acting valves. A limiting value of 2 seconds is assigned to the group.

As needed, primarily after rework or repair, the individual reference values and the group reference value are re-established. This group reference value is used as a common reference value for each valve in the group. The valve stroke-time test uses switch-actuation-to-red-light-out (closed indication) for open-to-close stroke time. The stroke-time of the slowest valve is observed and recorded. Typically, the slowest valve is not always the same component within the group. If the slowest valve exceeds the acceptance criterion (i.e., +/- 50% of the group reference value), the group is declared inoperable.

Corrective action is then taken, per ISTC-5133, Stroke Test Corrective Action.

The group limiting value of 2 seconds is significantly below the Updated Final Safety Analysis Report (UFSAR), Table VI-3b, Primary Containment Isolation Valves Lines Entering Free Space of the Containment, maximum operating time of 60 seconds. While some performance degradation is masked by this testing methodology, nuclear safety will not be compromised. Prior to any valve degrading and exceeding the UFSAR maximum operating time of 60 seconds, the acceptance criterion would be significantly exceeded, and corrective action would be taken. The proposed alternate testing method provides an adequate capability to monitor and detect individual valve degradation prior to exceeding the UFSAR maximum operating time of 60 seconds. This method provides an equivalent level of quality and safety compared to the Code required individual valve stroke-timing.

5.

Proposed Alternative and Basis for Use NMPNS proposes to establish valve pair reference values, group reference values, and group acceptance criteria. Stroke-timing of the valve groups will record the slowest operating valves corresponding stroke-time. NMPNS will then compare the slowest valve stroke-time to the acceptance criterion to determine the valve group operability status.

Corrective actions will be taken, as required, for exceeding the acceptance criterion.

Therefore, relief is requested pursuant to 10 CFR 50.55a(z)(1) based on the proposed alternative providing an acceptable level of quality and safety.

6.

Duration of Proposed Alternative This request, upon approval, will be applied to the NMPNS, Unit No. 1, fifth 10-year interval, which begins on January 1, 2019, and is scheduled to end on December 31, 2028.

7.

Precedent This relief request was previously approved for the fourth 10-year interval at NMPNS, Unit No. 1, as documented in NRC safety evaluation, Nine Mile Point Nuclear Station - Safety Evaluation of Relief Requests [CTNH202-VR-02] for the Unit No. 1 Fourth 10-Year and Unit No. 2 Third 10-Year Pump and Valve Inservice Testing Program (TAC Nos. MD9202 and MD9203), dated December 29, 2008 (ML083500039).

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

MS-VR Reactor Pressure Vessel Safety Valve Testing Page 1 of 5

1.

ASME Code Component(s) Affected The following Reactor Pressure Vessel Safety Valves are affected:

Component Description Class Category PSV-01-119A Reactor Pressure Vessel Safety Valve 1

C PSV-01-119B Reactor Pressure Vessel Safety Valve 1

C PSV-01-119C Reactor Pressure Vessel Safety Valve 1

C PSV-01-119D Reactor Pressure Vessel Safety Valve 1

C PSV-01-119F Reactor Pressure Vessel Safety Valve 1

C PSV-01-119G Reactor Pressure Vessel Safety Valve 1

C PSV-01-119H Reactor Pressure Vessel Safety Valve 1

C PSV-01-119J Reactor Pressure Vessel Safety Valve 1

C PSV-01-119M Reactor Pressure Vessel Safety Valve 1

C

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

Division 1, Mandatory Appendix I, Inservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants, paragraph I-1320, Test Frequencies, Class 1 Pressure Relief Valves, subparagraph (a) 5-Yr Test Interval, states:

Class 1 pressure relief valves shall be tested at least once every 5 yr, starting with initial electric power generation. No maximum limit is specified for the number of valves to be tested within each interval; however, a minimum of 20% of the valves from each valve group shall be tested within any 24-mo interval. This 20% shall consist of valves that have not been tested during the current 5-yr interval, if they exist. The test interval for any installed valve shall not exceed 5 yr. The 5-yr test interval shall begin from the date of the as-left set pressure test for each valve.

4.

Reason for Request

Pursuant to 10 CFR 50.55a, Codes and standards, paragraph (z)(1), an alternative is proposed to the Class 1 pressure relief valve test requirements of the ASME OM Code Mandatory Appendix I, I-1320(a). The basis of the request is that the proposed alternative would provide an acceptable level of quality and safety.

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

MS-VR Reactor Pressure Vessel Safety Valve Testing Page 2 of 5 The nine (9) reactor pressure vessel safety valves provide Code-required overpressure protection to the reactor pressure vessel and the Class 1 reactor recirculation system and are located on the reactor vessel head inside the primary containment. In the event of main steam isolation valve (MSIV) closure, the safety valves are designed and sized to limit the pressure rise to 110% of the design pressure.

The Dresser Model 3777QA, spring-loaded safety valves have shown exemplary test history at NMPNS, Unit 1 (NMP1), as shown in MS-VR-01 Table 1, Reactor Head Safety Valve Test Results. However, given the current 24-month operating cycle, NMP1 is required to remove and test approximately fifty (50) percent of the safety valves every refueling outage (i.e., alternating between either four or five of the nine each outage), so that all valves are removed and tested every two refueling outages. This ensures compliance with the ASME OM Code requirements for testing Class 1 pressure relief valves within a 5-year interval. Approval of extending the test interval to 6 years with a grace period of six months would reduce the number of safety valves tested at NMP1 over three (3) refueling outages by at least four.

Code Case OMN-17, Alternative Rules for Testing ASME Class 1 Pressure Relief/Safety Valves, was published in the ASME OM-2009 Code Edition. This Code Case has not been approved for use in Regulatory Guide (RG) 1.192, Operations and Maintenance Code Case Acceptability, ASME OM Code, nor incorporated in 10 CFR 50.55a by reference; however, the NRC has allowed licensees to use ASME Code Case OMN-17, provided all requirements are met. Code Case OMN-17 allows the Owner to extend the test frequencies for Class 1 pressure relief valves to a 72-month (6-year) test interval, with a 6-month grace period, providing the requirements of the Code Case are satisfied. The Code Inquiry from Code Case OMN-17, references, in part, ASME OM Code 2001 Edition through the 2006 Addenda of Mandatory Appendix I, paragraph I-1320. NMP1 is preparing to implement the ASME OM-2012 Code Edition and has verified no revisions were made to Mandatory Appendix I in this edition, since the OMa-2006 Addenda of the ASME OM Code, that would affect the implementation of Code Case OMN-17. NMP1 currently meets or exceeds all the requirements specified in Code Case OMN-17.

5.

Proposed Alternative and Basis for Use As an alternative to the Code-required 5-year test interval per Mandatory Appendix I, paragraph I-1320(a), Exelon proposes that the subject Class 1 Reactor Pressure Vessel Safety 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. 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 the 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

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

MS-VR Reactor Pressure Vessel Safety Valve Testing Page 3 of 5 prior to returning the valve(s) to service. Each valve shall be disassembled and inspected prior to the start of the 72-month interval. Disassembly and inspection prior to the implementation of Code Case OMN-17 may be used.

The safety valve testing and maintenance cycle at NMP1 consists of removal of the safety valve 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, seat leakage and set pressure testing. Prior to the return of a complement of safety valves for installation in the plant, the valves are disassembled and inspected to verify the internal surfaces and parts are free from defects resulting from time related degradation or service induced wear prior to the start of the next test interval. During this process, any identified adverse conditions are corrected; damaged or worn parts, springs, gaskets and seals are replaced and the valve seats are lapped, if necessary; and the valve is reassembled.

Following reassembly, the valves set pressure is recertified with an acceptance criterion of +/-1%. This existing process is in accordance with ASME OM Code Case OMN-17, paragraphs (d) Maintenance and (e) Disassembly and Inspection.

After recertification testing, the safety valves are stored at the test facility for future use.

The storage area is inspected and maintained to the requirements of ANSI/ASME N45.2.2, Packing, Handling, Shipping, Storage and Handling of Items for Nuclear Power Plants, which will minimize the potential for any valve degradation.

NMP1 has reviewed the as-found set pressure test results for all of the safety valves tested since March 2011 as shown in Attachment 1. No safety valve tested exceeded the as-found set pressure test acceptance criterion of +/-3%. However, as required by Code Case OMN-17 paragraph (c), Requirements for Testing Additional Valves (same as Appendix I, I-1320(c)), NMP1 will expand the scope of safety valve testing upon an as-found set-pressure (first test actuation) failure. Specifically, for each valve tested for which the as-found set-pressure (first test actuation) exceeds the greater of either the plus/minus tolerance limit of the Owner established set-pressure acceptance criteria or +/-3% of valve nameplate set-pressure, two additional valves shall be tested from the same valve group.

Based on the valve performance history and safety valve maintenance practices, there is continued assurance of valve operational readiness, consistent with ASME OM-2012 Code, Mandatory Appendix I. Therefore, the proposed alternative for testing the subject Class 1 safety valves 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, including a 6-month grace period, would maintain acceptable valve operational readiness and 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 NMP1 Fifth 10-Year Interval, which begins on January 1, 2019, and is scheduled to end on December 31, 2028.

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

MS-VR Reactor Pressure Vessel Safety Valve Testing Page 4 of 5

7.

Precedence

1.

A similar relief request was authorized for use during the third 10-year interval at NMP2, as documented in letter from NRC (T. L. Tate) to Exelon Generation Company, LLC (B. Hanson), Nine Mile Point Nuclear Station, Unit 2 - Safety Evaluation by the Office of the Nuclear Reactor Regulation for Relief Request MSS-VR-01, Revision 1 (CAC No. MF5773), dated January 29, 2016 (ML15345A006).

[Authorized the 6-month grace afforded by Code Case OMN-17.]

2.

A similar relief request was authorized for use via MSS-VR-01 during the third 10-year interval for NMP2, as documented in letter from NRC (M. G. Kowal) to Nine Mile Point Nuclear Station, LLC (K. J. Polson), Nine Mile Point Nuclear Station - Safety Evaluation of Relief Requests [MSS-VR-01] for the Unit No. 1 Fourth 10-Year and Unit No. 2 Third 10-Year Pump and Valve Inservice Testing Program (TAC Nos.

MD9202 and MD9203), dated December 29, 2008 (ML083500039). [Authorized the Extended Test Interval to 6 years for Safety Valve testing.]

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 Inservice Testing Program (TAC Nos. MF2509 and MF2510), dated April 30, 2014 (ML14094A051).

8.

References

1.

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

2.

ANSI/ASME N45.2.2, Packing, Handling, Shipping, Storage and Handling of Items for Nuclear Power Plants

3.

RG 1.192, Operations and Maintenance Code Case Acceptability, ASME OM Code, Revision 1, dated August 2014

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

MS-VR Reactor Pressure Vessel Safety Valve Testing Page 5 of 5 MS-VR-01 Table 1 Reactor Pressure Vessel Safety Valve Test Results Safety Valve Tested Serial No.

Setpoint (psig)

As-Found Setpoint Test Results (psig)

Max Set Pressure

+3% (psig)

Min Set Pressure-3% (psig)

Accept/

Reject Time from last test (Years)

Refueling Outage 24, March 2017 PSV-01-119C BK6524 1218 1200 1254.5 1181.4 Accept 4

PSV-01-119F BK6254 1245 1239 1282.3 1201.8 Accept 4

PSV-01-119H BR08508 1218 1198 1254.5 1181.4 Accept 4

PSV-01-119J BK6325 1227 1226 1263.8 1190.1 Accept 4

Refueling Outage 23, March 2015 PSV-01-119A BK6267 1236 1214 1273.0 1198.9 Accept 2

PSV-01-119B BK6317 1227 1220 1263.8 1190.1 Accept 2

PSV-01-119D BL6303 1254 1233 1291.6 1219.3 Accept 2

PSV-01-119G BL6280 1236 1225 1273.0 1198.9 Accept 2

PSV-01-119M BK6250 1218 1214 1254.5 1181.4 Accept 2

Refueling Outage 22, March 2013 PSV-01-119A BK6535 1236 1239 1273.0 1198.9 Accept 2

PSV-01-119B BK6253 1227 1195 1263.8 1190.1 Accept 2

PSV-01-119C BK6520 1218 1236 1254.5 1181.4 Accept 2

PSV-01-119D BK6292 1254 1248 1291.6 1219.3 Accept 2

PSV-01-119F BK6297 1245 1230 1282.3 1201.8 Accept 2

PSV-01-119G BK6256 1236 1203 1273.0 1198.9 Accept 2

PSV-01-119H BK6521 1218 1235 1254.5 1181.4 Accept 2

PSV-01-119J BK6319 1227 1196 1263.8 1190.1 Accept 2

PSV-01-119M BK6522 1218 1194 1254.5 1181.4 Accept 2

Refueling Outage 21, March 2011 PSV-01-119A BK6267 1236 1273 1273.0 1198.9 Accept 2

PSV-01-119B BK6317 1227 1259 1263.8 1190.1 Accept 2

PSV-01-119C BK6524 1218 1223 1254.5 1181.4 Accept 2

PSV-01-119D BK6303 1254 1268 1291.6 1219.3 Accept 2

PSV-01-119F BK6254 1245 1251 1282.3 1201.8 Accept 2

PSV-01-119G BK6280 1236 1235 1273.0 1198.9 Accept 2

PSV-01-119H BR08508 1218 1245 1254.5 1181.4 Accept 2

PSV-01-119J BK6325 1227 1239 1263.8 1190.1 Accept 2

PSV-01-119M BK6250 1218 1246 1254.5 1181.4 Accept 2

Comments for MS-VR-01 Table 1:

Maintenance History: Prior to installation, each valve was fully disassembled, inspected, reassembled and as-left recertified by NWS Technologies, Spartanburg, South Carolina.

Number of as-found setpoint tests results acceptable: 27 Number of as-found setpoint tests results unacceptable: 0

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2),

RBCLC-PR Alternate Frequency for Pump Testing Page 1 of 3

1.

ASME Code Component(s) Affected The following Reactor Building Closed Loop Cooling (RBCLC) Pumps are affected:

Component Description Class Group PMP-70-01 Reactor Building Closed Loop Cooling Water #11 3

A PMP-70-02 Reactor Building Closed Loop Cooling Water #12 3

A PMP-70-03 Reactor Building Closed Loop Cooling Water #13 3

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

ISTB-3400, Frequency of Inservice Tests, states An inservice test shall be run on each pump as specified in Table ISTB-3400-1.

Table ISTB-3000-1, Inservice Test Parameters, provides the parameters for Flow Rate (Q) and Differential Pressure (P) for Group A pump testing.

Table ISTB-3400-1, Inservice Test Frequency, provides Group A Test frequency for Group A pumps as Quarterly.

ISTB-5121, Group A Test Procedure, states, in part, that Group A tests shall be conducted with the pump operating as close as practical to a specified reference point and within the variances from the reference point as described in this paragraph. The test parameters shown in Table ISTB-3000-1 shall be determined and recorded as required by this paragraph.

4.

Reason for Request

Pursuant to 10 CFR 50.55a, Codes and standards, paragraph (z)(2), an alternative is proposed to the pump test requirements of the ASME OM-2012 Code. The basis of the request is that compliance with the Code requirements results in hardship or unusual difficulty with no compensating increase in the level of quality and safety.

The RBCLC system is not a fixed resistance system. For the RBCLC system, no pump test loops or individual pump flow instrumentation is installed. Individual pump flow can only be determined by measuring system flow rate. The system flow rate and differential pressure are a function of the number of pumps running and system heat loads. During normal plant operations, system heat loads prevent removing the RBCLC system from service. Operating conditions do not permit single pump operation at repeatable test

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2),

RBCLC-PR Alternate Frequency for Pump Testing Page 2 of 3 conditions to allow individual pump parameters (i.e., flow rate and differential pressure) to be measured.

Therefore, during normal plant operation, operating a single RBCLC pump at a fixed reference condition (per ISTB-5121) to perform a Group A test (per ISTB-3400) would require reducing system heat loads and may result in a plant shutdown to cold shutdown conditions. Complying with the Code would require NMP1, to enter cold shutdown conditions every quarter where RBCLC system operating conditions allow single pump operation. Cold shutdown reduces system heat loads sufficiently to allow single RBCLC pump operation at a fixed reference condition and thus allows measurement of individual pump parameters (i.e., flow rate and differential pressure).

Obtaining flow rate and differential pressure measurements (parameters required by Table ISTB-3000-1 for an individual RBCLC pump on a quarterly basis poses a significant hardship (plant shutdown).

Alternatively, compliance could be achieved by a major system redesign and modification such as installation of individual pump test loops with flow instrumentation. This would allow a single pump to be removed from the system flow path and operated on a test flow path at Code required fixed reference conditions. Such a major system modification would be costly and burdensome with no compensating increase in the level of quality or safety.

5.

Proposed Alternative and Basis for Use Quarterly, during normal system operation, vibration (V) shall be measured for each RBCLC pump. During cold shutdowns, all the applicable parameters for a Group A test from Table ISTB-3000-1 (flow rate (Q), vibration (V), and differential pressure (P)) shall be measured for each RBCLC pump. The comprehensive test specified in Table ISTB-3400-1 will also be performed biennially. The testing alternative described above will still allow an adequate determination of pump operational readiness and permit detection of component degradation.

Therefore, relief is requested pursuant to 10 CFR 50.55a(z)(2) based on the determination that compliance with the Code required Group A pump test requirements cannot be achieved without major system modifications resulting in a hardship or unusual difficulty without a compensating increase in the level of quality and safety and the proposed alternative provides reasonable assurance of pump operational readiness.

6.

Duration of Proposed Alternative This request, upon approval, will be applied to the NMP1, fifth 10-year interval, which begins on January 1, 2019, and is scheduled to end on December 31, 2028.

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2),

RBCLC-PR Alternate Frequency for Pump Testing Page 3 of 3

7.

Precedent This relief request was previously approved for the fourth 10-year interval at NMP1, as documented in NRC safety evaluation, Nine Mile Point Nuclear Station - Safety Evaluation of Relief Requests [RBCLC-PR-01] for the Unit No. 1 Fourth 10-Year and Unit No. 2 Third 10-Year Pump and Valve Inservice Testing Program (TAC Nos. MD9202 and MD9203), dated December 29, 2008 (ML083500039).

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

MSS-VR Reactor Pressure Vessel Safety Relief Valve Testing Page 1 of 9

1.

ASME Code Component(s) Affected The following main steam safety relief valves (MSSRVs) are affected:

Component Description Class Category 2MSS*PSV120 MAIN STEAM SRV 1

C 2MSS*PSV121 MAIN STEAM SRV (ADS)*

1 C

2MSS*PSV122 MAIN STEAM SRV 1

C 2MSS*PSV123 MAIN STEAM SRV 1

C 2MSS*PSV124 MAIN STEAM SRV 1

C 2MSS*PSV125 MAIN STEAM SRV 1

C 2MSS*PSV126 MAIN STEAM SRV (ADS) 1 C

2MSS*PSV127 MAIN STEAM SRV (ADS) 1 C

2MSS*PSV128 MAIN STEAM SRV 1

C 2MSS*PSV129 MAIN STEAM SRV (ADS) 1 C

2MSS*PSV130 MAIN STEAM SRV (ADS) 1 C

2MSS*PSV131 MAIN STEAM SRV 1

C 2MSS*PSV132 MAIN STEAM SRV 1

C 2MSS*PSV133 MAIN STEAM SRV 1

C 2MSS*PSV134 MAIN STEAM SRV (ADS) 1 C

2MSS*PSV135 MAIN STEAM SRV 1

C 2MSS*PSV136 MAIN STEAM SRV 1

C 2MSS*PSV137 MAIN STEAM SRV (ADS) 1 C

• ADS = Automatic Depressurization System

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, paragraph I-1320, Test Frequencies, Class 1 Pressure Relief Valves, (a) 5-Yr Test Interval, states, Class 1 pressure relief valves shall be tested at least once every 5 yr, starting with initial electric power generation. No maximum limit is

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

MSS-VR Reactor Pressure Vessel Safety Relief Valve Testing Page 2 of 9 specified for the number of valves to be tested within each interval; however, a minimum of 20% of the valves from each valve group shall be tested within any 24-mo interval. This 20% shall consist of valves that have not been tested during the current 5-yr interval, if they exist. The test interval for any installed valve shall not exceed 5 yr. The 5-yr test interval shall begin from the date of the as-left set pressure test for each valve.

4.

Reason for Request

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

Appendix I, Section l-1320(a) of the ASME OM-2012 Code states, in part, that Class 1 pressure relief valves shall be tested at least once every 5 years, starting with initial electric power generation. This section also states a minimum of 20% of the pressure relief valves are tested within any 24-month interval and that the test interval for any individual valve shall not exceed 5 years. The required tests ensure that the Safety Relief Valves (SRVs), which are located on each of the Main Steam (MS) lines between the reactor vessel and the first isolation valve within the drywell, will open at the pressures assumed in the safety analysis.

The SRVs are Dikkers / Model G471-6/125.04 (18 SRVs installed) in four (4) steam lines.

The SRVs are designed to actuate by either of two modes - the safety mode or the ADS mode. In the safety mode, the valve will open when reactor pressure exceeds a specific spring set-pressure. In the ADS mode, the valve will automatically open upon receipt of an overpressure signal (seven of the eighteen valves support this ADS function). The proposed changes do not impact either the safety mode of operation or the ADS mode of operation.

The SRVs have shown acceptable test history at NMPNS Unit 2 (NMP2) as described in Section 5 below. However, given the current 24-month operating cycle, NMP2 would be required to remove and test approximately half of the SRVs every refueling outage in order to ensure that all valves are removed and tested in compliance with the ASME OM-2012 Code requirements for testing Class 1 pressure relief valves within a 5-year interval. With a 5-year interval, NMP2 would be required to remove all 18 SRVs over 2 refuel cycles (i.e.,

4 years). However, consistent with the previously approved alternative MSS-VR-01, Revision 1 (ML15345A006), approval of extending the test interval to 6.5 years will reduce the number of SRVs removed during an individual outage, such that the full scope of 18 SRVs are replaced over 3 refuel cycles (i.e., 6 years, plus 6 months grace). This is consistent with the test interval and grace period described in ASME Code Case OMN-17, Alternate Rules Class 1 Pressure/Safety Valves, and continues 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 produced the unintended consequence of additional testing burden. Without Code relief, the incremental outage work due to the inclusion of the additional 2 - 3 SRVs per outage would be contrary to the principle of maintaining radiation dose As Low As Reasonably Achievable (ALARA). The removal and

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

MSS-VR Reactor Pressure Vessel Safety Relief Valve Testing Page 3 of 9 replacement of the additional 2 - 3 SRVs per outage without Code relief results in an additional exposure of approximately 2 - 4 Rem each outage. Additionally, the grace period allows for flexibility in the scheduling of as-left and as-found set-pressure testing, which is based on a test-to-test frequency.

In accordance with 10 CFR 50.55a(z)(1), Exelon requests approval of an alternative to the 5-year test interval requirements of ASME OM Code, Appendix I, Section l-1320(a) for the SRVs at NMP2. Exelon requests that the test interval be increased from 5 years to 6.5 years. All other requirements of the applicable ASME OM Code would be met.

5.

Proposed Alternative and Basis for Use As an alternative to the Code required 5-year test interval per Appendix I, paragraph I-1320(a), Exelon 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. 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.

As-found testing using steam and subsequent valve maintenance are currently performed at an off-site test facility. Subsequent to completion of as-found testing, each SRV in the removed complement is disassembled to perform inspections and a complete valve overhaul. Any SRV that failed the as-found set-pressure test is inspected to determine the cause of the test failure. Valve overhaul is performed to ensure that parts are free of defects resulting from time related degradation or service induced wear. All identified adverse conditions are corrected, the disc and seats are lapped, and the valve is reassembled. Each SRV is then recertified for service through inspection and testing consistent with ASME OM Code requirements, including set-pressure, seat tightness, stroke time and disc lift verifications, solenoid coil pick up/drop out, and air actuator integrity tests.

After recertification testing, the SRVs are stored at the test facility for future use. The storage area is inspected and maintained to the requirements of ANSI/ASME N45.2.2, Packing, Handling, Shipping, Storage and Handling of Items for Nuclear Power Plants, which will minimize the potential for any valve degradation.

The SRV as-found set-pressure test data in MSS-VR-01 Tables 1 and 2 demonstrates that the maintenance practices previously employed by NMP2 were effective. In the Spring 2010, prior to the RFO12 refueling outage, testing of the SRVs began at an offsite testing facility. These results are reflected in MSS-VR-01 Table 2.

Only one as-found set-pressure test failure (2002) has been experienced during the time period encompassed by the data in MSS-VR-01 Table 1. Note that testing performed on SRVs removed during these RFOs utilized nitrogen, with a correlated set-pressure. The data in MSS-VR-01 Table 1 also illustrates that SRVs that have exceeded 6 years

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

MSS-VR Reactor Pressure Vessel Safety Relief Valve Testing Page 4 of 9 between tests still demonstrated acceptable as-found set-pressure test results.

Additionally, there were a total of four (4) as-found set-pressure failures experienced in 2010 (MSS-VR-01 Table 2), which were determined not to be a result of set-pressure drift or a hardware degradation issue. The cause, which was determined to be common among the four (4) SRVs, was minor inaccuracies associated with the correlation used to establish the as-left set-pressure using nitrogen, and the as-found set-pressure using saturated steam.

During RFO13 (Spring of 2012), RFO14 (Spring of 2014), and RFO15 (Spring of 2016),

there were no as-found set-pressure failures identified (MSS-VR-01 Table 3).

Based on the above valve performance history and SRV maintenance practices, there is continued assurance of valve operational readiness, as required by ASME OM-2012 Code, Mandatory Appendix I, paragraph I-1310(b). Therefore, this proposed alternative for testing the subject Class 1 pressure relief valves 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, including a 6-month grace period, 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 NMP2 Fourth 10-Year Interval, which is scheduled to begin January 1, 2019, and conclude on December 31, 2028.

7.

Precedence

1.

This relief request was previously authorized for use during the third 10-year interval at NMP2, as documented in letter from NRC (T. L. Tate) to Exelon Generation Company, LLC (B. Hanson), Nine Mile Point Nuclear Station, Unit 2 - Safety Evaluation by the Office of the Nuclear Reactor Regulation for Relief Request MSS-VR-01, Revision 1 (CAC No. MF5773), dated January 29, 2016 (ML15345A006).

2.

This relief request was previously authorized for use via MSS-VR-01 during the third 10-year interval for NMP2, as documented in letter from NRC (M. G. Kowal) to Nine Mile Point Nuclear Station, LLC (K. J. Polson), Nine Mile Point Nuclear Station -

Safety Evaluation of Relief Requests for the Unit No. 1 Fourth 10-Year and Unit No. 2 Third 10-Year Pump and Valve Inservice Testing Program (TAC Nos. MD9202 and MD9203), dated December 29, 2008 (ML083500039).

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 Inservice Testing Program (TAC Nos. MF2509 and MF2510), dated April 30, 2014 (ML14094A051).

4.

This alternative was authorized for use at NMP2 for the remainder of the second 10-year interval via letter from NRC (M. Banerjee) to Niagara Mohawk Power Corporation (J. H. Mueller), Nine Mile Point Nuclear Station, Unit 2 - Alternative [GV-RR-07] to

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

MSS-VR Reactor Pressure Vessel Safety Relief Valve Testing Page 5 of 9 American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Regarding Inservice Testing of Main Steam Safety/Relief Valves (TAC No. MB0290),

dated April 17, 2001 (ML010880286).

8.

References

1.

ASME OM Code Case OMN-17, Alternate Rules Class 1 Pressure/Safety Valves

2.

ANSI/ASME N45.2.2, Packing, Handling, Shipping, Storage and Handling of Items for Nuclear Power Plants

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

MSS-VR Reactor Pressure Vessel Safety Relief Valve Testing Page 6 of 9 MSS-VR-01 Table 1 Main Steam SRVs Test Results for NMP2 Refueling Outages RFO7 (2000) through RFO11 (2008) (1)

SRV Tested Serial No.

Set-Pressure (psig)

As-Found Set-Pressure Test Results (psig)

Correlated Set-Pressure (psig)

Correlated Max Set-Pressure

+3% (psig)

Correlated Min Set-Pressure

-3%

(psig)

Accept/

Reject Time from Last Test (Years)

Refueling Outage 11, April 2008 2MSS*PSV126 160965 1195 1212 1215 1248 1181 Accept 6.77 2MSS*PSV127 160956 1205 1220 1224.9 1258.7 1190.9 Accept 6.79 2MSS*PSV128 160972 1165 1162 1184 1217 1151 Accept 8.42 2MSS*PSV131 160961 1175 1200 1194 1227 1161 Accept 6.77 2MSS*PSV132 160915 1185 1182 1205 1238 1171.4 Accept 6.77 2MSS*PSV135 160964 1195 1197 1215 1248 1181 Accept 10.1 Refueling Outage 10, April 2006 2MSS*PSV120 160935 1185 1203 1205 1238 1171 Accept 6.44 2MSS*PSV121 160966 1195 1224 1215 1248 1181 Accept 6.40 2MSS*PSV122 160951 1185 1222 1204 1238 1171 Accept 8.11 2MSS*PSV125 160968 1185 1194 1205 1238 1171 Accept 8.12 2MSS*PSV129 160971 1205 1225 1225 1258 1191 Accept 6.42 2MSS*PSV133 160958 1165 1176 1184 1217 1151 Accept 4.78 Refueling Outage 9, April 2004 2MSS*PSV123 160960 1175 1191 1195.2 1228 1162 Accept 7.19 2MSS*PSV124 160974 1175 1193 1195.2 1228 1162 Accept 4.43 2MSS*PSV130 160936 1195 1193 1215.5 1249 1181 Accept 7.21 2MSS*PSV134 160954 1205 1225 1225 1259 1191.6 Accept 7.92 2MSS*PSV136 160973 1175 1189 1195.2 1228 1162 Accept 6.11 2MSS*PSV137 160905 1205 1239 1225.7 1259.7 1191.6 Accept 6.12 Refueling Outage 8, March 2002 2MSS*PSV121 160939 1195 1219 1214 1248 1180 Accept 4.09 2MSS*PSV126 160967 1195 1189 1214 1247 1180 Accept 5.88 2MSS*PSV127 160955 1205 1201 1224 1258 1190 Accept 5.90 2MSS*PSV128 160903 1165 1176 1184 1216 1151 Accept 4.08 2MSS*PSV129 160904 1205 1220 1224 1258 1190 Accept 4.08 2MSS*PSV132 160953 1185 1181 1204 1237 1171 Accept 5.16 2MSS*PSV134 160970 1205 1192 1224 1258 1190 Accept 5.18 2MSS*PSV135(2) 160976 1195 1170 1214 1247 1180 Reject 5.16 2MSS*PSV131(2) 160962 1175 1186 1194 1227 1161 Accept 5.17 2MSS*PSV133(2) 160959 1165 1169 1184 1216 1151 Accept 5.18

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

MSS-VR Reactor Pressure Vessel Safety Relief Valve Testing Page 7 of 9 MSS-VR-01 Table 1 (Continued)

Main Steam SRVs Test Results for NMP2 Refueling Outages RFO7 (2000) through RFO11 (2008) (1)

SRV Tested Serial No.

Set-Pressure (psig)

As-Found Set-Pressure Test Results (psig)

Correlated Set-Pressure (psig)

Correlated Max Set-Pressure

+3%

(psig)

Correlated Min Set-Pressure

-3%

(psig)

Accept/

Reject Time from Last Test (Years)

Refueling Outage 7, March 2000 2MSS*PSV120 160915 1185 1219 1204 1238 1171 Accept 3.90 2MSS*PSV121 160965 1195 1231 1215 1248 1181 Accept 3.90 2MSS*PSV122 160950 1185 1222 1204 1238 1171 Accept 3.94 2MSS*PSV123 160963 1175 1208 1194 1227 1161 Accept 3.92 2MSS*PSV124 160906 1175 1189 1194 1227 1161 Accept 3.92 2MSS*PSV125 160952 1185 1220 1204 1238 1171 Accept 3.90 2MSS*PSV128 160958 1165 1193 1184 1217 1151 Accept 3.94 2MSS*PSV129 160956 1205 1214 1225 1258 1191 Accept 3.16 2MSS*PSV135 160975 1195 1244 1215 1248 1181 Accept 3.88 2MSS*PSV136 160961 1175 1221 1194 1227 1161 Accept 3.93 2MSS*PSV137 160954 1205 1222 1225 1259 1191 Accept 3.88 Notes: (1) Testing was performed at the NMP2 onsite test facility using nitrogen as the test medium.

(2) SRV 2MSS*PSV135 (SN 160976) failed the as-found set-pressure test (relieved early) during Refueling Outage 8. Two additional valves (2MSS*PSV131 (SN 160962) and 2MSS*PSV133 (SN 160959)) were tested per Code requirements, and both passed. The cause for this failure was determined to be set-pressure drift.

Minor adjustments were made to restore the set-pressure to the acceptance range.

No additional causes for the set-pressure drift were found during valve maintenance. The valve was refurbished and re-certified.

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

MSS-VR Reactor Pressure Vessel Safety Relief Valve Testing Page 8 of 9 MSS-VR-01 Table 2 Main Steam SRVs Test Results for NMP2 Refueling Outage RFO12 (2010) (1)

SRV Tested Serial No.

Set-Pressure (psig)

As-Found Set-Pressure Test Results (psig)

Max Set-Pressure

+3%(psig)

Min Set-Pressure

-3% (psig)

Accept/

Reject Time from Last Test (Years)

Refueling Outage 12, April 2010(1) 2MSS*PSV120 160953 1185 1229 1220.5 1149.4 Reject 7.38 2MSS*PSV121 160967 1195 1226 1230.9 1159 Accept 7.29 2MSS*PSV122 160952 1185 1217 1220.5 1149.4 Accept 9.76 2MSS*PSV123 160914 1175 1193 1210 1139.7 Accept 13.03 2MSS*PSV124 160906 1175 1212 1210 1139.7 Reject 9.70 2MSS*PSV125 160950 1185 1220 1220.5 1149.4 Accept 9.75 2MSS*PSV126 160939 1195 1229 1230.9 1159 Accept 7.44 2MSS*PSV127 160905 1205 1234 1241 1174.4 Accept 5.30 2MSS*PSV128 160903 1165 1196 1199.9 1130 Accept 7.46 2MSS*PSV129 160904 1205 1234 1241 1174.4 Accept 7.46 2MSS*PSV130 160976 1195 1226 1230.9 1159 Accept 7.4 2MSS*PSV131 160974 1175 1192 1210 1139.7 Accept 5.29 2MSS*PSV132 160969 1185 1197 1220.5 1149.4 Accept 3.31 2MSS*PSV133 160959 1165 1214 1199.9 1130 Reject 7.44 2MSS*PSV134 160955 1205 1211 1241 1174.4 Accept 7.39 2MSS*PSV135 160936 1195 1222 1230.9 1159 Accept 5.30 2MSS*PSV136 160962 1175 1204 1210 1139.7 Accept 7.46 2MSS*PSV137 160970 1205 1245 1241 1174.4 Reject 7.36 Notes: (1) All 18 SRVs were removed and replaced with pre-tested valves in Refueling Outage 12 (2010). The testing was performed at an offsite test facility using saturated steam as the test medium.

EXELON GENERATION COMPANY, LLC IST PROGRAM - RELIEF REQUEST Nine Mile Point Nuclear Station, Unit 2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1),

MSS-VR Reactor Pressure Vessel Safety Relief Valve Testing Page 9 of 9 MSS-VR-01 Table 3 Main Steam SRVs Test Results for NMP2 Refueling Outages RFO13 (2012), RFO14 (2014), and RFO15 (2016)

SRV Tested Serial No.

Set-Pressure (psig)

As-Found Set-pressure Test Results (psig)

Max Set-Pressure

+3% (psig)

Min Set-Pressure

-3% (psig)

Accept/

Reject Time from Last Test (Years)

Refueling Outage 15, March 2016 2MSS*PSV120 160968 1185 1150 1220.55 1149.45 Accept 6

2MSS*PSV121 160966 1195 1222 1230.85 1159.15 Accept 6

2MSS*PSV124 160973 1175 1174 1210.25 1139.75 Accept 6

2MSS*PSV125 160951 1185 1175 1220.55 1149.45 Accept 6

2MSS*PSV127 160971 1205 1188 1241.15 1168.85 Accept 6

2MSS*PSV128 160958 1165 1167 1199.95 1130.05 Accept 6

Refueling Outage 14, March 2014 2MSS*PSV122 160935 1185 1181 1220.55 1149.45 Accept 4

2MSS*PSV123 160963 1175 1170 1210.25 1139.75 Accept 4

2MSS*PSV126 160964 1195 1164 1230.85 1159.15 Accept 4

2MSS*PSV134 160954 1205 1201 1241.15 1168.85 Accept 4

2MSS*PSV135 160975 1195 1199 1230.85 1159.15 Accept 4

2MSS*PSV137 160957 1205 1190 1241.15 1168.85 Accept 4

Refueling Outage 13, April 2012 2MSS*PSV129 160956 1205 1207 1241.15 1168.85 Accept 2

2MSS*PSV130 160965 1195 1196 1230.85 1159.15 Accept 2

2MSS*PSV131 160961 1175 1166 1210.25 1139.75 Accept 2

2MSS*PSV132 160915 1185 1182 1220.55 1149.45 Accept 2

2MSS*PSV133 160972 1165 1139 1199.95 1130.05 Accept 2

2MSS*PSV136 160960 1175 1168 1210.25 1139.75 Accept 2