Text

License Amendment Request to Revise Technical Specification Surveillance Requirement 3.4.14.1 and Proposed Inservice Testing Alternative ALT-VR-02
	ML23121A267
Person / Time
Site:	Vogtle
Issue date:	05/01/2023
From:	Brown R; Southern Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	NL-23-0217
	Download: ML23121A267 (1)
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Regulatory Affairs 3535 Colonnade Parkway Birmingham AL 35243 205 992 5000 0D\, 2023 10 CFR 50.90 NL-23-0217 Docket Nos.: 50-424 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification Surveillance Requirement 3.4.14.1 and Proposed Inservice Testing Alternative ALT-VR-02 Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) requests an amendment to Facility Operating License Nos. NPF-68 and NPF-81 for the Vogtle Electric Generating Plant (VEGP) Units 1 and 2. The amendment request proposes to revise Technical Specifications (TS)

Surveillance Requirement (SR) 3.4.14.1. Specifically, the proposed change is a request to revise TS SR 3.4.14.1 to allow surveillance frequencies in accordance with the Inservice Testing Program.

Pursuant to 10 CFR 50.55a(z), the application also proposes an alternative to the testing frequencies in the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code, by adoption of 10 CFR 50 Appendix J performance-based testing frequencies. provides a description and assessment of the proposed license amendment. provides the existing TS pages marked to show the proposed changes. Attachment 2 provides revised (clean) TS pages. Attachment 3 provides existing TS Bases pages marked to show the proposed changes for information only. Attachment 4 provides a list of pressure isolation valves subject to testing under SR 3.4.14.1. Attachment 5 provides Inspection history for VEGP Pressure Isolation Valves. provides the request for an alternative to ASME OM Code.

Approval of the proposed amendment is requested by August 15, 2024, to support the Fall 2024 refueling outage for Unit 1. The proposed changes will be implemented within 90 days of issuance of the amendment. The proposed LAR revisions will be implemented prior to the Alternative.

In accordance with 10 CFR 50.91, SNC is notifying the state of Georgia of this license amendment request by transmitting a copy of this letter to the designated state official.

This letter contains no regulatory commitments. If you have any questions or if additional information is needed, please contact Amy Chamberlain at 205.992.6361.

U.S. Nuclear Regulatory Commission NL-23-0217 Page 2 I declare under penalty of perjury that the foregoing is true and correct.

Executed on the VW day of 0D\, 2023.

Respectfully submitted, R.Keith Brown Regulatory Affairs Director Southern Nuclear Operating Company RKB/dsp/cbg : Evaluation of the Proposed Change. Proposed Technical Specification Changes (Mark-Up) : Revised Technical Specification Changes. Proposed Technical Specifications Bases Changes (Mark-Up)

- For Information Only Attachment 4. Pressure Isolation Valves Subject to Testing Under SR 3.4.14.1. Pressure Isolation Valve Leakage Test History : Proposed Inservice Testing Alternative ALT-VR-02 cc: Regional Administrator, Region II NRR Project Manager - Vogtle 1 & 2 Senior Resident Inspector - Vogtle 1 & 2 State of Georgia Environmental Protection Division RType: CVC7000

Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification Surveillance Requirement 3.4.14.1 and Proposed Inservice Testing Alternative ALT-VR-02 Evaluation of the Proposed Change CONTENTS 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specifications Requirements 2.3 Reason for the Proposed Change 2.4 Description of the Proposed Change

3.0 TECHNICAL EVALUATION

3.1 Elimination of 18 Month Frequency 3.2 Elimination of Nine Month Frequency 3.3 Elimination of the Event-Based Frequency 3.4 Proposed Frequency is Consistent with Vogtle 3 and 4 TS 3.5 Conclusion

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 No Significant Hazards Consideration Determination Analysis 4.3 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

Attachments:

1.

Proposed Technical Specification Changes (Mark-Up) 2.

Revised Technical Specification Changes 3.

Proposed Technical Specifications Bases Changes (Mark-Up) - For Information Only 4.

Pressure Isolation Valves Subject to Testing Under SR 3.4.14.1 5.

Pressure Isolation Valve Leakage Test History

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1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Southern Nuclear Operating Company, Inc. (SNC) requests an amendment to Renewed Facility Operating License Nos. NFP-68 and NFP-81, for Vogtle Electric Generating Plant (VEGP), Units 1 and 2, respectively.

The proposed amendment would revise Technical Specification (TS) 3.4.14, "RCS Pressure Isolation Valve (PIV) Leakage," Surveillance Requirement (SR) 3.4.14.1, to only require testing at the frequencies specified in the Inservice Testing Program. The Inservice Testing Program complies with the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code).

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation The Reactor Coolant System (RCS) PIVs are two normally closed valves in series which separate the high-pressure RCS from attached lower pressure systems, such as the Residual Heat Removal (RHR) system, the Safety Injection (SI) system, and the Accumulators. Failure of both PIVs in a connection or excessive PIV leakage could lead to overpressurization of the low-pressure piping or components that could lead to a system rupture, potentially resulting in a loss of coolant accident (LOCA) outside of containment. The 1975 NRC WASH-1400 study (NUREG-75/014) identified potential intersystem LOCAs as a significant contributor to plant risk.

The low pressure systems connected to the RCS are designed and operated to limit and detect leakage past PIVs. PIV leakage would be detected by increasing low-pressure system level, temperature, or pressure indications or the lifting of relief valves.

The PIVs are required to be tested by both the TS and the ASME OM Code. A list of the PIVs subject to testing under SR 3.4.14.1 is provided in Attachment 4.

10 CFR 50.55a(f) requires operating plants to meet the inservice test requirements set forth in the ASME OM Code and addenda. ASME OM Code-2004 Edition with Addenda through OMb-2006, is the applicable Code for the VEGP 4th 10-Year Inservice Testing (IST) interval ending May 31, 2027. The following ASME OM Code requirement applies to seat leakage testing of PIVs subject to SR 3.4.14.1:

ISTC-3630 Leakage Rate for Other Than Containment Isolation Valves. Category A valves with a leakage requirement not based on an Owners 10 CFR 50, Appendix J program, shall be tested to verify their seat leakages within acceptable limits. Valve closure before seat leakage testing shall be by using the valve operator with no additional closing force applied.

(a) Frequency. Tests shall be conducted at least once every 2 years. to NL-23-0217 Evaluation of Proposed Change

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2.2 Current Technical Specifications Requirements VEGP TS Section 1.1 provides the following definition:

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

VEGP TS SR 3.4.14.1 states:

SURVEILLANCE FREQUENCY SR 3.4.14.1

NOTES--------------------------

1.

Not required to be performed in MODES 3 and 4.

2.

Not required to be performed on the RCS PIVs located in the RHR flow path when in the shutdown cooling mode of operation.

3.

RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.

Verify leakage from each RCS PIV is equivalent to 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS pressure 2215 psig and 2255 psig.

In accordance with the INSERVICE TESTING PROGRAM, and 18 months AND Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months (except for valves HV-8701A/B and HV-8702A/B)

AND For systems rated at less than 50% RCS design pressure, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation (except for valves HV-8701A/B and HV-8702A/B) to NL-23-0217 Evaluation of Proposed Change

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2.3 Reason for the Proposed Change The proposed change will align the VEGP TS PIV testing frequency with the ASME OM Code.

The VEGP TS SR 3.4.14.1 requires testing in accordance with the Inservice Testing (IST)

Program (i.e., the ASME OM Code as required by 10 CFR 50.55a(f)). In addition, the SR requires performances:

x Every 18 months, x

Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months (except for valves HV-8701A/B and HV-8702A/B), and x

For systems rated at less than 50% RCS design pressure, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation (except for valves HV-8701A/B and HV-8702A/B)

These additional performance requirements restrict the implementation of the Inservice Testing (IST) Program testing in accordance with the Code and approved Code cases, and application of alternatives in accordance with 10 CFR 50.55a(z) with prior NRC approval.

The testing history of the PIVs demonstrates that more frequent testing is not necessary to ensure the PIVs can perform their specified safety functions.

Eliminating the additional Frequencies would eliminate unnecessary testing, which would reduce occupational radiation exposure and potentially reduce the length of refueling outages.

The existing PIV testing frequency is inconsistent with the frequency for the same testing performed for VEGP Units 3 and 4. There are no significant differences between the Units 1 and 2 PIVs and the Units 3 and 4 PIVs; however, the Units 3 and 4 PIV testing (SR 3.4.15.1) does not require the more frequent testing specified in the Units 1 and 2 TS.

2.4 Description of the Proposed Change The proposed change will revise VEGP TS SR 3.4.14.1 to require performance in accordance with the Inservice Testing Program and all other frequency requirements are removed. The SR 3.4.14.1 Frequency will state:

In accordance with the INSERVICE TESTING PROGRAM. provides TS markups. provides the clean retyped TS.

The TS Bases are revised to reflect these changes. Markups of the TS Bases are provided in for information only. to NL-23-0217 Evaluation of Proposed Change

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3.0 TECHNICAL EVALUATION

The proposed change:

x Eliminates the requirement to periodically perform the SR every 18 months in addition to the Frequency specified in the ASME Code; x

Eliminates the requirement to perform the SR prior to entering Mode 2 whenever the unit has been in Mode 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months; and x

Eliminates the requirement to perform the SR within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation for systems rated at less than 50% RCS design pressure.

The elimination of these Frequencies also removes the need for the exceptions for valves HV-8701A/B and HV-8702A/B.

The proposed change to SR 3.4.14.1 has no effect on the TS PIV allowable leakage limits (i.e.,

the Surveillance Requirement itself). With the proposed change, only the frequency of the testing is affected, and RCS PIV testing would continue to be performed in accordance with the IST Program and the ASME Code.

3.1 Elimination of the 18 Month Frequency The proposed change would remove an overly restrictive 18 month Surveillance Frequency for performing SR 3.4.14.1. The current SR relies on the IST Program to establish the required methods for measurement, the method and criteria for evaluating results, any associated corrective actions, the personnel qualification requirements, and the requirements for record keeping and submittal of reports. However, even though the ASME Code establishes a testing frequency for PIV testing of 2 years, the TS includes a more restrictive requirements to perform the SR every 18 months.

The leakage rate testing performance of each VEGP PIV is shown in Attachment 5, "Pressure Isolation Valve Leakage Test History." The PIV testing performance has been excellent and there is no evidence that performing the SR at the IST Program frequency would result in the SR not being met.

The NRC regulations give precedence to the ASME Code when there are conflicts between the TS and the Code. 10 CFR 50.55a(f)(5)(ii), "IST program update: Conflicting IST Code requirements with technical specifications," states, "If a revised inservice test program for a facility conflict with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program."

In summary, the proposed change is acceptable because the more frequent performance requirement of 18 months is inconsistent with the ASME Code, could unnecessarily limit plant operation, and is not supported by the operating history. In addition, the proposed change is consistent with the regulations. to NL-23-0217 Evaluation of Proposed Change

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3.2 Elimination of the Nine Month Frequency The SR 3.4.14.1 Frequency requires the SR to be performed, "Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months." This has the effect of requiring performance of the SR every 9 months, but only if there is a mid-cycle outage of sufficient length to perform the test.

In addition, PIV testing to meet the 9 month frequency requirement prior to entering Mode 2 has been required once in the past 16 years, and the testing returned satisfactory results.

As discussed above, elimination of the more frequent performance of the SR is acceptable because the ASME Code, which provides all other aspects of the testing, specifies a longer testing frequency, and the more frequent testing is not warranted given the VEGP PIV test performance. There is no need to perform the testing every 9 months, as permitted by plant conditions, in order to verify that PIV leakage is within the limit.

3.3 Elimination of the Event-Based Frequency The SR 3.4.14.1 Frequency requires the SR to be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation for systems rated at less than 50% RCS design pressure. The purpose of the performance is to verify that the PIVs are closed or seated after being actuated. This test is redundant to other ASME OM Code testing (e.g., exercise testing per ASME OM Code Subsection ISTC-3520 and position indication verification per ASME OM Code Subsection ISTC-3700) and is unnecessary. PIV testing to meet the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post-actuation frequency has been required once in the past 10 years at VEGP, with satisfactory results.

In addition, there are other readily available indications that a PIV has failed to close or seat, such as low-pressure system level, temperature, or pressure indications or the lifting of relief valves. Performing the PIV leakage testing for this purpose is unnecessary and may result in higher occupational exposure.

3.4 Proposed Frequency is Consistent with the Vogtle 3 and 4 TS VEGP Units 3 and 4 are of the Westinghouse AP1000 design. The VEGP Units 3 and 4 TS 3.4.15, "RCS Pressure Isolation Valve (PIV) Integrity," is identical to the equivalent generic specification in the NRC-approved AP1000 Design Control Document, Revision 19, incorporated by reference into 10 CFR 52 Appendix D, Section III.A.

VEGP Units 3 and 4 SR 3.4.15.1 require performance of the same leakage testing, under the same conditions, and with the same acceptance criteria as the Units 1 and 2 SR.

SR 3.4.15.1 Verify leakage of each RCS PIV is equivalent to 0.5 gpm per nominal inch valve size up to a maximum of 5 gpm at an RCS pressure 2215 and 2255 psig.

The VEGP Units 3 and 4 SR 3.4.15.1 Frequency is "24 months," with no additional Frequencies requiring testing within 18 months, 9 months, or 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after actuation. In the to NL-23-0217 Evaluation of Proposed Change

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Units 1 and 2 TS, a fixed frequency such as "24 months" would be replaced with a reference to the Surveillance Frequency Control Program. Unlike Units 1 and 2, the VEGP Units 3 and 4 TS do not include a Surveillance Frequency Control Program. Given that the testing is required by the ASME OM Code and the current Units 1 and 2 TS Frequency, the proposed change retains the current SR Frequency that references the Inservice Testing Program.

The absence of the 18 month, 9 month, and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after actuation Frequencies in the VEGP Units 3 and 4 TS is attributable to differences between the Units 1 and 2 and Units 3 and 4 the design and licensing bases. SECY-90-016, "Evolutionary Light Water Reactor (LWR)

Certification Issues and their Relationship to Current Regulatory Requirements," dated January 12, 1990, presented the NRC staff's recommendations concerning proposed departures from current regulations for the evolutionary advanced light water reactors (ALWRs), such as the AP1000 design. One of the issues identified was protection from an intersystem LOCA.

SECY-90-016,Section II, Preventative Feature Issues, E, "Intersystem LOCA" stated the following in part regarding Pressurized Water Reactors (PWRs):

Future evolutionary ALWR designs can reduce the possibility of a loss-of-coolant accident (LOCA) outside containment by designing (to the extent practicable) all systems and subsystems connected to the reactor coolant system (RCS) to an ultimate rupture strength at least equal to the full RCS pressure.

For both BWRs and PWRs, EPRI states that low-pressure systems which could be over-pressurized by the RCS should be designed with sufficient margin to withstand full RCS pressure without structural failure.

For PWRs, relief valves sized to protect against overpressure transients, should be provided on the RHR system. RHR suction valves should be provided with permissive interlocks to prevent opening if RCS pressure exceeds RHR design pressure.

The staff concluded that designing, to the extent practicable, low-pressure systems to withstand full RCS pressure is an acceptable means for resolving this issue. However, the staff believes that for those systems that have not been designed to withstand full RCS pressure, evolutionary ALWRs should provide (1) the capability for leak testing of the pressure isolation valves, (2) valve position indication that is available in the control room when isolation valve operators are deenergized and (3) high-pressure alarms to warn control room operators when rising RCS pressure approaches the design pressure of attached low-pressure systems and both isolation valves are not closed. Imposition of these requirements exceed Commission regulations and guidance: therefore, the staff recommended that the Commission approve these positions for evolutionary ALWRs.

The staff noted that for some low-pressure systems attached to the RCS, it may not be practical or necessary to provide a higher system ultimate pressure capability for the entire low-pressure connected system. The staff will evaluate these exceptions on a case-by-case basis during specific design certification reviews. to NL-23-0217 Evaluation of Proposed Change

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The June 26, 1990, version of the SECY also stated:

The Commission (with all Commissioners agreeing) has approved the staff's position on intersystem LOCA provided that, as recommended by the ACRS, all elements of the low pressure system are considered (e.g., instrument lines, pump seals, heat exchanger tubes, and valve bonnets.)

While no definitive documentation could be located, the most likely reason the VEGP Units 3 and 4 TS do not contain the additional frequency requirements that appear in the VEGP Units 1 and 2 TS is due in part to satisfying the additional requirements discussed in SECY-90-016.

However, a comparison of the relevant VEGP Units 1 and 2 design to the VEGP Units 3 and 4 design determined that the simplified frequency is equally applicable to Units 1 and 2.

SECY-90-016,Section II, Preventative Feature Issues, E, "Intersystem LOCA," provides the following criteria for evaluating the design and administrative provisions that serve to limit and detect leakage past PIVs:

1)

Relief valves sized to protect against overpressure transients, should be provided.

2)

RHR suction valves should be provided with permissive interlocks to prevent opening if RCS pressure exceeds RHR design pressure.

3)

The capability for leak testing of the pressure isolation valves.

4)

Valve position indication that is available in the control room when isolation valve operators are deenergized.

5)

High-pressure alarms to warn control room operators when rising RCS pressure approaches the design pressure of attached low-pressure systems and both isolation valves are not closed.

These criteria are evaluated for each VEGP Units 1 and 2 low-pressure systems protected by PIVs.

x RHR System (Suction Side)

The RHR System is isolated from the RCS on the suction side by motor-operated valves. Leakage past the isolation valves would be detected by the lifting of RHR pump suction relief valves, accompanied by increasing pressurizer relief tank level, pressure, and temperature indications and alarms on the main control board. This satisfies Criteria 1 and 5.

Each inlet line to the RHR System is equipped with a relief valve to prevent RHR System overpressurization during plant startup, shutdown, and cold shutdown decay heat-removal operation. Each valve has a relief capacity of 900 gpm at a set pressure of 450 psig. An analysis of the capability of the RHR System relief valve to prevent overpressurization in the RHR System considered all credible events, including normal to NL-23-0217 Evaluation of Proposed Change

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operating conditions, infrequent transients, and abnormal occurrences. The analysis confirmed that one relief valve has the capability to maintain the RHR System maximum pressure within the ASME Code limits. This satisfies Criterion 1.

Indication of PIV leakage is also provided by an alarm on open hot leg valves and high RCS pressure. This satisfies Criterion 5.

Criterion 4 is not met for the RHR Loop Suction motor operated valves, but meeting this criterion is unnecessary because the valves are locked with a padlock in a tripped closed position to prevent unauthorized opening of the valve during the operating cycle.

TS SR 3.4.14.2 requires that operators, "Verify RHR System suction isolation valve interlock prevents the valves from being opened with a simulated or actual RCS pressure signal 450 psig at a frequency in accordance with the Surveillance Frequency Control Program." This satisfies Criterion 2.

These valves are currently and will continue to be tested in accordance with the IST Program, which satisfies Criterion 3.

x Accumulators The accumulators are isolated from the RCS by check valves. Leakage past these valves and into the accumulators is detected by redundant control room accumulator pressure and level indications and alarms. This satisfies Criterion 5.

Over pressure protection of the accumulators is provided by relief valves with a lift setpoint of 700 psig and a capacity of 1500 scfm. This satisfies Criterion 1.

Criterion 2 is not applicable to the accumulators.

Criterion 4 is not applicable to the accumulator PIV check valves; however, there are motor operated valves upstream of these check valves which are provided with power lockout and restoration from the motor control center. Valve position indication is provided in the control room. This satisfies criterion 4.

These valves are currently and will continue to be tested in accordance with the IST Program, which satisfies Criterion 3.

x RHR System (Discharge Side)

The RHR pump is isolated from the RCS by check valves and by a normally closed motor-operated valve. Any leakage past these valves will eventually pressurize the RHR discharge header and the pump suction header through the normally open RHR pump miniflow isolation valves. A continued increase in RHR pump discharge pressure will be indicated in the control room and ultimately result in lifting of relief valves in the suction header. This satisfies Criteria 1 and 5.

Each inlet line to the RHR System is equipped with a pressure relief valve designed to to NL-23-0217 Evaluation of Proposed Change

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prevent RHR System overpressurization assuming the most severe overpressure transients. These relief valves protect the system from inadvertent overpressurization during plant startup, shutdown, and cold shutdown decay heat-removal operations.

This satisfies Criterion 1.

Criterion 2 is not applicable to the RHR System discharge PIVs.

Each discharge line from the RHRS to the RCS is equipped with pressure relief valves, designed to relieve the maximum possible leakage through the valves isolating the RHR System from the RCS. These valves lift at 600 psig with relief capacity of 20 gpm.

This satisfies Criterion 1.

Criterion 4 is not applicable to the RHR to RCS Hot Leg Isolation check valves; however, there is a motor operated valve upstream of these check valves which are provided with power lockout, restoration and valve position indication from the control room. This satisfies criterion 4.

These valves are currently and will continue to be tested in accordance with the IST Program, which satisfies Criterion 3.

x Safety Injection Pump The Safety Injection pump discharge piping is isolated from the RCS by check valves and normally closed motor-operated valves. Leakage past these valves could pressurize the safety injection pump discharge header, resulting in control room indication of increasing pressure and, eventually, lifting of relief valves at 1750 psig and a capacity of 20 gpm. This satisfies Criteria 1 and 5.

Criterion 2 is not applicable to the safety injection pumps.

Criterion 4 is not applicable to the SI to RCS Hot Leg Isolation check valves; however, there is a motor operated valve upstream of these check valves which are provided with power lockout, restoration and valve position indication from the control room. This satisfies criterion 4.

These valves are currently and will continue to be tested in accordance with the IST Program, which satisfies Criterion 3.

In addition, the VEGP Units 1 and 2 design provides additional protection from an intersystem LOCA:

x The RHR discharge to the cold legs from the RHR heat exchangers is isolated by a motor-operated valve with power lockout, restoration and valve position indication from the control room.

x All PIVs are normally closed check valves without external operating mechanisms except for motor operated valves HV-8701A/B and HV-8702A/B. RHR System suction isolation valve interlock prevents the opening of HV-8701A/B and HV-8702A/B if RCS to NL-23-0217 Evaluation of Proposed Change

pressure 450 psig. The opening of both, in series, valves with RCS pressure 450 psig would require multiple failures.

x All PIVs are located inside containment.

A summary table comparing the VEGP Units 1 and 2 and VEGP Units 3 and 4 relevant requirements and design follows:

Comparison of VEGP Units 1 and 2 and VEGP Units 3 and 4 PIVs and Requirements VEGP Units 1 and 2 VEGP Units 3 and 4 PIV Surveillance Requirements SR 3.4.14.1 Verify leakage from each RCS PIV is equivalent to 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS pressure 2215 psig and 2255 psig.

SR 3.4.15.1 Verify leakage of each RCS PIV is equivalent to 0.5 gpm per nominal inch valve size up to a maximum of 5 gpm at an RCS pressure 2215 and 2255 psig.

PIV Valve Types and Sizes In series check valves ranging in sizes of 2" to 10".

12" shutdown cooling motor operated gate valves in series.

In series check valves ranging in sizes of 6" to 8".

10" shutdown cooling motor operated gate valves in series.

PIV Valve Pressure Ratings PIV design pressure of 2500 PSI at 650 oF with nominal pressure rating greater than 1500OEV, stainless steel, Class 1 valves.

PIV design pressure of 2485 psig at 680 oF with nominal pressure rating greater than 1500OEV, stainless steel, Class 1 valves.

Valve Seating Forces1 Check valves:

2 = 7,084 lbs.

6" = 63,759 lbs.

8 = 113,349 lbs.

10 = 177,107 lbs.

Shutdown cooling valves:

12" = 255,035 lbs.

Check valves:

6" = 63,759 lbs.

8" = 113,349 lbs.

Shutdown cooling valves:

10" = 177,107 lbs.

All of the VEGP Units 1 and 2 system piping and components protected by PIVs are at a minimum ASME Class 2, Seismic Class 1, Safety Class 2, Quality Group B which is comparable to the qualifications of the VEGP Units 3 and 4 piping and components which are ASME Class 2 or 3, Quality Group B or C, respectively.

Under Valve Seating Forces, the one difference between VEGP Units 1 & 2 and Units 3 & 4 is the shutdown cooling valves. The difference is insignificant because VEGP Units 1 & 2 shutdown cooling valves are locked with a padlock in a tripped closed position to prevent unauthorized opening of the valve during the operating cycle as previously stated in the SECY-90-016 section.

1 This comparison is a simple calculation using 2 x peak PIV test pressure, where 'r' = 1/2 the valve diameter.

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In summary, the intersystem LOCA protection of VEGP Units 1 and 2 satisfies the expectations in SECY-90-016,Section II, Preventative Feature Issues, E, "Intersystem LOCA," for ALWR plants, and comparable to the relevant design features of VEGP Units 3 and 4. As a result, the absence of the additional PIV testing frequencies of 18 months, 9 months, and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after actuation in the VEGP Units 3 and 4 TS is equally applicable to the VEGP Units 1 and 2 SR.

3.5 Conclusion The 18 month, 9 month, and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after actuation Frequencies of SR 3.4.14.1 are unnecessary and can be removed from the VEGP Units 1 and 2 TS because the remaining requirement to perform the SR in accordance with the IST Program will ensure the PIVs are capable of performing their specified safety function. The revised Frequency is consistent with the ASME OM Code, the regulations, and is supported by the excellent operating history of the PIVs. In addition, elimination of the additional testing frequencies is consistent with the NRC-approved VEGP Units 3 and 4 TS and the AP1000 DCD, and is equally applicable to the VEGP Units 1 and 2 design.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Guidance 10 CFR 50.36 The NRC's regulatory requirements related to the content of the TS are set forth in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, "Technical specifications." This regulation requires that the TS include items in the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings, (2) limiting conditions for operation, (3) SRs, (4) design features, and (5) administrative controls. The regulation does not specify the particular requirements to be included in a plant's TS.

Per 10 CFR 50.36(c)(3), SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

The proposed change will not eliminate any SRs. Therefore, the requirements of 10 CFR 50.36(c) continue to be met.

10 CFR 50.54(jj)

Structures, systems, and components subject to the codes and standards in 10 CFR 50.55a must be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed.

The proposed change does not alter the design, fabrication, erection, construction, or inspection of any components. The proposed change will continue to test the PIVs to the quality standards commensurate with the importance of the safety function being performed.

Therefore, the requirements of 10 CFR 50.54(jj) continue to be met.

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10 CFR 50.55a In accordance with 10 CFR 50.54, the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50.

These requirements include inservice testing of pumps and valves at nuclear power reactors pursuant to the ASME OM Code as specified in 10 CFR 50.55a(f). Paragraph (f) of 10 CFR 50.55a states in part that systems and components of pressurized water-cooled nuclear power reactors must meet the requirements of the ASME Boiler and Pressure Vessel (BPV)

Code and ASME OM Code as incorporated by reference into 10 CFR 50.55a. The proposed change will not affect the current TS requirement to test the PIVs in accordance with the ASME OM Code. Therefore, the requirements of 10 CFR 50.55a continue to be met.

10 CFR Part 50, Appendix A, General Design Criteria (GDC) 1, 14, 30, 32, 54, 55 Appendix A, GDC 1, Quality standards and records, states in part:

Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions.

Appendix A, GDC 14, Reactor coolant pressure boundary, states:

The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

Appendix A, GDC 30, Quality of reactor coolant pressure boundary," states Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical.

Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.

Appendix A, GDC 32, Inspection of reactor coolant pressure boundary, states:

Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.

Appendix A, GDC 54, Piping systems penetrating containment, states: to NL-23-0217 Evaluation of Proposed Change

E1-1

Piping systems penetrating primary reactor containment shall be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. Such piping systems shall be designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits.

Appendix A, GDC 55, Reactor coolant pressure boundary penetrating containment, " states:

Each line that is part of the reactor coolant pressure boundary and that penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:

(1)

One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2)

One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3)

One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or (4)

One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment. Isolation valves outside containment shall be located as close to containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety.

Other appropriate requirements to minimize the probability or consequences of an accidental rupture of these lines or of lines connected to them shall be provided as necessary to assure adequate safety. Determination of the appropriateness of these requirements, such as higher quality in design, fabrication, and testing, additional provisions for inservice inspection, protection against more severe natural phenomena, and additional isolation valves and containment, shall include consideration of the population density, use characteristics, and physical characteristics of the site environs.

The proposed change does not alter the design of any plant components. Therefore, compliance with the Appendix A GDCs, as described in the VEGP Updated Final Safety Analysis Report, is not affected by the proposed change. to NL-23-0217 Evaluation of Proposed Change

E1-1

Conclusion SNC has evaluated the proposed changes against the applicable regulatory requirements described above. Based on this evaluation, there is reasonable assurance that the health and safety of the public will remain unaffected following the approval of the proposed change.

4.2 Significant Hazards Consideration Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company, Inc. (SNC) hereby requests a revision to the Technical Specifications (TS) for Vogtle Electric Generating Plant, Units 1 and 2 (VEGP). The proposed amendment would revise the Surveillance Requirement (SR) Frequency for Reactor Coolant System (RCS) pressure isolation valve (PIV) operational leakage testing to only require testing at the frequency specified in the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code).

SNC has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

(1)

Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment would revise the SR Frequency for RCS PIV operational leakage testing to only require testing at the frequencies specified in the Inservice Testing Program, which complies with the ASME OM Code. RCS PIV testing is performed during a plant shutdown and is not an initiator to any accident previously evaluated. The RCS PIV operational testing acceptance criteria are not affected by the proposed change. The RCS PIVs will continue to be tested to ensure leakage is within the TS allowable leakage limits. As a result, the consequences of any accident previously evaluated are unchanged.

Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

(2)

Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed amendment would revise the SR Frequency for RCS PIV operational leakage testing to only require testing at the frequencies specified in the Inservice Testing Program, which complies with the ASME OM Code. RCS PIV operational testing is only performed during a plant shutdown. The testing methodology and acceptance criteria remain unchanged. The proposed change does not involve a physical change to the plant or the manner in which the plant is operated or controlled. to NL-23-0217 Evaluation of Proposed Change

The proposed change does not alter the design function or operation of the RCS PIVs.

The proposed change does not alter the ability of the RCS PIVs to perform their design function. Since pressure boundary leakage is an evaluated accident, the proposed change does not create any new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.

Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3)

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed amendment would revise the SR Frequency for RCS PIV operational leakage testing to only require testing at the frequencies specified in the Inservice Testing Program, which complies with the ASME OM Code. The proposed change does not affect the initial assumptions, margins, or controlling values used in any accident analysis. The amount of allowed leakage is not increased. The proposed change does not affect any design basis or safety limit or any Limiting Condition for Operation.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based upon the above evaluation, SNC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

4.3 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

S SNC has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined by 10 CFR 20, or it would change an inspection or surveillance requirement. However, the proposed changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

E1-1 to NL-23-0217 Evaluation of Proposed Change

E1-1

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs be prepared in connection with the proposed amendment.

6.0 REFERENCES

None to NL-23-0217 Evaluation of Proposed Change

Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification Surveillance Requirement 3.4.14.1 and Proposed Inservice Testing Alternative ALT-VR-02 Proposed Technical Specification Changes (Mark-Up)

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RCS PIV Leakage 3.4.14 Vogtle Units 1 and 2 3.4.14-4 Amendment No. 158 (Unit 1)

Amendment No. 140 (Unit 2)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 (continued)

For systems rated at less than 50% RCS design pressure, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation (except for valves HV-8701A/B and HV-8702A/B).

SR 3.4.14.2 Verify RHR System suction isolation valve interlock prevents the valves from being opened with a simulated or actual RCS pressure signal 450 psig.

In accordance with the Surveillance Frequency Control Program

Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification Surveillance Requirement 3.4.14.1 and Proposed Inservice Testing Alternative ALT-VR-02 Revised Technical Specification Changes

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RCS PIV Leakage B 3.4.14 Vogtle Units 1 and 2 B 3.4.14-7 REVISION 39 BASES SURVEILLANCE SR 3.4.14.2 (continued)

REQUIREMENTS valves from being opened is set so the actual RCS pressure must be

< 450 psig to open the valves. This setpoint ensures the RHR design pressure will not be exceeded. To ensure that the RHR relief valves will not lift, the actual interlock setpoint used in performing the surveillance is < 365 psig, and takes into consideration various allowances for relief valve setting variation, transmitter elevation, and the total instrument channel uncertainty. The total instrument channel uncertainty is calculated in accordance with reference 10, and the allowance for process instrumentation (rack drift) is 1%. Once the interlock setpoint is initially reached, administrative controls ensure that the RHR suction isolation valves are closed prior to reaching an RCS pressure that could cause the RHR suction relief valves to open.

Due to the bistable reset design, the valves could be opened at a pressure above the interlock setpoint, but below the reset pressure.

The administrative controls ensure that the valves will not be opened if RCS pressure exceeds 365 psig after RCS pressure has decreased below the interlock setpoint.

REFERENCES

10 CFR 50.2.

10 CFR 50.55a(c).

10 CFR 50, Appendix A, Section V, GDC 55.

WASH-1400 (NUREG-75/014), Appendix V, October 1975.

NUREG-0677, May 1980.

FSAR Section 5.4.

ASME Code for Operation and Maintenance of Nuclear Power

Plants (OM Code).'HOHWHG

WCAP-16294-NP-A, Rev. 1, Risk-Informed Evaluation of

Changes to Technical Specification Required Action Endstates

for Westinghouse NSSS PWRs, June 2010.

10 CFR 50.55a(f).'HOHWHG

WCAP-11269, Rev. 1, Westinghouse Setpoint Methodology for

Protection Systems.

Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification Surveillance Requirement 3.4.14.1 and Proposed Inservice Testing Alternative ALT-VR-02 Pressure Isolation Valves Subject to Testing Under SR 3.4.14.1

, Page 1 Pressure Isolation Valves Subject to Testing Under SR 3.4.14.1 Notes:

1)

All listed valves are included in the scope of the Vogtle Unit 1 and 2, 4th Interval lnservice Testing Plan.

2)

The listed PIVs are all of the design pressure of 2500 PSI at 650 with nominal pressure rating greater than 1500#, stainless steel, Class 1 valves.

Valve Number Description System1 ASME Code Class OM Code Category2 1/2-HV-8701A/B RHR Pump Supply Isolation From RCS Loops RCS 1

A 1/2-HV-8702A/B RHR Pump Supply Isolation From RCS Loops RCS 1

A 1/2-1204-U4-143 1/2-1204-U4-144 1/2-1204-U4-145 1/2-1204-U4-146 SI To RCS Cold Leg Isolation Check Valve SI 1

A/C 1/2-1204-U4-120 1/2-1204-U4-121 1/2-1204-U4-123 1/2-1204-U4-122 SI to RCS Hot Leg Isolation Check Valve SI 1

A/C 1/2-1204-U6-079 1/2-1204-U6-080 1/2-1204-U6-081 1/2-1204-U6-082 SI Accumulator Discharge Check Valve SI 1

A/C 1/2-1204-U6-083 1/2-1204-U6-084 1/2-1204-U6-085 1/2-1204-U6-086 SI/RHR Cold Leg Admission Check Valve SI 1

A/C 1/2-1204-U6-147 1/2-1204-U6-148 1/2-1204-U6-149 1/2-1204-U6-150 RHR To RCS Cold Leg Isolation Check Valve SI 1

A/C 1/2-1204-U6-128 1/2-1204-U6-129 RHR To RCS Hot Leg Isolation Check Valve SI 1

A/C 1/2-1204-U6-126 1/2-1204-U6-124 1/2-1204-U6-127 1/2-1204-U6-125 RCS Hot Leg SI Admission Check Valve SI 1

A/C 1 SI - Safety Injection RCS - Reactor Coolant System RHR - Residual Heat Removal 2 A - Valves for which seat leakage is limited to a specific maximum amount in the closed position A/C - Check valves for which seat leakage is limited to a specific maximum amount in the closed position.

Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification Surveillance Requirement 3.4.14.1 and Proposed Inservice Testing Alternative ALT-VR-02 Pressure Isolation Valve Leakage Test History

, Page 1 PIV Leakage Test History Component Number Date of Test Measured Value (gpm)

Maximum Allowable Leak Rate (gpm)

VEGP Unit 1 RHR Pump Supply Isolation From RCS Loops 1-HV-8701B Fall 2021 0.3 5.0 Spring 2020 0.5 Fall 2018 0.8 Spring 2017 0.0 Fall 2015 0.9 Spring 2014 2.1 Fall 2012 0.4 Spring 2011 0.4 1-HV-8701A Fall 2021 0.0 5.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0 Spring 2011 0.0 1-HV-8702B Fall 2021 0.0 5.0 Spring 2020 0.0 Fall 2018 0.2 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0 Spring 2011 0.0 1-HV-8702A Fall 2021 0.0 5.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0 Spring 2011 0.0 SI To RCS Cold Leg Isolation Check Valve 1-1204-U4-143 1-1204-U4-144 1-1204-U4-145 1-1204-U4-146 Fall 2021 0.0 1.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.3 Spring 2011 0.1 SI to RCS Hot Leg Isolation Check Valve 1-1204-U4-120 Fall 2021 0.0 1.0 Spring 2020 0.0

, Page 2 PIV Leakage Test History Component Number Date of Test Measured Value (gpm)

Maximum Allowable Leak Rate (gpm)

Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0 Spring 2011 0.0 1-1204-U4-121 Fall 2021 0.0 1.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0 Spring 2011 0.0 1-1204-U4-123 Fall 2021 0.0 1.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0 Spring 2011 0.0 1-1204-U4-122 Fall 2021 0.0 1.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0 Spring 2011 0.0 SI Accumulator Discharge Check Valve 1-1204-U6-079 Fall 2021 0.1 5.0 Spring 2020 0.1 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.1 Fall 2012 0.0 Spring 2011 0.1 1-1204-U6-080 Fall 2021 0.0 5.0 Spring 2020 0.1 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0

, Page 3 PIV Leakage Test History Component Number Date of Test Measured Value (gpm)

Maximum Allowable Leak Rate (gpm)

Spring 2011 0.0 1-1204-U6-081 Fall 2021 0.0 5.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0 Spring 2011 0.0 1-1204-U6-082 Fall 2021 0.0 5.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0 Spring 2011 0.0 SI/RHR Cold Leg Admission Check Valve 1-1204-U6-083 Fall 2021 0.0 5.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0 Spring 2011 0.0 1-1204-U6-084 Fall 2021 0.0 5.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0 Spring 2011 0.0 1-1204-U6-085 Fall 2021 0.0 5.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.1 Fall 2012 0.0 Spring 2011 0.0 1-1204-U6-086 Fall 2021 0.0 5.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0

, Page 4 PIV Leakage Test History Component Number Date of Test Measured Value (gpm)

Maximum Allowable Leak Rate (gpm)

Fall 2015 0.0 Spring 2014 0.2 Fall 2012 0.0 Spring 2011 0.0 RHR To RCS Cold Leg Isolation Check Valve 1-1204-U6-147 1-1204-U6-148 Fall 2021 0.0 3.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0 Spring 2011 0.0 1-1204-U6-149 1-1204-U6-150 Fall 2021 0.0 3.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0 Spring 2011 0.0 RHR To RCS Hot Leg Isolation Check Valve 1-1204-U6-128 Fall 2021 0.0 4.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0 Spring 2011 0.0 1-1204-U6-129 Fall 2021 0.0 4.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0 Spring 2011 0.0 RCS Hot Leg SI Admission Check Valve 1-1204-U6-126 Fall 2021 0.1 3.0 Spring 2020 0.0 Fall 2018 0.1 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.1

, Page 5 PIV Leakage Test History Component Number Date of Test Measured Value (gpm)

Maximum Allowable Leak Rate (gpm)

Spring 2011 0.0 1-1204-U6-124 Fall 2021 0.2 3.0 Spring 2020 0.3 Fall 2018 0.4 Spring 2017 0.3 Fall 2015 0.3 Spring 2014 0.2 Fall 2012 0.4 Spring 2011 0.3 1-1204-U6-127 Fall 2021 0.1 3.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.1 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.1 Spring 2011 0.1 1-1204-U6-125 Fall 2021 0.1 3.0 Spring 2020 0.0 Fall 2018 0.0 Spring 2017 0.0 Fall 2015 0.0 Spring 2014 0.0 Fall 2012 0.0 Spring 2011 0.0 VEGP Unit 2 RHR Pump Supply Isolation From RCS Loops 2-HV8701B Spring 2022 0.0 5.0 Fall 2020 0.8 Spring 2019 0.0 Fall 2017 0.2 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0 Fall 2011 0.2 2-HV8701A Spring 2022 0.5 5.0 Fall 2020 0.5 Spring 2019 0.0 Fall 2017 0.0 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0 Fall 2011 0.0 2-HV8702B Spring 2022 0.0 5.0 Fall 2020 0.0 Spring 2019 0.0

, Page 6 PIV Leakage Test History Component Number Date of Test Measured Value (gpm)

Maximum Allowable Leak Rate (gpm)

Fall 2017 0.1 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0 Fall 2011 0.0 2-HV8702A Spring 2022 0.0 5.0 Fall 2020 0.0 Spring 2019 0.0 Fall 2017 0.0 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0 Fall 2011 0.0 SI To RCS Cold Leg Isolation Check Valve 2-1204-U4-143 2-1204-U4-144 2-1204-U4-145 2-1204-U4-146 Spring 2022 0.0 1.0 Fall 2020 0.0 Spring 2019 0.0 Fall 2017 0.0 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0 Fall 2011 0.0 SI to RCS Hot Leg Isolation Check Valve 2-1204-U4-120 Spring 2022 0.5 1.0 Fall 2020 0.0 Spring 2019 0.0 Fall 2017 0.0 Spring 2016 0.0 Fall 2014 0.1 Spring 2013 0.0 Fall 2011 0.1 2-1204-U4-121 Spring 2022 0.1 1.0 Fall 2020 0.0 Spring 2019 0.0 Fall 2017 0.4 Spring 2016 0.0 Fall 2014 0.1 Spring 2013 0.0 Fall 2011 0.1 2-1204-U4-123 Spring 2022 0.0 1.0 Fall 2020 0.0 Spring 2019 0.0 Fall 2017 0.0 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0

, Page 7 PIV Leakage Test History Component Number Date of Test Measured Value (gpm)

Maximum Allowable Leak Rate (gpm)

Fall 2011 0.2 2-1204-U4-122 Spring 2022 0.0 1.0 Fall 2020 0.0 Spring 2019 0.0 Fall 2017 0.0 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0 Fall 2011 0.1 SI Accumulator Discharge Check Valve 2-1204-U6-079 Spring 2022 0.0 5.0 Fall 2020 0.0 Spring 2019 0.0 Fall 2017 0.0 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0 Fall 2011 0.0 2-1204-U6-080 Spring 2022 0.0 5.0 Fall 2020 0.1 Spring 2019 0.1 Fall 2017 0.1 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0 Fall 2011 0.1 2-1204-U6-081 Spring 2022 0.3 5.0 Fall 2020 0.0 Spring 2019 0.3 Fall 2017 0.0 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0 Fall 2011 0.1 2-1204-U6-082 Spring 2022 0.0 5.0 Fall 2020 0.0 Spring 2019 0.0 Fall 2017 0.0 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0 Fall 2011 0.0 SI/RHR Cold Leg Admission Check Valve 2-1204-U6-083 Spring 2022 0.0 5.0 Fall 2020 0.0 Spring 2019 0.0

, Page 8 PIV Leakage Test History Component Number Date of Test Measured Value (gpm)

Maximum Allowable Leak Rate (gpm)

Fall 2017 0.4 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.1 Fall 2011 0.1 2-1204-U6-084 Spring 2022 0.0 5.0 Fall 2020 0.0 Spring 2019 0.0 Fall 2017 0.1 Spring 2016 0.0 Fall 2014 0.2 Spring 2013 0.1 Fall 2011 0.1 2-1204-U6-085 Spring 2022 0.1 5.0 Fall 2020 0.0 Spring 2019 0.1 Fall 2017 0.0 Spring 2016 0.0 Fall 2014 0.1 Spring 2013 0.1 Fall 2011 0.1 2-1204-U6-086 Spring 2022 0.0 5.0 Fall 2020 0.0 Spring 2019 0.0 Fall 2017 0.0 Spring 2016 0.1 Fall 2014 0.1 Spring 2013 0.2 Fall 2011 0.2 RHR To RCS Cold Leg Isolation Check Valve 2-1204-U6-147 2-1204-U6-148 Spring 2022 0.0 3.0 Fall 2020 0.0 Spring 2019 0.0 Fall 2017 0.4 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0 Fall 2011 0.1 2-1204-U6-149 2-1204-U6-150 Spring 2022 0.0 3.0 Fall 2020 0.0 Spring 2019 0.0 Fall 2017 0.4 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0 Fall 2011 0.0

, Page 9 PIV Leakage Test History Component Number Date of Test Measured Value (gpm)

Maximum Allowable Leak Rate (gpm)

RHR To RCS Hot Leg Isolation Check Valve 2-1204-U6-128 Spring 2022 0.0 4.0 Fall 2020 0.0 Spring 2019 0.0 Fall 2017 0.0 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0 Fall 2011 0.0 2-1204-U6-129 Spring 2022 0.0 4.0 Fall 2020 0.0 Spring 2019 0.0 Fall 2017 0.0 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0 Fall 2011 0.0 RCS Hot Leg SI Admission Check Valve 2-1204-U6-126 Spring 2022 0.0 3.0 Fall 2020 0.0 Spring 2019 0.0 Fall 2017 0.0 Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0 Fall 2011 0.1 2-1204-U6-124 Spring 2022 0.1 3.0 Fall 2020 0.1 Spring 2019 0.1 Fall 2017 0.3 Spring 2016 0.1 Fall 2014 0.0 Spring 2013 0.1 Fall 2011 0.1 2-1204-U6-127 Spring 2022 0.0 3.0 Fall 2020 0.0 Spring 2019 0.0 Fall 2017 0.7 Spring 2016 0.6 Fall 2014 0.4 Spring 2013 0.0 Fall 2011 0.1 2-1204-U6-125 Spring 2022 0.0 3.0 Fall 2020 0.0 Spring 2019 0.3 Fall 2017 0.0

, Page 10 PIV Leakage Test History Component Number Date of Test Measured Value (gpm)

Maximum Allowable Leak Rate (gpm)

Spring 2016 0.0 Fall 2014 0.0 Spring 2013 0.0 Fall 2011 0.0

Enclosure

Proposed lnservice Testing Alternative ALT-VR-02 Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification Surveillance Requirement 3.4.14.1 and Proposed Inservice Testing Alternative ALT-VR-02

Vogtle Electric Generating Plant - Units 1 and 2 Proposed lnservice Testing Alternative ALT-VR-02

1. American Society of Mechanical Engineers (ASME) Code Component(s) Affected Table 1, PIVs Subject to Leakage Rate Testing in Accordance with TS SR 3.4.14.1 Valve Number Units 1 and 2 Description System1 ASME Code Class OM Code Category2 1/2-HV-8701A/B RHR Pump Supply Isolation From RCS Loops RCS 1