ML19204A349

R. E. Ginna Nuclear Power Plant: License Amendment Request - Increase the Main Steam Safety Valve Lift Setpoint Tolerance

ML19204A349
Person / Time
Site:	Ginna
Issue date:	07/23/2019
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML19204A349 (14)

Text

Exelon Generation July 23, 2019 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001 R. E. Ginna Nuclear Power Plant 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.90 Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244

Subject:

License Amendment Request-Increase the Main Steam Safety Valve Lift Setpoint Tolerance In accordance with 1 O CFR 50.90, Exelon Generation Company, LLC (EGC) requests proposed changes that would modify Technical Specification (TS) Section 3.7.1 ("Main Steam Safety Valves (MSSVs)"). The proposed change revises TS Surveillance Requirement (SR) 3.7.1.1 to increase the allowable as-found MSSV lift setpoint tolerance from +1%/-3% to +1.4%/-4%.

The change is proposed in order to reduce an unnecessarily restrictive surveillance requirement. The proposed change will not impact the reliability of the MSSVs or adversely impact their ability to perform their safety function. The change will reduce the number of TS MSSV surveillance test failures for early lift pressure and preclude the submittal of previously reportable Licensee Events Reports (LERs) to the NRC due to setpoint drift.

The proposed changes have been reviewed by the Ginna Plant Operations Review Committee in accordance with the requirements of the EGG Quality Assurance Program.

EGC requests approval of the proposed amendments by February 24, 2020, to support implementation during Ginna refueling outage G1 R42. Once approved, these amendments shall be implemented within 30 days of issuance. This implementation period will provide adequate time for documents to be revised using the appropriate change control mechanisms. contains the evaluation of the proposed changes. Attachment 2 provides the marked up TS and Bases pages. The Bases pages are being provided for information only.

In accordance with 1 O CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), EGC is notifying the State of New York of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

There are no commitments contained in this submittal.

Should you have any questions concerning this letter, please contact Jessie Hodge at

U.S. Nuclear Regulatory Commission License Amendment Request -

Increase the MSSV Lift Setpoint Tolerance July 23, 2019 Page2 (610) 765-5532.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 23rc1 day of July 2019.

Respectfully, tf '-'-'-'

• J T _j__, J yi-- ~

James Barstow Director, Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: 1) Evaluation of Proposed Changes

2) Markup of Technical Specifications and Bases Pages cc: USNRC Region I, Regional Administrator USNRC Senior Resident Inspector, Ginna USNRC Senior Project Manager, Ginna A. L. Peterson, NYSERDA

ATTACHMENT 1 Evaluation of Proposed Changes

SUBJECT:

Increase the Main Steam Safety Valve Lift Setpoint Tolerance 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

Page 1

ATTACHMENT 1 Evaluation of Proposed Changes 1.0

SUMMARY

DESCRIPTION In accordance with 1 O CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Renewed Facility Operating License No. DPR-18 for the A. E. Ginna Nuclear Power Plant (GNPP). The proposed change revises Technical Specification (TS) Surveillance Requirement (SR) 3.7.1.1 to increase the allowable as-found Main Steam Safety Valve (MSSV) lift setpoint tolerance from + 1 %/-3% to + 1.4%/-4%.

2.0 DETAILED DESCRIPTION The proposed change revises TS SR 3.7.1.1 to increase the as found lift pressure tolerance from + 1 %/-3% to+ 1.4%/-4% for the affected MSSVs. These MSSVs numbers are 3509, 3511, and 3515 of Steam Generator (SG) A, and 3508, 3510, and 3512 of SG B. The affected MSSVs have a set pressure of 1140 psig, whereas the unaffected MSSVs (3513 on SG A and 3514 on SG B) have a set pressure 1085 psig. provides the existing TS page marked-up to show the proposed change. Marked-up pages showing corresponding changes to the TS Bases are provided in Attachment 2 for information only. The TS Bases changes will be processed in accordance with the GNPP TS Bases Control Program (TS 5.5.13).

3.0 TECHNICAL EVALUATION

Ginna has four MSSVs for each steam line. The first valve lifts at 1085 psig, and the remaining three valves are set to lift at 1140 psig. The MSSVs basic design function is to limit the secondary system pressure to less than 110% of design pressure, in accordance with the ASME code. Reactor Coolant System (RCS) heat removal I overpressure protection is an additional MSSV design function by providing a heat sink for removal of RCS energy, if the preferred (but non-safety related) condenser heat sink is not available. The MSSV design includes staggered setpoints so that only the needed valves will actuate. Staggered setpoints reduce the potential for valve chattering that is due to steam pressure insufficient to fully open all valves following a turbine/reactor trip.

To minimize unnecessary maintenance and testing based on conservative acceptance criteria, this change justifies increasing the as-found acceptance range for the associated MSSVs.

The increase is achieved by using available margin in Ginna's design analyses. This change is limited to the six MSSVs that are set at 1140 psig. The tolerance for the two MSSVs that are set at 1085 psig is NOT changed here. Tech Spec Surveillance Requirement SR 3.7.1.1 currently requires (+1%, -3%) tolerance for all as-found lift settings. This change does NOT alter the Tech Spec SR 3.7.1.1 as-left setting requirement of(+ 1 %, -1 %). The following as-tound surveillance requirements will apply after approval of the License Amendment Request (LAA) associated with this design change:

Page 2

VALVE NUMBER SGA SGB 3509 3508 3511 3510 3515 3512 3513 3514 ATTACHMENT 1 Evaluation of Proposed Changes LIFT SETIING PSIG 1140 (+ 1.4%, - 4%)

1140 (+ 1.4%, - 4%)

1140 (+ 1.4%, - 4%)

1085 (+ 1%, - 3%)

Explanation of the new tolerance follows. The Loss of Load (LOL) analysis applied a tolerance of+ 1.4% and bounds the high-side tolerance. Expanding the tolerance beyond 1.4% would require significant investment in re-analysis efforts. Successful test results are expected with a high-side tolerance of 1.4%, given historical performance of the MSSVs. Ginna targets as-left settings that are biased low (but within the normal +/-1%), because the MSSVs have significantly more as-found margin on the low side. The as-found tolerance is expanded here to

(-4%) to avoid testing failures from normal drift, given that Ginna attempts to leave the valves in the low end of the (+1%, -1%) as-left range. The -4% is acceptable, because no analyses limit the low-side tolerance, and, if the setting drifts low within the tolerance, then overall safety margin would be improved during an event.

The MSSVs continue to have staggered setpoints such that there is a negligible increase in the potential for valve chattering. Ginna has more separation between first and second MSSVs to open compared to other U.S. pressurized water reactors. Separation to avoid chattering is based on nominal setpoints. The increase from -3% to -4% as-found tolerance has a negligible influence on chattering potential, and Ginna's margin to avoid chattering continues to be greater than other plants by inspection of other plant MSSV Tech Specs.

The MSSVs are required to reseat when pressure is reduced. The setpoint tolerance changes do not degrade the valves ability to reseat.

Table 1, below, shows the MSSV setpoint tolerances that are applied in each UFSAR chapter 15 event. Table 1 shows that all Chapter 15 accident analyses that are sensitive to MSSV setpoints apply the high-side tolerance to achieve conservative results. All Chapter 15 events that model MSSVs apply a tolerance of 1.4% or higher with a single exception: the Small Break Loss Of Coolant Accident (SBLOCA) analysis applies 1 % tolerance for the MSSV setpoints.

However, as described following Table 1, the effect of this change on the SBLOCA analysis results is negligible, and a change notice against the SBLOCA analyses is to be posted allowing the 1.4% tolerance for the 1140 psig setpoint MSSVs (3508, 3509, 3510, 3511, 3512, 3515). All Chapter 15 analyses continue to be bounding after increasing the as-found tolerance from 1 %

to 1.4%.

Table 1 identifies that Chapter 15 accident analyses methods for events sensitive to secondary steam release through the MSSVs achieve conservative results by opening the MSSVs at pressures above nominal. The two MSSV functions potentially affected by this change are secondary system overpressure protection and RCS heat removal. By increasing the low-side Page 3

ATTACHMENT 1 Evaluation of Proposed Changes tolerance to -4%, it is possible that the MSSVs could lift earlier than with the existing -3%

margin. This results in a potential for removing heat earlier in events than previously.

Removing heat earlier results in an improvement in protection against secondary system overpressure and an improvement in RCS temperatures. Therefore, in terms of heat removal, there is a net overall margin improvement associated with the tolerance change.

Steam release for dose for non-Steam Generator Tube Rupture (SGTR) events is not affected by MSSVs potentially lifting at a lower setting. This is because dose for steam releases for non-SGTR events are predicted with a method that uses bounding heat inputs and SIG mass releases as described in References 1 and 2; this method is insensitive to MSSV settings.

Steam release for dose for SGTR events is not affected by MSSVs potentially lifting at a lower setting, because the Atmospheric Relief Valves (ARVs) are credited for SGTR events. Only the first set of MSSVs would open for SGTR events, if ARVs were not available, such that the MSSV's being changed could not affect SGTR analyses.

Page4

UFSAR 15.1 15.1.1 15.1.2 15.1.3 15.1.4 15.1.5 15.1.6 15.2 15.2.1 15.2.2 15.2.3 15.2.4 15.2.S 15.2.6 15.2.7 15.3 15.3.1 15.3.2 15.4 15.4.1 15.4.2 15.4.3 15.4.4 15.4.5 15.4.6 15.S 15.6 15.6.1 15.6.2 15.6.3 15.6.4 15.6.4.1 15.6.4.2 15.7 15.8 15.8 Event ATTACHMENT 1 Evaluation of Proposed Changes INCREASE IN HEAT REMOVAL BY SECONDARY SYSTEM Decrease in Feedwater Temperature Increase in Feedwater Flow Excessive Load Increase (10%)

Inadvertent Opening of a Steam Generator Relief/Safety Valve Steam System Piping Failure - Zero Power (Core Response only)

Combined Steam Generator ARV and Feedwater Control Valve Failures DECREASE IN HEAT REMOVAL BY SECONDARY SYSTEM Steam Pressure Regulator Malfunction or Failure that Results in Decreasing Steam Flow Loss-of-External-Electrical Load Turbine Trip Loss-of-Condenser Vacuum Loss-of-Offsite-AC-Power to the Station Auxiliaries Loss-of-Normal Feedwater (LONF)

Feedwater System Pipe Breaks DECREASE IN RCS FLOW RATE Flow Coast Down Accidents (RCP failure or RCP power failure)

Locked Rotor Accident REACTIVITY AND POWER DISTRIBUTION ANOMALIES UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY WITHDRAWAL FROM A SUBCRITICAL CONDITION UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY WITHDRAWAL AT POWER STARTUP OF AN INACTIVE REACTOR COOLANT LOOP CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION RUPTURE OF A CONTROL ROD DRIVE MECHANISM HOUSING-ROD CLUSTER CONTROL ASSEMBLY EJECTION ROD CLUSTER CONTROL ASSEMBLY DROP INCREASE IN RCS INVENTORY DECREASE IN REACTOR COOLANT INVENTORY INADVERTENT OPENING OF A PRZR SAFETY OR PORV SMALL LINES CARRYING PRIMARY COOLANT OUTSIDE CNMNT STEAM GENERATOR TUBE RUPTURE PRIMARY SYSTEM PIPE RUPTURES SBLOCA LBLOCA RADIOACTIVE RELEASE FROM SUBSYSTEM OR COMPONENT ANTICIPATED TRANSIENT WITHOUT SCRAM ATWS NOTE 1: See discussion of 1140 psig valve setpoint tolerance change for SBLOCA following this table.

TABLE 1 Page 5 MSSV Tolerance NOT modeled

+1.5 %

NOT modeled Bounded by stm piping failure (UFSAR 15.1.5).

NOT modeled

+1.5 %

Bounded by LOL (UFSAR 15.2.2).

+1.4 %

Bounded by LOL (UFSAR 15.2.2).

Bounded by LOL (UFSAR 15.2.2).

+1.5%

+1.5 %

+1.5 %

NOT modeled

+1.5 %

NOT modeled

+1.5 %

NOT required NOT modeled NOT modeled MSSVs do NOT actuate.

NOT modeled

+1.5%

NOT modeled MSSVs NOT actuated

+ 1.0% (NOTE 1)

MSSVs NOT modeled MSSVs NOT modeled

+1.5%

ATTACHMENT 1 Evaluation of Proposed Changes Design analyses that were reviewed are shown under the References. Chapter 15 analyses are shown in Table 1. The SBLOCA analysis requires a change notice be posted to clarify that 1.4% tolerance may be used in future revisions of the SBLOCA model for the 1140 psig setpoint valves. Reference 3 documented the Westinghouse review of the SBLOCA analyses for this change including plots of steam pressure for 0.8-inch, 1.5-inch, 2-inch, and 3-inch equivalent diameter breaks. Ginna's limiting SBLOCA break size is 2-inch. The 1140 psig setpoint valves are predicted to NOT lift for break sizes of 2-inches and under as discussed in Reference 3.

The 1140 psig setpoint valves are predicted to lift for approximately 5 seconds for the 3-inch break case well before any boil-off core uncovery. For non-limiting SBLOCA events (3-inch and larger), slight event-timing changes (on the order of seconds) may occur, but the changes would NOT result in these breaks becoming more limiting than the 2-inch break per Reference 3. The change will have o° F peak cladding temperature (PCT) impact per Reference 3, and an action item is tracking 1 OCFR50.46 reporting for this evaluation.

The potential effect of the change on cyclic thermal and pressure loads associated with design transients was reviewed in Reference 4. Design transient cyclic loading is discussed in UFSAR sections 3.9.1and5.1.5 and evaluated in References 5, 6, and 7. Reference 4 concluded that there might be a 1.1° F increase in steam temperature due to the change and that this has a negligible effect on the design transients analyses.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The ASME OM Code Appendix I requires as found testing acceptance criteria be either Owner defined or +/- 3% of nominal setting. Ginna is establishing new Owner defined acceptance criteria which is permitted by the Code.

The proposed change has been evaluated considering the following and determined that applicable regulations and requirements continue to be met:

GOG 2, as it relates to safety-related portions of the system being capable of withstanding the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, and floods.

GOG 4, with respect to safety-related portions of the system being capable of withstanding the effects of external missiles and internally generated missiles, pipe whip, and jet impingement forces associated with pipe breaks.

GOG 5, as it relates to the capability of shared systems and components important to safety to perform required safety functions.

1 O CFR 50.63, as it relates to the ability of a plant to withstand for a specified duration and then recover from an SBO.

GOG 27 and 28, as they relate to the RCS being designed with an appropriate margin to ensure that acceptable fuel design limits are not exceeded and that the capability to cool the core is maintained.

Page 6

ATTACHMENT 1 Evaluation of Proposed Changes GDC 34, as it relates to the system function of transferring residual and sensible heat from the reactor system in indirect-cycle plants.

GDC 57, as it relates to the requirement that lines penetrating the primary containment boundary and neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere have at least one locked-closed, remote-manual, or automatic isolation valve2 outside containment.

4.2 Precedent The USNRC has approved similar license amendments related to increasing the main steam SRV/SV lift setpoints. Recent examples for BWRs include:

1) Pilgrim Nuclear Power Station (License Amendment No. 235 issued by USN RC letter dated March 28, 2011 - ADAMS Accession No. ML110650009). As part of this change, this plant revised the setpoint in addition to the tolerances also resulting in more extensive analysis required.

2) Quad Cities Nuclear Power Station, Units 1 and 2 (License Amendment No. 235 and 230 issued by USNRC letter dated November 1, 2007 -ADAMS Accession No. ML073060079). It is noted that this plant contains fuel from more than one fuel vendor which required more extensive transient evaluation. Ginna has only one fuel vendor that performs the reload analysis. This change also revised the SLC System boron enrichment which required more extensive A TWS analysis.

3) Limerick Generating Station, Units 1 (License Amendment No. 137 issued by USNRC letter dated November 10, 1999 - ADAMS Accession No. ML993250099} and 2 (License Amendment No. 98 issued by USNRC letter dated May 17, 1999-ADAMS Accession No. ML011560845).

4) LaSalle County Station, Units 1 and 2 (License Amendment No. 232 and 218 issued by USN RC letter dated December 19, 2018 - ADAMS Accession No. ML18278A030}.

4.3 No Significant Hazards Consideration Exelon Generation Company, LLC (Exelon) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 1 O CFR 50.92, "Issuance of amendment," as discussed below.

1.

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

Response: No The proposed change revises Technical Specification (TS) Surveillance Requirement (SR) 3.7.1.1 to increase the allowable as-found Main Steam Safety Valve (MSSV) lift setpoint tolerance from + 1 %/-3% to + 1.4%/-4%. As summarized in Section 3.0, increasing the applicable MSSV tolerance has been evaluated for the Small Break Loss Of Coolant Accident (SBLOCA) analysis of record but this change does not affect the limiting SBLOCA Page 7

ATTACHMENT 1 Evaluation of Proposed Changes results. However, this change does not alter the manner in which the valves are operated.

Consistent with current TS requirements, the proposed change continues to require that the MSSVs be adjusted to within +/-1 % of their nominal lift setpoints following testing. Since the proposed change does not alter the manner in which the valves are operated, there is no significant impact on reactor operation.

The proposed change does not involve a physical change to the valves, nor does it change the safety function of the valves. The proposed TS revision involves no significant changes to the operation of any systems or components in normal or accident operating conditions and no changes to existing structures, systems, or components. The proposed amendments do not change any other behavior or operation of any MSSV, and therefore, has no significant impact on reactor operation. They also have no significant impact on response to any perturbation of reactor operation including transients and accidents previously analyzed in the Updated Final Safety Analysis Report (UFSAR).

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

2.

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

Response: No The proposed change revises TS SR 3.7.1.1 to increase the allowable as-found MSSV lift setpoint tolerance from + 1 %/-3% to + 1.4%/-4%. The proposed change to the setpoint tolerance does not adversely affect the operation of any safety-related components or equipment. The proposed amendments do not involve physical changes to the applicable MSSVs, nor do they change the safety function of the MSSVs. The proposed amendments do not require any physical change or alteration of any existing plant equipment. No new or different equipment is being installed, and installed equipment is not being operated in a new or different manner. There is no alteration to the parameters within which the plant is normally operated. This change does not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. No alterations in the procedures that ensure the plant remains within analyzed limits are being proposed, and no changes are being made to the procedures relied upon to respond to an off-normal event as described in the UFSAR. As such, no new failure modes are being introduced.

The change does not alter assumptions made in the safety analysis and licensing basis.

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

3.

Do the proposed changes involve a significant reduction in a margin of safety?

Response: No The margin of safety is established through the design of the plant structures, systems, and components, the parameters within which the plant is operated, and the establishment of the setpoints for the actuation of equipment relied upon to respond to an event. The proposed change does not modify the safety limits or setpoints at which protective actions Page 8

ATTACHMENT 1 Evaluation of Proposed Changes are initiated, and does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above evaluation, Exelon concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 1 O CFR 50.92(c), and accordingly, a finding of no significant hazards consideration is justified.

4.4 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 the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

Exelon 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 in 1 O CFR 20, "Standards for Protection Against Radiation." However, the proposed amendment does 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. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 1 O CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review,"

paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, paragraph (b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

6.0

1.

2.

3.

4.

5.

6.

7.

REFERENCES DA-NS-08-051, Ginna Locked Rotor Accident Offsite and Control Room Doses CN-CRA-04-56, Steam Release for Dose for R.E. Ginna Unit 1 Extended Power Uprate NF-CB-19-044, "SBLOCA Evaluation for a Change to the MSSV Lift Setpoint Pressure Tolerance for R.E. Ginna and 10 CFR 50.46 Reporting Text, dated April 1, 2019 NF-CB-19-052, "Evaluation of MSSV Lift Pressure Tolerance Impact on the R.E. Ginna Design Transients", dated April 22, 2019 WCAP-16483-P, "Ginna Station Extended Power Uprate Project NSSS Engineering Report," September 2006 CN-SCS-04-36, Revision 1, "RGE - Design Transients for 19.5% Uprated Conditions,"

July 2004 CN-SCS-04-60, Revision 0, "RGE - Secondary Side Design Transients for 19.5%

Uprated Conditions," July 2004 Page 9

ATTACHMENT 2 Markup of Technical Specifications and Bases Pages Revised Pages 3.7.1-2 83.7.1-4 Page 10

MSSVs 3.7.1 SURVEILLANCE REQUIREMENTS SR 3.7.1.1 SURVEILLANCE FREQUENCY

- NOTE-Only required to be performed in MODES 1 and 2.

Verify each MSSV lift setpoint specified below in In accordance accordance with the INSERVICE TESTING PROGRAM.

with the Following testing, lift settings shall be within+/- 1 %.

INSERVICE VALVE NUMBER SGA SGB 3509 3508 3511 3510 3515 3512 3513 3514 LIFT SETTING (psig I 1 %, 3%)

1140 (psig +1.4%, -4%)

1140(psig+1.4%, -4%)

1140 (psig +1.4%, -4%)

1085 (psig + 1 %, -3%)

TESTING PROGRAM R.E. Ginna Nuclear Power Plant 3.7.1-2 Amendment 00, 124 Page 11

SURVEILLANCE REQUIREMENTS REFERENCES SR3.7.1.1 MSSVs B3.7.1 This SR verifies the OPERABILITY of the MSSVs by the verification of each MSSV lift setpoint in accordance with the INSERVICE TESTING PROGRAM. The ASME Code (Ref. 3), requires that safety and relief valve tests be performed in accordance with Appendix I of ASME OM Code-1998 (Ref. 4). According to Reference 4, the following tests are required:

a.

Visual examination;

b.

Seattightness determination;

c.

Setpoint pressure determination (lift setting);

d.

Compliance with owner's seat tightness criteria; and

e.

Verification of the balancing device integrity on balanced valves.

The ASME Standard requires that all valves be tested every 5 years, and a minimum of 20% of the valves be tested every 24 months. The ASME Code specifies the activities and frequencies necessary to satisfy the requirements. This SR allows a+ 1.4% and -~%for valves 3509, 3508, 3511, 3510, 3515, and 3512 and +1% and-3% forvalves3513 and 3514 setpoint tolerance for OPERABILITY; however, the valves are reset to+/--

1 % during the Surveillance to allow for drift.

This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. The MSSVs may be either bench tested or tested in situ at hot condtions using an assist device to simulate lift pressure. If the MSSVs are not tested at hot conditions, the lift setting pressure shall be corrected to ambient conditions of the valve at operating temperature and pressure.

1.

UFSAR, Section 10.3.2.4.

2.

UFSAR, Section 15.2.

3.

ASME Code for Operation and Maintenance of Nuclear Power Plants.

4.

Appendix I of ASME OM Code-1998.

R.E. Ginna Nuclear Power Plant B 3.7.1-4 Revision81