SBK-L-08005, License Amendment Request 07-04 Application for Amendment to Delete Post-Maintenance Testing Surveillance Requirement for Containment Isolation Valves

License Amendment Request 07-04 Application for Amendment to Delete Post-Maintenance Testing Surveillance Requirement for Containment Isolation Valves

ML080440304
Person / Time
Site:	Seabrook
Issue date:	02/08/2008
From:	St.Pierre G Florida Power & Light Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR 07-004, SBK-L-08005
Download: ML080440304 (14)

Text

FPL Energy Seabrook Station SFPL Energy P.O. Box 300 Seabrook, NH 03874 Seabrook Station (603) 773-7000 February 8, 2008 SBK-L-08005 Docket No. 50-443 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Seabrook Station License Amendment Request 07-04 "Application for Amendment to Delete Post-Maintenance Testing Surveillance Requirement for Containment Isolation Valves" In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), FPL Energy Seabrook, LLC (FPL Energy Seabrook) is submitting License Amendment Request (LAR) 07704 for an amendment to the technical specifications (TS) for Seabrook Station.

This LAR proposes a change to the Seabrook Station Technical Specifications to delete Surveillance Requirement 4.6.3.1, which specifies post-maintenance testing requirements for containment isolation valves. The proposed change will eliminate unnecessary testing and provide flexibility in determining the proper post-maintenance test. Appropriate testing will continue to ensure operability of containment isolation valves following maintenance activities. The Enclosure contains FPL Energy Seabrook's evaluation of the proposed amendment and includes a mark-up of the TS pages showing the proposed changes.

As discussed in the enclosed LAR, the proposed change does not involve a significant hazard consideration pursuant to 10 CFR 50.92. A copy of this letter and the enclosed LAR has been forwarded to the New Hampshire State Liaison Officer pursuant to 10 CFR 50.91(b). FPL Energy Seabrook has determined that LAR 07-04 meets the criteria of 10 CFR 51.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement. The Station Operation Review Committee has reviewed this LAR.

an FPL Group company

United States Nuclear Regulatory Commission SBK-L-08005 / Page 2 FPL Energy Seabrook requests NRC Staff review and approval of LAR 07-04 with issuance of a license amendment by February 28, 2009 and implementation of the amendment within 90 days.

Should you have any questions regarding this letter, please contact Mr. James M.

Peschel, Regulatory Programs Manager, at (603) 773-7194.

Very truly yours, FPL Energy Seabrook, LLC.

Gene F. St. Pierre Site Vice President

Enclosure:

FPL Energy Seabrook's Evaluation of the Proposed Change cc:

S. J. Collins, NRC Region I Administrator G. E. Miller, NRC Project Manager, Project Directorate 1-2 W. J. Raymond, NRC Senior Resident Inspector Mr. Christopher M. Pope, Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305

Seabrook Station AFFIDAVIT The following information is enclosed in support of this License Amendment Request:

0 Enclosure FPL Energy Seabrook's Evaluation of the Proposed Change I, Gene St. Pierre, Site Vice President of FPL Energy Seabrook, LLC hereby affirm that the information and statements contained within this License Amendment Request are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.

Sworn and Subscribed before me this

j.

dayof r-C.-6Jra,2

,2008 Gen Notary Pblic ie St. Pierre

ENCLOSURE FPL Energy - Seabrook's Evaluation of the Proposed Change

Subject:

License Amendment Request 07-04, "Application for Amendment to Delete Post-Maintenance Testing Surveillance Requirement for Containment Isolation Valves" 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

Attachment Technical Specification Page Markups 1

1.0

SUMMARY

DESCRIPTION This LAR proposes a change to the Seabrook Station Technical Specifications (TS) to delete Surveillance Requirement (SR) 4.6.3.1, which specifies post-maintenance testing requirements for containment isolation valves (CIV). This change will eliminate unnecessary testing and provide flexibility in determining the proper post-maintenance testing for CIVs being returned to service following maintenance activities. Appropriate testing will continue to ensure operability of CIVs after maintenance, repair, or replacement work is performed on the valves.

2.0 DETAILED DESCRIPTION Seabrook Station Technical Specification SR 4.6.3.1 (below) will be deleted with this proposed change as shown in the Attachment.

4.6.3.1 Each containment isolation valve shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair, or replacement work is performed on the valve or its associated actuator, control, or power circuit by performance of a cycling test and verification of isolation time.

Condition the Proposed Amendment is Intended to Resolve SR 4.6.3.1 requires that each containment isolation valve shall be demonstrated operable prior to returning the valve to service after maintenance, repair, or replacement work is performed on the valve or its associated actuator, control, or power circuit by performance of a cycling test and verification of isolation time.

While the purpose of SRs is to verify operability of TS-required equipment, performing a SR as a post-maintenance test may not be necessary in all instances.

The type and scope of work performed on a component will determine if the maintenance activity could adversely affect operation of the component and should be the basis for determining the post-maintenance test. However, in instances involving maintenance that has no potential to either render a CIV inoperable or adversely affect the CIV's ability to function, SR 4.6.3.1 specifies unnecessary testing by requiring a post-maintenance stroke test and verification of isolation time. For example, a maintenance activity that applies lubricant to a valve stem would neither render the valve inoperable nor adversely affect the valve's ability to function. Nonetheless, following such a maintenance activity, SR 4.6.3.1 unnecessarily requires a valve stroke test and verification of isolation time.

The proposed change will eliminate SR 4.6.3.1 because a stroke test and verification of isolation time following maintenance on a CIV is not always warranted. The change will minimize unnecessary testing and provide flexibility 2

in determining the appropriate post-maintenance test for a CIV based on the type and scope of work performed.

3.0 TECHNICAL EVALUATION

Description of the Containment Isolation System The containment isolation system is comprised of the valves, piping and actuators required to isolate the containment following a loss of coolant accident or main steam line rupture. The system is specifically designed to isolate lines penetrating the containment, which are not required for operation of the engineered safety feature systems, in the event of a loss of coolant accident or a main steam line break. The system establishes and maintains isolation of the containment from the outside environment to prevent the release of fission products and to ensure the public is protected in accordance with 10 CFR 100 guidelines. Upon receipt

.of a containment isolation signal, automatic containment isolation valves actuate to the closed position with valve stroke times that ensure any radioactive release to the environment during a design basis accident is within the limits of 10 CFR 100.

Evaluation Seabrook Station TS 3.6.3 requires operable CIVs in Modes 1 through 4. In addition to the post-maintenance testing required by SR 4.6.3.1, two additional SRs demonstrate operability of the CIVs. On an 18-month frequency, SR 4.6.3.2 verifies that each CIV actuates to its isolation position on a containment isolation signal. Further, SR 4.6.3.3 confirms that the isolation time of each CIV is within limits when tested pursuant to TS 4.0.5 (Surveillance Requirements for Inservice Inspection and Testing of ASME Code Class 1, 2, and 3 Components).

SR 4.0.1 stipulates that SRs must be met during the Mode or other specified condition in the applicability of a limiting condition for operation. When maintenance activities could either render a CIV inoperable or adversely affect the CIV's ability to function, SR 4.0.1 would require performing the appropriate SRs prior to returning the valve to operable status. Furthermore, implicit in the definition of operability, post-maintenance testing is already required to the extent necessary to ensure a TS-required component is operable after undergoing maintenance. As a result, SR 4.6.3.1 is an overly restrictive and duplicative statement of requirements that already exist in the TS.

Deleting TS 4.6.3.1 will provide flexibility in determining the appropriate post-maintenance test based on the work performed. Post-maintenance testing is controlled by Seabrook Station procedures that specify appropriate testing following maintenance activities. When maintenance activities could adversely affect the operability of a component, post-maintenance testing will include 3

performing the appropriate TS SRs prior to returning the component to an operable status.

The plant procedure that controls post-maintenance testing states that the scope of such testing should be based on the extent of maintenance performed. Further, it defines a satisfactory post-maintenance test as one that verifies the ability of a component or system to perform its intended function. The guidance for post-maintenance testing includes provisions for verifying the operability of TS required equipment and confirming that maintenance and testing activities do not invalidate SRs. The owners of various station programs (Motor-Operated Valves, Inservice Testing, etc.) and system components are responsible for identifying in station documents test requirements associated with their programs and components. The plant procedure directs using these documents when developing post-maintenance test requirements. Included in the post-maintenance testing procedure for 47 categories of plant equipment are Component Test Guides that consider test requirements based on the extent of maintenance performed on the components. For example, in the case of motor-operated valves, the guidance indicates that following maintenance on valve packing, the post-maintenance test should include a valve exercise and stroke time test. On the other hand, after painting, cleaning, grease inspection or other cosmetic maintenance, no testing is required.

Senior Reactor Operator licensed personnel (SROs) in the Operations Department are responsible for determining the operability of TS-required equipment affected by maintenance activities. Prior to authorizing the commencement of work, an Operations SRO reviews the work activity and post-maintenance test. The purpose of this review is to ensure that the testing specified will restore operability of the equipment affected by the maintenance activity. At the conclusion of the work activity, SRO approval of the completed work includes a determination that the post-maintenance test has restored the equipment to an operable status.

The proposed change is consistent with the treatment of CIVs in NUREG-1431, Revision 3, Westinghouse Standard Technical Specifications (STS) [Reference 1].

SR 3.6.3.5 in the STS states "Verify the isolation time of each power operated containment isolation valve is within limits". This SR is strictly a periodic requirement with no reference to the conduct of maintenance. With the deletion of Seabrook Station SR 4.6.3.1, remaining SRs 4.6.3.2 and 4.6.3.3 retain requirements for verification of operability of CIVs that are equivalent to those contained in the STS.

Summary FPL Energy Seabrook proposes to delete SR 4.6.3.1 on the basis that existing SRs 4.6.3.2 and 4.6.3.3 adequately demonstrate operability of the CIVs, including a verification of isolation time. After performing maintenance on CIVs, an 4

evaluation of the post-maintenance test requirements in accordance with station procedures, will consider the need for, and perform as appropriate, a valve stroke test and verification of isolation time.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements / Criteria 0

10 CFR 50.36 [Reference 2] requires that the technical specifications include surveillance requirements; which relate 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 eliminates the post-maintenance surveillance requirement for containment isolation valves. However, remaining SRs 4.6.3.2, 4.6.3.3, and 4.0.5 continue to satisfy the requirements of 10 CFR 50.36. The function of the CIVs is to actuate to their fisolation position on a containment isolation signal, and the remaining SRs confirm that the valves automatically actuate to the isolation position within the required time.

10 CFR 50.55a [Reference 3] specifies the inservice testing requirements for ASME Code Classl, 2, and 3 pumps and valves.

The proposed elimination of TS 4.6.3.1 makes no changes to the inservice testing requirements for C1Vs. The valves will continue to be tested in accordance with 10 CFR 50.55a as stipulated by Seabrook Station TS 4.0.5, surveillance requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components.

Criterion 16 of Appendix A to 10 CFR Part 50 --Containment design, includes a requirement for the reactor containment and associated systems to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

The integrity of containment is unaffected by the proposed change. The CIVs will continue to establish and maintain isolation of the containment from the outside environment to prevent the release of fission products and to ensure the public is protected in accordance with 10 CFR 100 guidelines 5

4.2 Precedent On December 21, 2001, the NRC approved the deletion of SR 4.6.3.1, post-maintenance testing surveillance requirement of CIVs, in Amendment 200 for Millstone Unit 3 [Reference 4]. Similarly, on August 31, 2006 Salem Units 1 and 2 received Amendments 274 and 255, respectively, which deleted the SR for post-maintenance testing of CIVs [Reference 5].

4.3 Significant Hazards Consideration No Significant Hazards Consideration In accordance with 10 CFR 50.92, FPL Energy Seabrook has concluded that the proposed changes do not involve a significant hazards consideration (SHC). The basis for the conclusion that the proposed changes do not involve a SHC is as follows:

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

The proposed amendment to the technical specifications, which is consistent with NUREG-1431, "Standard Technical Specifications, Westinghouse Plants," removes the surveillance requirement related to post-maintenance testing of containment isolation valves (CIVs).

Surveillance requirements are not initiators of accidents; consequently, the proposed change does not significantly increase the probability of an accident previously evaluated.

The proposed change does not alter the requirements regarding operability of CIVs, and appropriate testing will continue to confirm the operability of these valves following maintenance activities.

The CIVs will continue to be tested in a manner and at a frequency that demonstrates they remain capable of performing their intended safety function. As a result, the proposed amendment does not significantly affect the consequences of an accident previously evaluated.

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

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

The proposed change does not introduce any new accident scenarios, failure mechanisms, or single failures. The change does not add new equipment to the plant, does not modify or remove existing equipment, 6

and does not significantly change the operation of the plant. The ability of any operable equipment to perform its specified safety function is unaffected by this change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3.

The proposed changes do not involve a significant reduction in the margin of s afe ty.

The proposed change does not alter the initial conditions or results of any accident analyses. The operability requirements, performance, and design of the CJVs are unchanged with this proposed change. The CIVs will continue to meet the design bases for the containment isolation system as described in the Seabrook Station UFSAR. The proposed amendment will minimize unnecessary testing of CIVs. Therefore, the proposed change does not involve a significant reduction in the margin of safety.

Based on the above, FPL Energy Seabrook concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 5 0.92 (c), and accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusions In conclusion, based on the considerations discussed previously, (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

The proposed amendment does not involve (i) a significant hazards consideration, (i1) 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 10 CFR 51.22 (c) (9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement of environmental assessment need be prepared in connection with the proposed amendment.

7

6.0 REFERENCES

1. NUREG-1431, Standard Technical Specifications Westinghouse Plant, Revision 3.

2.

10 CFR 50.36, Technical Specifications.

3. 10 CFR 50.55a, Codes and Standards.

4. Millstone Nuclear Power Station, Amendment No. 200 to License No. NPF-49, December 21, 2001 (Accession number ML013330062).

5. Salem Nuclear Generating Station, Amendment No. 274 to License No. DPR-70, Amendment No. 255 to License No. DPR-75, August 31, 2006 (Accession Number ML062140500).

8

Attachrnent Proposed Technical Specification Change (mark-up)

Refer to the attached markup of the proposed changes to the Technical Specifications.

The attached markup reflects the currently issued version of the Technical Specifications.

At the time of submittal, the Technical Specifications were revised through Amendment No. 116. Pending Technical Specifications or Technical Specification changes issued subsequent to this submittal are not reflected in the enclosed markup.

Listed below are the license amendment requests that are awaiting NRC approval and may impact the currently issued version of the Technical Specifications affected by this LAR.

Date LAR Title FPL Energy Seabrook Letter Submitted None The following Technical Specifications are included in the attached markup:

Technical Specification Title Page 3/4 6-16 3/4 6-17 (info only)

TS 3.6.3 Containment Isolation Valves 9

.CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 Each containment isolation valve shall be OPERABLE*.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With one or more of the isolation valve(s) inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and:

a.

Restore the inoperable valve(s) to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or

b.

Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one deactivated automatic valve secured in the isolation position, or

c.

Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual valve or blind flange; or

d.

Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.3.1 ach containment isolati n valve shallbe demonstrated OPERABLE prior to I

ithe valv servic etmaintea repair,r lacemetor is *efrrned tonhe valve its assope.ted ac ontrol, or,.poder circuit byp~erfor" mance of a cycling test and erification of isolation time.

• Locked or sealed closed valves may be opened on an intermittent basis under administrative control.

SEABROOK - UNIT 1 3/4 6-16 Amendment No,&

~A~'c11-1-1,11S Ale CONTAINMENT SYSTEMS CONTAINMENT ISOLATION VALVES SURVEILLANCE REQUIREMENTS 4.6.3.2 Each containment isolation valve shall be demonstrated OPERABLE during shutdown at least once per 18 months by:

a.

Verifying that on a Phase "A" Isolation test signal, each Phase "A" Isolation valve actuates to its isolation position,

b.

Verifying that on a Phase "B" Isolation test signal, each Phase "B" Isolation valve actuates to its isolation position, and

c.

Verifying that on a Containment Purge and Exhaust Isolation test signal, each purge and exhaust valve actuates to its isolation position.

4.6.3.3 The isolation time of each power-operated or automatic containment isolation valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5.

SEABROOK - UNIT 1346-7AedntN.4 3/46-17 Amendment No. 34