ML20217G646

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Application for Amend to License DPR-28,revising Station SW & Alternate Cooling Sys Requirements
ML20217G646
Person / Time
Site: Vermont Yankee Entergy icon.png
Issue date: 04/23/1998
From: Reid D
VERMONT YANKEE NUCLEAR POWER CORP.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20217G651 List:
References
BVY-98-52, NUDOCS 9804290181
Download: ML20217G646 (10)
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VERMONT YANKEE NUCLEAR POWER CORPORATION 185 Old Ferry Road, Brattleboro, VT 05301-7002 (802) 257-5271 April 23,1998 BVY 98-52 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

SUBJECT:

Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271)

Technical Specification Proposed Chance No. 200 Revision to Station Service Water and Alternate Cooline System Requirements.

In accordance with 10 CFR 50.90, Vermont Yankee Nuclear Power Corporation (VYNPC) hereby requests a change to Appendix A of the Vermont Yankee Facility Operating License to:

(1) Replace the allowance for continued operation with two inoperable SSW subsystems with a more conservative requirement to shutdown the unit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, (2) Relocate certain Station Service Water (SSW) and Residual lleat Removal Service Water (RHRSW) testing details to the Technical Requirements Manual (TRM), and (3) Replace references to SSW " subsystem" with " essential equipment cooling loop" ,

to more accurately reDect the VYNPS design l

The TRM will be incorporated by reference into the Final Safety Analysis Report (FS AR) upon implementation of this change. In addition, the Bases will be revised to omit statements which ,

could imply that the ACS could provide adequate heat removal capability following a postulated I accident. Further, Bases additions will be made which provide certain operability clari6 cations  !

relative to affected Speci6 cations.

l VYNPC has determined that this change does not involve a signincant hazards consideration l

l pursuant to 10CFR 50.92(c). A description of the amendment request is provided in Attachment

l. The no signincant hazards determination in support of the proposed Technical Specincation change is provided in Attachment 2. Attachment 3 provides the proposed revised Technical Speci6 cation pages.

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VERMONT YANKEE NUCLEAR POWER CORPORATIr N i

U. S. Nuclear Regulatory Commission Docket No. 50-271 BVY 98-52 Page 2 VYNPC requests review and approval of the proposed license amendment and Bases changes by 10/1/98.

Should there be any questions pertaining to this request, please contact our Licensing Manager, Mr. Gautam Sen at 802-258-4111.

Sincerely, VERMONT YANKEE NUCLEAR POWER CORPORATION I

Donald A. Reid Senior Vice President, Operations Attachments cc: USNRC Region 1 Administrator USNRC Resident inspector- VYNPS USNRC Project Manager- VYNPS _ , , , .

VT Department of Public Service g,gAfg

[ NOTAm ("

6 STATE OF VERMONT ) 1 g PUBllC

)ss WINDHAM COUNTY )

Then personally appeared before me, Donald A, Reid, who, being duly sworn, di t Operatens, of Vermont Yankee Nuclear Power Corporation, that he is duty authonzed to exec _

' , h ." President, he p

oing document in the name and on the behatf of Vermont Yankee Nuclear Power Corporation, and that the statemeEs'idsieTn are true to the best of his knowledge and belef. 4 AL =

Safly A. Sanditrum, Notary Public My Commission expires February 10,1999 l

VERMONT YANKEE NUCLEAR POWER CORPORATION U. S. Nuclear Regulatory Commission -

Docket No. 50-271 BVY 98-52/ Attachments Page 1 of 8 ATTACHMENT 1 DESCRIPTION OF AMENDMENT REQUEST Description This amendment request proposes changes to the existing requirements for the RIIR Service Water (RHRSW), Station Service Water (SSW) and Alternate Cooling Tower Systems (ACS) as identified in Technical Specifications (TS) 4.5.C and 3/4.5.D.

- Specifically, the changes proposed are as follows:

(1) . Page 105, Specifications 3.5.D.3 and 4.S.D.3: This requirement is revised to delete the existing allowance for 7 days of operation after both SSW subsystems are made or found to be inoperable.

(_2) ~ Page 103, Specification 4.5.C.1, and Page 104, Specification 4.5.D.1: These requirements have been revised to relocate testing information related to pump flow and pressure testing characteristics for the RHRSW 'and SSW Systems, respectively, to the TRM.

(3)' . Pages 104 and 105, Specifications 3.5.D.1,3.5.D.2,3.5.D.3,4.5.D.2,4.5.D.3 and associated Bases: All reference to SSW " subsystem" has been replaced by " essential equipment cooling loop" to more accurately reflect VYNPS design and operation. In addition, certain operability clarifications have been made to the Bases relative to affected )

Specifications.

(4) Page 111, Bases for Specification 3.5.D: The Bases have been revised to omit statements, which imply that the ACS could provide adequate heat removal following a postulated accident. Other Bases additions have been made which include certain operability clarifications relative to affected Specifications.

Backaround In November 1995 (Ref.1) and again in June 1996 (Ref. 2), Vermont Yankee communicated its intent and progress on fully converting the Vermont Yankee Custom Technical Specifications (CTS) to the Improved Standard Technical Specifications (ITS). It was intended that the i proposed changes contained herein would be addressed at that time. The original schedule for i submittal was identified as September 1997 with implementation following startup from the Spring 1998 refueling outage. The ITS conversion effort has been impacted significantly by the

VERMONT YANKEE NUCLEAR POWER CORPORATION 1

U. S. Nuclear Regulatory Commission Docket No. 50 271 l BVY 98-52/ Attachments Page 2 of 8 l

, reassignment of engineering resources to Design Bases Documentation development, FSAR l update and emergent work issues.

Regulatory Guidance (Ref. 3) related to Technical Specifications, which are not consistent with the safety analysis states that "the stafTposition is that upon discovering such conditions, the licensee should take the appropriate action to put the plant in a safe condition (such as imposing more conservative administrative limits), and also take action (such as requesting a license amendment) so that the TS represent the minimum requirements."

Since discovering that the SSW System requirement Bases are incorrect, administrative controls have been implemented to require a shutdown if both SSW Subsystems are made or found to be inoperable. These controls assure that operation is not continued based on the incorrect assumption that the ACS is capable of removing post-accident heat loads.

Reason / Basis for Chance l For Chance No.1 Both the SSW Subsystems and the ACS are required to be operable by TS 3.5.L).l. The SSW Subsystems are required to provide the heat removal capability necessary for a shutdown and to l remove post-accident heat loads follmving a postulated accident. The SSW Subsystems are l redundant to the extent that a single failure will not result in a loss of this safety function. The ACS provides the necessary heat removal capability for a safe shutdown following a loss of the normal service water source and capability due to either a failure of the Vernon Dam on the Connecticut River, flooding of the SSW pump room, or a fire in the SSW pump room. However, the ACS does not have sufficient capacity to remove post-accident heat loads as implied in the Bases for TS 3.5.D. The ACS is designed to provide heat removal capability following a reactor shutdown and is designed to accept the worst case non-accident shutdown cooling loads. The ACS is designed to be aligned and operated in a controlled fashion; two hours are required to properly align the system. Because of the time required for proper alignment, the ACS is not designed to accept the consequences of a design basis loss-of-coolant accident. The current j requirements, i.e., TS 3.5.D.3, include an allowance for continued operation with both SSW l l Subsystems inoperable. This allowance is incorrectly based on the assumption that the ACS is  ;

! able to fulfill the post-accident heat removal requirements when both SSW Subsystems are l unavailable. Since the ACS is not capable of fulfilling this backup role, the allowance for seven days of operation with both SSW Subsystems inoperable is removed, and a requirement to j shutdown the unit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is provided in its place. l l

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VERMONT YANKEE NUCLEAR POWER CORPOR ATION U. S. Nuclear Regulatory Commission Docket No. 50-271 BVY 98-52/ Attachments Page 3 of 8 l

The allowance for continued operation for seven days with an inoperable ACS is not affected by this change. The low probability of either a dam failure, Gre in the SSW pump room or flooding in the SSW pump room which would require use of the ACS for shutdown of the unit provides the basis for this allowed outage time. This basis is not affected and, therefore, no changes are necessary for the ACS requirements.

For Chance No.2 NUREG-1433, Rev.1, does not establish and maintain procedurally related testing details in the Improved Technical Speci6 cations for the SSW or RHRSW systems. Therefore, TS 4.5.C.1 and TS 4.5.D.1 testing details for the RHRSW and SSW systems, respectively, are proposed to be relocated to the TRM under the control of 10 CFR 50.59. These controls are adequate to ensure l the required testing is performed to verify operability. As such, these relocated details are not required to be in the Technical Speci6 cations to provide adequate protection of the public health and safety. Changes to these relocated requirements in the TRM will be controlled by 10 CFR 50.59.

For Chance No.3 The VYNPS SSW system consists of four SSW pumps, associated valves and piping, one nonessential equipment cooling loop, and two essential equipment cooling loops and additional essential and nonessential equipment cooling loads. The essential equipment cooling loops provide redundant capability for analyzed accidents or transients. Two operable SSW pumps with one or both essential equipment cooling loops in operation will provide adequate cooling for analyzed accidents or transients. The wording of Specincations 3.5.D.1,3.5.D.2,3.5.D.3, 4.5.D.2,4.5.D.3 and associated Bases has therefore been revised as necessary to replace

" subsystem" with " essential equipment cooling loop" to more accurately reflect VYNPS design  :

and operation.

i For Chance No.4  :

The Bases for TS 3.5.D state: "The SSW Subsystems and the Alternate Cooling Tower System provide alternate heat sinks to dissipate residual heat after a shutdown or accident. Each SSW Subsystem and the ACS provides sufficient heat sink capacity to perform the required heat dissipation. The Alternate Cooling Tower System will provide the necessary heat sink in the event both SSW Subsystems become incapacitated due to a loss of the Vernon Dam with subsequent loss of the Vernon Pond." As indicated above, the Srst two sentences are incorrect

in their discussion of the capabilities of the Alternate Cooling Tower System and they are proposed to be revised to correctly reDect the capabilities of this system. Further Bases additions l

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VERMONT YANKEE NUCLEAR POWER CORPORATION J

U. S. Nuclear Regulatory Commission Docket No. 50-271 BVY 98-52/ Attachments Page 4 of 8 )

have been made which provide certain operability clarifications relative to affected Specifications.

References

1. Letter, VYNPC to USNRC, BVY 95-122, dated November 13,1995.
2. Letter, VYNPC to USNRC, BVY 96-82, dated June 21,1996.
3. SECY-97-035, dated February 12,1997.

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VERMONT YANKEE NUCLEAR POWER CORPORATION U. S. Nuclear Regulatory Commission Docke.t No. 50-271 BVY 98-52/ Attachments Page 5 of 8 ATTACllMENT 2 NO SIGNIFICANT llAZARDS CONSIDERATION

' Description This amendment request proposes changes to the existing requirements for the RiiR Service Water (RHRSW), Station Service Water (SSW) and Alternate Cooling Tower Sysiems (ACS) as identified in Technical Specifications (TS) 4.5.C and 3/4.5.D.

Specifically, the changes proposed are as follows:

(1) Page 105, Specifications 3.5.D.3 and 4.5.D.3: This requirement is revised to delete the existing allowance for 7 days of operation after both SSW subsystems are made or found to be inoperable.

.(2) Page 103, Specification 4.5.C.1, and Page 104, Specification 4.5.D.1: These requirements have been revised to relocate testing information related to pump flow and pressure testing characteristics for the RliRSW and SSW Systems, respectively, to the TRM.

(3) Pages 104 and 105, Specifications 3.5.D.1,3.5.D.2,3.5.D.3,4.5.D.2,4.5.D.3 and associated Bases: All reference to SSW " subsystem" has been replaced by " essential equipment cooling loop" to more accurately reflect VYNPS design and operation. In addition, certain operability clarifications have been made to the Bases relative to affected Specifications.

I (4) Page 111, Bases for Specification 3.5.D: The Bases have been revised to omit statements, which imply that the ACS could provide adequate heat removal following a postulated accident. Other Bases additions have been made which include certain operability clarifications relative to affected Specifications.

In accordance with the criteria set forth in 10 CFR 50.92, the Vermont Yankee Nuclear Power Corporation has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:

FOR CHANGE NO.1

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

VERMONT YANKEE NUCLEAR POWER CORPORATION

' U. S. Nuclear Regulatory Commission Docket No. 50-271.  !

BVY 98-52/ Attachments l Page 6 of 8 j This change deletes the existing allowance for 7 days of operation after both Station Service Water (SSW) subsystems are made or found to be inoperable. At least one subsystem of the SSW System is required to be operable to mitigate the consequences of a design basis accident. Therefore, with both subsystems inoperable, the unit is required to shut down. Current Technical SpeciHcations (TS) erroneously allow 7 days of operation afler both SSW subsystems are made or found to be inoperable before requiring _

that the reactor be placed in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This allowance is incorrectly based on the assumption that the Alternate Cooling Tower System (ACS) is able to fulHil the post-accident heat removal requirements when both SSW Subsystems are made or

' found to be inoperable. Since the ACS is not capable of fulfilling this backup role, the l allowance for seven days of operation with both SSW Subsystems inoperable is removed, and a requirement i.o shutdown the unit is provided in its place. This proposed change -

deletes the allowance for 7 days of operation in this condition, and instead requires an orderly shutdown to be initiated and the reactor to be placed in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Since the same amount of time is allowed to conduct the required shutdown, this change will not significantly increase the consequences of any previously analyzed accident. In addition, the SSW system is not considered to be the initiator of any previously analyzed accident. Therefore, this change will not significantly increase the probability or consequences of any previously analyzed accident.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

This change will not physically alter the plant (no new or different types of equipment will be installed). The changes in methods governing normal plant operation are consistent with the current safety analysis assumptions. Therefore, this change will not create the possibility of a new or different kind of accider.t from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

l ,

l This change deletes the existing allowance for 7 days of operation after both SSW  !

subsystems are made or found to be inoperable. At least one subsystem of the SSW i System is required to be operable to mitigate the consequences of a design basis accident. j

~

Therefore, with both subsystems inoperable, the unit is required to be shut down.

Current TS requirements erroneously allow 7 days of operation after both SSW subsystems are made or found to be inoperable before requiring that the reactor be placed in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This allowance is incorrectly based on the assumption l that the ACS is able to fulfill the post-accident heat removal requirements when both SSW Subsystems are inoperable. Since the ACS is not capable of fulfilling this backup

VERMONT YANKEE NUCLEAR POWER CORPORATION U. S. Nuclear Regulatory Commission Docket No. 50-271 BVY 98-52/ Attachments Page 7 of 8 role, the allowance for seven days of operation with both SSW Subsystems inoperable is removed, and a requirement to shutdown the unit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is provided in its place.

Therefore, elimination of the allowance for 7 days of operation with both SSW subsystems inoperable does not involve a significant reduction in a margin of safety.

FOR CHANGE NO.2

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change relocates testing information details for the Residual Heat Removal Service Water (RHRSW) and Station Service Water (SSW) systems, respectively, to the Technical Requirements Manual (TRM) under the control of 10 CFR 50.59. These controls are adequate to ensure the required testing is performed to verify operability. As such, these relocated details are not required to be in the Technical Specifications to provide adequate protection of the public health and safety. Changes to these relocated requirements in the TRM will be controlled by 10 CFR 50.59. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed change will not impose or eliminate any requirements and adequate control of the information will be maintained. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a. margin of safety?

The proposed change will not reduce a margin of safety because the simple relocation of l testing details from the TS to the TRM has no impact on any safety analyses assumptions.

l Since any future changes to these requirements will be evaluated per the requirements of l 10 CFR 50.59, no reduction in a margin of safety will be allowed. Therefore, this change does not involve a significant reduction in the margin of safety.

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VERMONT YANKEE NUCLEAR POWER CORPORATION U. S. Nuclear Regulatory Commission Docket No. 50-271 BVY 98-52/ Attachments Page 8 of 8 FOR CHANGE NO.3 l

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

This change proposes to revise the wording of Station Service Water (SSW)

Specifications to replace " subsystem" with " essential equipment cooling loop" to more accurately reflect VYNPS design and operation. At least two SSW pumps and one essential equipment cooling loop of the SSW System are required to be operable to 4 mitigate the consequences of a design basis accident. Since this proposed change represents no change to existing requirements, this change will not significantly increase the consequences of any previously analyzed accident. In addition, SSW is not considered to be the mitiator of any previously analyzed accident. Therefore, this change will not I significantly increase the probability or consequences of any previously analyzed accident.  !

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2. Does the change create the possibility of a new or different kind of accident from any  !

accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed change will not impose or eliminate any requirements and adequate control of existing requirements will be maintained. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The proposed change continues to provide the previous margin of safety regarding the capability to remove post-accident heat loads. At least two SSW pumps and one essential equipment cooling loop will be required to be operable or the unit will be required to be shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Since this is the same basis both before and after the change, this change will not involve a significant reduction in a margin of safety.

Based on the above discussion, we have determined that this change does not constitute a significant hazard consideration as defined in 10CFR 50.92(c).

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