ML20206H362

Application for Amend to License DPR-28,to Delete Specific Leak Rate Requirements of TS 3.7.A.4 & 4.7.A.4 for Main Steam Line Isolation Valves

ML20206H362
Person / Time
Site:	Vermont Yankee
Issue date:	05/06/1999
From:	Wanczyk R VERMONT YANKEE NUCLEAR POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20206H365	List: ML20206H362 ML20206H377
References
BVY-99-68, NUDOCS 9905110184
Download: ML20206H362 (10)

Text

e j VERMONT YANKEE ,

y NUCLEAR POWER-CORPORATION ]

I 185 Old Ferry Road, Brattleboro, VT 05301 7002 l (802) 257-5?71 l May 6,1999 BVY 99-68 U.S. Nuclear Regulatory Commission l A'ITN: Document Control Desk Washington, DC 20555

Subject:

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

Technical Specification Proposed Change No. 215 Removal of Main Steam Line Isolation Valve Leakane Snecifications Pursuant to 10CFR50.90, Vermont Yankee (VY) hereby proposes to amend its Facility Operating License, DPR-28, by incorporating the attached proposed change into the VY Technical Specifications.

This proposed change deletes the specific leak rate requirements of Technical Specifications 3.7.A.4 and 4.7.A.4 for the main steam line isolation valves.

In the transition from 10CFR50 Appendix J, Option A to Option B, as approved per Facility Operating License Amendment No.152, the specific ac::eptance criteria for the main steam line isolation valves was retained. However, with the advent of the Primary Containment Leak Rate Testing Program Plan (PCLRTP) created to implement Option B, the specific valve leak rate requirements (acceptance criteria) of the Technical Specifications could be deleted and the leak rate limits controlled in the implementing program plan. VY is now proposing to exercise this option to remove the specific leak rate acceptance criteria for the main steam line isolation valves from the Technical Specifications since the criteria is redundant, is obsolete per 10CFR50 Appendix J, and constitutes a burden with no significant gain in safety. Main steam line isolation valve leakage is a component of the Technical Specificatiori 6.15, Primary Containment Leak Rate Testing Program, combined local leak rate acceptance criteria and will continue to be governed by that specification. The specific main steam line isolation valve leak rate criteria will be relocated to the PCLRTP, which implements Technical Specification 6.15. ,

Attachment i to this letter contains supporting information and the safety assessment of the proposed change. Attachment 2 contains the determination of no significant hazards consideration. Attachment 3 i

provides the mark up version of the current Technical Specification page. Attachmer.t 4 is the retyped Technical Specification page.

VY has reviewed the proposed Technical Specification change in accordance with 10CFR50.92 and 00 concludes that the proposed change does not involve a significant hazards consideration.

9905110184 990506 31 PDR ADOCK 05000271 P PDR ,

Vnunoxr Y^m NmAn I*owen Counmu nn BVY 99-68 \ Pcgi 2 Pursuant to 10CFR50.91(b)(1), we have provided a copy of this proposed change and the associated no significant hazards consideration to the appropriate State of Vermont representative.

VY has also determined that the proposed change satisfies the criteria for a categorical exclusion in accordance with 10CFR51.22(c)(9) and does not require an environmental review. Therefore, pursuant to 10CFR51.22(b), no environmental impact statement or environmental assessment needs to be prepared for this change.

We request that the Staff issue the subject license amendment no later than September 1999, in order to ensure implementation prior to the next scheduled refueling outage in October 1999.

If you have any questions on this transmittal, please contact Mr. Jim Devincentis at (802) 258-4236.

Sincerely, VERMONT YANKEC NUCLEAR POWER CORPORATION j

] /2f .

Robert J. Wanc k '(/

Director of Safety and Regulatory Affairs STATE OF VERMONT ) i S hgD i

)ss C WINDHAM COUNTY '

) NOTAM i I

Then personally appeared before me, Robert J. Wanczyk, who, being duly sworn, di e thfW8klGet fof Safety and Regulatory Affairs of Vermont Yankee Nuclear Power Corporation, that he i au ized to cut and file the foregoing document in the name and on the behalf of Vermont Yankee Nucle er ion -

that the statements therein are true to the best of his knowledge and belief. Y4/

~

Sally A. 5fandstrum, Notary Public My Commission Expires February 10,2003 Attachments cc: USNRC Region ! Administrator USNRC Resident inspector - VYNPS USNRC Project Manager - VYNPS Vermont Department of Public Service

Yl?HMONT YASKlili NL'CI.liAllI'OWillt COHl*0HATION Docket No. 50-271 BVY 99 Attachment i Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 215 Removal of Main Steam Line Isolation Valve Leakage Specifications Supporting Information and Safety Assessment of Proposed Change 1

i i

1

BVY 99-68'/ Attachment I / Page l'

' numi YAmm Nectr^u Powim com onArus INTitODUCTION Option B to 10CFR50 Appendix J was implemented at Vermont Yankee (VY) per Facility Operating-License AmendmentNo.1528 With the advent of the Primary Containment Leak Rate Testing Program Plan (PCLRTP) created to implement Option B, the specific valve leak rate requirements (acceptance criteria) of the Technical Specifications could be eliminated. . The individual valve leak rates are components of the combined local leak rate as governed by Technical Specification 6.15, and control of the specific valve leak rate limits is provided by the PCLRTP that implements Technical Specification - 6.15.

Therefore, the following paraphrased text "the leakage from any one isolation valve shall not exceed 5 percent of the allowable leak rate" was removed from the VY Technical Specifications 3.7.A.4 and 4.7.A.4 per Amendment 152. However, the specific acceptance criteria for the main steam line isolation valves was retained in the transition from 10CFR50 Appendix J, Option A to Option B.

Technical Specifications 3.7.A.4 and 4.7.A.4 contain main steam line isolation valve leakage acceptance criteria of 15.5 scf/hr at 44 psig (Pa) and 11.5 scf/hr at 24 psig (Pt) . The 15.5 scf/hr is equal to the-calculated value of 0.05La (i.e., the pre Amendment 152 individual valve leakage rate limit discussed above, where Laequals 0.8 wt%/ day (309.890 scf/hr]). The original and present Technical Specification surveillance requirement leakage acceptance criterion for the main steam line isolation valves is 11.5 scf/hr, when tested at a reduced pressure of 24 psig (P t). During the post rule making-implementation period for the original 10CFR50 Appendix J (circa 1972), VY requested relief from.the -

requirement to test the main steam line isolation valves at 44 psig (Pa). In the Safety Evaluation dated '

August 19,19832, VY was granted relief to continue the testing of the main steam line isolation valves by pressurizing between the inboard and outboard valves at a reduced pressure of 24 psig (P t ). VY will continue to utilize a reduced pressure test at 24 psig (Pt ). Pressurizing the main steam line isolation valves between the inboard and outboard valves.

VY is now proposing to exercise the option to remove from Technical Specifications the-specific acceptance criteria for the main steam line isolation valves. Main steam line isolation valve leakage is a component of the Technical Specification 6.15 combined local leak rate acceptance criteria and, as such, will continue to be governed by that specification. The specific main steam line isolation valve leak rate criteria will be relocated to the PCLRTP. .

Maintaining the specific main steam line isolation valve leak rate criteria in Technical Specifications is redundant, is obsolete per 10CFR50 Appendix J, and constitutes a burden with no significant gain. in safety. Due to the prescriptive requirement, maintenance is required whenever a leak rate is even slightly above the specific acceptance criterion (i.e.,11.5 scf/hr). If the combined local leak rate acceptance criterion of 0.6La (maximum or minimum pathway basis depending on operating mode as prescribed by Technical Specification 6.15) is not challenged, the maintenance is unnecessary. Maintenance conducted to fulfill the redundant specification results in significant worker exposure and increased outage duration.

1 Letter USNRC to VYNPC, NVY 98-24," Issuance of Amendment No.152 to Facility Operating License No. DPR 28, VYNPS (TAC No. M99264)," dated February 26,1998.

{ Letter USNRC to VYNPC, NVY 83-192,." Exemption from Cenain Requirements of Section 50.54(o) and Appendix J of 10 CFR 50," dated August 19,1983.

f 4 b

+ w a

BVY 99-68 / Attachment I / Page 2 -

The PCLRTP implements the requirements of 10CFR50, Appendix J Option B, and is consistent with the U.S; Nuclear Regulatory Commission Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program," dated September 1995. The technical methods and techniques for performing the Types A, B, and C tests are in accordance with the ANSI /ANS 56.8-1994, " Containment System Leakage Testing Requirements." The program plan follows the Nuclear Energy Institute's " Industry Guideline for l Implementing Performance Based Option of 10CFR50, Appendix J,"(NEI 94-01, Revision 0) for both l the Type A and the Types B and C Test Programs. Additionally, the program plan implements the requirements of 10CFR50.55a, " Codes and Standards," Section (b)(2)(vii) Inservice testing of containment isolation valves to be analyzed in accordance with 4.2.2.3(e) and corrective actions performed in accordance 4.2.2.3(f) of Part 10 of ASME/ ANSI OMa-1988 Addenda to ASME/ ANSI OM-1987.

The documents that provide the requirements and the approved guidelines for implementation of the Option B leak rate testing program, permit the establishment of Administrative and Corrective Action Limits for the determination of valve leak rate performance. When a Corrective Action Limit is exceeded, the containment isolation valves "...shall be declared inoperable and either repaired or replaced." Corrective Action Limits are established and controlled in the PCLRTP for all Type C local leak rate tested valves, including the main steam line isolation valves. Presently the Corrective Action Limit for a main steam line isolation valve is the Technical Specification acceptance criteria of 11.5 scf/hr. Administrative controls will be established to limit, prior to entering a mode of operation where Primary Containment Integrity is required, the combined total leak rate of the main steam line -

isolation valves based on the maximum pathway and consistent with the analyses of record, including the main control room habitability analysis'^5 Appropriate Corrective Action Limits will be established for the valves in accordance with the PCLRTP.

The analysis and the repair or replacement process required by the ASME/ ANSI OMa 1988', Part 10 paragraph 4.2.2.3(f) and the corrective action and causal determination process of NEI 94-01, Revision 0, are included in the PCLRTP. The PCLRTP and other plant procedures control the repair and subsequent retest requirements for all leak rate tested valves, including the main steam line isolation valves.

SAFETY ASSESSMENT The proposed change is not a safety concern and can be implemented without endangering the public health and safety because: .

1. Regulatory Guide 1.163 provides specific guidance concerning the implementation of Option B to 10CFR50 Appendix J. The Regulatory Guide does not prescribe a specific leak rate for a main steam line isolation valve. It does provide directions for the testing intervals for these valves. The proposed Technical Specification change is consistent with the Regulatory Guide.

8 Letter VYNPC to USNRC, FVY 81 8," Submittal of Information on NUREG 0737, Item Ill D.3.4: Control Room Habitability," dated January 12,1981.

• Letter USNRC to VYNPC, NVY 82-22," Safety Evaluation Report for IILD.3.4 Control Room Habitability Requirements (NUREG-0737)," dated Febnary 24,1982.

8 Vermont Yankee Final Safety Analysis Report, Revision 15, Section 14.9.1.5.

• ASME/ ANSI OMa-1988 Addenda to ASME/ ANSI OM 1987," Operation and Maintenance of Nuclear Power -

Plants."

"""" Y^** NN^" I'"wn CouronAms

,BVY 99-68 / Attachment I / Paga 3

2. The present testing requirements of Option B of 10CFR50 Appendix J, contained in the PCLRTP,

' nd based on the guidelines of NE! 94-01, Revision 0, continue to provide reasonable assurance that a

the leakage through the primary containment and components penetrating the primary containment will not exceed the allowable leakage rates in Technical Specification 6.15 and that the integrity of the containment structure is maintained during its service life. Neither Option B of 10CFR50 Appendix J nor NEI 94-01, Revision 0, requires a specific leak rate for the main steam line isolation valves.

3. Main steam line isolation valve leakage is a component of the Technical Specification 6.15 combined local leak rate acceptance criteria and, as such, will continue to be governed by that specification. The specific main steam line isolation valve leak rate criteria will be relocated to the Primary Containment Leak Rate Testing Program Plan (PCLRTP), which implements Technical Specification 6.15.

4. Prior to Facility Operating License Amendment 152, the main steam line isolation valve maximum pathway leak rate test results were added to Types B and C tests summation for evaluation compared to the 0.6La maximum pathway acceptance criteria. With the implementation of Amendment 152,  ;

the main steam line isolation valve maximum pathway leak rate test results are added to Types B and )

C tests summation for evaluation and for comparison to the 0.6La maximum and minimum combined leak rate acceptance criteria. The relocation of the specific leak rate acceptance criteria for the main steam line isolation valves does not change the calculated leak rate results that are compared to the 0.6La maximum and minimum combined leak rate acceptance criteria.

5. The radiological consequences of the design basis loss of coolant accident are dependent upon containment leakage rates. The leakage rate limitations assumed in the safety analyses are not impacted by this change.

e

'* vel 4MONT YANNI E NL'CI. EAR POWEl4 COHl'OHATION

., Docket No. 50 271 BVY 99-68 Attachment 2 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 215 Removal of Main Steam Line Isolation Valve Leakage Specifications Determination of No Significant Hazards Consideration

VruuoNT hwun Nrcumn Powen Coneonmos -

,BVY 99-68 / Attachment 2 / Page i Pursuant to 10CFR50.92, Vermont Yankee (VY) has reviewed the proposed change and concludes that the change does not involve a significant hazards consideration since the proposed change satisfies the criteria in 10CFR50.92(c).

1. The coeration of Vermont Yankee Nuclear Power Station in accordance with the oronosed amendment will not involve a significant increase in the orobability or consecuences of an accident oreviously evaluated.

l The proposed change does not involve a change to the plant design or operation. As a result, the proposed change does not affect any of the parameters 'or conditions that contribute to the initiation of any accidents previously evaluated. Thus, the proposed change cannot increase the probability of any accident previously evaluated.

The proposed change does not affect the leak tight integrity of the containment structure that is designed to mitigate the consequences of a loss-of-coolant accident (LOCA). The primary containment must maintain functional integrity during and following the peak transient pressures and temperatures that result from any LOCA, thereby limiting fission product leakage following the accident. Because the proposed change does not alter any of the fission produit leak rate assumptions used in the design basis LOCA analysis, the analyzed consequences of the Loss of Coolant Accident are not changed.

Based on the above VY has concluded that the proposed change will not result in a significant increase in the probability or consequences of any accident previously evaluated.

2. The operation of Vermont Yankee Nuclear Power Station in accordance with the oroposed amendment will not create the possibility of a new or different kind of accident from any i accident oreviously evaluated.

The proposed change does not involve a change to the plant design or operation. As a result, the proposed change does not affect any prameters or conditions that muld contribute to the initiation of any accident. This change eliminates redundant acceptance criteria from the Technical Specifications.

The methods of performing the tests are not changed. No new accident modes are created by the removal of the acceptance criteria. No safety-related equipment or safety functions are altered as a result of this change. The removal of the acceptance criteria has no influence over nor does it contribute to, the possibility of a new or different kind of accident or malfunction from those previously evaluated.

Based on the above VY has concluded that the proposed change will not create the possibility of a new or different kind of accident from those previously evaluated.

3. The operation of Vermont Yankee Nuclear Power Station in accordance with the orooosed amendment will not involve a significant reduction in a margin of safety.

The removal of the acceptance criteria does not impact the margin of safety. The 0.6La maximum and minimum pathway leak rate acceptance criteria provide the previously analyzed margin of safety. The testing method for detemiining the leak-tightness of the main steam line isolation valves has not changed. The leak rate test results are presently added to the Types B and C tests summation. The 0.6La maximum and minimum pathway leak rate acceptance criteria and the programmatic Corrective Action Limits provide assurance that component degradation does not impact the assumptions used to determine, nor provide a reduction in, the analyzed margin of safety.

. VrnuoNT nskun Necuan Powtu Connmxuos BVY 99-68 / Attachment 2 / Page 2 '

Based on the above VY has concluded that the proposed change will not cause a significant reduction in a margin of safety.

tidEnfies On the basis of the above, VY has determined that operation of the facility in accordance with the-proposed change does not involve a significant hazards consideration as defined in 10CFR50.92(c), in that it: (1) does not involve a significant increase in the probability or consequences of an accident previously evaluated;(2) does not create the possibility of a new or different kind of accident from any accident previously evaluated; and (3) does not involve a significant reduction in a margin of safety.

In making this detennination, Vermont Yankee has also reviewed .the NRC examples of license amendments considered not likely to involve significant hazards considerations as provided in the final adoption of 10CFR50.92 published in the Endgral Renister. Volume 51, No. 44, dated March 6,1986.-

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Vr.unoNr YANsat Necu:Au Powat Colu>on mon Docket No. 50-271 BVY 99-68 Attachment 3 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 215 .

I Removal of Main Steam Line Isolation Valve Leakage Specifications Marked-up Version of the Current Technical Specifications q

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