ML20209E089

TS Change Request 269 for License DPR-16,changing Component Surveillance Frequencies to Indicate Frequency of Once Per Three Months

ML20209E089
Person / Time
Site:	Oyster Creek
Issue date:	07/07/1999
From:	Roche M GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20209E088	List: ML20209E082 ML20209E089 ML20209E095
References
NUDOCS 9907140144
Download: ML20209E089 (8)

Text

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GPU Nuclear, Inc.

Oyster Creek Nuclear Generating Station Facility License No. DPR.16 Technical Specification Change Request No. 269 Docket No. 50-219 l

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Applicant hereby submits a change to Appendix A Technical Specification pages 4.41,4.4-2, 4.4-3,4.5-3 and 4.5 12.

By:

Michael B. Roche i Vice President and Director Oyster Creek l

l Sworn and Subscribed to before me thisitiday of Qy. I W1 ,

Mu f Ai Notary Public ofNJ GERALDINE E.UivlN NOT g M 3LC 0F W AlWEY p m p TI1isooo 9907140144 990707 PDR ADOCK 05000219 P PDR

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l GPU Nuclear. inc.

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U.S. Route M South NUCLEAR Post Office Box 388 Forked River, NJ 087310388 Tel 609-9714000 i

July 7, 1999 1940-99-20187 The Honorable William J. Boehm Mayor of Lacey Township 818 West Lacey Road Forked River, NJ 08731

Dear Mayor Boehm:

Enclosed herewith is one copy of Technical Specification Change Request No. 269 for the j Oyster Creek Nuclear Generating Station Operating License, j This document was filed 'with the United States Nuclear Regulatory Commission on July 7, 1999.

Very truly yours, b

Michael B. Roche -

Vice President and Director Oyster Creek

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United States of America Nuclear Regulatory Commission In the Matter of )

) Docket No. 50-219 GPU Nuclear, Inc. )

Certificate of Service This is to certify that a copy of Technical Specification Change Request No. 269 for Oyster Creek Nuclear Generating Station Operating License, filed with the U.S. Nuclear Rcgulatory Commission on July 7,1999, has this day of July 7,1999 , been served on the Mayor of Lacey Township, Ocean County, New Jersey by deposit in the United States mail, addressed as follows:

l The Honorable William J. Boehm Mayor of Lacey Township 818 West Lacey Road Forked River,NJ 08731 l l

l By:

Michael B. Roche i Vice President and Director  ;

Oyster Creek i 1

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. 1940-99-20187 Page 1of5 Attachment 1 Oyster Creek Nuclear Generating Station Technical Specification Change Request No. 269 1.- Change Reauested GPU Nuclear requests that Appendix A Technical Specification (TS) pages 4.4-1,4.4-2, 4.5-3 and 4.5-12 be revised to reflect a surveillance frequency of once per three months for specifications 4.4.A.1, 4.4.A.2, 4.4.C.1, 4.4.D.1, 4.4.F (isolation valves only) and 4.5.F.5.a. In addition, pages 4.4-1 and 4.4-2 incorporate editorial format changes while page 4.4-3 haa been revised to accommodate the expanded text.

II. Discussion of Proposed Change TS 4.4.A.1 addresses the periodic testing requirements employed to verify the operability of Core Spray System pumps. Currently, TS 4.4.A.1 specifies a surveillance frequency of once/ month and, in addition, after major maintenance and prior to startup following a refueling outage.

TS 4.4.A.2 addresses the periodic testing requirements employed to verify the operability of Core Spray System motor operated valves. Currently, TS 4.4.A.2 specifies a surveillance frequency of once/ month.

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TS 4.4.C.1 addresses the periodic testing requirements employed to verify the operability of j Containment Cooling System pumps. Currently, TS 4.4.C.1 specifies a surveillance frequency of once/ month and, in addition, aRer major maintenance and prior to startup following a refueling outage.  ;

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TS 4.4.D.1 addresses the periodic testmg requirements employed to verify the operability of l Emergency Service Water System pumps. Curmntly, TS 4.4.D.1 specifies a surveillance l frequency of once/ month and, in addition, aRer major maintenance and prior to startup l following a refueling outage.

l TS 4.4.F.1 addresses the periodic testing requirements employed to verify the operability of j Fire Protection System pumps and isolation valves. Currently, TS 4.4.F.1 specifies a j surveillance frequency of once/ month and, in addition, after major maintenance and prior to startup following a refueling outage.

.TS 4.5.F.5.a addresses the periodic testing requirements employed to verify the operability of Pressure Suppression Chamber - Drywell Vacuum Breakers. Currently, TS 4.5.F.5.a specifies a surveillance frequency of once each month and following any release of energy which would tend to increase pressure to the suppression chamber.

Increasing the pump and valve test interval from once/ month to once/3 months, as described in Section I above, will reduce the risk of plant transients due to surveillance testing, reduce personnel radiation exposure and reduce equipment degradation.

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. 1940-99-20187 Page 2 of 5 Accordingly, GPU Nuclear is proposing to change TS 4.4.A.1,4.4.A.2,4.4.C.1,4.4.D.1 and 4.5.F.5.a. such that surveillance testing is performed once per t! 'e months. In addition, with respect to the isolation valve operability requirements of TS 4.4.F.1, GPU Nuclear is proposing to change TS 4.4.F such that isolation valve operability surveillance testing is performed once per thrm nonths. The existing pump operability surveillance requirement of once/ month in TS 4.4.F.1 shall not be changed.

The existing additional surveillance requirement to verify operability after major maintenance and prior to stanup following a refueling outage that applies to specifications 4.4.A.1,4.4.C.1,4.4.D.1 and 4.4.F will not be changed. With respect to TS 4.5.F.5.a, the additional surveillance requirement to perform an operability test following any release of energy which would tend to increase pressure to the suppression chamber will not be changed.

The capitalization of text and the repositioning of column headings, incorporated on pages 4.4-1 and 4.4-2 respectively, are editorial in nature and do not alter the content or meaning of the text and have no effect on safe plant operations.

III. Safety Assessment The purpose of the Core Spray System is to provide for the removal of the decay heat from the core following a postulated Loss-of-Coolant Accident (LOCA), so that fuel clad melting is prevented for the entire spectrum of postulated LOCAs. The Core Spray I System consists of two loops each containing two main pumps, two booster pumps, two sets of psrallel isolation valves inside and outside the drywell, a spray sparger, and i associated piping, instrumentation and controls. The Fire Protection System is connected to each of the two Core Spray System loops. The purpose of this connection is to provide a backup supply of cooling water to the spargers.

1 The Containment Spray and Emergency Service Water Systems comprise the Containment Ileat Removal Systems for the OCNGS. The Containment Heat Removal ,

Systems are designed to reduce containraent pressure and temperature following a Design l Basis LOCA by removing thermal energy from the containment atmosphere. These systems also serve to limit offsite doses by reducing the pressure differential between the I containment atmosphere and the external environment. The Containment Spray System consists of two redundant loops which deliver water from the suppression pool to the 4 spray headers in the drywell and torus. Each loop consists of two pumps in parallel, two heat exchangers in parallel, two drywell spray headers and a torus spray header. Cooling I water to the tube side of the Containment Spray System heat exchangers is delivered by  ;

the Emergency Service Water System from the ultimate heat sink. There are two Emergency Service Water System pumps per Containment Spray loop.

The primary containment pressure suppression system, in parc, consists of a torus-to-drywell vacuum relief system, which prevents suppression pool water from backing up into the drywell during various reactor coolant and suppression system condensation modes, and which limits negative pressure differentials on the drywell in conjunction with the torus vacuum relief system.

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. 1940-99-20187 Page 3 of 5 l

Surveillances are performed to verify the operability of systems and components, and to

-verify that variables are within specified limits. Currently, specifications 4.4.A.1, 4.4.A.2,4.4.C.1,4.4.D.1,4.4.F.1 and 4.5.F.5.a stipulate a surveillance frequency of once per month for verifying the operability of their respective components. TSCR No. 269 proposes to alter the pump operability surveillance requirements of specifications 4.4.A.1,4.4.C.1 and 4.4.D.1 such that they specify a surveillance frequency of once per three months and, in addition, alter the valve operability surveillance requirements of specifications 4.4.A.2,4.4.F and 4.5.F.5.a such that they specify a surveillance frequency of once per three months.

The proposed surveillance frequency is consistent with ASME Code Section XI, IWP-3400, IWV-3411 and IWP-3521 (1986 Edition with no addenda, the Code of record for Oyster Creek ) which requires in-service testing of the active function pumps and valves every three months as well as being consistent with BWR Standard Technical Specifications, NUREG-1433 " General Electric Plants, BWR/4", which requires surveillance testing of valves in the ECCS system once every ninety two days. The proposed change is also consistent with the recommendations of ~ NUREG-1482

" Guidelines for Inservice Testing at Nuclear Power Plants". The risk of a plant transient due to surveillance testing, personnel radiation exposure and equipment degradation will be reduced as a result of the proposed increase in the surveillance test interval. In addition, a review of the Oyster Creek Inservice Test surveillance test data for the last three years demonstrates that the pumps and valves are reliable. The proposed changes i are compatible with plant operadng experience and are consistent with the guidance of Generic Letter 93-05, "Line-Item Technical Specifications Improvements To Reduce Surveillance Requirements For Testing During Power Operation". Based on the above, the proposed change in surveillance fr:quency will not adversely affect system .

availability and reliability nor will it adversely affect nuclear safety or safe plant l operation. l

. i IV. Information Supporting a Finding of No Sienificant Hazards Consideration GPU Nuclear has concluded that the proposed change to extend the surveillance interval for verifying pump and valve operability, for the system components delineated above, from once per month to once per three months does not involve a Significant Hazards Consideration. In support of this determination, an evaluation of each of the three (3) standards set forth in 10 CFR 50.92 is provided below.

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. I940-99-20187 Page 4 of 5

1. The pronosed TS chance does not involve a sienificant increase in the orrbability or conseauences of an accident previously evaluated.

The proposed surveillance interval change does not alter the actual surveillance requirements, nor does it alter the limits and restrictions on plant operations. The reliability of systems and components relied upon to prevent or mitigate the consequences of accidents previously evaluated is not degraded by the proposed change to the surveillance interval. Assurance of system and equipment availability is maintained. The proposed change does not alter any system or equipment configuration.

Based on the above, the proposed change does not significantly increase the probability or consequences of an accident previously evaluated.

2. The proposed TS chance does not create the possibility of a new or differer,t kind of accident from any accident previously evaluated.

The proposed surveillance interval change does not alter the actual surveillance requirements, nor does it alter the limits and restrictions on plant operations.

Assurance of system and equipment availability is maintained. The proposed change does not alter any system or equipment configuration nor does it introduce {

any new mechanisms which could contribute to the creation of a new or different kind of accident than previously evaluated.

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3. The proposed TS chances do not involve a significant reduction in a marein of sr.fety.

The proposed change extends the surveillance interval for verifying the operability of the specified pumps and valves from once per month to once per three months. The proposed change does not alter the actual surveillance requirements, the limits and restrictions on plant operations nor the design, function or manner of operation of any structures, systems or components.

System availability and reliability are maintained. Accordingly, the proposed TS change does not involve a significant reduction in a margin of safety.

1 V. Information Sunnorting an Environmental Assessment An environmental assessment is not reanired for the proposed change since the proposed change conforms to the criteria for " actions eligible for categorical exclusion" as specified in 10 CFR 51.22(c)(9). The proposed change will have no impact on the environment. The proposed change does not involve a significant hazards consideration as discussed in the preceding section. The proposed change does not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite. In addition, the proposed change does not involve a significant increase in individual or cumulative occupational radiation exposure.

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. I940-99-20187 Page 5 of 5 l . VI. Conclusion The proposed changes have been reviewed in accordance with Section 6.5 of the Oyster Creek Technical Specifications and it has been cor.cluded there are no unreviewed safety l questions. As discussed above, using the standards in 10 CFR 50.92, GPU Nuclear has determined that there are no Significant Hazards Considerations involved with the proposed changes.

VIL IMPLEMENTATION It .is requested that the amendment authorizing this change become effective upon issuance.

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