ML20151B467

Safety Evaluation Supporting Amend 34 to License NPF-38

ML20151B467
Person / Time
Site:	Waterford
Issue date:	03/30/1988
From:	Office of Nuclear Reactor Regulation
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ML20151B465	List: ML20151B463 ML20151B467
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NUDOCS 8804080251
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(,,g NUCLEAR REGULATORY COMMISSION 9

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

'N SUPPORTING AMENDMENT NO. 34 TO FACILITY OPERATING LICENSE NO. NPF-38 LOUISIANA p0KCR AND LIGHT COMPANY i

WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382

1.0 INTRODUCTION

Ry application dated December 10, 1987, Louisiana Power and Light Company (LP&L or the licensee) requested changes to the Technical Specifications (Appendix A to Facility Operatino License No. NPF-38) for Waterford Steam Electric Station, Unit 3.

The proposed charges would revise Technical Specification 3.5.2, "ECCS subsystens - Tavg Greater than 350'F" and Technical Specification 3.5.3, "ECCS Subsystems -

Tavg less than 350*F" by adding a note to the Applicability section of both Technical Specifications to indicate that two Emergency Core Cooling Systen (ECCS) subsystens are required to be operable when Reactor Coolant Systen (RCS) average terperature is equal to or greater than 500*F.

In addition, the proposed change would also revise the title of the Technical Specifications such that it conforms to typical nomenclature.

Py letter dated March 24, 1988, the licensee further modified the Basis section to eddress the above changes.

2.0 DISCUSSION The changes proposed by the licensee would revise Technical Specification 3.5.? and 3.5.3 such that a note would be added to the Mode 3 applicability statement that will require both ECCS subsystens to be operable any time the RCS averege temperature is equal to or greater than 500'F regardless of the pressurizer pressure.

Also, the licensee would change the title cf the Technical Specification subsections to reflect mode of operation rather than average coolant temperature.

3.0 EVALUATION Currently Technical Specification 3.5.2 requires two independent ECCS subsystems to be operable when the reac.or is in Fodes 1, 2, and 3; c

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however, the requirements of this Technical Specification in Mode 3 are applicable only if the pressurizer pressure is equal to or greater than i

1750 psia. Techical Specification 3.5.3 currently requires one ECCS subsystem to be operable if the reactor is in Medes 3 and 4 with a require-rent that the pressurizer pressure is less than 1750 psia in Mode 3.

The 8804000251 000330 I

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2-proposed change to both Technical Specificatiers are similar in that a note will be added to the Mode 3 applicability statement that will require both ECCS subsystens to be operable any tine the RCS average temperature is eoual to or greater than 500*F. The intent of these Specifications is to ensure there will be sufficient emergency core cooling capability available in the event of a loss of coolant accident (LOCA) coincident with a single failure that results in the loss of one ECCS subsystem. The Waterford 3 Cycle 2 safety analysis has shown that borated water from the High Pressure Safety Injection (HPSI) Syster is required to prevent the core from becoming critical during an uncontrolled RCS cooldown (i.e., a steam line break) from greater than 500*F. Therefore, the licensee must ensure that at least one train of the HPSI system is available to mitigate the consequences of a postulated steam line break accident initiated from an RCS average temperature of 500'F or greater. The proposed change will accomplish this by requiring two ECCS subsystems to be operable whenever the average RCS terperature is equal to or greater than 500'F. Therefore, even if one ECCS subsystem is essumed to fail, one train of HPSI will be available to inject borated water into the RCS during a steam line break.

The staff cnneludes that the proposed changes to Technical Specifications 3.5.2 and 3.5.3 constituto an additional restriction on plant operation to increase the margin of safety, and are, therefore, acceptable.

In addition to the above, the proposed change will also revise the title of Technical Specsifications 3.5.2 and 3.5.3.

The current title describes the Technical Specification in terms of average ecolant temperature.

It is standard practice to refer to plant conditions in tems of operating Moder rather than average coolant temperature. Therefore, the proposed change would revise the titles such that they confom to Technical Speci-fication nomenclature and are acceptable.

4.0 CONTACT WITH STATF 0FFICIAL The NRC staff has advised the Administrator, Nuclear Energy Division, Office of Environmental Affairs, State of Louisiana of the proposed determination of no significant hazards consideration. No coments were received.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment relates to changes in installation or use of a facility i

corponent located within the restricted area as defined in 10 CFR Part 20.

The staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or ceulative occupational radiation exposure. The Comission has previously issued a proposed finding that this amendrent i

involves no significant hazards consideration and there has been no public coment on such finding.

AccordinD y, the amendment meets the eli 1

criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9)gibility l

Pursuantto10CFR51.22(b),noenvironmentalimpactstatementorenviron-mental assessment need be prepared in ectnection with the issuance of this crendment.

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6.0 CONCLUSION

l Based upon its evaluation of the proposed changes to the Waterford 3 Technical Specifications, the staff has concluded that:

there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and such j

activities will be conducted in compliance with the Commission's regulations and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

The staff, therefore, concludes that the proposed changes are acceptable, and are hereby incorporated into the Waterford 3 Technical Specifications.

Dated: March 30,1938

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Principal Contributor:

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CONFORfMNCE TO GENERIC LETTER 83-28, ITEM 2.2.1--

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Eng/neer/ng EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-7.c,j RELATED COMPONENTS: WATERFORD-3

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l CISCLAIMER Th s boos wee preoered se en eccount of wort econsored by en egycy of the Umted states Govwnment. Nestnw the ureted States Get not any agency trereof, nor on, of tror empovees. meses any wwrenty, empress or umseed, or enownee any Wel Whty or reeooeht>tfy fCW t% SCCWecy, CompdeteneM, or geeNWees of any eformeton, soceratue, prodwet et procese Geooted. or represente thet ste use windd not efange smetery owned rgets o gronces herein to any spechc conewool o

oroovet, croceos. or saves ey trece name, trm, menuseetwor, or othennes.

ooes not necessenN constitwto or emoN its encoreement, recomrreneeton, or fevonne Dy the Urwted $tette GChemtrent or any egywy throof The woneg erel oportorW of owthors encreened herein 00 not wW eute or renect those of the Ursted $tene Govenment or any agency trereof-l

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l TECHNICAL EVALUATION REPORT CONFORMANCE TO GENEWIC LETTER 83-28. ITEM 2.2.1--

EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS:

WATERFORD-3 Docket No. 50-382 I

l R. VanderBeek 1

Published April 1987 i

Idaho National Engineering Laboratory EG6G Idaho. Inc.

Idaho Falls Idaho 83415 i

Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Under DOE Contract No. OE-AC07-761001570 l

i FIN No. 06001 1

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A8STRACT i

This EGM Idaho. Inc report provides a review of the submittals for the Waterford Steam Electric Station. Unit No. 3 for conformance to Generic Letter 83-28, Item 2.2.1.

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TAC No. 51705 J

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FOREWORD This report is supplied as part of the program for evaluating 4

licensee / applicant conformance to Generic Letter 83-28 ' Required Actions l

Based on Generic Implications of Salem ATWS Events.' This work is being conducted for the U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of PWR Licensing-A, by EG66 Idaho, Inc.

The U.S. Nuclear Regulatory Commission funded this work under the authorization 86R 20-19-10-11-3, f!N No. 06001.

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1 Docket No. 50-382 TAC No. 57705 i

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f CONTENTS 1

• ABSTRACT..............................................................

11 FOREWORD..............................................................

11) 1.

INTRODUCTION.....................................................

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REVIEW CONTENT AND FORMAT........................................

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ITEM 2.2.1 - PROGRAM............................*.................

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3.1 Guideline..................................................

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3.2 Evaluation.................................................

3 3.3 Conclusion.................................................

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ITEM 2.2.1.1 - IDENTIFICATION CRITERIA...........................

5 4.1 Guideline..................................................

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i 4.2 Evaluation.................................................

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0 4.3 Conclusion.................................................

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ITEM 2.2.1.2 - INFORMATION HANDLING SYSTEM.......................

6 5.1 Guideline..................................................

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5.2 Evaluation.................................................

6 5.3 Conclusion.................................................

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ITEM 2.2.1.3 - USE OF EQUIPMENT CLASSIFICATION LISTING...........

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6.1 Guideline..................................................

7 6.2 Evaluation.................................................

's 6.3 Conclusion.................................................

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ITEM 2.2.1.4 - MANAGEMENT CONTROLS...............................

8 7.1 Guideline

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8 7.2 Evaluat' ion.................................................

8 7.3 Conclusion.................................................

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ITEM 2.2.1.5 - DESIGN VERIFICATION AND PROCUREMENT...............

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8.1 Guideline

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4.2 Evaluation.................................................

9 8.3 Conclusion................................................

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ITEM 2.2.1.6 "IMPORTANT TO SAFETY' COMPONENTS..................

10 9.1 Guideline..................................................

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10. CONCLUSION.......................................................

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11. REFERENCES.......................................................

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l CONFORMANCE TO GENERIC LETTER 83-28. ITEM 2.2,1--

EQUIPMENT CL ASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS:

WATERFORD-3 1.

INTRODUCTION On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Sales Nuclear Power Plant failed to open upon an autos tic reactor trip j

signal from the reactor protection system.

This incident was terminated l

m nually by the operator about 30 seconds after the initiation of the autos tic trip signal.

The failure of the circuit breakers was determined l

to be related to the sticking of the undervoltage trip attachment.

Prior f

to this incident, on February 22, 1983,.t Unit 1 of the Salem Nuclear Power Plant, an autos tic trip signal was generated based on steam generator low-low level during plant startup.

In this case, the reactor was tripped u nually by the operator almost coincidentally with the autos tic trip.

Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (EU0), directed the staf f to investigate and report on the generic implications of these occurrences at Unit 1 of the Salem Nuclear Power Plant. The results of the staff's inquiry into the generic j

implications of the Salem unit incidents are reported in NUREG-1000,

' Generic Implications of the ATWS Events at the Salem Nuclear Power

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Plant.' As a result of this investigation, the Commission (NRC) requested I

(by Generic Letter 83-28 dated July 8, 1983 ) all licensees of operating reactors, applicants for an operating license, and holders of construction permits to respond to generic issues raised by the analyses of these two ATWS events.

i This report is an evaivation of the' responses submitted by Louisiana Power and Light for Waterford Steam Electric Station, Unit No. 3 for item 2.2.1 of Generic Letter 83-28.

The actual documents reviewed as a part of this evaluation are listed in the references at the end of this

report, i

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REVIEW CONTENT AND FORMAT

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i Item 2.2.1 of Generic Letter 83-28 requests the licensee / applicant. to j

i submit, for staff review, a description of their programs for i

j classification of their safety.related equipment includes supporting

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information, in considerable detail, es indicated in the guidelines j

j preceding the evaluation of each sub-item.

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As previously stated, each of the six sub-items of item 2.2.1 is i

j evaluated in a separate section in which the guideline is presented; an l

evaiestion of the licensee's/ applicant's response is made; and conclusions about its accept 4t. *>- Jre drawn, j

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IT[M 2.2.1 - PROGRAM 3

3.1 Guideline Licensees anti applicants should conftrm that an equipment classification program exists which provides assurance that all i

safety-related components are designated As safety-related on all plant l

l documents, drawings and procedures and in the information handling system that is used in accomplishing safety-related activities, such as work l

orders for repair, maintenance and surveillance testing ind oruers for l

replacement parts.

Licensee and applicant resoonses which address the l

features of this program are evaluated in the remainder of this report.

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3.2 Evaluation i

t The licensee for Waterford Steam Electric Station, U41! No. 3 provideo 2

a response to Generic Letter 83-28 with submittals dated Novembo., 1963 and November 15, 1985.3 These submittsls included infora tion that descrtbes their safety-related equipment classift ation program.

In the review of the licensee's response to this item, it was assumed tha; the information and documentation supporting this program is available for audit upon request.

The itcensee has provided a description of the equipment classification program for tne identification of safety-related activities I

for repair, maintenance, and procurament,

& wever, the response does not directly confirm that all components designated as safety-related in the MEL/Q-list are also properly designated on plar.1 documents, piecedures and in the information handling systems used for safety-related activities.

However,thelicenshe'sresponsetoItems2.2.1.2and2.1.1.3indicatethat the documents used to control safety-related activities from start to

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finish are approortately marked as safety-releted.

This is (,ascussed in Sections 5.2 and 6.2, We consider this to be acceptable, i

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We have reviewed the licensee's information and, in general, find that l

the licensee's response is adequate.

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ITEM 2.2.1.1 - IDENTIFICATION CRITERIA 4.1 Guideline The appitcant or licensee should confirm that their program used for equipment classification includes criteria used for identifying components as safety-related.

4.2 Evaluation The licensee's response states that safety-related structures, systems, and components are identified as safety-related based on the criteria specified in the project management procedure PMP-321 "Determination of Tafety/Q-Level Components for the MEL/Q-List".

The procedure was nct ir.cluded in the response; however, review of Section 3.2 of the FSAR identified these criteria.

t 4.3 Conclusion The licensee's response to this item is considered to be complete and is acceptable.

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ITEM 2.2.1.2 - INFORMATION HANDLING SYSTEM 5.1 Guideline The licensee or applicant should confirm that the program for equioment classification includes an information handling system that is used to identify safety-related components.

The response should confirm that this information handling system includes a list of safety-related equipment and that procedures exist which govern its development and l

validation.

5.2 Evaluation l

The licensee's response states that the Q-list is maintained current j

by a dedicated staff whose activities are governed by project management proce ure PMP-321.

This procedures is being updated to include requirements for Q-List maintenance activities.

The Q-List information for components in the plant is entered in the data base and validated in accordance with project management procedure PMP-320.

5.3 Conclusion The licensee's response to this item is considered to be complete and is acceptable.

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ITEM 2.2.1.3 - USE OF EQUIPMENT CLASSIFICATION LISTING 6.1 Guideline The licensee's or applicant's description should confirm that their program for equipment classification includes criteria and procedures which govern how station personnel use the equipment classification information handling system to determine that an activity is safety-related and what procedures for maintenance, surveillance, parts replacement and other activities defined in the introduction to 10 CFR 50, Appendix 8, apply to safety-related components.

6.2 Evaluation The licensee's response identifies the use of the Q-list, and Administrative procedures in the determination of safety-related activities in the areas of parts replacement, storage, maintenance, modification.

testing, and surveillance.

Collectively, these documents contain the controls to ensure that s4fety-related equipment is identified and handled in an appropriate manner.

6.3 Conclusion i

The licensee's response to this item is considered to be complete and is acceptable.

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ITEM 2.2.1.4 - MANAGEMENT CONTROLS 7.1 Guidelines The applicant or licensee should confirm that the management controls used to verify that the procedures for preparation, validation and routine utilization of the information handling system have been followed.

7.2 Evaluation The licensee's response states that the management controls established for activities related to the development, validation and maintenance of the Q-List are covered by procedures and instructions which are prepared, reviewed, and approved in accordance with project management procedure PMP-001, Preparation and Revision of Project Management Procedure / Instructions".

The management controls established for activities related to the routine utilization of the Q-List are governed by Administrative procedure UNT-1-002 and QP-5-001, ' Instructions, Procedures and Drawings."

7.3 Conclusion The licensee's response to this item is considered to be complete and is acceptable.

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ITEM 2.2.1.5 DESIGN VERIFICATION AND PROCUREMENT J

8.1 Guideline The applicant's or licensee's submittal should document that past usage demonstrates that appropriate design verification and qualification testing is specified for the procurement of safety-related components and parts.

The specifications should include qualification testing for expected safety service conditions and provide support for the applicant's/ licensee's receipt of testing documentation to support the limits of life recommended by the supplier.

If such documentation is not available, confirmation that the present program meets these requirements should be provided, 8.2 Evaluation

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The licensee's response states that specifications imposed upon the j

vendor are referenced on the Purchase Order Requisition based on either previous orders for the same equipment or specifications supplied by 1

Engineering.

Standard Clauses in UNT-8-001 are used to ensure that technical and quality requirements are specified consistently for safety and quality related equipment orders.

8.3 Conclusion The licensee's response for this item is considered to be complete and is acceptable.

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ITEM 2.2.1.6 "IMPORTANT TO SAFETY" COMPONENTS 9.1 Guideline The Generic Letter 83-28 states that the licensee's equipment classification program should incluc4 (in addition to the safety-related components) a broader class of components designated as "Important to Safety." However, since the Generic Letter does not require the applicant / licensee to furnish this information as part of their response, review of this item will not be performed.

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CONCLUSION Based on our review of the licensee's response to the specific requirements of Item 2.2.1, we find that the information provided by the licensee to resolve the concerns of Items 2.2.1 of Generic Letter 83-28 is acceptable.

Item 2.2.1.6 was not reviewed as noted in Section 9 of this report.

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REFERENCES 1.

NRC Letter, D. G. Eisenhut to all Licensees of Operating Reactors, Applicants for Operating License, and Holders of Construction Permits, l

"Required Actions Based on Generic Implication of Salem ATWS Events (Generic Letter 83-28),' July 8, 1983.

2.

Louisiana Power and Light letter, K. W. Cook to 0. G. Eisenhut, NRC, November 4,1983, W3P83-3911, 4-3-A20.02.02, 3-A1.01.04, L.02.

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Louisiana Power and Light letter, K. W. Cook to G. W. Knighton, NRC, November 15, 1985, W3P85-3158, A4.05,NQA.

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BISUOGRAPHIC DATA SHEET EGG-NTA-7446 sii..it.ver o~ o. v..svi.u j fif L4.NQ,93 7 16.

J kg.yl Sk.NE CONFORMANCE TO GENERIC LETTER 83-28, ITEM 2.2.1--

EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMP 0NENTS: WATERFORD-3

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This EG&G Idaho, Inc., report provides a review of the submittals from Louisiana Power and Light regarding conformance to Generic Letter 83-28, Item 2.2.1 for the Waterford Steam Electric Station, Unit No. 3.

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