ML20215L863

Final Conformance to Generic Ltr 83-28,Item 2.2.1 - Equipment Classification for All Other Safety-Related Components,Brunswick 1 & 2, Informal Rept

ML20215L863
Person / Time
Site:	Brunswick
Issue date:	04/30/1987
From:	Vanderbeek R EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC
Shared Package
ML20215L815	List: ML20215L811 ML20215L863
References
CON-FIN-D-6001 EGG-NTA-7195, GL-83-28, TAC-53657, TAC-53658, NUDOCS 8706260196
Download: ML20215L863 (21)

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' DISCLAIMER This book was prepared as an account of work sponsored by an agency of the United

' States Government. Neither the United States Govemment nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product or process disclosed, or repre 'ents that its use would

. not infringe privately owned rights.' References herein to any specific commercial

. product, process; or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring -

by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof, i

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.E GG-NTA-7195-30 y-T.-

TECHNICAL EVALUATION REPORT-CONFORMANCE TO GENERIC LETTER 83-28 ITEM 2.2.1--

EQUIPMENT CLASSIFICATION FOR ALL'OTHER SAFETY-RELATED COMPONENTS:

BRUNSWICK-l'AND -2 Docket Nos. 50-325 and 50-324 R. VanderBeek Published April 1987 Idaho National Engineering Laboratory EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 S'

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

20555 Under DOE Contract No. DE-AC07-761001570 FIN No. 06001

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ABSTRACT

This EG&G Idaho, Inc., report provides a review of the submittal from Brunswick Steam Electric Plant Unit Nos. 1 and 2 for conformance to Generic Letter 83-28, Item 2.2.1.

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- Docket Nos. 50-325 and 50-324 l

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TAC Nos. 53657 and 53658 11

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i FOREWORD This report is supplied as partlof the program'for evaluating-licensee / applicant conformance to Generic. Letter 83-28 " Required-Actions 4

Based on Generic Implications.of Salem ATWS Events." This work'is being

- conductedLfor the U.S. Nuclear Regt;latory Commission, Office'of Nuclear

. Reactor Regulation, Division of PWR Licensing-A, by EG&G Idaho, Inc.

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The U.S. Nuclear Regulatory Commission funded this work under the

- authorization B&R 20-19-10-11-3, FIN No. 6001.'

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Docket Nos. 50-325 and 50-324 l

TAC Nos. 53657 and 53658 i

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CONTENTS l

ABSTRACT..............................................................

ii-4 FOREWORD..............................................................

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1. -

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

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2.

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

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3.

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

3 3.1.

' Guideline..................................................

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1 3.2' Eva1uation'.................................................

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.s 3.' 3 Conclusion..................................................

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

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4.1 Guide 1.*ne~.......................

5.

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

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4;3 Conclusion.................................................

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5.

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

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

6 5.2 Evaluation.................................................

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

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1 6.

ITEM 2.2.1.3 - USE_OF' EQUIPMENT CLASSIFICATION LISTING...........

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

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6.2 Evaluation.................................................

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1 6.3 Conclusion.................................................

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

9 7.1 Guideline...................................................

9 7.2 Evaluation.................................................

9 7.3 Conclusion..................................................

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ITEM 2.2;1.5 - DESIGN VERIFICATION'AND PROCUREMENT...............

10 8.1'

. Guideline...................................................

10 8.2 Evaluation.................................................

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8.3 Conclusion...........~......................................-

10-9..

ITEM 2.2.1.6

"IMPORTANT TO SAF ETY" COMPONENTS..................

11 9.1

. Guideline..................................................

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

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' 11. REFERENCES.......................................................

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l 2CONFORMANCE TO GENERIC LETTER 83-28 ITEM 2.2.1--

EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS:

8RUNSWICK-1 AND -2 L

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'1.

INTRODUCTION 1

1 On February 25, 1983, both of'the scram circuit breakers'at Unit 1 of

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the Salem Nuclear' Power Plant failed to open upon an automatic reactor trip-

. signal from the reactor protection system.

This incident was terminated manually.by the operator about 30 seconds.after the initiation of the l

automatic trip signal. The failure of;the circuit breakers was determined t'o be related to the sticking of the undervoltage trip attachment.

Prior j

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to this incident, on February 22, 1983, at Unit.1 of. the Salem Nui. lear Power Plant, an' automatic' trip signal was generated' based on stea'n generator low-low' level during plant start-up. 'In this case ~,'the reactor ll was tripped manually by'the operator almost coincidentally with the automatic trip.

j following these incidents, on February-28, 1983, the NRC Executive Director for Op'erations (EDO), directed the staff to investigate and report on the gener.ic implications of these occurrences at Unit 1 of the Salem q

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,.

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" Generic Implications of the ATWS Events at the Salem Nuclear: Power i

l 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 j

permits to respond to generic issues raised by the analyses of these two-I 1

ATWS events, j

l This report is an evaluation of the response submitted by Carolina Power and Light Company for Brunswick Steam Electric Plant, Unit-Nos. l'

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and 2 for Item 2.2.1 of Generic Letter 83-28.

The actual document reviewed as part.of this evaluation is listed in the references at the end of this report.

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REVIEW CONTENT AND FORMAT

'l Item 2.2.1 of Generic Letter 83-28 requests the licensee / applicant to l

submit,_for staff review, a description of their programs for r

classification of their_ safety-related equipment that includes supporting information, in considerable detail, as indicated in the guideline preceding the evaluation of each-item.-

'As previously stated, each of the six items of Item 2.2.1 is evaluated in a separate section in which the guideline is presented; an evaluation of the licensee's/ applicant's response is made; and conclusions about its

~ acceptability are drawn, i

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ITEM 2.2.1 - PROGRAM 3.1 Guideline Licensees and' applicants s'hould confirm that an equipment classification program ~ exists which provides' assurance that all safety-related components are designated as' safety-related on all plant-documents, drawings and. procedures and in the:information handling system that is used.in accomplishing safety-related activities, such as work orders for repair', maintenance and surveillance testing and orders for.

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replacement parts.. Licensee and applicant responses which address the I

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

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'3.2 Evaluation I

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.The licensee for Brunswick Steam Electric Plant, Unit Nos. 1 and 2 provided a response to Generic Letter 83-28 on' November 7, 1983.2 This l

submittal included information that describes their safety-relat.ed

- equipment classification program.

In the review of the lice.;ee's response to this item, it was assumed that the information and documentation f

supporting this program is available for audit upon request.

The-licensee has provided a description of the equipment classification program and. controls associated with the identification of

' safety-related activities for repair, maintenance, and procurement.

However, the response does not directly confirm that all components designated as safety-related in the Q-list are also properly designated on plant documents, procedures and in the information handling systems used i

for safety-related activities.

However, the licensee's response to Item 2.2.1.2 and 2.2.1.3 indicate that the documents used to control safety-related activities from start to finish are marked as

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a safety-related.

This is discussed in. Sections 5.2 and'6.2 of this report.

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We consider this to be acceptable.

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Conclusion:

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

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~thetlicenseefs~ response is adequate.

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ITEML2.2.1.1 - IDENTIFICATION CRITERIA 4.1 Guideline The' criteria for. identifying components-as safety-related should be presented; This should include description of means for handling

-*1 sub-components or parts as well as procedures for initiating-the identification of components.as safety-related.or non-safety related-if.no previous classification' existed.

4.2 ' Evaluation

-Theulicensee's response provided', from Volume YI. Book 2, of the Plant j

Operating Manual, the criteria for determining.which plant items are safety-related.-_The' response also identified the criteria used where i

. clarification of.the actual' physical boundary of a' safety-related. component i

is. required.

4.3 Conclusion The licensee's response to this item is considered to.be' complete and I

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'is' acceptable, j

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ITEM 2.2.~1.2'- INFORMATION' HANDLING' SYSTEM 5.1 Guideline The.11:ensee.or applicant should confirm that the program for equipment classification-includes an information handling system that-is used to idantify 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 validation.

5.2 ' Evaluation i

.The licensee's response infers that the Information Handling System is l

the procedure by'which.the Q-list is maintained. The Q-list identifies the.

j portions of systems, the instruments and the special components of systems-E which are. safety related. -The Q-list is part of the plant operating

. manual, which is a controlled document and was developed under the guidance-of ANSI /ANS-52.1-1978'and the information contained within'the FSAR.

The licensee states that the'present safety classification is adequate until the new detailed computerized qualification classification. data base is,

I established.' The response indicates that a review of the Reactor Protection System found that proper Q-list' classification was being utilized.

5.3 conclusion u

.We-find 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;0F EQUIPMENT CLASSIFICATION LISTING 6.1 Guideline 1

1 The licensee's description should show how station personnel Use the

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. equipment classification information handling, system to determine:

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.(a) when'an activity is safety-related,-and (b) what procedures are to be "used for maintenance work, routine surveillance testing, accomplishment of-

design changes. and performance of special tests or studies. We should be able.to gain confidence.from our review that there will be no confusion Labout when activity is safety-related.

6.2 Evaluation

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The licensee's. response-provides a description of the plant procedures

. governing maintenance, modifications, and procurement. activities. These-I procedures provide.the criteria involving Q-list components'or systems before any' work begins. Maintenance procedure MP-14 requires that the-safety-related components and the specific maintenance instructions be identified on the. Work Request and Authorization.

Procedures MP-03,10, and 16 require maintenance instructions be written for maintenance on

'Q-List (safety-related) equipment except in specific-cases.

The response

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states tnat where no procedure exists for a Q-list related component, the Maintenance Engineer prepares written work. instructions for the job in

.accordance with maintenance procedures MP-03, MP-10, MP-12, HP-16, MP-26, and MP-30 (or other approved procedures).

Store procedure SK-01 is stated to provide the controls that ensure Q. listed materials are properly-Adentified, procured, received,' inspected and stored.

Plant modification

. activities are stated to be controlled by engineering procedure ENP-03.

This procedure requires that safety-related equipment be identified as Q-List and the requirements of the Corporate QA manual and other applicable C

documents be implemented in the modification procedures.

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

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t' (7.. ITEM 2.2.1.4 - MANAGEMENT CONTROLS 7.1 Guideline Managerial controls that'will be used by the. licensee to verify that the information handling system for. equipment classification has been l prepared according to the approved procedures, that-its' contents have been validated, that it is being maintained current, and that it is being used to determine equipment classification as intended shall be described. The description of these controls shall be in sufficient detail for the staff to' determine that they are in place and.are workable.

U L7.2' Evaluation The li'censee's response states.that the. plant operating manual provides: the controis on the procedures and instructions -af fecting.Q-list components and, as such,:the procedures.and instructions are controlled

-documents and must receive certain reviews and approvals before implementation or: revision ~ to verify their technical correctness and conformance to applicable requirements.

The basis for a change, a safety.

analysis,-QA review, two-party. technical and safety review and Plant Nuclear Safety Committee (PNSC) and plant general manager approvals are required for'new or revised procedures.

This is-part of the means by which the management assures itself that the information system is being controlled.

In addition periodic QA surveillance.and audits are stated to provide assurance to management of the correctness of the programs and f

implementation.

7.3 Conclusion l

7-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 8.1 Guideline The licensee's submittals shall show that the specifications for procurement of replacement safety-related components and parts require that verification of design capability and evidence of testing that qualifies 7

the components and parts for. service under-the expected conditions for the

-serviceLlife specified by the supplier.is included.

8.2 Evaluation q

The. licensee's response states that plant Q-list requisitions are-l

reviewed by the Quality Engineer / Specialist to verify.that the proper

.j specifications, technical requirements and quality' assurance. requirements-l

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.have been stipulated. These requisitions are then given a second review by

-a quality assurance group to ensure that Q-list parts are being' properly

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.-specified when ordered. The response states th'at Procedures NPCD-P-0066 and NPCD-P-0075 contain the guidelines used.for approval of requisitions

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for construction procured items.

Also that safety-related items must

' nclude certification requirements and provisions of 10 CFR Part'21 on the j

i purchase order when required.

In addition, the response states'that 1) a j

technical review, 2) a-safety analysis, 3) QA review, 4) an' independent j

nuclear safety review by the On-Site Nuclear Safety group, 5) a plant nuclear safety committee review, 6) general manager's approval and 7) the corporate nuclear safety review of the safety analysis.along with routine management reviews will assure that the appropriate procurement.

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' requirements are specified for Q-listed components associated with plant modifications.

8.3 Conclusion 1

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

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ITEM P.2.1.6' "IMPORTANT TO SAFETY" COMPONENTS-9.1 Guideline Generic Letter 83-28 states that the licensee's or applicant's

. equipment classification program should. include.(in addition.to the l

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safety-related components) a broader-class'of components designated as "Important to Safety." However, since the generic letter does not require the'11censee or applicant to furnish this information as part.of their.

response,. review of this item will not be performed.

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10. CONCLUSION.

4 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 Item 2.2.1 of Generic Letter 83-28 is'

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acceptable.

Item 2.2.1.6 was not reviewed ~as'noted in Section'9 of this r epor t'..

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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,

" Required Actions Based on Generic Implications of Salem ATWS Events'

.(Generic. Letter 83-28)," July 8, 1983.

2.

Carolina Power and Light Company letter, P. W. Howe to D. G. Eisenhut, NRC, November 7, 1983.

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'l see instauctions on tus atvensa a gg Ave SLANat 3 fif bt AN0 8utfif tt CONFORMANCE TO GENERIC LETTER 83-28, ITEM 2.2.1--

EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED i

• oA's apoar m*utro COMPONENTS:

BRUNSWICK-1'AND -2' l

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EG8G Idaho, Inc.

P. O. Box 1625 Idaho Falls, ID' 83415 D6001 10 SPONSORING ORGAN 14 A flON NAWS ANO.sAltsNO A00mtse thes4str to C.*A 11s. 7VPE OP MEPORT Division of PWR Licensing - A Office of Nuclear Reactor Regulation

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U. S. Nuclear Regulatory Commission Washington,.DC 20555 1159PPLEW4NT ARY NOTSS o A..r. Aci,m This EG&G Idaho, Inc., report provides a review cf_.the submittal from the Carolina Power and Light Company regarding conformance ?.o Generic Letter 83-28, Item 2.2.1 for the Brunswick Steam Electric Plant, Unit Nos.1 and 2.

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