ML20245C256

Conformance to Generic Ltr 83-28,Item 2.2.1-Equipment Classification for All Other Safety-Related Components: Ginna, Final Informal Rept

ML20245C256
Person / Time
Site:	Ginna
Issue date:	07/31/1987
From:	Udy A EG&G IDAHO, INC.
To:	NRC
Shared Package
ML17250A934	List: ML20245C256
References
CON-FIN-D-6001 EGG-NTA-7241, GL-83-28, TAC-53675, NUDOCS 8708130123
Download: ML20245C256 (17)

Text

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TECHNICAL EVALUATION REPORT CONFORMANCE.TO GENERIC LETTER B3-28, ITEM 2'.2.1--

EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS: GINNA.

f Docket No. 50-244 Alan C. Udy l

Published July 1987 t ]:

Idaho National Engineering Lacoratory EG&G Icaho, Inc. .)

Idaho Falls, Idaho 83415' l

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Washington, D.C. 20555 1 Under DOE Contract No. DE-AC07-76ID01570 FIN No. D6001

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ABSTRACT This EG&G Idaho. Inc., report provides a review of the submittals from the R. E. Ginna Nuclear Power Plant for conformance to Generic Letter 83-28, Item 2.2.1.

i Docket No. 50-244 TAC No. 53675 ii

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FOREWORD

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1 This-report is suoplied as part of the program for evaluating licensee / applicant.conformance.to Generic Letter 83-28 " Required Actions Based on Generic Implicatic'ns of- Salem ATWS Events." :This work 11's being conducted for' the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of. Engineering and System' Technology,-by'EG&G Idaho, Inc., Electrical, Instrumentation and Control Systems Evaluation Group.

t The U.S. Nuclear Regulatory Commission funded this work under the authori:ation B&R 10-19-11-3, FIN No. D6001.

Docket No. 50-244 TAC No. 53675 iii

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E's. CONTENTS 4*'

l ABSTRACT.....'.......................................................... 11 FOREWORD ............................................................... iii

1. INTRODUCTION ....................................................,. I-

.. 2. REVIEW CONTENT AND FORMAT ........................................ :2

3. ITEM 2.2.1 - PROGRAM ............................................. 3-3.1 Guideline ......... ...................................;.... 3-3.2 3.3 Evaluation .. ..............................................

~ Conclusion .

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4. ITEM 2.2.1.1 - IDENTIFICATION CRITERIA ........................... 4 4.1 2

' Guideline ................................................... 4 4.2 Evaluation ................................................. 4 4.3 ' conclusion ................................................. 14

5. . ITEM 2.2.1. 2 -' INFORMATION HANDLING SYSTEM . . . . . . . . . . . . . . . . . . . . . . . 5 u5.I' Guideline .................................................. 5 5.2 Evaluation ................................................. 5

,5.3 Conclusion .................................................. 5

6. ITEM 2.2.1.3 - USE OF EQUIPMENT CLASSIFICATION LISTING . . . .. . . . . . . 6 6 .1 ' Guideline .'....... .......................................... 6 6.2 Evaluation .. ............................... .............. 6 6.3 Conclusion ................................................. 7-

'7; ITEM 2.2.1.4 - MANAGEMENT. CONTROLS ............................ .. 8 7.1 Guideline .................................................. 8 7.2 ' Evaluation ............................................. ... 8 7.3 Conclusion ............................... ................. 8

8. ITEM 2.2.1.5 - DESIGN VERIFICATION AND PR0CUREMENT'............... 9 8.1  : Guideline .................................................. 9 8.2 Evaluation ................................................. 9 8.3 Conclusion ................................................. 9

9. ITEM 2.2.1.6' "IMPORTANT TO SAFETY" COMPONENTS .................. 10 9.1 Guideline .................................................. 10

10. CONCLUSION ....................................................... 1*

11. REFERENCES.................................. ... ............ .... 12 iv l

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.. CONFORMANCE TO GENERIC LETTER 83-28, ITEM 2.2.1--

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EOUIPMENT CLASSIFICATION FOR ALL OTHER

, ' SAFETY-RELATED COMPONENTS: 'GINNA-

1. INTRODUCTION On February 25,1983,-both of the scram circuit breakers at Unit 1 of-the Salem Nuclear Power Plant failed to open upon an automatic reactor trip

.- signal from the reactor protection system. This incident was terminated g - manually by the operator abcut 30 seconds after the' initiation of the automatic trip ~ signal'. The failure of the circuit breakers was determined to be related to the sticking of the undervoltage trip attachment. Prior.

to this incident, on February 22, 1983, at Unit 1 of the Salem Nuclear Power Plant, an automatic trip signal was generated based on. steam

. generator low-low level during plant startup. In this case, the reactor was tripped manually by the operator almost coincidentally with the ,

automatic trip. '

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Following these . incidents, on February 28,1983, the NRC Executive

- Director for Operatiens' (EDO), directed the NRC staff to investigate and report on the generic' implications of tnese occurrences at Unit'l of the Salem Nuclear. Power. Plint. The results of the ~ staff's inquiry into the generic: implications of the Salem unit incidents are reported in NUREG-1000, " Generic Implications of the ATWS Events at the Salem Nuclear Power Plant." As a result of this investigation, the Commission (NRC) 1 requested (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 the generic issues raised by the analyses'of these two ATWS events.

This report is an evaluation of the responses submitted by the Rochester Gas and Electric Corporation, the licensee for the R. E. Ginna Nuclear Power Plant, for Item 2.2.1 of Generic Letter 83-28. The documents reviewed as a part of this evaluation are listed in the references at the end of this report.

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. 2. REVIEW CONTENT AND FORMAT Item 2.2.1 of Generic Letter 83-28 requests the. licensee or applicant to submit, for the staff review, a description of their programs for safety-related equipment classification including supporting information,

'in considerable detail, as indicated in the guideline section for each sub-item within this report.

As previously indicated, each of the six sub-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 the programs'of the licensee or applicant for safety-related ecuipment classification are drawn. --

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T:. 3. 11 TEM 2.2.1 - PROGRAM L

3.1 Guideline L,

I L Licensees and applicants should confirm that an equipment g

! , classification program exists which provides assurance that all safety-related components are cesignated as safety-related on all plant l -Occuments, crawings ano procecures and in the information nandling system l-that is used in accomolishir.g safety-related activities. such as work orders for recair,. maintenance and surveillance testing and orders for replacement parts. . Licensee and applicant responses wnich address the features of this.

program are evaluated in the remainder of this report,

, 3.2 Evaluation The licensee for the R. E. Ginna Nuclear Power Plant responded to these requirements with submittals d'ated November 4, 1983,2 August 23, 19853  !

and June 4,: 1987. These submittals include information that describes their safety-relatec e;uioment classification program. In the review of the licensee's resconse to this itsm, it was assumed that the information and cccumentation support.ing inis program is available for audit upon request.

We.have reviewed this information and note the following.

The licensee states that the list of safety-related ecuioment (Appendix A to the station Quality Assurance Manual) will oe used. This is periodically updated and is a controlled cocument. The licensee has stated that'this listing of safety-related parts and components is the information nandling system referred to.

3.3 Conclusion We have reviewed the licensee's submittals and, in general, find that the licensee's response is adequate.

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p.  ? 4 ITEM 2.2.1~.'1 -' IDENTIFICATION. CRITERIA

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,4.1 Guideline ri LThe applicant or licensee should confirm.that their program used, for

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equipment classification includes criteria.used'for identifying components as' safety-related.

4.2 Evaluation TheLlicensee's response states that safety-related equipment is~

defined as' that' necessary to assure.

1. The integr.ity'of.the reactor coolant' system pressurz boun'dary, _ ,

2. The capability to shut down the reactor or to maintain it in a

. shutdown condition or w

3. LThe~ prevention.or mitigation of the consequence of accidents thatL could result in potential offsite exoosures.

4.3 Conclusion .

We find that tne criteria used in-the identification of safety-relate'd components meets the requirements of Item 2.2.-l.1 anc are acceptacle.

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5. ITEM.2.2.1.2 - INFORMATION HANDLING SYSTEM 5.1 Guideline The licensee or applicant should confirm that the program;for

. equipment classification includes an information handling system that is useo to identify safety-rblated components. The response should confirm that this information handling system includes allist of safety-related ecuipment and tnat;procecures exist which govern its development and valication.

5.2 Evaluation The licensee's submittals identify the information handling system as consisting of a manual' listing of safety-related ccmponents. The licensee's description states that its development and validation are  ;

controlled by a formal review and approval procedure. Changes to the J.

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listing (Appendix A of the station's Quality Assurance Manual) are prepared by the responsible engineer in tne Engineering Department, are : reviewed by Ouality Assurance,'and are approved by the Manager of the Nuclear Engineering Department.

5.3- Conclusion We find that the informatio'n contained in the' licensee's submittals is sufficient for us to conclude that the licensee's information handling system for equipment classification meets the guideline requirements.

Therefore, the information provided by the licensee'for this item is acceptable.

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6. ITEM 2.2.1.3 - USE'0F EQUIPMENT CLASSIFICATION LISTING  ;

l< 6.1 Guideline 1

-i The licensee's or applicant's description should confirm that their program.for equipment classification includes criteria and procedures whien 4

govern how station personnel use the equipment classification information handling system to determine that an activity is safety-related and what j

procedures for maintenance, surveillance, parts replacement and other activities defined in the introduction to 10 CFR 50, Appendix B, apply to safety-related components. l 6.2 Evaluation The licensee states that the Quality Assurance Manual and its .

appendices are the mechanism used in determining whether an activity, system, or component is safety-related. Procedure A-1603 imposes the use a of Appendix A in determining wnether an activity is safety-related or not.

This procedure is used in requesting maintenance work on safety-related equipment, maintenance work requests (MWR), and trouble reports. It also describes the routing and disposition of the MWR. The licensee also briefly cescribes the following procedures.

A-502 Plant procedure adherence requirements Plant procedura content and format requirements l

A-600 Procedures regarding enanges (temporary and permanent) to i administrative procedures A-52 Tracks the allowable out-of-service time for equipment A-801 Control of accepted material parts and components 6

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' A-S02-. Identification and marking of ~ accepted material parts and components

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i- A-401 Control of! procurement document's A-701 Receipt and acceptance'of' material.'and parts A-1501 Control of nonconforming items These crocedures give the guidance and criteria'n'ecessary.to determine the safety-related status of an activity and provide > control for the attivities involved.

6.3 Conclusion We find that the description of plant administrative controls and procedures' meets the requirements of this item and is, therefore, acceptable',

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i ITEM 2.2.1.4 - MANAGEMENT CONTROLS 7.1 Guideline

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The applicant or licensee should confirm that the' management controls i used'to verify that the procedures for preparation, validation and routine >

utilization of tne information handling system have'been followec.

7.2 Evaluation l

Tne licensee's response states that managerial controls are part of the plant administrative. procedures, requiring plant operations review, committee review and superintendent approval. The licensee's responses describe how managerial cont'rols (wnich'are audited semiannually)-are used to assure'that'the equipment classification information handling system has been properly prepared, that its contents have been validated, that it is being maintained current and that it is being used to determine equipment g classification as intended. Other controls are document reviews,' quality control surveillance, quality assurance audits, and by work in progress monitoring. Maintenance work orders and procedure data packages are' reviewed by.the Maintenance Manager to ensure that administrative controls are adhered to.

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7.3 Conclusion We find that the management controls used by tne licensee assure that the information handling system is maintained, is current and is used as intended. Therefore, the licensee's resconse for this item is acceptable.

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, 8. ITEM 2.2.1.5 DESIGN VERIFICATION AND PROCUREMENT

'8.1 Guiceline The applicant's or licensee's submittal should document that past usage demonstrates that appropriate design verification and qualification testing is soecified for the procurement of. safety-related components and parts. 'The specifications should include qualification testing for l 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 Evaluatien Reference 2 states that the present design control process provides for the required design verification, qualification testing and

- documentation to support the life limits.

8.3 Conclusion Although the licensee did not specify the design criteria applied to this item, we conclude that the licensee has addressed the concerns of this.

item and, therefore, we find the response for this item acceptaole. l

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

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Generic Letter 83-28 states that tne licensee's or applicant's

' equipment classification program 'should include (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 licensee or applicant to furnish this information as part of their response, review of tnis item will not ce performed; 4

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..' ' Bas'ed on our review of the licen'see's response-to'the' specific .

requirements of-Item 2.2.1;'we: find'that' the.in' formation provided by the -

licensee'to resolve the concerns'of.~ Items' 2.2.'1;1', 2.2.1.2, 2.2.1.3,

.2.2.1.4, and 2.2 1.5 meet the requirements of Generic Letter 83-28 and.is

.~ a cc eptabl e. . Item 2.2.1.6.was not reviewed as noted in Section 9'1-

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$ '11. REFERENCES i:

e 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. Rochester Gas and Electric Corporation letter, J. E.' Maier to Director of' Nuclear Reactor Regulation, NRC, " Generic Letter 83-28,"

r November 4, 1983.

- 3. . Rochester Gas and Electric Corporation letter, R. W. Kocer to Director of Nuclear Reactor Regulation, NRC, " Generic Letter 83-28,"

August 23, 1985.

L 4 Rochester Gas and Electric Corporation letter, R. W. Kober to C. Stahle, NRC, "NRC' Letter of March 27, 1987 from C. Stahle to

' R. Kober Request fer Additional Information on Generic Letter 83-28, EItem 2.2," June 4, 1987.

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EG&G Idaho. Inc.

P. O. Box 1625 ' a ** 'aa' a*** a Idaho Falls, ID 83415 D6001 l .. o..o. ... o. ... . . . .o. .. . ... .. 6... . o o. . , . , . c , . .. , , o, o. ,

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l This EG&G Idaho, Inc., report provides a review of the submittals from the Rochester Gas and Electric Company regarding conformance to Generic Letter 83-28, Item 2.2.1 for the R. E. Ginna Nuclear Generating Plant.

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