ML20246L177

Conformance to Generic Ltr 83-28,Item 2.2.1-Equipment Classification for All Other Safety-Related Components:Nine Mile Point 1 & 2

ML20246L177
Person / Time
Site:	Nine Mile Point
Issue date:	02/28/1989
From:	Udy A EG&G IDAHO, INC.
To:	NRC
Shared Package
ML17056A304	List: ML20246L177
References
CON-FIN-D-6001, CON-IIT07-437-91, CON-IIT7-437-91 EGG-NTA-7275, GL-83-28, NUREG-1455, TAC-53692, NUDOCS 8909060275
Download: ML20246L177 (21)

Text

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ENCLOSURE A l F ,

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. ), February 1989 k

' TECHNICAL EVALUATION REPORT

f lI idaho l National CONFORMANCE TO GENERIC LETTER 83-28, ITEM 2.2.1--

EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-Eng/neering RELATED COMPONENTS: NINE MILE POINT-1/-2 Laboratory

1. "#E# Alan C. Udy by the U.S. 1 Depanment \

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

EQUIPMENT CLASSIFICATION.FOR ALL OTHER SAFETY-RELATED COMPONENTS:

NINE MILE POINT-1/-2 l

Docket Nos. 50-220/50-410 Alan C. Udy Published February 1989 Idaho National Engineering Laboratory EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 l h

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Prepared for the U.S. Nuclear Regulatory Commission i Washington, D.C. 20555 Under DOE Contract No. DE-AC07-76ID01570 ,

FIN No. D6001 )

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l ABSTRACT This EG&G Idaho, Inc., report provides a review of the submittals for Unit Nos. I and 2 of the Nine Mile Point Nuclear Station for conformance to Generic Letter 83-28. Item 2.2.1.

1 Docket Nos. 50-220/50-410 TAC No. 53692 ii i

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1 FOREWORD.

I This report is supplied as part of the program for evaluating )

licensee / applicant conformance to Generic Letter 83-28 " Required Actions 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 Engineering and System Technology, by EG&G Idaho, Inc., NRC Regulatory Technical Assistance Unit. l The U.S. Nuclear Regulatory Commission funded this work under the authorization B&R 20-19-30-11-3, FIN No. D6001.

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Docket Nos. 50-220/50-410 TAC No. 53692 iii

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CONTENTS

, AB S T RA C T . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . 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 Evaluation ................................................. 3 3.3 Conclusion ................................................. 4

4. ' ITEM 2.2.1.1 - IDENTIFICATION CRITERIA ........................... 5 4.I' Guideline ........ ......................................... 5 4.2 Evaluation ............................... ................. 5 4.3 Conclusion ................................................. 5 I

5. ITEM 2.2.1.2 - INFORMATION HANDLING SYSTEM ....................... 6 5.1 Guideline ....... .......................................... 6 5.2 Evaluation ................................................. 6 5.3 Conclusion .................................................. 7

6. ITEM 2.2.1.3 - USE OF EQUIPMENT CLASSIFICATION LISTING ........... 8 1

6.1 Guideline . ........................................... .... 8 6.2 Evaluation ........ .................................. ..... 8 6.3 Conclusion ............... ............. . . . . . . . . . . . . . . . . . . . 8-

7. ITEM 2.2.1.4 - MANAGEMENT CONTROLS .. ............ . . . . 9 7.1 Guideline ................ ................................. 9 7 -. 2 Evaluation . .. ...................................... . . 9 7.3 Conclusion ....... ........ ............. .......... ...... 9

8. ITEM 2.2.1.5 - DESIGN VERI FICATICM AND PROCUREMENT . . . . . . . . . . . . . . . 10 8.1 Guideline ......... ............................... . ..... 10 8.2 Evaluation ........................ .......................... 10 8.3 Conclusion ... .... .... ...... ..... .. ............ . ... 10

9. ITEM 2.2.1.6 "IMPORTANT-TO-SAFETY" COMPONENTS ............... .. 11 9.1~ Guideline .............................................. . 11

10. CONCLUSION .. . . . . . . . 12 4.. -: - :fw-:..c .

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

. EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS:

NINE MILE POINT-1/-2

• 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 manually by the operator about 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 attact. ment. 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.

Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (EDO), directed the NRC staff to investigate and report on the generic implications of these occurrences at Unit 1 of the Salen Nuclear Power Plant. The results of the staff's inquiry into the generic implications of the Salem incidents are reported in NUREG-1000,

" Generic Implications of the ATWS Events at the Salem Nuclear Power Plant." As a result of tnis investigation, the Commission (NRC) requestec (by Generic Lette- 83-28 dated July 8,19831 ) all licensees o? 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 Niagara Mohawk Power Corporation, the licensee for the Nine Mile Point Nuclear Station,.for Iter 2.2.1 of Generic Letter 83-28. The documents reviewed as a'part of this evaluation are listed in the References Section at the end

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2. REVIEW CONTENT AND FORMAT Item 2.2.1 of Generic Letter 83-28 requests the licensee to submit a detailed description of their programs for safety related equipment classification for staff review. Detailed supporting information should also be included in'the description, as indicated in the guideline section for each item within this report.

As previously indicated, each of'the six sub-items of Item 2.2.1 is l evaluated in a separate section in which the guideline is presented; an evaluation of the licensee's response is made; and conclusions about the programs of tne licensee for safety-related equipment classification are draun.

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. 3. ITEM 2.2.1 - PROGRAM 3.1 Guideline Licensees should confirm that-an equipment classification program is in place which will provide assurance that all safety-related components are designated as safety-related on plant documentation. The program should provide assurance that the equipment classification information handling system is used so that activities that may affect safety-related components are designated safety-related. With this use of the information handling system, personnel are aware that they are working on safety-related components and are directed to and are guided by safety-related procedures and constraints. Licensee responses that address the features of this program are evaluated in the remainder of this report.

3.2 Evaluation l The licensee for the Nine Mile Point Nuclear Station responded to these requirements with submittals dated November 8, 1983 2 (Unit 1), April 10, 19843 (Unit 2), December 20, 19834 (Unit 2), and Decembe- 31, 19855 (Unit 1). The responses for Unit 2 were revised and  !

consolidated on April 15, 1986.6 The responses for Unit I were revised and consolidated on October 17, '988.7 These submittals include information that describes the licensee's safety-related equipment classification program. In the review of the licensee's responses to this item, it was assumed that the information and documentation supperting this program is available for audit upon request.

The licensee's component classification program for Unit I consists of the controlled 0-list, the computer database component Q-list, and the controlled process and instrumentation diagrams (P&ID) that are specifically controlled as color-coded, 0-list coded drawin95-The licensee's component classification program for Ui.it 2 consists of l tne "Ecuipmert and Structure Classification List" (Sec*. ion 3.2 of the 3 l l

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,, Final. Safety; Analysis Report)~. This list was developed by the architect-engineer and is being maintained by the lead modification engineer.

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

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

4.2 Evaluation The licensee states that the Unit 1 Q-list system was developed and verified in 1985. The licensee refers to ANSI /ANS-52.1-1983, " Nuclear Safety Criteria for the Design of Stationary Boiling Water Reactor Plants." Regulatory Guide 1.26, Re/ision 3, is also used in quality group classification. Engineering procedure EP 020 was used in the original determination of whether a component is-stfety-related. Engineering Procedure EP 190 is used to initiate identification of the safety-related or nonsafety-related status of components or ; arts where no such j classification existed previously.

Reference 6 gives the criteria used for identifying safety-related equipment and components for Unit 2. A component is considered safety-related if it is required to assure: (a) the integrity of the reactor' coolant system pressurc boundary, (b) the capability to achieve and to'maintair a safe shutdown, or (c) the capability to prevent or to mitigate the consequences of an accident which would result in potential offsite exposures.

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4.3 Conclusion The licensee's responses to this item are considered to be complete.

Therefore, the licensee's responses for this item are acceptable.

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5. ITEM 2.2.1.2 - INFORMATION HANDLING SYSTEM e

5.1 Guideline The licensee should confirm that the program for equipment classification includes an information handling system that is used to

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identify safety-related components. The response should confirm that this information handling system includes a list of safety-related equipment and that procedures exist to govern its development and validation.

5.2 Evaluation The licensee described how the Q-11st was originally prepared in accordance with Engineering Procedure EP 20 for Unit 1. The Q-list for Unit 1 is included as an appendix to that procedure. Engineering Procedure EP 190 provides the means for entering new safety-related items into the Q-list or for changing the classification of existing entries of the Q-list. Procedure NT-110.B is the document that controls the Q-list, the Q-list database, and the Q-list color-coded drawings; a change in any of these requires the signature of the Manager of Licensing. The approved change is logged, filed, and distributed to all users by the Nuclear Engineering Document Control Department under Procedure NE2-014 The Q-list database is changed only with t'he approval of the lead engineer for safety analysis. The Nuclear Staff Services Department enters'the change in the database, and each change is verified by a licensing engineer. The Records Management Department microfilms each change in the database for a permanent record. Procedure NT-110.0 controls changes to the Q-list color-coded drawings. The lead engineer for safety analysis must approve

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of these changes and is also responsible for distributing the updated cra vv i ng s .

The Q-list is part of the Unit 2 FSAR, Table 3.2-1. This list is being incorporated into the Master Equipment List, an on-line computer

-database. The licensee': rub 'tta's de c dbed the deve'oprent of the Master Ecuip nent List. The lead modification engineer is responsible for I

validatier, verification, anc updating the database. Information from 6

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.any database user, when shown to require update of the database, will be

. verified before the new information is entered into the database. The licensee described the continuing effort to validate the Master Equipment List, and how the users have access to the database.

5.3 Conclusion The licensee's responses are considered to be complete. Therefore, the licensee's responses for this item are acceptable.

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6. ITEM 2.2.1.3 - USE OF EQUIPMENT CLASSIFICATION LISTING 6.1- Guideline The licensee's description should confirm that the program for i equipment classification includes criteria and procedures that govern how ,

station personnel use the equipment classification information handling system to determine that an activity is safety-related. The description should-also include the procedures for maintenance, surveillance, parts replacement,.and other activities defined in the introduction to 10 CFR 50, l

Appendix B, that apply to safety-related components.

6.2 Evaluation The licensee's responses describe how procedures call for the use of the Q-list to determine when an activity is safety-related. The licensee

. states that they have procedural checks and balances that use the Q-list to preclude non-safety-related procedures being applied to safety-related equipment. Site Administrative Procedures are being revised to identify the Q-list as the sole source of safety classification information and to assure that adequate evaluations of proposed setpoint changes have been made. The Quality. Assurance Program is responsible for assuring that the correct procedures are used for Unit I safety-related activities. For Unit 2, the shif t supervisor is the final administrative control. The shift supervisor ensures that the correct procedures are used, that the Q-list is consulted, and that all procedural controls have been followed.

6.3 Conclusion We find that the licensee's description of plant administrative controls and procedures meets the requirements of this item. Therefore, the licensee's responses for this item are acceptable. ,

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. 7. ITEM 2.2.1.4 - MANAGEMENT CONTROLS 7.1 Guideline l'

The licensee should briefly describe the management controls that are

$ used to verify that the' procedures for preparation, validation, and the routine use of~the information handling system have been, and are being,

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followed.

7.2 Evaluation L The licensee's responses state that their Quality Assurance Program serves as the method of managerial control. Audits and inspections,_on both scheduled and unscheduled bases, are used to verify' the preparation, the validation, and the routine use of the information handling system and to assure that safety-related activities and their implementation are

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7.3 Conclusion l

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

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8. ITEM 2.2.1.5 - DESIGN VERfFfCATf0N AND PROCUREMENT 8.1 Guideline The licensee's submittals should document that past usage demonstrates l 4

that appropriate design verification and qualification testing are i

. specified for the procurement of safety-related components and parts. The , g i specification should include qualification testing for the expected ))

l safety-service conditions and should provid2 support for the licensee's j receipt of testing documentation to support the limits of life recommended by the supplier. If such documentation is not available, confirmation that l the present program meets these requirements should be provided. 1 l

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l 8.2 Evaluation The licensee states that engineering procedures for Unit 1, such as l EP 90, " Controlled Design and Design verification," and EP 100, " Control of l Procurement Activities," contain requirements concerning design verification and qualification testing. The licensee also describes Procedure NEL-015.M. " Procurement Requirements Evaluation and Dedication Planning"; ND-180, " Spare Part Equivalency Evaluations," the procurement requirements evaluation form, and the purchase order.

For Unit 2, the licensee refers to project procedures, which have requirements for design verification and equipment qualification and review and approval of the test documentation that are similar to the Unit I requirements. The licensee, in Reference 6, states that the Unit 1 procecures will be updated to include Unit 2 for this item.

8.3 Conclusion We conclude that the licensee has addressed the concerns of this item. Therefore, the licensee's responses for this item are acceptable.

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L 9. ITEM 2.2.1.6 "IMPORTANT-TO-SAFETY" COMPONENTS l 9.1 Guideline 1

Generic Letter 83-28 states that the licensee'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 to furnish this information as part of their response, this item will not be reviewed.

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10. 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 these concerns meets the requirements of Generic Letter 83-28 and is acceptable. Item 2.2.1.6 was not reviewed, as noted in Section 9.1.

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11. REFERENCES L 1. Letter, NRC (D. G. Eisenhut) to'all Licenseer. 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. Letter,' Niagara Mohawk Power Corporation (T. E. Lcapges) to NRC, 1- November 8, 1983 (Unit 1). ,

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3. Letter, Niagara Mohawk Power Corporation (G. K. Rhode) to NRC (A. Schwencer), April 10, 1984, NMP2L 0029 (Unit 2).

4. Letter, Niagara Mohawk Power Corporation (T. E. Lempges) to NRC (E. G. Adensam), December 20, 1985, NMP2L 0566 (Unit 2).

5. Letter, Nisgara Mohawk Power Corporation (C. V. Mangan) to NRC, December 31, 1985, NMP1 L0008 (Unit 1).

6. Letter, Niagare Mohawk Power Corporation (T. E. Lempges) to NRC (E. G. Adensam), April 15, 1986, NMP2L 0687 (Unit 2).

7. Letter,. Niagara Mohawk Power Corporation (C. D. Terry) to NRC, October 17, 1988, NMPIL 0314 (Unit 1).

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

EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS: NINE MILE POINT-1/-2

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Alan'C. Udy * ** " "*:" a s o February 1989 -

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EG8G Idaho. Inc.

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Idaho Falls ID 83415 D6001

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Division of Engineering and System Technology Technical Evaluation Report Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission ' " " ' * * ' ""' *"'

Washington, DC 20555 2 .. a ...... t..

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This EG&G 16ho, Inc., report provides a review of the submittals from the Nine Mile Point Nuclear St.ition regarding conformance to Generic Letter 83-28, Item 2.2.I.

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, ENCLOSURE 3 o,

SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE F/CILITY NAME Nine Mile Point, Units 1&2

SUMMARY

OF REVIEW / INSPECTION ACTIVITIES

' The SICB completed its review of the Item 2.2.1 of Generic Letter 83-28 submitted by the Niagara Mohowk Power Corporation for Nine Mile Point, Units 1&2.: The staff finds the licensee's responses for this item to be acceptable.

NARRAT:YF DISCUSS 10A 0F LICENSEE PERFORMAliCE - FUNCTIONAL AREA The responses to Generic Letter 83-28,' Item 2.2.1 submitted by the licensee.

required additional information to responc to staff concerns. All concerns were resolved irc 6 satisO . tory manner.

Author: L. Tran DATE: S/21/89 i

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