ML20205N953

Conformance to Generic Ltr 83-28,Item 2.2.1 - Equipment Classification for All Other Safety-Related Components: Millstone-3

ML20205N953
Person / Time
Site:	Millstone, Haddam Neck, 05000000
Issue date:	01/31/1988
From:	Udy A IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC
Shared Package
ML20205N941	List: ML20205N936 ML20205N943 ML20205N948 ML20205N953 ML20205P551
References
CON-FIN-D-6001 EGG-NTA-7384, GL-83-28, TAC-60395, NUDOCS 8811070121
Download: ML20205N953 (18)

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EGG-NTA-7384 TECHNICAL EVALUATION REPORT T

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

ECUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS:

MILLSTONE-3 -

c Docket No. 50-423 Alan C. Udy Published January 1988 Idaho National Engineering Laboratory EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Comission Washington 0.C. 20555 Under DOE Contract No. OE-AC07-761001570

FIN No. 06001 B B //op ty " up

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4 ABSTRACT i

This EG&G Idaho, Inc., report provides a review of the submittals from Unit No. 3 of the Millstone Nuclear Power Station.for conformance to Generic Letter 83-28, Item 2.2.1.

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Docket No. 50-423 TAC No. 60395 11

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FOREWORD c

This report is suppited 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., Electrical, Instrumentation and Control Systems Evaluation Unit. ,

The U.S. Nuclear Regulatory Commission funded this work under the i authorization B&R 20-19-10-11-3, FIN No. 06001. L l

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Docket No. 50-423 i TAC No. 60395  ;

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O O-CONTENTS ABSTRACT .............................................................. 11 FOREWORD .............................................................. iii

1. INTRODUCTION ..................................................... I

2. REVIEW CONTENT AND FORMAT ........................;............... 2

3. ITEM 2.2.1 - PR03 RAM ............................................. 3 3.1 Guideline .................................................. 3 3.2 Evaluation ............................................ .... 3 3.3 Conclusion ........................................... ..... 4 4 ITEM 2.2.1.1 - IDENTIFICATION CRITERIA ........................... 5 4.1 Guideline ......................................... ........ 5 4.2 Evaluation ................................................. 5 4.3 Conclusion ................................................. 6 i 5. ITEM 2.2.1.2 - INFORMATION HANDLING SYSTEM ....................... 7

! 5.1 Guideline .................................................. 7 l 5.2 Evaluation ................................................. 7 5.3 Conclusion ................................................. 7

6. ITEM 2.2.1.3 - USE OF EQUIPMENT CLASSIFICATION LISTIN3 . . . . .. . . . . . 0 6.1 Guideline .................................................. 8 6.2 Evaluation ................................................. 8 l 6.3 Con:1usion ................................................. 8

7. ITEM 2.2.1.4 - MANAGEMENT CONTROLS ............................... 9 l

l 7.1 Guideline .................................................. 9 l 7.2 Evaluation ................................................. 9 l 7.3 Conclusion ................................................. 9

8. ITEM 2.2.1.5 e OESIGN VERIFICATION 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 .................. Il 9.1 Guideline ................................................... 11

10. CONCLUSION ....................................................... 12

11. REFERENCES ....................................................... 13 iv l

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

. EQUIPMENT CLASSIFICATION FOR ALL C"N CAFETY-RELATED COMpCNENTS:

MILLSTONE-3

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 I 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 stick og of the undervoltage trip attachment. Prior to this incident, en February 22, 1983, at Unit 1 of the Salem Nuclear Pcwer Plant, an autematic trip signal was generatec based on steam l generator lew-lew level during plant startup. In this case, the reactor was trippec manually by the operator almost coincidentally with the autCmatic 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 Salem Nuclear Power Plant. The resuits of the staff's incuiry 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 this investigation, the Commission (NRC) requested 1

(by Generic Letter 83-28 dated July 8, 1983 ) all licensees if o7erating reactors, applicants for an operating license, and holders of construction permits to respond to the generic issues raised by the analyses of these two ATuis events, j This report is an evaluation of the response submitted by the Northeast Utilities Company, the licensee for the Millstone Nuclear Power Station, for Item 2.2.1 of Generic Letter 83-28. The document reviewed as a part of this evaluation is listed in the references at the end of this report.

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2. REVIEU CONTENT AND FORMAT Item 2.2.1 of Generic Letter 83-28 requests the licensee / applicant submit, for staff review, a description of their programs for safety-related equipment classification including supporting information, i in considerable detail, as indicated in the guideline section for each item l within this report.

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  !

I the licensee's/ applicant's response is made; and conclu11ons about the i licensee's or applicant's program for safety-related eauipment ,

i classification are drawn.

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3. Item 2.2.1 - PROGRAM 3.1 Guideline Licansees and applicants should confirm that an equipment classification program exists that provides assurance that all safety-related components are cesignated as safety-related on plant '

documentation and in the information handling system that controls safety-related activities. The purpose of this program is to ensure that personnel performing activities that affect safety-related c0mponents are  !

aware that they are working on safety-related components and are guided by safety-related procedures and constraints. Features of this program sre evaluated in the remainder of this report.

3.2 Evaluation The licensee for Unit 3 of the Millstone Nuclear Power Station responded to these requirements with submittals datec November 8, 1983 2 and March 13, 1957.3 These submittals include information that describes the Millstone-3 safety-related equipment classification program. In the review of the licensee's response to this item, it was asi,umed that the information and documentation succorting this program is available for audit  !

uoon recuest.

The licensee's computer-cased Production Maintenance Management System  ;

(PMwS) and the Material, Ecutoment and Parts List (MEPL) make up the  ;

equipment classification program. Both the PMMS and the MEPL use Nuclear  !

Engineering and Operations Procedures (6.10 and 6.01) for control. Thus, the Nuclear Engineering and Operations Department has the responsibility for both the PMMS and the MEPL.

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Work orders are generated by the PM S and include a designation for safety-related work activities. The plant procedures for surveillance ,

testing, administrative control procedures, drawings, and purchase documents also designate safety-related equipment.

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

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

4.2 Evaluation The licensee's response provided tne criteria for the original and ongoing identification of systems, structures, and components as safety-related. The criteria consist of a listing safety-relate i systems, structures, and components, and includes issocations and supports. The listing consists of 19 items, most of which contain sub-items. The licensee states that these criteria are currently used to identify safety-related components in accordance with quality assurance i procedures.

Generic Letter 83-28 identifies as safety-related those structures, systems, and cc. oonents that as'ure (following a desig1 ccsis event)

(1) the integrity of tne reac*or coolant bouncary, (2) the caeability to snut down the reactor a9e to maintati it in a safe shutdown condition, and (3) the capacility to p.* event or to mitigate consequential offsite exposures.

The licensee defines (but the definition is not limited to) as i

safety-related those systems, structures, and components, including associated foundations, supports and auxiliary systems, that (1) are a portion of the reactor coolant prissure boundary, (2) are used for emergency core cooling, reactor shutdown, residual heat removal and cooling water systems that support the previous systems, and (3) reactivity control systems, primary to secondary containment, portions of radioactist waste control systems, and systems that might contain radioactive materials and 4 whose postulated failure could result in offsite exposures.

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4.3 Conclusion

• We find that the criteria used to identify saf3ty-related components encompasses the defit ition for infety-related that is part of the generic letter. Therefore tl e identification criteria meets the recuirements of Item 2.2.1.1.

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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 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 validation.

5.2 Evaluation l

The licensee's submittals identify the hard-copy MEPL as the information handling system that lists safety-related structures, systems, j co ponents, anc parts. The PMMS is a computerized data base that will i eventually replace the MEPL. Currently the two systems co-exist, and the FEPL is the governing document. The licensee's description of the systems included the methods used for the development of these systems, the process by which new safety-related items are entered, how changes in the classification of listed items are made, how listed items are verified, how unauthorized changes to the listing are prevented, and how the listing is maintained anc distributec to online computer terminal users. Revisions to both the MEPL and the FMMS are cortrolled by the MEPL engineer.

5.3 Conclusion We find that the information centained in the Itcensee'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 OF EQUIPMENT CLASSIFICATION LISTING ,

6.1 Guideline The licensee's or aplicant's description should confirm that the program for ecuipm*nt classification includes criteria and procedures that govern how station personnel use the eQuiement classification information .,

handlir.g 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, Appendix B, that apply to safety-related ccmponents. I 6.2 Evaluation Tne licensee states that cocument centrol during ccnstructicn has been nder the responsibility of the architect engineer, Stone & Webster. The -

licensee states that the arenitect engineer has procedural and management l l

l control to assure that vender information is reviewed. The licensee, upon receipt cf vendor information, maintains it in library files and distributes 't to the resoonsible personnel fer procedure and test preparation.

The licentee states that either the MEPL or the PMMS is consulted to ceter-nne the safety-related status of the abeve work activities. Work orcers control these activities. These work orders are generated by the PMS . The licensee has provided a listing of Actinistrative Control Procedures that address these activities. The Work Order designates a specific activity as safety-related or not and designates the procedures to be used in the above activities.

6.3 Conclusien We find that the licensee's description of plant actinistrative controls and procedures meets the requirements of this iten and is, therefore, acceot.ble.

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7.  ! TEM 2.2.1.4 - MANAGEMENT CONTROLS i

. 7.1 Guideline The applicant or licensee should briefly describe the management controls'that are used to verify that the procedures for preparation, validation, and routine utilization of the information handling system have been ano are being followed.  :

i 7.2 Evaluation  !

The licensee's response states that their Ovality Assurance Program Topical Report serves as the method of managerial control. CAD 18.0 (audits), as implemented by NOA 1.14, is used to verify the preparation, valir.ation, and reutine use of the information handling system. Quality assurance reviews and attits occur en a schedulec casts and assurv that the programs and their implementation are ccreect.

l 7.3 .C.o.n cl u si on We find that the management controls used by the licensee assure tnat i the information handling system is mairtained, is current. and is used as t

l intended. Therefore, the licensee's response for this item is acceptable.

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8. ITEM 2.2.1.5 - DESIGN VERIFICATION AND PRCCUREMENT '

8.1 G}p deline The spplicant's or licensee's subtittal should document that past usage demenstrates that appropriate design verification and qualification testing are specified for the crocurunent of safety-related components and  !

parts. Tbt specifications should include qualification testing for ,

expected safety service conditions and should 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 I available, confirmation that the present program meets these requirements sr.ould be previced.

i 8.2 Evaluation The licensee's submittals specify that Quality Assurance Program procedures CAP 4.0 arc 7.0 satisfy the requirements of this guideline, Procedure GEC 2.01 (Generatien Engineering and Construction Division) '

augments these procedures by stipulating testing requirerents and acceptan:e criteria. Procedure NE0 3.06 (Nuclear Engineering anc Operations) also supplements the Quality Assurance program pec:a: ores :j assuring that the purchase documer. s have the proper Ovality Assurance i c:cuments attached to them.

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8.3 Cenelusion i The licensee's response for this item is considered to be complace.  !

I The information provided addresses the concerns of this item and is ,

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9. ITEM 2.2.1.6 "!MPORTANT TO SAFETY" COMPONENTS l t

9.1 Guideline r Generic Letter 83 28 states that the licensee's equipment  !

classification program should include (ir addition to the safety-related i components) a broader c1sss of components designated as'"Important to i safety.- He.ever, since sne generic 1etter eoes not reautre the iteensee l

d to furnish this information as part of their response, this item will not [

! .e revie.ee. j I

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10. CONCLUSION '

! Based on our review of the licensee's response to the specific re wirements of Item 2.2.1, we find that the information provided by the l licenseetoresolvetheconcernsofIteis 2.2.1.1, 2.2.1.2, 2.2.1.3,  !

I 2.2.1.4, and 2.2.1.5 meets the requirements of Generic Letter 83-28 and is I I

acceptable. Item 2.2.1.6 was not reviewed, as noted in Section 9.1. .

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11. REFERENCES

1. Letter, NRC (O. G. Eisenhut) to all Licensees of Opurating Reactors, i A:plicants for Operating License, and Holders of Construction Permits, (

"Recuired Actions Based on Generic Implications of Salem ATW5 Events (Generic Letter 83-23)," July 8,1953. j

2. Letter, Northeast Utilities Co cany (W. G. Counsil) to NRO (O. G. Eisenhut), "Response to Generic Letter 83-28 Generic Implications of Salem ATW5 Events " November 8, 1923. A03381,  !

. Attachments 4 and 5. l

. 3. Lotter, Northeast Utilities Company (E. J. Mroezka) to NRC, .

"Generic Letter 83-28, Item 2.2, 'Ecuipment Classification',"  !

March 13, 1957, A06176, A06384, A06385, B12375, i I

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7,"M' BleuoGRAPHic DATA SHEET EGG-NTA 7384

.........4..... .......

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

EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED *** i's.

COMPONENTS: MILLSTONE-3 . . ' ' " * *,' < o-....

.~'-*+ January 198A A1an C. Udy . ..'. ****. ..

- January 1988

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EG&G Idaho Inc.

P. O. Box 1625 'TN - -

Idaho Falls. 10 83415 06001

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! . Division of Engineering and System Technology Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission ...co.... - e Washington, DC 20555 l

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Thi. EG&3 Idaho. Inc., report provides a review of the submituls from Unit rio. 3.

the Millstone Nuclear Power Station regarding confomance to Generic Lett;, 63-28 l Item 2.2.1. >

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