ML20214R463

Comformance to Generic Ltr 83-28,Item 2.2.1 -- Equipment Classification for All Other Safety-Related Components: Crystal River-3, Final Informal Rept

ML20214R463
Person / Time
Site:	Crystal River
Issue date:	04/30/1987
From:	Udy A EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC
Shared Package
ML20214R457	List: ML20214R455 ML20214R463
References
CON-FIN-D-6001 EGG-NTA-7315, GL-83-28, TAC-53664, NUDOCS 8706080153
Download: ML20214R463 (19)

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i EGG-NTA-7315 1j April 1987 i

4 INFORMAL REPORT I

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idaho i Motional

i CONFORMANCE TO GENERIC LETTER 83-28, ITEfi 2.2.1--

Eng/neering EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-Laboratory RELATED COMPONENTS: CRYSTAL RIVER-3 Managed. .,

by the U.S. i Department Alan C. Udy ofEnergy i

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uo. ou$l/$ls'*! U.S. NUCLEAR REGULATORY COMMISSION 8706080153 DR 870417 ADOCK 05000302

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e DISCLAIMER This book was prepared as an account of work sponsored by an agency of the Uruted States Government. Neither the Uruted States Government not any agency thereof, nor any of tneer employees, makes any warranty, express or imphed, or assurnes any legal liabikty or responsibihty for the accuracy, completeness, or usefulness of any information, apparatus, product or process disclosed, or represents that its use would not infnnge pnvately owned nghts. References herein to any specife commeroal product, process, or servce by trade name, trademark, manufacturer, or otherwise, does not necessanly constitute or imply its endorsement, recommendation, or favonng by the United States Government or any agency thereof. The views and opiruons of authors expressed herein do not necessanty state or reflect those of the United States Government or any agency thereof.

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.7 EGG-NTA-7315 4

TECHNICAL EVALUATION REPORT ~

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

EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS:

CRYSTAL RIVER-3 l Docket No. 50-302 Alan C. Udy i'

4 Published April 1987 s,

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Idaho National Engineering Laboratory l EG&G Idaho, Inc.

2 Idaho Falls, Idaho 83415 1 .

i Prepared for the i U.S. Nuclear Regulatory Commission

! Washington, D.C. 20555 ,

Under DOE Contract No. DE-AC07-76ID01570 j FIN No. 06001 i

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

Docket No. 50-302 TAC No. 53664 11 1

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. . FOREWORD 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 PWR Licensing-A, by EG&G Idaho, Inc., NRR and I&E Support Branch.

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

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authorization B&R 20-19-10-11-3, FIN No. 06001.

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4 Docket No. 50-302

. TAC No. 53664 tii

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CONTENTS ABSTRACT .............................................................. 11 FOREWORD .............................................................. iii

1. INTRODUCTION ..................................................... 1

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

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

4. ITEM 2.2.1.1 - IDENTIFICATION CRITERIA ........................... 4 4.1 Guideline .................................................. 4 4.2 Evaluation ................................................. 4 4.3 Conclusion ................................................. 4

5. ITEM 2.2.1.2 - INFORMATION HANDLING SYSTEM ....................... 5 5.1 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 ................................................. 6

7. ITEM 2. 2.1. 4 - MANAGEMENT CONTRO LS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 7.1 Guideline .................................................. 7 7.2 Evaluation ................................................. 7 7.3 Conclusion ................................................. 7

8. ITEM 2.2.1.5 - DESIGN VERIFICATION AND PROCUREMENT ............... 8 8.1 Guideline .................................................. 8 8.2 Evaluation ................................................. 8 8.3 Conclusion ................................................. 8

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

10. CONCLUSION ................................................ ..... 10 :

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11. REFERENCES ....................................................... 11 ;

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

EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS:

CRYSTAL RIVER-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 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 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.

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 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 this investigation, the Commission (NRC) requested 1

(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 Florida Power Corporation, the licensee for Unit No. 3 of the Crystal River Station, 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 equipment classification are drawn.

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3. ITEM 2.2.1 - PROGRAM 3.1 Guideline Licensees and applicants should 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 replacement parts. Licensee and applicant responses which address the features of this program are evaluated in the remainder of this report.

3.2 Evaluation ,

The licensee for Crystal River Unit 3 responded to these requirements with a submittal dated November 4, 1983.2 This response was revised on July 31, 1984.3 These submittals include information that describes the licensee's safety-related equipment classification program. In the review of the licensee's response to this item, it was' assumed that the information and documentation supporting this program is available for audit upon request.

The licensee states that their Safety Listing is the information handling system referred to. The Safety Listing is consulted in preparing ,

work requests and work packages to identify safety-related components, activities and procedures.

3.3 Conclusion We have reviewed the licensee's information and, in general, 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 their program used for l equipment classification includes criteria used for identifying components .

as safety-related.

4.2 Evaluation The licensee's response gives the criteria for identifying 4

safety-related equipment and components. A component is considered -

safety-related if it is required to assure: (a) the integrity of the reactor coolant system pressure boundary, (b) the capability to achieve and maintain a safe shutdown or (c) the capability to prevent or to mitigate

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the consequences of an accident which could result in potential offsite i exposures.

4.3 Conclusion i

We find that the criteria used in the identification of safety-related components meets the requirements of Item 2.2.1.1 and are acceptable.

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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  !

i equipment and that procedures exist which govern its development and validation.

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5.2 Evaluation W

The Itcensee describes how the Safety Listing was originally prepared

in accordance.with the Architect / Engineer's (Gilbert Associates Inc. (GAI))

] Procedures Manual. The Safety Listing was verified by review, comment and j approval signature by GAI and the licensee's engineering groups,. plant t

staff and the Quality programs Department, as well as approvals by the j- managers for the Nuclear Engineering Department and the Production  :

Engineering Department. Revisions are controlled by safety-related engineering procedure (SREP)-1, " Safety Identification Design Input ,

Requirements." Revisions and new entries are prepared by a design engineer and reviewed, verified, and approved by supervisors, the Quality Program Department, the Nuclear Plant Manager and the Nuclear Engineering Department Manager. The description describes-how the Safety Listing'is maintained.

5.3 Conclusion We find that the information 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. l

~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 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 4 activities defined in the introduction to 10 CFR 50, Appendix B, apply to safety-related components.

6.2 Evaluation The licensee's responses describe the utilization of the Safety Listing to determine when an activity is safety-related. The licensee states the Plant Operating Quality Assurance Manual Compliance Procedure CP-113, " Procedure for Handling and Controlling Work Requests and Work Packages," requires the consultation of the Safety Listing in determining-if an activity is safety-related. This procedure is used in preparing any work requests for those activities identified by this sub-item. The reviews called out by CP-113 insure that the proper procedures are used for maintenance work, routine surveillance testing, accomplishment of design changes, performance of engineering support work, accomplishment of setpoint changes and the performance of special tests and studies.

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

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7. ITEM 2.2.1.4 - MANAGEMENT CONTROLS 7.1 Guideline 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 their method of managerial control includes Nuclear Engineering Department supervisory approval of any l revisions to the Safety Listing and approvals of forms for safety classification review and for classification of items and services and for l approvals of procurement packages and design change packages. All work requests have supervisory approval of the Nuclear Plant Department.

Changes and revisions to the Safety Listing and to the SREP-1 controlling procedure for the Safety Listing require managerial approvals. Audits by the Quality Programs Department are also used to verify the preparation, validation and routine use of the information handling system and to assure that safety-related activities and their implementation are correct.

7.3 Conclusion 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 response for this item is acceptable.

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8. ITEM 2.2.1.5 - DESIGN VERIFICATION AND PROCUREMENT 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 4 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 The licensee states that the Nuclear Procurement and Storage Manual contains design verification and qualification testing requirements for both replacement parts and new equipment. The requirement for the vendor to submit evidence of testing is specifically addressed by the licensee, and requires qualification reports frem the vendor.

8.3 Conclusion We consider the licensee's response for this item to be complete. The information provided addresses the concerns of this item and is acceptable.

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9. ITEM 2.2.1.6 "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 safety-related components) a broader class of components designated as "Important to Safety." However, since the generic letter does not require i the licensee or applicant to furnish this information as part of their  ;

response, review of this itcd will not be performed. l i

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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 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 .

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

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11. 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. Florida Power Corporation letter, G. R. Westafer to Director of Nuclear Reactor Regulation, NRC, " Generic Letter 83-28, Required Actions Based on Generic Implications of Salem ATWS Events,"

November 4, 1983, 3F1183-03. '

3. Florida Power Corporation letter, G. R. Westafer to Director of Nuclear Reactor Regulation, NRC, " Updated Response to Generic Letter 83-28," July 31,1984, 3F0784-21.

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- April  ! 1987 7 PtxFOmWING OmsANi2 ATIO80 N AMG AND MatuMG AcontSS ifarnesele Comp 8 PROJECT (TAsatruvomst uNat NutettR EG8G Idaho, Inc.

P. O. Box 1625 ""'"'"'""'"

Idaho Falls, ID 83415 D6001

10. SPONSOR *NG ORGANIZ ATION NAME AseO MastsNG ADOntse rfaraser te cesw 11a TYPE OP REPORT Division of PWR Licensing - A Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission a PeaiOO CoventO u~.-u Washington, DC 20555 12 SuPPLautNT AMv NOTES 13 a85TR ACT (J00 = ores er 'esas This EG&G Idaho, Inc., report provides a review of the submittals from the Florida Power Corporation regarding conformance to Generic Letter 83-28, Item 2.2.1 for Crystal River-3.

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