ML20055E514
| ML20055E514 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/02/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20055E513 | List: |
| References | |
| GL-83-28, NUDOCS 9007120069 | |
| Download: ML20055E514 (4) | |
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i SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i
FLORIDA POWER CORPORATION, ET AL.
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CRYSTAL RIVER UNIT NO. 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302 GENERIC LETTER 83-28, ITEM 2.2.1 EOUIPMENT CLASSIFICATION PROGRAMS FOR ALL SAFETY-RELATED COMPONENTS i
1.0 INTRODUCTION
i Generic Letter 83-28 was issued by the NRC on July 8,1983. This letter listed actions to be taken by licensees and applicants based on the generic implications of the Salem ATWS events.
Item 2.2.1 of the letter states that licensees and applicants shall describe their program for classifying all safety-related components other than reactor trip system (RTS) components. Specifically, the licensee / applicant's submittal was required to contain information describing (1) the criteria used to identify these components as safety-related;,(2) the information handling system which identifies the components as safety-related; (3) the manner in which station personnel use this-information handling system to control activities affecting these components;.(4) management controls that are used to verify that the information handling system is prepared, maintained, validated, and used in accordance with approved procedures; and (5) design-verification and qualification testing requirements that are part of the specifications for procurement of safety-related components.
The licensee for the Crystal River Unit No. 3 Nuclear Generating Plant submitted original and revised responses to Generic Letter 83-28 -Item 2.2.1 on November 4, 1983 and July 31, 1984.
2.0 EVALUATION In th'ese sections the licensee's responses to the program and each of five sub-items are individually evaluated against guidelines developed by the staff and conclusions are drawn regarding their individual and collective acceptability.
2.1 Identification Criteria Guideline: The licensee's response should describe the criteria used to identify safety-related equipment and components.
(Item 2.2.1.1) i 9007120069 900702 PDR ADOCK 05000302 P
' Evaluation:
The_ licensee's responses contained their criteria for identifying safety-related equipment and components. These criteria state that a com is considered to be safety-related if it is required to assure (a)ponent
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the integrity of the reactor coolant system pressure boundary, (b) the capa-bility _to achieve and maintain a safe shutdown, or (c) the capability i
to prevent or to mitigate the consequences of an accident which could result in potential'offsite exposures, i
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Conclusion:==
.t We find the stated criteria meet the staff ~4 requirements _and are acceptable.
2.2 Information Handling System Guideline: The licensee's response 1 should confirm that the equipment i
classification program includes an information handling system that is used.to identify safety-related equipment and components. Approved procedures which govern its development, maintenance, and validation should exist.
(Item 2.2.1.2)
Evaluation; The licensee, in its responses, described how the information handling system (safety listing) was prepared,- verified, and approved. Revisions and new entries are prepared and controlled according to a safety-related engineering procedure-(SREP-1) that provides for-review, verification, and approval of all such changes.
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Conclusion:==
We conclude that the licensee's information handling system meets staff requirements and is' acceptable.
2.3 Use of Information Handling System Guideline: The licensee's response shouli confirm that their equipment classification program includes crite-it and procedures that govern the i
use of the information handling syste* to determine whether an activity is safety-related and that safety-related procedures for maintenance, i
surveillance, parts replacement, and other activities defined-in the t
introduction to 10 CFR Part 50, Appendix B, are applied to safety-related components. (Item 2.2.1.3) i Evaluation:
s The. licensee's response describe how the safety listing (is used according to their Plant Operating QA Manual Compliance Procedure CP-113) to prepare 1
work requests for activities such as maintenance work, routine surveillance i
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3-testing, accomplishment of design changes, performance of engineering support work, accomplishments of.setpoint changes,..and performance of special tests and studies.
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Conclusion:==
We conclude that the licensee-has described plant administrative controls and procedures that meet staff requirements for this item and are acceptable.
2.4 Management Controls Guideline: The licensee / applicant should confirm that management controls used to verify that.the procedures for preparation, validation,, and routine use of the information handling system have been and are being followed.
(Item 2.2.1.4)
Evaluation:
4 The licensee's responses stated that changes and revision to the safety listing and its controlling document. require management review and approval.
All work requests, procurement packages, and design change packages also require' management review and approval..Further, audits by the Quality-Programs Department are used by management to verify that preparation, validation, and routine use of the information handling syst1m is done correctly.
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Conclusion:==
We conclude that the licensee has described acceptable means by which.
their management is informed of the correct use of the information handling system and these means are acceptable.
2.5 Design Verification and Procurement Guideline:.The licensee / applicant's response should document that past usage demonstrates that appropriate design verification and qualification testing is specified for the procurement of safety-related components and parts. The specifications should include qualification testing for.
expected safety service conditions and provide support for licensee's receipt of testing documentation that supports the limits of life recom-7 mended by the supplier.
If such documentation is not available, confir-l-
mation that the present program meets these requirements should be provided.
(Item 2.2.1.5) l l
. Evaluation:
l The licensee states that their nuclear procurement and storage manual contains design verification and qualification testing requirements for both new and replacement parts and equipment. Vendors are specifically i
required by this procedure to submit qualification reports with all j
shipments of safety-related parts and equipment.
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Conclusion:
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We find the licensee's procedures meet the staff requirements for this l
item and are acceptable.
2.6 "Important To Safety" Components Guideline: Generic Letter 83-28 states _ that licensee / applicant equ'ipn.3nt classification programs should include, in addition to the safety-relate:1-l components, a broader class of components designated'as "Important to Safety." However, since the Generic Letter does not require the licensee /
applicant to furnish this infoi1 nation as part'of their response, staff review of this sub-item will not be performed.
(Item 2.2.1.6) 2.7 Equipment Classification Program Guideline: Licensees / applicants should confirm that.they have an equipment classification program that provides assurance that all safety-related components are designated as safety-related on plant documents such as drawings, procedures, system descriptions, test and maintenance instruc-tic..s, operating procedures,'and information handling systems.. This will i
ensure that personnel who perform activities that affect such safety-related i
components are aware that they are working on safety-related components 1
and are guided by safety-related procedures and constraints. -(Items 2.2.1).
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Evaluation:
l The licensee's responses to these requirements were contained in a submittal dated November 4,1983 and in a revised submittal dated July 31, 1984.
These submittals described the licensee's program for identifying and classifying safety-related equipment and components which meet the staff requirements.as indicated in preceding sub-item evaluations.
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Conclusion:==
We concluded that the licensee's program addresses the' staff _ concerns regarding equipment and component classification'and is acceptable.
3.0
SUMMARY
AND CONCLUSION i
Based on our review of the licensee's submittals as summarized above, and of i
our contractor's Technical Evaluation Report, we conclude that the licenseek-responses to Generic Letter 83-28 Item 2.2.1, are acceptable.
I Dated: July 2, 1990 Principal Contributor:
D. Lasher i
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April 1987 INFORMAL REPORT 1
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CONFORMANCE TO GENERIC LETTER 83-28. ITEM 2.2.1--
EQUIPMJNT CLASSIFICATION FOR ALL OTHER SAFETY-
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Laboratory j
RELATED COMPONENTS:
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EGG-NTA-7315 TECHNICAL EVALUATION REPORT i
CONFORMANCE TO GENERIC LETTER 83-28, ITEM 2.2.1--
I E0VIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS:
l' CRYSTAL RIVER-3 I.
Docket No. 50-302 Alan C. Udy Published' April 1987~
Idaho National Engineering Laboratory EG&G Idaho, Inc.
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Prepared for the U.S. Nuclear Regulatory Commission.
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20555 Under 00E Contract No. OE-AC07-76ID01570 FIN No. 06001
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i ABSTRACT This EG&G Idaho, Inc... report provides a review of the submittals from-
. I Unit No. 3 of the Crystal River Station for'conformance to Generic 4
Letter 83-28, Item 2.2.1.
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Docket No.60-302 TAC No. 53664 h
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FOREWORD j
This report is suoplied as part of the program for evaluating 1
licensee / applicant conformance to Generic-Letter.83-28 " Required Actions Based on Generic-Implications oof 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.
l The U.S. Nuclear Regulatory' Commission funded this work under the authorization B&R 20-19-10-11-3, FIN No. 06001.
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Docket No. 50-302 TAC Jh). 53664 w-jjj l-r
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CONTENTS-A B S T RA C T.............. -............................................. -,..
11 FOREW0RD...............................................................
iii 1.
INTRODUCTION.........................................'...........
1 2.
R EV I EW CON T EN T AN D FO RMAT...................... -....................
2 a
3.
I T EM 2. 2.1 - P ROG RAM..............................................
3 6-f 3.1 Guideline..................................................
3 3.2 Evaluation................................
3 3.3' Conclusion.................................................
3L 4.
ITEM 2.2.1.1 - IDENTIFICATION CRITERIA...........................
4 4.1 Guideline.........,,........................................
4 4.2 Evaluation.................................................
'4 4.3 Conclusion.................................................
4 1
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ITEM.2.2.1.2 - INFORMATION' HANDLING SYSTEM.......................
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5.1 Guideline..................................................
5 5.2 Evaluation.................................................
'S 5.3-Conclusion.................................................
5 6.
ITEM 2.2.1.3 - USE OF EQUIPMENT CLASSIFICATION LISTING...........
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.6.1 Guideline..................................................
6 6.2 Evaluation.................................................
6 6.3 Conclusion.........................................-........
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ITEM 2.2.1.4 - MANAGEMENT CONTROLS...............................
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.......................................................
10-11.
REFERENCES.......................................................
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CONFORMANCE TO GENERIC LETTER 83-28,'fTEM 2.2.1--
EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMP 0NENTS:
CRYSTAL RIVER-3 1.
INTRODUCTION On Februcry 25' 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant failed to open upon an automatic reactor trir 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
.I generator low-low level during plant startup.
In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip.
3 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 M tha Salem Nuclear Power Plant. The results-of the staff's inquiry into the f
generic implications of the Salem incidents are reported in NUREG-1000,
" Generic Implications of the ATWS Events'at the Salem Nuclear Power
-i 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 ratsad 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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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.
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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
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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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ITEM 2.2.1 - PROGRAM 3,1 Guideline Licensees and applicants should confirm that an equipment.
classification program exists which provides assurance that all
'i 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-
.i replacement parts.' Licensee and applicant responses which' address the features of this program are evaluated.in the' remainder of this report.
i 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 l
the licensee's response is adequate.
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ITEM 2.2.1.1 - IDENTIFICATION CRITERIA 4.1 Guideline
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l The applicant or licensee should confirm that their program used for
~ equipment classification includes criteria used for identifying components i
as safety-related.
4.2 Evaluation j
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The licensee's' response gives the criteria for identifying i
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safety-related equipment and components. 'A component is considered 1
i safety-related if it is required..to assure:
(a):-the. integrity of the l
l reactor coolant system pressure' boundary, (b) the capability to achieve and t-t maintain a safe shutdown or-(c) the capability to prevent'or to mitigate
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l the consequences of an accident which could result in potential offsite exposures.
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4.3-Conclusion 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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ITEM 2.2.1.2 - INFORMATION HANDLING SYSTEM 5.1 Guideline i
The licensee.or applicant.should confirm that the program for
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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 i
The licensee 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 by1 review, comment and approval signature by GAI and the licensee's engineering groups, plant staff and the Quality Programs Department,-as well as approvals by the 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 Req ui ren,a n ts. " Revisions and new entries are prepared by a design engineer-and reviewed. verified, and approved tar 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.
Therefore, the information provided by the licensee for this item is acceptable.
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ITEM 2.2.1.3 - USE OF EQUIPMENT CLASSIFICATI0N' LISTING i
6.1 Guideline l
The licensee's or applicant's description should confirm that their~
c 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--
I activities defined in the ir.troduction to 10 CFR 50, Appendix B,-apply to 1;
safety-related components.
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6.2 Evaluation 3
The licensee's responses describe the utilization.of the Safety i.
Listing to determine when an activity is safety-related. The' licensee' I
states the Plant Operating Quality Assurance Manual _ Compliance Procedure CP-113, " Procedure for Handling-and-Controlling Work Requests'and Work
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Packages," requires the consultation of the Safety Listing in< determining if an activity is safety-related. -This procedure.is'used in preparing'any-l work requests for those activities identified by this sub-item. The l
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 t
We find that the licensee's description of plant. administrative controls and procedures meets the requ rements of this. item and is, therefore, acceptable.
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ITEM 2.2.1.4 --MANAGEMENT CONTROLS' l
7.1 Guideline l
i, The applicant or licensee -should confirm that the management. controls l
used to verify that the procedures for preparati)n, validation and routine j
utilization of,the -information handling system Lave.been followed.
l 7.2-Evaluatica s
The licensee's response states that their method of managerial control; i
includes Nuclear Engineering Department supervisory approval of any revisions to the Safety Listing andiapprovals of forms for safety i
classification review and for classification of items and services and for j
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 i
the information handling system is maintainec, is current and is used as intended. Therefore, the licensee's response for this item is acceptable.
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ITEM 2.2.1.5'- DESIGN VERIFICATION AND PROCUREMENT-y -
I 8.1 Guideline l
The applicant's or; licensee's submittal should document that past, l_
-usage demonstrates that. appropriate design verification and qualification testing is specified for the procurement of-safety-related' component's and parts. The; specifications should include qualification testing' for' expected safety service conditions land 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 from 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, l
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ITEM 2.2.1.6:
"IMPORTANT-TO-SAFETY'i COMPONENTS 9.1 Guideline 1
i Generic: Letter 83-28 states that the licensee's or applicant's-equipment classification. program should include (in addition to-the safety-related ccmponents) a broader class of' components designated as-L "Important to Safety." However,-since the generic letter does not require the licensee or applicant to furnish tnis information as part of,their response, review of this item.will not be performed.
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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, 2.271.2, 2.2.1.3, 4
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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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 Nuct'ar Reactor Regulation, NRC, " Generic: Letter 83-28, Required 3
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 I
Nuclear Reactor Regulation,.NRC,~" Updated Response to Generic
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Letter-83-28," July 31,.1984, 3F0784-21.
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EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS:
CRYSTAL RIVER-3
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Washington, DC 20555
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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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