ML20214R397
| ML20214R397 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 03/31/1987 |
| From: | Udy A EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20214R373 | List: |
| References | |
| CON-FIN-D-6001 EGG-NTA-7283, GL-83-28, TAC-53700, TAC-53701, NUDOCS 8706080111 | |
| Download: ML20214R397 (19) | |
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EGG-NTA-7283 March 1987 4
INFORMAL REPORT
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Idaho National i
CONFORMANCE TO GENERIC LETTER 83-28, ITEM 2.2.1, EnE nserinE EQUIPMENT CLASSIFICATION (ALL OTHER SAFETY-Laboratory -
RELATED COMPONENTS), PEACH BOTTOM-2 AND -3 Managed.
by the U.S.
Department Alan C. Udy ofEnergy 0
E G n G.o 94 work performed unser Prepared far the DOE Contract l
uo ouComansa U.S. NUCLEAR REGULATORY COMMISSION 8706080111 870311 DR ADOCK 05000277 PDR l
l DISCLAIMER This book was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any ency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability 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 spoofic commercal product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessanly constitute or imply its endorsement, recommendation, or favonng by the United States Govemment or any agency thereof. The views and opinions of authors expressed herein do not necessanly state or reflect those of the United States Govemment or any agency thereof.
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EGG-NTA-7283
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TECHNICAL EVALUATION REPORT CONFORMANCE TO GENERIC LETTER 83-28, ITEM 2.2.1, EQUIPMENT CLASSIFICATION (ALL OTHER SAFETY-RELATED COMPONENTS),
PEACH BOTTOM -2 *.;JD -3 Docket Nos. 50-277/50-278 Alan C. Udy i
Published March 1987 i
Idaho National Engineering Laboratory EG&G Idaho, Inc.
Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Under DOE Contract No. DE-AC07-76IO01570 FIN No. 06001
ABSTRACT This EG&G Idaho, Inc., report provides a review of the submittal for Unit Nos. 2 and 3 of the Peach Bottom Atomic Power Station on conformance-to Generic Letter 83-28, Item 2.2.1.
Docket Nos. 50-277/50-278 TAC Nos. 53700/53701 11
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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 Regulatio;., 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 authorization B&R 20-19-10-11-3, FIN No. D6001.
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4 Docket Nos. 50-277/50-278 TAC Nos. 53700/53701 111
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e CONTENTS s
ABSTRACT..............................................................
11 FOREWORD..............................................................
111 1.
INTRODUCTION.....................................................
1 2.
REVIEW CONTENT AND FORMAT........................................
'2 3.
I T E M 2. 2.1 - P R O G R AM.............................................
3 3.1 Guideline..................................................
3 4
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 s
3 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 CONTROLS...............................
7 7.1 Guideline..................................................
7 7.2 Evaluation.................................................
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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" COMP 0NENTS..................
9 6
9.1 Guideline..................................................
9 10.
CONCLUSION.......................................................
10 11.
REFERENCES.......................................................
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1 CONFORMANCE TO GENERIC LETTER 83-28. ITEM 2.2.1 EQUIPMENT CLASSIFICATION (ALL OTHER SAFETY-RELATED COMPONENTS),
PEACH BOTTOM -2 AND -3 1.
INTRODUCTION 8
On February 25, 1983, both of the scram circuit breakers at Unit 1 of
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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 (E00), 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 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 response submitted by the j
Philadelphia Electric Company, the licensee for the Peach Bottom Atomic 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 the report.
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 stated, 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 concerning the acceptability of the program of the licensee or applicant 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 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 i
The licensee for the Peach Bottom Atomic Power Station responded to these requirements with a submittal dated November 4, 1983.2 This submittal included information that describes their 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 the information source used to identify safety-related parts and components is their Q-list. The submittal states.
that the Q-list is the single controlling document that identifies safety-related structures, systems and components.
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 equipment classification includes criteria used for identifying components e
as safety-related.
4.2 Evaluation The licensee's response gives the criteria for identifying 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 the consequences of an accident which could result in potential offsite exposures. The licensee states that these criteria are in conformance with the 10 CFR 50, Appendix 8, requirements.
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 - INf0RMATION HANDLING SYSTEM 5.1 Guideline l
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
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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 The licensee states that the Q-list identifies the safety-related components. This list is verified by the Mechanical Project Engineer.
4 The licensee's description indicates that the Q-list originated with the-Architect / Engineer, and includes the methods used from the time the Q-list was turned over to the licensee for development ~and. validation; the-process by which new safety-related items are entered; how changes in classification are made, how listed items are verified; how unauthorized changes are prevented; and how the Q-list is maintained and distributed to users as a single, official, consistent and unambiguous. document.
The Q-list is maintained and controlled by Engineering and Research Department Procedure (ERDP) 3.2.
Deletions from the Q-list are made in accordance with ERDP 3.3.
The Q-list amendments ind revisions are provided
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to users with a return receipt requirement.
The returr. receipts are logged by the Mechanical Project Engineer, thus assuring a single, official, 4
consistent and unambiguous document.
5.3 Conclusion 3
i We find that the information contained in the licensee's submittal is i
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 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 activities defined in the introduction to 10 CFR 50, Appendix 8, apply to safety-related components.
6.2 Evaluation The licensee states that the Maintenance Request Form (MRF) system determines whether activities are classified as safety-related or non-safety-related. This system has been updated to a computer-based Computerized History and Maintenance Planning System (CHAMPS), which automates MRF preparation and tracking.
The safety-related status of structures, systems, components and parts is entered onto the MRF by the CHAMPS. The CHAMPS is updated by direct access to the current Q-list.
The licensee has described plant administrative controls and procedures that govern maintenance, modification and procurement activities.
These controls assure that the Q-list status of components and systems is known before any maintenance, testing, design changes, engineering support, setpoint changes or special tests or studies are initiated.
6.3 Conclusion e
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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ITEM 2.2.1.4 - MANAGEMENT CONTROLS l
I 7.1 Guideline 1
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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 the information handling system have been-followed.
i 7.2 Evaluation The licensee's submittal describes the managerial controls that are applied 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 classification as intended.
These controls are maintained by the Engineering and Research Department management and are operated through the Mechanical Project Engineer.
Surveillance and audits are performed by the-Quality Assurance Program. Checks and balances within the Electric Production Department Quality Assurance Division' audit Quality Assurance-departmental procedures, audits and surveillances, in addition-to audits of plant procedures and of the Maintenance Request Form System.
7.3 Conclusion l
We find that the licensee's description meets the requirements of this item and is, therefore, acceptable.
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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.
- r testing is specified for the procurement of safety-related components and
- parts. The specifications should include qualification testing for expec ed sa ety service con it ons and provide support for the-t f
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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 lists seven Engineering and Research Department Procedures (ERDP) that verify the appropriate use-of replacement parts and insure the-technical and quality requirements, including documentation, verification of design capability and evidence of. testing, are included in' the purchase specifications. The procedures listed are ERDP 3.4, 4.4, 4.5, 4.6, 6.2, 6.3, and 7.1.
The licensee included a brief overall description of each procedure.
8.3 Conclusion We consider the licensee's response for this item to be complete..The j
information provided address the concerns of this item and is acceptable.
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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 doas not require the licensee or' applicant to furnish this information as part of their response, revleu 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, Equipment Classification Program for All Other Safety-Related Components, 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 8ased on Generic Implications of Salem ATWS Events (Generic Letter 83-28)," July 8, 1983.
~2.
~ Philadelphia Electric Company letter. S. L. Daltroff to D. G. Eisenhut, NRC, November 4, 1983.
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