ML20245H903
| ML20245H903 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 08/15/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20245H896 | List: |
| References | |
| GL-83-28, NUDOCS 8908170346 | |
| Download: ML20245H903 (3) | |
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NUCLEAR REGULATORY COMMISSION
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO GENERIC LETTER 63-38, ITEM 2.1 (PART 1)-
EQUIPMENT CLASSIFICATION (RTS COMPONENTS)
OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1 DOCKET NO. 50-285 INTRODUCTION AND
SUMMARY
On February 25,19L0, both of the scram circuit breakers at Unit I 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 I of the Salem Nuclear Power Plant, an automatic trip signal was generated based on steam' generator low-low level during plant start-up.
In this case, the reactor was tripped manually by the operator almost coin-cidentally with the automatic trip.
Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (EDO), directed the 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, i
I theCommission(NRC) requested (byGenericLetter83-28datedJuly8,1983) all licensees of operating reactors, applicants for an operating license, and holders of construction permits to respond to generic issues raised by the analyses of these two ATWS events.
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This report is an evaluation of the response submitted by Omaha Public Power-District, the licensee for the Fort Calhoun Station, Unit 1, for Item 2.1 (Part 1) of Generic-Letter 83-28. The actual. documents reviewed as part of this evaluation are listed in the references at the end of the report, Item 2.1 (Part 1) requires the licensee to confirm that all Reactor Trip System components are identified, classified and treated as safety-related as indicated in the following statement:
Licensees.and applicants shall confirm that all components whose functioning is. required to trip the reactor are identified as sMy-related on documents, procedures, and information handling systems used in the plant to control. safety-related activities, in-cluding maintenance, work orders, and parts replacement.
EVALUATION The licensee for the Fort Calhoun Station, Unit I responded to the requirements 2
3 of Item 2.1 (Part 1) with submittals dated November 4,1983 and May 24, 1985,
The first submittal described their interim program for identifying and classi-fying RPS components which was based on their " Interim Electrical CQE List."
It also described their plan for improving the system to confirm that the list of safety-related components was complete and to verify that plant documents adequately
4 control activities that affect safety-related RPS components. The second sub-mittal stated that their improvement program had,been completed; confirmed that the CQE List was accurate and complete; and'that plant documents did adequately control activities associated with safety-related RPS components.
CONCLUSION Based on our review of these responses, we find the licensee's statements confirm that a program exists for identifying, classifying and treating components that are required for performance of the reactor trip function as safety related. This program meets the requirements of Item.2.1 (Part 1) of the Generic Letter 83-28, and is therefore acceptable.
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.
Letter, W. C. Jones, Omaha Public Power District, to D. G. Eisenhut, NRC, November 4, 1983.
3.
Letter, R. L. Andrews, Omaha Public Power District, to J. R. Miller, NRC, May 24, 1985.
Dated: August 15, 1989 Principal Contributor: Don R. Lasher
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CONFORMANCE TO GENERIC LETTER 83-28 ITEM 2.1 (PART 1) EQUIPMENT CLASSIFICATI0N'(RTS COMPONENTS)
FORT CALHOUN MILLSTONE UNIT 2 NINE MILE POINT UNIT 2 FORT ST. VRAIN 4
W-R. Haroldsen Published July 1986 EG&G Idaho, Inc.
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Prepared for the U.S. Nuclear Regulatory Comission
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Washington, D.C.
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Under DOE Contract No. DE-AC07-76ID01570 FIN Nos. D6001 and D6002 t
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ABSTRACT This EG8G Idaho, Inc.. report provides a. review of the. submittals f rom selected operating and applicant plants for conformance to Generic Letter 83-28, Item 2.1 f Part 1). The following plants are included in this review.
Plant Name Docket Number TAC Number Foxt Calhoun 50-285 52839 Millstone Unit 2 50-336 52855 Nine Mile Point Unit 2 50-410 OL Fort St. Vrain 50-267 52840 l
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FOREWORD This report is supplied as part of the program for evaluating licensee / applicant confomance 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 Comission, Office of Nuclear Reactor Regulation, Division of PWR Licensing-A, by the EG8G Idaho, Inc.
The U.S.- Nuclear' Regulatory Comission funded this work under the authorization B&R 20-1.9-10-11-3 and 20-19 40-41-3, FIN Nos. D6001 and D6002.
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i-CONTENTS ABSTRACT..........,.............'......................................
11 FOREWORD..............................................................
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I NTRODUCTI ON JUO
SUMMARY
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P LANT RESP ON SE EV ALU ATI ON S.......................................
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2.1 ForCalhoun................................................
3 2.2 Conclusion.................................................
3 2.3 Millstone 2................................................
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2.4 Conclusion.................................................
4 2.5-N i ne M il e P oi n t 2..........................................
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-2.6' Conclusion.................................................
5 2.7 Fort St. Vrain.............................................
6 2.8 Conclusions................................................
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GENERIC REFERENCES...............................................
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INTRODUCTION AND
SUMMARY
l On February 25, 1983, bcth of the scram circuit breakers at Unit 1 of the Salem Wuclear 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 f ailure of the circuit breakers was determined to be related to the sticking of the undervoltage trip attachment. Prior to the incident, on February 22, 1983, an automatic trip signal was generated at Unit 1 of the Salem Nuclear Power Plant 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 of Operations (EDO), directed the 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 1 incidents are reported in NUREG-1000,
" Generic Implications of the ATWS Events at the Salem Nuclear Power Plant."I As a result of this investigation, the Commission (NRC) requested (by Generic Letter 83-28, dated July 8, 1983)2 all licensees of operating reactors, applicants for an operating license, and holders of construction pemits to respond to generic issues raised by the analyses of these two ATWS events.
This report is an evaluation of the responses submitted from a group of reactor plants f or Item 2.1 (Part 1) of Generic Letter 83-28. The results of the reviews of several plant responses are reported on in this document to enhance review efficiency. The specific plants reviewed in this report were selected based on convenience of review. The actual documents which were reviewed for each evaluation are listed at the end of each plant evaluation. The generic documents referenced in this report are listed at the end of the report.
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Part 1.of item 2.1 of Generic Letter 83-28 requires the licensee or applicant to confirm that all reactor trip system components are identified, classified, and treated as safety-related as indicated in the following statement:
Licensees and applicants shall confirm that all components whose functioning is required to trip the reactor are identified as
'i safety-related on documents, procedures, and information handling systems used in the plant to control safety-related activities, l
including maintenance, work orders, and parts replacement.
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PLANT RESPONSE EVALUATIONS 2.1 Fort Calhoun, 50-285 TAC No. 52839 The licensee for the Fort Calhoun Nuclear Plant (Omaha Public_ Power District) provided responses to Items 2.1 (Part 1) of Generic Letter 83-28 in submittals dated November 4,1983 and May 24, 1985. The first submittal-described an interim program for identifying and classifying components of the reactor protective system. This system was based on the " Interim Electrical CQE List." A plan was described for upgrading the system, to provide confirmation that the list of safety-related components was
- complete and to review applicable procedures to insure that plant documents adequately control activities that relate to reactor protective system components.
The second submittal stated that review of the reactor protective system equipment list had been completed and the' CQE List was found to be accurate and complete.. This. list has been integrated into the plant information data base. The submittal also states that the review of reactor protective sys, tem procedure had been completed to verify the adequacy of controls. The controls are said to meet the requirements of the Generic Letter.
i 2,2 Conclusions Based on the review of the licensee's submittal, we find that the components necessary to perform reactcr trip are classified as safety-related and that activities relating to safety-related components are controlled by procedures which reflect the necessary requirements for handling safety-related components. We, therefore, find that the licensee's responses meet the requirements of Item 2.1 (Part 1) of Generic Letter 83-28 and are acceptable.
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'4 References.
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Letter, W. C.. Jones, Omaha. Public. Power District, to D. G.- Eisenhut, NRC, November 4,>1983.
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Letter, R. L. Andrews, Omaha Public Power District, to J. R. Miller, NRC, May 24,1985. -
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-2.3 Millstone Unit 2, 50-336, TAC No. 52855
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The licensee for Millstone Unit 2 (Northeast Nuclear Energy Co.)
' responded to the requirements of Item 2.1 (Part 1.) in submittals dated
, November 8,.1983 and May 9,1985. The submittals state that all components whose function is required to. trip. the reactor are identified as Category 1 (safety'related). on their Material,-Equipment and Parts List (MEPL) and that safety-related activities on these components including maintenance, work orders and parts replacement will be completed using Category 1 controls.
2.4 Conclusion Based on the review of. the licensee's submittals we find that the licensee's responses confim that the components necessary to perform reactor trip are classified safety related and that all activities relating to these-components are designated as safety related. These responses,.
therefore, meet the requirements of Item 2.1 (Part 1) of Generic Letter 83-28, and are acceptable.
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References L
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Letter, W. G. Counsil, Northeast Nuclear Energy Co., to D. G. Eisenhut, NRC, November 8,1983.
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Letter, J. F. Opeka, Northeast Nuclear Energy Co., to J. A. Zwolinski, NRC, May 9,1985.
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,,v 2.5 Nine Mile Point' Unit 2, 50-410, (OL)
The applicant for Nine Mile Point Unit 2 (Niagra Mohawk Power Corporation) provided responses to Item 2.1 (Part 1) of Generic Letter 83-28 in submittals dated April 10, 1984, December 20, 1985 and' April 15,1986. The last of these submittals states that an equipment classification list (Q-list) is used at Nine Mile Point Unit 2 to identify safety-related components including those components which contribute to the reactor trip function.
Administrative controls consisting of documents, procedures and information handling systems are used to control safety-related activities including maintenance, work requests, parts replacements and modifications. These activities require classification information derived f rom the 0-list. A Computerized Infomation handling system is being developed which will include information on all equipment installed at the plant.
2.6 Conclusion l
Based on the review of the licensee's submittal, we find that the components necessary to perfom reactor trip are classified as safety-related and that activities relating to safety-related components are controlled by procedures which reflect the necessary requirements for handling safety-related components. We, therefore, find that the licensee's response meets the requirements of Items 2.1 (Part 1) of Generic Letter 83-28 and is acceptable.
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References 1.
Letter, G. K. Rhode, Niagra Mohawk Power Co., to A. Schwencer, NRC, April 10,1984.
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Letter, T. E. Lempges, Niagra Mohawk Power Co., to A. Schwencer, NRC, December 20, 1985.
3.
Letter, T. E. Lempges, Niagra Mohawk Power Co., to E. G. Adensam, NRC, April 15,1986.
2.7 Fort St. Vrain, 50-267, TAC No. 52840 The licensee for the Fort St. Vrain Nuclear Plant (Public Service Company of Colorado) responded to the requirements of Item 2.1 (Part 1) of Generic Letter 83-28 in submittals dated November 4,1983 and June 12, 1985. The first submittal states that the licensee was engaged in a
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project to highlight Class I electrical equipment required for safe shutdown including reactor trip system components on schematic drawings.
j An audit has been completed by the licensee to verify that safety-related components are indicat'ed as such on a computerized safety-related component list and on the process and instrumentation diagrams.
Components are not necessarily labeled as safety-related within safety-related procedures but provide appropriate guidance for activities related to safety-related components.
The June 12, 1985 submittal states that the plant maintenance procedures were being rewritten and that all reactor trip system components would be identified as safety-related in the revised procedures. The revised procedures were expected to be completed by January 3,1986.
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2.8-Conclusions Based on the review of the licensee's submittal, w find that the components necessary to perfonn reactor trip are classified as safety-related and that activities relating to safety-related components are controlled by procedures which reflect the necessary requirements for handling safety-related components. We, therefore, find that the licensee's responses meet the requirements of Item 2.1 (Part 1) of Generic Letter 83-28 and are acceptable.
I References i
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Letter, O. R. Lee, Public Service Company of Colorado, to l
- 0. G. Eisenhut, NRC, November 4,1983.
2.
Letter, J. W. Gahm, Public Service Company of Colorado, to Regional Administrator, Region IV, NRC, June 12, 1985.
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GENERIC REFERENCES 4
1.
Generic Implications of ATWS Events at the Salem Nuclear Power Plant, NUREG-1000,..Yolume 1 April'1983; Yolume 2, July 1983.
2.
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 l
(Generic Letter 83-28)," July 8,1983.
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