ML20246B815
| ML20246B815 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley, Millstone, Monticello, Nine Mile Point, Fermi, Oyster Creek, Hope Creek, Limerick, Duane Arnold, LaSalle, 05000000 |
| Issue date: | 04/30/1989 |
| From: | Farmer F EG&G IDAHO, INC. |
| To: | NRC |
| Shared Package | |
| ML20246B820 | List: |
| References | |
| CON-FIN-D-6001 EGG-NTA-7470, EGG-NTA-7470-R01, GL-83-28, NUDOCS 8905090179 | |
| Download: ML20246B815 (22) | |
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EGG-NTA-7470 Revision 1 o
TECHNICAL EVALUATION REPORT w+
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CONFORMANCE TO ITEM 4.5. ' 0F GENERIC LETTER 83 DUANE ARNOLD' ENRICO FERMI-2 '
HOPE CREEK' -
LASALLE COUNTY-1/-2' LIMERICK-1/-2
. MILLSTONE-V MONTICELLO NINE MILE POINT-1/-2 OYSTER CREEK
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F. G. Farmer 7
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Alan C. Udy l
Published April 1989 Idaho National Engineering Laboratory EG&G Idaho, Inc.
.1 Prepared t> the U.S. Nuclear Regulawry Commission Washington, D.C.
20555 Under DOE Contract No. DE-AC07-761001570 FIN Nos. D6001 and D6002 B&R 20-19-19-11-3 l
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SUMMARY
This EG&G Idaho, Inc. report provides a review of the submittals for some of the General Electric (GE)-supplied nuclear plants for conformance to Generic Letter 83-28, Item 4.5.2.
Tnis item deals with the on-line functional testing.of Reactor Trip System components.. Where special circumstances and proper justification exist, alterr.atives to on-line testing can be permitted. The report includes the following plants, all
.GE, and is in partial fulfillment of the following TAC Nos.:
plant Docket Number TAC Number Duane Arnold 50-331,
53979 Enrico Fermi-2 (OL) 50-341 N/A Hope Creek 50-354 61478 LaSalle County-1 50-373 53994 o
LaSalle County-2 50-374 53995 Limerick-1 50-352 56263 Limerick-2 (OL) 50-353 N/A Millstone-1 50-245 53999 Monticello 50-263 54001 Nine Mile Point-1 50-220 54002 Nine Mile Point-2 (OL) 50-410 N/A Oyster Creek 50-219 54008 l
B&R 20-19-19-11-3 FIN Nos. D6001 and D6002 11
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I PREFACE This report is provided 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 conducted for the U. S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Engineering and System Technology, by EG&G Idaho, Inc., Regulatory and Technical Assistance Unit.
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. CONTENTS
SUMMARY
11 PREFACE.............................................................
iii 1.
INTRODUCTION.....................................................
1 2.
REVIEW REQUIREMENTS.............................................
2 3.
GROUP REVIEW RESULTS.
4 4.
REVIEW RESULTS FOR DUANE AF.NOLD.......,...........................
5 4.1 Evaluation................................................
5 4.2 Conclusion...............................................
5 5.
REVIEW RESULTS FOR ENRICO FERMI-2...............................
6 5.1 Evaluation................................................
~6 5.2 Conclusion................................................
6 6.
REVI EW RESULTS FOR HOPE CREEK...................................
7 6.1 Evaluation................................................
7 6.2 Conclusion................................................
7 7.
REVIEW RESULTS FOR LASALLE COUNTY-1/-2..........................
8 7.1 Evaluation................................................
8 7.2 Conclusion.................................................
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8.
REVIEW RESULTS FOR LIMERICK-1/-2................................
9 8.1 Evaluation................................................
9 8.2 Conclusion................................................
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9.
REVI EW RESU LTS FOR MI LLSTONE-1..................................
10 9.1 Evaluation................................................
10 9.2 Conclusion................................................
10 10.
REVIEW RESULTS FOR MONTICELLO...................................
11 10.1 Evaluation................................................
11 iv
l 10.2 Conclusion................................................
11
- 11. REVIEW RESULTS FOR NINE MILE POINT-1............................
12 1
s 11.1 Evaluation..........................-.....................
12 11.2 Conclusion................................................
12 12.
REVIEW RESULTS FOR NINE MILE POINT-2............................
13 12.1 Evaluetion................................................
13 12.2 Conclusion................................................
13 13.
REVIEW RESULTS FOR OYSTER CREEK.......'..........................
14 13.1 Evaluation................................................
14 13.2 Conclusion........*........................................
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- 14. GROUP CONCLUSION................................................
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15.
REFERENCES......................................................
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CONFORMANCE 'N3 ITEM 4.5.2 0F GENERIC LETTER 83-28 DUANE ARNOLD ENRICO FERMI-2 HOPE CREEK LASALLE COUNTY-1/-2 LIMERICK-1/-2 MILLSTONE-1 MONTICELLO NINE MILE POINT-1/-2 OYSTER CREEK 1.
INTRODUCTION 1
On July 8, 1983, Generic Letter 83-28 was issued by D. G. Eisenhut, Director of the Division of Licensing, Office of Nuclear Reactor Regulation, to all. licensees of operating reactors, applicants for operating licenses, and holders of construction permits. This. letter included required actions based on generic implications of the Salem ATWS events.
These requirements have been published in Volume 2 of NUREG-1000,
" Generic Implications of ATWS Events at the Salem Nuclear Power Plant."2 This report documents the EG&G Idaho, Inc., review of the submittals of some of the GE plants including Duane Arnold, Enrico Fermi-2, Hope Creek, LaSalle-1/-2, Limerick-1/-2, Millstone-1, Monticello, Nine Mile Point-1/~2, and Oyster Creek for conformance to Item 4.5.2 of Generic Letter 83-28.
The submittals from the licensees utilized in these evaluations are referenced in Section 15 of this report.
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I 2.
REVIEW REQUIREMENTS Item 4.5.2 (Reactor Trip System Reliability - System Functional Testing - On-Line Testing) requires licensees and applicants with plants not currently designed to permit en-line testing to justify not making modifications to permit such testing. Alternatives to on-line testing will be considered where special circumstances exist and where the objtetive of high reliability can be met in another way.
Item 4.5.2 may be interdependent with Item 4.5.3 when there is.a need to justify not performing on-line testing because of the peculiarities of a particular design.
All portions of the Reactor Trip System (RTS) that do n'ot have on-line testing capability will be reviewed under the guidelines for this item.
Maintenance and testing of the Reactor Trip Breakers (RTBs) are also excluded from this review, as they are evaluated under Item 4.2.
This review of the licensee / applicant submittals will:
1.
Confirm that the licensee / applicant has identified those portions of the Reactor Trip System that are not on-line testable.
If the entire Reactor Trip System is verified to be on-line testable, no further review is required.
2.
Evaluate modifications proposed by licensees / applicants to permit
.on-line testing against the existing criteria for the design of the protection systems for the plant being modified.
3.
Evaluate proposed alternatives to on-line testing of the Reactor Trip System for acceptability based on the following:
a.
The licensee / applicant submittal substantiates the impracticality of the modifications necessary to permit on-line testing, and 1
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- b.. High Reactor Tr4 3ystem availability (t.,* parable to that which -
would be pc, 'ble with on-line testing) is achieved in another j
way. Any uch proposed alternative must be described in detail i
sufficient to permit an independent evaluation of the basis and analysis provided in lieu of performing on-line testing. Methods that may be used to demonstrate that the objective of high reliability has been met may include the following:
1.
Demon'stration by systematic analysis that testing at shutdown intervals provides essentially equivalent reliability to that obtained by on-line testing at shorter intervals.
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- 11. Demonstration that reliability equivalent to that obtained by on-line testing is accomplished by additional redundant-and diverse components or by other features.
iii. Development of a maintenance program based on early replacement of critical components that compensates for the lack of on-line testing.
Such a program would require analytjgal justification supported by test data.
iv. Development of a test program that compensates for the lack of on-line testing, e. g., one which uses trend analysis and
. identification of safety margins for critical parameters of safety-related components.
Such a program would rec re analytical justification supported by test data.
4.
Verify the capability to perform independent on-line testing of j
the reactor trip system breaker undervoltage and shunt trip attachments on CE plants.
Information from licensees and applicants with CE plants will be reviewed to verify that they require independent on-line testing of the reactor trip breaker undervoltage and shunt trip attachments.
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3.
GROUP REVIEW RESULTS
-The relevant submittals from each of the GE rer: tor plants were reviewed to det9rmine compliance with Item 4.5.2.
first, the submittals from each plant were reviewed to establish that Ite.n 4.5.2 was specifically addressed. Second, the submittals were evaluated to determine the extent to which each of the GE plants complies with the staff guidelines for Item 4.E.2.
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4.
REVIEW RESULTS FOR DUANE ARNOLD 4.1 Evaluation The Iowa Electric Light and Power Company, the licensee for Duane Arnold, provided their responses to Item-4.5.2 of the generic letter on February 29, 1984 and April 30, 1985.
In those responses, the licensee states that the Reactor Protection System (RPS) design complies with all applicable regulatory requirements for the reactor trip system, and includes a summary description of the on-line functional testing that is performed on the RPS and the testing intervals used.
The licensee's response states that' Arnold does not perform on-line testing of the backup scram valves because testing during operation would cause a plant scram; and the valves will be independently tested during each refueling outage.
4.2 Conclusion In as much as the Reactor Protection System includes those components necessary to trip the. reactor, we find that the licensees stated position on Item 4.5.2 of the generic letter, including their justification for not performing periodic on-line testing of the backup scram valves, meets the requirements of Item 4.5.2 of Generic Letter 83-28 and is, we believe, acceptable.
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REVIEW RESULTS FOR ENRICO FERMI-2 5.1 Evaluation Detroit Edison, the applicant for Fermi-2, provided their response to Ites 4.5.2 of the generic letter on April 30, 1985.
In that response, the applicant affirms that Fermi-2 is designed to permit on-line testing of the Reactor Trip System.
The applicant's response states that Fermi-2 does not perform on-line testing of the backup scram logic and valves because testing during operation would cause a plant scram; and the backup scram logic and valves are independently tested during each refueling outage.
5.2 Conclusion In as much as the Reactor Protection System includes those components necessary to trip the reactor, we find that the applicant's stated position on Item 4.5.2 of the generic letter, including their justification for not performing periodic on-line testing of the backup scram logic and valves, meets the requirements.of Item 4.5.2 of Generic Letter 83-28 and is, we believe, acceptable.
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REVIEW RESULTS FOR HOPE CREEK 6.1 Evaluation The Public Service Electric and Gas Company, the licensee for Hope Creek, responded to Item 4.5.2 of the generic letter on March 30, 1984.
In that response, the licensee confirms that Hope Creek will perform periodic on-line testing of the Reactor Trip System.
The licensee's response states that Hope Creek does not perform on-line testing of the backup scram valves because testing during operation would cause a plant scram; and. the valves are independently tested during each refueling outage.
6.2 Conclusion In as much as the Reactor Protection System includes those components necessary to trip the reactor, we find that the licensee's stated position on Item 4.5.2 of the generic letter, including their justification for not.
performing periodic on-line testing of the backup scram valves, meets the requirements of Item 4.5.2 of Generic Letter 83-28 and is', we believe, j
acceptable.
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REVIEW RESULTS FOR LASALLE COUNTY-1/-2 7.1 Evaluation Commonwealth Edison, the licensee for LaSalle-1/-2, responded to Item 4.5.2 of the generic letter on November 5,1983 and June 1,1984.
In those responses, the licensee confirms that on-line functional testing of the Reactor Trip System 1s allowed during normal plant operation.
The licensee's response states that LaSalle-1/-2 does not perform on-line testing of the reactor mode switch or the backup scram logic and solenoid valves because' testing during operation would cause a plant scram; and the backup scram logic and solenoid' valves are independently tested during each refueling outage.
7.2 Conclusion In as much as tFo Reactor Protection System includes those components necessary to trip the reactor, we find that the licensee's stated position on Item 4.5.2 of the generic letter, including their justification for not performing periodic on-line testing of the backup scram logic and solenoid valves, meets the requirements of Item 4.5.2 of Generic Letter 83-28 and is, we believe, acceptable.
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9$f' Av,,R, I t W %as#7 IMAGE EVALUATION a e ,04 / ((@%' /(g //o//"q[f' $f/* TEST TARGET (MT-3) \\\\\\//// p7pp \\\\\\\\ p' //// <s 1.0 lt a E y W g =? 1 .: a: llllN k' l,l l.8 1.25 1.4 l 1.6 4 150mm 4 6" gl%y//, / fb se+f4;u ///// ,,y, y gj fs i 8. REVIEW RESULTS FOR LIMERICK-1/-2 8.1 Evaluation ~ \\ Philadelphia Electric Company, the licensee for Limerick-1 and applicant for Limerick-2, responded to Item 4.5.2 of the gener.r letter on 1 May 8, 1984. In that response, the licensee / applicant confirms that the Limerick Reactor Protection System design permits on-line testing of the RPS. 1 The licensee's/ applicant's response states that Limerick does not perform on-line testing of the backup scram valves because testing during operation would cause a plant scram; and the valves are independently tested during each refueling octage. 8.2 Conclusion In as much as the Reacter Protection System includes these components necessary to trip the reactor, we find that the licensee's/cpplicant's stated position on Item 4.5.2 of the generic letter, including their justification for not performing periodic on-line testing of the backup scram valves, meets the requirements of Item 4.5.2 of Generic Letter 83-28 and is, we believe, acceptable. l 9 l l l 9. REVIEW RESULTS FOR MILLSTONE-1 9.1 Evaluation Northeast Utilities, the licensee for M111 stone-1, responded to Item 4.5.2 of the generic letter on November 8, 1983. In that response, the licensee states that the M111 stone-1 Reactor Trip System, with the exception of the backup scram valves, is designed to allow on-line testing, and that such tests are performed at the frequencies defined in the Technical Specifications. The normal backup scram valves and the ATWS backup scram valves are tested during refueling outages. These components cannot be tested during operation because they are the final control elements in the RTS and on-line testing of these valves would c'~se a scram. au 9.2 Conclusion The licensee has confirmed that the RTS components are tested on-line with the exception of the backup scram valves and the ATWS scram valves. The licensee is justified in not testing these components on-line and has committed to functionally test these components on a refueling basis. Based on these responses, we find the licensee's submittal ~s regarding Item 4.5.2 of Generic Letter 83-28 acceptable. i l l r ) 10 f 1 1 10. REVIEW RESULTS FOR MONTICELLO 10.1 Evaluation Northern States Power Company, the licensee for Monticelle, responded to Item 4.5.2 of the generic letter on November 14, 1983. In that response, the licensee affirms that on-line functional. testing of the reactor trip system is being performed at Monticello, with the exception of the backup scram valves. The licensee states that functional testing of the backup scram valves is performed as part of the plant prestart testing, which is performed prior to restart from each refueling outage. 10.2 Conclusions In as much as the Reactor Protection System includes those components necessary to trip the reactor, we find that the licensee's/ applicant's stated position on Item 4.5.2 of the generic letter, including their justification for not performing periodic on-line testing of the backup scram valves, meets the requirements of Item 4.5.2 of Generic Letter 33-28 and is, we believe, acceptable. 11 _ _ ___-______=__ - _ _ 11. REVIEW RESULTS FOR NINE MILE POINT-1 11.1 Evaluation The Niagara Mohawk Power Corporation, the licensee for Nine Mile Point-1, responded to the generic letter on November 8, 1983, July 31,1984, and December 31, 1984. The licensee's responses confirm that on-line function 61 testing of the Nine Mile Point I reactor trip system is ps.' formed on a regular basis. The licensee's response states that Nine Mile Point-1.does not perform on-line testing of the backup scram valves because testing during operation would cause a plant scram. In Reference 16, the licensee commits to testing those valves on a refueling basis. 11.2 Conclusion The licensee has confirmed that the RTS components are tested on-line with the exception of the backup scram valves. The licensee is justified in not testing these components on-line and has committed to functionally test these components on a refueling basis. Based on these responses, we ~ found the licensee's sdbmittals regarding Item 4.5.2 of Generic Letter 83-28 acceptable. em 12 12. REVIEW RESULTS FOR NINE MILE POINT-2 { 12.1 Evaluation The Niagara Mohawk Power Corporation, the applicant for Nine Mile Point-2, responded to the generic letter on April 10, 1984, December 20, 1985, and April 15, 1986. The applicant's responses affirms that Nine Mile Point-2 is designed to permit on-line functional testing of the Reactor Protection System, with the exception of the. backup scram valves. The applicant's response states that Nine Mile Point-2 will ps *' ?m functional testing of the backup scram valves during refueling out.,,s'. 12.2 Conclusion In as much as the Reactor Protection System includes those components necessary to trip the reactor, we find that the applicant's stated position; on Item 4.5.2 of the generic letter, including their justification for not performing periodic on-line testing of the backup scram valves, meets the requirements of Item 4.5.2 of Generic Letter 83-28 and is, we believe, acceptable. m 13 4 ~ -13. REVIEW RESULTS FOR OYSTER CREEK '{ 13.1 Evaluation 4 GPU Nuclear Corporation, the licensee for Oyster Creek, responded to-the generic letter on November 14, 1983. The licensee's response confirms that, with the exception of the scram pilot valves and backup scram valves, on-line functional testing is currently being performed on the Oyster Creek Reactor Trip System.' The licensee's response states that Oyster Creek will provide justification for the adequacy of current functional tests of the scram pilot valves and backup scram valves. 13.2 Conclusion We find that the licensee's stated position on Item 4.5.2 of the generic letter is unacceptable, as the licensee has not provided justification for not performing periodic on-line testi g of the scram n pilot valves or the backup scram valves, and has not confirmed that the backup scram valves are tested on at least a refueling outage basis. W l 14 1
- 14. GROUP CONCLUSION We conclude that the licensee / applicant responses for the listed GE I
plants for Item 4.5.2 of Generic Letter 83-28 are acceptable, with the exceptions of the justification and confirmation needed from Oyster Creek. e h h i l 15 i 15. REFERENCES 1. Letter, NRC (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 1 Letter 83-28)," July 8,1983. 2. Generic" Implications of ATWS Events at the Salem Nuclear Power Plant NUREG-1009, Volume 1, April 1983; Volume 2, July 1983. 3. Letter, Iowa Electric Light and Power Company (R. W. McGaughy) to NRC (Harold R. Denton), February 29, 1984. 4. Letter, Iowa Electric Light and Power Company (R. W. McGaughy).to NRC (Harold R. Denton), April 30, 1985. 5. Letter, Detroit Edison (W. H. Jens) to NRC, " Detroit Edison Response to Generic Letter 83-28," November 3,1983. 6. Letter, Public Service Electric and Gas Company (R. L. Mitt 1) to NRC, " Response to. Generic Letter 83-28," March 30,1984. 7. Letter, Commonwealth Edison Company ( P. L. Barnes) to NRC (Harold R. Denton), November 5, 1983. 8. Letter, Commonwealth Edison Company (P. L. Barnes) to NRC (Harold R. Denton), June 1, 1984. 9. Letter, Philadelphia Electric Company (V. S. Boyer) to NRC (D. G. Eisenhut), " Additional Response to Generic Letto 83-28," May 8, 1984. 10. Letter, Northeast Utilities (W. G. Counsil) to NRC (D. G. Eisenhut), " Response to GeneHe Letter 83-28, Generic Implications of Salem ATWS Events," November 8, 1983. 11. Letter, Northeast Utilities (W.G. Counsil) to NRC (D.G. Eisenhut), " Response to Generic Letter 83-28, Generic Implications of Salem ATWS Events," March 16, 1984, B11053. 12. Letter, Northern States Power Company (D. Musolf) to NRC, " Generic Implications of Salem ATWS Events (Generic Letter 83-28)," November 14, 1983. 13. Letter, Niagara Mohawk Power Corporation (T. E. Lempges) to NRC, November 8, 1983. 14. Letter, Niagara Mohawk Power Corporation (G. K. Rhode) to NRC, April 10, 1984. 16 .o p 1 [ ~ July 31,1984.
- 15. Letter, Niagara Mohawk Power Corporation (T. E. Lempges) to NRC, 16' Letter, Niagara Mohawk Power Corporation (C.V. Mangan) to NRC (D. B.- V,assallo), December 31, 1984.
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- 17. Letter).,_ Niagara Mohawk Power Corporation (T. E. Lempges) to NRC (E. G.
Adensam), December 20, 1985. 18. Letter, Niagara Mohawk Power Corporation (T. E. Lempges) to NRC, July 31,1984. 19. Letter, GPU Nuclear Corporation (P. B. Fiedler) to NRC, "0yster Creek Generating Station," November 14, 1983. 4 = 17