ML20210P278
| ML20210P278 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 02/05/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20210P221 | List: |
| References | |
| GL-83-28, NUDOCS 8702130349 | |
| Download: ML20210P278 (2) | |
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Enclosure 1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION DOCKET NOS. 50-266/301 POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 GENERIC LETTER 83-28, ITEM 4.5.2 REACTOR TRIP SYSTEM RELIABILITY, ON-LINE TESTING
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INTRODUCTION AND
SUMMARY
Generic Letter 83-28 was issued by NRC on July 8,1983, indiceting actions to be taken by applicants and licensees based on the generic implications of the k Salem ATWS events. Item 4.5 states a staff position which requires on-line '
functional testing of the reactor trip system, including independent testing of the diverse trip features of the reactor trip breakers, for all plants.
Item 4.5.2 requires applicants and licensees with plants not currently designed to permit this periodic on-line testing to justify not making modifications to permit such testing. By letters dated November 7, 1983, and April 11, 1986, the licensee, Wisconsin Electric Power Company, responded to the staff position regarding Item 4.5.2 of Generic Letter 83-28. Our review of this response finds it to be acceptable.
EVALUATION The licensee states that modifications have been made to the Point Beach plant to allow on-line testing of the Reactor Trip System, including independent testing of the undervoltage and shunt trip attachments. The staff has determined under review of Item 4.3 of Generic Letter 83-28, the design modifications to be acceptable.
CONCLUSION The staff finds that the Point Beach Nuclear Plant, Units 1,2, is designed to permit on-line functional testing of the reactor trip system, including independent testing of the diverse trip features of the reactor trip breakers. Thus, the applicant meets the staff position of Item 4.5.2 of Generic Letter 83-28.
Principal Contributors:
A. Toalston 8702130349 870205 PDR ADOCK 05000266 P PDR Date: FEB 0 51537
EGG-NTA-7459 TECHNICAL EVALUATION REPORT REACTOR TRIP SYSTEM RELIABILITY CONFORMANCE TO ITEM 4.5.2 OF GENERIC LETTER 83-28 JOSEPH M. FARLEY NUCLEAR PLANT UNITS 1 AND 2 R. E. GINNA NUCLEAR POWER PLANT HADDAM NECK PLANT INDIAN POINT UNIT NO. 2 INDIAN POINT 3 NUCLEAR POWER PLANT KEWAUNEE NUCLEAR POWER PLANT NORTH ANNA UNITS 1 AND 2 POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNITS 1 AND 2 F. G. Farmer Published January 9, 1987 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-AC07-76ID01570 FIN Nos. D6001 and D6002 Q O th\ 7QblilO-V\
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ABSTRACT This EGkG Idaho, Inc. report provides a review of the submittals for com2 of the Westinghouse (W) nuclear plants for conformance to Generic Lstter 83-28, Item 4.5.2. The report includes the following plants, all W;ctinghouse, and is in partial fulfillment of the following TAC Nos.:
Plant Docket Number TAC Number Joacph M. Farley Unit 1 50-348 53980 Jo2cph M. Farley Unit 2 50-364 53981 R. E. Ginna Nuclear Power Plant 50-244 53985 Hiddam Neck Plant 50-213 53987 Indian Point Unit 2 50-247 53990 Indian Point 3 Nuclear Power Plant 50-286 53991 Kcwnunee Nuclear Power Plant 50-305 53992 North Anna Unit 1 50-338 54003 North Anna Unit 2 50-339 54004 Point Beach Nuclear Plant Unit 1 * -
50-266
- 54013 - ,
Point Beach Nuclear Plant Unit 2 50-301 54014 Prcirie Island Nuclear Generating Plant Unit 1 50-282 54015 Prairie Island Nuclear Generating Plant Unit 2 50-306 54016 I
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FOREWORD This report is provided as part of the program for evaluating liccnsee/ applicant conformcnce to Generic Letter 83-28, " Required Actions Occcd on Generic Implications of Salem ATWS Events." This work is conducted for the U. S. Nuclear Regulatory Commission, Office of Nuclear Reactor R;gulation, Division of PWR Licensing-A by EG&G Idaho, Inc.
The U. S. Nuclear Regulatory Commission funded the work under the cuthorization, B&R 20-19-19-11-3, FIN Nos. D6001 and D6002.
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CONTENTS ABSTRACT ............................................................. ii FOREWORD ............................................................. iii
- 1. INTRODUCTION .................................................... 1
- 2. REVIEW REQUIREMENTS ............................................. 2
- 3. GROUP REVIEW RESULTS ............................................ 5
- 4. REVIEW RESULTS FOR JOSEPH M. FARLEY NUCLEAR PLANT UNITS 1 AND 2 ................................................... 6 4.1 Evaluation ................................................ 6 4.2 Conclusion ................................................ 6
- 5. REVIEW RESULTS FOR R. E. GINNA NUCLEAR POWER PLANT .............. 7 5.1 Evaluation ................................................ 7 5.2 Conclusion ................................................ 7
- 6. REVIEW RESULTS FDR HADDAM NECK PLANT ............................ B 6.1 Evaluation ................................................ B j 6.2 Conclusion ................................................ 8
- 7. REVIEW RESULTS FOR INDIAN POINT UNIT NO. 2 ...................... 9 7.1 Evaluation ................................................ 9 i
7.2 Conclusion ................................................ 9
- 0. REVIEW RESULTS FDR INDIAN POINT 3 NUCLEAR POWER PLANT ........... 10 l 8.1 Evaluation ................................................ 10 8.2 Conclusion ................................................ 10
- 9. REVIEW RESULTS FOR KEWAUNEE NUCLEAR POWER PLANT ................. 11 9.1 Evaluation ................................................ 11 9.2 Conclusion ................................................ 11 l 10. REVIEW RESULTS FOR NORTH ANNA UNITS 1 AND 2 ..................... 12 l
! 10.1 Evaluation ................................................ 12 iv i
10.2 Conclusion ................................................ 12
- 11. REVIEW RESULTS FOR POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 ..... 13 11.1 ' Evaluation ................................................ 13 11.2 Conclusion . . . . . . . . . . . . . . . . ................................ 13
- 12. REVIEW RESULTS FOR PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNITS 1 AND 2 ................................................... 14 12.1 Evaluation . . . . . . . . . . . . . . . . ................................ 14 12.2 Conclusion ................................................ 14
- 13. GROUP CONCLUSION ................................................ 15
- 14. REFERENCES ...................................................... 16 e
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TECHNICAL EVALUATION. REPORT- -_
REACTOR TRIP SYSTEM RELIABILITY CONFORMANCE TO ITEM 4.5.2 OF GENERIC LETTER 83-28 JOSEPH M. FARLEY NUCLEAR PLANT UNITS 1 AND 2 R. E. GINNA NUCLEAR POWER PLANT HADDAM NECK PLANT INDIAN POINT UNIT NO. 2 INDIAN POINT 3 NUCLEAR POWER PLANT KEWAUNEE NUCLEAR POWER PLANT NORTH ANNA UNITS 1 AND 2 POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNITS 1 AND 2
- 1. INTRODUCTION On July 8, 1983, Generic Letter 83-285 was issued by D. G. Eisenhut, DirGctor of the Division of Licensing, Office of Nuclear Reactor Regulation, to all licensees of operating reactors, applicants for operating licenses,-
cnd holders of construction permits. This letter included required actions bcOsd on generic implications of the Salem ATWS events. These require,ments hava been published in Volume 2 of NUREG-1000, " Generic Implications of ATWS Ev:nts at the Salem Nuclear Power Plant."*
This report documents the EG&G Idaho, Inc. review of the submittals of coma of the Westinghouse plants including Farley Units 1 and 2, Ginna, Hrddam Neck, Indian Point Units 2 and 3, Kewaunee, North Anna Units 1 and 2, Point Beach UniAs 1 and 2 and Prairie Island Units 1 and 2 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 14 of this report.
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- 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 on-line testing to justify not making modifications to permit s'uch testing. Alternatives to on-line testing will ba considered where special circumstances exist and where the objective of high reliability can be met in another way. Item 4.5.2 may be intsrdependent with Item 4.5.3 when there is a need to justify not parforming on-line testing because of the peculiarities of a particular danign.
All portions of the Reactor Trip System that do not have on-line tacting capability will be reviewed under the guidelines for this item.
Mnintenance and testing of the Reactor Trip Breakers are also excluded from thic review, as they are evaluated under Item 4.2. This review of the licsnsee/ 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 f or the plant being modified.
- 3. Evaluate proposed alternatives to on-line testing of the Reactor Trip System for acceptability based on the following:
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- a. The licensee / applicant submittal substantiates the impracticality of the modifications necessary to permit on-line testing, and
- b. High Reactor Trip System availability (comparable to that which would be possible with on-line testing) is achieved in another way. Any such proposed alternative must be described in detail 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 f ollowing:
- 1. Demonstration by systematic analysis that testing at shutdown intervals provides essentially equivalent reliability to that obtained by on-line testing at shorter intervals.
ii. Demonstration that reliability equivalent to that obtained by on-line testing is accomplished by additional redundant and diverse components or by other f eatures.
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 analytical Justification supported by test data.
iv. Development of a test program that compensates for the lack i of on-line testing, e. g., one which uses trand analysis and l identification of safety margins for critical parameters of safety-related components. Such a program would require i analytical Justification supported by test data.
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- 4. Verify the capability to perfarm independent on-line testing of 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 en-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 Westinghouse reactor plants w;ra reviewed to determine compliance with Item 4.5.2. First, the cubmittals from each plant were reviewed to establish that Item 4.5.2 was cp;cifically addressed. Second, the submittals were evaluated to determine tha extent to which each of the Westinghouse plants complies with the staff guidelines for Item 4.5.2.
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- 4. REVIEW RESULTS FOR JOSEPH M. FARLEY NUCLEAR PLANT UNITS 1 AND 2 4.1 Evaluation ,
Alabama Power Company, the license for Farley 1 and 2, provided their rceponse to Item 4.5.2 of the Generic Letter on November 4, 1983. In that rccponse, the licensee states that Farley performs on-line testing of the undsrvoltage and shunt trip attachments to the reactor trip breakers.
It is not clear from the licensee response that Farley performs on-line tccting of the reactor trip system; however, the licensee's Technical Cpscifications require monthly operability testing of all portions of the RTS, which implies this testing is performed on-line.
In a Safety Evaluation Report issued on September 20, 1983, the NRC confirmed that the shunt and undervoltage trips are independently tested en-line.
4.2 Conclusion We find that the licensee is required to periodically tes't all portions of the RTS on-line, and that the shunt and undervoltage trips are ind: pendently tested on-line, which meets the staff's position and is, we b311 eve, acceptable.
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- 5. REVIEW RESULTS FOR R. E..GINNA NUCLEAR POWER PLANT 5.1 Evaluation Rochester Gas and Electric, the licensee for Ginna, provided their rccponse to Item 4.5.2 of the Generic Letter on November 4, 1983. In that rceponse, the licensee states that Ginna will perform on-line testing of the R;cctor Trip System, including independent testing of the undervoltage and chunt trip attachments to the reactor trip breakers. The licensee further etctes that the on-line testing will be conducted on an annual or refueling bacis.
5.2 Conclusion We find the applicant's statement of the extent to which they will p;rform on-line testing of the RTS meets the staff position on Item 4.5.2 of tha Generic Letter and is, we believe, acceptable. The licensee's proposal to conduct such testing at annual or refueling intervals will be evaluated cc part of the resolution of Generic Letter 83-28, Item 4.5.3, and Generic Lctter 85-09.
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REVIEW RESULTS FOR HADDAH NECK PLANT"-
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6.1 Evaluation Northeast Utilities, the licensee for Haddam Neck, responded to Ittm 4.5.2 of the Generic Letter on November 5, 1983 and October 18, 1985.
In those responses, the licensee states that Haddam Neck was not designed to pcrmit performance of on-line testing of the Reactor Trip System. The H:ddam Neck design does not include bypass breakers, which are necessary to pcrmit on-line tripping of the reactor trip breakers without tripping the reactor. The licensee states that installation of the equipment required to modify the plant would be very difficult to accomplish because of the lack cf cpace in the switchgear room, and that the Haddam Neck design provides cicultaneous operation of both shunt and undervoltage trip attachments. The licensee also states that maintenance and inspection of the RTBs revealed no indications of failure to trip during the past 19 years of operation.
6.2 Conclusi on We find that the licensee's justification for not installing the modifications necessary to permit on-line testing of the Reactor Trip System ct Haddam Neck is acceptable, in view of the cost and difficulty of inctalling the necessary equipment and of the satisfactory history of rocctor trip reliability at the plant.
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- 7. REVIEW RESULTS FOR INDIAN POINT UNIT'NO. 2 7.1 Evaluation Consolidated Edison, the licensee for Indian Point 2, responded to Itca 4.5.2 of the Generic Letter on November 4, 1983. In that response, the '
liccnsee states that Indian Point 2 is designed to permit performance of cn-line testing of the Reactor Trip System, including independent on-line tcating of the shunt and undervoltage attachments.
7.2 Conclusion '
We find that the licensee's statement that they perform on-line' testing i
cf the RTS, including independent on-line testing of the shuntiane J, y l,
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,? . - o und:rvoltage attachments, meets the staf f position on Item 4.5.'2' of, the i .
Gen ric Letter and is, we believe~,' acceptable. .
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- 8. REVIEW RESULTS FOR INDIAN POINT 3 NUCLEAR POWER PLANT 8.1 Evaluation t' The New York Power Authority, the licensee f or Indian Point 3, i rc-hponhsd to Item 4.5.2 of the Generic Letter on November 7, 1983. In that rc:peque the licensee states that Indian Point 3 is designed to permit p;rformance of on-line testing of the Reactor Trip System and commits to on-line testing of the reactor protection system, including testing of the undcrvoltage and shunt trip attachments. However, it is not clear from the ,
rc ponse that the licensee can perf orm independent verification of the optrability of the diverse trip features.
8.2 poncl us i erj, i'
We find that the licensee's statement that they-will perform on-line tccting of the RTS meets the staf f position on Item 4.5.2 of the Generic Lctter and is, we believe, acceptable. However, the licensee should confirm rthOt the IrIdian Point
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,. 3 ,on-line testing includes the capability to perform ipdependant verification' of the operability of the diverse trip features.
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- 7 , 9. REVIEW RESULTS FOR KEWAUNEE NUCLEAR POWER PLANT
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1 1 Wisconsin Public Service Corporation, the licensee for Kewaunee,
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recponded to Item 4.5.2 of the Generic Letter on December 7, 1983, and April _13, 1984. In those responses, the licensee states that the Kewaunee plcnt performs on-line testing of the reactor trip breakers, specifically -
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including testing of the breaker undervoltage(trip attachment, and that Kcwnunee plans to implement a design change which will:. allow > independent
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verification of the operation of the shunt trip attachment.- .The licensee's Tcchnical Specificatiobs 3
require that on-line testing ofs al'1 portions the Racetor Trip System be per' formed on a periodic basis.
9.2 Conclusion
, 3 We find that the liciensee's responses did not clearly state that the entire Reactor Trip System i could be functionally tested. However, since the Tcchnical Specifications do require that a11' portions the RTS be p riodically tested, which implies that they are tested on-line, we believe thace requirements and the licensee *s commitment to perform independent cn-line testing of the undervoltage and shunt trip attachments meet the etcff's position on Generic Letter 83-28, Item 4.5.2, and are, we believe, ccccptable.
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- 10. REVIEW RESULTS FOR NORTH ANNA UNITS 1 AND 2 1 10.1 Evaluation l 1
Virginia Electric and Power Company, the licensee for North Anna, )
rocponded to Item 4.5.2 of the Generic Letter on November 4, 1983. In that rceponse, the licensee states that at North Anna, procedures are being rcvised to include independent testing of the diverse trip features, and '
that Item 4.5.2 of the Generic Letter is not applicable.
10.2 Conclusion We find the licensee's statement that Item 4.5.2 is not applicable to ha confirmation that North Anna performs on-line testing of the RTS, that thio confirmation meets the staff position on Item 4.5.2 of the Generic Lotter and is, we*believe, acceptable.
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- 11. REVIEW RESULTS FOR POINT BEACH.NUCLEGR_RLANT, UNITS 1;AND 2 l 11.1 Evaluation l Wisconsin Electric Power Company, licensee for Point Beach Units 1 cnd 2, responded to the Generic Letter on November 7, 1983. The licensee's rocponse states'that Point Beach will make modifications to permit them to pcrform on-line testing of the Reactor Trip System, including independent on-line testing of the shunt and undervoltage trip attachments.
11.2 Conclusion We find that the licensee's statement that they will make modifications to permit them to perf orm on-line testing of the RTS meets the staf f position on Item 4.5.2 of the Generic Letter and is, we believe, acceptable.
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- 12. REVIEW RESULTS FOR PRAIRIE _ ISLAND NUCLEAR GENERATING PLANT UNITS 1 AND 2 12.1 Evaluation Northern States Power Company, the licensee for Prairie Island Units 1 cnd 2 submitted a response ~to Item 4.5.2 of the Generic Letter on Novnmber 4, 1983. In that response, the licensee states that Point Beach is dasigned to permit on-line testing of the Reactor Trip System, that on-line tc= ting is performed monthly, and that the ability' to functionally tett the div rse trip features will be in place upon completion of the automatic chunt trip actuation modification. The licensee's July 6, 1984, letter daccribing the Prairie Island shunt trip attachment actuation modification confirms that shunt and undervoltage trip attachment testing is both on-line cnd independent.
12.2 Conclusion We find that the licensee's statement that Point Beach Units 1 and 2 cro designed to permit on-line testing of the RTS meets the staff position on Item 4.5.2 of the Generic Letter and is, we believe, acceptable.
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- 13. GROUP CONCLUSION __
We conclude that the licensee / applicant responses for the listed W;stinghouse plants for Item 4.5.2 of Generic Letter 83-28 are acceptable, with the exception that Indian Point 3 must provide the confirmation addressed in the plant specific review.
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- 14. REFERENCES s
- 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. Generic Imolications of ATWS Events at the Salem Nuclear Power Plant NUREG-1000, Volume 1, April 1983; Volume 2, July 1983.
- 3. Alabama Power letter to NRC, F. L. Clayton to Director, Nuclear Reactor i Regulation, " Response to Generic Letter 83-28," November 4, 1983.
i 4. Rochester Gas and Electric letter to NRC, John E. Maier to Director, Nuclear Reactor Regulation, NRC, November 4, 1983.
- 5. Northeast Utilities letter to NRC, W. G. Counsil to Darrel G. Eisenhut, Directcr, Division of Licensing, November 8, 1983.
,6. Northeast Utilities letter.to NRC, W. G. Counsil to Darrel G. Eisenhut, Director, Division of Licensing, October 18, 1983.
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- 7. Consolidated Edison Company letter to NRC, John D. O'Toole to D. G. Eisenhut, Director, Division of Licensing, November 4, 1983.
- 8. New York Power Authority letter to NRC, J. P. Bayne to D. G. Eisenhut, Director, Division of Licensing, " Required Actions Based on Generic Implications of Salem ATWS Events (Generic Letter 83-28, dated July 8, 1983)," November 7, 1983.
- 9. Wisconsin Public Service Corporation letter to NRC, C. W. Geisler to D. G. Eisenhut, Director, Division of Licensing, " Generic Implications of Salem ATWS Events (Generic Letter 83-28)," December 2, 1983.
- 10. Wisconsin Public Service Corporation letter to NRC, C. W. Geisler to D. G. Eisenhut, Director, Division of Licensing, " Generic Implications
- of Salem ATWS Events (Generic Letter 83-28)," April 13, 1984.
17." Virginia'Eiectric and Power Company letter to NRd,'W.
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L. Stewart, to Harold R. Denton, Director, Office of Nuclear Reactor Regulation,
" Response to Generic Letter 83-28," November 4, 1983.
- 12. Wisconsin Electric Power Company letter to NRC, C. W. Fay, to H. R. Penton, Director, Office of Nuclear Reactor Regulation, " Required Actions Based on Generic Implications of Salem ATWS Events,"
November 7, 1983.
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- 13. Northern States Power Company letter to NRC, D. M. Musolf, to Director, Office of Nuclear Reactor Regulation, " Generic Implications of Salem ATWS Events," November 4, 1983.
- 14. Northern States Power Company letter to NRC, D. M. Musolf, to Director, Of fice of Nuclear Reactor Regulation, " Description of Modifications for Aaiomatic Actuation of Scram Breaker Shunt Trip Attachment,"
July 6, 1984.
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