ML20211B320
| ML20211B320 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 05/30/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20211B318 | List: |
| References | |
| GL-83-28, NUDOCS 8606110590 | |
| Download: ML20211B320 (3) | |
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ENCLOSURE SAFETY EVALUATION REPORT GENERIC LETTER 83-28, ITEMS 3.1.3 AND 3.2.3 POST-MAINTENANCE TESTING (RTS COMPONENTS, ALL OTHER SAFETY-RELATED COMPONENTS)
EDWIN I. HATCH NUCLEAR PLANT, UNITS 1,2 DOCKET NO.: 50-321/366 INTRODUCTION AND
SUMMARY
1 Generic Letter 83-28 describes intermediate term actions to be taken by licensees and applicants to address the generic issues raised by the two ATWS events that occurred at Unit 1 of Salem Nuclear Power Plant.
This report is an evaluation of the responses submitted by Georgia Power Company, the licensee for the Edwin I. Hatch Nuclear Plant, Units 1, 2 for Items 3.1.3 and 3.2.3 of the Generic Letter. The actual documents reviewed as part of this evaluation are listed in the references at the end of this report.
The requirements for these two items are identical with the exception that Item 3.1.3 applies these requirements to the Reactor Trip System components and Item 3.2.3 applies them to all other safety-related components.
Because of this similarity, the responses to both items were evaluated together.
REQUIREMENT Licensees and applicants shall identify, if applicable, any post-maintenance test requirements in existing Technical Specifications which can be demonstrated to degrade rather than enhance safety. Appropriate changes to these test re-quirements, with supporting justification, shall be submitted for staff approval.
8606110590 860530 PDR ADOCK 05000321 P
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. EVALUATION i
The licensee for Hatch, Units 1, 2 responded to these requirements with 2
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submittals dated November 7, 1983, February 29, 1984, and January 31, 1986,
The two early submittals indicated that the licensee was confusing the require-ments of Items 3.1.3 and 3.2.3 with those of Item 4.5.3.
A request for addi-4 tional information clarifying the requirements of Items 3.1.3 and 3.2.3 was sent to the licensee and the licensee responded with the submittal of January 31, 1986. The licensee stated in these submittals that there were no post-maintenance testing requirements in Technical Specifications for either reactor trip system or other safety-related components which degraded safety.
CONCLUSION Based on the licensee's statement that no post-naintenance test requirements were found in Technical Specifications that degraded safety, we find the licensee's responses acceptable for Items 3.1.3 and 3.2.3 of Generic Letter 83-28.
l REFERENCES i
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NRC Letter, D. G. Eisenhut to all Licensees of Operating Reactors, Applicants for Operating License, and Holders of Construction Permits, 1
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" Required Actions Based on Generic Implications of Salem ATWS Events (Generic Letter 83-28)," July 8, 1983.
. REFERENCES (CONT.)
2.
Georgia Power Company letter, L. T. Gucwa to J. F. Stolz, Chief, Operating Reactors Branch No. 4, Division of Licensing, NRC, "NRC Dockets 50-321, 50-366, Operating Licenses DPR-57, NPF-5, Edwin I. Hatch Nuclear Plants Units 1, 2, Status Report on Salem Generic Requirements," November 7, 1983, NED-83-546.
3.
Georgia Power Company letter, L. T. Gucwa to J. F. Stolz, Chief, Operating Reactors Branch No. 4, Division of Licensing, NRC, "NRC Dockets 50-321, 50-366, Operating Licenses DPR-57, NPF-5, Edwin I. Hatch Nuclear Plant Units 1 and 2, Respnnse to Generic Letter 83-28, Salem Requirements," February 29, 1984, NED-84-054.
4.
NRC letter, J. F. Stolz, Chief, Operating Reactors Branch No. 4, Division of Licensing, to J. T. Beckham, Georgia Power Company, April 4, 1985.
5.
Georgia Power Company letter, L. T. Gucwa to D. Muller, Project Director, BWR Project Directorate No. 2, NRC, January 31, 1986.
t EGG-EA-6912 CONFORMANCE TO GENERIC LETTER 83-28 ITEMS 3.1.3 and 3.2.3 BROWNS FERRY UNIT NOS. 1, 2, and 3, BRUNSWICK UNIT NOS. I and 2, COOPER, DUANE ARNOLD, FITZPATRICK, HATCH UNIT NOS. I and 2, PEACH BOTTOM UNIT NOS. 2 and 3, VERMONT YANKEE R. O. Haroldsen Published February 1986 EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission o
Washington, D.C.
20555 Under DOE Contra-t No. DE-AC07-76ID01570 o
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ABSTRACT This EG&G Idaho, Inc., report provides a review of the submittals for several nuclear plants for conformance to Generic Letter 83-28, items 3.1.3 and 3.2.3.
The specific plants reviewed were selected as a group because of similarity in type and applicability of the review items. The group includes the following plants:
Plant Docket Number TAC Numbers Browns Ferry 1 50-259 52982, 53820 Browns Ferry 2 50-260 52983, 53821 Browns Ferry 3 50-296 52984, 53822 Brunswick 1 50-325 52985, 53823 Brunswick 2 50-324 52986, 53824 Cooper 50-298 52991, 53829 Duane Arnold 50-331 52997, 53835 FitzPatrick 50-333 53000, 53838 Hatch 1 50-321 53006, 53844 Hatch 2 50-366 53007, 53845 Peach Bottom 2 50-277 53027, 53866 Peach Bottom 3 50-278 53028, 53867 Vermont Yankee 50-271 53055, 53894 i
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FOREWORD This report is supplied as part of the program for evaluating licensee / applicant conformance to Generic Letter 83-28 "Reauired 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 Regulation, Division of PWR Licensing-A by EG&G Idaho, Inc., NRR and 1&E Support Branch.
The U.S. Nuclear Regulatory Commission funded the work under the authorization, B&R 20-19-19-11-3, FIN No. D6001.
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CONTENTS A B S TR A CT..............................................................
11 FOREWORD..............................................................
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INTRODUCTION.....................................................
1 2.
REVIEW REQUIREMENTS..............................................
3 3.
GROUP R E V I EW R E SULTS.............................................
4 4.
REVIEW RESULTS FOR BROWNS FERRY UNIT NOS. 1, 2, and 3............
6 4.1 Evaluation..................................................
6 4.2 Conclusion..................................................
6 5.
REVIEW RESULTS FOR BRUNSWICK UNIT NOS. 1 and 2...................
7 5.1 Evaluation..................................................
7 5.2 Conclusion..................................................
7 6.
REVIEW RESULTS FOR COOPER........................................
8 6.1 Evaluation..................................................
8 6.2 Conclusion..................................................
8 7.
REVIEW RESULTS FOR DUANE ARNOLD..................................
9 7.1 Evaluation..................................................
9 7.2 Conclusion..................................................
9 8.
REVIEW RESULTS FOR FITZPATRICK...................................
10 8.1 Evaluation..................................................
10 8.2 Conclusion..................................................
10 9.
REVIEW RESULTS FOR HATCH UNIT NOS. I and 2.......................
11 9.1 Evaluation..................................................
11 9.2 Conclusion.................................................
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- 10. REVIEW RESULTS FOR PEACH BOTTOM UNIT NOS. 2 and 3................
13 10.1 Evaluation...............................
13 10.2 Conclusion.................................................
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- 11. REVIEW RESULTS FOR VERMONT YANKEE...............................
14 11.1 Evaluation.................................................
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11.2 Conclusion.................................................
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- 12. GROUP CONCLUSION.................................................
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- 13. REFERENCES.......................................................
16 TABLES TABLE 1...............................................................
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CONFORMANCE TO GENERIC LETTER 83-28 ITEMS 3.1.3 and 3.2.3 FOR BROWNS FERRY UNIT NOS. 1, 2, and 3, BRUNSWICK UNIT NOS. 1 and 2 COOPER, DUANE ARNOLD, FITZPATRICK, HATCH UNIT NOS. I and 2 PEACH BOTTOM UNIT N05. 2 and 3 VERMONT YANKEE 1.
INTRODUCTION I
On July 8, 1983,' Generic Letter No. 83-28 was issued by D. G.
Eisenhut, Director of the Division of Licensing, Nuclear Reactor Regulation, to all licensees of operating reactors, applicants for operating licensees, and holders of construction permits. This letter included reouired 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 from Browns Ferry Unit Nos. 1, 2 and 3. Brunswick Unit Nos. 1 and 2, Cooper, Duane Arnold, FitzPatrick, Hatch Unit Nos.1 and 2, Peach Bottom Unit Nos. 2 and 3, and Vermont Yankee for conformance to items 3.1.3 and 3.2.3 of Generic Letter 83-28.
The submittals from the licensees utilized in these evaluations are referenced in Section 13 of this recort.
These review results are applicable to the group of nuclear plants previously identified because of their similarity. These plaats are
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similar in the following respects:
1.
Tney are operating General Electric boiling water reactors.
2.
They utilize the Mark 1 Containment and Pressure Suppression Systems.
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They are 1967 (Model 4) reactors.
4.
They utilize two Class lE power system trains.
4 5.
They use relay logic in the Reactor Trip Systems.
An item of concern identified for any one of these plants is assumed to be potentially significant for all of the plants in the group.
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REVIEW REQUIREMENTS Item 3.1.3 (Post-Maintenance Testing of Reactor Trip System (RTS)
Components) reouires licensees and applicants to identify, if applicable, any post-maintenance test reauirements for the RTS in existing technical specifications which can be demonstrated to degrade rather than enhance safety.
Item 3.2.3 extends this same requirement to include all other safety-related components. Any proposed technical specification changes resulting from this action shall receive a pre-implementation revicw by NRC.
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GROUP REVIEW RESULTS The relevant submittals from each of the reactor plants previously named were reviewed to determine compliance with items 3.1.3 and 3.2.3 of the generic letter. First, the submittals from each plant were reviewed to determine if these two items were specifically addressed. Second, the submittals were checked to determine if there was any post-maintenance test recuirements specified by the technical specifications that were suspected to degrade rather than enhance safety. Last, the submittals were reviewed for evidence of special conditions or other significant information relating to the two items of concern. The results of this review are summarized in Table 1.
All of the plants responded specifically to the two items. With the exception of Browns Ferry, Unit Nos. 1, 2 and 3, all of the licensees indicated that there had been no items identified in the technical specifications relating to post-maintenance testing that could be demonstrated to degrade rather than enhance safety.
The submittal from FitzPatrick makes reference to an expected proposal from the boiling water reactor owners' group (BWROG) to make some changes to the technical specifications. These changes are expected to affect surveillance testing freauency, out-of-service intervals for testing and post-maintenance testing. This is a reference to a BWROG response to Generic Letter 83-28, Item 4.5.3.
3 The BWROG is presently addressing Generic Letter 83-28, Item 4.5.3 i
which may result in proposed changes to the technical' specification recuirements for surveillance testing freauency and out-of-service intervals for testing. The primary concern of Item 4.5.3 is the f
surveillance testing intervals.
Items 3.l.3 and 3.2.3 are specifically i
directed at post-maintenance test reauirements. These concerns are essentially independent. However, the evaluations of these concerns are coordinated so that any correlation between these concerns will be adeauately considered. Since no specific proposal to change the technical saecifications has been submitted, there is no identifiable need at this time for correlating the reviews of item 4.5.3 with this review.
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TABLE 1 Were Items 3.1.3 and 3.2.3 Addressed in the Licensee / Applicant
Response
Plants Submittal Findings Acceptable Canments Browns Ferry Yes No conclusion No Evaluating 1,2&3 work discontinued.
Brunswick Yes No Tech. Spec. Items Yes 1&2 1dentified that degrade safety Cooper Yes No post-maintenance Yes test reauirements that degrade safety Duane Arnold Yes No Tech. Spec. Items Yes identified that degrade safety FitzPatrick Yes No Tech. Spec. Items Yes R.eference identified that made degrade safety to BWROG for possible proposal to change the Tech. Specs.
Hatch I & 2 Yes No post-maintenance Yes test requirements that degrade safety Peach Bottom Yes No Tech. Spec. Items Yes Response is 2&3 identified that very brief.
degrade safety Vermont Yes No Tech. Spec. Items Yes They will Yankee identified that continue to degrade safety review.
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REVIEW RESULTS FOR BROWNS FERRY UNIT NOS. 1, 2, and 3 4.1 Evaluation Tennessee Valley Authority, the licensee for Browns Ferry Unit Nos. 1, 2, and 3, provided a response to items 3.1.3 and 3.2.3 of Generic Letter 83-28 on November 7, 1983.4 Within the response, the licensee's evaluation for item 3.1.3 is that they are not aware of any program currently being done by GE to specifically determine if current testing recuirements degrade the reliability of the Reactor Trip System eauipment, no. oo they have any information indicating that existing testing degrades rather than enhances safety. The licensee's response to item 3.2.3 is that it is not their philosophy to propose changes in technical specifications which are perceived to degrade rather than enhance safety.
If such should be determined, appropriate changes would be submitted along with supporting justification.
4.2 Conclusion The licensee's responses to items 3.1.3 and 3.2.3 of Generic Letter 83-28, are considered inadeauate by the staff. These items will be held open pending receipt and assessment of 1nformation responsive to these Concerns.
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REVIEW RESULTS FOR BRUNSWICK UNIT NOS. I and ?
5.1 Evaluation Carolina Power and Light Company, the licensee for Brunswick Unit Nos. I and 2, provided a response to items 3.l.3 and 3.2.3 of Generic Letter 83-28 on November 7, 1983.5 Within the response, the licensee's evaluation for items 3.1.3 and 3.2.3 is that, following a review of the Brunswick standard technical specifications, no testing reouirements which tended to degrade the safety of either the Reactor Trip System components or that for the other safety-related components were identified. The scoDe of the review included only those recuirements which were clearly defined i
in technical specifications as post-maintenance tests. The licensee committed to continue the review for items 3.1.3 and 1'.2.3.
5.2 Conclusion Based on the licensee's statement that they have reviewed their technical specification reauirements to identify any post-maintenance testing which could be demonstrated to degrade rather than enhance safety and found none that degraded safety, we find the licensee's response acceptable.
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REVIEW RESULTS FOR COOPER 6.1 Evaluation The Nebraska Public Power District, the licensee for the Cooper Nuclear Station, provided a response to items 3.1.3 and 3.2.3 of Generic Letter 83-28 on November 4, 1983.6 However, that response did not contain results of their review as reouired by the generic letter. A 7
subsecuent response on July 30, 1985 contains more specific information.
In this response the licensee states that a search was made of the Technical Specifications that revealed no post-maintenance test reautrements that would clearly degrade rather than enhance safety.
6.2 Conclusion The licensee's responses to items 3.l.3 and 3.2.3 meet the reauirements of Generic Letter 83-28 and are acceptable.
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REVIEW RESULTS FOR DUANE ARNOLD 7.1 Evaluation Iowa Electric Light and Power Company, the licensee for the Duane Arnold Energy Center, provided an initial response to items 3.1.3 and 3.2.3 of Generic Letter 83-28 on November 7, 1983.8 This response was later superceded by a response on February 29, 1984.9 Within the response, the licensee's evaluation for items 3.1.3 and 3.2.3 is that the review of the existing post-maintenance test recuirements in the technical specifications shows that there are no testing reauirements found that are preceived to degrade the plant safety for the re8Ctor trip system and safety-related components.
7.2 -Conclusion Based on the licensee's statement that no recuirements have been found in the technical specification test recuirements that degrade rather than enhance safety, the staff concludes that the licensee's response is adecuate and acceptable.
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REVIEW RESULTS FOR FITZPATRICK 8.1 Eva16ation The New York Power Authority, the licensee for the James A. FitzPatrick Nuclear Power Plant, provided an initial response to items 3.1.3 and 3.2.3 of Generic Letter 83-28 on November 9, 1983.10 This response was finalized on June 29, 1984.II Within 'he response, the t
licensee states that no post-maintenance test requirements have been identified in their technical specifications that degrade safety.
8.2 Conciusion Based on tne licensee's statement that no post-maintenance test reauthements have been identified in the technical specifications that degrade safety, the staff concludes that.the licensee's response is adeouate and acceptatile.
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REVIEW RESULTS FOR HATCH UNIT NOS. 1 and 2 9.1 Evaluation Georgia Power Company, the licensee for the Edwin I. Hatch Nuclear Plant, Unit Nos. I and 2, provided an initial response to items 3.1.3 and 3.2.3 of Generic Letter 83-28 on November 7, 1983.12 At the time of this response a complete review of maintenance procedures and technical specifications had not been conducted. The evaluation of existing reauirements in the technical specifications which can be demonstrated to degrade rather than enhance safety is expected (by the licensee) to be addressed generically by the BWR Owners Group. On February 29, 1984,13 the licensee provided an update on their conformance and further information related to plans and schedules for any needed improvements.
For item 3.1.3, the licensee specifies that the BWR Owners Group Technical Specification Improvement Comittee is expected to eventually address all aspects of plant technical specifications. The program is expected to be complete within two years. For item 3.2.3, the licensee stated that testing requirements for the diesel generators were degrading their reliability. The licensee has proposed technical specification that were scheduled for NRC submittal in March 1984.
The licensee evidently confused the reauirements of items 3.1.3 and 3.2.3 for an evaluation of post-maintenance testing with the reauirements of item 4.5.3 for evaluating intervals for on-line functional testing and with the reauirements of Generic Letter 84 15 regarding surveillance end functional testing of diesel generators.
Items 3.1.3 and 3.2.3 of Generic Letter 83-28 are restricted to post-maintenance testing of Reactor Trip System Components. The staff is not aware of any effort by the BWR Owners Group to address items 3.1.3 and 3.2.3.
A request for additional information more specific to the concerns of items 3.1.3 and 3.2.3 was transmitted to the licensee on April 4, 1985.I4 15 A response dated January 31, 1986 provided the reouested information.
In che response, the licensee confirmed that a review of the Technical 11
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I Specifications nad been conducted and that no post-maintenance test recuirements were identified for either reactor trip systems components or safety-related components that degrade safety.
9.2 Conclusion The licensee's responses to items 3.1.3 and 3.2.3 meet the reauirements of Generic Letter 83-28 and are acceptable.
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- 10. REVIEW RESULTS FOR PEACH BOTTOM UNIT NOS. 2 and 3 10.1 Evaluation Philadelphia Electric Company, the licensee for the Peach Bottom 1
Atomic Power Station, Unit Nos. 2 and 3, provioed a response to items 3.1.3 and 3.2.3 of Generic Letter 83-28 on November 4, 1983.16 Within the response, the licensee's evaluation for items 3.1.3 and 3.2.3 is that there are no existing testing reauirements of the technical specifications identified which degrade safety in the reactor protection system or other safety-related components.
10.2 Conclusion Based on the licensee's statement that no existing test requirements of the technical specifications have been identified that degrade safety in the reactor protection system or other safety related components, the staff concludes that the licensee's response is adeauate and acceptable.
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- 11. REVIEW RESULTS FOR VERMONT YANKEE 11.1 Evaluation The Vermont Yankee Nuclear Power Corporation, the licensee for Vermont
- /ankee, provided a response to items 3.1.3 and 3.2.3 of Generic Letter 83-28 on November 7, 1983.17 Within the response, the licensee's evaluation is that, at the present time, there are no proposed technical specification changes recuired for items 3.1.3 and 3.2.3.
The licensee has committed to continuously review technical specification recuirements and identify the need for revision in the future. A complete technical specification review to clarify wording and provide consistency with all technical specification sections has been planned. Any identified deficiencies would be the subject of future proposed technical specification changes.
11.2 Conclusion Based on the licensee's statement that no items have been identified in the technical specifications that degrade saf.ety, the staff concludes that the licensee's response is adequate and acceptable.
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- 12. GROUP CONCLUSION With the exception of Browns Ferry Unit Nos.1, 2 and 3, the staff concludes that the licensee's responses are adequate and acceptable.
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- 13. REFERENCES 1.
NRC Letter, D. G. Eisenhut to all Licensees of Operating Reactors, Applicants for Operating License, and Holders of Construction Permits, "Reauired Actions Based on Generic implications of Salem ATWS Events (Generic Letter 83-28)", July 8, 1983.
2.
Generic Implications of ATWS Events at the Salem Nuclear Power Plant, NUREG-1000, Volume 1, April 1983; Volume 2, July 1983.
3.
BWR Owners' Group Responses to NRC Generic Letter 83-28 Item 4.5.3, General Electric Company Proprietary Information, NEDC-30844 January 1985.
4.
Tennessee Valley Authority letter, L. M. Mills to H. R. Denton, Director, Office of Nuclear Reactor Regulation, NRC, " Response to Generic Letter 83-28", November 7, 1983.
5.
Carolina Power and Light Company letter, P. W. Howe to D. G. Eisenhut, k
Director, Division of Licensing, NRC, " Generic Implications of Salem ATWS Events for Brunswick Steam Electric Plant, Unit Nos. I and 2",
November 7, 1983.
6.
Nebraska Public Power District letter, L. G. Kund to D. G. Eisenhut, Director, Division of Licensing, NRC, " Response to Generic Letter 83-28, 'Recuired Actions Based on Generic Implications of Salem ATWS Events.' NRC Docket No. 50-298, DPR-46.", November 4, 1983.
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Nebraska Public Power District letter, J. M. Pilant to D. B. Vassalo, r
Nuclear Reactor Regulation Division of Licensing, NRC Generic Letter 83-38. Item 3.1.3 " Post Maintenance Testing" (Reactor Trip System Components) NRC Doc'ket No. 50-298, DPR-46, July 30, 1985.
8.
Iowa Electric Light and Power Company letter, R. W. McGaugby to H. R. Denton, Director, Office of Nuclear Reactor Regulation, NRC, "Duane Arnold Energy Center, Docket No. 50-331, Op License No:
DPR-49, Generic Letter 83-28: 'Reouired Actions Based on Generic Implications of Salem ATWS Event'," November 7,1983, NG-83-3824.
9.
Iowa Electric Light and Power Company letter, R. W. McGaughy to H. R. Denton, Director, Office of Nuclear Reactor Regulation, NRC, "Duane Arnold Energy Center, Docket No 50-331, Op License No:
DPR-49, Generic Letter 83-28:
'Reauired Actions Based on Generic Implications of Salem ATWS Event'," February 29, 1984, NG-84-0825.
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- 10. New York Power Authority letter, J. P. Bayne to D. B. Vassallo, Chief, Operating Reactors Branch No. 2, Division of Licensing, NRC,
" James A. FitzPatrick Nuclear Power Plant, Docket No. 50-333, Response to Generic Implications of Salem ATWS Events (Generic Letter 83-28)",
November 9, 1983, JPN-83-92.
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- 11. New York Power Authority letter, J. P. Bayne to D. B. Vassallo, Chief, Operating Reactors Branch No. 2 Division of Licensing, NRC,
" James A. FitzPatrick Nuclear Power Plant, Docket No. 50-333, Generic Letter 83-28, Reouired Actions Based on Generic Implications of Salem ATWS Events", June 29, 1984, JPN-84-42.
- 12. Georgia Power Company letter, L. T. Gucwa to J. F. Stolz, Chief, Operating Reactors Branch No. 4 Division of Licensing, NRC, "NRC Dockets 50-321, 50-366, Operating Licenses DPR-57, NPF-5, Edwin I Hatch Nuclear Plants Units 1, 2, Status Report on Salem Generic Reauirements", November 7, 1983, NED-83-546.
- 13. Georgia Power Company letter, L. T. Gucwa to J. F. Stolz, Chief, Operating Reactors Branch No. 4, Division of Licensing, NRC, "NRC Dockets 50-321, 50-366, Operating Licenses DPR-57, NPF-5, Edwin I. Hatch Nuclear Plant Units 1 and 2,' Response to Generic Letter 83-28, Salem Reouirements," February 29, 1984, NED-84-054.
- 14. NRC letter, J. F. Stolz, Chief, Operating Reactors Branch No. 4, Division of Licensing, to J. T. Beckham, Georgia Power Company, April 4, 1985.
- 15. Georgia Power Company letter, L. T. Gucwa to D. Miller, Project Director, BWR Project Directorate No.
2., NRC, January 31, 1986.
- 16. Philadelphia Electric Company letter, S. L. Daltroff to D. G. Eisenhut, Director, Division of Licensing, NRC, " Generic Letter 83-28, Recuired Actions Based on Generic Implications of Salem ATWS Events", November 3, 1983.
- 17. Vermont Yankee Nuclear Power Corporation letter, W. P. Murphy to D. B. Vassallo, Chief Office of Nuclear Reactor Regulations, NRC,
" Generic Letter 83-28, Generic Implications of the Salem ATWS Events",
November 7, 1983, 2.C.2.1, FVY 83 117.
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., &,,r..e, a This EG8G Idaho, Inc. report provides a review of the submittals for several nuclear plants for conformance to Generic Letter 83-28, Items 3.1.3 and 3.2.3.
The specific plants reviewed were selected as a group because of similarity in type and applicability of the review items. The group includes the following plants:
Browns Ferry, Units 1,2 and 3 Fitzpatrick Brunswick, Units 1 and 2 Hatch, Units 1 and 2 Cooper Peach Bottom, Units 2 and 3 Duane Arnold Vermont Yankee i
Revision 2 of this report reflects new information for the Hatch Unit Nos.1 and 2.
Review work has been discontinued on Browns Ferry Unit Nos. 1, 2 and 3.
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