ML20066G727
| ML20066G727 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, 05000356  |
| Issue date: | 11/15/1982 |
| From: | Crouse R TOLEDO EDISON CO. |
| To: | Eisenhut D, Eusenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.1, TASK-1.A.1.3, TASK-2.E.1.1, TASK-2.E.1.2, TASK-2.E.4.1, TASK-2.E.4.2, TASK-2.K.2.08, TASK-2.K.2.13, TASK-2.K.3.03, TASK-2.K.3.12, TASK-TM 873, GL-82-16, GL-82-17, GL-82-23, TAC-49481, TAC-49482, TAC-49730, NUDOCS 8211230022 | |
| Download: ML20066G727 (5) | |
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TOLEDO
%mm EDISON Ro4Ano P. CnousE Docket No. 50-346 vao,n.on
%K4f License No. NPF-3
"' S' emaai Serial No. 873 November 15, 1982 Director of Nuclear Reactor Regulation Attention:
Mr. Darrell G. Eisenhut, Director Division of Licensing United States Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Eisenhut:
This is in response to your letter dated September 20, 1982 (Log No. 1093) relating to NURP.G-0737 Technical Specifications (Generic Letter No. 82-16).
Toledo Edison's response as applicable to Davis-Besse Nuclear Power Station Unit 1 is provided in the attachment to this letter.
Very truly yours,
,4 P C m a /tt m RPC:SCJ:RLW Attachment l
cc: DB-1 Resident NRC Inspector l
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THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO, OH!O 43652
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8211230022 82111
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Docket No. 50-346 License No. NPF-3 Serial No. 873 November 15, 1982 ATTACHMENT Toledo Edison Davis-Besse Nuclear Power Station Unit 1 Response To Enclosure 1 of NRC Letter of September 20, 1982 Item (1)
(I.A.1.1.3 STA Training)
Our July 2,1980 letter provided model Technical Specifications for TMI lessons learned Category "A" items.
Included were Technical Specifications that specified the qualifications, training and on-duty requirements for the Shift Technical Advisors (STA). STA training requirements are under considera-tion by the Commission. Further guidance will be provided pending a decision on the requirements by the Commission.
Response Toledo Edison's letter of September 16, 1980 (Serial No. 650) responded to your letter of July 2, 1980. The Technical Speci-fications for qualifications, training and on-duty requirements for Shif t Technical Advisor (STA) have now been incorporated in our Technical Specifications, therefore no additional change is required.
Item (2)
(I.A.I.3 Limit Overtime)
On June 15, 1982 we transmitted to licensees of operating plants a revised version of the Commission's Policy Statement on nuclear power plant staff working hours.
In the same letter we also transmitted revised pages of NUREG-0737 (Item I. A.I.3).
The administrative section of the technical specifications should be revised to require procedures that follow the policy statement guidelines. An acceptable specification would be "the amount of overtime worked by plant staff members performing safety-related functions must be limited in accordance with the NRC Policy Statement on working hours (Generic Letter No. 82-12),"
or following the model Technical Specifications in Enclosure 2.
Response Toledo Edison has revised administrative procedures to limit overtime. This item is considered implemented and closed by NRC Inspection Report 81-10 dated June 19, 1982 (Log No. 1-519).
Item (3)
(II.E.1.1 Short Term Auxiliary Feedwater System Evaluation)
The objective of this item is to improve the reliability and performance of the auxiliary feedwater (AFW) system. Technical Specifications depend on the results of the licensee's evaluation and the staff review, and are being developed separately for each plant. The limiting conditions of operation (LCO's) and surveillance requirements for the AFW system should be similar to other safety-related systems.
1 Docket No. 50-346 License No. NPF-3 Serial No. 873 November 15, 1982 Response As identified in Toledo Edison's response dated September 14, 1982 (Serial No. 857) to the NRC letter dated August 3, 1982 (Log No. 1052) concerning Item II.E.1.1, no changes to the Technical Specifications are required.
Item (4)
(II.E.1.2 Safety Grade AFW System Initiation and Flow Indication The AFW system automatic initiation system was to have been control grade by June 1,1980 and safety grade by July 1,1981; the AFW system flow indication was to have been control grade by January 1,1980 and safety grade by July 1,1981. The control grade requirement was part of the short term lessons learned activities, and model Technical Specifications were included with our July 2,1980 letter. These Technical Specifications are considered adequate as Technical Specifications for the safety grade requirement.
Response The AFW system automatic initiation system has always been safety grade at Davis-Besse Unit 1.
In our response to your letter dated July 2, 1980, we proposed Technical Specification changes (now issued) to address the AFW system flow indicator.
Therefore, no additional Technical Specification changes are required.
Item (5)
(II.E.4.1 Dedicated Hydrogen Penetrations)
Plants that use external recombiners or purge systems for post-accident combustible gas control of the containment atmos-phere should provide containment penetrations dediented to that service.
In satisfying this item, some plants may have to add some additional piping and valves.
If so, these valves should be subjected to the requirements of Appendix J, and the Technical Specifications should be modified accordingly.
Response Davis-Besse utilizes the hydrogen dilution system for post accident combustible gas control. Dedicated Hyorogen Penetra-tions for this system are part of the original design of Davis-Besse as referenced in our December 30, 1980 letter (Serial No. 670).
No Technical Specification changes are required.
Item (6)
(II.E.4.2.5 Containment Pressure Setpoint)
The containment pressure setpoint that initiates containraent isolation must be reduced to the minimum compatible with normal operating conditions. Most plants provided Jut.tification for not changing their setpoint and we approved their justification by separate correspondence. The remaining plants must submit a change to the Technical Specifications with the lower containment pressure setpoint and provide justification if this setpoint is more than 1 psi above maximum expected containment pressure during normal operation.
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Docket No. 50-346 License No. NPF-3 Serial No. 873 November 15, 1982 Response The containment isolation pressure setpoint for Davis-Besse Unit I meets the requirements stated in NUREG-0737 per NRC Safety Evaluation Report dated April 14, 1982 (Log No. 961). Therefore no action is required.
Item (7)
(II.E.4.2.6 Containment Purge Valve)
Model Technical Specifications are oeing sent separately to each plant as part of the overall containment purge review. These l
Technical Specifications include the requirement that the containment purge valves be locked closed except for safety-related activities, verified closed at least every 31 days, and be subjected to leakage rate limits.
Response Toledo Edison will review the model Technical Specifications i
when issued.
Item (8)
(II.E.4.2.7 Radiation Signal On Purge Valves)
The containment purge valves must close promptly to reduce the J
amount of radiation released outside containment following a release of radioactive materials to containment. Technical Specifications should include the requirement that at least one radiation monitor that automatically closes the purge valves l
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upon sensing high radiation in the containment atmosphere be operable at all times except cold shutdowns and refueling outages.
If not operable, either the plant should be proceeding to cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the purge valves should be closed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Model Techotcal Specifications are provided in Enclosure 2 in Standard Pechnical Specifications format for those plants that are using safety-grade components to satisfy the requirement.
I Response The radiation signal on purge valves is part of the Safety Features Actuation System (SFAS) original design and no additional Technical Specifications are required.
(Item 9)
(II.K.2.8 Upgrade B&W AFW System)
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Acceptance criteria for proposed Technical Specification are identical to that described in (2) and (3) above.
(Toledo ddison Note: NRC Statement should reference items-(3) and (4) above).
I Response See response to items (3) and (4) above.
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Toledo Edison Note: Item 10 is missing from Enclosure 1 of NRC Generic Letter 82-16, i
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e Docket No. 50-346 License No. NPF-3 Serial No. 873 November 15, 1982 Item (11) (II.K.2.13 B&W Thermal - Mechanical Report)
Licensees of B&W operating reactors are required to submit by January 1,1981 an analysis of the thermal-mechanical conditions in the reactor vessel during recovery from small breaks with an extended loss of all feedwater. Technical Specifications, if required, will be determined following staff review.
Response Toledo Edison letter dated March 3, 1982 (Serial No. 790) responded to the thermal-mechanical conditions in the reactor vessel during recovery from a Small Break Loss of Coolant Accident. No further action is required.
Item (12) (II.K.3.3 Reporting SV and RV Failures and Challenges)
NUREG-0660 stated that safety and relief valve failures be reported promptly and challenges be reported annually. The sections of your Technical Specifications that discuss reporting requirements should be accordingly changed; model Technical Specifications are given in Enclosure 2.
Note that an acceptable j
alternative would be to report challenges monthly.
l Response Toledo Edison committed to report challenges of reactor coolant j
safety and relief valves in its monthly operating reports in our i
letter dated June 26, 1980 (Serial No. 624). This will be j
back-reported to commence April 1, 1980. Failures of such valves shall be reported pursuant to Section 6.9.1.7 of the i
" Davis-Besse Nuclear Power Station Unit 1, Technical Specifi-cation, Appendix A".
No Technical Specification change is required.
l' Item (13) (II.K.3.12 Anticipatory Trip on Turbine Trip) l Licensees with Westinghouse-designed operating plants have l
confirmed that their plants have an anticipatory reactor trip upon turbine trip. Many of these plants already have this trip in their Technical Specifications.
For those that do not, the anticipatory trip should be added to the Technical Specifications.
Model Technical Specifications are included in Enclosure 2 in l
the format of Standard Technical Specifications.
l Response In our letter dated November 25, 1981 (Serial No. 757), we proposed revised Technical Specifications for the Safety Grade Anticipatory Reactor Trip System (ARTS). Toledo Edison will l
resubmit a revised application for license amendment by March 1, 1983.
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