ML20080G154
| ML20080G154 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/16/1983 |
| From: | Westafer G FLORIDA POWER CORP. |
| To: | Stolz J 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.2.10, TASK-2.K.3.03, TASK-2.K.3.12, TASK-2.K.3.13, TASK-TM 3F-0983-06, 3F-983-6, GL-82-16, NUDOCS 8309200084 | |
| Download: ML20080G154 (5) | |
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.P..o..w..e...r September 16,1983 3F-0983-06 Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 Generic Letter 82-16 NUREG-0737 Technical Specifications
Dear Sir:
Attached is an update of the status of the items in Generic Letter 82-16. This supplements our letter of March 4,1983, on the same subject.
Sincerel,
/ f nY G. R. Westafer[ w Manager Nuclear Operations Licensing & Fuel Management Attachment RMB/TD/feb cc:
Mr. James P. O'Reilly Regional Administrator, Region 11 Office of Inspection & Enforcement U.S. Nuclear Regulatory Commission 101 Marietta Street N.W., Suite 2900 g
Atlanta, GA 30303 8309200084 830916
'l PDR ADOCK 05000302 P
PDR General Office 3201 Th rty-fourtn street souin. P O. Box 14042. St. Petersburg. Fiorda 33733 e 813-866-5151
D ATTACHMENT Item 1:
STA Training (I.A.I.l.3)
Our July 2,1980 letter provided model TSs for TMI lessons learned Category "A" items.
Included were TSs that specified the qualifications, training and on-duty requirements for the Shift Technical Advisors (STA). STA training requirements are under the consideration by the Commission.
Further guidance will be provided pending the decision on the requirements by the Commission.
Response 1:
The Commission has not yet provided further guidance; FPC will evaluate that guidance when received.
Item 2:
Limit Overtime (I.A.I.3)
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 administration 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 TSs in Enclosure 2.
Response 2:
The NRC, in its SER dated December 7,1981 approved the FPC method of implementing the NRC Policy Statement on working hours (i.e.,
limiting overtime) in the CR-3 Administrative Instructions. Therefore, FPC considers this item complete without the need to encorporate the limitations in the Crystal River Unit 3 (CR-3) Technical Specifications (TSs).
Item 3:
Short Term Auxiliary Feedwater System Evaluation (ll.E.1.1)
The objective of this item is to improve the reliability and performance of the auxiliary feedwater (AFW) system. TSs 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.
Response 3:
FPC submitted Technical Specifiction Change Request No. 82 (Letter #3F-0683-12,06/22/83) to verify, at least once per 31 days, that each valve in the EFW flow path is in its correct position In addition, we are discussing, with your staff, further Technical Specification changes to improve EFW reliability and performance.
Item 4:
Safety Grade AFW System Initiation and Flow Indication (II.E.1.2)
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 TSs were included with our July 2,1980 letter. These TSs are considered adequate as TSs for the safety grade requirement.
Response 4:
The safety-grade Emergency Feedwater Initiation and Control System is scheduled for installation in refuel outage V (scheduled for Spring 1985). Associated Technical Specification changes will be requested at that time.
Item 5:
Dedicated Hydrogen Penetrations (ll.E.4.1)
Plants that use external recombiners or purge systems for post-accident combustible gas control of the containment atmosphere should provide containment penetrations dedicated 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 3, and the TSs should be modified accordingly.
Response 5:
The CR-3 design includes redundant, single failure proof, dedicated hydrogen penetrations.
No additional piping and valves are required to satisfy this item.
Hence, no additional Technical Specification modifications are required.
Item 6:
Containment Pressure Setpoint (II.E.4.2.5)
The containment pressure setpoint that initiates containment isolation must be reduced to the minimum compatible with normal operating conditions. Most plants provided justification for not changing their setpoint and we approved their justification by separate correspondence.
The remaining plants must submit a change to the TSs 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.
Response 6:
The NRC provided a Safety Evaluation dated December 10, 1982:
that " concluded that the (present) containment pressure setpoint is acceptable".
Item 7:
Containment Purge Valve (II.E.4.2.6)
~
Model TSs are being sent separately to each plant as part of the overall containment purge review.
These TSs 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 7:
FPC has closed the purge and vent valves. We will verify every 31 days that the valves are locked closed when in Modes 1, 2, 3, and 4.
Hence, in a safety evaluation dated April 6,1983, the NRC concluded "that you (FPC) are in compliance with positions 6 and 7 of Item II.E.4.2".
Item 8:
Radiation Signal on Purge Valves (II.E.4.2.7)
The containment purge valves must close promptly to reduce the amount of radiation released outside containment following a release of radioactive materials to containment.
TSs should include the requirement that at least one radiation monitor that automatically closes the purge valves 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 TSs are provided in in Standard Technical Specifications format for those plants that are using safety-grade components to satisfy the requirement.
Response 8:
(See response 7.) Because the containment purge valves are closed (and verified), automatic closing of the valves on high radiation signal is not required.
Item 9:
Upgrade B&W AFW System (II.K.2.8)
Acceptance criteria fcr proposed TSs are identical to that described in 2 and 3 above.
Response 9:
See the appropriate responses 3 and 4 of this attachment.
Item 10: Anticipatory Trip - B&W Plants (II.K.2.10)
Licensees with B&W designed operating plants should examine the need to include anticipatory reactor trip (s) to their Technical Specifications.
Response 10: FPC submitted Technical Specification Change Request No. 82 (Letter #3F-0683-12,06/22/83) to include:
1.
Anticipatory Reactor Trip - Main Turbine; and 2.
Anticipatory Reactor Trip - Both Main Feedwater Pumps.
Item 11: B&W Thermal-Mechanical Report (II.K.2.13)
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. TSs, if required, will be determined following staff review.
Response 11: FPC submitted the required report on January 30,1931 and is waiting for the NRC staff to complete its review.
Item 12: Reporting SV and RV Failures and Challenges (ll.K.3.3)
NUREG-0660 stated that safety and relief valve failures be reported promptly and challenges be reported annually.
The sections of your TSs that discuss reporting requirements should be accordingly changed; model TSs are given in Enclosure 2.
Note that an acceptable alternative would be to report challenges monthly.
Response 12: FPC committed (letter dated July 21,1980) to report challenges to safety and relief valves in the CR-3 Annual Report. Safety and relief valve failure reporting is required by CR-3 Technical Specification 6.9.1.8.b.
Therefore, the requirements of this item have been implemented and a Technical Specification Change is not appropriate.
Item 13: Anticipatory Trip on Turbine Trip (ll.K.3.12)
Licensees with Westinghouse designed operating plants have confirmed that their plants have an anticipatory reactor trip upon turbine trip. Many of these plants already have this trip in their TSs. For those that do not, the anticipatory trip should be added to the TSs. Model TSs are included in Enclosure 2 in the format of Standard Technical Specifications.
Response 13: This item applies to Westinghouse designed plants. See reponse to item 10.
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