ML20056G794

From kanterella
Jump to navigation Jump to search
Forwards Technical Basis for Crystal River Unit 3 Plant Specific Procedure to Resolve NRC Generic Ltr 87-02 & Florida Power Corp Plant Specific Procedure for Seismic Verification of Nuclear Plant Equipment, Rev 0
ML20056G794
Person / Time
Site: Crystal River Duke Energy icon.png
Issue date: 08/27/1993
From: Beard P
FLORIDA POWER CORP.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20056G795 List:
References
REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR 3F0893-12, 3F893-12, GL-87-02, GL-87-2, GL-88-20, NUDOCS 9309070175
Download: ML20056G794 (3)


Text

i Florida Power CORPORATION N U $0S August 27, 1993 3F0803-12 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555

Subject:

Generic Letter 87-02, Supplement 1 Verification of Seismic Adequacy of Equipment in Older Operating Nuclear Plants

References:

1. FPC to NRC letter, 3F0793-06, dated July 1, 1993
2. Generic Letter 88-20, Supplement 4, dated June 28, 1991 (3N0691-23)
3. FPC to NP,C letter, 3F0992-02, dated September 4, 1992

Dear Sir:

In Reference 1, Florida Power Corporation (FPC) committed to submit the Plant Specific Procedure for Seismic Verification of Nuclear Plant Equipment for Crystal River 3 (CR-3). The procedure is provided as attachment 1. This procedure will be used to resolve Unresolved Safety Issue A-46 for CR-3. This procedure will also be used to resolve the seismic portions of the Individual Plant Examination for External Events (IPEEE) as requested by the NRC in Reference 2. Also provided as attachment 2 is the Technical Basis for the Crystal River 3 Plant Specific Procedure to Resolve NRC Generic Letter 87-02.

This document contains the detailed, point by point basis and justification for the departures from the Generic Implementation Procedure (GIP) produced by the Seismic Qualification Utility Group and approved by the NRC Staff.

Instances where FPC is unable to meet the single failure requirement of the GIP and the Plant Specific Procedure because of plant design are documented in atticnment 3 to this letter.

7!....iI) i l

CRYSTAL RIVER NUCLEAR PLANT: Powerline Road . P.O. Box 219

  • Crystal River, Florida 34423 0219 + (904) 7954486 9309070175 N DR ADOCK930827 05000302 }j p

,.'_ PDR b

U. S. Nuclear Regulatory Commission 3F0893-12 Page 2 FPC continues cn schedule to complete the walkdown in 1995 and submit the final report before December 31, 1995 as committed in Reference 2. Work on the walkdown packages will begin shortly and some of the walkdowns are expected to be conducted during 1994.

Sincerely, M

. if. Beard, Jr.

Senior Vice President Nuclear Operations PMB/AEF:ff Attachments ve: Regional Administrator, Region II NRR Project Manager Senior Resident Inspector Director, Nuclear Licensing, FP&L Co.

ATTACHMENT 1 TO 3F0893-12 DEVICES AND FUNCTIONS SUBJECT TO SINGLE FAILURES In addition to the departures from the GIP documented in the Technical Basis document, the design features of CR-3 will not allow FPC, in every case, to meet the guidance in the GIP and the Plant Specific Procedure related to the requiremeat to assume a single failure. These departures are as follows:

1. The atmospheric dump valves are not sufficiently large to be considered redundant. Given a failure (closed) of either valve, HOT SHUTDOWN cannot be achieved within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
2. The main steam line isolation function is not redundant unless credit is taken for the turbine throttle valves. The turbine throttle valves are not st.ismically qualifiable using the Plant Specific Procedure methodology.

Given a failure (open) of any of these main steam line isolation valves and j the concurrent failure of the turbine throttle valves, the affected steam generator would have emergency feedwater isolated and boil dry. This would have the same affect on the achievable cooldown rate as Item 1 above.

3. To cool down to HOT SHUTDOWN requires depressurization below 600 psig. This requires closure of the core flood tank discharge isolation valves (CFV-5 and CFV-6) to prevent discharge of the tanks. Failure (open) of CFV-5, CFV-6, or loss of the Engineered Safeguards Motor Control Center 3AB would prevent this.

l l

! . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ _____ _ _