ML20043F160

Provides Supplemental Info on Reactor Bldg Flooding,Per Util 900517 Ltr Describing Resolution Plan Being Pursued.Util Has Limited Vol of Water Contributed by Borated Water Storage Tank & Sodium Hydroxide Tank to Flood Level

ML20043F160
Person / Time
Site:	Crystal River
Issue date:	06/04/1990
From:	Boldt G FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
3F0690-04, 3F690-4, NUDOCS 9006140204
Download: ML20043F160 (3)

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.'C O R P 0 H A T 6 0 N H June 4, 1990 3F0690-04 l U.S. Nuclear Regulatory Commission Attention:

Document Control Desk 'I Washington, D.C. 20555

Subject:

Crystal River Unit-3 l Docket No. 50-302 '

Operating License-No. DPR-72 Supplemental Information on Reactor Building Flooding

Dear-Sir:

Florida - Power Corporation's (FPC) letter dated May 17, 1990 described the L resolution plan that is being pursued for the Reactor Building (RB) flooding

. issue. FPC is continuing with those activities to completion before CR-3 I restarts' from this refueling outage. The approach which FPC has taken to resolve 1 l

the RB flooding issue is to limit the volume of water contributed by the Borated '

Water Storage Tank (BWST) and the Sodium Hydroxide (Na0H) Tank .to the ' flood level. The implementation of this approach has identified two other issues which-have an impact on Crystal River Unit 3 (CR-3). Those issues and the actions planned by FPC are discussed below.

FPC's evaluation of.the calculations supporting LOCA mitigation identified the 1 H  : possibility of a wider range of pH conditions than previously analyzed. This issue is ag_t adversely affected by the change in initiation of the recirculation flow path which resolves the-flood level issue. These result from various small

" break loss of coolant accidents (SBLOCA) creating different BWST and Na0H Tank l drawdown rates than were assumed for large break LOCAs (LBLOCA). The drawdown rates of liquid from the BWST and the NaOH Tank are functions of the hydraulic conditions for each tank. A LBLOCA at CR-3 will actuate the Reactor Building

Spray (BS). System at a high RB pressure of 30 psig. The CR-3 drawdown analysis assumes that the BWST and the Na0H Tank drawdown at the flow rates necessary to mitigate a LBLOCA. Depending upon the break size, a SBLOCA may not actuate the BS System until some time after the break occurs. This delay in actuation of 9006140204 900604 DR ADOCK 05000302 PDC ,(g 3 POST OFFICE BOX 219 + CRYSTAL RIVER, FLORIDA 32629-o219 + (904) 563-2943 ,

A Florida Progress Company

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June 4, 1990 3F0690-04 Page 2.

the BS System could create a difference in the relationship of the hydraulic heads on the BWST and Na0H Tank. Consequently, the drawdown rates could be different from those assumed in the LBLOCA. The resultant spray pH as a function of time is therefore not completely defined. This lack of clearly defined spray pH range is being evaluated by FPC under the appropriate reporting requirements.

FPC has discussed this situation with the B&W Nuclear Services Company and the B&W Owners Group. The lack of consideration of SBLOCA pH calculations could be a generic problem. Conversely, this situation may not be explicitly part of ,

associated licensing or design bases. The present plant configuration provides j adequate protection for LBLOCA pH control. l l

One ; solution to long term pH control with either a SBLOCA or LBLOCA as the I initiating event is the use of tri-sodium phosphate (TSP) stored in baskets j located in the RB sump. Dilution of the TSP by the borated water from all sources will establish proper pH balance. The Na0H Tank and its contents would no longer be necessary to perform an accident mitigating function. FPC is I evaluating whether to install such TSP baskets.

The second issue discovered during calculational verification efforts is that-l the Control Room Habitability Dose is adversely effected by this change in RB flood volume if overly conservative failure postulations are not also changed.

The doses prescribed by 10 CFR 50, Appendix A, GDC 19-Control Room, are 5 rem whole body, or its equivalent to any part of the body. Although CR-3 was not reviewed against the Standard Review Plan (SRP), FPC considered SRP 6.4, " Control Room Habitability System," and SRP 15.6.5, Appendix B, " Radiological Consequences of a Design Basis Loss-of-Coolant - Accident: Leakage from Engineered Safety Feature Components Outside Containment," as guidance in developing the Control Room Habitability Study submitted by FPC's letter dated June 30, 1987. SRP 6.4 suggests a thyroid dose limit of 30 rem to assure that the GDC'19 limit is met.

The 30-day control room thyroid dose given in the Control Room Habitability Study submitted in FPC's letter dated June 30, 1987 is 26.5 rem. The Control Room l Habitability Study postulated a gross failure of a passive component which causes L a 50 gpm leak for 30 minutes at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The dose contribution frora this passive failure is directly related to the radionuclide concentration in the sump fluids. This is increased by reducing the amount of non radioactive fluids (BWST and NaOH tank) discharged to sump producing an estimated thyroid dose of l 38.6 rem without credit for iodine removal by the filter system. The SRP L requires such a postulation if the plant does not provide an Engineered Safety Features '(ESF) filtration system. This assumption was overly conservative for CR-3. CR-3 does have a filtration system associated with the areas containing

-the ESF systems and such passive failures have not been considered as part of o the CR-3 licensing basis. FPC chose to not take any credit for the filtration l- system in previous analyses because acceptable doses ~ could be demonstrated l without it. The revised main control room dose analysis continues to postulate an operational leakage from the ESF systems of 4510 cc/hr. This-leakage is twice the Technical Specification limit and is the value used in the Control Room

! Habitability Study. This analysis produces a 30-day control room thyroid dose of 22.5 rem without any credit for iodine removal by the filter system.

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1 June'4, 1990l 3F0690-04 Page 3 If you have any questions, please contact Mr. Kenneth R. Wilson, Manager, Nuclear Licensing at.(904) 563-4549.

. Sincerely,.  :

G.L. oldt, Vice President Nuclear Production ,

GLB:JWT xc: Regional Administrator, Region-II

-Senior Resident Inspector P

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