ML20214Q788

Application for Amends to Licenses DPR-53 & DPR-69,changing Tech Specs to Extend Allowed out-of-svc Period for Diesel Generator 12 in Order That Maint Can Be Performed W/O Requiring Unscheduled Cold Shutdown.Fee Paid

ML20214Q788
Person / Time
Site:	Calvert Cliffs
Issue date:	09/19/1986
From:	Tiernan J BALTIMORE GAS & ELECTRIC CO.
To:	Thadani A Office of Nuclear Reactor Regulation
Shared Package
ML20214Q790	List: ML20214Q788 ML20214Q792
References
NUDOCS 8609260314
Download: ML20214Q788 (6)

Text

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O BALTIMORE GAS AND ELECTRIC l CHARLES CENTER R O. BOX 1475 BALTIMORE, MARYLAND 21203 JOSEPH A.TIERNAN Vict PRESIDENT NUCLEAR ENERGY September 19,1986 U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D. C. 20555 ATTENTION: Mr. Ashok C. Thadani, Director PWR Project Directorate #8 Division of PWR Licensing-B

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Emergency Request for Amendment and. Waiver M

REFERENCE:

(a) Letter from Mr. E. 3. Butcher (NRC), to Mr. A. E. Lundvall, Jr.,

(BG&E), dated October 7,1985, NUREG-0737 Item II.K.3.25 (b) Letter from Mr. A. E. Lundvall, Jr. (BG&E), to Mr 3. R. Miller (NRC), dated November 30, 1984, Reactor Coolant Pump Seal Integrity Following Lass of Offsite Power Gentlemen:

The Baltimore Gas and Electric Company hereby requests an Amendment to its Operating License Nos. DPR-53 and DPR-69 for Calvert Cliffs Unit Nos.1 & 2, respectively, with the submittal of the proposed changes to the Technical Specifications.

PROPOSED CHANGE (BG&E FCR 86-138)

Change pages 3/4 8-1, 8-2, and 8-3 of the Unit I and 2 Technical Specifications as shown on the marked-up copies attached to this transmittal.

DISCUSSION Problem Description This change proposes to add an extension to the allowed out-of-service period for No.12 diesel generator (DG) such that maintenance can be effccted without requiring an unscheduled cold shutdown. yo adequately perform the repairs and allow for Jr contingencies, we would need 4m additional days of outage time for the swing diesel generator.

d**T (Sec. .Tnseet A)

We have discovered a crack in No.12 cgj,ipdgr I We estimated that J 4.r+es seven days will be required to repairlTt e UGMpi We are er on No.12 currently DG.Statement in Action 3.8.1.1.b, and by 0600, September 20, 1986, we must have a unit in cold shutdown. k 8609260314 860919 PDR ADOCK 05000317 Y 1 k

P PDR~ f gg

INSERT A i-A We have obtained a portable DG which is currently onsite. This DG is a 1000 kW,

! 480 volt Cummins Model KTA-50-G1 diesel generator. This DG would be electrically I connected to the vital 480 volt A.C. bus #14A for Unit No. 1 or #24A for Unit No. 2. Each bus could power one charging pump for its respective unit.

Plant personnel shall be trained through a walkthrough to electrically connect i the portable DG to the appropriate bus. Additionally, plant personnel shall r be trained through a walkthrough on the procedure to use the 69 kV SMECO off-site power circuit to power the emergency plant loads via the #23 bus.

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Mr. Ashok C. Thadani September 19,1986 Page 2 Therefore, more days of outage time are necessary to complete repairs. We need to g keep No.12 DG out-of-service until 0600,9, = du ^', ; E Septe4e, so, s.

Qualitative Analysis We have calculated the change in plant risk during a 10-day period in which No.12 diesel generator is unavailable. The results of this (PRA) analysis show that there is no significant change in overall plant core melt frequency during a 10-day outage. We believe that this conclusion is to be expected, based upon the,results of the PRA dpf evaluation performed for the 10-day outage, and hat the risk for mE more days is still JFinsignificant. If the DG is out-of-service ays, the probability of a core damage '

.JWevent during the periop is calculated to 8E When comparing this to the previous 10-day risk of 5x10- we feel there is no significant change in risk as a result of extending the outage six days.

On a loss of offsite power, No.12 DG starts, but only automatically aligns to a specific unit when a LOCA occurs and a Safety Injection Actuation Signal (SIAS) is initiated. If there is loss of offsite power (LOSP) without a LOCA, the operators me instructed to line-up No.12 DG to a specific unit. The Interim Reliability Evaluation Program (IREP),

which analyzed only a single unit at Calvert Ci unavailabilities that there is a probability of (P 5 tx) that 10 pfs, indicates via the calculated the operator align No.12 DG to Unit I and that there is an overall probability (P of2)1.2 thatx 10 yill fail No.12 DG will not be available to provide power to Unit 1 because it is needed on Unit 2. Additionally, period is about 7 x(P10 3).

jhe probability that it will fail to start and run during the dem Unavailability of No.12 DG = Pi+P2+P3 The net result is that the probability of No. I'2 DG failing to take a role in mitigpting the consequences of an offsite power event, as modeled in IREP, is about 2.4 x 10- (24% or 1 in 4). Clearly, because of this high unavailabi! ty assumed for the swing diesel generator, the' contribution of its failure to overall core melt frequency is very small.

This tends to reinforce our. confidence in the accuracy of the very small calculated change seen between our model results and those obtained from IREP.

Deterministic Analysis A deterministic analysis was performed assuming a total loss of offsite power (LOSP) while No.12 DG is out-of-service, and a failure of one of the two remaining DGs to start. In addition, we assumed a reactor coolant pump seal leak, as described in Reference (a), of 40 gpm in the unit opposite the unit with the remaining operable DG.

The analysis calculated the time to core uncovery. Our analysis of this hypothetical event showed that the decay heat load can be removed by natural circulation and that the Reactor Coolant System (RCS) pressure will not exceed the Pressurizer safety valve setpoint as long as Steam Generator inventory is maintained. The initial RCS inventory, excluding the Pressurizer, was conservatively assumed to shrink to a volume consistent with a temperature of 212 F and a pressure of 2500 psia. Calculations show there is adequate fluid volume without makeup flow to keep the core covered for at least 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />. We believe that in all probability, offsite AC power can be restored in most cases within four hours.

Mr. Ashok C. Thadani September 19,1986 j Page 3 i

, The four-hour restoration time assumed above is supported by several NRC and industry sources. In NUREG-1109, " Regulatory Analysis for the Resolution of Unresolved Safety Issue A-44, Station Blackout," the median restoration time for offsite power was given as about one-half hour, with 90% of the losses being restnred in three hours or less.

, NUREG-1032, " Evaluation of Station Blackout Accidents at Nuclear Power Plants,"

i includes graphs that display the probability of restoring offsite power following plant-

. centered and grid-related losses. Figures A.2a, A.26, and A.4 in NUREG-1032 show that

! greater than 90% of the losses can be restored within four hours. These data are ,

consistent with the conclusions of NSAC-103, EPRI's study M loss offsite power events.

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In addition, our grid system studies show that at least one of our 500 kv power lines will be restored within four hours following a loss of offsite puer. The system restoration study for PEPCO's electrical system was reviewed. In this study, SMECO is identified as a portion of PEPCO's load that would be restored early, within two hours if possible, i during their system restoration. Therefore, we feel a foe hour restoration time for offsite power is both realistic and reasonable.

i Offsite AC Power Sources a

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<dT We will maintain the 69 kv SMECO power circuit operable onsite dumg this)0 day maintenance period. We will energize the 13 kv bus No. 23 with SMECO power by shutting breaker No. 252-2301, isolate it from all other buses, and open the feder from i No. 23 bus to our warehouse. The warehouse will still be powered by SMECO. but through alternate feeders. The appropriate procedures exist and the operators poaass i adequate knowledge to perform such an electrical line-up.

l We will ensure that no unnecessary planned maintenance will be performed offsite on the l two 500 kv lines that would jeopardize the operability of the power sources. We also

acknowledge the proposed change in the Limiting Condition for Operation in the marked-

up Technical Specifications.

Reactor Coolant Pump Seal Leakage i Regarding NUREG-0737 Item II.K.3.25, " Reactor Coolant Pump Seal (RCP) Integrity Following Loss of Offsite Power," the industry's position has been that seal cooling water flow is not required to maintain RCP seal integrity. This position was reiterated in Reference (b). In fact, Reference (a) found BG&E in compliance with this NUREG-0737 Item.

I j Based on industry data, a two-hour loss of seal cooling could damage seal non-metallics i

that may initiate RCP seal leakage on the order of 0-10 gpm. However, there is no apparent reason to believe that a loss of RCP seal cooling to an idle pump can directly violate the reactor coolant pressure boundary integrity. Therefore, the seal cooling water supply is not required to assure seal integrity following a loss of offsite power. We l- do not believe that a complete loss of seal function will occur as the result of a loss of I offsite power.

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Mr. Ashok C. Thadani September 19,1986 Page 4 Hurricane Considerations 8 During the out-of-service period for No.12 DG, we will institute the hurricane policy described in Attachment 2 which provides our actions based on the given condition. In the event that the conditions listed in Item D. in Attachment 2 are no longer present, we plan to take actions (consistent with our normal procedures) to bring both units back on-line. This policy will apply only during this change.

Summary of Mitigating Features As discussed previously, BG&E has a unique 69 kv tie line. This power source is capable of handling all of the safe shutdown loads at the site (it has the load-carrying capacity of two DGs) and it is a fully qualified GDC-17 power source. Aligning this source, such that it is readily available, compensates somewhat for the unavailability of No.12 DG.

In the unlikely event of a LOSP followed by a failure of No.11 or No. 21 DG while No.12 DG is out-of-service (as described earlier), two full-capacity steam-driven AFW pumps would be available initially to feed the " blacked-out" unit. The recently installed cross-connect between Unit I and Unit 2 motor-driven AFW pumps gives us the ability to feed one unit from the other unit.

The Calvert Cliffs DC electrical power system (including the batteries, the battery chargers, and the inverters) is common to Units 1 and 2. In the event that only a single DG remained operable following a LOSP, it could provide battery charging that serves both units. Sufficient battery capacity would be available for four hours.

DETERMINATION OF SIGNIFICANT HAZARDS This proposed change has been evaluated against the standards in 10 CFR 50.92 and has been determined to involve no significant hazards considerations, in that operation of the facility in accordance with the proposed amendment would not:

(i) involve a'significant increase in the probability or consequences of an accident previously evaluated; or an .Jhd to d.ys of iaer?'Al%

<MT This change, allowing 2 f--- t 2: d :n r _ - for No.12 DG, does not signi:icantly increase the probability or consequences of an accident previously evaluated. Due to the commonality of our plant design and the insignificant increase in risk, the probability of J/>f retaining AC power remains virtually the same with or without a)A 10- j day outage of No.12 DG. )

1 (ii) create the possibility of a new or different type of accident from any accident previously evaluated; or No new or different type of accident will be created by this proposed change. A LOSP and Station Blackout have been evaluated.

Mr. Ashok C. Thadani September 19,1986 Page 5 (iii) involve a significant reduction in a margin of safety.

This change does involve an incremental reduction in the margin of safety in that thgIswing DG is proposed to be removed from service

< /y/7~ for up to a Ba,05y period. However, this reduction is not considered significant in that a probabilistic risk assessment performed reveals only a negligible reduction in the margin of safety.

i FEE DETERMINATION Pursuant to 10 CFR 170.21, we are including BG&E Check No. 1907020 in the amount of l $150.00 to the NRC to cover the application fee for this request.

1 Very truly yours, h

J STATE OF MARYLAND:

TO WIT:

CITY OF BALTIMORE :

Joseph A. Tiernan, being duly sworn states that he is Vice President of the Baltimore Gas and Electric Company, a corporation of the State of Maryland; that he provides the foregoing response for the purposes therein set forth; that the statements made are true and correct to the best of his knowledge, information, and belief; and that he was authorized to provide the response on behalf of said Corporation.

WITNESS my Hand and Notarial Seal: c wc/ k <3/[v Notary Public My Commission Expires: Old7 , #8

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' </ Date JAT/SRC/ dim Attachments cc: D. A. Brune, Esquire

3. E. Silberg, Esquire S. A. McNeil, NRC T. Foley, NRC T. Magette, DNR f