ML20090K465
| ML20090K465 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 05/11/1984 |
| From: | Farrar D COMMONWEALTH EDISON CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| Shared Package | |
| ML20090K451 | List: |
| References | |
| 8600N, CAL, NUDOCS 8405240214 | |
| Download: ML20090K465 (9) | |
Text
{{#Wiki_filter:c. .[ N Commonwe:lth Edison / C ) One First N:tional Plaza. Chicaoo. litges s ) Chicago, Illinois 60690 Address Reply to: Post Office' Box 767 PRINCIPAL STAEF_ .g Y, ^? 4A i lDP @/ ij/RA DE V LGN VRA DN4SF [A0 May 11, 1984 5 SGA ML INF File s?6c: Mr. James G. Keppler Regional Administrator U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, IL 60137
Subject:
Quad Cities Station Units 1 and 2 125 Volt DC Battery Operability NRC Docket Nos. 50-254 and 50-265 Reference (a): J. G. Keppler letter to Cordell Reed dated May 7, 1984 (Confirmatory Action Letter).
Dear Mr. Keppler:
In the Reference (a) Confirmatory Action Letter, we were required to provide your offices with written justification for the continued operation of our Quad Cities facility based on a battery profile analysis which demonstrates the actual capabilities of the station's batteries are within the accident analysis requirements. Also, Quad Cities Station must implement procedures to reduce the 125 Volt DC loads below 62.3 amperes within 30 minutes upon the loss of the associated battery chargers. The enclosed attachment to this letter addresses the battery profile for the worst case of the following:
- 1) No Break
- 2) Small Break
- 3) Large Break Detailed operating sequences are shown in attachment B.
When the design events analyzed in the FSAR (Small Break, Large Break) are evaluated against the battery capacity, a factor of 9 exists in the battery capacity. The no break scenario appears to be the most severe due to the four hour need, for DC power to allow for the normal cooldown rate (700F/HR) to drop to 2800F, at which tirce RHR could be initiated and maintained without DC power. .This particular worst case scenario was not considered in the original design basis of the' plant. However, with the procedure being implemented to reduce DC load to 62.3 amps within 30 minutes the battery capacity will meet the requirements of this scenario. e405240214 e40518 PDR ADOCK 050002g P g
% If you have any further questions regarding this matter, please contact this office. Very truly yours, Dennis L. Farrar Director of Nuclear Licensing EJR:mnh cc: NRC Resident Inspector - Quad Cities Attachment 8600N b I I L L To I ..n,.-
s, ',.s 5-11-84 QUAD CI' TIES DC SYSTEM ANALYSIS NRC RESPONSE Based on the attached scenarios, (Attachment "B") battery capacity calculations have been performed to determine whether the battery has sufficient capacity to energize the DC System loads for the required duration. The load cycle utilized for these calculations is illustrated in Attachment "A". This load cycle has been modified from the load cycle used in the original battery sizing calculation to more accurately reflect present loading and to reflect the new station procedure that reduces the 125 volt Direct Current Loads below 62.3 amperes within 30 minutes on loss of the associated chargers.
== Conclusion:== Based on the calculation, the batteries were found to have sufficient capacity to energize the system loads for 4 hours, which envelops the requirements for (a) the no line break scenario and (b) the small and large break scenarios. i e 4 O
- O e
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.-.-. =_ 125 'VDC <'ontrol Battery Attachment A ,V 5-11-84 4 'I'otal Elapsed Times 1 1/4 1/2 1 ,2-1/2 3 4 min. I!r Hr Hr Hr Hrs Hrs __npxt next,next next nek.q 1st 181-1/4 1/2 1-1/2 1/2jnext Tabulation of Loading: 14in. mins Mr. Hr. Hr. Hr 1 Hr. Escape Lighting ~ 100 100 100 4 Annunciator Relay. Cabinet & i 15 15 15 15 15 15 . Visual.'.nnunciator p Indicating Lamps & Auxiliary ! ; 42 42 42 37 37 37 37 Relays RICI. Valve l ~ 5 5. s d 15 15 i Pidnt S.irens- -Electrot-:stic Relief Valves 32 4 4
- l-]-l-l Trip OC3S -(Sw. Yd.)
120 j-{- l-l- Trip ricid. ACB . 10 s l-
- -l-Trip ACBS-
'120 l-Trip 'furb. dts. 10 I-j- l-l 2 l-1 - l 'Close. ACBS 40_ Standby Dicsel Field l14O l-l- )- ,- {- l Flashir.g t i i t i i HPCI Turbine Controls' 5 5 5 ,5
- 5
,5 l5 e E Tip Systela Sitear Verives 50 l' I f, HPCI Turb. Drain Valves 5 l 5'. 5 ~5 j5 5 ~ 5 l. ( l s . Total Discharge Currents:~' 709 191. 171' G2 62 62 62 x t .Y W m g-1 ..h'
QUAD' CITIES STATION UNIT 2 5/11/84 ^" ~ " - i LOSS OF BATTERY CIIARGERS + -- i Operating ~ Sequence for LOOP Without LOCA t I e Auto'discon.nect safety buses 23-1 and 24-1 from buses 23 and 24 4 e Auto trip loads supplied by 23-1 and 24-1 1 e Auto trip loads on 23 and 24 (includes feedwater pump trip) e Scram - (Note 1) 2 e ' Main Steam Line Isolation from Condenser (Note 2) e Generator Trip ~ e Diesel Generator Auto Start-i e Breakers auto close energizing 23-1 and 24-1 from DG e Auto Sequence Loads onto 23-1 and 24-1 Auto open Target Rock and Electromatic Relief Valves i e - o' Auto Initiate HPCI and RCIC (Note 3) e Auto. trip HPCI (Note'4) p e Manual stop RCIC-(Note 5) Manually-control RCIC and Electromatic relief valves to I e depressurize (Note 6) I Manual closure' of breakers from 23-1 to 23 and 24-1 to 24 e Manual start service RBCCW System, Service Water Pump and 1 e CRD' pump. Manually start RHRS for Suppression Pool Cooling e Manually realine RHRS for Shutdown Cooling e F Note 1: On low-voltage to protection system. Note 2: Turbind stop and bypass valve closure due to loss of i condenser vacuum. - Note 3: On low-low water level. Note 4: On high water level approximately 5-10 minutes after initiation. Note 5: After refilling reactor approximately 5-10 minutes after [ initiation. f Note 6: A four-hour cooldown is assumed based on a 70*F/hr 2 cooldown rate from 550'F (Reac, tor operating temperature) to 280*F (RHR initiation temperature). (
Reference:
Abnormal Procedure QGA-12), i i L .-w. y-e ~,, -.--y- ,--,-,,,,-v.,, -w, ,---w ,--w-- .--,-ey---. ,,vy,----w .~,,-we,mme,e -e
QUA7 CITIES STATION UNIT 2 5/11/84 loc? 09 P ATTERY CHARGSRS Operating Sequ .s. rqr LOOP with Small Break LOCA (Note 7) e Auto disconnect safety buses '23-l' and 24-1 from buses 23 and 24 e Auto trip loads supplied by 23-1 and 24-1 e Auto trip loads on 23 and 24 (' includes feedwater pump trip) e Scram,(Note'l) 6 e Main S' team Line~ Isolation from Condenser (Note 2) ~ e Generator Trip j Diesel ~ Gen'rator Auto Start e e BreaNers auto close energizing 23-l 'and 24-1 from DG e e Auto Sequence Loads onto 23-1 and 24-1 e Auto initiate HPCI and RCIC or ADS e Auto open admission valves for CS/LPCI (. Note 3) e Auto close HPCI steam isolation valves (Note 4) e Auto close RCIC steam isolation valves (Note 5) e Manually close breakers from 23-1 to 23 and 24-1 to 24 Manually'real'ine RHRS from LPCI to containment cooling mode e (Note 6) Note 1: On' low-voltage to protection system. Note 2: 3EIV closure er turbine stop and bypass valve closure due l to loss of condenser vacuum. Note 3: Reactor pressure equals 325 psig. Note 4: Reactor prer'sure equals 100 psig. Note 5: Peactor pressure equals 50 psig. Note 6: Approximately 10 minutes afte'r initiation. Note 7: FSAR Figure 6.2.17 shows that reactor pressure is reduced to approximately 120 psig in less than 10 minutes. Fifteen minutes is conservatively assumed for depressurization for de load on battery. l w p - y -,u --e, a
i 1 ... ~ / f' Rctc (Nore1) % 'O oy NPCI ec 0 04 7 iu54 AFL/6F wwer l (Aro re 2) 1 AC S YS TGM i ^ l l lo m w. / y y 4 \\ tissE (Hons) I istA.10R ECcS DC LOAD $ VS T/MG - L O O P w t rHOL/7" l.0CA ~ 1 j NO TC / ~ Reic DC loa DS ARE coNSE A VA TtvEd.y AS.synsCO k \\ 70 BC EGVAL ro ritt N/'ct oc s.oADs, G 1 \\ i oo Aso rs 2 ; FOR LOAO CALcub ATION PURPOSES 3 T//6 TorAL RELIEF VALVE OC LOAO WAS PLACEP lN T//6 f*/ns r 1/2. 140l/K o f THE LOA D C YC L E. t
CETZES ST&TXOM UNET 2 5/11/84 LOSS OF BATTERY CHARGERS Operating Sequence for LOOP with Large Break LOCA (Note B) Auto disconnect sarety buses 23-1 and 24-1 from buses 23 and 24 e Auto trip loads supplied by 23-1 and 24-1 e e Auto trip on 23 and 24 (includes feedwater pump trip) e Scram (Note 1) Main Steam Line Isolation from Condenser (Note 2) e k Generator Trip e Diesel Generator Auto Start Breakers auto close energizing 23-1 and 24-1 from DG e e Auto Sequence Loads onto 23-1 and 24-1 e Auto Initiate HPCI and RCIC or ADS (Note 3) Auto open admission valves for CS/LPCI (Note 4) e e Auto close HPCI steam isolation valves (Note 5) e Auto close RCIC steam isolation valves (Note 6) Manually close breakers from 23-1 to 23 and 24-1 to 24 e Manually realine RHRS from LPCI to containment cooling mode e (Note 7) Note 1: On low-voltage to protection system. Note 2: MSIV closure or turbine stop and bypass valve closure due to loss of condenser vacuum. Note 3: On large breaks HPCI, RCIC, or ADS will actuate.
- However, no credit is taken for their use because rapid depressuri-zation causes closure of system isolation valves.
LPCI and CS will completely handle core cooling. Note 4: Reactor pressure equals 325 psig. Note 5: Reactor pressure equals 100 psig. Note 6: Reactor pressure equals 50 psig. Note 7: Approximately 10 minutes af ter initiation. Note,8: FSAR Figure 6.2.13 shows that reactor pressure is reduced to 150 psig in less than 7 minutes. Fifteen minutes is ' conservatively assumed for depressurization for de load on battery.
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