ML20033C439
| ML20033C439 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 11/10/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20033C437 | List: |
| References | |
| NUDOCS 8112030215 | |
| Download: ML20033C439 (8) | |
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.c-SAFETY EVALUATION REPORT EFFECT OF DC POWER SUPPLY FAILURE U
ON ECCS PERFORMANCE BRUNSWICK UNITS 1 AND 2
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BACKGROUND It has been recognized that the lor; of a direct current (DC) power supply could disable several' emergency core cooling system components and thereby could result in a. limiting singl.e failure condition ~ for some BWR recirculation line breaks.. In
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Reference 1, General Electric has submit [ed the results of a study of a DC power.
4 source failure specifically for BWR/3's and 4's.
The staff has under generic re-view the justification that the ECCS performance continues to be adequate.,
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By Reference 2, Carolina Power and Light Company (CP&L) was regt.asted to address this concern specifically for Brunswick Steam Electric Plant Units 1 and 2.
In addition, CP&L was asked to address the loss of ECCS equipment due to water spillage.
CP&L responded to this request in References 3 and 4.
This Safety Evaluation Report is plant specific for Brunswick 1 a'nd 2 and is not intended as a generic review.
REGULATORY STAFF EVALUATION The General Electric study (Reference 1) was conducted for two plant categories; LPCI mod and non-LPCI mod.
(Non-LPCI mod plants have retained LPCI loop selection logic). The GE study covered both the large and small break regions and analyses were performed with the 1977 GE ECCS evaluat, ion model described in NED0-20566,
" General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50 Appendix K."
The licensee's response to the staff's request to 81120 215 811 10 PDR A K 050 P
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address a DC power source failure provided a comparison of the Brunswick ECCS c40ip-Nn't' remaining operable with the results obtained from the GE study for each
.. hypothesized accident.
Brunswick I and 2 are BWR/4's with modified low pressure coolant injection (LPCIT' systems. The LPCI is designed for automatic operation following a br.eak.in.one of the reactor recirculation loops. The LPCI logic is required to open the injection v'alves to both recirculation loops and close the recirculation pump discharge valves and the' discharge bypass valves in both reciiculation loops. The valves in th'e
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dan 5[ged r'ecirculation l'oop are left open t'o allow continued depris'surization of' hhe
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primary system. The assessment provided by the licensee is 'therefore based on
~~ comparison with the GE analysis for LPCI-mod plants. The results of'the licsnsee's si[riey are summarized in the attached Tables 1 and 2.
The results of the comparison with the GE analysis are provided in Table 3.
The design of the Brunswick plant is such that the LPCI system is a subsystem of
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the Residual Heat Removal (RHR) system.
Each unit is equipped with two RHR loops with LPCI mode injection into the pump discharge side of each primary recirculation line.
Each RHR loop has two RHR pumps for a total of four per unit.
Under normal conditions, the two loops of the RHR system are cross connected by a single header, making it possible to supply either loop from the pumps in the other loop.
The 125/250 volt DC power supply system consists of two batteries and four battery chargers per unit which supply electric circuitry and switches required for operation and surveillance of the system.
The two batteries for each unit have engineered safety feature control loads for the two units distributed among them so that re-dundant subsystems on each unit have separate normal power supplies. All critical loads are supplied by redundant sources of DC power.
Load switching'is done manually.
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l The HPCI system is not a redundant system and has its control power suppl.ied by one bus for each unit'. Therefore the single failure of one DC power supply wo'. tid.elimi-nate the HPCI system entirely. 'This is identified in' Tables 1 and 2 for~ Brunswick Unit 2 as a failure of Battery 2A. The Core Spray and RHR sytems aviredundancy and are only partially affected by a control power failure.
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The Licensee's response included the effects of environmental c'onditions '(water spillage from break) on the ECCS equipment available. The HPCI, core spray pumks and RHR a
pumps,are located o~utside the; primary containment and drywell.-As stajedjin the FSAR,
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all components of.the HPCI;and core spray. system.that are required for6 system opera-a9.
tion have taken into consideration the normal and aiccidentlenvironment's in which they
-1 must operate. Possible failures are considered in the failure a'n'alyif's summarized in Table 1 (discharge break) 'by considering the possibility that_RHR pump operation may
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.ii be affected by water _sp113 age._.Since_ redundant safety-related equipme.nt 'within the primary containment is designed and installed in two divisions withLphysicai separa-
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tion exir Ung between the divisions (Reference 4), only onefdivision'(two'out of four RHR pumps} is assumed to be affected. Although not completed by the licensee, the suction line break summarized in Table 2 can be extended in a similar manner to include the failure of one RHR division due to water spillage. The summary provided in Table 3 includes the combined effects of DC power source failure and water spillage. The effects of spillage and harsh environment on Class 1E equipment within primary contain-ment are considered by the staff as part of the Environmenta1' Qualifications Branch revie{.for Brunswick 1 and 2.
Existing evaluation of BWR ECC systems have assumed that the worst single failure for a large break has been the failure of a low pressure coolant injection _(LPCI) system
-injection valve, and a failure of the high pressure coolant injection (HPCI) system I
pump for small breaks. The General Elect [ic assessment of the effects of a DC power source failure showed that for large breaks the LPCI injection valve failure remains the limiting failure when compared to the DC power source failure
4-Based on the results of the licensee's review of ECCS equipment availability, the
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staff concludes that the consequences of a DC power source failure are not limiting fcr large breaks.
Fcr small breaks, the General Ele'ctric assessment of the. effects of a DC power
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source failure can have worse consequences than the HPCI pump failure as$umption.
The. remaining operable equipment for a DC :ower source failure was detemined to be~
one ' core spray component, one LPCI pump and one automatic;depressurizai.fon system component. The Brunswick combination as detemined by the licensee exceeds the GE combination by the HPCI component.- Since the HPCI component is of primary signifi.-
cance 'for small breaks, the staff concludes that the Brunswick combination is
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acceptatile.
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r REGULATION STAFF POSITION We have reviewed the response to the NRC request for a review of the ECCS equipment availability with a DC power supply failure provided by Carolina Power and Light
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Company for Brunswick Units 3 and 2.
Based on the results of our review as discussed in the above staff evaluation, we conclude that the Brunswick ECCS design is adequ' ate
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to accomodate a DC power source failure in combination. with equipment loss due to; water spillage'.
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TABLE 1
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- PRES 12fT DESIGN-DISCllARCX LINE BREAK IAOP-A ANALYSIS FOR t
YAILU;14 OY' ONE (1) BA*ITERY (D.C. POWEX) UNIT 2 s
/AILung LOSS DUE *U LOSS DUE TO LOSS IXJE 'lV IDSS DUE TO RUNNING CCtfrHOL 1VK DiERGENCY LPCI/UIJ ECTION PIPE BREAK PAILuitz,
PWR SYSTEM VLv.
PAILURE/IO OPEN gaa t t.
2A D/C #3 C.S. Pmp 2A Unit 2 4KV SWGR. E3 RRR Pump 2A RHR Pump 2C C.S. Pump 25 Div. I l@CI RIDL Ftap 25 RHR Ptsap 20
.s ADS Ita t t. '2A D/C #4
.C.S.
P.tnap 2B s
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RHR Ptamp 2A C.S. Ptump 2A.
Unic 2 4L] SVCR. E4 RllR Pump 25
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Div. II RHR Pump'2C RHR Ptsap 2D ADS 1
llPCI lil I'
g ecc. lA D/C d1 RHR Pump 2C g nit 1 4KV SVCR. El
- RHR Pt=np 2A U
C.S. Pump 2Ai
. D'i v. I-C.S. Pump 25 4 RilR Ptsap 2B 6 lutR Psaap 2D ADS ILPCI l
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.D e t t. 13 D/C 72 RH1 Ptamp 2D
. Unit 1 4KV SWCR. E2 RHR Pump 25 RIIR: Pump 2A C.S. Pump 2A
'Div. II RHR Pt::sp 2C C.S. Pump sS ;
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I ADS IIPCI
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TABI.E 2
.,PRESENT DESIGN-SUCTION LI5E BREAK LOOP-A ANAI.TSIS P04 FAILURE OF ONE (1)IBATTE.
(D.C. PWXR) UNIT 2 e
,t PAILUME
~.03S DUE TO IDSS DUP. TO 12352.DUE'IO LOSS DUE TO
- RUNNING,
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FAILURE /ID OPDi i
1 nacc. 2A
.1/C, f 3 C.S. Firap 2A C.S. Pissp 28 ADS
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Un1e 2
.KV WCR. E3 RitR Purap; 2A, R)(R Pump 2B Div. I ILPCI RilR Pump 2C RllR Pursp 2D 4
- su uste. 25 is/Ci 74 C. S. Pt.
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C.S. Ptzsp 2A ADS Untc 2 d.KV SVCit. E4 RHR IW; a
RJIR Pump 2A IIPCI Div. I'I Ri!R Pump 2C f-Rl!R Pump 2D It a c t.
lA li/C 71 RiLR Pump 2C Ri!R Pump 2A C.S. Pump 2A ADS
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Uptt 1 4 KV Sk'CE. El C.S. Pump 25 HPCI D l,v. I RHR Pump 2B RHR.Pursp 2D ita t t. IB 1/C 12 RIIR Pump'2D RitR Ptsap 2B C. S. Pum p 2 A ADS
.unit 1 4KV SVC2. E2 C.S. Puup 25 10'CI j
.Div. II j'
RJIR Pump 2A l
RilR Pump 2C i
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LOSS OF DC POWER AVAILABLE ECCS EQUIPMENT COMPARISON OF BRUNSWICK 1/2 WITH GE ANALYSIS 3
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GE BNPS SMALL BREAK Suction and Discharge 1 CS + 1.LPCI + ADS 1 CS +.1 LPCI g, ;,.
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'A ADS + HPCI:
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LARGE BREAK
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- ." Discharge '
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' -- ~ 2 C$ 5HPdI'+ ADS
. Suction
.1 CS + 3 LPCI 2 CS +2 LPCI + HPCI
+ ADS e
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REFERENCES i
t 1.
Letter, R. E. Engle.(GE) to P. S. Check, (NRC), "DC Power Source Failure for_
BNR/3 and 4," dated November 1,1978 (with attachment).
S 2..ll.etter,T.A.Ippolito:(NRC)toJ.A. Jones (CP&L),"EffectofaDCPower
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-3ource Failure on ECCS Performance," dated April 25, 1980 (withenclosure).
- 3. ' Cette'r,' E. E.'Utley (CP5L) t6 i. A. Ippolito (NRC), "Effect of DC Power Supply
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, Failure on ECCS,Perfonnance," dated September 11, 1980 (with attachments).
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- 4.. Letter, E. E. Utley (CP&L) to T. A. Ippolito (NRC), "Effect of Spillage on -
ECCS Performance," dated December 10, 1980.
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