ML19209B694

Responds to 790917 Ltr Re Interaction Between nonsafety- Grade & safety-grade Sys.Forwards Results of GE Comprehensive Generic Assessment.Only Potentially Adverse Sys Interaction Involves Reactor Vessel Head Vent

ML19209B694
Person / Time
Site:	FitzPatrick
Issue date:	10/05/1979
From:	Early P POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	Harold Denton Office of Nuclear Reactor Regulation
References
JPN-79-63, NUDOCS 7910100327
Download: ML19209B694 (7)

Text

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- POWER AUTHORITY OF THE STATE OF NEW YORK to CoLUMaus CIRCLE NEU/ YORK. N. Y. 1C019 (212) 397 6200

&. October 5, 1979 JPN-79-63 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation Nuclear Regulatory Commission Washington, D. C. 20555

Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Non-Safety / Safety Grade Systems lie;eraction

Reference:

NRC Letter to All Ope.ating Light Water Reactors dated September 17, 1979

Dear Mr. Denton:

This letter responds to the above Reference regarding the interaction between non-safety grade systems and safety grade systems at the James A. FitzPatrick Plar ~ .

Geaeral Electric has completed a comprehensive generic assessment of the subject, the results of which are enclosed as Attachment 1. In addition, our Architect Engineer, Stone &

Webster, was requested to confirm that equipment locations assumed in the GE evaluation are appropriate for the FitzPatrick Plant, and that all non-safety grade systems were accounted for.

Thi; has been confirmed.

As shown on the Attachment, except for the reactor vessel head vent valves discussed below, the evaluation has not identi-

'ied any impact on safety actions or analysis conclusions which sould increase the consequences (calculated peak cladding tem-perature, peak containment pressure, peak suppression pool tem-perature, or radiological release) of any SAR events. In particular, the evaluation concludes that:

(1) No previously identified safety actions would be negated by the failure of non-safety equipment due to environmental effects of high energy pipe breaks (HEPB's);

QP 4

1135 293 h\

7910 1 00 3( l' P

Mr. Harold R. Denton U. S. Nuclear Regulatory Commission -2~

(2) No previously identified safety limits would be violated by the subject effects.

As noted previously, the only potentially ad erse non-safety system i . chion identified involves the reactor vessel head vent, wnr... As comprised of a two inch line with two air-operated isolation valves in series. These valves are closed during power operation. The probability of a LOCA steam environment causing both isolation valves to open at the start of the event is exceedingly small. However, to bound this worst case, GE has executed an analysis assuming a LOCA concurrent with simultaneous opening of the two vent line isolation valves. Depending on the size of the LOCA, a worst case analysis indicates that there could be a 10 0 F impact on Peak Clad Temperature. This is con-sidered an insignificant change, as defined in 10CFR50, Appendix K.

A later opening of the head vent isolation valves would reduce this impact further.

A number of observations should be made even in light of the successful evaluation of systems interaction.

(1) The criteria and suggested NRC Staff evaluation basis involved in this assessment are new, recently evolved requirements from RG 1.70, Rev. 2.

Previous plant design bases for non-safety equip-ment established a " fail as is" mode rather than the present " fail in worst position' . This is a rather arbitrary and extremely conservative re-quirement.

(2) Evaluation of plant safety in regard to HEPB's have been conducted in recent years. Comprehensive analyses were submitted to the NRC Staff and their approval was documented. Reevaluation here for more severe criteria has confirmed the previous safety audit.

In summary, the evaluation submitted herein concludes that no environmentally induced f ailures of no - safety grade systems lastalled at the FitzPatrick Plant effc- to a significant degree, the ability of safety grade systes to perform their intended safety functions on any safety analyses executed for the FitzPatrick Plant.

1135 294 .

' .Mr . Harold R. Denten U. S. Nuclear Regulatory Commission ,

On this basis, continued operation of the FitzPatrick Plant is justified, operating license.

without any modification to the existing NRC Very truly yours, o

, w L Paul J. Early[

Assistant Chibf Engineer-Projects I Subscribed to and sworn this day of October, 1979. ,

Notary Public RUTH G. ZAPF Netary Public. State of Nevi York No. 30G3428 Orplified en Nassau County Cnmnussion D:: ires March 30,19_

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, . ATTACHMENT 1 EFFF"T OF NON-SAFETY SYSTEM FAILURSS (PO LATED DUE TO ADVERSE ENVIRONMENT)

ON .iERFORMANCE OF SAFETY EQUIPMENT Table 1 attached summarizes the effect of non-safety system failures on the perfcrmance of safety equipmer.t. The table identi-fies the non-safety systems installed at the FitzPatrick Plt.nt and the effect of their postulated failure on safety system per-formance for a variety of postulated high-energy pipe breaks, locations, and sizes.

Entries in the table are as follows:

1. Environmental induced malfunction may provide adverse response, i.e., increase in previously reported peak

. Drywell pressure

. Wetwell pressure

. Suppression pool temperature

. Fuel clad temperature

2. Environmental induced malfunction will not provide adverse response.

3. System is qualified for adverse environment.

4. System will not experience adverse environment.

5. No conceivable system failure can affect response.

Location abbreviations in the table are as follows:

DW - Dry well RB - Reactor building (including crescent area)

TB - Turbine building CR - Control room RR - Relay room MG - Motor generator room ii3S 296

TABLE 1 ENVIRONMENTAL INTERArTION FEE 0 WATER LOCA RWCU RCIC llPCI MAIN STEAM LINE:

inside Breaks Inside Inside Reacto r Turbine Reactor Turbine location Small Large Bldg. Bldg. Inside Bldg._ Bldg. Small Large Outside outside Outside RECIRC SYS1EM 4 2 4 4 2 2 4 4 4

. P uu.p s DW 2 2 4

' 4 3 4 4 3 3 4 4 4

. Valves & Operators DW 3 3 4 4 4 4 4 4 4 4 4 4 4 4

.MG Sets MG 4 4 4 4 4 4 4 4 4 4

.MCC RB 4 4 4 4 4 4 4 4 4 4 4

. Speed Control Systen CR/MG 4 4 4 4 4 4 4 4 4 4 4 4 4

. Control Inst. Transmi tters RB 4 4 4 FEEDWATER DELIVERY SYSTEM 4 4 2 4 4 4 4 4

. flow Elements TB 4 4 4 2 4 2 4 4 2 2 4 4 4

.tevel DW/RB 2 2 4 4 4 2 4 4 4 4 4

. Pun.p s TB 4 4 4 2

• 4 2 4 4 2 4 4 4 4

. Valves & Operators TB 4 4 4 4 4 4 4 4 4 4 4 4 4 4

.MCC TB 4 4 4 4 4 4 4 4 4 4

. Flow Control System CR 4 4 4 4 4 4 2 4 4 2 4 4

.fW lieating TB 4 4 4 4 2 2 2 4 2 2 4 2 2

. Instrun.ent Ai r RB/TB 4 4 4 4 4 4 4 2 2 4 2 2

. Control Inst. Transmitter RB/lB 4 TURBINE PRESSURE CONTROL 4 4 2 4 4 4 4 4

. Bypass Valves TB 4 4 4 2 4 2 4 4 2 4 4 4 4 4

. Pressure 5ensors TB 4 4 4 4 4 4 4 4 4 4

.Lontrol System CR/RR 4 4 4 4

- NfuTRON MONITORING SYSTEM U 2 4 2 2 4 2 2 2 4 4

.LPRMs & Cables DW/RB/CR/FR 2 2

' ty- 4 4 4 4 4 4 4 4

.APRMs & Cables CR 4 4 4 4 2 2 4 2 2 2 4 4

.RPIS DW/RB/CR/RR 2 2 2 4 2 2 4 2 2 2 4 4 I' ;. .TIP CW/RD/CR/RR 2 2 2 4 4 4 4 4 4 4 4 4

'-C) 4 4 4 4

.kod elock Moni tor CR N

TABLE 1 ENVIR0r3tE"9L 'NTERACTION MAIN STEAM llNE: FEEDWATER LOCA RWCU RCIC HPCI Inside Breaks Inside Inside Reactor Turbine Reactor Turbine location Small La rge Bldg. _ Bldg._ Inside Bldg. Bldg. Small gr3e. Outside Outside Outside REACTOR PROTECT 10ft SYSTEM

. Turbine Scram '

TB 4 4 4 2 4 4 2 4 4 4 4 4

.MG Set TB 4 4 4 4 4 4 4 4 4 4 4 4 REACTOR MAfiUAL C0f41ROL SYSTEM RB/CR/RR 4 4 4 4 4 4 4 4 4 4 4 4 SRV SYST[M (Hun AI)S) DW/RB/CR/RR 3 3 3 4 3 3 4 3 3 4 4 4 RBCCW SYSTEM DW/RB 2 2 2 4 2 2 4 2 2 4 4 4 TBCCW SYSTEM TB 4 4 4 2 4 4 2 4 4 4 4 4 RWCU DW/RB 3 3 2 4 3 2 4 3 3 2 2 2 SUPPRESSION POOL

. Temperature Monitoring RB/ TORUS /CR 4 4 2 4 4 2 4 4 4 2 2 2

.tevel Monitoring RB/ TORUS /CR 4 4 2 4 4 2 4 4 4 2 2 2 CIRCULATING WATER SYSTEM INTAKE /TB 4 4 4 2 4 4 2 4 4 4 4 4

! (Non Safety)

HVAC SYSTEM ALL 2 2 2 2 2 2 2 2 2 2 2 2 AC AUXIt!ARY ELECTRIC RB/TB 4 4 4 4 4 4 4 4 4 4 4 4

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CONDENSATE TRANSFER & STORAGE TB/RB 4 4 2 3 4 2 2 4 4 2 2 2 LA (J- MAIN TURBINE & CONTROLS TB 4 4 4 2 4 4 2 4 4 4 4 4

. MAIN CONDENSER & CONTROL TB 4 4 4 2 4 4 2 4 4 4 4 4 Q INSTRUMENT AIR SYSTEM CO *

. Compressors TB 4 4 4 2 4 4 2 4 4 4 4 4

. Piping & Controls TB/RB/DW/CR/ 2 2 2 2 2 2 2 2 2 2 2 2 RR

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TABLE 1 ENVIRnNMENTAL INTERACTION MAIN STEAM LINE FEEDWATER 10CA RWCU RCIC llPCI

_Inside Breaks Inside Inside Reactor Turbine Reactor Turbine location Small la rge Bl.dg . Bldg. Inside Bldg. Bldg. Small La rqe Outside Outside Outside 2 2 4 2 2 4 4 2 2 2 flRE PRUTECTION SYSTEM TB/RB/CR 4 4 2 4 4 2 4 4 4 4 4 4 CRD llYDRAullC SYS1EM (NON SCRAM) RB 4 4 4 4 2 4 4 1 1 4 4 4 RV ItEAD VENT (Note A) DW 2 2 2 4 3 2 4 3 3 2 2 2 SLC SYSILM DW/RB/CR/RR 3 3 NOTE: A. The potentially adverse impact involves air operated reactor vessel head vent isolation valves. The adverse impact is negligible as discussed in the forwarding letter.

ANY HIGH ENERGY BREAK SYSTEM l

Lighting 5 l

Coninuni ca tion 5 Service Air Lines 5

)';

Equip. Drain Piping Drywell Temp. Monitoring 5

5 under Vessel Maintenance Equip. 5 Process Computer 5 Area Radiation Monitoring 5 i Process Radiation Monitoring 5 Sampling Systems 5 j 5

. Maintenance Monorails

' Environs Monitoring 5 Deminerallied Water 5 Potable Water 5 Screen Wash 5

~ 5

' liydrogen Cooling

< LN Coridenser Priming 5 Stator Cooling 5

, (f 5 Offgas

, Radwas te 5 Fuel Pool Cooling 5

% Makeup Treatment 5 NO Lube Oil System . 5