ML20212M870

Summary of 860808 Meeting W/Util & Westinghouse to Discuss Progress on Treat Computer Code to Resolve SER Open Item on long-term Cooling.Viewgraphs Encl

ML20212M870
Person / Time
Site:	South Texas
Issue date:	08/20/1986
From:	Kadambi N Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8608270336
Download: ML20212M870 (32)

Text

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-l NUCLEAR REGULATORY COMMISSION W ASmNGTON, D. C. 20655

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/ AUG g01gp Docket Nos.: 50-498 and 50-499 l APPL.ICANT: Houston I.ighting and Power Comapny FACILITY: South Texas Project, Units 1 and 2

SUMMARY

OF MEETING HELD ON AUGUST 8,1986 TO DISCUSS j THE TREAT COMPUTER CODE.

I The applicant requested this meeting to inform the staff on the progress being

?~ made on the use of the TREAT code to resolve the SER open item on long-term '

cooling at South Texas. Enclosure 1 provides the listing of the meeting participants. Enclosure 2 provides the hand-outs at the meeting.

Discussion:

The applicant and. Westinghouse described the methodology being used for the analyses and the rationale for selection of parameters which will be used.

The analyses using these parameters will be submitted on the South Texas docket. The staff will conduct the review to support the South Texas licensing effort. The submittal on TREAT is not meant to be a replacement of the evaluation model under 10 CFR 50 Appendix K, but the applicant will request NRC approval of the claim that the code complies with Appendix K in limited areas.

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N. Prasad Kadambi, Project Manager PWR Project Directorate #5 Division of PWR I.icensing-A e60e270336 860820 8 DR ADOCK 0500

MEETING ON TREAT CODE August 8, 1986 N. P. Kadambi NRC/NRR/DPLA Jerry N. Wilson NRR/DPLA/ PARS B. Mann ~NRR/DPLA/ PARS Carl F. Berlinger NRR/DPLA/RSB R. lobel NRR/DPLA/RSB

l. Bell NRR/DPLA/RSB W. R. Spezalietti W STP licensing B. S. Marty W Nuclear Safety r- A. C. Cheung W Nuclear Safety '

T. G. Roberson Fl&P Engineering ,

J. S. Phelps Fl&P licensing L. Schlazen Fl&P Engr.

Eric Frantz W - Nuclear Safety t

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HLaP-H-NRC LONG TERM COOLING MEETING

,' AUGUST 8, 1986 DEMONSTRATE PURSUANT TO NRC QUESTIONS 440.38, 440.39 THA ACHIEVE LONG TERM COOLING FOR A SPECTRUM OF CURRENT DESIGN USING SAFETY GRADE EQUIPMENT

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SAFETY GRADE QUALIFIED EQUIPMENT A. QUALIFIED RHR SYSTEM B.

SUFFICIENT QUANTITY OF SAFETY GRADE AUXILIARY FEEDWATER C. HHSI/LHSI D. CHARGING SYSTEM E. SG/PRZR PORV II.

CASES CHOSEN TO DEMONSTRATE THE OBJECTIVE A. SMALL BREAK LOCAS (2) .7" AND 1.5" B. SECONDARY SIDE BREAKS

1. g FEED LINE/ MAIN STEAM LINE BREAK

2. l(

BOUNDS ISOLATABLE SMALL BREAK LOCA  !

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DESIGN BASIS FOR IONG TERM COOLING f

LOCA LOCA LOCA Isolatable Steam 14CA Feedwater i

3/8" - 1"- 1" - 4" ,

>4" Break

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N MITIGATION HHSI HHSI HHSI HHSI HHSI g g a l

LONG TERM COOLING with HHSI HHSI LHSI AFW AFW k

Hose Connection AFW to AFWST kg LONG TERM .

COOLING with LHSI 'LHSI LHSI RHR RHR 1 Qualified. RHR '

RHR Pump i .

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u SOUTH TEXAS PROJECT LONG TERM COOLING ANALYSIS WITH TREAT A PRESENTATION TO THE US NRC ERIC FRANTZ WESTINGHOUSE - NUCLEAR SAFETY 4

AUGUST 8, 1986 i

STP FSAR OPEN ITEMS AND NRC QUESTIONS o

RSB QUESTION 440.38 REQUIRES STP TO 10CFR PART 50.46 CRITERIA FOR LONG T SMALL BREAK LOCA RECIRCULATION REQUIRED ,

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RCS PRESSURE AB0VE LHSI PUMP SHUT 0FF HEA ~

PROVIDE LONG TERM SBLOCA ANALYSES TO  :

DECAY HEAT REMOVAL USING SAFETY GRA WATER SOURCES s

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QUESTION REMOVAL:

440.39 REQUIRES STP TO DEMO y FOR QUALIFIED EQUIPMENT ONLY I i

FOR ISOLATABLE AND UN-ISOLATABLE LOCA '

TO INCLUDE POST-LOCA C00LDOWN PERIOD

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CONSIDER SUMP NOT AVAILABLE FOR IS0LABL

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MODELING ASSUMPTIONS FOR STP LONG TERM COOLING ,

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O SOUTH TEXAS SPECIFIC o 102 PERCENT POWER i o 120 PERCENT ANS-5, 1971 DECAY HEAT o LIMITING SINGLE FAILURE o REC 0VERY WITH SAFETY GRADE EQUIPMENT I

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FEEDLINE BREAK OBJECTIVES o DEMONSTRATE C00LDOWN & DEPRESSURIZATION TO RHR CONDITIONS WITH SAFETY GRADE EQUIPMENT y -

o FOR AFWST DEPLETION, BOUND ALL SECONDARY BREAK, ISOLABLE RCS BREAK AND NO BREAK SCENARIOS HEATUP EVENT - HIGHEST RCS TEMPERATURES ESTABLISHED AFW SPILL THROUGH BREAK S

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..s Feedline Break Scenario for Long Term Cooling Remarks /

Event Time (sec) TyDe of Action Feedline Break, SG No. 1 10 FSAR Low-low SG Level Reactor Trip with 30 FSAR-automatic Loss of Offsite Power AFW Delivery - 540 gpm to SG 2 90 FSAR-automatic 675 gpm spill from SG 1 MSIVs Close on Low SL Pressure SI 561 FSAR-aatomatic Pressurizer PORVs or Safety 1300 automatic Valves Cycle

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Isolate AFU to SG 1 (E-2, Step 4) 1890 Req'd op. Act.

per Q211.52-2 i

Control AFW to SG 2 When Req'd Continuous, as  !

(E-1, Step 3 et. al.) req'd t o keep ,

level in NR SI Termination (E-1 to ES-1.1) 2400 SG 2 Pressure Reaches Safety Valve 2616 FSAR-automatic I Setpoint i

Stabilize Thot with SG 2 PORV 3000 Req'd Op. Act.  !

(dump to 1000-1100 psia) (ES-1.1) per Q211.52-2 l l

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l Feedline Break Scenario for Long Term Cooling (Continued) l Remarks /

Event Tipe (sec) Tyne of Action Establish Charging and , 4500 Req'd Op. Act.

RCP Seal Flow (align to BAT) for c.d. and boration Establich RV Head Vent Path before 7200 May be needed to control przr level Complete AFW Cross-Connect to SG 3 before 12000 Reg'd Op. Act.

(per E-1 or ES-1.1) do before c.d.

End of Hot Standby 14400 4 hrs T '

Start 25 F/hr cooldown using 14400 Req'd Op. Adt.

SG 2 and 3 PORVs Finite AFWST Operate Charging for Makeup 14400-end Req'd Cp. Act.

and Boration Switch Makeup Supply to RWST approx 30000 Use T.S. limit of one BAT End of Cooldown to 350 F, 58000 Depress. to Operate RV Head Vent and RHR cut-in.

one Przr PORV to depressurize.

End of Transient 68000 RHR Conditions

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EXPECTED AFWST DEPLETIONS FOR FEEDLINE BREAK SCENARIO AFWST USED (GALLONS) 120% ANS-5 1971 ANS-5.1-1979 DECAY HEAT DECAY HEAT T ~

AFW SPILL, 675 GPM FOR 30 MIN 20,000 20 000 PRELIMINARY ANALYSIS RESULTS 4 HRS HOT STANDBY 98,000 86,000 C00LDOWN TO 350*F (25"F/HR MAX) 252,000 194,000 l 3 HRS DEPRESSURIZE TO RHR (MAX) 45,000 35,000 TOTAL AFWST DEPLETION 415,000 335,000 AVAILABLE AFWST VOLUME 445,000 445,000 MARGIN 30,000 110,000 9

i CONCLUSIONS FOR FEEDLINE BREAK t

o DEMONSTRATE ABILITY TO C00LDOWN AND DEPRESSURIZE TO RHR CONDITIONS WITH SAFETY GRADE EQUIPMENT y - .

o DEMONSTRATE AFWST MARGIN WHEN RHR CONDITIONS REACHED i

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..s 0.7 INCH BREAK OBJECTIVES o DEMONSTRATE THAT RHR CONDITIONS CAN BE REACHED FOR A CASE g WiiERE SI FLOW WILL MAINTAIN RCS PRESSURE AB0VE RHR CUT-IN.

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C0tlDITIONS C00LDOWN CAPABILITY LIMITED PRESSURIZER PORV USE TO REFILL 3CS STOPPING OF HHSI PUMPS LilSI AND RHR FOR LONG TERM COOLING AFWS'T CAPABILITY

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0.7 Inch Diameter Cold Leg Break Scenario Remarks /

Event Time (sec) Typt_of Action' O.7" CL Break, Loop 2 0 Initiating 102% Constant Power Operation 0-197 Event Reactor Trip, Loss of Offsite Power 197 Automatic AFW Injection at 120 F -

257 Automatic (540 gpm to SG 2, 540 gpm to SG 3)

Control AFW to SG 2 and 3 When Req'd Continuous, as (E-1, Step 3 et. al.) req'd to keep level in NR SI Starts Injecting to CL 2 and 3 1740 Prcs < 1500 psi Start 50 F/hr Cooldown 2100 .

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Using SG PORVe on SG 2 and 3'

• Dump Steam from Inactive SGs 1 & 4 7000-7100 Req'd prior Using SG PORVs to RCS Depress Open One Pressurizar PORV 8000-8050 to Restore Level Stop one HHSI Pump - Loop 2 (Based on 172 F subcooling) .- 14000 Stop last HMSI Pump to Allow LHSI 18000 j (Based on Thot < 380 F)

• Open One Pressurizer PORV 18200-18500 l to Rectore Level with Subcooling and Level Indicated LHSI to Loop 3 Starts to Inject 18300 Automatic Depressurize Inactive SGs 20000-20200 If req'd End of Transient Modeled 24000 (RHR Conditions Established)

• If the inactive SGs are above 400 psia at this time, they will need to be depressurized bsfore the RCS depressurizes to less than the LHSI shutoff head pressure (300 psia). A 2000 sec delay will be assumed before completing the final actions.

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! 0.7 INCH. BREAK

SUMMARY

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i o DEMONSTRATE ABILITY TO C00LDOWN AND RE-ESTABLISH PRZR LEVEL

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j o DEMONSTRATE SEQUENTIAL STOPPING 0F HHSI PUMPS TO

DEPRESSURIZE RCS i

i o MAKEUP CAN BE PROVIDED BY ONE CCP OR LHSI

) o RHR CUT-IN CONDITIONS ESTABLISHED I

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1.5 INCH BREAK OBJECTIVES o DEMONSTRATE THAT RHR CONDITIONS CAN BE REACHED AND BREAK /LHSI CAPABLE OF REMOVING DECAY HEAT

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o BOUND LARGER BREAKS WITH RESPECT TO BREAK ENERGY REMOVAL '

o BOUND SMALLER BREAKS WITH RESPECT TO ABILITY TO ESTABLISH

RHR l

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1.5 Inch Diameter Cold Leg Break Scenario Remarks /

Event Time (sec) Tyne of Actio.D f

1.5" Cold Leg Break on Loop 2 0 Initiating 102% Constant Power Operation 0-46 Evint Reactor Trip, Loss of Offsite Power 46 Automatic AFW Injection at l'20 F 106 Automatic (540 gpm to SG 2, 540 gpm to SG 3)

SI Starts Injecting to Cold Leg 3 400 Prcs < 1500 psi l

Control AFW to SG 2 and 3 When Req'd Continuous, as

.. (E-1, Step 3 et. al.) req'd to keep level in NR l =

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Start 50 F/hr Cooldown 1500 -

Using SG PORVs on SG 2 ar.d 3 i

Accumulators in Loops 1 and 3 7000 Automatic, ,

Start to Inject -

600 psia Depressurize Inactive SGs with PORVs 10000-10200 400 psi delta (to 450 psia or same as SG 2 and 3) between SGs RWST Switchove'r (350,000 gallons used) 12000 Automatic I

High-head SI Temperature of 220 F (Psat(3 psig), no delay assumed) l Stop HHSI to Allow LHSI (Thot < 380 F) 17000 SI Reduction Start one LHSI pump (Loop 3) 17500 i

}' Continue Cooldown to < 350 F 18000-22000 RHR Cut-in End of Transient Modeled 24000 RHR Conditions (possible inactive SG and Break /LESI depressurization will also will remove be included) decay heat.

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1.5 INCH BREAK

SUMMARY

o BREAK /LHSI DECAY HEAT REMOVAL DEMONSTRATED FOR THIS AND j LARGER BREAKS I .

o RHR CUT-IN CONDITIONS DEMONSTRATED FOR THIS AND SMALLER

, BREAKS 7

1 o C00LDOWN AND STOPPING HHSI ENABLES RCS DEPRESSURIZATION TO LHSI AND RHR INITIATION PRESSURE-i o SWITCH 0VER T0. RECIRCULATION REQUIRED l

o LHSI PROVIDES LONG TERM MAKEUP I

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..g 1.5" Break 0.7" Break Feedline Break Type of Accident Condition 3 Condition 3 Condition 4 EculDment Used 675 gpm spill AFW Flow 540 gpm-SG2 540 gpm-SG2 540 gpm split Capabiliti. 540 gpm-SG3 540 gpm-SG3 to SG 2 & 3 HHSI fic. Spill Lcop 2 Inject to SIAS but no Assumed Inject to 3 Loops 2 & 3 injection (LP)

Fail Loop 1 Fail Loop 1 LHSI flow Loop 3 Loop 3 (or 2) Not used <

Inject to Not used Not used Accumulators Loops 1 & 3 EOPs permit EOPs permit Spill Loop 2 Isolation Isolation

' ' Makeup during

! r- Charging Flow Not used Not used. cooldown.'

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Possible long Boration and term makeup seal injection (optional) in hot standby SG PORVs SG 2 SG 2 SG 2 Active Loops SG 3 SG 3 SG 3 SG PORVs SG 1 & 4 SG 1 & 4 SG 1 & 4 Inactive Loops Used 2.8 hr Used 1.9 hr Not Used and when used after Rx trip after Rx trip PRZR PORV Not used Used Used l I

RCFCs Used ,Used Used I containment Spray Not Req'd Not Req'd Possibly Req'd for Short Term May be needed

Vessel Head Vent Not used Not used during hot standby and for RCS depress. to RHR cut-in.

RHR System Optional Can be placed Can be placed Operation and nin after 7 hr in service in service time to place in (Break /LHSI after 7 hr after 19 hr service cooling is r.dequate) l Table 2 -1 Summary of Safety Grade Equipment Used for the Long Term Cooling Recovery

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1. 1.5" CL Break with Min SI

2. 0.7" CL Break with Min SI (w/o spill)

3. Feedline Break l l

A Max SI ' g SG PORVs SG PORVs SG PORVs PRZR PORVs PRZR PORVs Stop all HHS!

Stop all HHSI Stop all HHSI LHSI Makeup 3 HHSI < -

Charging or LHSI Makeup LHSI/ Break LHSI Makeup RHR Cooling Cooling-RHR Cooling -

T' 2 HHSI 4)3 1( ) )

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1 HHSI "

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Min SI - -

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O. .375 .75 1.0 1.5 Break Size Diameter (Inches) e Figure 1-1 STP Fecovery Actions and Long Term Cooling Modes for varicus Break sizes and High-head SI Configurations

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S1ML M b I QUEST *0N 440.38 PAGE I OF in (Section E.3. & 15.6.5)

a. Demonstrate that the STP ECCS meets 10 CFR Part 50.46 criteria for long l .- term decay heat removal in the event of a small break LOCA of a size such that recirculation would be required but the RCS pressure either remains above the low head safety injection (LH5I) pump shutoff head or recovers after loss of the sacondary heat sink. An examination of Figures 6.3-1 -

throug>r$.3-frdoes not indicate that the STP ECCS is designed for hign head recirculation combined with decay heat removal by the RHR heat exchangers, i.e., there are no apparent provisions for routing l

recirculation flow from the RHR heat exchangers to the HH5I pumps.- Also,.

as described in Appendix 5.4A " Cold Shutdown Capability," the steam generators have a 1,imited supply of safety grade secondary water' supply, since there is not a safety grade backup to the auxiliary feedwater storage tack (AFST). Therefore, provide long term analyses for a spectrum

! of small break LOCAs that demonstrate that decay heat can be adeqwitely removed and the RC5;depressurized using only safety grade equipment and

.- water sources, assuming loss of offsite power and the most severe single failure. If credit is taken for operator actions, the STP emergency response guideline (ERG) sequence of operator actions should be followed.

Justify the timing of operator actions if they are less conservative than i? '

those recomended in ANSI N-660 for a condition IV event.  ;

~ b. In a conference call held on March.8.1985, the applicant indicated to NRC that for small break LOCAs the combined heat sink capacity of the RW57 and l

the steam generators would provide core cooling for approximately 18 1 hours, after which the reactor containment fan coolers (RCFCs) would provide an adequate heat sink for decay heat removal. No credit is taken for heat removal by the RHR heat exchangers. Provide a detailed explanation of the mechanism of energy removal from the RCS after less of the secondary heat sink and supporting analyses that demonstrate that

. energy can be adequately removed to meet the acceptance criteria of 10 CFR i

Part 50.46. We are concerned that for very small break LOCAs (e.g.,

1 inch) energy would not be adequately removed from the RCS for a considerable period of time after the accident. Thus, WCAr-9600, " Report on Small Break Accidents for Westinghouse N555 System" June 1979, l indicates that for 1 inch breaks the break can remove all the decay heat 4

only after about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and that prior to that time, auxiliary feedwater is required to maintain the heat sink.

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l QUESTION 440.39

.PAGE3 O (Section 6.3) a, It is stated in 10 CFR Part 50.46(b)(5) that, for long term cooling, 'the calculated core temperature shall be maintained at an acceptably low value l and decay heat shall be removed for tne extended In period order ofto time assurcrequirec by the-long-term radioactivity remaining in the core."

this, heat removal for this extended period must utilize equipment that is -

fully qualified for the environmental conottiens that prev acciderft'.

core with qualified equipment only, following all sizes Include consideration of the post-LOCA couldown period in your response.

and the f act that for isolated LOCAs, the sump would not be available for long term cooling. -

b. Discuss whether the RHR pumps are qualified for theIfenvironmental the RHR effect of the large and small break LOCAs and steam line breaks.

- pumps are not qualified discuss how long term mitigation of these accidents would be accomplished.

EdNRREa w_

i APPLICATION OF TREAT CODE FOR SPECIFIC STP LONG TERM SMALL LOCA RECOVERY ANALYSIS PRESENTED T0 r -

THE a NRC STAFF BY 4

AUGUSTINE CHEUNG

MANAGER, OPERATIONAL SAFEGUARDS ANALYSIS

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NUCLEAR SAFETY DEPARTMENT WESTINGHOUSE ELECTRIC CORPORATION

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!- AUGUST 8, 1986 f

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BACKGROUND o TREAT IS USED IN STP LONG TERM SMALL LOCA RECOVERY ANALYSES REAL TIME FEATURE SUITABLE FOR LONG TERM ANALYSES 7 -

SIMULATION OF OPERATOR ACTIONS MADE EASY DUE TO -

INTERACTIVE FEATURE HAS BEEN USED FOR WOG POST-LOCA EMERGENCY RESPONSE GUIDELINE DEVELOPMENT o T!!E STAFF REQUIRES A TOPICAL REPORT ON TREAT TO DEMONSTRATE PARTIAL CONFORMANCE OF TREAT WITH 10CFR50.

APPENDIX K FOR LONG TERM COOLING ANALYSIS H

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IRANSIENT REAL-TIME ENGINEERING ANALYSIS 100L (TREAT) o ADVANCED THERMAL HYDRAULIC SYSTEM CODE 7 . DERIVED FROM EARLIER VERSION OF NOTRUMP ,

TWO PHASE NON-EQUILIBRIUM (5 EQUATIONS WITH DRIFT FLUX)

GLOBAL COMPRESSIBLE (SINGLE PRESSURE FOR STEAM PROPERTIES)

B03GN CONCENTRATION o REAL-TIME AND INTERACTIVE ON MINI COMPUTER o ON-LINE OUTPUT / GRAPHICS DISPLAY o HAS BEEN USED EXTENSIVELY FOR E0P ANALYSES e

APPROACH -

o SHOW TREAT CONFORM WITH 10CFR50 APPENDIX K FOR LIMITED APPLICATION SBLOCA WITHOUT CORE UNC0VERY ..

LONG TERM COOLING RECOVERY ANALYSIS o ANALYZE FIRST HOUR OF SMALL LOCA REC 0VERY TRANSIENT FOR STP WITH NRC APPROVED EVALUATION MODEL'NOTRUMP o COMPARE TREAT AGAINST NOTRUMP RESULTS DEMONSTRATE ACCEPTABLE RESULTS .

PERFORM SUBSEQUENT LONG TERM COOLING ANALYSES o DOCUMENT RESULTS IN A REPORT l

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, FOCUS OF TREAT /NOTRUMP COMPARISON REPORT o DOCUMENT TREAT MODELS o EVALUATE STATUS OF TREAT CONFORMANCE WITH 10CFR50 APP. K

y. , FOR LIMITED APPLICATION -

SBLOCA WITHOUT CORE UNC0VERY (TWO PHASE FLOW, -

NATURAL CIRCULATION)

LONG TERM COOLING REC 0 VERI o COMPARE TREAT AGAINST NOTRUMP FOR STP APPLICATION WITH A 1.5" SBLOCA 1 -

TWO PHASE NATURAL CIRCULATION FLOW ASYMMETRIC LOOP CONDITIONS ABILITY TO MODEL OPERATOR ACTIONS (C00LDOWN, .

DEPRESSURIZATION .... ETC.)

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SUMMARY

OF'10CFR50 APP'. K CONFORMANCE STATUS o TREAT IN CONFORMANCE WITH APP. K REQUIREMENTS IN AREAS

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THAT-IMPACT

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SBLOCA (WITHOUT CORE UNC0VERY) PHENOMENA I

o AREAS OF NOT STRICTLY CONFORMING INCLUDE:

LARGE BREAK LOCA PHENOMENA.(BLOWDOWN, REFILL, REFLOOD, HEAT TRANSFER)

PHENOMENA ASSOCIATED WITH PROLONGED CORE UNC0VERY (METAL WATER REACTION, CLADDING RUPTURE, MOMENTUM FLUX FOR AREA CHANGE) b i - .

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' w CONCLUSION i .

3 o TREAT HAS THE NECESSARY AND REQUIRED T/H MODELS ,

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o TREAT CONFORMS WITH'10CFR50 APP..K REQUIREMENT FOR l PROPOSED LIMITED APP!ICATION i

SBLOCA WITHOUT CORE UNC0VERY LONG TERM COOLING RECOVERY o A REPORT HAS BEEN GENERATED SUITABLE FOR REVIEW BY .

THE-NRC TO CLOSE OUT THE OPEN ITEM RELATED TO TREAT's USAGE-S s

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Mr. J. H. Goldberg Houston lighting and Power Company South Texas Project cc:

Brian Berwick, Esq. Resident Inspector / South Texas Assistant Attorney General Project Environmental Protection Division c/o U.S. Nuclear Regulatory Commission P. O. Box 12548 P. O. Box 910 Capitol Station Bay City, Texas 77414 Austin, Texas 78711 Mr. Jonathan Davis Mr. J. T. Westermeir Assistant City Attorney Manager, South Texas Project City of Austin Houston lighting and Power Company P. O. Box 1088 P. O. Box 1700 Austin, Texas 78767 Houston, Texas 77001 Ms. Pat Coy

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- Mr. H l.-Peterson Citizens Concerned About Nuclear-Mr. G. Pokorny Power City of Austin 5106 Casa Oro -

P. O. Box 1088 San Antonio, Texas 78233 Austin, Texas 78767 Mr. Mark R. Wisenberg Mr. J. B. Poston Manager, Nuclear licensing Mr. A. Von Rosenberg Houston lighting and Power Company City Public Service Boad P. O. Box'1700 P. O. Box 1771 Houston, Texas 77001 San Antonio, Texas 73296 Mr. Charles Halligan Jack R. Newman, Esq. Mr. Burton L. Lex Newman & Holtzinger, P.C. Bechtel Corporation 1615 L Street, NW P. O. Box 2166 -

Washington, D.C. 20036 Houston, Texas 77001 Melbert Schwartz, Jr., Esq. Mr. E. R. Brooks Baker & Botts Mr. R. L. Range One Shell Plaza Central Power and light Company Houston, Texas 77002 P. O. Box 2122 Corpus Christi, Texas 78403 Mrs. Peggy Buchnrn Executive Director Citizens for Equitable Utilities, Inc.

Route 1, Box 1684 Brazoria, Texas 77422 e

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• Fouston lighting & Power Company South Texas Project cc:

Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Office of Executive Director for Operations -

611 Ryan Plaza Drive, Suite 1000 ,

Arlington, Texas 76011 Mr. lanny Sinkin, Counsel for Intervenor Citizens Concerned about Nuclear Power, Inc.

Christic Institute i 1324 North Capitol Street Washington, D.C. 20002 licensing Representative Fouston lighting and Power Company l

Suite 1309 7910 Woodmont Avenue Bethesda, Maryland- 20814 .

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AUGo0 St%f, Docket.Nos.: 50-498 and 50-499 APPLICANT: Houston lighting and Power Comappy FACILITY: South Texas Project, Units 1 and 2

SUMMARY

OF MEETING HELD ON AUGUST 8,1986 TO DISCUSS THE TREAT COMPUTER CODE.

The applicant requested this meeting to inform the staff on the progress being made on the use of the TREAT code to resolve the SER open item on long-term .

. cooling at South Texas. Enclosure 1 provides the listing of the meeting participants. Enclosure 2 provides the hand-outs at the meeting.

Discussion:

The applicant and Westinghouse described the methodology being used for the analyses and the rationale for selection of parameters which will be used.

The analyses using these parameters will be submitted on the South Texas docket. The staff will conduct the review to support the South Texas licensing effort. The submittal on TREAT is not meant to be a replacement of-the evaluation model under 10 CFR 50 Appendix K, but the applican,t will j request NRC approval of the claim that the code complies with Appendix K in limited areas.

4 N. Prasad Kadambi, Project Manager PWR Project Directorate #5 Division of PWR licensing-A l

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k'i . s 8/to/86 ,

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s (c, - y9s AUG 20 3

\ o Meeting Summary Dis'ribution t ROschet E diiOsl*F11F ( NRC Participants

!' NRC'PDR-'* *"- " ' " * " ~

l.ocal PDR N. P. Kadambi i PD#5 Reading File Jerry N. Wilson

. J. Partlow B. Mann V. Noonan Carl H. Berlinger Project Manager P. Kadambi R. l.obel i

0 ELD L. Bell E. Jordan

, B. Grimes

) ACRS (10)

M. Rushbrook 2-l i

cc: Licensee and Plant Service list 4

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