ML20215B637

Forwards Info on SER Open Item 16 & Confirmatory Item 13 Re long-term Cooling,Including Westinghouse Long-Term Cooling Analysis for South Texas Project & Repts WCAP-11232 & WCAP-11297.WCAP-11232 Withheld (Ref 10CFR2.790)

ML20215B637
Person / Time
Site:	South Texas
Issue date:	09/30/1986
From:	Wisenburg M HOUSTON LIGHTING & POWER CO.
To:	Noonan V Office of Nuclear Reactor Regulation
Shared Package
ML19292F992	List: ML20210Q647 ML20210Q685 ML20215B637 ML20215C022 ML20215C032 ST-HL-AE-1767, Comparison of Treat & Notrump Small Break LOCA Transient Results
References
ST-HL-AE-1767, NUDOCS 8610060980
Download: ML20215B637 (16)

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The Light Company n ,m,,, upo,e P,m m n.,x noo n.,m>,,.wxas ,,om aes.mn September 30, 1986 ST-HL-AE-1767 File No.: G9.18 Mr. Vincent S. Noonan, Project Director PWR Project Directorate #3 U. S. Nuclear Regulatory Commission Washington, DC 20555 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 SER Open Item #16 and Confirmatory Item #13; Long Term Cooling

Reference:

Letter ST-HL-AE-1724 dated August 20, 1986; M. R. Wisenburg to V. S. Noonan

Dear Mr. Noonan:

This letter transmits information with respect to South Texas Project Safety Evaluation Report (SER) open item #16 and confirmatory item #13. These items involve the use of the Westinghouse (}{) TREAT computer program to evaluate STP's long term cooling capabilities in the event of a small break loss-of-coolant accident (LOCA), a non-isolable LOCA and a secondary break.

The analysis has been completed and demonstrates that the STP emergency core cooling system (ECCS) meets the requirements of 10CFR50.46. The analysis, which was discussed with the NRC staff in August 1986 (meeting minutes were provided in the reference), is provided in Enclosure 1.

Enclosure 2 includes associated revisions to the FSAR which will be incorporated in a future FSAR amendment.

Also provided is a comparison of the TREAT and NOTRUMP Westinghouse programs which was used to benchmark TREAT against applicable 10CFR50.46 criteria. Enclosure 3 contains the following:

1. 5 copies of WCAP-ll232, " Comparison of the TREAT and NOTRUMP Small Break LOCA Transient Results." (Proprietary)

2. 5 copies of WCAP-11297, " Comparison of the TREAT and NOTRUMP Small Break LOCA Transient Results." (Non-proprietary)

Also enclosed is a Westinghouse authorization letter (CAW-86-084), Proprietary Informatior. Notice, and accompanying affidavit.

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• ST-HL-AE-1767 Houston Lighting &c Pbwer Company File No.: G9.18 Page 2 As itam (1) of Enclosure-3 contains information proprietary to Westinghouse Electric Corporation, it is supported by an affidavit signed by Westinghouse, the owner of the information. The affidavit. sets forth the basis on which.the information may be_ withheld from public disclosure by the

-Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.790 of the Commission's regulations.

Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10CFR Section 2.790 of the Commission's regulations. Cortespondence with respect to the proprietary aspects of the Application for Withholding or the supporting Westinghouse affidavit should reference CAW-86-084 and should be addressed to R. A. Wiesemann, Manager, Regulatory and Legislative Affairs, Westinghouse Electric Corporation, P. O. Box 355, Pittsburgh, Pennsylvania 15230.

Based on the information provided in the enclosures Houston Lighting &

Power Company considers the aforementioned items to be " closed". If you should have any questions on this matter, please contact Mr. J. S. Phelps at (713) 993-1367.

Very truly yours, M. R. Wis urg Manager, N clear Lic s ng JSP/yd Enclosures 1: Long Term Cooling Analysis for South Texas Project 2: FSAR revisions 3: Comparison of the TREAT and NOTRUMP Small Break Transient Results Five copies each of WCAP-ll232 (Proprietary) and WCAP-11297 (Non-proprietary)

L1/NRC/na

, Houston Lighting Se Pbwer Company ST-HL-AE-1767' File No.: G9.18 Page 3

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Hugh L. Thompson, Jr. , Director Brian E. Berwick, Esquire Division of PWR Licensing - A Assistant Attorney General for Office of Nuclear Reactor Regulation the State of Texas U.S. Nuclear Regulatory Commission P.O. Box 12548, Capitol Station Washington, DC 20555 Austin, TX 78711 Robert D. Martin Lanny A. Sinkin Regional Administrator, Region IV Christic Institute Nuclear Regulatory Commission 1324 North Capitol Street 611 Ryan Plaza Drive, Suite 1000 Washington, D.C. 20002 Arlington, TX 76011 Oreste R. Pirfo, Esquire N. Prasad Kadambi, Project Manager Hearing Attorney U.S. Nuclear Regulatory Commission Office of the Executive Legal Director 7920 Norfolk Avenue U.S. Nuclear Regulatory Commission Bethesda, MD 20814 Washington, DC 20555 Claude E. Johnson Charles Bechhoefer, Esquire Senior Resident Inspector /STP Chairman, Atomic Safety &

c/o U.S. Nuclear Regulatory Licensing Board Commission U.S. Nuclear Regulatory Commission P.O. Box 910 Washington, DC 20555 Bay City, TX 77414 Dr. James C. Lamb, III M.D. Schwarz, Jr. , Esquire 313 Woodhaven Road Baker & Botts Chapel Hill, NC 27514 One Shell Plaza Houston, TX 77002 Judge Frederick J. Shon Atomic Safety and Licensing Board J.R. Newman, Esquire U.S. Nuclear Regulatory Commission Newman & Holtzinger, P.C. Washington, DC 20555 1615 L Street, N.W.

i Washington, DC 20036 Citizens for Equitable Utilities, Inc.

c/o Ms. Peggy Buchorn Director, Office of Inspection Route 1, Box 1684

, and Enforcement Brazoria, TX 77422 U.S. Nuclear Regulatory Commission Washington, DC 20555 Docketing & Service Section I Office of the Secretary T.V. Shockley/R.L. Range U.S. Nuclear Regulatory Commission Central Power & Light Company Washington, DC 20555 P.O.' Box 2121 (3 Copies)

Corpus Christi, TX 78403 Advisory Committee on Reactor Safeguards H.L. Peterson/G. Pokorny U.S. Nuclear Regulatory Commission

! City of Austin 1717 H Street I P.O. Box 1088 Washington, DC 20555 j Austin, TX 78767 J.B. Poston/A. vonRosenberg City Public Service Board l P.O. Box 1771 San Antonio, TX 78296

• Enclosures 1 & 2 Only L1/NRC/na Revised 5/22/86 L.

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ATTACHMENT QL ST HL AE- 069

. STP FSAR PAGE I 0F /JA ,

monitoring. To accommodate this additional RCS inventory, letdown may be dis-charged from the reactor vessel head vent line with letdown routed to the pressurizer relief tank. The head vent throttling valves are controlled through the Qualified Display Processing System (QDPS), described in Section 53 7.5.6.

The safety-related cooldown is accomplished by increasing the steam release from the SG PORVs to attain a rate of primary side cooling of approximately 25'F/hr. The SG PORVs are also controlled through the QDPS. In conjunction l53 with this portion of the cooldown, the charging pumps are used to deliver water to make up for primary system contraction due to cooling. Makeup is also required for inventory control in the event the reactor vessel head vent path is periodically opened to provide head cooling. ^ alternati"e te head cerling is te previde er 9 *-^"r >eek peried fellering-RCS :: ldcur nd prier 3--

"CS d: pre::urie: tier 54 l) ^

Th: ::fety gred: .FST h:2 rd:quet: ::pecity te rcer- * -

:det: the identified :::h p ried. Upon approaching the end of this phase of cooldownyfRCS temperature of approximately 350*d, the RCS is depressurized to approximately 350 psig by venting the pressurizer through the safety-related pressurizer PORVs.

To ensure that the accumulators do not repressurize the RCS, the accumulator discharge valves are closed prior to the RCS pressure dropping below the accumulator discharge pressure. Each accumulator is provided with a Class 1E solenoid-actuated valve to ensure that the accumulator may be vented through the nitrogen supply header should the accumulator discharge isolation valve fail. A branch line inside the Containment with a parallel set of Class lE valves allows venting the nitrogen header to Containment atmosphere. 38 s'

Actuation of the SIS is precluded by use of the pressurizer low pressure and excessive cooldown signal blocks.

When the reactor coolant temperature and pressure are reduced to approximately 350' and 350 psig, respectively, the second phase of cooldown starts with the RHRS being placed in operation. Since loss of the non-safety-grade instrument l53 air system results in a loss of the air supply to the flow control valves that are normally used to limit the initial RHRS cooldown rate, the operator may choose to use only one of the RHR subsystems as a means to control cooldown rate. Should a single failure occur, such as that of an RHRS component, precluding operation of one of the RHR subsystems, the operator could elect to use a fully operational RHR train. Cooldown would continue using the fully operational RHR train (s), until the failed equipment or component could be made available. A failurs mode and effects analysis for cold shutdown operations is provided in Table 5.4.A-2.

Cooldown of the RCS is continued using available RHR trains and following cooldown rate limits. The time required to reach the cold shutdown conditions (see definition in Technical Specifications) depends upon the number of RHR trains available, and the CCW and ECW temperatures.

5.4.A-3 Amendment 53

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TA8tE 5.4.a 1 (Continind)

COMPLIANCE COMPARIsou WITN BRANCM TECM4ICAL POSITION R$e 5 1 Design Requirements process and (Systese

_of_8fP ass 5-1 PostIbte Solutton for Recomumended laptementation Degree of STP or ComponentL Full Comptionee _ for Class 2 plants

• M t,lence**

V. Test requirement Rtm tests and confine-( ing onetysis to meet Compt fence required. Meets the intent of Meet RG 1.68 for PWRs, requirement. RG 1.68. Test date

( test plus enelysis for and analysis for e i

cooldown inder natural plant simiter in circulation to confirm design to STP will edegaste mining verify adequete end cooldown within mixing and cooldoom timits specified in toder naturet circute-Emergency Operating tion conditions Procedsres. ,

(seetIon 14.2).

VI. Operationet procedsre Develop procedares and Compilence regsf red. Generic Procedures es Meet RG 1.33. For PWRs, Information from tests developed by the West-and onetysis.

include specific proce- Inghouse owners Group dJres and information for witI be used as the cooldown tnder naturet basis for plant specific circulation. procedures. 38 v!I. Auxillery Feedseter Emergency feedwater Frce tests and onetysta Stpply Compilence ullt not be

, stypt y obtain conservative required if it is shown the AFSt usebte cepecify of 525,000 sets is ade-y estimate of sumillery that en adequate etternate m Seismic Category I feedwater etwty to quote to styport 4 hrs et st@pty for euxillery Seismic Category I source hot stenchy and a 10 hr y meet requirments and le aveltable. en feedwater for et provide Seismic Cate- cooldown to RNR cut-in 54 7, teest four hours gory I supply. conditions) ottowed by at hot *'- -  ; *= S g 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> soak period plus cooldown to prior to actual eMe ini-residJet heat removal M tieti with a mergin Tor cut in teacerature con engencies. fne AFST meets Seismic Category I

\ (f rottowed ~_ seat period by en 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />

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• The innlementation for Class 2 ptents does not resutt in a mejor inpact while providing additlonal capabitity to gn to cold shutdown.

The major ippact resul t s f rren the rerpsirement for safety-relate <f stram efm p watves.

• • SIP fetit within the rategory of Class 2 plant as definerf by Section M, "Irrtmentation," of Franch Terbnical Pas i t ion PsR 5 1. Devit im 2.

ATTACHMENT .t.

ST HL-AE-176 7 PAGE 3 0Fi A STP FSAR The AWS is also designed for the following normal plant operations. l39 10.4.9.1.1 Plant Cold Startup: The AWS is designed to back up the main W system during plant startup in the event the main W system and/or the

• startup SGFP is unavailable.

10.4.9.1.2 Plant Hot Shutdown: The AWS is designed to back up the main W system during plant hot shutdown (or hot standby) in the event the main W system and/or the startup SGFP is unavailable. The AWS can be used as a means of continuous W supply even if this condition is maintained for L, extended periods. W is continuously supplied from the AFST, which during normal operation receives required makeup from the demineralized water storage tank (DWST). The DVST in turn is supplied by water from wells through the

. demineralizers, as shown on Figures 9.2.3-1 and 9.2.6-1.

10.4.9.1.3 Plant Cold Shutdown: The AWS is designed to back up the main W system when achieving plant cold shutdown.

10.4.9.2 System Description. One AWS is provided for each unit. The piping diagram is shown on Figure 10.4.9-1. The system includes an adequate l39 water storage, redundant pumping capacity to supply the SGs, associated piping, valves, and instrumentation.

The AWS supplies water to the SGs, where it is converted into steam by the heat transferred frcm the primary coolant that removes decay heat from the reactor core and heat generated in the primary coolant loop by the reactor C,. coolant pumps. 39 The AFST provides water to the AW pumps. It is a concrete, stainless steel lined, tank with a usable capacity of 525,000 gallons based on the following ilk plus a margin for contingencies; 31 e maintaining the plant in hot standby for four hours, then e cooling down the primary system to 350*F j N N

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The cooldown rate is 50*F/hr with one RCP operating or 25'F/hr with natural

circulation. During normal cooldown the rate is limited to 100*F/hr due to l39 l structural limits of the RCS components.

i Four A W pumps, each with independent motive power supplies, are provided to l46 comply with redundancy requirements of the safety standards, both for equipment and power supplies. Pump ch*aracteristics are given in Table 10.1-1. l39 Three horizontal, centrifugal, multistage, electric motor-driven pumps supply one SG each. Each pump motor is supplied power from a separate engineered safety bus, and the power supply is separated throughout. ,

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l 10.4-29 Amendment 54

ATTACHMENT J-ST44L AE l?6 ?

PAGE 4 OF 2, STP FSAR

7. Uncoupling of Human Errors This study assumes that test and maintenance activities are staggered. That is, redundant AFWS components are not tested by the same personnel on the same shift, but in general, tests and/or maintenance of redundant componepts involve time and/or personnel changes (e.g., different personnel and shifts, or the same personnel on a different day, etc.) In addition, a double-check procedure is assumed to assure the correct status of locked open valves after test and maintenance. This signif-icantly reduces the probability of human error in two or more trains simultaneously. Given that test and maintenance activi-ties are staggered and the use of a double check-procedure, it

. is reasonable to assume that human errors for test and mainte-nance are uncoupled.

For the above reasons, the evaluation does not consider con-current disabling of multiple trains because of human error in conjunction with test or maintenance to be a credible failure scenario.

8. Technical Specification The auxiliary feedwater system design is evaluated in accor-dance with the STP Technical Specifications (Ref. 7).

Train A - Availability is assumed to be degraded since there is 54 no Technical Specification requirement on Train A.

Trains B, - Operable except for the scenarios C, and D illustrated in the fault trees in Section 10A.3.2.

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. 9. HVAC Support The motor driven auxiliary feedwater pump rooms are cooled by safety-related HVAC units powered by their respective trains.

The turbine driven pump room is cooled by a Train A HVAC unit, however, the turbine driven pump is qualified for operation following the loss of all HVAC. Consistent with NUREG-0611 methodology, HVAC support to the pumps is not considered in this evaluation, i

10. Auxiliary Feedwater Storage Tank The AFWST capacity is sufficient to allow the RCS to remain at hot standby for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> fol16ved by a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> cooldown-f d r 4F1r' lim cr-b --- bd at which point further RCS cooldown is per-formed by the residual heat removal system. If additional quantities are needed, water can be provided to the AFWST from the demineralized water storage tank, the condenser hot well, or an alternate onsite source. The AFWST has level instrumen-tation with control room indication and annunciation'to warn operators of low AFWST water inventory.

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10A-4 Amendment 54

ATTACHMENT Ps ST-HL AE lHo'l PAGE 5 OFIA wiiL o ye M' v'r6u <ut. g STP FSAR , stro .M

4. Auxiliary Fe dwater Storage Tank The Seismic Category I auxiliary feedwater storage tank pro-vides water to the AIV pumps. It is a concrete, stainless steel line 519,000 g;11;r tank rhi;h has sufficient capacity to allow t RCS to remain at hot standby for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> followed by a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> cooldown - ' er ? i:__ :::'; ;:ri;d'at which point further RCS cooldown ks performed by the residual heat removal system.

The AFWST is designed to withstand environmental design condi-tions, including floods, earthquakes, hurricanes, tornado load-ings, and tornado missiles. The AFWST is designed so that no single active failure will preclude the ability to provide water to the AFW system. Each train has a dedicated suction line from the AFWST to the AFW pumps. The water level in the AFWST is indicated in the control room as well as at the auxiliary shutdown panel. A low level alarm is also provided in the control room.

10A.2.3 Emergency Operation The AFWS is designed for automatic actuation in an emergency. Any of the following conditions automatically starts the three Class 1E motor-driven Pumps:

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, . 1. Two out of four channels showing low-low water level in any steam generator

2. Safety injection signal

3. 4.16 kV bus undervoltage. The AFW pump is started in conjunction with diesel generator starting and load sequencing.

Water is not automatically fed to the steam generator until condition 1 or 2 above exists.

The turbine-driven auxiliary feedwater pump starts automatically on any of the following signals:

1. Two out of four channels showing low-low water level in any steam generator

2. Safety injection signal A one-inch bypass line with a normally closed solenoid operated valve (FV0143) and orifice is provided around the steam inlet valve (MS0143). This bypass valve (FV0143) opens upon receipt of either of the above signals to supply steam to the turbine and allow the turbine to reach governor control speed.

After a time delay to allow governor control speed to be reached, the steam inlet valve is opened which allows rated steam flow to the turbine. This arrangement precludes an overspeed trip due to excessive steam flow prior to governor warmup. This bypass line is not dependent upon AC power to operate.

10A-7 Amendment 54

- ATTACHMENT 3 STP FSAR ST-HL AE IPo'l PAGElo OF i 1.

Question 440.30N With regard to the information in Appendix 5.4A " Cold Shutdown Capability" identify the most limiting single failure with regard to cooldown capability and verify that the statement of Table 5.4A-1 that the auxiliary feedwater storage tank (AFST) " capacity of 500,000 gallons is adequate to support 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at hot standby conditions followed by 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> cooldown to RHR cut in condition with a margin for contingencies" considers this failure.

Response W The most limiting failure regarding cooldown time is the loss of "A" train AC power, which results in the loss of two steam generator PORVs. RHR cut-in conditions can be achieved with this failure-M-hours after reactor trip based on maintaining hot standby for four hours followed by a8 ten hour natural circulation cooldown,and ther en eight heur ::ch period ' Approximately

%~ 0 -447,000 gallons of water would be added to the effective steam generators 54 during this period. .

ed a ld Specifically the AFST sizing considers: 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at hot standbyA # hour natural circulation cooldown " L m seak period?' It also considers possible level instrument error, water delivered to a faulted steam generator, water lost through the turbine lube oil cooler, various small system water losses (ie.,

flange or pump seal leakage) h d a margin against vortex formation. The usable volume in the AFST above the suction nozzles is 525,000 gallons.

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Vol.2 Q&R 5.4-5N Amendment 54

ATTACHMENT ;L.

ST.HL-AE l?67 STP FSAR PAGE 9 0F IJ-Question 440.38N

a. Demonstrate that the STP ECCS meets 10 CFR Part 50.46 criteria for long 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 (LHSI) pump shutoff head or recovers after loss of the secondary heat sink. An examination of Figures 6.3-1 through 6.3-5 does not indicate that the STP ECCS is designed for*high-

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head recirculation combined with decay heat removal by the RHR heat exchangers, i.e., there are no apparent provisions for routing recircula-tion flow from the RHR heat exchangers to the HHSI pumps. Also, as described in Appendix 5.4.A " Cold Shutdown capability," the steam genera-tors have a limited supply of safety grade secondary water supply, since there is not a safety grade backup to the auxiliary feedwater storage tank (AFST). Therefore, provide long term analyses for a spectrum of small break LOCAs that demonstrate that decay heat can be adequately removed and the RCS 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 those recommended 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.

RWST and the steam generators would provide core cooling for approxi-mately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />, after which the reactor containment fan coolers (RCFCs) would provide an adequate heat sink for decay heat removal. No credit is l taken for heat removal by the RHR heat exchangers. Provide a detailed explanation of the mechanism of energy removal from the RCS after loss of the secondary heat sink and supporting analyses that demonstrate that energy can be adequately removed to meet the acceptance criteria of 10 CFR Part 50.46. We are concerned that for very small break LOCAs (e.g.,

, l~ 1 inch) energy would not be adequately removed from the RCS for a consi-derable period of time after the accident. Thus, WCAP-9600, " Report on Small Break Accidents for Westinghouse NSSS System" June 1979, indicates that for 1 inch breaks the break can remove all the decay heat 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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Vol.2 Q&R 6.3-8N Amendment 49

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ATTACHMENT A ST HL-AE 1%7 PAGE R OF19.

Insert X Larga Break LOCA For large break LOCA (breaks greater than 1 sq. ft.) the break will cause a significant Reactor Coolant System (RCS) depressurization. 3reaks of this size are not isolable so the sump is used for long term cooling and makeup.

Breaks considered large breaks will have sufficient energy removal through the break to sump flow path to remove decay heat' energy. Sufficient make-up capability to keep the core adequately cooled and to meet 10CFR Part 50.46 (b)

(5) requirements is provided. Containment heat removal will be provided in

, the STP design by both containment fan coolers and low head safety injection (LHSI) recirculation flow which is cooled by the RHR heat exchangers.

Equipment relied upon is fully qualified for the environmental conditions that prevail during the accident.

Small Break LOCA As result of the accident at Three Mile Island Unit 2, Westinghouse performed extensive analyses that focused on the behavior of small break loss of coolant accidents (SBLOCA) for the Westinghouse NSSS. The purpose of the

, analyses was to demonstrate adequacy of the Westinghouse NSSS design in mitigation and long term recovery from a range of breaks classified as small breaks (less than 1 sq. ft. area).

. The results of the analyses were reported in WCAP-9600, " Report on Small Break Accident for Westinghouse NSSS System," dated June 1979. The "Small j Break Evaluation Model" at that time consisted of the WFLASH thermal-hydraulic i

code and the LOCTA fuel rod model. The analyses were performed for generic application using a standard 4-loop Westinghouse designed, a standard 3 loop and standard 2 loop depending on the nature of the study and which plant type was expected to be bounding. The conclusions are applicable for all Westinghouse designs, including STP with exceptions as described in the following.

STP SBLOCA Design Features i STP has a three train low pressure SI system consisting of three high head SI (HHSI) pumps, three DiSI pumps, and three accumulators. Each train is aligned to a separate RCS loop. The pressure ranges for the SI pumps follow:

HHSI: 0 - 1445 psig LHSI: 0 - 283 psig 1

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ATTACHMENT L ST.HL AE 1969

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Insert X (Continued)

For recirculation, the LHSI and HHSI pumps take suction directly from the sump.

The LHSI pump flow passes through the RHR heat exchanger and is cooled before entering the RCS.

The plant has three motor driven auxiliary feedwater (AFW) pumps and one turbine driven auxiliary feedwater pump. The normal system alignment connects each AFW pump directly to one steam generator. The steam does not have a common header, but cross connections exist in the AFW lines. The valves in the cross connections are normally closed and fail closed. Two motor driven AFW pumps and the turbine driven AFW pump are required operable by the Technical Specifications.

The Auxiliary Feedwater Storage Tank (AFST) has a useable capacity of 525,000 gallons.

Non-safety grade sources of condensate grade make-up to the AFST are:

o Demineralized Water Storage Tank - One 1,000,000 gallon storage tank shared between units.

o Secondary Make-up Tank - One 300,000 gallon storage tank per unit o Condenser Hotwell - about 100,000 gallons per unit While these tank volumes are not covered by Technical Specifications or other administrative controls, it would be very improbable to have less than 500,000 gallons of condensate grade water available for each unit.

The limiting single failure for the STP design will result in the loss of one train of safety injection (1 LHSI and HHSI pump) and one AFW pump.

Since one AFW pump is allowed out-of-service for maintenance, this will result in the ability to feed two steam generators.

The STP design provides means to remove energy through the steam generators (AFW and atmospheric relief valves), through Containment steam condensation (fan coolers) and through the RHR heat exchangers (LHSI pumps and RHR heat exchangers). In this way energy is removed from containment sump water (RHR heat exchangers) so that relatively cool water will be continued to be supplied as make-up and for decay heat removal.

For all break sizes, heat is removed from the core by the break and steam generators. AFW is required for secondary inventory and heat removal until the break is able to remove all the decay heat or the RHR System is placed in operation. The break removes energy from the RCS because the makeup water from the RWST is relatively cold and can absorb energy before exiting the RCS.

The WCAP-9600 analyses with consideration of STP design features and STP analyses of long term cooling discussed in the report titled "Long Term Cooling Analysis for South Texas Project" demonstrate decay heat removal capability for SBLOCA. The Long Term Cooling Report was transmitted in HL&P letter ST-HL-AE-1767 dated September 30, 1986.

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ATTACFDAENT 3, ST-HL AE- libi Insert X (Continued)

SBLOCA Response The initiating event is the break. If the break is 3/8" or less equivalent diameter and the charging system and feedwater system are available, the event is classified as a leak since normal charging flow would be sufficient to keep up with leak flow without a significant RCS depressurization. There would not be an automatic reactor trip or safety injection signal.

For breaks larger than 3/8", automatic reactor trip and safety injection

, will occur due to RCS depressurization caused by the loss of primary inventory. After reactor trip and safety injection initiation, safety injection pump flow provides makeup to the RCS and maximum peak clad temperature will remain below 10CFR50.46 Appendix K criteria.

For breaks greater than 3/8" and less than 1.5", SI flow can match break flow so no significant RCS depressurization or core uncovery will occur. At the point where SI flow matches break flow, the mitigation phase of the

, accident ends and a long term decay heat removal phase begins. The operator will cool down and depressurize to below the shutoff head pressure of the LHSI l pumps (283 psig). This will be accomplished using the steam generator PORVs for i cooldown and pressurizer PORVs in combination with HHSI flow termination for depressurization. The detailed actions will be provided in the STP Emergency i Procedures which are based on the WOG Emergency Response Guidelines. The RHRS i

will be available to provide heat removal at RCS pressures below 350 psig and temperatures below 350 F. Adequate long term decay heat removal will be provided by LHSI pump flow through an RHR heat exchanger in addition to RHRS operation.

! For breaks from 1.5" to 4", the operator will cool down and depressurize i the RCS to a pressure below the shutoff head pressure of the LHSI. The combined heat sink capacity of the Refueling Water Storage Tank and the steam i generators would provide core cooling until the containment fan coolers and 1 the RHR heat exchangers via LHSI pumps provide an adequate heat sink for decay removal.

For breaks greater than 4", the decay heat will be rencved by the break and the containment fan coolers and the RHR heat exchangers via LHSI pumps. No i operator action is required.

For isolable breaks, the operator will cool down and depressurize the RCS

via a sufficient quantity of auxiliary feedwater to RHRS cut-in conditions of i

350 psig and 350 F. Adequate long term decay heat removal will then be provided via the Residual Heat Removal System, f

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Ll/NRC/na Revised 5/22/86 1

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ATTACHMENT 2.

ST HL AE- 06')

+ STP FSAR PAGE 11 OF 19-Question 440.39N

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 and decay heat shall be removed for the extended period of time required by the long-term radioactivity remaining in the core." In order to assure this, heat removal for this extended period must* utilize equipment that is fully qualified for the environmental conditions that prevail during the accident. Please demonstrate that decay heat can be removed from the STP core with qualified equipment only, following all sizes of LOCAs, including all LOCAs which could be subsequently isolated by the operator. Include consideration of the post-LOCA cooldown period in your response, and the fact that for isolated LOCAs, the sump would not be available for long term cooling.

b. Discuss whether the RHR pumps are qualified for the environmental effects of the large and small break LOCAs and steam line breaks. If the RHR pumps are not qualified discuss how long term mitigation of these acci-dents would be accomplished.

Response

• 111 A p e *

- M == ndment.

a. See the response to NRC Question 440.038N.

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b. The RHR pumps are qualified for the containment environment following a DBA, including small LOCA and secondary breaks. The NRC was notified regarding the qualification of the RHR pumps via ST-HL-AE-1684 dated June 17, 1986.

Vol.2 Q&R 6.3-9N Amendment 49

. ATTACFDAENT QL ST-HL AE 1967 PAGEla-OF19-EMERGENCY CORE COOLING SYSTEMS 3/4.5.6 RHR LIMITING CONDITION FOR OPERATION 3.5.6 Three independent Residual Heat Removal (RHR) loops shall be OPERABLE with each loop comprised of:

a. One OPERABLE RHR pump,

b. one OPERABLE RHR heat exchanger, and

c. one OPERABLE flowpath capable of taking suction from its associated RCS, hot leg and discharging to its associated RCS cold leg.

APPLICABILITY: Modes 1, 2 and 3 ACTION:

a. With one RHR loop inoperable, restore the required loop to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,

b. With two RHR loops inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. With three RHR loops inoperable, immediately initiate corrective action to restore at least one RHR loop to OPERABLE status as soon as possible.

SURVEILLANCE REQUIREMENTS 4.5.6 Each RHR loop shall be demonstrated OPERABLE pursuant to the requirements of specification 4.0.5 South Texas Project 3/4 5-11 Sept. 30, 1986

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Enclosure 3 e

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