ML20136G621
| ML20136G621 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 11/18/1985 |
| From: | Kadambi N Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8511220402 | |
| Download: ML20136G621 (43) | |
Text
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,. ,f j NOV 181985 Docket Nos: 50-498 and 50-499 APPLICANT: Houston Lighting & Power Company FACILITY: ~ South Texas Project, Units 1 and 2
SUBJECT:
SUMMARY
OF MEETING HELD ON OCTOBER 24, 1985 TO DISCUSS LONG TERM COOLING The applicant requested this meeting to discuss long term cooling issues which are associated with responses to several requests for additional information as well as forthcoming amendments to the FSAR. The attendees at the meeting are shown in Enclosure 1. The applicant provided draft copies of responses which were used as the subject for discussion. These drafts are enclosed as R Enclosure 2.
' Discussion:
The applicant informed the staff that the design of the South Texas Project will incorporate fully qualified RHR pumps. However, due to unavailability of the required pump motors in time for the current licensing schedule, staff approval was requested for a temporary design which involved making a hose connection between the essential cooling pond and the auxiliary feed-water storage tank (AFWST) in the event of an accident which depleted the AFWST.
The interim and long term design basis was presented in a table shown as Enclosure 3. The' staff indicated that the request would be_ considered as part of the ongoing licensing review. Dr. Sheron stated that, if approved, the hose connection, supported by training and procedures, would form the temporary licensing basis for the plant, with no equipment qualification -
exemption being required for the RHR pumps in the interim period. The applicant connitted to address operator action times for the various accident scenarios.
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/s/
N. P. Kadambi, Project Manager
.- Licensing Branch No. 3 l Division of Licensing F
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Enclosures:
As stated
! Distribtuion:
L cc: See next page Docket File 50-498/499 l PRC System BSheron
!" LB#3 Reading OELD l *Previously concurred on by: NPKadambi 0 JLee
- DL:LB#3 *RSB r DL:LBf3 $
NPKadambi/es BSheron WKnight n l
11/11/85 11/12/85 11/(T/85 I .
l i- 851122O402 851118 PDR ADOCK 05000498 A PDR
"/eyn ttag% UNITED STATES 8 i NUCLEAR REGULATORY COMMISSION 5 g WASHINGTON, D. C. 20555
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Docket Nos: 50-498 and 50-499 APPLICANT: Houston Lighting & Power Company FACILITY: South Texas Project, Units 1 and 2
SUBJECT:
SUMMARY
OF MEETING HELD ON OCTOBER 24, 1985 TO DISCUSS LONG TERM COOLING The applicant requested this meeting to discuss long term cooling issues which are associated with responses to several requests for additional information as well as forthcoming amendments to the FSAR. The attendees at the meeting are shown in Enclosure 1. The applicant provided draft copies of responses which were used as the subject for discussion. These drafts are enclosed as Enclosure 2.
Discussion: -
The applicant infonned the staff that the design of the South Texas Project will incorporate fully qualified RHR pumps. However, due to unavailability of the required pump motors in time for the current licensing schedule, staff approval was requested for a temporary design which involved making a hose connection between the essential cooling pond and the auxiliary feed-water storage tank (AFWST) in the eyes.t of an accident which depleted the AFWST.
The interim and long term design basis was presented in a table shown as Enclosure 3. The staff indicated that the request would be considered as part of the ongoing licensing review. Dr. Sheron stated that, if approved, the hose connection, supported by training and procedures, would form the temporary licensing basis for the pir.t. with no equipment qualification exemption being required for the RMR pumps in the interim period. The applicant committed to address operator action times for the various accident scenarios.
O JA.
N. P. Kadambi, Project Manager
, Licensing Branch No. 3 Division of Licensing i
L
Enclosures:
As stated cc: See next page l
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Mr. J. H. Goldberg Houston Lighting and Power Company South Texas Project
~ CC'- 1 Brian Berwick, Esq. Resident Inspector / South Texas Assistant Attorney General Project Environmental Protection Division c/o U.S. Nuclear Regulatory Commission P. 0. 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 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 78296 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 '
l 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 l Mrs. Peggy Buchorn Executive Director Citizens for Equitable Utilities, Inc.
Route 1, Box 1684 Brazoria, Texas 77422 I
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. .' e Houston 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 Citizens Concerned About Nuclear Power 3022 Porter Street, NW #304 Washington, D.C. 20008 Mr. S. Head, Representative Houston Lighting and Power Company Suite 1309 7910 Woodmont Avenue Bethesda, Maryland 20814 S
J
, s' Enclosure 1 HL&P AND NRC LONG TERM COOLING MEETING 10/24/85 NAME ORGANIZATION ;
Jack Bailey HL&P Engineering Hulbert Li NRC/ICSB Bernard Mann NRC/RSB Jeff Phelps HL&P Licensing Raj P. Goel NRR/DSI/ASB Robert Jakub W Nuclear Safety Bruce Monty W Nuclear Safety Jerry Wilson NRC/ASB/RSB Bill Watson .Bechtel Engineering Mark Wisenburg HL&P Licensing N. P. Kadambi NRC/NRR/DL Bill Spezialetti W Nuclear Safety Robert E. Sweeney Ebasco - HL&P Bethesda Office George W. Knighton NRC/DL/LB3 Tad Marsh NRC Brian Sheron NRC*
Errol Dotson HL&P*
- Part time
/
, i ENCLOSURE 2 DRAFT RESP 0f:SES TO RAI'S 5
0
% 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 PER cut in l condition with a margin for contingencies" considers this failure.
Resoonse O_; ::q - u_ :: "'- anastion will be provided in a later *** d--. .
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J Vol.2 Q6R 5.4 5N Amendment 49
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Question 440.30 Response The most limiting failure regarding cooldown time is the loss of "A" train AC power, which results in the loss of two team generator PORV's. RHR cutin conditions can be achieved 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after reactor trip based on maintaining hot standby for four hours followed by a ten hour natural circulation cooldown and then a six hour soak period. Approximately 420,000 gallons of water would be added to the steam generators during this period.
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 standby, 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> natural circulation cooldown, 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> soak period. It also considers possible level instrument error, water lost through the turbine lube oil cooler, various small system water losses (ie. flange or pump seal leakage) and a margin against vortex formation. The net useable volume in the AFST is 445,000 gallons.
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. STP FSAR The AFWS is also designed for the following normal plant operations. 39 10.4.9.1.1 Plant Cold Startup: The Afv5 is designed to back up the main FW system during plant startup in the event the main FW systes and/or the startup SGFP is unavailable.
-' . 10.4.9.1.2 Plant Ect Shutdown: The AFWS is designed to back up the main sy systes during plant hot shutdown (or hot standby) in the event the main FW system and/or the startup SGFF is unavailable. The AFWS can be used as a 45 seans of continuous FW supply even if this condition is maintained for extended periods. FW is continuously supplied from the AFST, which during normal operation receives required makeup from the desin.ralised water storage tank (DVST). Thn- DWST in turn is supplied by water from wells through the domineralizers, as shown on Figures 9.2.3-1 and 9.2.6-1.
10.4.9.1.3 Plant cold shutdown: The AFWS is designed to back up the main FW system when achieving plant cold shutdown.
10.4.9.2 System Description. One AFWS is provided for each unit. The piping diagram is shown on Figure 10.4.9-1. The system includes an adequate l 39 water storage, redundant pumping capacity to supply the SGs, associated piping, valves, and instrumentation.
The AFWS' supplies water to the SCs. where it is converted into steam by the heat transferred from the primary coolant that removes decay heat from the reactor core and heat generated in the primary coolant loop by the reactor l 39 coolanc, pumps. ,, ,
) The AFST provides water to, the AFV pumps. It is a concrete, stainless steel lined. 500.000 gallon tank With capacity based *on:
31 e- maintaining the plant in hot standby for four hours, then
-- " f .~.-f e cooling down the' primary system to 3,5,0*F,
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The cooldown rare is 507/Wr wTttrone RCP operating or 25'7/hr with natural circulation. During normal cooldown the rate is limite.d to 100*F/hr due to l39 structural limits of the RCS components.
Four AFW pumps, each with independent active power supplies, are provided to l46 comply with redundancy requirements of the safety standards, both for equipment and power supplies. Pump characteristics are given in Table 10.1-1. [39 Three horizontal, centrifugal, multistage, electric actor-driven pumps supply one 50 each. Each pump motor is supplied power from a separate engineered safety bus, and the power supply is separated throughout.
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10.4-29 Amendment 46
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STP FSAR TABLE $.4.A-1 COMPLEANCE COMPARI$ alt WITN BRANCN TECIRelCAL 7051710N RSS 5-l ,-
4 e Destan seguiremente Process and [5yeten Fossible Setutten for secommended Implemenvatten Degree of STF of BTP B5s S-1 or Ceepenent) Fell Compifance for Class 2 plantes Compliance *n W. Test regelrement Boa teste edJ confirm. Ceepliance regetred. Iseets the intent of tag analysts to meet BC l.64. Test data Meet BC a.&0 for FWWe, regelresent. and analyste for a test P I wo analyste for plant sletter in coeldown under natural deelen to STP will strculatten to esefire verify adequate adequate statsg etsing and emeldoen and coeldown within ander natural carcela.
- Itatte spectfled in tien eendittene Faergency Operating (Sectlen 14.1).
Procedures.
VI. Operettenal precedure Develop procedeces and Compliance regelred. Senerte freredores se taformatten free teste developed by the West =
teeet RC 1.33. For FWRe. and analyele. Inghouse Owners Creep include spectifs prece- will be used as the deres and soformatten for beste for plant specifte u cooldown under natural precedures.
- a. etweeletten.
1 911. Assillery Feedwater Emergency feedvetor From teete end analyste f*eepitence will not be The AFST espeelty of Supply supply obtain eeneervative regelred if it to shove 500.000 mets to adegeste setteste of mentitary that en adequate alternate to support 4 bra et het 34 Seteste Category I feedwater supply to Setente Category I sentee atendby followed by espply for austilary
- eest regetroente and to evettable. 10 hre coeldownj e put feedveter for et provide Setente Cate- cut ineenettienekwitha least four hours g gery I supply, earsin for contisseneles.
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W an The AyST meets Salente at het _ -- y Category I regeltenente plus reeldown to al best removal (Seetten 10.4.7) cut-tag booed se long. . .
est time for only onette er sely effette power a.d see e4 sto,1.
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- The implementattee for Close I plante does not result la e major topact while providing addittenal cepehtitty to go to cold ehetdown.
The major topect results free the requirement for safety-related stese deep volves.
ee STP f alle althin the careparv of a Claae y plaat an defined hv tectlen M. "fpplementattan." of 9 ranch Techntent Pasttenn S AS g-1. Sevtegan 2.
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STP M Ek t#M S ~ 2' - 4 C.O d\ S 5.2.2.9 System Reliability. The reliability of the pressure-relieving devices is discussed in Section 4 of Reference 5.2-1. . g 5.2.2.10 Testing and Insp'etion. e Testing and inspection of the over-pressure protection components are discussed in Subsection 5.4.13.4' dnd Chapter 14.
5.2.2.11 RCS Pressure Con'. col During Low Temperature Operatio,n_.
Administrative procedures are structured to aid the operator in controlling Reactor Coolant System Pressure during low temperature operation. However to provide a backup to the operator, an automatic system is provided to g intain pressures within allowable limits.
M 5.2.2.11.1 System Operation: Each of the two pressurizer power-operated relief valves is supplied with actuation logic to ensure that a completely automatic and independent RCS pressure control back-up feature is provided for the operator during low temperature operations. This system provides the capability for additional RCS inventory letdown, thereby maintaining RCS pressure within allowable limits. Rafer to Sections 5.4.7. 5.4.10, 5.4.13, 7.6 and 9.3.4 for additional information on 9 i
ECS pressure and inventory control during other modes of operation. Q211
.. 02 co. Nalk usly monitor RCS The basicand temperature function pressureofconditions, the system logic with the logicis4 to contin [ armed whenever plant operation is at a temperature below 350*F. An auctioneered system #
temperature will be continuously converted to an allowable pressure and thencomparedtotheactualRCSpressure.f is comparison will provide an actuation signal to the power-operated relief valves when requirad, to
%i prevent pressure-temperature conditions from exceeding the allowable g inits. -
See Section 7.6 for a further discussion of system logic.
5.2.2.11.2 Evaluation of Low Temperature overpressure TransientcX.*ete-ASME Section III. Appendix G. establishes guidelines and limits for RCS Pressure primarily for low temperature conditions ($350*F).
Transient analyses were performed to determine the maximum pressure for the postulated (credible) worst case mass input and heat input events.
I ansient. wn1 h wou d occur no. frequ6pt'y during m 31
% h ico_e maof inp pres. m q the a plan pe ti would M p p' de ering uput than e al to ve letd isolatio'swith gal / min ov'ever c
he ging '
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wa ehdsen to )(ov de ad gional dh tem flex, ilitifory ensure coner
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' The heat input analysis was performed for an inadvertent reactor coolant pump start assuming that the RCS was water solid at the initiation of the event sud
]C that a 50*F mismatch existed between the RCS and the secondary side of' the oteam generators.
(At lower temperatures, the mass input case is the limiting transient condition.) .
5.2-4 Amend:ent 31
INSERT A Analyses have shown that one pressurizer power operated relief valve is sufficient to prevent violation of these limits due to anticipated mass and heat input transients. Redundant protection against a low temperature overpressure event is provided through the use of two pressurizerher cperated relief valves to mitigate potential pressure transients. t The automatic system is required only during low temperature water solid cperation when it is manually armed and automatically actuated. p lb f l As described in subsection 5.4.13, the STP PORVs are safety related. They an e designed in accordance with the ASME Code and are qualified via the tiestinghouse pump and valve operability program which is described in seclim
'z. Offsite power is not required for the system to
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function. Theactuationlogicinthesystemistestable1.ThePORVsarenot - " j"
- 6 l'owever, they are capable of being Cxercised tested as required with theby reactor the ASME at Code power,an)d the STP Technical Specifications.
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j The system logic will first annunciate a main control board alam whenever the measured RCS pressure approact;es the allowable pressure. . .
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0940n/RHF/10-85 *
'INSERI C The mass input transient is divided into two parts for plant operation in Mode 4 (above 2000F) and Mode 5 (1 2000F). In Mode 4, the mass input transient assumes the operation of one high head SI pump and one centrifugal charging pump delivering normal charging flow with letdown isolated. In Mode 5, the mass input transient assumes the operation of one centrifugal pump with letdown isolated. t A GNV "
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s Both hact input cnd mass input ennlyses took into account thz single failure criteria and therefore, only one Power Operated Relief Valve (PORV) was assuned to be available for pressure relief. The evaluation of the transient 31 results conclude that the allowable limits will not be exceeded and therefore jj will not constitute an impairment to vessel integrity and plant' safety. ( j Q. L
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5.2.3 *> Reactor Coolant Pressure Boundary Materials i 5.2.3.1 Material Specifications. Material specifications.used for the principal pressure-retaining applications in each component constitut-ing the RCPB are listed in Table 5.2-2 for ASME Class 1 primary components and Table 5.2-3 for ASME Class 1 and 2 auxiliary components. These tables also include the unstabilized austenitic stainless steel material specifica-tions used for components in systems required for reactor shutdown and for emergency core cooling. The unstabilized austenitic stainless steel material for the reactor vessel internals which are required for emergency core cooling for any mode of normal operation or under postulated accident conditions and for core structural load-bearing members are listed in Table 5.2-5. All of the materials utilized conform with the material specification
. requirements and include the special requirements of the ASME Code, Section III, plus addenda and code cases as are applicable and appropriate to meet Appendix B of 10CFR50. The listed specifications in Table 5.2-3 l are respresentative of those materials utilized.
The welding materials used for joining the ferritic base materials of the
'RCPB conform to or are equivalent to ASME Material Specifications SEA 5.1 5.2, 5.17, 5.18, and 5.20. They are test,ed and qualified to the require-ments of ASME Code, Section III. In addition, the ferritic materials of the reactor vessel belt line are restricted to the following maximum linits of copper, phosphorous, and vanadium to reduce sensitivity to irradiation embrittlement in service. .
i' t 5.2-4a A endmer.t 31 v . , , - - _ - , , , , , , - - , , - , , . - . . - , - - - - , . - - , _ - , , - - . , , - - + , - . . . , , _ , , _ _ ,-----,-------,---,,,---cn .-,-n- - --,-------,----e,-
c INSERT D 5.2.2.11.3 Administrative Procedures l Although the system described in Section S.2.2.11.1 is provided t.o; maintain RCS pressure within allowable limits, administrative procedures minimize the potential for and the consequences of any transient that could actbate the overpressure relief system. The following discussion highlights these procedural controls. Of primary importance is the basic method of operation of the plant. Normal plant operating procedures will maximize the use of a pressurizer steam bubble during periods of low pressure, low temperature operation. This steam bubble will dampen the plants' response to potential transient generating inputs, providing easier pressure control with the slower response rates. A steam bubble substantially reduces the severity of potential pressure transients, such as reactor coolant pump induced heat input, and slows the rate of pressure rise for others. In conjunction with the alarms discussed in Section 7.6, this provides reasonable assurance that most potential transients can be terminated by operator action before the overpressure relief system ! actuates. ' However, for those modes of operation when water solid operation may still be possible, procedures will further highlight precautions that minimize the severity of, or the potential for, developing an overpressurization transient. The following precautions or measures are considered in developing cperating procedures. l a. Whenever the plant is water solid and the reactor coolant pre'ssure is l being maintained by the low pressure letdown control valve, letdown flow normally bypasses the normal letdown orifices. In addition, all three l 1etdown orifices normally remain open. L
- b. If all reactor coolant pumps have stopped for more than 5 minutes during plant heat up and the reactor coolant temperature is greater than the charging and seal injection water temperature, a steam bubble will be formed in the pressurizer prior to restarting a reactor coolant pump.
This precaution minimizes the pressure transient when the pump is started and the cold water previously injected by the charging pumps is circulated through the warmer reactor coolant components. The steam bubble will acconnodate the resultant expansion as the cold water is rapidly warmed.
- c. If all reactor coolant pumps are stopped.and the RCS is being cooled down by the residual heat exchangers, a nonuniform temperature distribution may <
occur in the reactor coolant loops. Prior to restarting a reactor coolant pump, a steam bubble will be formed in the pressurizer or an acceptable temperature profile will be demonstrated.
- d. During plant cooldown, all steam generators will normally be connected to the steam header to assure a uniform cooldown of the reactor coolant loops.
- e. At least one reactor coolant pump will normally remain in service until l
! the reactor coolant temperature is reduced to 1600F. tkN. [D, l V l l 0940n/RHF/10-85
x, , . These special precautions back-up the normal cperaticnal mode of maxinizing periods of steam bubble operation so that cold overpressure transient 1 prevention is continued during periods of transitional operations. These precautions do not apply to reactor coolant system hydrostatic testing. The specific plant configurations of emergency core cooling system testing and alignment will also highlight procedural recommendations to preven) develeping cold overpressurization transients. During these limited periods of plant cperation, the following precautions / measures are considered in developing the cperational procedures:
- a. To preclude inadvertent emergency core cooling system actuation during heatup and cooldown, procedures require blocking the pressurizer pressure, i and 1_ pteem lir.e pressere signal actuation logic at 1,900 psig.
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- b. During further cooldown, closure and power lockout of the accumulator isolation valves and power lockout of the nonoperating charging pumps and ss; safety injection pumps will be performed at 1,000 psig, approximately
'h,3; ; 425 F RCS conditions, providing additional back-up to step a above.
- c. The reconnended procedure for periodic emergency core cooling system pump performance testing will be to test the pumps during normal power operation or at hot shutdown conditions. This precludes any potential for developing a cold overpressurization transient.
Should cold shutdown testing of the pumps be desired, the test will be done when the vessel is open to atmosphere, again precluding overpressurization potential. fc s o testin Ti the' eactpr-viFssel ci sed is necess11% the pr e es r utre rgency cor c -Ting. system p ps arge 1ve l ure an. RHR ignment to ate potential ncy core c ing L- svstem nu put and to pr de back-up benefit of the RHRS Igj Ey='":..
- d. SIS circuitry testing, if done during cold shutdown, requires nonoperating safety injection pumps power lockout to preclude developing cold overpressurization transients.
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[ .' . RESPONSE TD QUESTION 440.39A l Leroe Break LOCA For large break LOCA (breaks greater than 1 sq. ft.) the break will ecuse a significant Reactor Coolant System (RCS) depressurization. Breaks of this size are not isolatable so the sump is used for long term cooling cnd makeup. All breaks considered large breaks will have sufficient enrrgy removal through the break to sump flow path to remove decay heat cncrgy. Sufficient make-up capability to keep the core adequately cooled cnd to meet 10CFR Part 50.46 (b) (5) requirements is provided. Containment hast removal wil,1,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. All equipment- relied upon is f ully qunlified for the environmental conditions that prevail during the cccident. Smm11 Break LOCA As a result of the accident at Three Mile Island Unit 2, Westinghouse parformed 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 W2ctinghouse NSSS design in mitigation and long term recovery f rom a range of breaks classified as small breaks (less than 1 sq. ft area). The results of the anal yses were reported in WCAP-9600, " Report on Small Br eak Accident f or Westinghouse NSSS System," dated Jurse 1979. The "Small Breal Evaluati on' Model " at that time consisted of the WFLASH
Yhermal-hydrculic coda cnd th] LOCTA fugl rod mod 21. Th2 cnnlycan were parformed for censric rpplicction using a ntcndtrd 4-loop Wsztinghouse dacign, a standard 3 loop and standard 2 loop depending on the nature of th? study and which plant type was expected to be bounding. The conclusions are applicable for all Westinghouse designs. STP SBLOCA Desian Features STP has a three train low pressure SI system consisting of three HHS1 pumps, three LHSI pumps, and three accumulators. Each train is aligned to e ceparate RCS loop. The pressure ranges for the SI pumps follow: HHSI
}&W O - b6ES PSIG LHSI O - 283 PSIG For recirculation, the LHSI and HHSI pumps take suction directly from the cump. The LHSI pump flow passes through the RHR heat exchanger and is cooled before entering the RCS.
l The plant has three motor driven auxiliary feedwater pumps and one , turbine driven auxiliary feedwater pump. The normal system alignment ccnnects each AFW' pump directly to one steam generator. The system does not have a common header, but cross connections exist in the AFW lines. Tha valves in the . cross connections remain closed normally.
.The limiting single failure for the STP design will result in the loss of one train of safety injection (1 LHSI and HHS1 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, even though successful cooling can be accomplished by feeding one steam generator.
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. Th2 STP dvsign provid s maans to rcmove Gnsrgy through the steam cinerators (long term AFW) , through containment steam condensation (fan coolers) and through the RHR heat exchangers (LHSI). In this way energy is rnmoved from containment atmosphere (primarily steam condensation) and 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 n, team generators. AFW is required for secondary inventory and heat removal until the break is able to remove all the decay heat. The break rcmoves energy from the RCS because the makeup water from the RWST is relatively cold and can absorb energy bef ore exiting the RCS. The WCAP 9600 analyses with , consideration of STP design features demonstrate d: cay heat removal capability for SBLOCA. SBLOCA Response The initiating event is.the break. If the break is 3/8" or less l in equivalent diameter and the charging system and feedwater system are l j ovcilable, 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 cafety injection signal. I 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 saf ety injection initiation, saf ety injection pump flow provides make-up to the RCS pressure and , manimum peal, clad temper atures will remain below 10CFR50.46 Appendin F. criteria. m , - -
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For breaks greater than 3/8" and less then 1 inch, SI flow can l Catch break flow so no significant RCS depressurization or core unccvery will occur. At the point where SI flow matches break flow the i aitigation phase of the accident ends and a long term stable condition is reached. i The long term stable condition is that the steam generators provide heat removal capability in conjunction with break flow. Heat rcmoval from the steam generators occurs through the safety valves. A eupply of auxiliary feedwater is available for the long term. Break sizes between 1 and 4 inches reach an equilibrium RCS prcesure above the shutoff pressure of the LHS1 pumps (283 psig). Stocm generator heat removal is necessary until the break can remove c11 decay heat. For a one inch cold leg break, the break can remove all decay heat approximately 24 hours after the accident is initiated (WCAP 9600, Sec 3.1). For a four inch co?d leg break, all decay heat i' ccn be removed through the break in less than 300 seconds into the trcnsient (WCAP-9600, Sec. 3.2). Energy is removed through the break to containment where containment fan coolers remove energy to component cooling water. If containment sump recirculation is established, the HHSI pumps are available for makeup. .The water used by the HHSI pump will be at sump temperature and the break alone may not be able to rcmove all decay heat depending on decay heat levels at sump I rccirculation switchover time. If the break cannot remove all decay hast, the steam generators would be relied upon until the break can rcmove all energy. I
For breaks greater than 4," the RCS will depressurized to below"~ tha LHSI shutoff pressure and all decay heat will be removed from the RCS through the break. The RHR heat exchangers will be sufficient to provide cooled makeup water to the RCS via LHS1 pump flow. For isolatable break cases, the steam generators would be relied upon to provide decay heat removal for an extended period following tGrmination of the break. By providing a supply of make-up to the AFW storage tank, decay hast can be removed from the STP core with qualified equipment only, following all sizes of LOCAs, including all LOCAs which could be cubsequently isolated by the operato.r. STP has committed to provide make v; from the essential cooling pond (SRP) for up 30 days utilizing hone reel connections from the EGP to the AFW storage tank. Beyond this time normal sources of AFW will be restored. The hose reel connections are to be utilized temporarily (through STPs initial 2 rofueling outages) until qualification of the residual heat removal cyntem (RHRS) can be completed. The qualification and procurement of RHR pumps program, is expected to require up to 3 years and is currently in process. The following discussion describes the SBLOCA ronponse when RHRS is available. For breaks greater than 4" the decay heat will be removed by the break to the containment fur coolers and the RHR heat exchangers via LHS1 pumps. No operator action is required.
6 , For breaks from 1" to 4", the operator will cool down and l l d: pressurize the RCS to a pressure below the cut-in pressure of the LHSI. This will be accomplished using the steam generator PORVs for cooldown and Pressurizer PORVs in combination with HHS1 flow tcrmination for depressurization. The detailed actions will be provided in the STP Emergency Procedures which are based on the WOG Emnrgency Response Guidelines, Revision 1. The combined heat sink cepetity of the Refueling Water Storage Tank and the steam generators would provide core cooling for approximately 27 hours, after which the containment fan coolers and the RHR heat exchangers via LHSI pumps will provide an adequate heat sink f or decay removal . ., For breaks between 3/8" to 1", the operator will cool down and d: pressurize to below the cut-in pressure of the LHSI pumps (283 psig). l Tha RHRS will be available to provide heat removal at RCS pressures i balow 400 psig and temperatures below 350 F. Adequate long term decay h;st removal will be provided by LHSI pump flow through an RHR heat cxchanger in addition to RHRS operation. 4 4 4 I y- - - - - - - - - - - . .- . , _ . - - _ _ - , - - , . _ . . - - - _ _ _ - -,
" ~
'g . STP FSAR I l Question 440.54N , State whether the STP ADO and PA analyses were performed for all operational - nodes. If not, or the assumption is made that Mode 1 bounds all the others, please review each AOO and FA to provide assurance that all ' equipment and systems relied upon for AOO or PA mitigation whose availability and oper-cbility is assured by the STP Technical Specifications in Modes 1 and 2 can ' also be relied on to provide mitigation in other modes. If this assurance can not be provided, then provide a detailed accounting of what systems, equip-cant, and protective functions were assumed for these modes, a justification of why the Modes 1 and 2 analyses are bounding, and a confirmation from the cpplicant that the technical specifications applicable in Modes 3, 4, and 5 vill be consistent with and provide the same level of intended protection as the technical specifications in Modes 1 and 2. If differences exist between the Modes 1 and 2 analyses and those for other modes, these should be dis-cussed in detail. - Rispense IA review of all STP anticipated operational occurrence (AOO) and postulated ' cecident (PA) analyses for all modes.and a discussion of the bounding analysis will be completed. A detailed confirmation that the Technical Specifications
. _'N (cpplicableinModes3,4,and5areconsistentwithModes1and2 vill be done. The results will*be available in the third quarter analyses of 1985. -- /
SEFA Mo{ Vcl. 2 QiR 15.0 12N A=endment 44
~
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4 4 4 1-t i i
Response
A Most Aoos and pas are analyzed for occurrences in in Modes 1 and 2, i since these are usually the modes in which the most severe consequences y could possibly result. Since the Technical Specifications are based upon these analyses, and are written to ensure the availability of required protection logic and. equipment in Modes 1 and 2, this respon=e will concentrate on transients which are postulated to occur while the plant is in any of the suberitical
- operational modes (jModes 3, 4, and
~
5 ) '. Generally, the occurrence of an Aob or PA when the plant is in a suberitical mode will not result in consequences more severe than those which would result in Modes 1 and 2. This is due mainly to the reduced temperature and pressure conditions characteristic of suberitical modes. In some cases, certain AOos or pas cannot occur, or cannot produce a significant transient (i.e., a transient which would challenge plant safety limits), and therefore, protection is not always required to the same degree as in Modes 1 and 2.
- W,f Each Aoo and PA has been reviewed with attention'to occurrences in modes which are not identified in the FSAR, and to the protection requirements and availability in these modes. The results and onclusions of this reviev.are given belev: _
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D MW 5
h INSERT A Each A00 and PA has been reviewed with attention to occurrences in modes which are not identified in the FSAR, and to the protection operability requirements in these modes. This review included contideration of the applicable Technical Specification to assure that the protection and operability required by the analysis was assured by the present Technical Specifications. The results and conclusions of this review are given below. f
- i 0962n:40/AN0/10-85
.,g- ; -
a . i Feedvater System Malfunction This A00 increases the core heat removal rate, which reduces the core temperature, leading to an increase in power generation (due to the < negative moderator temperature coefficient) and a consequential , reduction in thermal margin. The heat removal rate may be increased l oither by an increase in feedwater flow, or a decrease in feedwater temperature. Analyses or evaluations for both cases, in Modes 1 and 2, cre presented in *"'*" c.We.pk c- if-In Modes 1 and 2, protection is available fro = the power range high ! neutron flux trip. If the increased heat removal is due to abnormally
- high feedwater flow, then turbine trip and feedwater isolation will cecur when a high steam generator water level setpoint is reached.
The feedwater malfunction, associated with a drop in feedwater temperature, is not a concern below Mode 2, because there is no i pre-heating of feedvater in those modes.
- l. c.eW The feedwater;(malfunction which causes an increase in feedvater flow, postulated to result from the failing open of a feedvater control'
. valve, is ill defined below Mode 2, s'ince the main feedwater sys' tem would probabl be secured. Even if the main feedwater system were in-operation in ode 3, the flow would n o e m'Et 2 be controlled via
- i. the feedwater bypass valves, not the feedwater control valves, since l- the (smaller) bypass valves provide much better control under low flow l conditions. Therefore, failure of a main feedwater control valve in Mode 3 is not likely. The assumption of a failed-open feedvater bypass l:
valve, in Mode 3 and below, would result in a relatively slow transient I due to the lower feedwater flow rate. cp.d l' . . In suberitic 'z6das, the increased heat removal rate would cause an , increase in the audible count rate and possib y the astude(renge flux -tipt w en alarm to sound, alerting the operator to the increase in neutron flux.' If no operator action is taken, then any withdrawn rods - will be automatically inserted when the source range high. neutron flux trip setpoint is reached. In Modes 5 and 6, when the RCS is cold, any increase in heat removal rate would not be meaningful, nor would there be any viable feedback to the core, since the main heat removal path will be via the RER systa=. In general, the potential for. serious consequences.fresulting fro: cooldown events in the suberitical modes is low, since the RCS is relatively cool (usually less than the no-load temperature); and the core is shutdown. . O e
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- Excessive Load Increase The excessive load increase, in Mode 2, vill not be as~ opening severe of asaeither steam increase, or the the Mode 1 excessive load generator safety or relief valve (which is analyzed in Mode 2). This Aoo is an increase in steam flow (load), usually lo pere' int, which may er may not generate a reactor trip signal, depending upon the plant and k' protection system characteristics. Modein1 Mode analyses are presented in 1 is considered limiting, the c~y
- 16. -FSARn- An excessive load increase cince an excessive loadpower increase at full power would put the plant at level. Load increases at less than full the highest achievable power, or during startup (Mode 2), would not reach as high a power level before trip.
In Mode 3, the excessive load increase may be considered to be a simple oteam release, since there can be no load, per se, when the turbine is off-line and the core is suberitical. The Mode 3 load increase would be less limiting than the Mode 1 or Mode 2 case, since the core is already
'cuberitical. Automatic safety injection actuation may not be available, necessary.to if it is blocked 1ar the operator (blocking is depressurize). Movaver, the RCS must be borated to the cold shutdown concentration prior to blocking SI, in order to prevent a return to criticality in the event of a cooldown. .. .
The Mode 4 situation is bounded by Mode secondary 3, since pressure and and systems are ta=perature conditions in the primary reduced. Also, a cooldown in ModeAt4 vill so=e not be aggaravated by the point in Mode 4, the RER Cddition of auxiliary feedwater. cyste= will be placed in service, disconnecting the steam generators from the heat removal path. In Modes 5 and 6, the residual heat ra= oval systa= should be in operation. Any steam release, if possible, would have little or no effect upon the core. Spurious Opening of a steam Generator Safety or Relief Valve The Condition II steam line break, or the spurious opening of a steam generator safety or relief valve, also affects the core like a load increase; but the analysis assumptions that are applied are different. to be an The condition II steam line break is usually assumed unisolatable, uncontrolled steam release .which causes a non-uniform core cooldown (typical of an open safety valve) during the period im=ediately following a reactor trip which instry.s all but the most reactive RCCA. The resultingmargin reactivity excursion may and return be large enough the core to critical, to overcoce the shutdown especially when there is little or no decay heat (with power peaking in the region of the stuck RCCA) . The Condition 77. sten: line break is analyzed in Mode 2, and the assurptions used lead to a more severe transient than would result from a loa'd increase in Mode 1.
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- 4. '
MC f of this In Mode 1, prior to reactor trip, the transient characteristicip signal, Aoo are similar,to the excessive load increase.delta-T overpower A reactor logi After the if needed, w4l result from the roactor trip, the concern becomes a possible return to crit ality with the most reactive RCCA stuck in the fully withdrawn position, leading to high local power levels. However, a post-trip return to criticality ic less likely when this AOO occurs in Mode 1 than in Mode 2, because which tends to retard the there will be more decay heat present, cooldown. Mode i steamline break is discussed in WCAP-9226. In Mode 3, results are expected to be better than the Mode 2 case, temperature and flow conditions would be less limiting. cince pressure, An occurrence in Mode 4 would be even less severe than in Modes 2 or 3, due to the lower initial RCS temperature, and an effective decoupling, of the secondary system from the primary system as the reactor coolant pumps are removed from service and the residual heat removal system is ctarted. Automatic SI actuation is available through Mode 3, until the RCS is borated and the SI is blocked (see excessive load increase SI pump must be operable in discussion) . One high-head Mode 4, available for activation by the operator, if needed. Any cooldown in Modes 5 and 6 is meaningless, since the RCS is already cold, and the RER system effectively decouples the steam generators from the core. Steam Line Rupture The steam line rupture is a Condition IV avant, opening producing a greater uncontrolled steamorrelease than the spurious of a steam generator safety relief valve (above); but the relative effects in the various modes, and require =ents for protection equirr2nt are the . came. This is the most severe cooldown event. , Stea= T1ov Reduction . In the case of South Texas, there are no steam pressure regulators whose malfunction or failure could cause a steam flow transient. He j cafety analyses or protection, in any mode, are required. Loss of Electrical Load This Aoo can occur only in Mode 1, since the. turbine is off-line in all i other modes. Loss of electrical load is bounded by the turbine trip (below), which is analyzed and reported in th; TS?2dfor Mode 1. C.%p%r s T l I
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1 I 1 i i l Turbine Trip .. I This Aoo lls defined only in Mode 1, since the turbine would be off-line l below Mode 1, and bounds the loss of electrical load (above), since oteam flow is terminated more rapidly by a turbine trip than by a loss of load. 1 Spurious MSIV Closure i l The Mode 1 case is limiting, which itself is bounded by the turbine trip Aoo (above) . In Modes 2, 3, and 4, the plant any be cooling down ! via steam dumping to the condenser. MSIV closure in these modes may j prevent the use of the condenser, and require atmospheric steam i dumping. There is no steam flow below Mode 4, since cooldown is l continued via the residual heat re= oval system. Only the availability 1 of a means to dump steam to the atmosphere (including PORVs and safety I valves) is required, for decay heat removal, in the event that the l MSIVs close while the plant is in any mode above Mode 5. . 4 I Loss of Condense.r Vacuum . The full power case is bounded by the turbine trip Aoo. Loss of the [ condenser vacuum, while the plant is below Mode 1, may require decay ! heat removal via atmospheric steam dumping, until some time in Mode 4, l when the residual heat removal system is placed in operation. Loss of AC Power The Icss of AC power results in the loss of primary coolant flow and , cain feedwater flow. It must be shown that decay heat can be removed, via natural circulation in the reactor coolant system, to the steam generators, which are supplied with auxiliary f3edwater. Therefore, the full power case (maximum decay heat) is limiting. At least one i ouxiliary feedwater pump is required (and is available), in Modes 1 through 3, for decay heat removal. , In Mode 4, the transition is made from steam du= ping to the residual heat removal system for further cooldown. Although the auxiliary feedwater pumps are not required to be available in this mode, it is reasonable to assume that, during cooldown operations, the reactor i l operator would continue to feed the steam generators with auxiliary
- feedwater, well into Mode 4, until the RCS pressure decreases to a level low enough to activate the Residual Heat Re= oval System.
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l Loss of Teedwater / This A00 results in a heatup and pressurization of the RCS. Therefore, cn occurrence at full power would result in the most severe consequences. , In Modes 2 and 3, the auxiliary feedvater pumps are available yend4-s_____, . for startup and decay heat removal. Not all the auxiliary feedwater pu=ps are required for decay heat removal, and loss of all auxiliary feedwater pu=ps is not likely. If none of the auxiliary feedwater pumps are operable, then the diecbM require the operator to restore at least one pump as soon as,possible g g At some time in Mode 4, the Recidual Meat Removal System will be pla'ced in service, and cooldown via the steam ' generators will not be necessary. Prior to the operation of the RHR Systsem,.the operator is assumed to be using the auxiliary feedwater pumps, even though they are not required to be availa Mo see y of AC Power). Feedvater Line Rupture This PA may occur any time the steam generator is pressurized. An RCS heatup and occurrence in Mode 1 would cause the grcatest pressurization. Therefore, the Mode 1 case is analyzed, and bounds events in Modes 2, 3, and 4. - Auxiliary feedwater is required through Mode 3 for decay heat removal. In Mode 4, the lov levels of decay heat and pri=ary and secondary side temperature and pressure will result in a relatively minor, slow transient. Below Mode 4, the question becomes moot, since the steam generators are no longer required for decay heat removal. Partial Loss of Flo.w The loss of a reactor coolant pu=p reduces the heat removal rate fron the primary to the secondary coolant system, thereby causing a heatup in the RCS. An occurrence at full power would produce a greater heatup than would an occurrence at no-load (Mode 2). Below Mode 2, when the core is suberitical, it is com=on to have one or more reactor coolant pumps out of service, since full flow is 2, no longer required. Loss of a even if it is the only purp in reactor coolant pu=p - below Mode service, would still be bounded by either the partial loss of flow in Mode 1, or the complete loss of flow in Mode 1 (below).
- - - - - - - - ---- - - - - - - ~ - - -- , < -
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l Loss of Flow A3 in the partial loss of flow, the most severe case is an occurrence in Mode 1. However, the loss of all reactor coolant pumps means that the only mechanism available for decay heat removal from the core is via natural circulation. Therefore, ired through adequate natural circulation and Mode 3. Auxiliary feedvater, cuxiliary feedwater are re 01though not required by tec spegg, is assumed to be available until the residual heat removal system can be placed in service (Mode 4). This event, and its protection equirements are similar to the Loss of AC Power event (discussed previous y . ,4pf . Locked Rotor and Reactor coolant Pump Shaft Break These pas are similar to the partial loss of flow (above),as far as the limiting modes and required protection equipment ara concerned. RCCA Withdrawal from Suberitical C.h@r- C presents an analysis for this A00 in Mode 2. An occurrence in 75AW* Th Mode 3, 4, or 5, with two or more reactor coolant pumps in operation would be bounded by the analysis in Mode 2. This is based upon the-FGAkE the Cnalysis assumption that reactor trip does not occur until power-range (low se.tting) high neutron flux setpoint at maximum is reached, and speed (72 that two banks are withdrawn sequentially otep/ min) . These conservativesome assu=ptions result in the core returning power prior to trip. Therefore, the to critical and generating important consideration, as a primary system flow rate becomes an
- factor in DNB evaluation. (Note that, in Mode 3, the the reactor W require trip two reactor coolant pumps to be in operation whenever, breakers are closed). ug gt-Mowaver, in Modes 3, 4, and 5, the source range high neutron flux trip will be available to terminate the event, by , tripping any withdrawn and withdrawing rods, before any*significant power level cou14pe attained.
Therefore, DNS and primary system flow rate needrnot be considered. Also, the reactivity insertion rate would be slowerg when in any of the Cuberitical modes, withdrawal since a single failure in the red control system could cause the of only one bank, and its withdrawal rate would be expected to be slower than the maxiet: rod speed which is possible when in automatic rod control (and is assumed in theHFEAR'k. CA"/ ' I 3-analysis).
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-_______-__;- ---------7 RCCA Withdrawal at Power This Aoo is defined only in Mode 1.
Dropped RCCA Bank Since the dropping of an RCCA bank will perturb the core only if there io some significant neutron flux level, this event is analyzed only in Mcde 1. A less severe case can be postulated at the low power level of Mcde 2. Dropping an RCCA bank while in any of the suberitical modes,-if cny are withdrawn, would have no effect (i.e., no DNB concern). Dropped RCCAs - - 50e dropped RCCA bank (above).
- Single Rod Withdrawal The lin'iting case is an occurrence while in Mode 1. An occurrence in any of the suberitical modes would have no effect. If the shutdown targin requirements are satisfied, then no single rod withdrawal would insert enough reactivity to attain criticality, since the shutdown cargin requirements are determined assuming the most reactive RCCA is fully withdrawn. .
Static Rod Misalignment A3 in the dropped RCCAs and dropped RCCA bank, this event would have no offect in the absence of a critical neutron flux. The limiting case, and analysis, is for Mode 1, which bounds Mode 2. There.is no DNB i concern in any of the suberitical modes. StartupofahInactiveboop Tor plants without loop isolation valves (e.g. South Texas), the consequences of this event are directly related to the te=perature difference between the inactive loop vessel inlet and the core. Rolatively cold water Vould enter the coreF g after the reactor coolant pump in the inactive loop is started up, ar.d cause a reactivity oxcursion. Therefore, the most severe consequenc'es are incurred when the plant is operating at the maximum per=issable power level with a loop out of service. This is the Mode 1 case which appears in the FSAR. The Mode 2 case, when starting up, is bounded. Startup of an inactive
g ,,,;.. Icep whilo in cny ofcora tnatGmporaturr,, cumcritical aceco woule havo rolotivoly little DincQ th3r0 would b3 littlO Cr no offect upon the tcuperature difference between active and inactive loops. Below Mode 4, the RHR system would be in operation. Boron Dilution This Aoo is analyzed or evaluated in every mode. f l Fuel Assembly Misloading This event, like the rod misalignment events, is meaningful only in the , presence of a critical neutron flux. Mode 1 behavior is presented in
- tu ?S W which bounds the Mode 2 startup case.
Cakyk e t s~ RCCA Ejection
- is C.k& d.
Mode 1 and 2 cases are analyzed der the 70AT. If shutdown margin requirements are met, then the ejection of a rod while in Modes 3 and 4 would not insert sufficient reactivity to attain criticality, since the chutdown margin requirem,ents are determined assuming that the most reactive RCCA is fully withdrawn. In Mode 5, ejection is impossible, cince the RCS is depressurized. , Accidental ECCS Actuation In Mode 1, this event is analyzed for its effect upon the core. A opurious$5I signal should cause an immediate reactor trip. Delivery of cafety injection fluid to the core would also cause shutdown. However, if the SI signal dor.s not generate a reactor trip signal, pumps then do not there have would be no effect, since ?.he south TexasHs1 cufficient head (rated at cr.ly Soo psi) to inject into the RCS at normal operating pressure. At pressures below the normal operating pressure, in Mode 3, the SI Oystem has the potential to pressurize the RCS to the shutoff head of the)MSI pumps (1,{0c psia). When RCSI,, temperature then only drops to the level one high-head necessary to arm COMS 81 pumpis(Mode 4) permitted to be available, and the RCS may become pressurized to the pressurizer relief valve setpoint, which is 9 set to provide cold overpressure protection. In Mode 4 spurious SI actuation is not likely, since most automatic sI signals , 7 are blocked. , CVCS Malfunction The boron dilution aspects of thisaddition event are covered in the boron transient, this event is dilution AOO (above). As n' zass addressed by the cold overpressure tech specs (Modes 4 and 5).
l'- / Spuricu3 Cp ning of O Proscurizar Roliof cr Sofoty Volvo cencarn becomes o When analyzed oc o doproesurizaticn ov0nt, the of the minimum DNBR criterion. Therefore, this AOO possible violation 10 analyzed in Mode 1, which also bounds Mode 2. DNB is not a realistic modes. In Mode 5, this A00 is concern in any of the suberitical inconsequential, since the RCS is already depressurized. *- The loss of RCS inventory aspects of this AOO are considered as part of the Small Break loss of coolant Accident (below). i
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Small Break Loss of Coolant Accident' f , LATER Steam Generator Tube Rupture deferred, pending the findings of 'the The
- tube rupture question Group is study in progress. Results are expected Westinghouse Owners cround the and of this year.
Large Break Loss.of Coolant ,. , LATER
-" Conclusion x )
Protection pas is available,inallsuberiticalmodes,forallapplicable{ Aoos and (except where notedll, to a level which is judged to be g consistent with Modes 1 and 2, considering the potential. consequences, or PA, transient-initiating' failures are of each Aoo components and systems, in which removed from service, as modes and the RCS postulated to occur, are are reduced. This permits the disarming of temperature and pressure It is our protection systems, when they are no longer required. be available, protection vould engineering judgment that adequrste or by operator action, to a level that is either by automatic means the protection available in Modes 1function and 2, generally consistentconsequential with reduction in protective considering the r requirements, as modes are reduced below Mode 2. I A - W
INSERT B Therefore it is con' eluded that appropriate protection for all applicable A00s and pas (except where noted), is available and assured by Technical Specifications for all operable modes. The reduction in the operability requirements of Technical Specifications below Modes 1 and 2 are consistent with the reduction in the severity and potential consequences of each A00 and PA in these lower modes. Many components and systems, in which transient-initiating failures are postulated to occur when in Modes 1 and 2, are removed f rom service as the RCS temperature and pressure is reduced below Mode 2. This permits the disarming of protection systems when they are no longer required. Nevertheless, it is our conclusion that adequate protection is assured, and would be available by automatic actuation or operator action at a level that is consistent with the protection available in Modes 1 and 2, considering the reduction in the protection requirements below Mode 2. 1 . l r i ~ I I l l l . 1 1 OM?n:41/AN D/_10-8 5. . _ _ _ _ _ ___ _ _ _ _ _ _ _ _ _ _ . __ _ . _ . _ . _ _ . _ _ _ _ . _ _ _____.____
_ _ _ _ _ . _ . _ _._ . _ _ _ _ . . _.m _ _ - . - _ _ . __ BoR.oAJ
. STP FSAR -
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l
. Question 440.67N .
4
. Provide the following informat*.1on with regard to the "CVCS Malfunction that l Results in a Decrease in Boron Concentration in the Reactor Coolant" analysis-l
- a. For each operational mode, list the alarms and indications that would i alert the operators to the occurrence of a BDE, and verify their redun- '
dancy. Also describe any automatic mitigation systems. Confirm,that your technical specification will require two alarms to be operable during all shutdown and refueling modes.
- b. The FSAR states that the maximum dilution flow during startup and hot standby is 382 gpa based on operation of two reactor makeup water (RMW) pumps while the RCS is at 2250 psi. For this dilution flow rate, the minimum time for loss of shutdown margin is 19.6 minutes,
- t
- 1. Please confirm that you will impose technical specification limits
*to ensure that RCS pressure, when accounting for instrument error, will not be dropped below 2250 psi in either of these two modes.
- 2. Please provide analyses of boron dilution events in modes 4, 5, and
- 6. How do you intent to ensure RCS pressure never drops below the pressure corresponding to the maximum dilution flow assumed in your analysis? Our concern is that the SRP Section 15.4.6 criterion of i 15 minutes (30 minutes for Mode 6) for minimum time availability before shutdown margin is lost will be met with maximum dilution flows assuming operation of two charging pumps and two RMW pumps at
] minimum RCS pressure for the particular mode analyzed. , I c. The FSAR states that valve CV0298 in the CVCS will be locked closed during refueling. Discuss whether' additional valves should also be 7 locked closed for redundancy, f-Demonstrate that all possible dilution flow paths have locked closed valves, and confirm that the tech specs will contain this information. a 4 . r i i Response f*
- i f
Qe MS4h k l h, W b-1 1 ] Vol. 3 Q&R 15.4-2N Amendment 49 - I
; .. l' INSERT A The following information describes the alarms and indicators available to alert the operators to the occurrence of a boron dilution event.
There are several different alarms / indicators which would alert an gperator of a boron dilution event at STP, as follows:
- 1) High Flux Level Alarm - During startup and shutdown conditions, the high flux level at shutdown alarm is visually and audibly annunciated in the control room when the setpoint of 1/2 decade above the background level is reached. The audible alarm is also given inside the containment as the containment evacuation alarm.
- 2) Source Range Audible Counter .An isolated output from the pulse amplifier for the source range instrumentation provides an audible tone proportional to the selected source range channel count rate.
- 3) Flow Differential Alarm - As described in FSAR Section 9.3.4.1.3, in the event of a boric acid makeup fault during automatic makeup, boric acid and demineralized water flow indication and flow deviation alarms are provided to alert the operator.
- 4) Overtemperature AT - Rod Inser Limit Alarms - During full-power operation, with the reactor in automatic control, the power and temperature increase from boron dilution results in insertion of the RCCAs and a decrease in the shutdown margin. The rod insertion limit alarms (low and low-low settings) provide the operator with adequate time (on the order of 26.9 minutes) to determine the cause of dilution, isolate the reactor grade water source, and initiate reboration before the total shutdown margin is lost due to dilution.
$ith the reactor in manual control and if no operator action is taken, i the power and temperature rise will cause the reactor to reach'the overtemperature AT trip setpoint. The boron dilution accident,in this l
6392N:0258N
o case is essentially identical to a RCCA withdrawal accident. The maximum reactivity insertion rate for boron dilution is approximately 1.72 pcm/sec and is seen to be within the range of insertion rates analyzed. Prior to the overtemperature tit trip, an overtemperature tit alarm and turbine runback would be actuated. There is adequate time available (on the order of 17.1 minutes) after a reactor trip fo,r the operator to determine the cause of dilution, isolate the reactor grade water sources, and initiate reboration before the reactor can return to criticality.
- 5) Neutron Flux Shutdown Monitor Alarm - A qualified safety grade neutron flux shutdown monitor is provided to measure the countrate from the qualified class lE extended range neutron flux monitors (refer to Table
- 7. 5 - 1 ) . The shutdown monito'r prevides an alarm when the countrate increases by an amount equal to the alarm ratio that has been set into wn; tor.
the shutdown 4 Redundant monitors, each with its own alarm are provided. With the above instrumentation, the reactor operator can be alerted to any reduction in shutdown margin. 1 6392N:0258N
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YSV$T h SWSWT '
- BORON DILUTION ANALYSIS
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i 1.0 CVERVIEW I N-"---
~ ~ " - alysis %g to demonstrate that the South Texas Project .
- m. o g requireme s M Oe. ;t: f;rd .e de- Mer. Geetur. M.O.5) regarding boron dilution events If any administrative changes are necessary As to part of maintain regulatory comp 11anc,e, these changes will be identified.
this effort. a probabilistic analysis willPlant be performed to evaluate the response to each credible probability of boron dilution events. initiator will be modeled The to obtain the probab probabilistic dilution event resulting in inadvertent criticality. analysis will identify whers'the South Texas Project is susceptible to boron dilstion events, and thus allow Houston Lighting & Power Company to insure that administrative requirements are sufficient to reduce the probability of boron dilution events to an acceptable level. , A j.. p l li!
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*l ,. s . STP FSAR ,, Question 440.68N Describe or reference th's anal tical model used in the BDE calculations. '
Discuss the degree of conservatis:n of this model, including that of scram times, moderator and Doppler c.oefficients, and mixing of coolant.
Response
L y N W W '> W 6 =r W w As described in the response to. Question 440.67, a probabilistic analysis is being performed to evaluate the probability of boron dilution events. The' analysis will begin with a detailed Failure ' Mode and Effects Analysis (FMEA) to identify potential boron dilktion initiators. The FMEA will provide a detailed evaluation of the CVCS system to identify potential equipment l faults or operator errors which could result in hn advertent dilution of reacto'r coolant system boron .concentrathn. Shu'tdown modes 3 (hot standby), 4 (hot shutdown), 5 (cold shutdown), and 6 (refuelin g wi 1 analyzed.
< The' frequency of eact .redible boron dilution event wS11 be calculated using
_ industry accepted equipment failure and human er.ror probabilities. Maximum dilution flow rates for each initiator will beat:t:d 8- =i+5er Ye=+i 6vuus esii;t, r nr =ue+*a -";' :=.*' Minimum shutdown boron concentration and ' shutdown margin, for each shutdown mode, will-be obtained from the #- - 9
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utility. technical specifications Additionally, minjmum reactoror agreed upo'n'by coolanti system West,inghouse volume will be' and C . P jdentified for each shutdown mode. Using this information, the time to alarm annunciation of a boron dilution event and timeito criticality will be calculated to show compliance with etaa:e.- -^v 5,44) requirements. 3 Additionally, a probabilistic analysis of boron.dilu on events will be performed. Event tree modeling will be employed to calculate the frequency of boron dilution events which result in unplanned criticality. This analysis will include an evaluation of alarm reliabilities, and a probabilistic eval.uation of the o.perator response to the boron dilution event. g s 3 i/K $ 5.% k
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NOV 18 E85 MEETING
SUMMARY
DISTRIBUTION
.s . Docke t, flo( s) ;; 150-498/499 -
NRC PDR Local PDR NSIC PRC Systen LB3 Reading Attorney, OELD GWXnighton Project Manager N. P. Kadambi JLee NRC PARTICIPANTS H. Li B. Mann R. Goel J. Wilson N. P. Kadambi G. W. Knighton T. Marsh B. Sheron bec: Applicant & Service List}}