ST-HL-AE-1495, Forwards Response to Listed Draft SER Open Items Re Finalized Cold Overpressure Mitigation Sys analysis.Marked- Up FSAR Pages Will Be Incorporated Into Future FSAR Amend

Forwards Response to Listed Draft SER Open Items Re Finalized Cold Overpressure Mitigation Sys analysis.Marked- Up FSAR Pages Will Be Incorporated Into Future FSAR Amend

ML20137H769
Person / Time
Site:	South Texas
Issue date:	11/20/1985
From:	Wisenburg M HOUSTON LIGHTING & POWER CO.
To:	Knighton G Office of Nuclear Reactor Regulation
References
CON-#485-278 OL, ST-HL-AE-1495, NUDOCS 8512020469
Download: ML20137H769 (8)

Text

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The Light NE ilomion 1.ighting & l'ower l'O. Ilos 1700 llumion. In.n 77001 (7 th 22M 'll November 20, 1985 ST HL- AE 1495 File No.: C9.17 Mr. George W. Knighton, Chief Licensing Branch No. 3 Division of Licensing U. S. Nuclear Regulatory Comnission Washington, DC 20555 South Texas Project Units 1 and 2 Docket Nos. STN S0-498, STN $0 499 Et.1ITH111J a_DSEE/.ISAILit e m s.L C 04S

Dear Mr. Knighton:

The attachment enclosed provides STP's response to Draft Safety Evaluation Report (DSER) or Final Safety Analysis Report (IS All) items.

The item number listed below correrpond to those assigned on STP's internal list of items for completion wht.h includes open and confirmatory DSER items, STP FSAR open items and open hRC questions. This list was given to your Mr. N. Prasad Kadambi on October 8, 1985 by our Mr. H. E.

Powell.

• The attachment includes mark ups of FSAR pages which will be incorporated in a future FSAR amendment unless otherwise s.oted below.

The items which are attached to this letter are:

Attachment Item No.* Subject 1 D S.2 8 CVCS Fump Operation Design Basis D S.2 9 Cold Overpressure Hitigation System (CONS)

D 5.2 10 IIHS! Pumps and Accumulator Isolation Valve Lockout Note 1: Inconsistencies may exist between the attached revision to FSAR Section S.2.2.11 and the response to Question 440.23N.

The response to Question 440.23N will be updated at a later time to agree with the information provided herein.

• Legend D DSER 0l'>en Item C DSF.R Confirmatory Item F FSAR Open Item Q FSAR Question llesponse Item M//

L1/DSER/a7 (10120D0469 H$ lipo DN ADOCK 0000 0 #[

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t Houuon Ughting & Fuwer Company ST-ML AE 1495 File No.: C9.17 Fase 2 Note 2: Information provided in this letter is based on a completed .

CONS analysis. This supercedes statements in ST ML AE 1445, dated October 31, 1985 which indicates that the COMS analysis has not been finalised.

If you should have any questions concerning this matter, please contact Mr. Powell at (713) 993-1328.

Very truly yours, t

y. ,

h( .s ,t.u ..iA e A N. R. W1 nburg Manager,NuclearLickning JSF/b1 Attachments: See above I

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ST ML AE 1495 File No.: 09.17 Fase 3 cc Nugh L. Thompson, Jr., Director Brian E. Berwick, Esquire Division of Licensing Assistant Attorney General for Office of helear Reactor Regulation the State of Temas l U.S. Nuclear Regulatory Commission F.O. Dos 12548 Capitol Station . i Washington, DC 20555 Austin, TX 78711 Robert D. Martin !anny A. Sinkin l Regional Administrator, Region IV 3022 Porter Street, N.W. e304 Nuclear Regulatory Commission Washington, DC 20008 411 Ryan Flasa Drive, suite 1000 Arlington, TX 74011 Oreste R. Firfo. Esquire Nearing Attorney N. Frasad Radambi, Project Manager Office of the Esecutive Legal Director U.S. Nuclear Regulatory Commission U.S. helear Regulatory Commission 7920 Norfolk Avenue Washington, DC 20555 Bethesda, ND 20814 j Charles techhoefer, Esquire  !

Claude E. Johnson Chairman, Atomic Safety & i senior Resident Inspector /STP Licensing Board '

c/o U.S. Nuclear Regulatory U.S. helear Regulatory Commission Commission Washington, DC 20555 i F.O. Som 910 Bay City, TX 77414 Dr. James C.14mb, III l 313 Woodhaven Road N.D. Schwers, Jr., Inqui.e Chapel Hill, NC 27)14, taker & Botts

• 3 One Shell Flasa Judge Frederick J. Shon Houston, TX 77002 Atomic Safety and Licensing Board U.S. helear Regulatory Commission J.R. Newman, Esquire Washington, DC 20b5 newman & Nottsinger, P.C. '

1615 L Street, N.V. Mr. Ray Goldsteinl Esquire Washington, DC 20036 1001 Vaughn Building  ;

807 Brasos Director, Office of Inspection Austin, TX 78701

and Enforcement U.S. Nuclear Regulatory Commission Citisens for Equitable Utilities, Inc.

Washington, DC 20555 c/o Ma. Peggy suchorn Route 1 Bos 1684 E.R. Brooks /R.L. Range Brasoria, TX 77422 Central Power & Light Company F.0. Boa 2121 Docketing & service section Corpus Christi, TX 78403 Office of the Secretary U.S. helaar Regulatory Commission j N.L. Peterson/0. Fokorny Washington, DC 20S$5 City of Austin (3 Copies)

F.O. Boa 1088 Austin, TX 78747 Advisory Committee on Reactor Safeguards U.S. helear Regulatory Consission J.B. Poston/A. ventosenberg 1717 N Street City Public service Board Washington, DC 205S5 F.0, Bos 1771 i

San Antonio, TX 78296 Revised 9/2S/85 L1/DesR/a7

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• STF FSAR ATTACHMEN .Z ST HL AE /

_PAGE / 0 s 5.2.2.9 tretes Relfabtlity. The reliability of the pressure-relieving devices is discussed in Section 4 of Reference 5.2-1. ,

, 3.2.2.10 Testina and inspection. Testing and inspection of the ever-i pressure protection composeate are dioeveeed in Sebeection 5.4.13.4 dad

! Chapter 14 S.2.2.11 BCs Preseure control Durina t,ow Teeperature operation. l Adelaistrative procedures are structured to and the operator in controlling j

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taatter Coolant systes Pressure during low toeperature operation. Bowever to provide a backup to the operator, as autoestic eystem is provided to

g 1stato pressures within allowable limits.

S.2.2.11.1 5petes operation: Bach of the two pressuriser power-operated relief valves is supplied with actuation logic to ensure that a

, completely automatic sed ladependent RCS preneure control back-wp feature is provided for the operator during law temperature operatione. This orates provides the capability for additional RCS taventory letdown, thereby maintaiste* RCS preer.ure withis alloweble limite. Refer to ,

Sectione 3.4.7. 3.4.10. S.4.13. 7.6Jdad 9.3.4 for additional information en 0211 l RC5 preneure and tavestery sentrol during other modes of operation. i

! 02 l

wudalli ely monitor RCS The toeperature basicand function pressureofconditions, the systen logic with the logic is4to contia[ armed whenever plan l 1

operation is at a temperature below 350'F. As auctioneered systee #

temperature will be continuously converte(to en allowable pressure and i

then compared to the actuel gCS pressure.)c i: : ,- 9 w111 provide an ,

actuation elsnel to the power-operates relief valves when required', to Q' prevent pressure-temperature condittese free esc'eeding the allowable g g imite.

I I. See Section 7.6 for a further discueston of oysten logic.

I S.2.2.11.2 _tvelustion of 1aw Teoperature overpressure Transientok *pMt=

$Df[5ection!!!.AppendiaG.establishesguidelinesandlimitsforRCSpressure primarily for low toeparature conditions (as350*F). .

Transient analyses were performed to determine the assimum pressure for the

,..tulated <cremle) worst case -.. urut a.d heat u rut event..

31

%W as of a Ico iniw pref....

pe

.n. Jus vntah wou d occgr ao (ti would we lete freIueetyduringthe isolati with c sing h*K p e de erlag\p1 aput than 1 to gel /ste owever he '

p[ ases p

ut ena gwee rio d suming own tool n with a che g

, operat s la a - afigura .produ( ysisus ieryr s.\ e -

% W 4 rete = f1).stkil.itJ.Jer gressure conte E more well y and ,

reNeverecost attonweegdeento ov de ad giona1 i A ,, , _

7 l The heat input snelysis was performed for sa inadvertent reacter seelast puer start seeweing that the RCS was water solid at the initiation of the event and that a 50*F alematch esisted between the BCS and the secondary side of the

} steam generators. (At lower toeperatures, the asse input esse is the lietting transient condition.)

3.2-4 Amendment 31

ATTACHWENT 1 STRAE /WC

. Insert A PAGE.2 0F d- _

Analyses have shown that one pressuriser 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 tempersture overpressure event is provided through the use of two pressuriser power operated relief valves to mitigate potential pressure transients.

The automatic systes is required only during low temperature water solid

• operation when it is manually armed and automatically actuated.

As described in Section 5.4.13, the STP PORVa are safety related and Class 1E powered. They are designed in accordance with the ASNE Code and are qualified via the Vestinghouse pump and valve operability program which is described in Section 3.9.3.2.1 and are seismically qualified as described in section 3.10.N and environmentally qualified as described in section 3.11.N.

Offsite power is not required for the system to function. The actuation logic in the system is testable on line. The PORVs are not exercised with the reactor at power, however, they are capable of being tested as required by the ASMZ Code and the STP Technical Specifications. Chapter 7.6 provides details of the design of the FORV interlocks for low temperature operation.

i Insert 3 The systes logic will first annunciate a main control board stars whenever the measured RCS pressure approaches the allowable pressure. Upon further increase in the RCS pressure, the systes 1 . .

Insert C The mass input transient is divided into two parts for plant operation in Mode 4 (above 200'F) and Mode 5 (less than or equal to 200*F). 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 S, the mass input transient assumes the operation of one centrifugal charging pump with letdown isolated L1/08F.R/a 7

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ATT ACHMENT .2.

ST.HL.AE. Ht97 PAGE~5 0F dr

. Both h:st input c d mass 1:put cnalyses ts:k into cecou_t the 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 jjgg will not constitute an impairment to vessel integrity and plant safety.

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l 2.3 Reactor Coolant pressure Boundary Materials .

5.2.3.1 Material specifiestions. Material specifications used for

• the principal pressure-retaining applications in each component constitut-1rg the RCFB are listed in Table 5.2-2 for ASME Class 1 primary components and Table 5.2-3 for A5ME Class 1 and 2 auxiliary components. These tables also include the unstab111:ed austenitic stainless steel material.specifica-tions used for components in systems required for reactor shutdown and for energency core cooling.

The unstabillied 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 utill ed conform with the material specification requirements and include the special requirements of the ASME Code.

Section 111. plus addenda and code cases as are applicable and appropriate to meet Appendix B of 10CTR30. The listed specifications in Table 5.2-3 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 Naterial Specifications STA 5.1 5.2. 5.17. 5.18. and 5.20. They are tested and qualified to th( require-ments of ASME Code. Section !!!. In addition, the ferritic materials of the reactor vessel belt line are restricted to the following maximum limits of copper, phosphorous, and vanadium to reduce sensitivity to irradiation embrittlement in service. -

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5.2-4a Amendment 31

ATTACHMEN GL ST HL AE

, Inenrt D PAGE # 0 5 5.2.2.11.3 Administrative Procedures Although the system described in Section 5.2.2.11.1 is provided to maintain RCS pressure within allowable limits, administrative procedures minimize the potential for the consequences of any transient that could actuate 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 specify that the use of a pressurizer steam bubble during periods of low pressure, low temperature operatior. is preferred.

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 operating procedures:

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a. Whenever the plant is water solid and the reactor coolant pressure is being maintained by the low pressure letdown control valve, letdown ficw normally bypasses the normal letdown orifices.

b. If all reactor coolant pumps have stopped for more than 5 minutes during plant heat up after filling and venting has been completed 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 accommodate the resultant expansion as the cold water is rapidly warmed.

c. If the 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.

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ATTACHIEN .2 ST HL Al / W Insnrt D (Cont'd) -PAGE ar O ef- ~

These special precautions back-up the normal operational mode of maximizing periods of steam bubble operation so that cold overpressure transient 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 prevent developing cold overpressurization transients. During these limited periods of plant

• operation, the following precautions / measures are considered in developing the procedures:

a. To preclude inadvertent emergency core cooling system actuation during heatup and cooldown, procedures require blocking the pressurizer pressure, and excessive cooldown protection signal actuation logic below the P-ll setpoint.

b. During further cooldown, closure and power lockout of the accumulator isolation valves will be performed at 1,000 psig. When the RCS temperature is reduced to or below 350*F, a maximum of one centrifugal charging pump and one HHSI pump is allowed operable' by Technical Specifications. The LHSI pump does not impact the COMS analysis because of the low differential shutoff head (approximately 315 psi).

c.

The recommended procedure for periodic emergency core cooling system pump performance testing vill be to test the pumps during normal power operation 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 vill be done when the vessel is open to atmosphere, again precluding overpressurization potential.

If cold shutdown testing with the reactor vessel closed is necessary, the procedures require emergency core cooling system pump discharge valves to be closed. .

d. SIS circuitry testing, if done during cold shutdown, requires the nonoperating safety injection pumps that are not to be tested have power locked out to preclude developing cold overpressurization transients.

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