ST-HL-AE-2263, Forwards Annotated Revs to FSAR Sections 6.2.1.4 & 10.4.7, Tables 6.2.4-2 & 15.1-1 & Figure 6.2.4-1 (Sheets 6-9) Re Closure Time for Main Feedwater Isolation Valves.Changes Result from Startup Testing.Values Will Be Incorporated

Forwards Annotated Revs to FSAR Sections 6.2.1.4 & 10.4.7, Tables 6.2.4-2 & 15.1-1 & Figure 6.2.4-1 (Sheets 6-9) Re Closure Time for Main Feedwater Isolation Valves.Changes Result from Startup Testing.Values Will Be Incorporated

ML20215G889
Person / Time
Site:	South Texas
Issue date:	06/17/1987
From:	Wisenburg M HOUSTON LIGHTING & POWER CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
ST-HL-AE-2263, NUDOCS 8706230426
Download: ML20215G889 (14)

Text

... ..

The Light Company u- uew- e.um umma.m mm msmo June 17, 1987 ST-HL-AE-2263 File No.: G9.1 10CFR50 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Annotated FSAR Revisions Regarding Closure Time for the Main Feedwater Isolation Valves Enclosed are annotated revisions to FSAR Sections 6.2.1.4 and 10.4.7, Tables 6.2.4-2 and 15.1-1, and Figure 6.2.4-1 (Sheets 6 through 9) regarding the closure time for Main Feedwater Isolation Valves (MFIV). These changes are the result of startup testing performed at STP and are being incorporated to envelope the tested vaines.

The impact of this change in valve closure time (from 5 seconds to 10 seconds) on affected accident analyses in FSAR Chapters 6 and 15 was evaluated.

The safety evaluations were completed on an analysis-by-analysis basis and addressed the increase in MFIV closure time. For each of the evaluations the appropriate approach to assess the impact (e.g. calculations, use of existing sensitivity studies, or limited / selective reanalysis) was determined and utilized. For each of the applicable safety analyses in the FSAR, the safety evaluations confirmed the acceptability of operating STP Units 1 & 2 with an MFIV closure time of 10 seconds.

The evaluations for the LOCA related transients in FSAR Chapters 6 and 15 indicated that there is no adverse effect on the FSAR results. Each of the affected non-LOCA transients has been evaluated with respect to the DNB acceptance criterion or to the quantity of secondary mass and energy released, depending on the specific transient. The most limiting of the affected events are the steamline break transients. Reanalysis of these events with the increased MFIV closure time indicates that the DNB acceptance criterion is satisfied. The conclusions in the FSAR for the other affected transients with respect to DNBR remain valid and there is no increase in the quantity of mass and energy released following a postulated main steamline break with this increased closure time.

The documentation associated with the safety evaluations is maintained in the Westinghouse offices and is available for audit.

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E7062'O426 070617 0 PDR A

ADOCK 05000498 pop  !

( i)

Ifouston Lighting & Power Company June 17, 1987 ST-HL-AR-2263 File No.: G9.1 Page 2 1

These changes will be incorporated in.a future FSAR amendment. If you  !

should have any questions on this matter, please contact Mr.'J..S. Phelps at

(713).993-1367. .

1 M. R Wisenbur Mana er, Enginee g and Licensing WW/JSP/ma

Attachment:

Revised FSAR Sections 6.2.1.4 and 10.4.7; Tables 6.2.4-2 and 15.1-1; Figure 6.2.4-1 (Sheets 6 through 9)

ST-HL-AE-2263 File No.: G9.1 Houston Lighting & Power Company Page 3 cc:

Regional Administrator, Region IV.- M.B. Lee /J.E. Malaski Nuclear Regulatory Commission City of Austin 611 Ryan Plaza Drive, Suite 1000' P.O. Box 1088 Arlington, TX' 76011 Austin, TX 78767-8814 N. Prasad Kadambi, Project Manager A. von Rosenberg/M.T. Hardt U.S. Nuclear Regulatory Commission City Public Service Board 7920 Norfolk Avenue P.O. Box 1771 Bethesda, MD 20814 San Antonio, TX 78296 Robert L. Perch, Project Manager Advisory Committee on Reactor Safeguards U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue 1717 H Street Bethesda, MD 20814 Washington, DC 20555 Dan-R. Carpenter Senior Resident Inspector /0perations c/o U.S. Nuclear Regulatory Commission-P.O. Box 910 Bay City, TX .77414 Claude E. Johnson Senior Resident Inspector / Construction c/o U.S. Nuclear Regulatory Commission P.O. Box 910 Bay City, TX 77414

~M.D. Schwarz, Jr., Esquire Baker & Botts One Shell Plaza Houston, TX 77002 J.R. Newman, Esquire Newman & Holtzinger, P.C.

1615 L Street, N.W.

Washington, DC 20036 T.V. Shockley/R.L. Range Central Power & Light Company P. O. Box 2121 s Corpus Christi, TX 78403

STP FSAR ATTACHMENT ST.HL.AE. 27_g PAGE / OF ll 6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures Inside the Containment. Following a postulated main steam line break or a main feedwater line break inside the Containment, the contents of one SG will be released to the Containment. Most of the contents of the other SGs will be isolated by the main steam isolation valves (MSIVs) and main feedwater isolation valves. Cotitainment pressurization following a secondary side rupture depends on how much of the break fluid enters the Con-tainment atmosphere as steam. Main steam line break flows can be pure steam or two-phase, while main feedwater line break flows are two-phase. With a pure steam blowdown, all of the break flow enters the Containment vapor space atmosphere. With two-phase blowdown, part of the liquid in the break flow boils off in the Containment and is added to the vapor space atmosphere, while the remaining liquid falls to the sump and contributes nothing to Containment pressuri::ation. For main steam line break cases with large break area, steam cannot escape fast enough from the two-phase region of the ruptured SG, and the two-phase level rises rapidly to the steam line nozzle. A two-phase blowdown results. The duration of this blowdown is short, therefore reducing primary-to-secondary heat transfer, and the break flow is largely liquid.

For main steam line break cases with small break areas, steam can escape fast enough from the two-phase region of the SG with the ruptured line that the level swell does not reach the steam line nozzle, and a pure steam blowdown results. Because of the pressure-reducing effects of active and passive Con-tainment heat sinks, the highest peak Containment pressure resulting from a main steam line break for a given set of initial SG conditions occurs for that case where the break area is the maximum at which a pure steam blowdown can occur. For conservatism, the main steam line break analysis assumed only pure steam blowdown for all break sizes and power levels.

Main steam line isolation is initiated on the following signals: high-2 Containment pressure, low steamline pressure or low-low T (above P-ll setpoint),highnegativesteamlinepressurerate(belowtMD-11setpoint), Si and manual. Main feedwater line isolation is initiated by SG High-High water level, excessive cooldown protection signal, reactor trip in conjunction with low T and SI. Both the MSIVs and the main feedvater isolation valves are4 2 fully *No, sed in 5 seconds, e-- Incea 1-gg4g The Auxiliary Feedwater System functions automatically following a secondary 49 system line break to assure that a heat sink is always available to the RCS by supplying cold feedwater to the SGs. For conservatism, it was assumed that the Auxiliary Feedwater System attains full flow to the SG immediately follow- I ing feedwater isolation. In addi: ion, the analysis includes the flashing of the volume of fluid located between the main Feer' water isolation valve and the 49 affected SG. This fluid then flows through the affected SG and into the Containment.

The feedwater enters the SG in the two-phase region; therefore, main feedwater line break cases always result in two phase blowdowns through smaller size lines and do not produce peak Containment pressures as severe as main steam line break cases.

To permit a determination of t.he effect of main steam line break upon Contain- 14 ment pressure, a spectrum of break sizes was assumed to occur inside the Con- l tainment, downstream from the integral steam line flow restrictors and up j stream of the MSIVs. Unrestricted critical flow from the rupture was assumed.

6.2-21 Amendment 57

r ATTACHMENT ST HL AE. 226 3 PAGE 2-OF,_1l Insert (Pg. 6.2.21)

This analysis has been evaluated with a main feedwater isolation valve closure time of 10 seconds. The results have been found to be acceptable as described in Reference 6.2.1.4-2.

I- .

STP FSAR ATTACHMENT ST.HL.AE 11' 3 l

,PAGE 3 0F )\

starting at time zero and continuing until manuallyTatopped by the plant l operator. Although operator action after 10 minutes following the break is

-) anticipated, it is conservatively assumed in this analysis that auxiliary l

{ feedwater is manually terminated after 30 minutes.

The Auxiliary Feedwater System design is such that otly one auxiliary feed-water pump feeds each SG. The maximum flow that be delivered to a de-pressurized SG is 1210 gpm (pump runout flow). .ma flow is assumed, although the AFW flow control valve, in combination with the AFW flow element and the ,

i Qualified Display Processing System (QDPS); (described in Sections 7.5 and 49 10.4.9), begin to limit and control AW flow delivered to the SG at approx-imately 550 gpm.

l The mass and energy release data presented are conservatively based on an auxiliary feedwater addition of 1210 gpm to the SG with the broken line from 0 to 1800 seconds.

Fluid Stored in the Feedwater Piping Prior to Isolation.

The blowdown data were determined assuming a value of 260 fts of unisolated volume in each main feedline.

14 l Fluid Stored in the Steam Piping Prior to Isolation. Q222.

06 All the steam in the steam lines up to the check valves upstream of the tur-l bine steam chest (6520 fts) is assumed to be released to the Containment fol-

' i lowing the break for the case of the main steam line isolation valve (MSIV) failure. l Availability of Offsite Power Loss of offsite power following a steam line rupture would result in tripping of the reactor coolant pumps and steam driven main feedwater pumps, and delay 49.

of auxiliary feedwater initiation due to standby DG starting and sequencer loading delays. Each of these occurrences aids in mitigating the effects of the steam line break releases by either reducing the fluid inventory available to feed the blowdown or reducing the energy transferred from the primary cool-ant system to the SGs. Thus, blowdown cases which occur in conjunction with a LOOP are less severe than cases where offsite power is available. Therefore, 49 blowdown has been determined assuming offsite power is available. However, for purposes of determining the activation time of the CSS, the main steam 53 line breaks are conservatively assumed to occur simultaneously with a LOOP.

Safety System Failures

1. Failure of Main Feedwater Line Isolation Valve I f l There are two valves in series /open during power operation in-each feed-f water line and both areCdesigned) to close within 7 see after the feed-f water isolation setpoint is reached (2 see instrtmentation response time l including delays and 5 sec valve closing time).

)

)

6.2-25 Amendment 57

av

. i ATTACHMENT STP FSAR ST.HL AE 2261 PAGE 9 OF ll REFERENCES (Continued)

Section 6.2: ,

6.2.1.3-1 Shepard, R.M., H.W. Massie, R.H. Mark, and P.J. Doherty,

" Westinghouse Mass and Energy Release Data for. Containment Design," WCAP-8264-P-A, Proprietary (June 1975) and WCAP-8312-A, Revision 1, Nonproprietary (June 1975).

g, 6.2.1.4 1 Coets, J.M., " Marvel-A Digital Computer-Code for Transient Analysis of a Multiloop PWR System," WCAP-7909, June 1972.

6.2.1.5-1 Bordelon, F.M., Massie, H.W., Jr., Zordon, T.A., " Westinghouse' I Emergency Core Cooling System Evaluation Model Summary", 52 WCAP-8339, June 1974.

usNR.C (o ,Z . \. % - ?_ Housh n L '3Wg a @w leb 4 o 5T - M L- A L- 7 2b3 dcd<d be 17 i W B7.

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l 6.2-61 Amendment 51

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.STATTACHMENT

.) TABLE 6.2.4-2 (Continusd) HL AE U O PAGE 5 OF ii CONTAINMENT ISOLATION VALVES MAXIMUM ISOLATION TIME FENETRATION VALVE NO. FUNCTION (SECONDSJ OTHER AUTOMATIC VALVES (Continued)

(Steam Line Isolation Actuated Valves)

M-1 FV-7903A Main Steam Line Drains 5 ML.vaS orHail Auro (Fm

• Ca + rrnATs c.Irv-

.TsokLfi +wofad Vatves)

M-6 FV- 1 1 Feedwater Isolation F 10 M-83 FV-7192 Feedwater Isolation - SG- 5 P4aAa.ar<.A 6ypa.*

M-7 FV 7142 Feedwater Isolation fr /O M-84 FV-7191 Feedwater Isolation - Srr 5 Paale.o c44 SyP 58 M-8 FV-7143 Feedwater Isolation 5-10 M-94 FV 7189 Feedwater Isolation- SG 5 fanke.a.Lar. B ypn M-5 FV-7144 Feedwater Isolation X/0

)

M-95 FV-7190 Feedwater Isolation SG- 5 fm Bype M-5 FV-7145A Feedwater Isolation 5 Valve Bypass M-8 FV-7146A Feedwater Isolation 5 ,

Valve Bypass l M-7 FV-7147A Feedwater Isolation 5 V.alve Bypass M-6 FV-7148A Feedwater Isolation 5 Valve Bypass OTHER AUTOMATIC VALVES (Auxiliary Feedwater Initiation Actuated Valves)

M-62 FV-4150 Steam Generator Blowdown 35 M-65 FV-4151 Steam Generator Blowdown 35 6.2-237L Amendment 58 S

f ATTACHMENI ST-HL AE. zz 63 PAGE 6 0F it 1 8

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STP FSAR ATTACHMENT

• ST4%.AE Itb3 Two valvas era providzd in each FW lina antering the Ccc._GE M . .._.io.m 0F. t)2ne one closest to the penetration is a swing check valve while the upstream valve is

, a hydraulically operated, fail-closed stop valve. This arrangement satisfies

the requirements for isolation for postulated accident conditions (see Chapter 15).

Io With loss of flow in the normal direction, the check valve closes to prevent outflow from the Containment u til the stop valve can be closed. The stop valve is designed to close in seconds or less. Thus, failure in the non-safety class portion of the FW system has no effect on the safety of the l39 reactor, which can be shut down in an orderly manner; neither will it result in the release of a significant amount of radioactive material to the environ-I ment. In the unlikely event of a piping rupture in the FW system, with the 39 M

i 1

i I

l s

I 10.4-20a Amendment 45

l STP FSAR ATTACHMENT ST-HL AE. 2h3 PAGE ll 0F ll _

TABLE 15.1-1

_} -

TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH CAUSE AN INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM i

Accident Event Time (sec)

Excessive feedwater One main feedwater 0.0 flow at full power control valve fails fully open Minimum DNBR occurs 27 18 High high steam generator 132 57 water level signal generated Turbine trip occurs 135 due to high-high steam generator. water level signal Reactor trip occurs 137 Feedwater isolation valves 139 -j(-

, close (due to high-high steam

) generator water level signal)

Excessive increase in secondary steam flow

1. Manual Reactor 10% step load increase 0.0 Control (Minimum moderator feedback)

Equilibrium conditions 150 18 reached (approximate time only)

2. Manual Reactor 10% step load increase 0.0 Control (Maximum moderator feedback)

Equilibrium conditions 100 reached (approximate time only)

• b MSm G. C. /. 4 %N -Pe oc/wN MctiLcL vct/v4.

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15.1-19 Amendment 57 i