ML20038B881

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Forwards Info Necessary to Close Open Issues Re FSAR & Tech Specs Changes in Primary Containment Isolation Valve Closure Time,Primary Containment Accident Pressure & HPCS Flow Rate. Changes Will Be Included in Next Amend to FSAR
ML20038B881
Person / Time
Site: LaSalle  
Issue date: 11/24/1981
From: Sargent C
COMMONWEALTH EDISON CO.
To: Schwencer A
Office of Nuclear Reactor Regulation
References
2943N, NUDOCS 8112090224
Download: ML20038B881 (37)
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Text

,

e' Commonwealth Edison

~'

one Fir *A Fluional Pttzt, Chicago. Ilknois -

C 5 Address Reply io: Post Office Box 767 Chicago, Illinois 60690 Novembe r 24, 1981 h @//'N Mr. A. Schwencer, Chief Licensing Branch #2 t

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Division of Licensing f

U.S. Nuclear Regulatory Commission

-] h Or / j',.

Washington, DC 20555

Subject:

LaSalle County Station Units fdnd 7,p% %/s

/

Proposed Technical ~Specificatign E,

Primary Containment Isolation \\yql

\\ \\,/).

f Valve Closure Time; Primary Containment Accident Pressure Change; High Pressure Core Spray Flow Rate NRC Docket Nos. 50-373/374

Dear Mr. Schwencer:

The. purpose of this transmittal is to provide information necessary to close open issues regarding the LaSalle County Station Technical Specification.

The following information is provided in this transmittal:-

1.

The FSAR change and associated proposed Technical Specification change reflecting the change in Primary Containment accident pressure from 32.5 psic to 39.6 psig.

2.

The FSAR change and proposed Technical Specification change reflecting a new isolation closure time of 5 seconds instead of I second.

This change applies to several containment monitoring valves and one Drywell Pneumatic Valve.

3.

The FSAR reference and proposed Technical Specification change reflecting a High Pressure Core Spray Flow of 6250 gallons per minute instead of 6350.

The proposed FSAR changes will be included in the next Amendment to the FSAR.

If there are any questions in'this regard, please contact this office.

Very truly yours,

(

C. E. Sa rgent Nuclear Licensing Administrator.

Enclosures cc:

NRC_ Resident Inspector LSCS

'8112090224 811124" PDR ADOCK 05000373

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,PDR.,_

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Etibg8 RECEIPT OF l,

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~.)I.iIllE'd'.L5 TABl E 3.6.3-1 (Continued) 1 gj 5

PRIMARY CONTAINMENT ISOLATION VALVES

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i

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APPLICABLE MAXIMUM p

OPERATIONAL ISOLATION TIME VALVE GROUP (a) -CONDITIONS (Seconds)

VALVE FUNCTION AND NUMBER

[

Automatic Isolation Valves (Continued) 11.

Containment Monitoting Valves 2

1,2,3 5

)

l ICM017A,B /

ICM018A.B'#

ICM019A,B ICM020A,8 /

' 1CM0218(h)/

1CM022A a

2 1CM0238

/

  • T(

ICM024A(h)b ICM025A t'

t.1CM026B(h)-

s T

ICM027 v O

ICM028 /

ICM029 /

ICM030 v ICM031 -

ICM032 /

ICM033 '

[.

1CM034 /

u mi 2

1,2,3 12.

Drywell Pneumatic Valves

< 40 IIN001A and B 7 30 11N017 lIN074 j4

" /se.

7 30 [

73 o

IIN075

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"[yt.t

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7 4

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su IIN018V p :r an g

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=

g 13.

RilR Shutdown Cooling Mode Valves

.6 1,2,3 5 41 1E12-F008

< 41 m

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D ~~~ U[]h h hh 3M.6 C40MIAINMENI SYSTEftS 3 3 /.l. 6.1 F111 MARY CONTAINMEfl1 _m,,y+ b PRIMARY CONTAINMENT IhTEGRITY LIMITING C0flDITIO!! FOR OPERATION 3.6.1.1 PRIliARY CONTAli MENT I!iTEGRITY shall be tr lintained. APPLICABILITY: OPERATIONf' CONDITIONS 1, 2,* and 3#, ACTION: Without PRIMARY CCilTAIN!!ENT INTEc71TY, restore PRIMARY CONT AIUMENT INTEGRITY within 1 hour er be in at least h0T SHUIDOWN within the next 12 hours and in COLD SHUTD0'ift within the felloaing 24 hours. SURVEILLANCE REQU!REMENTS

4. 6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated:

O Af ter each closing of each penetration subject to Type B testing, a. except the primary containment air Iccks,, if cpened following Type A or B test, by leak rate testing the seal with gas at Pa, Je6 ocia. gg and verifying that when the measured leakage rate for these seals is added to the leakage rates dM ermined pursuant tos5urveillance Requirement 4.6.1.2.d for all other Type S and C penetrations, the combined leakage rate is less than or aqual to 0.60 la. b. At least once per 21 days by verifying that all primry crntainment penetraticos " not capable of being closed by CPERABLE containment auto:ratic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in position. except as o mvided in Table 3.6. 3-1 of Speci fication 3.6. 3. c By verifying aach prirtary containment air lor' OPERABtE por Specification 3.6.1.3. d. By verifying the supprossion chamber OPERABLE por Specification 3.G.2.1.

  • See Special Test Exception 3.10.1
    • Except valves, blind flanges, and deactivated auton tic valves which are located inside the containment, and are locked, sealed or otherwise secured in the closed position.

These penetrations shall be verified closed during p each COLD SHUTOOWN except such verification need not be performed when the primary containmont has not been deinerted since the last verification or more often than once per 92 days.

  1. See Special Test Exception 3.10.7 909 5GD LA SALLE - UNIT 1 3M 01

O. c.u.su.c.w.[ey. n Cpt.s ]

y. n,

f CONTAI!DtENT SYSTEMS PRIhARY CONTAINMENT LEAKAGE. n !.IMITING CONDITION FOR OPER'ATION e 3.6.1.2 Primary containment lea ge rates shall be limited to: An overall integrated leakage rate of less than'or equal'to.L,, a. 0.635 percent by weight of the containment air per'24 hours:at'P, ~ 3

psig, b.

A conbined leakage rate of less than or equal.to 0.60 L, for all penetraticns and all' valves listed in. Table 3.6.3-1, except for main steam' isolation valves and valves which are hydrostatically leak tested per Table 3.6.3-1, subject to Type 3 and C tests when. pressurized to P, psig. 3

  • Less than or equal to 25 scf per hour for any one main steam

( c. isolation valve when tested'at 25.0 psig. O A combined leakage rate of less than or equal to 1 gpm times the d. ~ total number of ECCS and RCIC containment isolation valves in hydro-statically tested lines which penetrate the primary containnent, when tested at 1.10 P,, psig. APPLICABILITY: When PRIMARY CONTAINMENT INTEGRITY is required per Specification 3.6.1.1.- . ACTION: With: The measured overall integrated primary containment. leakage rate a. exceeding 0.75 L ' # a b. The measured combined leakage rate for all penetrations and all valves listed in Table 3.6.3-1, except for main steam isolation ^ valves and valves which are hydrostatically leak tested per Table 3.6.3-1, subject to Type B and C tests exceeding 0.60 L a The measured leakag'e rate exceeding 25 scf,per hour for any one l c. main steam isolation valve, or d. The measured combined leakage rate for all ECCS and RCIC containment isolation valves in hydrostatically tested lines which penetrate.the p primary containment exceeding 1 gpm times the total number of~such

valves,
  • Exemption to Appendix "J" of 10 CFR 50.

g N 2 8 19 91 LA SAL'tE - UNIT 1 3/4 6-2

~ _. i i .s V D .* h - L'ds;pnmf-r m r5 n. q r m dJa CONTAINMENT SYSTEMS L I LIMITING COND!710N FOR OPERATION (Continued) ACTION (Continued) ' restore: T.he overall integrated leakage rate (s).to less than or. equal to 0.75 a. L, and a 4 The combined leakage rate for all penetrations and all valves listed b. i in Table 3.6.3-1, except for main steam isolation valves and valves which-are hydrostatically leak tested per Table 3.6.3-1, subject = to Type'B and C tests to less than or equal to 0.60 L, and i a The leakage rate to less than or equal to 25 scf per hour for any l' c. one main steam isol.ation valve, and The combined leakage rate for all ECCS and RCIC containment isolation-d. valves in hydrostatically tested lines which penetrate the primary containment to less than or equal to 1 gpm times the total number of such valves, j prior to increasing reactor coolant system temperature abovd 200 F. SURVEILLANCE REQUIREMENTS The primary containment leakage ratas shall be demonstrated ~at the- ~

4. 6.1. 2 following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50 using the. methods and provisions ~of. ANSI N45s4 - 1972:

Three Type A Overall Integrated Containment Leakage Rate tests shall-a. be conducted at 40 t 10 month intervals during shutdcwn at Pg The third test of 3<f,(, D/T*psig, during each 10 year service period. I each set shall be conducted during the shutdown for the.10 year ~ plant inservice inspection. If any periodic Type A test fails to meet.75 L, the test schedule I b. a for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fall to meet.75 L,- a a l Type A test shall be performed at least every 18 months until two consecutive Type A tests meet.75 L, at which time'the above test' a schedule may be resumed. The accuracy of each type A test shall be verified by a supplemental I c. test which: Confirms the accuracy of the test _by verifying that the 1. difference between the supplemental data and the Type A test data is within 0.25 L ' a l Has duration sufficient to establish accurately the change in 2. g leakage rate between the Type.A test and the supplemental test. l' Requires the quantity of gas injected into the containment' or ~ l 3. . bled from the containment during the supplemental test to be equivalent to at least 25 pei :ent of the total measured leakage. i atP,,Jr.Tpsig. N4 3/4 6-3' OCT p g qq ..tir - UNI T'1. 'a a

l, CONTAINMENT SYSTEMS ~ ~' ' ' O SURVEILLANCE REQUIREMENTS (Continued) 97-psig*, at l d. Type B and C tests shall be conducted with gas at p, '2^.: a intervals no greater than 24 conths except for tests involving: 1. Air locks, ~ 2. Main steam line isolation valves, 3. Valves pressurized with fluid from a seal system, and 4 ECCS and RCIC containment isolation valves in hydrostatically l

  • ested line.s which penetrate the prinary containaent.

e. Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1. 3. f. Main steam line isolation valves shall be leak tested at least once per 18 months. g. Leakage f rom isolation valves that are sealed with fluid f rcm a se31 system may be excluded, subject to the provisicos of Arpendix<J, O Section III.C.3, when determining the combined leakage rate provided +gsealsystenandvalvesarepressurizedtoatleast1.10 a M psig, and the seal system capacity is adequate to naintain system prc:sure for at least ?0 days. h. ECC5 and RCIC containment isolation valves in hydrostatically tested l l'ines which penetrate the primary containment shall be leak tested at least once per 13 nonthr. i. The provisions of Specification 4.0.2 are not applicable te 24 o nth l or 40 10 month surveillance intervals. 'Unless a hyoraulic test is required per Table 3.6.3-1. l O 3EF 1 1961-LA SALT.E - UNIT 1 3/4 6-4

... m... i r" N CE-'lilbCI CONTAINMENT SYSTEMS b PRIMARY CONTAINMENT AIR LOCKS UMIi!NG CONDITION FOR OPERATICff 3.6.1.3 Each primary containment air lock shall be 0PERABLE.with: a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lack door shall be closed, and b. An overall air lock leakage rate of less than or equal to 0.05 La "t P, psig. a APPLICABILITY: OPERATIONAL CONDITIONS 1, 2* and 3. ACTION: a. 'dith one primary containment air lock decr inoperable: 1. Maintain at least the OPERABLE air lock dcor cle:ed and either restore the inoperable air lock dcor to CPERABLE status within 24 hours or lock the OPERASLE. air lock door closed. g;

2.. Operation may -hen continue until ;:erformance of the next required I

overall air lock leakage test provided that the OPERABLE air lock door is verified to be locknd closed at least ance per 31 days. [ 3. Otherwise, be in at least HOT SHUTDOWN within the na.xt 12 hcurs and l in COLD SHUTCOWN within the following 24 hours. 4 The provisions of Specification 3.0.4 are not ecplicable. l b. With the primary containment air lock inoperable, except as a result of an inoperable air lock door, maintain at least one air lack door closed; restore the inoperable air lock to OPERA 8LE status within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD ShdT00+N within the following 24 hours. ~ "See Special iest Exception 3.10.1. I l I a OCT 3 1 B50 LA 5ALLE - UNIT 1 3/4 6-5

t o l-(,((($ hhfh (-((. CONTAlta!ENT SYSTElls A SURVEILtdNCE REQUIREMENTS 4.6.1.3 Each primary containment. air lock shall be denonstrated OPERABLE: a. Within 72 hours following each closing, except when the air lock is- ] being used for multiple entries, then at least once per 72 hours', ti9' verifying seal leakage rate less than or equal to 5 scf per hour when j :G the gap between the door seals is pressurized to greater than or equal to 10 psig. $9b - - - 1. By conducting an' averall air lock leakage test at P,M psig, a an.d verifying that the overall ' air lock leakage rat.e is within its limit.* ~ ~ ~., ' ' ~ ~ 1. After eaqh opening, unless perfortred within the. previous G months', but at least once per 18 months *, anif 2. Prior to establishing PRIMARY C0!1TAINMENT INTEGRITY when maintenance has been performed on the air lock that could affect the air lock sealing capability.* A c c. _ At least once per 6 months by verifying that only one door in each ~ air lock can be coened at a time.** l Fine provisions of Specificatioc. 4.0.2 are not applicaule.

  • Exemption to 10 CFR 50, Appendix J.
    • Except that the inner dcor need not be opened to /e: 4f" iv erlock OPERABILITY when the primary contair. ment is inerted, providea that the inner door inter-lock is tested within 8 hours af ter the primary containment has t>een deinerted.

O LA SALLE - WIT 1 3/4 6-6

$ r N, 9. m_.( p..l Tu i V a g CONTAINMENT SYSTEMS BASES 3/4.5.2. OEPRESSURIZATION SYSTEMS The specifications of this section ensure that the primary containment pressure will not exceed the design pressure of 45 psig during primary system blowdown from full operating pressure. The suporession chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system. The sucpression chancer water volume must absort the associated decay and structural sensible heat released during reactor coolant system blowdown from 1020 psig. Since all of the gases in the drywell are purged into the suppression chamcer air space during a loss of coolant accident, the pressure of the liquid must not exceed 45 psig, the suppression chamcer maximum pressure. The design volume of the suppression chamcer, water and air, was obtained by considering that the total volume of reactor coolant and to be considered is discharged to the sucoression chamcer and that the drywell volume is purged to the suppression chamber. Using the minimum or maximum water volumes given in this specification,3 containment pressure during the design basis accident is approximately ~32.Ji F psig wnich f s below the design pnssun of 45 psig. Maximum water volume of 3 131,900 ft results jn a downcomer submergence of 12.4 ft and the minimu:n volume of 128,800 ft results in a submergence approximately 8 incnes less. The majority of the Bogeda tests were run with a submerged length of four feet and with comolete condensation. Thus, with respect to the downcomer submergence, this specification is adequate. Should it be necessary to make the suppression chamcer inoperable, this shall only be done as specified in Specification 3.5.3. Under full power operating conditions, blowdown from an initial suppression chamber water temperature of 90*F results in a water temperature of approximately 135'F immediately following blowdown which is below the 200*F used for complete l condensation via T-quencher devices. At this temperature and atmospheric pressure, the available NPSH exceeds that required by both the RHR and core spray pumps, thus there is no dependency on containment overpressure during the accident injection pnase. Experimuntal data indicates that excessive steam condensing loads can be avoided if the pean bulk temperature of the suppression pool is maintained belew 200*F during any period of relief valve operation with sonic conditions at the discharge exit for T quencher devices. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be deoressurized in a timely sanner to avoid the regime of potentially high suopression chamcer 1cadings. I l LA SALLE - UNIT 1 8 3/4 6-3 JAN 1 ; geg

.:.~. x_- fth h hN b CONTAINMENT SYSTEMS ( BASES 3/4.6.1.5 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained corrparable to the original design standards for the life of the facility. Structural integrity.is required to ensure that the containment will withstand the maximum pressure of 45 psig in the event of a LOCA. The measurement of containment tendon lift off force, the tensile tests of the tendon wires or strands, the visual examination of tendons, ancnorages and exposed interior and exterior surfaces of the containment, and the Type A leakage test are sufficient to demonstrate this capability. The surveillance requirements for demonstrating the containment's structural integrity are in compliance with the recommendations of Regulatory Guide 1.35 " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures," June 1974. 3/4.6.1.6 QRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE 7 N9 - 9 The limitations on drywell and suppression chamber-in,ternal pressure ensure that the containment peak pressure of 72$ psig does not exceed the (' design pressur'e of 45 psig during LOCA conditions or that the external pres-sure differential does not exceed the design maximum external pressure g differential of 5 psid. Thelimitof2.0psigforinitialpositiveg

r 5.ry :Ontair. ment pressure.ill limit the total pressure to E l psig which is 'ess,tnan the design pressure and is consistent witn tne ac:icent analysis.

3/4.6.1 7 DRYWELL AVERAGE AIR TEMPERATURE The limitation on drywell average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 340*F during LOCA conditions and is consistent with the accident analysis. 3/4.6.1.8 ORYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM The drywell and suppression chamber purge supply and exhaust isolation valves are required to be closed during plant operation except as required for inerting, de-inerting and pressure control. Until these valves have been d.monstrated capable of closing during a LOCA or steam line break accident, they shall be blocked so as not to open more than 50*. LA SALLE - UNIT 1 B 3/4 6-2 SEP 2 41981

1 1 ---

.- "f'

( 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive mate-rials free the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions. 3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE The limitations on primary containment leakage rates ensure that the i total containment leakage volume will not exceed the value assumed in the 3 q,6 a accident analyses at the peak accident pressure of it:+-psig, P. As an added conservatism, the measured overal,1 integrated leakage rati is 'urther limited to less than or equal to 0.75 t' during performance of the periodic taststoaccountforpossibledegradati8nofthecontainmentleakagebarriers between leakage tests. Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightnacs of the valves; therefore the special requirement for testing these valves. The surveillance testing for measuring leakage rates is consistent I with the requirements of Apcendix J to 10 wFR 50 with the exception of I exemotion(s) granted for win steam isolation valve leak testing and testing the airlocks after each opeling. l 3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS The timitations on closure and leak rate for the primary containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and the primary containment leakage rate given in Specifications 3.6.1.1 and 3.6.1.2. The specification makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured position during reactor operation. Only one closed door in each air lock is required to maintain the integrity of the containment. 3/4.6.1.4 MSIV LEAKAGE CONTROL SYSTEM Calculated doses resulting from the maximum leakage allowance for the main steamline isolation valves in the postulated LOCA situations would be a small fraction of the 10 CFR 100 guidelines provided the main steam line l system from the isolation valves up to and including the' turbine condenser remains intact. Operating experience has indicated that degradation has l cccasionally occurred in the leak tightr.ess of the MSIV's such that the scecified leakage requirements have not always been maintained continuously. The requirement for the leakage control system will reduce the untreated leakage from the isolation valves when isolation of the primary system and ( containment is required. LA SALLE - UNIT 1 B 3/4 6-1 M 2 " 19 9 t m -.-r--. .~# -.,,y-_.-%,,--- w.- .v,m-_-.,._ _._._y-

' * ^ ~ _ 5., ..5.. T mis tu m m.m s.:In OE ESS$3EE1$N tilF0!bE!5bM a Ic M'OM =Trute,3.s.3-1 (Continued) ? .g- - m PRIMARY CONTAINMENT ~150tATION VALVES-p r-MAX 1 HUM E: ISOLATION TIME VALVE GROUP (a) _ (Second VALVE FUNCTION AND NUMBER E Automatic Isolation! Valves-(Continued) w 19. Feedwater Testable Check Valves 2 NA: 1821-F032 A'and B \\ b. Manual Isolation Valves ffA - 1. 1FC086 NA. 2. IFC113 NA r 3. IFC114 s w NA ) 4. 1FCll5 T - 5. 1MC027- _NA< ~ NA <- 6. IMC033 i I NA. 7. 15A042 NAL 8. ISA04G i But > 3 seconds, i The iirovisions of Specification 3.0.4 are not applicable. i (a). See Specificat. ion 3.3.2,. Table 3.3.2-1, for isolation signal (s) that operates-each valve group. (b).May be opened on an intermittent basis under administrative control. (c) Not closed by SLCS actuation.. l (d) Not closed by Trip Functions Sa, b or c, Specification 3.3.2,' Table:3.3.2-L (e) Not closed by Trip Functions 4a, c, d, c.or f of Specification 3.3.2, Table 3.3.2-L-(f) Opens on an isolation signal. u, Q (g) Also closed by @ ell pressure-high signal. l (h) Not subject. to.iype C leakage; test. :,ealed with. fluid from a seal system. (i) llydraulic test. at pf.(o psig.1 ,~o 1

  • h.

m. g. e-

s r Pn00F & HEW CDP 7 CONTAINMENT SYSTEMS 3/4.6g PRDtARY CONTAINMENT AT140 SPHERE CONTROL i DRYWELL AND SUPPRESSICN CHAP 9ER HYOROGEN RECCfl01NER SYSTEMS l L' IMITEG CONDITION' FOR OPERATION .._m

3. 6. 6.1 Two independent drywell and suppression chamber hydrogen recombiner systeras shall be OPERA 8LE.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: With one drywell and/or suppression chamber hydrogen recombiner system inoperable. l restore the incoerable, system to OPERA 8LE status within,30 days or be in at least l HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS 4.6.6.1 Each drywell and suppression chamber hydrogen recembiner system shall l be demonstrated 0FERABLE: a. At least ones per 92 days by cycling each flow control valve and recirculation valvo through at least one complete cycle of full travel. ?< b. At least once per 6 months by verifying, during a recombiner system l functional test: i 1. That the heaters are OPERABLE by determining that the current l in each phase differs.by less thsn or equal to 5% from the other phases and is within 5% of the value observed in the original acceptance test, corrected for line voltage differences. 2. That the reaction chamber gas temperature increases to 1200 9'F within 2 hours. c. At least once per 18 months by: l j 1. Performing.a CHANNEL CALIBRAIION of all recombiner operating l instrumentation and control circuits. 2. Verifying the integrity of all heater electrical circuits by performing'a resistance to ground test within 30 minutes following the above required functional test. -The resistance to ground for j any heater phase shall bw greater than or equal to 100,000 ohms. d. By measuring the Icakage rate: l 1. As a part of the overall intSgrated leakage rate test required by Specification 3.6.1.2, or 2. Bymeasuringtneleakagerateofthesystemoutside5fthe containment isolation valves at P,,7gpsig, on the schedule O require'd by Spec 11(cation 4.6.1.2 an {ncluding the measured leakage as a part cf the leakage determined in accordance with Specification 4.6.1.2. I t.A SALLE - UNIT 1 3/4 6-37 JUl. 6 19 81

LSCS-FSAR AMENCMENT $ 6 MARC 2 TABLE 1.3-4 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS 8 PRIMARY COVTAINMENT* LA SALLE IIMMER 1 wpPss 2 RATCH 1 Type Over & under Over & under Over & under Pressure pressure pressure pressure suppression suppression suppression suppression Construction Concrete Concrete ASME Pressure Steel drywell post-tensioned post-tensicned vessel and with steel with steel suppression pool liner liner Drywell Frustum of Frustum of Frustum of Light bulb / coae upper cone upper cone upper steel vessel portion portion portion ' Pressure-suppression Cylindrical Cylindrical Cylindrical Torus / steel chamber lower portion lower portion lower portion vessel Pressure-suppression chamber internal design pressure, psig 45 62 45 56 Premeure-suppression chamber external design pressure, psi 5 2 2 Drywell internal design pressure, psii 45 45 45 56 Drywell external design pressure, psi 5 2 2 2 Drpell free volt e, ftJ 221,518 180,000** 202,242 146,240 Pressare-suppression N chamber free volane, ft3 165,100 93,000 144,166 110,950 Pressure-suppression pool water vola e, ft3 (min) 128,500 95,762 108,387 87,300 Submergence of vent pipe below pressare f' pool surf ace. 't 12.0 10 11.67 1.67 Design teTperature of drywell, "T 340 340 340 281 Design temperstare of pressure-suppression cnarter,

  • r 275 275 275 281 Downcomer vent pres-sure loss facter 1.9 2.17 1.9 6.21 Break area / total vent area 0.0105 0.008 0.0105 0.0194 Calculated c.awinum pressare aft-r bics-j$hI, f 7 down to drywoti, 40.4 37.2 46.5 psig 4-Pressure-suppressicn chamber, psi <

28 35.6 28 28 Initial pres 12re-suppression 23o1 temperature rise, *r 50 35 50 50 Leakage ra te, 9 free volure/ day a. 45 f1 psig and 340' r 0.635 0.635 0.5 at 200* r 1.2 at 59 pst5 f ^

  • Where appl.cible. Contatnment parameters are bastd on design power.
    • This value anc Lw!c s the vent volure.

1.3-9

dI LSCS-FSAR the drywell and suppression chamber, and are evenly distributed h around the suppression chamber air volume to prevent any W possibility of localized pressure gradients from occurring due to geometry. The vacuum relief valves are instrumented with redundant position indication and will be indicated in the main control room. The valves are provided with the capability for remote manual testing from a local instrument panel. This design provides adequate assurance of limiting the diffe ential pressure between the drywell and suppression chamber and assures proper valve operaticn and testing during normal plant operation. No vacuum relief valves are provided between the drywell and the reactor building atmosphere. The concrete centainment structure has the ability to accommodate subatmospheric pressures of approximately 5 psi absolute. 6.2.1.1.3 Desian Evaluation The key design parameters for the pressure suppression containment being provided for the La Salle County Station (LSCS) are listed in Table 6.2-1. (~N These desian paremeters are not determined from a single accident (.. event but from an envelope of accident conditions. As a result, there is no single design-basis accider.: (DBA) for this ( 39,4 containment system. 30.6 A maxi "a drywell and suppression chamber pressure of.275 psic and Glir psig, respectively is predicted near the end of the blowdown phase of a loss-of-coolant accident (LOCA) transi.. Approximately the same ceak pressure cccurs for either the brenh of a recirculation line or a main steamline. Both accidents are evaluated. The most severe drywell temperature condition is predicted for a small primary system rupture above the reactor water level that results in the blowdcwn of reactor steam to the drywell. Based upon the thermodynamic conditions this would produce high temperature steam ir. the drywell. In order to demonstrate that breaks snaller than the rupture of the largest primary system pipe will not exceed the containment design parameters, the blowdown phase of an intermediate size break is evaltated. Containment desicn conditions are not exceeded for t nis or the other break s:.2 's. All of the analyses assume that the prr. mary system and g'x contait ent are at the maximum normal operating conditions. ') References are provided that describe 'elevant experimental 6.2-7

/)Bfff LSCS-FSAR Case A - Offsite Power Available All ECCS equipment and containment spray operating. Case B - Loss of Offsite Power Minimum diesel power available for ECCS and containment spray. Case C - Same as Case B (excent no containment scrav) Initial Conditions for Accident Analyses Table 6.2-3 provides the initial reactor coolant system and containment conditions used in all the accident response evaluations. The tabulation includes parameters for the reactor, the drywell, the suppression chamber and the vent system. Table 6.2-4 provides the initial conditions and numerical values assumed for the recirculation line break accident as well as the sources of energy considered prior to the postulated pipe rupture. The assumed conditions for the reactor blowdown are also provided. The mass and energy release sources and rates for the containment response analyses are given in Subsection 6.2.1.3. Short-Tern Accident Resconse 39,6 The calculated containment pressure and temperature res nses for the recirculation line break are shown in Figures 6.2-2(and 6.2 ~ respectively. The calculated peak drywell pressure is 43E9 psig, which is below the containment design pressure of 45 psig. The suppressio/c, /2 /3/ noncondensables from the dryy,dcaurized by the carryover of

r. chamber is pr ell and hv heatup of the suppression pool.

As the vapor formed i,n the drywell is condensed in the suppression pool, the '.emperature of the suppression chamber water approaches 1400 F and hhe suppression chamber pressure stabilizes at approximately 43 psig. The drywell pressure stabilizes at z slightly higher pressure, the difference being equal to the dc wnconer submergence. During the FPV depressurization phase, most of the noncondensable gases in the drywell initially are forced into the suporession chamber. However, following the depressurization the noncondensables will redistribute between the drywell and suppression chamber via the vacuum breaker system. This redistribution takes place as preusure is decreased by the steam condensatien process occurring in the drywell. The LPCI and LICS systems supply suf ficient core cooling wa ter to control core heatup and limit metal-water reaction to less than p 3 0.2%. After the RPV is flooded to the height of the jet pump / nozzles, the excess flow discharges through the recirculation 6.2-12

A d[sf}P LSCS-FSAR line break into the drywell. This flow of water (steam flow is negligible) transports the core decay heat out of the RPV, through the broken recirculation line, in the form of hot water which flows into the suppression chamber via the drywell to suppression chamber vent system. This flow, in additien to heat losses to the drywell walls, provides a heat sink for the drywell atmosphere, causes a depressurization cf the containment, and redistributes the noncondensables as the steam in the drywell is condensed. ffra?! Sew,r Table 0. 2-5 provides the peak pressure, temperature, and timt paramaters for the re irculation line break as predicted for the conditions of Table 6 2-1 and in correspondence with Figures 6.2-2 and 6.2-3. The peak calculated drywell floor (deck) differential pressure is '1337 psid, which is 6357% below the design [dif ferential pressure of 25 psid. M70 .sustaixed u24. z, During the blowdown period of the LOCA, the pressure suppressicn vent system conducts the flow of the steam-water gas mixture in the drywell to the suppression pool for condensation of the steam. The pressure dif ferential between the drywell and suppression pool controls this flow versus time. Figure 6.2-4 provides the mass flow versus time relationship through the vent system for this accident. Lona-Tern Accident Resoonses In order to assess the adequacy of the containment following the initial blowdown transient, an analysie was made of the long-term temperature and pressure response follcwing the accident. The analysis assumptions are those discussed previously for the three cases of interest. The initial pressure response of the containment (the first 600 seconds af ter the break) is the same for each case. C se A - All ECCS Ecuienent ooeratina (with containment scrav) This case assumes that offsite a-c power is available to operate all cooling systems. During the first 600 seconds following the pipe break, the high-pressure core spray (HPCS), low pressure core spray (LPCS), and all three LPCI pumps are assumed operating. All flow is injected directly into the reactor vessel. After 600 seconds, both RHR heat exchangers ure activated to remove energy from the containment. During this mode of operation the flow from two of the LPC) pumps is routed through the RHR heat exchanger, where it is cocled before being discharged into the containment spray l eader. The containment pressure response to ttis set of conditions is shown as curve A in Figure 6.2-5. The corresponding drywell and / ) suppression pool temperature responses are shown as curve A in Figures 6.2-6 and 6.2-7. After the initial blowdown and 6.2-13 L

LSCS-FSAR /7AJ,7[ 6.2.1.1.3.1.2 Main steamline Break The sequence of events immediately following the rupture of a main steamline between the reactor vessel and the flow limiter has been determined. The flow on both sides of the break will accelerate to the maximum allcwed by critical flow considerations. In the side adjacent to the reactor vessel, the flow will correspond to critical flow in the 2.98-fte steamline cross section. Blowdown through the other side of the break can occur because the steamlines are all interconnected at a point upstream of the turbine by the bypass header. This intel connection allows primary system fluid to flow frnm the three unbroken steamlines, through the header ca.d back into the drywell via the broken line. Flow will be limited by critical flow in the 0.94-ft2 steamline flow restrictor. The total effective flow area is thus 3.92 ft2, which is the sum cf the steamline cross-sectional area and the flow restrictor area. Subsection 6.2.1.3 provides information on the mass and energy release rates. Immediately follcwing the break, the total steam flow rate leaving the vessel would be approximately 12,000 lb/sec, which exceeds the steam generati( n rate in the core of 4,500 lb/sec. This steam ficw to steam generation mismatch causes an initial depressurization of the reactor vessel at a rate of 50 psi /sec. The void formation in the reactor vessel water causes a rapid rise in the water level, and it is conservatively assumed that the water level reaches the vessel steam nozzles 1 recond after the break occurs. The water level rire time of 1 second is the minimum that could occur under any retcter operating condition. From that time on, a two phase mixture would be discharged from the break. During the first second of the blowdown, the bicwdem flow will consist of saturated reactor steam. This steam will enter the containment in a super-heated condition of approximately 3300 F. Figures 6. 2-9 and 6.2-10 show the pressure and temperature response of the drywell and containment during the primary system blowdown phase of the accident. Figure 6.2-10 shows that the drywell atmosphere temperature approaches 3300 F after 1 second of primary system steam blowdown. At that time, the water level in the vessel will reach the steamline nozzle elevation and the blowdown flow will changa to a two-phase mixture. This increased flow causes a more rapid drywell pressure rise. Ecwever, the peak differential pressure is (E29>psid, which occurs shortly after the vent clearing transient. As the blowdown preceeds, the primary system pressure and fluid inventory will decrease and this will result in reduced break {flowrates. M.R As a consequence, the flow rate in the vent system also starts to y~_) decrease, and this results in a decreasing differential pressure between the drywell and containment. 6.2-16

/}/jff[f i LSCS-FSAR M M a a RT D D g H P + y44 y y + 144 y W + y- =P y D D s s w which can be solved for the unknown air masses. The total pressures can then be determined. 6.2.1.1.4 Necative Pressure Desian Evaluation Containment negative pressure has been addressed in Chapter 3.0 and ir. the Design Assessment Rcport. 6.2.1.1.S Sucoression Pool Evrars Effects Protection Acainst Evcass Paths The pressure boundary between drywell and suppression chamber including the vent pipes, vent header, and downcomers are fabrica ted, ere cte d, and inspected by nondestructive examination methods in accordance with and to the acceptance standards of the ASME Code Section III, Subsection B, 1971 (Summer 197 2 Addenda). This special construction, inspection and quality control ensures the integrity of this boundary. The design pressure and temperature for this boundary was established at 25 psig and 3400 F, which is substantially greater than conditions during a (1 DBA. Actual peak accident dif ferential pressure and temperature acrose this boundary will be less than d23s3 psid and approxin7tely QiHZ2P during a LOCA. In ddition a stainless steel liner ha been provided between t e drywell and the wetwell as described i Chapter 3.0. _ ;pg, g -286 F - All penetrations of this boundary except the vacuum breaker seats, are welded. Allpenetrationsareavailableforperiodicvisual_f' inspection. Cmand su ftesaioy pot >{ /bytafull_ WOd V>t 9 & pen e fra.Oca J Reactor Blowdown Conditions and Ocerator Pbsconse In the highly unlikely event of a reactor depressurization to the drywell accompanied by a simultaneous open bypass path between the drywell and suppressicn chamber, several postulated conditions nay occur. For a given primary system break area, the maximum allowable leakage capacity can be determined when the containment pressure reaches the design pressure at the end of reactor blovdown. The most limiting conditions would occur for those primary system break sizes which do not cause rapid reacter depressurization. This corresponds to breaks of less than approximately 0.4 ft2 which require some operator acticn to terminate the reactor blowdown. Immediately after the postulated conditions given above for a /~) small primary system break, there would be a f airly rapid rise in 2 containment pressure as the noncondencable gases in the drywell are carried over to the suppression chamber. During this portion 6.2-25

AMENDMENT 44n 5 7 LSCS-FSAR MARCH 19)f integrity tests as described in Subsection 3.R.l.7, a preoperational containment leakage rate test is oreformed to verify that the actual containment leak rate does not exceed the design limits. In order to ensure a successful integrated leak rate test, local

  • leakage tests (Type B and C tests) are performed on penetrations and isolation valves, and repairs are made, if necessary, to ensure that leakage through the containment' isolation barriers does not exceed the design limits.

An integrated leakage rate test is then performed on the entire containment in order to determine that the total leakage (exclusive of MSIV leakage) through containment ~[ isolation barriers does not exceed the maximum allowable leakage rate of 0.635% per day at the calculated peak contain-ment internal pressure at d35) psig. The pertinent test l data, including test pressures and acceptance criteria, is presented in Table 6.2-23. L g Pretest requirements have been described in the pre-operational test abstract included in Chacter 14.0. As p$ stated therein, power operated isolation salves will be closed by their actuators prior to the start of the in-tegrated leakage rate test. During the integrated leak rate test the containment systems will be configured as follows, a. Reactor building closed cooling water - lined up for normal operation; isolation valves closed and system filled. b. Primary containment chilled water - lined up for normal operation; isolation valves closed and system filled. c. Resi al heat removal - lined up in low-pressure coolant injection standby mode, containment and suppression pool spray flow paths isolated, full flow test lines isolated, reactor head cooling flow path isolated, minimum flow isolated, shutdown cooling suction and discharge lines isolated, and condensate discharge from RHR heat exchangers shell side flow path isolated; system filled. (Note: Subsequent to commercial operation, the RHR system may need to be placed in the shutdown cooling mode to main-tain a consr. ant reactor vessel temperature. When this re-quirement e::is ts, one loop of the RHR system will be placed g in the shutdown cooling mode throughout the Type A test. ) d. Low-pressure core spray - system filled and isolated. a 6.2-58

AMENDMENT 5,41IfI J LSCS-FSAR JANUARYJM21 TABLE 6.2-1 CONTAINMENT DESIGN PARAMETERS SUPPRESSION DRYWELL CHAMBER A. Drywell and Suppression Chamber 1 1. Internal design pressure, psig 45 d5 2. External design pressure, psig 5 5 3. Drywell deck design differ-ential pressure, psid a) Downward 25 25 b) Upward 5 5 4. Desigt temperature, 'F 340 275 5. Drywell (including vents) (- ) net free volume, ft3 221,513 6. Design leak ratio, %/ day @ 45 psig 0.5 0.5 7. Suppression chamber free volume, ft3 165, 100 8. Suppression chamber water j volume J a) Minimum, ft3 128,800 ) b) Maximum, ft3 131,900 9. Pool cross-section area, ft 4999' ef [wk & 93 6. Water surf ace (cycluebs p? Co/aavs) a) kd dfd (/ cot ~syffs2 b) Total 5899 10. Pool depth (normal), ft 26.5 s u 6.2-62 L

LSCS-FSAR [' TABLE 6.2-1 (Cont'd) SUPPRESSION DRYWELL CHAMBER B. Vent System 1. Number of downconers 98 2. Internal downcomer diameter, in. 23.5 3. Total vent area, ft2M 295 i f jp ' g'f,y 4. Downcomer submergence, 7 -12.0 5. Downcomer loss factor

  • Q

_.s c93D fb bO$g) c y 4 Tlt e OcfttcLf //7)f///>f} Q YCO.. S $$e OfOfistj 5ize /brougl2 /kg clco)y&c7;fgr,oyofe[M covers (Ay/>u/s). 2h correspcxd/e, tos s 3 2. //oweveC, Stivou YW e ora /yse,s jac[br is eeyu<res Mal e.ri//axce /vssos, pip e lo sses am/ en/ /osses La base <f or; a. s;)y/e. '//Ie 4/ / fa c l5, o f 5. >3 uM/zect,g ien /ounesu//iug is a 6yter

arew, aos preusure.

oi<d, -/4srefore, a xvore. ouw uabie axchal s. 8 6.2-63 1

c LSCS-FSAR ffgg TABLE 6.2-3 (Cont'd) ( Suppression B. Containment Drywell Chamber 1. Pressure, psig 0.75 0.75 2. Inside temperature, F 135 100 3. Outside temperature, F 104 104 4. Relative humidity, % 20 100 5. Service water temperature, *F 100 100 128,200 6. Water volume, ft3 (zvj,,j;wy;,c) Q4h46fr 7. Vent submergence, k[z/AX m/) 42. ~ / /2 - 4 ,ry a %_) b 8 6.2-67

LSCS-FSAR AMENDMENT 2' MAY 197 TABLE 6.2-8

SUMMARY

OF ACCIDENT RESULTS FOR CONTAINMENT RESPONSE TO RECIRCULATION LINE AND STEA'4LINE BREAKS A. Accident Parameters RECIRCULATION

  • STEAMLINE LINE BREAK BREAK

'b 1. Peak drywell pressure, psig WA5-32 2. Peak drywell deck differential W" ;;2 //,R 17.5 pressure, psid 3. Time (s) of peak pressures, see -2^ M RR 11 4. Peak drywell temperature,

  • F
22$ B8(o 320 5.

Peak suppression chamber 1 -2S 30,[a 25 pressure, psig 6. Time of peak suppression chamber 50 50 pressure, sec 7. Peak suppression pool temperature 136.5 100 during blowdown, F 8. Peak suppression pool temperatura, 200 long term,

  • F 9.

Calculated drywell margin, % W /R 10. Calculated suppression chamber nS-3 R margin, % 11. Calculated deck differential )I 3,8 pressure margin, 1 See Figures 6.2-2 and 6.2-5 for plots of pressures vs time. v-See Figures 6.2-3 and 6.2-7 for plots of temperatures vs time. / 6.2-72

LSCS-FSAR MAY 19,81 AMEN fI TABLE 6.2-21 (Cont'd) These valves are under continuous leakage test because they are always subjected to a differential pressure acting across the seat. Leakage through these valves is continuonsly monitored by the pressure switches in the pump discharge lines, which have a low alarm setpoint in the main control room. 9, Evan though a special Jeakage test is not merited on these valves for the reasons discussed above, a system leakage test to meet the requirements of Type C testing and as hereinafter described will be performed to ensure the leak-tightness of the ECCS and RCIC systems. The systems will be prcssurized with water to a minimum _ pressure of Qjds psig (peak drywell accident pressure) with the system totally isolated from primary containment. A leakage rate for the entire system will then be determined and compared to an acceptance limit based on site boundary dose considerations (10 CFR 100: ECCS subsystem leakage not to J exceed 1 gpm times number of valves in the. subsystem tes.ted. I) 30. The leakages through the Main Steamline valves will not be included in establishing the acceptance limits for the ccm-bined leakage in accordance with the 10 CFR 50, Appendix J, 7.(, Type B and C tests. Because the Main Steamlines are provided with a leakage control system, the leakage through these valves will not be added into the combined leakage rate. This exclusion is in accordance with Article III.C.3 of 10 CFR 50, Appendix J. 31. Although only one isolation valve signal is indicated for these valves, the valves also receive automatic signals from various system operational parametcrs. For example, the ECCS pump ninimum flow valves clost automatically when adequate flow is achieved in the system; the ECCS test lines close automatically on receipt of an accident signal. Although these signals are not considered isolation signals; and are therefore, excluded from this table, there are other system opetation signals that control these valves to ensure their proper position for safe shutdown. Reference to tha logic diagr ams for these valves indicates which other signals ~ close these valve 3. 32. To satisfy the requirements of General Design Criterion 56 and to per;.~orm their function, these instrument lines have been designed to meet the requirements of Regulatory Guide 1.11 (Safety Guide ll). These lines are seismic Category I and terminate in instru-ments that are Seismic Category I. They are provided with manual iso.'ation valves and excess flow check valves. 6.2-93k )

O O O ^ f TABLE 6.2-23 CONTAINMENT LEAKAGE TESTING LEAK RATES -at Pa (t /24 hours) TYPE OF TEST DESCRIPTION CALCULATED MAXIMUM TEST PER APPENDIX J OF PEAK PRESSURE ALLOWADLE DESIGN PRESSURE OF 10 CFR 50 TEST Pa (psig) (La) (Ld) Pt (psig) A Integrated Leak Rate 82YS37,4 0.635(3) 0.5 12 4 39 f, B Local Penetration Leakage Rate MidTh (1) (1) 3 % 37,6 g-m n Y M' C Local Containment (1) (2) 0.1 SCFH per N Isolation Valve, MJ7,6 inch of nomi-3% 37, g, Leakage Rate nal valve size at 50 psig 7_ I4I MSIV Leakage Rate 32v5'Jf, 6 8 7 11.5 scfh 25 (1)The combined leakage rate of all penetrations and valves exclusive of MSIV leakage subject to Type B and C tests shall be less than 0.60 La, as specified in Appendix J y to 10 CFR 50. g (2)See Table 6.2-21. >o NM (3) Exclusive of the MSIV leakage rates. (4) Exemption to 10 CFR 50, as stated in III C.3 of Appendix J. l y N \\ Vg L

. AMEfiE"ENT 50 1 OCTOBER 1980 f o d s lf a e. E } ~ E s/ 1 n= \\ oi E5 / 3 3 \\ a v ( o \\ f_ - o {z $b \\ E b '~ l l i ?0 5 ev O / I m. O l i l l l l l n l o o o o o o o i (sind) Junssind LA SALLE COUNTY STATION j FINAL S AFETY AN AL YSIS REPOR T i FIGURE 6.2-2 RECIRCULATION LINE BREAK PRESSURE RESPONSE

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  • 4 c
  • I

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==4 9 2 1 I LJ ~ g 13 i / \\ [f) n 20. .t1 T- -/ 0 * ' i i~b'- I i i ' . ~ h? N. 10 S -10 0 ^ l000 .,2', 4 LOCTIME T,v,c -(scco.cs) L e.m 0 u pa.2

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THIS PME 0?E?) PENDIN3 REEL?T 3F IPEMRGENCYiCORE C000jIkqc$03. !.'..:)I P' "M y W'""U3[ (.h m--

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t SURVEILLANCE REQUIREMENTS 4.5.1 ECCS ditisions 1, 2 and 3 shall be demonstrated OPERABLE by: a. At least once per 31 days for the LPCS, LPCI and HPCS systems: 1. Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water. 2. Perfomance of 'a CHANNEL FUNCTIONAL TEST of the: a) Discharge line " keep filled" pressure alarm instrumentationi and b) Header delta P instrumentation. 3. Verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. b. Verifying that, when tested pursuant to Specification 4.0.5, each: 1. LPCS pump develops a flow of at least 6350 rpm against a test line pressure greater than or equal to.N psig. bg (S / 2. LPCI pump develops a flow of at least 7200 gpm against a test line pressure greater than or equal to 1 0 ig. 7 }) / 4 3.. HPCS pump develops a flow of at least (6350) gpa against a test' l line pressure greater than or equal to 370 psig. c. For the LPCS, LPCI and HPCS ' systems, at least once per 18 months: 1. Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded from this test. c LA SALLE - UNIT 1 3/4 5-4 M 28 1981 -.}}

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