ML20148T173

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Forwards Responses to NRC Questions Re Open Issues from Ser. Hypothetical Gate Valve Leakage on ECCS Suction Lines Off Containment Suppression Pool Does Not Present Safety Concern
ML20148T173
Person / Time
Site: LaSalle  
Issue date: 02/20/1981
From: Delgeorge L
COMMONWEALTH EDISON CO.
To: Youngblood B
Office of Nuclear Reactor Regulation
References
NUDOCS 8102250030
Download: ML20148T173 (33)
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{{#Wiki_filter:_ _ _ - _.._ C:mm:nw:cith Edison oni First Nation:t Plaza, Chicago, lil'nois Address Reply to: Post Office Bcx 767 Chicago, filinois 60690 February 20, 1981 Mr. B. J. Youngblood, Chief Licensing Branch No. 1 Division of Licensing U.S. Nuclear Regulatoly Commission Washington, DC 20555

Subject:

LaSalle County Station Units 1 and 2 Resolution of Safety Evaluation Report Issues NRC Docket Nos. 50-373/374

Dear Mr. Youngblood:

In response to questions from the NRC Staff, information is provided herein to resolve remaining issues in the LaSalle County Safety Evaluation Report. The areas addressed are the following: 1. Reactor Systems Branch - Enclosure 1 (a) NUREG-0737 Item II.K.3(46) (b) Post-LOCA ECCS Leakage (c) Flow Control Valve Closure Analysis 2. Power Systems Branch - Enclosure 2 (a) Degradeo Grid (undervoltage protection design) (b) Electrical Penetrations (medium voltage circuits) 2 ..,l y 3. Chemical Engineering Branch - Enclosure 3 ~

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(a) Control Room Carpet (fire rating) 4. Containment Systems Branch - Enclosure 4 L (a) Appendix J Testing (FSAR Table 6.2-21 Notes 29 & 39) 5. Quality Assurance Branch - Enclosure 5 \\ N\\ (a) Q-List Requirements g 6. Financial Qualification Branch - Enclosure 6 (a) Operating Cost Estitaate (LSCS 1 1935 & 1986'; (b) LSCS 1 Commercial Service date 8302250030 6I

4 Mr. B. J. Youngblood, Chief February 20, 1981 Page 2 These materials are addressed in the referenced enclosures. Changes will be made in a future FSAR amendment as discussed in the enclosures. If there are any further questions in this regard, please direct thern to this of fice, very truly yours, sf L. O. DelGeorge Nuclear Licensing Administrator Enclosures 1-4 cc: NRC Resident Inspector - LSCS 07378

Reactor Systems Branch (a) NUREG-0737 Item II.K.3.(46) Commonwealth Edison has participated in the BWR Owners Group Review of the concerns reaised by the ACRS consultant (Mr. Michelson). the R. H. Buchholz letter to D. F. Ross dated February 21, 1980 has been reviewed for applicability to the LaSalle County design and is judged to adequately resolve the subject concerns for LaSalle County Station. The NRC request for additional information contained in the P.S. Check letter to T. Kennan dated June 24, 1980, and the BWR Owners Group response in the D. B. Waters letter to D. G. Eisenhut dated January 31, 1981 do not alter this conclusion. In the event future evaluation of the conditions discussed in II.K.3(46) are undertaken by the BWR Owners Group, Commonwealth Edison will participate. This information will be documented in Appendix 1 of the FSAR in a future amendment. (b) Post LOCA ECCS Leakage Commonwealth Edison has considered the hypothetical leakage of gate valves on the ECCS suction lines off the containment suppression pool, and concludes that such leakage does not present a safety concern. The bases for this conclusion are discussed in detail in the attachment to this enclosure. (c) Flow Control Valve Closure Analysis Based upon conversations between the NRC Staff and General Electric on February 19, 1981, it is concluded that this issue is resolved. The issue presented concerned the effects of power supply failure to the P transmitter and FCV electronic controller on the LOCA analysis. It was verified by GE that the P transmitter and valve electronic controls are on separate 24 V ac power supplies fed from the same 120 V ac bus. Failure of the bus would result in valve failure in the as-is position. Because such a failure would not result in rapid closure, the reported LOCA analysis results are not effected.

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the leakage associated with one division of the ECCS will not 7,,, ; ' O.;.; cdversely' effect the other ECCS division [s due to flooding. ,a Ao described in section 3.4.1, the BCCS pumps are located in ccparate corner cubicles which are water-tight to an elevation (. m. (68667") equivalent to that associated with the resultant aquilibrium water level from a hypothetical suppression pool . rupture. Thus, even if flooding from an ECCS valve packing J 1g;y.. wdh. inak is assumed to drain the pool to a level at which the water:./~2Q, outside the primary containment equalises with that insidei [ there are no adverse effects on. plant saf'etye.. a. Tho second area of concern pertains to the ability to =mba-up .Q[. . v. tha volume of suppression pool water 'that wonid be lost.2from the postulated leakage and' the effect that. this. loss. of water level migh,t have on ECCS pump performance due to inadequate NPSH. It srust be pointed out that there are a myriad of sources of make-up water to the suppression pool or the primary system that could be utilized to replace the water lost through leakage. Two readily available sources are the CRD pumps, and the HPCS 1. 6 .-_.,,,._....,_._,.....,m........_ ,.,..__,,4

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I.scs-Fsam f** ~. w:: - - ..e s ,. ~.... .m ' T.,, , system, either of which has more than sufficient capacity to m( transfer water from the cycled condensate tank to the primary p, :- k,7 leakage. Other 1 system to make-up any water lost through sources which require no special access include the feedwater system, condensate / condensate booster system, and HPCS a If access to the reactor building is available, water leg pumps. lines. there are many manual valved connections between the BCCS _g:- connected to the pool or RPV and the cycled condensate sydtem. Therefore, adequate make-up is not a problem.' - It should also be noted that leakage from the suppression pool,. # .w ' i J 49, into the floor drain sumps would be pumped to the radweste c the water' is then returned to the..,,'_ system. After treatment, ? cycled condensate system where it would be available for-Therefore, a closed inventory of water would. leakage make-up. be avaiable to replenish any pool leakage. b Cs With regard to the effect of the decreased pool level on EC 6.3-8 pump performance, please refer to FSAR Figures 6.3.4u 12 the NPSE As can be seen, snd 6,3-11 and subsection 6.3.2.2.6. can be decreased available to the HPCS pump (governing case) by 5 feet, corresponding to a pool level decrease of this before any adyt:ese effects on ECCS pump operation same amount, i 4 .,,.:.,,--.,.n> ,.,n.. .n..e.n,

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as (,. LSCS-FSAR AMENDMENT @ 6 f.. 'v QpESTION 040.106 "Concer'ning the response relating to Regulatory Guide 1.63, address each of the following items. 7;__ The response states that the electrical penetrations (1) for the La Salle County Stations are designed with ~ oversized conductors through the penetration seals such that' they can ~ withstand any conceivable fault current versus time condition not interr,upted by the primary ~' - protective device. Provide a quantified meaning for the terms 'any conceivable fault current' for the 6900, 4160 and 480 volt electrical penetrations. Also quantify the maximum time duration for which this conceivable fault current is to be applied to each of the above three types of electrical pen-etrations. The response also states that 'should the backup (2) protective device also fail to clear the fault, the ratio of conductor size within the seals to the conductor size of the external circuit is such that failure of the electrical circuit external to the penetration seals is calculated to occur prior to failure of the penetration conductors.' (3) With regard to this item, provide the detail calcula-tions which demonstrate tha abov?. ( The response also appears to reference subsection (3) 3.8.1 for test information which verifies that the capability of the electrical penetrations circuits as stated in item 2 above has been confirmed by This subsection does not address the testing testing. Accordingly, of the penetration circuits in this manner. provide a short description of these tests as well as clearly stating the associated test results. l it is also In connection with this regulatory guide, the addition of a LOCA signal to trip (4) stated that the circuit overload protective device is one n:eans the requirements of Section 4 of available to meet IEEE Standard 317-1972. Identify all cases in the electrical design of the La Salle Stations whereby these requirements are met in this manner." RESPO!1SE 040.106-The requested infor.mation is provided in Tablealactrrrd h (1) 4160-voJ' cont dn e t g ote that LSCS_ hae_;-nppnet ;tionfhtSo were ruv f a the Gio o volt < t ectr a.t geweir.t:n.) toe. 1 90.!56-2. 7{ con de >,- c.l e r i 7. Joe.s not ta r.e a d u a m t s 3 e. *f o o9 e n i n. 1 c e o s - rc.3 r 1sar un u T 3 r rirney n J b d 9 e a p. relyi.n f., r rh e E; Et TN'*' bu+ Y a.+b e r h o. I F * " mt t ra o se r e4 f-(oE P A) p..r J K o m in Fia vr e o mi o Curve r S h w Ng 4[pk0 jg py ,w Q v, sr e p.v;ded ; y 7;J,res q q o, t o(,.3, c 'ro, toc-1,q 4o, joi xagg,. ( ? * *

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W LSCS-FSAR AMENDMENT 4D l \\ (2) and (3) Tests, which were conducted at the G-uld/ITE High Power Testing Laboratory in Chalfont, Pson. on Feb-ruary 28, 1978, demonstrated for the field cable /penetra- ~ tion combinations tested that the La Salle County Station design is valid without exception, provided that the penetration assembly conductors are at least one standard (AWG) size larger than'the connected field cable. The failure (fusing) of the field cabling with resulting inter-ruption of the test occurred before any damage occurred to the penetration assembly or pressure seal integrity. This was true for all cases where the relative size of cable followed the qualification combinations used at La Salle. Figure Q40.106-1 and Tables Q'sts and results. -r 40.106-2, 4 describe the te Q40.106-3, and Q40.106-4 The last column to the right in Table Q40.106-4 shows that for configurations identical to LSCS field / penetration cable interface sizings, the maximum thermal capacity of i the penetration conductors 1s f rom approximately(-lS9s~2 5 9 7-to approximate 1yyG.edt of the thermal load experienced ) '~dQr In~cj ~ ties t ing. Hence sufficient margin exists between N G b the penetration conductor's maximum thermal capacity and.the thermal load to field fusion for all the required LSCS field / penetration cable interface sizingsg [ ~ ---% a w t w & A ~ (4) The LSCS design utilizesA h'ii fusing propert'ibs of the cables external to theQ5enetration assemblies to meet ( the requirements of section 4 of IEEE 317-1971 and therefore, does not make use of a LOCA signal to trip circuit overcurrent ( protective devices. I {

- N g

( \\ \\ % AL S $ igurer Q Q o./o6 - 3, Q 40 /%-4 Q So.lo 6-4 Q So.lo'o-6 ar.d s f j \\ Q 40 /M.7 ill us t ra.t e. mt g le moxgin existr k e rween N c.( y

e. v (t c,v e r e n t p e,7g,c,9;on ud th e Case.rVat/V,

C b"* *l liWf c.@nay curve of th electrical / Pe nctr$ rio m ~ m_ s(, 040.10E-2 -___.__.-_.-.-_._._,,.._.g,-...,,,m..my=.5--.-n------. . ~..

~~~~ ~~ l ~ (.) =.

  1. 'l t

TABLE Q40.106-1 1 -l CONTAINMENT ELECTRICAL PENETRATION FAULT CURRENT MAGNITUDE AND A i \\ i ,3 1 TIME DURATION VALUE i it !) ] EXTERNAL (FIELD) INTERNAL TIME' CONDUCTOR CONDUCTOR FAULT CURRENT

  • DURATION i

PENETRATION TYPE SIZE SIZE _ amperes) (seconds) ( 's edi JA Vol-tdyG6wdEN --- - .',['.'.2.Z2:~T- ~ ~ ~ ' ~ ="7560':MCM -701RQ0 N 0,1,3W--==-= f (6900-V)- - -- -1600713dDTMCM*n r go h Low Voltage Power (480-V) 4/0 AWG -39 0' iCM 23,500 1.0 - Low Voltage Power (480-V)

  1. 2 AWG 250 MCM 17,500 1.0 i

h j2AWG 8,800 1.0 Low Voltage Power (480-V)

  1. 6 AWG i

1 o .I a Low Voltage Power (480-V)

  1. 10 AWG
  2. 2 AWG 3,600 1.0

( o 1 -pa Low Voltage Power and Control -

{

( 4 80-V) 414 AWG

  1. 10 AWG 1,400 1.0 l w

.)j j

1 id i

4 .~ ~ dcto @ i-ut:fdr-sWIMk HGR, !i r i 11. 1 t, 1 e k ~ n

i V C) %) %)... ' - .n -{ ,1 i 1 1 TABLE Q40.106-2

)

CONTAINMENT ELECTRIC PENETRATION ASSEMBLY (EPA) CABLE ) g FUSION QUALIFICATION TESTING 3 i a REF FLD EPA CURRENT DURATION TEST UIRE WIRE RMS TO NO. SIZE ** SIZE ** SYM AMPS FAIL (SEC) I t REMARKS 5 Al 514AWG fl0AWG 921* 0.620 5.26 (10 ) -Field cable failed near connection to power bus connection. Thermal damage to polyolefin heat shrink tubing near EPA /Fld cable splice on the fld. cable side. EPA feedthrough maintained 48 Psi 9 5 A2 ti4AWG $10AWG 1315* 0.250 ** 4.32 (10 ) Same as Al 2 A3

  1. 14AWG I10AWG 1340 0.310 5.57 (10 )

Fusible link in penetration circuit 5 5 as shown in Figure Q40.106-1. Link Q o L did not fuse. Field cable failure i near power bus connection. EPA feedthrough maintained 48'priig. g 5 A4

  1. 14AWG fl0AWG 1340 0.280 5.03 (10 )

Same as A3 0 B1

  1. 14 AWG I12AWG 1260 0.29 4.60 (10 )

Fusible link in circuit near connec-tion to power bus. Link did not fuse. Field cable failed on each side of fusible link.' EPA feedthrough' main- - tained 48 psig. yg Fusible link with electrical insulat- @W Q 5 B2 i14AWG

  1. 12AWG 1297*

0.29 4.90 (10 ) g ing tape in circuit near connection . a: .F to power bus. link did not fuse. gE H Field cable failed at power bus feedthrough maintainedhp connection. EPA 8"g{. 48 psig. 0 . n l 1 'i

i s

TABLE 040.106-2 (Cont'd) ( ,) ] REF FLD EPA CURRENT DURATION TEST WIRE WIRE RMS TO / f REMARKS SIZE ** SIZE ** SYM AMPS FAIL (SEC) I t t, 1:0. - 0 Fusible link in curciut near power bus ~ C1 fl0AWG 46AWG 1940 0.90 3.39 (10 ) connection. Link did not fuse. Field cable failed on each side of link. EPA j feedthrough maintained 48 psig. !f 6 ~ i C2

  1. 10AWG
  2. 6AWG 3280 0.295 3.17 (10 )

Field cable failed near both power bus l f connections. Flamh emjtting from field 'I cable insulation at\\ point of f ailure l l for approximately 10 seconds after fail - FPA feedthrouah heaintained 48 psig. ure. j 6 Same as C2 with no visible flame.but 1 C3 fl0AWG

  1. 6ANG 3300*

0.290 3.16 (10 ) much smoke emenating from field cable j jacketing. EPA feedthrough maintained j 48 psig. 4 t* o D1 010AWG fBAWG 3190 0.305 3.10 (10 ) Field cable failed at connection to one 8 i 6 side of power bus supply. Failure at y ^ EPA /Fld cable splice. Flame emitting M j o g frbm each end of field cable for 10 seconds after failure. EPA feedthrough o i maintained 48 psig. v' i i e 7 .I E1

  1. 6AWG
  2. 2AWG 1790 8.10 2.60 (10 )

Field cable fail.ed near power bus 1 connection at jacket.ing/ cable juncture. -l Visible flame and smoke from from field cable. EPA feedthrough pressure in-3 creased 5 psig above 48 psig. ]hl _{ 3z E2 56AWG

  1. 2AWG 4470 1.12 2.24 (10 )

Same as El. j@ 7 o tn E3

  1. 6MM 92AWG 7650 0.38 2.22 (10 )

Same as El. Electromagnetic forces noticeably bent penetration pigtails j,: l oa" F1 46AWG f4AWG 7940 ~ 0.38 2.40 (10 ) Same as E3. e j h 4 .r I i -~,

--9,.- s. 1 (r) .... _ _ _. _ _..(g. ll ' TABLE Q40.106-2 (Cont'd) J.. A 1 \\ REF FLD EPA CURRENT DURATION s I TEST WIRE WIRE RMS TO 2 NO. SIZE ** SIZE ** SYM AMPS FAIL (SEC) I t REMARKS 6 f' G1 fl0AWG

  1. 10AWG 1390 1.30 2.51 (10 )

Penetration conductor failed near connection to field cable. Inboard seals ] failed with EPA feedthrough pressure to increase tp 58 psig and then decreasing J to O psig rapidly.*** K1 84/0 24700* 1.86 1.13 (10') Cable fused with smoke and flame which l' ti self-extinguished in'approximately 10 seconds. l' K2 '82AWG 17800 0.30 9.51 (10 ) Same as K1 s 3 a e i' O M En 1 O li m i, t. Integrated average RMS Current. i

    • Reference Figure Q40.106-1 for Test Arrangement f

'. h

      • Although the feedthru inboard seal failed, contain-it ment integrity would have been maintained since the
)~

outboard feedthrough seal was undamaged and when tested possessed a leak rate that was less than h$ 1 x 10- ' sec/sec helium at 75 psig. e! ..M *r i ' *4 y ), eH W ~a a

  • O %8

+I i il I ,1

1 f TABLE Q40.106-3 PREDICTED FINAL TEMPERATURE THAT OCCURRED DURING CONTAINMEt!T ELECTRICAL PENETRATION 1.. ASSEMBLY (EPA) TESTING _ ~ EPA REF FLD 2 TEST WIRE SIZE WIRE SIZE Tf 'C Tf 'C NO. (AWG) (AWG) (AMP -SEC) (FIELD) (EPA) Al 14 10 5.2591 (10 ) . 2586 134 5 i A2t 14 10 4.3231 (10 ) 1601 110 Afg 14 10 5.5664 (10 ) 3013 143 -i 1 A4h 14 10 5.0277 (10 ) 2302 128 l 1 B1 14 12 4.6040 (10 ). 1854 349 - n2 14 12 4.8784 (10 ) 2134 379 C1 10 6 3.3872 (10 ) 2653 135 C2 10 6 3.1737 (10 ) 2242 127 C3 10 6 3.1581 (10 ) 2214 126-ou D1 10 8 3.1037 (10 ) 2120 376 .j Elb 6 2 2.5953 (10 ) 4455 164

t 2902 140

.] y E24 - 6 2 2.2379 (10 ) .,m E 6 2 2.2239 (10 ) 2853 139 .f d F1 6 4 2.3957 (10 ) 3511 499 1 G1 10 10 2.5117 (1C ) 1304 1304 1.1348 (10 ) 1566 ] K1 4/0 9.5052 (10 ) 1111 i' K2 2 c ? N $ CO71 fig ut o.-tion ' d N cal t. LS CS Eletdhenetr*&n

l i

cable interf.cc sig;n a A

.\\,

!!o I l

).. ' - ) O 1 c TABLi.}40.106-4 s THERMAL LOADING COMPARISON BETWEEN TEST VALUES AND MAXIMUM ,7 j CAPACITY VALUES FOR CONTAINMENT ELECTRICAL PENETRATION ASSEMBLY (EPA) t l i 2 i REF. ACTUAL FLD EPA g7 g - EPA TEST (I*t) ACT. CABLE CABLE, MAXIMUM EPA Ratio I NO. FROM TEST SIZE SIZE (I t) EPA CAPABILITY ** (I t) ACT. 6 Al* 5.26 (105) <14 AWG w :0 AWG.- 1.48 x 10 2.81 6 j A2* 4.32 (10 ) 14 AWG 10 AUG-1.48 x 10 <3.43< 6 A3* 5.57 (lN) 14 AWG 10 AWG 1.48 x 10 \\.2.66 6 x h A4* 5.03 (10 ) 14v AWG. 10' AWG3 1.48 x 10 24941 5 ) B1 4.60 (105) 14 LWG 12 AWG 5.87 x 10 .28 \\ 5 B2 4.80 (100) 14 AWG 12 AWG 5.'87 x 10 y,. 2 0 ^ 6 C1 3.39 (106) 10 AWG - 6 AWG 9.47 x 10 2.79 r.) 2 C2 3.17 (106) 10 AWG 6 AWG 9.47 x 10 2.99'. 6 6 ? C3 3.16 '(106) 10 AWG 6 AWG 9.47 x 10 3.00 6 -i H .j g D1 3.10 (106) 10 AWG 8 AWG 3.75 x 10 1.21 f 7 El* 2.60 (107) 61 AWG ' 2 AWGn 6.06 x 10

  • 2 a 3 3 8 Y

i j m 7 E2* 2.24 (10 ) 6 AWG 2 AWG 6.06 x 10 4.71' 3 E3* 2.22 (10 ) 6 AWG, 2'AWG' 6.06 x 10 ,2. 7 3' ~ Fl*** 2.40 (107) 6 AWG 4 AWG 2.40 x 10 - 1.00 6 Gl**** 2.51 (106) 10 AWG 0 Aht; x 10 f lh 10 6- 'Kl* 1.13 (109) 4/0 AWG.l 0-CM 0 l K: # 9.51 (10 ) 2 AWG 250 MCM 8.60 x 10 9.04 E .] h i

  • Configuration identical to LSCS field / penetration cable interface sizing.

'j

    • Based on maximum limiting conductor temperature of 499* C which manufacturer feels y

that' the insulation system can wi thstand under f ault condition. Thus ( 2 I t IT + 234 Ti = 18.33' C (65' F) l y j F-[W 0,9 Tg = 4W C D30' M. + 234 s T1 ( W

      • Case F1 is the baseline case for this analysis and is not a field / penetration cable W

5 interface that is used at LSCS. l

        • "ase Gl was a planned-configuration to demonstrate the damage potential of equal field /renetration cable sining.

The inboard seal was lost during this test, however, vii not r.: i 1 rod onle havo naintained containment integrity. ou t i eat.i r a; t

I h h e Q ~ l \\ .p D1 ij \\ ^ M B2 TYPICAL MUl,TI-\\ '.{) m i CONDUCTOR JACKETED H p POWER SUPPLY A1 A A3 A4 B1 C1 e i 7 BUS d FIELD CABLE, LOW-- VOLTAGE POWER AND CONTROL APPROX ' D El E2 E3 IMATELY 5' IN C2 LENGTH "q j {l / B1 ) B2 \\ FUSIBLE LINK !) USED IN TESTS o o C1 Cl AND 81 0NLY l1 o g .,i._, ' yl 8 G1 y d) I r ~ 1 d ,, S PENETRATION T 10 im ASSEMBLY El E2 E3 HEADER E >[ Sj M 0 c PLATE 0 0 r, U m 4 , l'1 O. O TYPICAL PENETRATION-1 n c> !i E R Sg fi g 9, }c O O / ASSEMBLY CONDUCTOR yZ o .] -t hh 5 G1 m a PRESSURE NOTE: o;- "m f GAGE Qi f' -t FUSIBLE LINK POINTS WHERE CIRCUIT. g[ I l "6 USED IN TESTS TYPICAL OF (8) FAILED DURING TESTING ARE me. i d z A3 AND A4 ONLY PENETRATION MODULE .NOTED BY CIRCLED TEST g-FEED THROUGH NUMBER, WITH REFERENCE y Jj ASSEMBLIES APPROX-TO TABLE Q40.106-2. IMATELY 2' IN LENGTH {. f,- i

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Chemical Engineering Branch' Based on discussions between Commonwealth Edison (C. Reed) and the NRC Staff (V. Benaraya) on February 19, 1981, the following summary of qualifications of the LaSalle County Control Room carpeting is provided: Manufacturer: Bigelow - Sanford Incorporated Type: Bigelow Growgrain 2, LF200r-F Qualification Tests 1.(a) 1973 UL Tested UL Fire Propogation Index of 0.83 =.10 inches 12 min. derived from the 12 gas jet test UL used. Result: carpet self extinguished upon removal of flame jets..Index (0.83) is excellent indicator of non-propogation characteristics. 1.(o) Bigelow Tests Flame propagation via Steiner Tunnel Test (ASTM E84) 79(A) and by Radient Panel Test (NFPA 253, ASTM E648) which tests ability for self-extinguishing. Result: Does not ignite in Tunnel Test. Self-extinguishes in Radient Panel Test. 2. ignition Test: Pill Test (Dept. of Commerce FF1-70) which is the ignition of explosive charge in contact with carpet. Result: Non Flammable 3. National Bureau of Standards Smoke Test (NFPA-258) Optical Standard of 450 or less acceptable Result: While flamming - 239 non-flamming - 255 l

1 l Containment Systems Branch In respanse to an NRC Staff request for clarification of the Table 6.2-21 octes 29 and 39 related to special leakage tests performed on certain ECCS suction lines to satisfy the requirements of 10 CFR 50 Appendix J, the attached proposed revision to those notes are provided. These revisions will be incorporated into a 'uture amendment to the FSAR. Specifically, the revision establishes a cumulative limit of 1 gpm per valve for system leakage tests; i.e. for systems having 30 valves, total system leakage should nct exceed 30 gpm. .=.-....

( P R.s P o s G D ) I ** AMENDMENT JW LSCS-FSAR JANUARY 1981 TABLE 6.2-21 (Cont'd) I 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 continuously monitored by the pressure switches in the pump discharge lines, which have a low alarm setpoint in the main control room. Even though a special leakage 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 pressurized with water to a minimun pressure of 32.5 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 b sed on s te bou dary b sds-A M fpmXN. M(10CFR100)*vk -A ,f dose oasiderations 30. Tie leakages throu. the Main Steamline valves will not be included in establishing the acceptance limits for the com-bined leakage in accordance with the 10 CFR 50, Appendix J, Type F,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 parameters. For example, the ECCS pump minimum flow valves close 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 operation signals that control these valves to ensure their proper position for safe shutdown. Reference to the logic diagrams for these valves indicates which other signals close these valves. 32. To satisfy the requirements of General Design Criterion 56 and to perform their function, these instrument lines have been designed to meet the requirements of Regulatory Guide 1.11 (Saf ety Guide 11). These lines are Seismic Category I and terminate in instru-ments that are Seismic Category I. They are provided with manual isolation valves and excess flow check valves. 6.2-93k

Quality Assurance Branch At a' meeting between Commonwealth Edison and the NRC Staff on February 19, 1981, the LaSalle County response to Q421.6 and 421.7.were discussed. Commonwealth Edison agreed to revise Table 3.2-1 to-include the following additional items: 1. Item XXXVIII - add reference to HVAC filter 2. Item V.9 - add data and modules term 3. Item XLV - add containment monitoring system 4. Item XLII - add PC and vacuum breaker valves. Indicate sample line included in Section III(Recire System). In addition, for those items not included in the table based on previous agreements with the Staff on the safety-classification of certain systems, structures and components supplemental documentation will be-submitted to allow the QA Branch to verify these agreements. This documentation can be provided by March 6, 1981. Internal discussion between the QA Branch and the other. Technical branches could close-tnis issue more quickly. i i _._ _ _..-._._.~.- _....-...--......-~ _.-_- _.---.-,-,.., -.

l Financial Qualification Branch (a) Operating Cost Estimates - LaSalle 1 (1985, 1986) Attached are the subject estimates requested by the NRC Staff on February 13, 1981. This information was telecopied to Mr. A. i Bournia on February 19, 1981. (b) LaSalle County Unit 1 - Commercial Service Date The LaSalle County Unit 1 comme'rcial service date is scheduled for April 1, 1982. This information was communicated to Mr. A. Bournia on February 17, 1981. 07378 b I I _.._m._.,_.

4 ATTACHMENT FOR ITEM NO. 1.a. ESTIMATED ANNU L COST OF 0FI MTING NUCLEAR GENERATING UNIT: LaSalle._ ;it_tTo. 1 FOR THE CAiJ. .: U.AR 1986 (thousands of dollars) Operation and maintenance extenses Nuclear power ceneration Nuclear fuel expense (plant factor 65 %).......... $ 61.159.000_ 556.000 Other operating expenses.. 861.000 Maintenance expenses.................... 62.556.000 Total nuclear power generation........... 92.000 Transmission expenses......... Administrative and ceneral extenses 1.862.000 Property and liability insurance. 5.ha?.000 ~ Other A.&G. expenses. 7.28h.000 Total A.&G. expenses. 69.932.000 TOTAL O&M EXPENSES 20.561.000 Depreciation expense...................... Taxes other than_ income taxes 4.916.000 Property taxes..................,..... Other.. 4.916.000 . Total taxes other than income taxes. 21.829.000 Income taxes - Federal..................... 1.977.000 Income taxes - other...................... 5.960.000 Deferred income taxes - net................... -(7.482.000) Investment tax credit adjustments - net............. 64.132.000 1 Return (rate of return: 12.24%)................ $181.825.000 TOTAL ANNUAL COST OF OPEF.ATION l

4 ATTACIEGNT FCK ITEM NO. 1.a. ESTIMATED ANNUIL COST OF OPERATING NUCLEAR GENERATING UNIT:_LaSalle Unit No. 1 FOR THE C h iDAR YEAR _1 m (thousands of dollars) Operation and maintenance exnenses Nuclear nower generation Nuclear fuel expense (plant factor 65 %).......... $ 58.370.000-510.000 Other operating expenses. 786.000-Maintenance expenses.................... 59.684.000 Total nuclear power generation........... 86.000 Transmission expenses..................... Administrative and ceneral exnenses 1.842.000 Property and liability insurance. 5.193.000 Other A.&G. expenses.................... 7 035.000 2 Total A.&G. expenses....... 66.805.000 TOTAL O&M EXPENSES................. 20.561.000 Depreciation expense., Taxes other than income taxes 4.772.000 Property taxes........................ Other. 4.772.000 .. Total taxes other than income taxes.... 21.602.000 Income taxes - Federal..................... 1.957.000 Income taxes - other...................... 8,360,000 Deferred income taxes - net. -(7,609,000) Investment tax credit adjustments - net............. 67.464.000_ Return (rate of return: 12.25%)................ $183,912,000 TOTAL ANNUAL COST OF OPERATION .,,-n-, -w,n, ---,}}

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