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Forwards Final Draft Responses to Commission 821005 Info Request Re GESSAR-II.Instrumentation & Control Sys Branch Responses Will Be Provided Following Ge/Nrc Meeting in Jan 1983
	ML20071P889
Person / Time
Site:	05000447
Issue date:	12/22/1982
From:	Sherwood G; GENERAL ELECTRIC CO.
To:	Eisenhut D; Office of Nuclear Reactor Regulation
References
	MFN-196-82, NUDOCS 8212290119
	Download: ML20071P889 (169)
	v • d • e

{{#Wiki_filter:GENER AL h ELECTRIC NUCLEAR POWER SYSTEMS DIVISION GENERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE, CALIFORNIA 95125 MFN 196-82 MC 682, (408) 925-5040 December 22, 1982 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 Attention : Mr. D.G. Eisenhut, Director Division of Licensing Gentlement:

SUBJECT:

IN THE MATTER OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT DOCKET N0. STN 50-447 FINAL DRAFT RESPONSES TO COMMISSION'S OCTOBER 5,1982 INFORMATION REQUEST Attached please find draft responses to the Commission's October 5,1982 information request on GESSAR II. Attachment No.1 summarizes the status of these responses. Except for the Instrumentation and Control Systems Branch responses, this transmittal completes the subject request. As indicated on Attachment No. 1 the Instrumentation and Control Systems Branch responses will be provided within two weeks following the GE/NRC-ICSB meeting to be held in mid-January 1983. An amendment is scheduled for February 1983 to formalize the responses. Sincerely, Glenn G. Sherwood, Manager Nuclear Safety & Licensing Operation GGS :td Attachments cc: F.J. Miraglia (w/o attachments) C.0. Thomas (w/o attachments) D.C. Scaletti L.S. Gifford (w/o attachments) 8212290119 821222 ) PDR ADOCK 05000447 A PDR Y )

ATTACHMENT NO.1 Status of Final Draft Responses to Commission's October 5,1982 Request Branch Final Draft Response Instrumentation and Control Systems Submittal within 2 weeks following mid-January 1983 GE/NRC-ICSB meeting Containment Systems Attachment Number 2 Quality Assurance Attachment Number 3 l l l

ATTACHMENT NO. 2 FINAL DRAFT RESPONSES TO CONTAINMENT SYSTEMS BRANCH QUESTIONS l l l

~ 480,01 i Appendix 1 to Section 522.1.1.C of the Standard Review plea (SAP) cortt.atus (5,2.1)! guidance on leakage tests and presents our position on surveillance re-quirements for Mark II cantainments. Some of these gu1A: lines are a priate for. and are being applied to. Mark III containments as well,ppro-specifically. Section B.f of Appendix ! to Section 4.2.1.1.C contains our ( position regarding the need for high and low-pressure leak tests including our position on the acceptance criteria for both types of tests (i.e., the measured leakage shall be 'Isss than 10 percent of the bypass leakage capacity). In addit. ion, surveillance requirements at each refueling cutage and vacuum relief valve position indications and alarus, and monthly operability to ts are discussed. State what provisions wit) be made in your proposed design to conform with our positions in the SRP cited above. Raw 3e i 9 s I ,q.g h. f. 4 2fle o 1 I ?%.r_rs4.cLLC,.2.Jo.gr,i_IsiJ4'5&re._ L4.3 c.o.,atain i Lu-a, ame c ommrea.e.w ctrs,- i 8 M 'p!:s-m aicego.avs.) so.n+ T*w.esr_e39 U* to A.b_. I I I_! l ( ll_iArse._ FoST-eftr.ed.lRf 84! i ! 4 ! I 'I _.f.% e_ n @ _p w.g va.gi\\i % ! !II i i i i lI l'f i !i ' ! i hw. -s e.s 16.7.. t. t.5144 rt.sr.eenew l trectyjttu i Lh_e Tv.sr ( i mwtsArio+ on I Aww^%C 3rdLYeGJ_._._ _ ' MLA i c.4c.6..x2Mr M il o _J.__TEW,.Mur.AmTM.WE ' t-E Amusi_ mmc 5 1 _v4.wA._*_-w__W.W._w rN.T- / I 5IG.vr.avP m_?_neo E /J_ L ! c eet_iG: Ptrovi6 eA5.( p ov s.swu W ' f rTtu p,i m or to e i i i w = m. w H ua.v h icA. I j.p_.',hggj..D1g.f Qf.t.b I i i 'I l f iili i : ! t / i i i %3s, +! nc45 L % 1. W l "4 (;.4 Jy c.u.M 5, W : 1s Wuzo;wmde1 3 VIrd R G E p E w W <a ! 11 w - m a.t L Is r % !-r a n j % v w q.t-t wr m. i I i i o w w m vne <a.s w v 4 e i_ e to - ii N-i : e I i W-,_ w. i w <rre,ic. n.e x s A c. w t % c.c __. 1i!!! !l I I l 1 i MWTt/g I'F t.'mDJtM tR' 39ya.nC.De4 6.0.l.(od_.3*L IV' ' i-i t i i f: i P%E.ALJo p!, asst.ne y f. 7 1J.4.Li_.Dr5c.mW~=i__ i_ i, ! i ms - -. n_ tw ~ , NMM Ghl@Nd ' i /e fryLWM&%Ee**.ct.J61. m 161EA_'3F elT' -- W N i wv wp s v a v tc.st. a c. . it - are. h.Ap.e t'or,TjiP V rerv c_titw.BruifT ' I LN i i i : 17 i 19WW_13Nk*tLh1 M24W I i T-tE' a&.co45_'-Wi> ; 'tmc g19 91 e_.r33 APfE.C.Le i\\_ t i i I C-m i sa ri w t i nur_._2xwi re e_nsmD._< k d i ; i I ij ! I I I

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,_,__~_uu-------- 238 NUCLEAR ISIJUID Rev. 0, 6.2.1.6.1.3.2 High-Pressure Leak-R1tte Test Q (. Isumediately following the high-pressure structural proof test, the drywell pressurization source is shut off and the change in dry-j well pressure and temperature is monitored for the next 30 minutes. The drywell pressure and temperature decay information is used to establish that the drywell leak rate is less than the allowable value. g. The drywell air-flow rate from the 1-hr structural test L holding period is used as a gross check on the drywell leak rate. Figure 6.2-37 shows the expected pressure decay rate for the drywell from the 30-psig starting point, the possible effect of temperature, and the calculated allowable and technical specifi-l c.ation lim 3ts. The figure demonstrates that adequate accuracy in I the drywell leak rate can be obtained by a 30-min

test.

The acceptance criterion for the high-pressure leak-rate test is demonstration that the drywell has a bypass A/ W of less than lot of the A/VE value for bypass capability under DBA conditions (i. e., 4 less than 10% of 4.3 ft2 ~ or d.43 ft ). TI4ts t% A. CMC.4 -oM T~sst l *$ 2 6.2.1.6.1.4 Post-Construction Drywell Test The drywell is subjected to periodic low proar de integrated leak rate tests to confirm continuing adequate leax tightness. The Q , frequency of these tests will be identified in the technical speci-fications. Th'e differential pressure selected for the periodic tests is sufficient to simulate controlling SBE conditions, but slightly less than the differential pressure required to clear the top, row of horizontal vents. That is, the head of suppression pool water above the top row of horizontal vents, under test conditions, is sufficient to seal the vents without having to install temporary Q closures. ) 6.2-70

238 NUCLEAR ISLAND Rev. 0 ( 6.2.1.6.1.4 Post-Construction Drywell Test (Continued) () The low pressure integrated leak rate tests are conducted as described below net N r M E :_W "!.TS N rN P " " l '$ ~~ (1) Reactor pressure vessel is closed. (2) The drywell head is installed and the upper containment pool filled to its normal operating level. s,- (3) The suppression pool is at normal level. (4) The drywell equipment hatch and personnel lock are closed. (5) Containment is vented to atmosphere. (j (6) High drywell pressure is simulated to achieve isolation of lines penetrating the drywell. (7) The compressors of the Hydrogen Mixing System are started and the drywell pressure is increased to approximately 3 psig. Figure 6.2-38 shows the expected rate of pressure rise in the drywell with both com-pressors operating. () (8) The compressors of the Hydrogen Mixing System are turned off, and the drywell temperature and pressure decay { rate in drywell and suppression pool water level are l monitored for 30 min. (9) Using the data from Step 8, it is determined that the (]) actual leak rate is less than the allowable leak rate. C) 6.2-71

b. e..t. / Inctrumentctica Requiremento (Continu:d) I N containment. Similar transmitters, which sense containment-to- ,.j shield-annulus differential pressure, are initiating, inputs to 4 the Containment Vacuum Relief System. W W u % tv E M E 8' YMS y Ap.st %wtc.w.AL9 1se tt. o?'sGe emTY AY te.AF *h Wot4M pace and containment RWCU room temperatures are inputs to the Leak Detection System. Four thermocouples are mounted at appror-4 ate elevations of the drywell space, and 12 thermo- _) couples monitor drywell HVAC differential temperatures. Six-teen thermocouples are mounted in the containment RWCU rooms. Four suppression pool-level sensors are immersed in the sup-l pression pool water, and the assocaited level trar}sducers are mounted above the water level. The level signals are transmitted to.wmu dystem icgic in the control room. Eighteen thermo-couples are immersed in the suppression poo1 wate'r. Suppression i pool temperature readouts and alarms are located in the CM.LIG1 foom. Two hyorogen analyzers are mounted in the drywell, and two z.: '..l in the containment. Each analyzer draws a sample from an appropriate area of the drywell or containment. ] Hydrogen concentration alarms and recorders are located in the control room.

u. ;. a a Lectors are mounted in the containment ventilation exhaust ducts.

Radiation monitors and containment isolation trip .s located in the control room. section 7.2 for a description of drywell pressure as an .% c; w input to t.he Reactor Protection System, and Section 7.3 for a de.scription of containment and drywell pressure, containment-to-. i 6.2-74 i l

i i 480.02 / You do not indicate la your FSAR whether you conducted node sensitivity studies in the subcompartment onelyses to determine if en adequate number (6.2.1)f'gofmodeswerechosen. Discuss the extent to which node sensitivity studies were performed. l l N5 PENSE l A wscn*w srv% ws. w+mee us'*

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480,03 Indicate how tosponents of the reactor butiding i. eating, ventilation and (6,2.1) air-conditioning (HVAC) system, when operating in the drywell cooling mode, are designed to prevent damage to engineered safety feature (ESF) equipment or the reactor coolant pressure boundary during a safe shutdown earthquake ($$E). Indir. ate whether this portion of the reactor building HVAC system is classified as an ESF system or is designed to seismic Category I criteria. If not, provide justification for your proposed design. kESPOVSC-- t_..L_L.__ ___.-._.L ._J _L. N ' _.t__[ L / A . d I _., '. _&.. W=. e.1,._ _ g _..'. ..f. n.cm b._. h_ -rh i 0.t; mS. ei 6._ c E S f p g..__- %.._..l u u .J.. l 1 .r A aL_., e-wr _a. _. l m.. a w r <m a m-. O,I.L. e,\\~ 9_. .e.~. n u l l l A. yx w at <-i=L,w DxtlT' I q__ I { y.pAf hsm s 1 ,V' h' f A MWhd U[ __i i M l'*h r- - l +- . l 1 1 A. -c. - %. W M M.E U1~.l e o -y 3 (& ti. lk p sm_--ak NA-_y & y _'_t: 1

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480.04 You state in Section 6. 2.1.1. 5.4 of your FSAR that the limiting case for (6.2.1)' steam bypass capacity is a very small reactor system break which will not automatically result in reactor depressurization. It is possible, however, that an intermediate break accident will actually produce the most signifi-cant drywell' to containment leakage prior tJ initiation of the containment sprays. Cospare the leakage rates which will result from an intermediate size break with those from the small break accident and demonstrate that the intermediate break is not more severe than the small break. DG TH\\s RE5WG wtu. (G /Achm ab PA M OF M k) M 5e ~1D Tac t ETilER FRon; JARREC.L-6, EtsEnHU7 70 d(YW

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480.05 Generic Technical Activity A-39, " Determination of Safety Relief Valve (SRV) (6.2.i) Pool Dynamic Loads and Temperature Limits for SWR Containment" was estab-iished to resolve, among other things, the concern about steam condensation ( behavior for Mark III containments. NUREG-0783, " Suppression Pool Temperature Limits for BWR Containments," presents the resolution of this issue by indicating: (1) staff-approved acceptance criteria related to suppression pool temperature limits; (2) events for which suppression pool temperature response analyses are required; (3) assumptions to be used in the analysis; "4 (4) requirements for the suppressi,on pool '.esperature monitoring system. '" acceptance criteria, for instance, require that the suppression pool .perature should not exceed 200"F for all plant transients involving operation of the safety / relief valves (SRV's) during which the steam flux ( l thr ough the quencher perforations exceeds 94 lbs/ft sec. The temperature limits for other values of steam flux are given in Section 5.1 of this NUREG report. Provide analyses which show the extent to which your proposed design meets the provisions of Section 5.1 through 5.7 of NUREG-0783, including the acceptance criteria, the events r9 quired to be analyzed, your assumptions l and the requirements for the suppression pool temperature annitoring g system. i R F.e po AJ S E j $ack~3 rou.m[ Prior f0 ke N.ssuay~ce cs> S ~ ~ ' ~ B.02 56-k793, ' Mayk W-Sha uda yc 23T t 4 N -f Ya bd etc a ad Lock a na (ysez-i te)(dc k a re Slw UcQ-pr~%rw.. -re. ec Je tEived Af A kn % ?:-0 724 Y-he. a Eta }s-s ~ _<g s s um lH ///$ "d'783. i i

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2 Sub 4RY OF INITIAL CONDITIONS, EVENT SEQUENCES AND Asst N TIONS DEPRESSURIZATION STUCK-OPEN SRV FROM ISOLATION _S ELL OREAK ACCIDENT PARADETEk5 EVtni 1(a) EVENT 1(b) EvtNT 2(a) EVtni Z(t) EVENT 3(a) EVtnT 3(b) Ouring Power During Hot > IDU t/hr luu t/hr Accident Norsel 1. INITIAL CON 01TIONS 5tandby Mode ' Mode 1.01 Reactor Power (% Nated): 1021 r = 1.02 Service Water Temp. (*F) Max. Plant Oata a too 2 = 1.03 Initial Pool Temp, T (*F) Max. Tech. Spec. =tes = 1.04 Initial Pool Volume (cu. ft) Min. Tech. Spec.*s2.'F## r 1.05 Orywell remfseAfus,p

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TABLE T (continued) OEPRESSURIZATICII STUCX-0 PEN SRV FROM ISOLATION _SM4LL $REAK ACCIODI PARAMETERS EVENTt. 1(a) [ VENT 1(b] EVENT 2(a) EVENT Z(b) EVENT 3 M [ VENT 3(b) 30 ASSUMPTIONS (Cont'd) 3.04 # Pr.s Available Yes Yes Yes Yes YFS.t. .Yes 3.05 RCIC Available Yes Yes Yes Yes 'ys.r Yes 3.06 Condensate Storage Tank Avail. Yes Yes Yes Yes Ns' Yes 3.07 Dryw11 Fan Coolers Available Yes Yes Yes Yes No Yes 3.08 RHR Heat Exchanger Duty Based on Maximum Observed Equilibrium Crud Buildup = 3.09 Number of RHR Loops Avail. 1 1 2 1 P. 1 2 3.10 SRV Capacities (% of ASME 122.5% 122.M 122.N 122.M 122.E 122.5. Rated) 3.11 Decay Heat Curve Decay Heat Curves For tentainmen belysis -4N5 W)e NOTES: II/A = Not Applicable + 4Pcs avastelele eyeept & swIl bmh is $e A \\cs Use e f 'M

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^ ' < ' ' '" 480.06 , Show, by sto results of an analysts, that the loss of the containment unti. ~(6.2.1) lation systes during normal operations will not causa the hsign opsratIng ~ tenditions to he Caceeded for safety-related eq';fpment inside the :entain- ~ If this criterion is not satisfied. it is our position that this ment. ventilation system should boiconsidered safety 3rade. ya sto N s s W -4 l .._ h. 0 N Yh' - -- -_-/ Y .4-. l a t p..,.._,_,, l ... J....!

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ww - to LQi3ED U.S.D.1.4 of your F5AR you indicato that 12 cvaltating the ) (5J.1) capability of your proposed containme.t 13 cithstand negative presture, a . small primary system break. is one of the transients which could result la l negative pressures withia the containment. It is our position tha't this small line break should consist of the largest unguarded pipe passing through the containment. * ".3dicate which pipe system line was modeled and the size of the upguarded, 'pe. pressure in the containment following inadvertent actuation of boih co tainment spray trains. Et,. E S Po N S E '5ee. Stesease. To quemew 4so.o4 1 4 e

1 ),, __.._.....v...-- repressurizo the containment following a post 31sted 1:ss-of-coolant accident (LOCA). You have proposed la NED0-21424 a represtrization limit of 90 percent cf the design pressure for the 30-day period following a LOCA. It is our position that the limit on the containeert repressurization for the 30-day period following a LOCA be 50 percent of the design pressure. Our basis is that this limit on the repressurization provides adequate margin and reasonable assurence that no problems will occur after that 30-day period. This position is consistent with our requirements for the contain-aent atmosphere dilution (CAD) systems, which also have a repressurization limit of 50 percent of the design pressure. Indicate whether you will adopt the lower repressurization limit of 50 percent. R ESPo d S E the nitrogen CAD system can be used to dilute the containment atmosphere following a LOCA in order to preclude a hydrogen combustion event. Because such dilution will pressurize the containment, it is essential that the addition of nitrogen be terminated at a pressure which wv11 not jeopardize containment integrity. NRC regulations now in effect (10CFR50.44) specify that containment repressurization be limited to 50 percent of the containment design pressure. However, therg is no apparent reason from a containment structural integrity point of view as to why repressurization should be limited to 50% of containment design pressure. The containment structure is a large, unfired pressure vessel. It is designed to withstand pressures which are at the design pressure for an indefinite period of time, or even cycled a few times between low and high (design) pressure. In this regard, it should be noted that the craf t interim hydrogen rule for Mark III plants allows for pressurization of that containment under non-accident cenditions to ASME Service Level A limits, which are signi-ficantly above this 50% design pressure limit.15 In addition, due to the large inherent containment design margin, the containment could withstand a pressure of 2 to 2.5 times design pressure and still retain its inte-grity. This conclusion was reached in the WASH-1400 Reactor Safety 1 Study 2 and the Limerick Probabilistic Risk Assessment (PRA).13 j);gF0q MAKE CVnS! DEME pundy 60G th70 h6 OW4!^'W I / gzgyg3 yggy pygTHff' JYl/wir/ d MT Mf

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In Tablo 6.2-10 cf your FSAR, yo*a list a relativ2 humidity of 10 parc2nt i [ 480.09 (6.2.1) fcr cn ir,itial condition la tha subcospartment analysis of th2 biological shield annulus. You indicate that the design margin for this case is 5.7 Provide the basis for choosing this value instead of the more percent. conservative value of zero percent relative humidity, which is endorsed in Item II.B.1 of Section 6.2.1.2 of the SRP. 51milarly, provide your basis for selecting a reJative humidity of 20 percent as one of the initial conditions for the subcoupartment analysis of the drywell head region. Res p on se. i 1-G e dei. &, s (L u'e(clo undx o O MQ Vet C f.,9N C b2Y D$ a u ?uh4al refa li w> /Anw Nifv' L/ / n.S oue ,e e ye- -_ u a /, ~, 4 m e-Y$u, A hu LM t' /A)<CL 5 > *-- Q J C,5% try Ye L 2-tM a /- --/p w/1 rl m ' a-e &n aAA % o f u 3 42 '7KfS r f f he hf N A ( YW'J T f/T"Lt.A -T !L ~ r 7L C.7 yow d s1aryrm afke-iu << e ssa -vt +aMe e7-+ h lLO A r% ')C6 Si*.JfS sh_/hk-- l5 C h9Y Y Pe lH A^ ~ fey e ~~iO k, va %AM Lusm 7 dc'/v_,_111.c- /, /n m eer d n k Jrve /xaus2_cf/ -fV e /. ~ a $)A l m O-O QL Y ? [k ?

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480.10 Disc:ss tith the aid tf dra51Egs, the containment spray capability [Th] (6.2.1) fc11owing additional inf rmation should be provided a. The locations of the spray headers relative to the intcrnal struct res. b, The arrangement of the spray nozzles on the spray headers and the expected spray pattern. ~ c. The type of spray nozzles used and the nozzle atomizing capability. d. The effect of drop residence time and drop size on heat removal effectiveness. _o c. E. S t* b nl C L ^ (c0 Tho positions of the containment spray headers are shown on drawings in GESSAR. Figure 1.2-9 and Figure 1.2-10, Reactor, Auxiliary and Fuel Building Arrangement Drawings. (b The flowrate of one containment spray loop is 5,250 gpm and is assumed

. Le initiated no sooner than 10 minutes after the accident (this analysis assumes that the sprays are activated by 13 minutes). The suppression pool water passes through the RER heat exchanger and is injected into the upper containment region. The spray will rapidly condense the stratified steam and would therefore create a homogeneous air-steam raixture in the containment.

[(._) 6PEC.LFlC SFO PrV NOMCL1 N O C p(> % M c'5 cJNL Ee PRowCEo By The Ao9tttesuT siwace. 'bsey A<te vewovfl DeeCevU', SCC UOvw (d) ?x2100 Fort Genera < Maid, CA The efficiency of.the sprays are based upon the local steam to air ratio as defined in Report BN-TOP 3, Schwartztrauber and Pervich, Performance and #,izing of Dry Pressure Containments, Dec.1972.

(eg.t30.10 'cMT) The following an.dysis provides an illustration of the methods used to calculate steam condensing capability under typical post LOCA conditions. For ~the conditions defined below the containment spray can condense about 38 lb/see of steam when the containment is at its design limits. The condensation capability was calculated using equation (1): U m, s - c c s a = p fg where: i s, = steam condensation rate 4, = spray flow rate = 73d lb/sec (5,250 spa = 1 RHR Pump) 9, = spray efficiency T,= containment temperature T, = spray temperature at the inlet h = latent heat gg The spray water temperature was calculated from equation (2) - _EllK ( p ~ p g sw} T, = T Cp s where: Tp = suppression pool temperature III = heat exchange capacity = 512 Beu/sec *F (84% of rat d) T,, = service water temperature = 100*F. e

i--- "revide a list cf all loads tsed in the design tf the RHR it.tako strainers. h.;.Z.2) Provide II.formattori uhtch demonstrates the capability of tha strainers to accosmodate the lydre@ nemic loads from the horizontal vents in the dry-mell wall. g c s e o ni s & + =, :-m _ ?_ .e A. + - A ' ofu. - > r ThR ND - ~_b.. J e - x. L : l' L e. m i =0 w~ A5mt cor__- 7. 7~1 - L.: ? .e 0 A-A =__ - s= f L:%_K <=_ ASmt w _k..e-+ L u &/e b

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/ *g 1 6.3.2.2.4 ECCS Pump Suppresi*n P nl Scrnen,. All ECCS pump 3 cucticn lina cod sma' the RCIC*+ pumps suction line will (* provided with strainers. Tlie strainer ( assemblies will consist of 200 perecent capacity corrosion resistant strainers. The following pumps may on occasion take suction from the suppression pool, and all will have the forementioned type of strainer. a. HPCS Pump b. RHR A, B, C Pump ] c. LPCS Pump d. RCIC Pump (non-ECCS) (::: sec o e.y=:::.cces ). g: M re,, t,: fa ap a ra H ng

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2::.u.6.ueu anu u... fr :ti: ?'.". '. (Vu mu u ;. "--' r rf ni'd. wo. T.,1 t). The HPCS and RC O pumps preferred source of water is from the condensate storage tank. In the event this is exhausted, the pumps will switch to the 0 suppression pool. Each ECCS pump will be supplied from its own suction line ar usnaossa n.c.se..rm c<s 1 ~o .2 and strainer assembly. The strainers v111 be designed to ASME Se6ser Class 2 3 F in accordance with ASME Boiler and Pressure Vessel Code Section III. The suppression pool strainer will be designed to operate under the following k l, conditions: c i Design Pressure 12 psig external Design Temperature 212*F A d 55MRtiippuusee 'The strainers shall not exceed the allowable 3 pressure drops with 50 percent of the strainer openings plugged at the conditions listed below. .:r seem mucs /Lu tL AC8C L/C5 EfCU Gka,A,B,C 3 5 (d R 3 NOZZLE SIZE 24"sch10S 24"sch10S 8"sch10S 24"sch10S 12"sch10s FLOW, GPM 7800 8520 725 7800 2200 MA)CIMUM f ALLOWABLE l PRESSURE,g DROP, PSI (+) 0.8 0.9 0.5 0.8 o */ INLET PRESSURE PSIG 6.1 6.1 6.1 6.1 2 Go* sec o sys% c c e M <A.y m

The RCICg "{* pump:suc-ion screens will be designed to the game cr 1

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zett a.. :..x.t. (ea nano) i The strainers will be located so that their center lines will 'oe at an c2evation which assures minimum submergence allowing the strainer to function cs designed throughout the entire period of emergency cooling. The strainer will also be located far enough above the suppression pool floor to prevent snv necumulated debris from being picked up. The strainer configuration b i e precludes the possibility of one large piece of debris blocking the entire ( scttlon. Adequate separation will be maintained between mach strainer to f assure tha; pump suction will perform as designed when the adjacent hc pumps are also taking suction from the suppression pool. The strainers will (w be designed to provide low approach velocity to prevent plugging. Initial tests will be conducted and base data colle:ted. During period'ic testing these variables will again be measured and compared to the base date. Trom the comparison, system conditions can be determined including the conditior .:.rainers. Provisions will be made for visual inspection of the s u.u r.c r s. f The strainer mesh openings shall be the maximum size possible without allowing the passage of any object with a diameter greater than 3/32-inch (spherical). The strainer desion shall also block the passage of any object which e.3ceeds 3/32-inch in any_one dimension (sliver)j' G TMs !s somctevi P FWorect Tae. pump cvcLwe seenwos aa.,,. mur sem aaeue,ea s sPw w L5 4 Foet-SPAc.w G., l d

480.12 Describe the operating characteristics and efficiencics of th] " fully (6.2.2) fouled" RHR, heat enchangers. ft. C S P 6 N.S & l w % % :l e ~ M 9 &r:L MwMwa &p w ' 6:us s M O. e., a/ bey & &p, dd' fY>W j yM nmaM&q W M 4 % 7#v%~-wh t~- P A A M-M;ed+ ~ y n M f p '. ~ s & ann: ,X k! & fJ'y Q M p f4 ek m h 4<@ A< 4 e* g D w; Sv7M e q 2co L

,Mocu.i3,6 j Describe with the aid of detailed drawtss, the principal design features g,;.t}t of the protective screen assemblies at wetwell suction points when the RHR / system is gerating in the containment cooling mode. Indicate the mesh sizing of the strainers and describe your analyses to determine the final screen sizes. Compare openings and spacings (your proposed strainer size with relevant flow path e.g., containment spray nozzles, the emergency core cooling system apargers and fuel channel spacings) to show that the strainer size you have chosen is, appropriate. k E. S (* b N S L n g.- - p - x + p =a l I _^ "^ ^ A' $^ " ^^ --w w v s h>_ k L y -___ _ -b2 e __ ^ ^ ^ -h i N M> bd. A4___.._ _._ _. / a d. e0 du ^ 'f f- - Y ^ nn_ f, j < J - - di c,. t. a.. t. g w e. - C ccS dn 5.m L g ru.. _ f5ge case 4Aa 4-so.n i / ~-

450.14 Discuss the heat transfer coefficients you used in the analys:s cf thm (6.2.3) secondary containment compartments. Provide the fallowizg additional inforestion: The heat transfer from the primary containment atmosphere to the a. primary structure. (Refer to tranch Technic.a1 Position C5B 6-1). b. The conductive heat transfer through the primary containment structure and the convective heat transfer to the secondary containment atmosphere. . The radiant heat transfer to the secondary containment. Discuss and list the heat loads which were included in your analysis of the secondary containment. Indicate how the heat sources will be accounted for in t time,he secondary containment tests to verify system capacities and drawdown ft.E.s P6 (d.S E- .N.Y kc2MS E h[o1/A D -- - f -.. ro4{ b3l% MtB4.y. 7$>. .. --....p.H ntec-y.y .. S 5 N1.c(h r l.?Cb.bdhki.5.7IhhY S0(e.Y ..A Vi..Y) c-(l.ttS..-.L3.d,E.Scaswd. T% ,e f. ... ~b2...- - ... te;5fcry7e o ... -...f..aes.V~2.cryl 48 O/b. f.CLA 32.e J.'.2d -. 2. YES 52 =: pesk. nox;, Me .............. _ f..... -. - z . s u w o u (a't1C M e - ----'L%e - -f JV f2Se'< _ -.2.y.-. 4E SSN.. E... -.ce-..c -...d 2, /< 7. .c. tEcm- .... Q SS t.t. tt^.e s. -.- - <n.E. -ice c' -

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480.15 You do not discuss in Section 6.2.333.1 of your FSAR, the details of th2 (6.2.3) secondary containment compartment pressurization analysis. Discuss this f~ onalysis, including the information listed in Items II.l.a through II.l.h in Section 6.2.3 of the SRP. v [t E S I b tV 3 L I -ssess_u<erf A Ha4. I is au i' af GesM# r to u,Cn-,a -w & Me_ua_ e K /~ + 4 ro u ; L E./ l< e-F SRP C.L3, Dew zua e f sk P &. 2. 3 grfa.-b:. s- +bd b ad -Fran.L prn +5e ar_;w n v A sec. Ja > a en>1$,wwef / / / sLli L e uEev i / 4.I 1*desa_'W~ Q W [ S_b3_ l~btg S.G. eon y tu cssun iz seamcw presen+mi bred e u . usa n - ekel,y a d /o.2. 3. E ine ve herN re lead fr~ au Arrh r y / Y $<4- (W k-W P22_ Y.Q W T-hygSev $w b VE wnfv fruhD1XfeY i

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SRP ASSE3SMENT SRP NO. I' 2 l H V PAGE.L C F.,3_. TIRE SFCI NDAkY t 1NTAlt U T Fl!Nf'TIOlMI. DFJ ICH [ EVI AT10*4 J'.STIFICATION R 1. Analyses of the pressure and temperature response of the secondary containment to a loss-of-coolant accident occurring in the primary containment should be based on the following guidelines: a. Ilent transfer from the primary to secondary containment shuuld be considered. (1) Ilest transfer from the primary CE does not use the coef-The design of the BINt pressure supprecolon containment atmosphere to the ficients given in Reference 4 containment assures that for des 1 n basis R primary containment structure for steam-air mixture. The IDCA there is no steam-air mixture in the should he calculated usinR coefficients used for heat containment. Heat transfer from primary a conservative heat transfer transfer eniculations in the containment to secondary containment areaa coefficients such as those RWR pressure suppression con-of Auxiliary Building (ECCS Rooms) and provided in Branch Technical tainment are based on Ref-Fuel Building is not considered in the Pnsition CSB 6-1 (Ref. 6). crencejf of CSB 6-1 for the initial transient analysis because of the portion of the steel con-Annulus and Shield ButIdtng separate these (2) Conductive heat transfer tainment above the suppres-areas from primary containment (see SAR through the primary contain-sion pool surface. Other Figures 6.2-43 through 6.2-50). The heat ment structure and convective coefficients are used fnr trcnsfer from the primary containment to heat transfer to the secondary water to steel and 100Z the secondary containment shield annulus cnntainment atmosphere should relative humidity air to volume is presented in Figure 6.5-5 and be ennsidered. steel on the containment side SAR Sections 6.5.1.3 and 6.5.1.3.1. i and steel to dry air on the (3) R.ediant heat transfer to the shield annulus side. Conduc-This portion of the SRP uses concepts and secondary cuntainment should tion, convection and radia-coefficients which apply to pWR contain-he considered, tion heat transfer coef-ments and not the BWR pressure suppression ficients are included in the containment. modeI.

SRP ASSESSMENT 2- - OF) pggg SRP NO. 611 REV SEC0t?DARY CONTAINHENT l'tJNCTIONA1. DES 1CN.(Contin sed) TITLE SRP ACCEPTANCE DNATim MF4 CATION CRITERIA ~ No cred'it is taken for heat transter vor b. Adishetic boundary conditions No Deviation secondarycontainmentexteriorstructurel should be assumed for the surface surfaces to the atmosphere. of the secondary enntainment struc-ture exposed to the outside environment. The reduction in naulus air space due to The compressive effect of primary No Deviation expansion of the free-standing steel con c. containment expansion nn the secon-tainment due to pressure within the prise dory containment atmosphere should containment has been included in calcula he cnnsidered. tions (see SAR Section 6.5.1.3.1). No Deviation The secondary containment inleakage is d '. Secondary containment intenhage designed and tested to be less than 1001 should be considered. of the free air volume per day (see SAR Section 6.5.1.3). No credit is taken for secondary contain No credit should be taken for No Deviation ment outleakage (see SAR Section 6.5.1.G e. secondary enntainment outicakage. l ,8 e e

~ - - ~ - - - - St.P ASSE:;SMENT erpeo, _ 6.2.3 IEV. - PAGE OF $. iLE SrCIWDA Y r"NT ' INHE', INCr10tlAT. ' ESTCH {Co itipued) 5PA E ^ DEVIATION JUSTIFICATION p,E I f. SecondarycontainNtresponseanal-Iso Devistinn Secondary contairusent response analysis is yees should be based on the assump-based on loss of offsite power and single tion of loss of offsite power and active failure (see SAR Figure 6.5-2 and the most severe single active fat-- Section 6.5.1.1). See SAR Section 6.5.1.3 lure in the emergency power system for assumptions used in secondary contain-(e.g., a diesel generator failure), isent response analysis. In tie primary containment heat removat systems, in the core cooling systems, or in the secondary con-tainment depressurization and fli-tration system. Any delay, due to system design, in actuating the secondary containment depressurfsa-l tion,and filtration system sInnstd be considered. l g. Ilest loads genernted within the 910 Deviation Equipment-generated hent loads are con-i necondary containment (e.g., equip-sidered in the secondary containment anal-ment locat fonds) should be ysis for ECCS pump rones and Fuct Railding considered. (see SAR Sections 6.5.1.3.2 and 6.5.1.3.3). h. Fan performance characteristics IIo Deviation Pan performence characteristics are een-should be considered in evaluating sidered in evaluating the depressurise-l the depressuris'ation of the secon-tion of the secondary containment (see I dary containment. SAR Section 6.5.1.3). l 5 I i e-

460.16 Provide the followizg additional information related to the potential (6.2.3) typass leakage paths given 12 Tabla 5,2-24 o For each air or water seal, demonstrate that a sufficient inirentory a. of fluid is available to maintain the seal for 30 days following a LOCA. (The suppression pool cannot be considered a water seal.) Describe the testing and the proposed Technical Specifications which will verify the assumptions used in your analyses. b. For each pat' h where water seals eliminate the potential for bypass leakage, provide a sketch to show the location of the water seal relative to the system isolation velves. tt. E S f o M S E- _ PAGs !! jqawstvied _ l { !l l l l l l l l l j !l ij'I i,: Afo-1 1 42o, ibid Au _IN Ac.rtvT-i MM.9ES_A.in: 59iE i.tym i TeL9.M.t - h &ttwin.__mWM E.1.I.Kr ED..._ __ _ dW 1 C _c J.hd dnsnW.Initg _PELHe >A 95t7.1 i M W H etagtjsa. w i V A.1.itL y1 t,r W N C L A.c,6 I AcT'h1M t/AbW_i_'AftL.LSff 4d?_id v_._Id..tL4_4.ti__ t 8 MAX t Mo M i bfE.irNw : f44Tt 5_LC35_. "2., c wl-) k l (* EthYI t 42.*r P G?.* k'C_(Z.JWC1.>l'Z " (*/.i!.1{ ___1 flC.Rf l-- I 46mitua. : yn LtLF % )t_Lig__At. L_.A T_.t43p:;c e I i STATIC..TCl*_f_fiM}1X-4_(IM.5.' V/s V@ I'Rf.'.*/*,U1.f.1._ l 12AT:N.9), t - i - ! i i i i i T4L Sl'hJ l_N4_.L 37Q. A.P.P W T/ d X._ S t' W 3L$.._, I i ' W ich e% ' 2 f 6999. y a o V.A:.NE.5,A1_ I Ic h y.,e u m.- L i t } I I l t ! I i i I I Sud '7 ) rac,jf I % gg.C i15:.rar g i i I8 I I i ! } ! t i ! i I . i t i i th a) %I AQ ! ~/ Q l 'l o t tiMqp_ _ _ i I i I j i '} ; j ,I l ! ! ! l 1 l I T d o! 4-MO4 I ! M.,$_n. Q ~1,. M c.,s.4.__ ___ __. i I i J ! ; j i ,'.j, I I! I $4 6-4Wia i i i j I II I i i ! l ! t 1 l' i i i i !. I i i j i t i l ! I i I i _.L ! ! I I I l i l_.] '! !I ii ! _] I _ _.I. i I

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!#rtify all openings for gaining access to the secondary containment. 400.17,, : 8 e (6. 7. ') ', Discuss the admiaistrative controls which will be exercised over them, Discuss the instrumentation uhtch will monitor the status of these openings. State whether positten indicators and alarms will have read-out and.alars capability.in the min sentrol room. gLsr6NSC ~ l l ALL C PEN 1 N GS FOR G alm i N G ACCESS TO TNG SECON D AR V CcN TAINMGMT Ant IDENTIFIED AND LIS TfD IN TA SLS (e.2-21 THE OP&NING.5 AND THEI A INSTRUMEN TATsoN wNseH HeMITO R THE LTATUS.0M THEst c!ENINes ARE 985605GE0 IN.siCTIONS C,2,3,2, 4, 2 l AND

6. 2. 5,'S

480,18 In Tabla 612-24 of your FSAR, you indicato that there are no patcntial (6,2.3) bypass paths. However, some of the you teko credit 12 this table are no@t sufficient for us to precludepa typass leakage. instance, the methods indicated by notes 6 on page 6.2-172 do not, by themselves, prec(2) C, (2) A (4), s paths. lude the existence p he Paths relyirg on these barriers alone should dared as potential bypnas paths.

19. light of our posf tion on could bypass the secondary containment and escapa entre environment.

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y.3'M A, ^ ~ m -r- ,.l._ .f. , f%, :_? 7 ~ 3 Table 6.2-24 ) EVAIfATI: J M POTEN7IAL ONTAINMElT BYPAS:1 LEAKAGE PATHS (Cont nued) x Line Size Penetrativg. Tet nr.. .lon Bypass Leakage Potential Prim try Cor tainem it. Penetration Containmerg Regi 1) BarrierI2I Bypass Path 4 O i' 54C. cLPCS Pump Discharge 12-in. S NA No 55C. LPCS Pump Test Line 12-in. S NA ^' .No - ~ 56C. LPCS SRV Discharge to Supprestilon Pool 2-in. S ~NA No f'

57C.

Air PositivewScal to Air System 3/4-in. S HA No j5EC. HPCS Pump Discharge 12-in. S (5), (6) C (5) No 59C. IIPCS Pump Suction 24-in. S (5), (6) (5), (6) No g60C. IIPCS SRV Discharge 12-in. S (5), (6) (5), (6) No g63C. RWCU Pump Suction From Recirc Pump S-in. S C No 64C. RWCU Return to Feedwater Line 6-in. S C No e 65C. (RWCU Discharge to Main Cor; denser 4-in. E C, (6), (3) No m 68C. Con tainment-Supply Purge (IIVAC) 42-in. E C, (3) No $69C. Con tainmen t-Exhaust. (HVAC) 42-in. E C, (3) No t j g 4 !,! 70C. Cont.ainment Vacuum Relief Outlet 24-in. S C No [ w 72C. Containment Vacuum Relief Outlet 24-in. S C No O 70C. Skimmer Drain to FPCC IO-in. E C, (3) No s< t.79C. Demineralizer to PPCC Pool 10-in. S L No r83C. 24-in. Pipe Spare 24-in. S NA No y 0,1C1 Instrument Line 3/4-in. S NA No e e 84C -34C4 Spares S NA No 2 ll4C. Drywell CRW Sump to CRW 3-in. E C, (3) No g .l115C. Drywell DRW Sump to DRW 3-in. E C, (3) No

ll6C, M4e. 12-in. Pipe Spares 12-in.

S NA No 10C. 24-in. Pipe Spare 24 :in. S NA No i119C. RWCU Backwash Drain 2-in. E C, (3), (4) No '120C. CCW To Containment 10-in. E C, (3) No 9121C. CCW Return from Containment 10-in. E C, (3) No I'124C. 12-in. Pipo Sparo 12-in. S NA No 125C. III Chilled Water to Containment 6-in. E C, f(3) No 126C. lH Chilled Water from Containment 6-in. E C, (3) No !-127C. Condensate Dist to Containment 6-in. E 120C1 3/4-in. Pressure Scnsing Line for ILRT 3/4-in. S C,$(3) No e 4 I 4 k;120C NA No 2' Upare S NA No o X VI C, 8 na.. Ptne Pge,x go,4 %wp_., IT-In E C. O)h) No 1 e ,w

?"%

r. ; _

m Table 6.2-24 EVALUATION OF POTENTIAL CONTAINMENT 'GYPASS LEAKAGE PATHS '-(Continued), Line Size ~ Penetrating Termination Bypass Leakage Fotential Primary Containment, Penetration Containment _ egion(l) Barrier (2) R 9ypass P g 129C. Service Air Distribution-i 4-in. E C, (3), (6) Nog 130C. Instrument Q r Distribution 3-in. E C, (3) No 131C. ADS Pneumatf.c Supply (Div 2) 1-in. S NA No-135C-136C 2-in. Spare 2-in. S NA No 137C. 24-in. Spare 24-in. S HA No 142C. Chilled Water to Drywell CLRS 6-19. E L, V No 143C. Chilled Water Return from Drywe)1 CLRS 6-in. E C, V, (3) No 145C1 ESW Line to H2 Mixing Blower System y / (Div 1) s 4-in. E (6); C, (6), V No en 145C2 ESW Line Return from H2 Nixing Blower e System (Div 1) z 3-in. E (6) C, (6), V No 'g 146C. 24-in. Pipe Spare 24-in. S NA-I 147C. 12-in. Pipe Spare 12-in. S NA ~ No\\ , @( g 148C. ADS Pneumatic Supply (Div 1) 1-in. S HA No 38 No ys P 156C. Spare x 157C. Spare }_ 158C. ESW Line'to H2 Nixing Blower System ~- , g j (Div 2) 3-in. E (6) . C, (6), V No t1 160C. Air to RCIC Turbine Exhaust Line 3-in. S C, (3) No 164C. RWrU Pump to Filter Damineralizer 6-in. S C No 165C. ESW Return from H2 Mixing Blower System (Div 2) 3-in. E (6) C, (6), V No 166C. Drywell Pressure Bicedoff Line 2-in. S C No 16Sc. Upper Containment Pool to !4ain condenser 12-in. E C, L No l 178C. Air Positive Seal to E51-F063 3/4-in. S NA No 319C. Suction to SPCU Pump 12-in. S C 75) - No 320C. SPCU Return to Suppression Pool 8-in. S (,gg) No A . 30 A e >. f2 C .0% i i

~ 238 NUCLEAR ISLAND rov. 0 Table 6.2-24 I EVALUATION OF POTENTIAL CONTAINMENT BYPASS LEAKAGE PATHS (Continued) NOTES: (1) Termination Region S = Secondary containment '3CCS Rooms or Fuel Building). Lines terminating vithin the secondary containment are not potential throughline leakage path. ? = '""ironmental, beyond secondary containment. Such lines [k either pass directly through the secondary containment to the environment, or are connected to branch lines which pass through the secondary containment to the environment. For either case, potential throughline leakage is pre-cluded by a combination of leakage barrier. (2) Bypass Leakage Barriers C = Redundtit Primary Containment Isolation Valves a = neouncant Secondary Coutainment Isolation Valves L = Water Leg Seal V = Vented to Secondary Containment with CLOC (Closed Loop &:tside Containment, see Subsection b.5.3.2.1) (3) Containment Seal Leakage Control System Provided. (4) Th4*d Isolation Valve (Remote Manual) Provided. ....r. generally operates in'a closed-loop mode, within ( I ', ~' the secondary containment. However, there are several lines such as flushing water, etc, which penetrate the secondary ...___.._ ant and offer a potential leakage path from the.pri-mary containment to environment. For such case, however, throughline leakage and bypass of the secondary containment j is precluded by the following: 'he line provides a source of makeup water to the RPV, a T# isolst4cn is necessary. mi b. If the line does not provide makeup to the RPV, isolation is provided by redundant valves at the secondary contain- .: a single valve with redundant solenoids. (6) Secondary containment leakage control is provided. Type of -- '- -- 4: shown in Figure 6.2-82 for each individual case. I 6.2-192 l

Ia Tabla 5 t-24 of your FSAR, you indicato the use of 1:atage control 480.1g ~ f6.2.3) systems es a method to preclude bypass leakage. However, lines relying en air and meter leakage control systems are actually bypass paths tp until the time when these systems become fully operational. Revise / Table 5.2-24 to indicate which lines are in this category. Indicate the amount of leakage espected to pass beyond the secondary containment from these lines. (t. g s e o W S C-t g 3 3 p A A 1 9 R I I .h.c_LWe s_t e.a_;rwp..M.E.24_ b utz.19:4

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I ~ '. ~. ' " - Q s ct:u.ru --Ku indicat; in youW5 Aft that neither the drywber(e system nor the (o.ic.3) drywell bleed-off ~ vent system is used to control the drywell pressure perturbations during normal speration. If either of these systems, or any other systess, will be operated through the standby gas treatment system (5GTS) in the event of a LOCA show that the 3GTS equipment will be able to withstand the LOCA pressure. p c s 70 niS E. 7hr. eL:__,,nx r..w i ~.1 A. y ; u u m-A 4A v%A MM = A 1L sa vt < - 4 _d a '++ L.OCA Mus. L _g A-... o 1 ->b-d. g/ e ? - _ n _ _. i U Cu u Y". "^k & _~- % %. u e oD g._ --._l J. - .c & A n Al h m_v1_42n M 4 d c. A. h 4-f m %4C 9 4: f. t 4, M. s .r m.y h. E i R eL=....oj n*. 4 4,, u.s. 4 & "' ~ " g i l l l

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l l 480.21 In Table 6.2 25 of your F5AR. you indicate that the purge exhaust line (6.2.4) isolation valve closure time is within 6 seconds whereas you indicate en page 1.5-171 that the purge valves are capable of full closure in 10 seconds at peak temperatures. Explain this apparent discrepancy. Indicate whether the purge valve closure time in excess of five seconds is factored into your dosage calculations, made pursuant to 10 CFR Part 100, in accordance with Branch Technical Position C53 6-4. 1 hNkt f A Y h A4 c nak4 aA-o-uae M4 6 M. 74 m a r-<-<ms a dm cn /. E - i,/ c1 6 MsaAe' LL wk/ 6o Z&& <rs <'ve && c/-p-ety =. & w / u y e y ca .svau-t hu>. S W,,zeslue-O% p(MDd 's ) & ~< ~ ~ m-o

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m ocra&Fy 238 NUCLEAR ISLAND Rcv. 0 (- IV.2.16 SRP Section 6.2.4 (BTP 6-4), Dated 11/24/75 h

Title:

BTP 6-4, Containment Purging During Normal Operation * (Category IV, Item B-5) This branch technical position pertains to system lines which can ...-_.... upen path from the containment to the environs during normal plant operation (e.g., the purge and vent lines of the

.t-im.t.cr.'. purge system).

It supplements the position taken in (' SRP Section 6.2.4. ~ Evaluation GESSAR Section 9.4.5, describes how the GESSAR design meets the STP. The containment purge system is designed to prevent any oignificant release of radioactive material following a loss-of-coolant or fuel-handling accident. Tne interconnectin'g piping of the penetration and the, values are Seismic Category I and Quality Group B. The space betwaen the two isolation valves allows a sealing air supply to the penetra-tion during a LOCA. This ensures that leakage out of the contain-ment can ne eliminated through the ventilation penetrations. A connection directly to the Standby Gas Treatment System serves tne drywell and containment when purging is required through the l otandby gas treatment and the isolation valves which are open 4 ([ during normal operation and capable of full closure in )6 seconds [8 6 -ature against the full pressure of the containment Ihe valves are fail safe which close on loss-of- .s actuating-air pressure. Remote manual controls also enable the .svlate the conta'inment ventilation systems. or......

" a LOCA, the signals f _r closure actuation include

( Tre: high drywell pressure and/or low-low reactor water level, O 1.8-171

480.22 Closed systems outside containment'having a post-7,ccident function become ,(6.2.4), Catensions of the containment houndary f;11owing a LOCA. Certaia of th:52 ~ ^ ' systems may c1so Le identified as one of the redundant containment is:- \\ lation harriers. Since these sys?.eas contain camponents whjch may leak, specify the leakase limits for each of these systems which becomes an extenston of the containment houndary followine a L-=- Provide a dis-cussion er your plans to leak test the systems either hydrostatically or pneumatically. Discuss how the leakage will be factored into your dosage calculations.mede pursuant to 10 CFR Part 100. Ic E. 5 P b N S Q Lankage testing of the closed ESF systems cutside - j iiiiiiiiie iiii i ieii i

I

! ii '-~ containment is performed in accordance with Section XI of the ASME Code as discussed in Section 6.6. There are no plans for specifying leakags limits. Any airborne [. radioactivity resulting from leakage from these ESF systems _ following a LOCA is processed through the standby gas _ treatment system (SGTS) prior to discharge to the_ environment, and the offsite doses from this source are _ small. This contributien bas been account.ed for in the _ , radiological assessment of the site. $ O M c4 w 5 '

15. G.E.S. )

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480.23 You state on pages 1.8-171 and 9.4-62 of your FSAR that the contairment (6.2.4) purge valves are open during normal operation and that containment pressure is controlled through these valves. However, we state our position on this matter in branch technical position CSB.6-4 that the use of large purge and vent lines should be restricted to cold shu'tdown conditions and refueling coutages. Provide your basis for purging the containment continuously in light of our position on this matter.

Response

Continuous purging of the Primary Containment outside of the drywell during reactor operation is required for access, inspection, and main-tneance associa ted with the Control Rod Drive - hydraulic control units (one per CRD); safety related instrument calibrations, water sampling of reactor water, suppression pool, and upper containment pools, Ra'CU system and feedwater. These activities involve several operating personnel occupying the primary containment during a significant portion of each shif t. The ventilation rate of 5,000 CFM provides an air change in the containment only every 3 hours and 45 minutes. This is minimal for controlling [ humidity, odors, and dilution of potential airbcrne radioactivity released from a small number of safety relief valve vents, RWCU system filter-demineralizer maintenance, and upper containment pool walls at the wet-dry line. When Branch Tecnnical Position CSB 6-4 was initially issued, the containment ver-ilation penetrations were modified to reduce their size 1. u

480.23 (continued) for normal operating continuous purge from 42-inch diameter to 18-inch diameter. The 42-inch penetration was open only during reactor shut-down and refueling to allow for higher ventilation flow rates when more operating personnel would be present and potential airborne radiation levels could be higher than during ncrmal operation. Fast closing isolation valves (6 seconds) were provided to close on LOCA signal or whenever the exhaust radiation sensors detected radiation levels high enough to exceed plant operating limits. During normal reactor operations with the 5,000 cpm ventilation rate, the airborne radiation levels must be less than 100FR20 limits inside the primary containment but outside the drywell. The drywell purge vents are closed during reactor operation and this is normal for all BWRs. The radiation monitors located in the exhaust duct are installed far enough away from the primary containment isolation valves so these valves can close before airborne radiation is released from the primary containment. As a further precaution, radiation monitors are also located near the upper containment pool surface 50 early detection of potential radioactivity can be detected during refueling when the reactor is open. A suppression pool cleanup system has been provided for the BWR6, Mark III to insure that radioactivity in the cool water can be kept low and reduce the amount of airborne radioactivity during abnormal plant operating events. 2.

480.23 (continued) This same topic was also an issue contained in NURG073~ Item II.E.4.2 which was addressed in Section 1A.29 6 "Cor ainment Isolation Dependa bili ty." Studies of suppression pool scrubbi.n; action has been found to retain considerable amounts of iodine in the event of a major pipe break and failure of fuel cladding. This same capability also applies to normal and abnormal reactor operations when potential safety relief valve simering occurs, and RCIC testing relea:Es condensed reactor steam to the suppression pool. All of these plant operating conditions have been studied and discussed with the stiff during the various reviews and licensing procedures. Periodic isolation valve testing is required during reactor operation to insure the ventilation isolation system is ever ready o function. This includes Appendix "J" leakage testing and radiation sensor test and cal i bration. Studies have been made to evaluate possible contairnent ventilation isolation during each day. The maintenance functions =ere planned such that each two shifts would schedule their work tack-to-back. At the end of the second shif t work, the containment could be isolated for a very short period because the containment would then require purging at a 25,000 cpm flow rate through the large 42-inch valees to insure the environment would be habitable for the following twe -ork shif ts. The small amount of time that containment isolation coulc be achieved, was l not cost beneficial for good plant operation and mai"mance. 1 3.

$60,24; Describe your provistens to ensure that debris will not become entrained (e.?.a) to the purge volves and prevent their closure. The guidance provided below represents acceptabla debris screen desipi crit:ria: a. The debris screen should be designed to seismic Category ? criteria and installed typically about one pipe diaerter away from the inner side of the inboard isolation valve. b. The piping between the debris screen and the valve should also be designed to seismic Category I criteria. c. The debris screen should be designed to withstand the differential pressure resulting from a LOCA. d. The debris screen openings typically should be about 2 inches by 1-3/16 inches. Q C S P b N 3 E-m A -- A A d e' O ' {- - d, AD' 'a _* _ w M k tv r - = U V ?WL ^ ^ - th 49s' ri"^ ' ' L' 't-l' - ^ >o my-Joe. .%.3 s ; a.-- +.>>. c -~. en.ct L et-D. -4 u. e to. o - & _+m 4f & n % __ w w + L & $ % k y M. (( - -L U. h.>

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480,25 F;r those lines required to apen f211ow1:g an accident and which have (6,2.4) cither two isolation valves outside containment or a sizgle remote-manual is31stion valve outside containment in a closed system, show that the i remotewmanual isolation valve outside the containment and the piping between the containment and the yalve is enclosed in a leak-tight or controlled leakage housing. Alternatively, show that the design of the piping up to and including the fiist isolation valve conforms to the ecceptance criteria of Section 3.6.2 of the SRP. In either case, discuss your proposed method to detect leakage from the valve shaft and/or bonnet sesis and piping. Discuss your proposed method to ter- ,ainate the leakage. g ( s fb N3 & All isolation valves located outside containment that are r open after an accident containment, a controlled leakage housing.(as shown in Table 6.2-25) equired to are within secondary bonnet sei.ls has net been pr:videC. Direct determination for le rygSz g $,pc7 Centnintent is. soc 5ted within du :: n ::E;CrtishtInct. ICCS r,yr. ten o.:tsice .ach ec.~.partisent hss lesh detectg devices With acoSpartme.:. alarms. ppropriate should any leakage (e.g., from a valve obsit) be detected by these (e. vices,, the plant operator can I

r. hut down the affected ECCS system and isolate it.

ing ECCS systemt The retai71-are capable of provid.ing adequate core cooling. 1

480.26 Provide a coupleto Caving for each closed system outside containment

(6.2.4) for which credit is claimed ia Tabla 6,2-24 as an is: 1stion barrier as well as a bypass leakage barrtr. Show all piping connections to the i

closed system up to a second isolation barrier, identify aT) lines connected to the closed systems which leave the secondary containment. Discuss the periodic laakage measurements which will be made on these closed systems. St CSf6ASE- ,L t t iI l I l 'l 1. I t iI I i 8 e = ? .._iTetc.__atirhd h %'t deanshe ss etVE-ti_ ' A.Ari.Ms o ! %it4_L4cDr hj_F.1.ILv.,1/.r., _E_4e.=.I.P O-i CeiG./cL.i3.r*? 4J-Ib 12 -4ule,1_ w_0 M,5'- lhZtan. onusys,anAtust_.sscewPnE. y@ reedrAuemets Astu ! uss14b'sMIED._Sh"DAlo-gag b o eswttug c4RekEht_Lhte(MfM1)_*G __Enrevv_.apal_ es wf.45%ss mer. N_Ennsat se_ u e, wieguae._;Anc ute,y.u'aureo l e._amha_.cgp em. i i I i 8 l 1 l I t i i i, I i i 1 8E*MT19 Ct_'_elFE_*!CM19L l.$ NEr$5i N_.915ctls$2A_

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W3-tUght emergency core cooling system p-(ECCS) pump rooms be sized so that if a fatizra of a seal or Csket on e 11.: taking tat:r from the suppression pool occurred inside th] pump th] volume of the suppression pool watcr needed to fill the room r" a. tould not reduce the suppression pool level telow the minimum drawdown level. Verify your conformance with our position on this matter by pro-viding the details of your analysis including the distance of the suction lines below the pool surface at the sinimum drawdown level. R C [/*6 (YS 0
h ECCS pump room is provided with leak detection capabilities discusse,$ in subsection 9.3.3.2.2.
If leakage from a seal gssk2t is detected in one of the pump rooms during normal nnt conditions, the remotely operated valve installed in the l ap cuction line would be closed, thereby isolating the leaking nponznt from the suppression pool water. There are no seals gaskats installed between the containment penetration and the >1e,tien valve. The only potential path for leakage of >pression pool. water into the ECCS pump rooms is through the ips' euction lines as these are the only lines which penetrate i containment at an elevation below the suppression pool water 41. Additional details of the ECCS pumps' suction lines are
svid2d 2n subsection 6.2.4.3.2.2.2.
rofore, the need to size the ECCS pump rooms so that the !.uma of suppression pool water needed to fill the ECCS pump 'm would not reduce the suppression pool level below the .imum drawline is not required due to the leak detection and lation capabilities incorporated in the design. The potential uction in suppression pool water inventory before detection isolation of a leaking seal or gasket in the pump room would insignificant. Suppression pool makeup water during normal nt condi+ia-9 fr f-em the condensate water storage tank. elevations of the ECCS pump suction lines and the suppression 1 minim tw&wn level are - (26'-0") and - (15'-3"), .pcctively. B O b l l
890.?O Verify that the reltf valve setpoints for the RHR A and B relief lines (penetrations 37C. 43C. and 40C). the LpCS rettf v41we setpoint (p:ne-tration 56C) ar.d high pressure core spray (WCS) pump test relisf valve setpoint (penetration 60C) are greater than 1.5 times the containeer.t design pressure. g C S PO N S & The setpoints for the relief valves are as follows: PENETRATIdN RELIEF VALVES SERVED SETPOIrir NO. 37C E12-F055A 500 psig' 39C E12-F055B 500 psig 40C E12-F036 75 psig 43C E12-F017 200 psig 3G; E21-F018, F031 600, 100 psig tidC E22-F035, F014 1560, 100 psig T4e, e-onMis me44 dei ep pc.n k m I S. 15 pri 3 e )
Specify whether the controVCCRfC3m 400,23 .(5,2.4) to reactor water cleanup (MEU) samp1] penet are engineered safety fcature systems or engineered safety feature system < elated. (Noindicationis given la Table 6,2-253) Provide schenstics for each line penetrating containment showing all connections and branch lines between tha isolation barriers inside and outside containment. These schematics should include, but need not he limited to, the piping and instrumentation drawings refer-enced in Table 6.2-25 of your FSAR. Provide justification indicating why test lines should not he treated as branch lines and included in the containment isolation valve tables and tested in accordance with Appendix ( J to 10 CFR part 50. 5 RcardddL a) The CRD and Demineralized Water to RWCU sample panel are not ESF in ' term of their function. Table 6.2,25 of the FSAP,will be revised to - include these as not ESF. b)andc) Schematics will be added to FSAR. Table 6.2-25 will be revised to include the branch lines. d) Tm m 444 r m otE Iou 48 d h* 3 * * * *d aos h an e., c u +s >.a-e-> d ase +rn a s e %.. ra ua N4- !%k "4 J 4-V 4(A 4.1
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m-.- eent integrated leak reta test, certata systems which are normally fi11cd (6,2.4)' cith wat:r and are opergting under past-LOCA conditions, need not be vented 13 the containment atmosphere. Provide a tabla listi:tg these systems. In addition to being normally filled with watcr. these systems.must satisfy the following criteria: (1) the system mast be protected against missiles and pipe whip; (2) the system est be designed to seismic Category I criteria; (3) the system est be classified Safety Class 2; and (4) the system mst hanin full ef water for 30 days, assuming a single active failure. State whether these systees meet these requirements. k 6.S P O (9 3 Q t- _4 i u.. 2 ie i', i i:: IMidS1S; C.aMSS.cP.!WF*! CQ1'5(M..iU4mMMEll!"" - (n>dPC.) MsW3 E445.._h3H1c,et MC ov'Ndq_ [v.CL t.A t*o:if.;-Loc.a /-#_.4!ano4% 4xG:, .Som.'iv.sr_n40m MtML6.7'r. T trim VJbiPy i
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PW.5m HER_7c' ! M tS M ic,. ;CatTq citef_ Tj ! t.QvALSTM_2.$mO47. S_ AiEL $rWW ~ _J._ ( LA+fe. 5 met.mteEMF s3G ' >I I! (5-3. ) G.wgi1.Cp To p;W,MA1pl.FWfc CG NUAI'EL. l I' J ._? rom._Ar_ LW.A st'_SQ_ vaaf 4 - U G E.__5,% w w E.Au-a_ I i(1LEkiR_.Le cfLA,.1b,t,L.' I I 3 ' M W4CPeeAvrrti x.asL.L"afe*v_Starcti. LLoc_.YnE%w.aKJg.e;c 4_'t*1_S mit-M I i .5 % T E M. g A T_t.r u ri o d. K4rugnyg: i i i i i i i i i i i
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lt _$_ W Yb 4EG.31 the secondary containment and, consequently, repres (5.2.4), l leakage paths. Examples of systems containing air are the containment and instrument air lines (penetration 129c and 130c). purge s times for these air-filled lines range from six to 20 seconds.The valve closure closing times should have been established on the basis of minimizing the These l, i release of containment atmosphere to the environs, mitigating the offsite t, radiological consequences and assuring that the ECC5 effectiveness is not degraded by a redustion in the containment back-pressure. discussion of the analyses you performed to evaluate these radiological provide a .. consequences and the effect on the containment atmosphere. valve closure times for these systems. Justify the ) i em i 1< < S /> e 6 1 I M $4A LAJ4 WO-2I n% w ms Qa cep of OW M + n u.
480.32 Describe the containment seaMeakage control system esing drawings f (6.2.4) clarity. Indicate how this system functions to prec1tde bypass laakage on system lines such as the service and irstrument air distribution lines which terminata keyond the secondary containment. Describe how this system interacts with the air and water positive leakage control systems , Indicate how and when this system is_ actuated. @ ES/*o.d&C [ tis Caio rAIAlmed7 /SotA7tM t/AL.vE l OVAGS Mk% Q't.7er ms AAE DC5ckW O f N Mc:Tttw G., G. "f. L oF GE55 Pin.E THt55E AAE. CO A*. PosiTsa. 9EM.rsotATteu VA.V. E. LcAwvd5 CMADI QQ Q)gyc.p n THE MoTu C5) o F TABL.s (, 4.-- 2.1-pas stEE:N R.ec o.o 7c. F b s. c La p. s i c.A Tt o m. O
GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. O l" s Table 6.2-24 (- EVALUATION OF POTENTIAL CONTAINMENT BYPASS ~ 12AKAGE PATHS (Continued) NOTES: 6;; Termination Region S = Secondary containment (ECCS Rooms or Fuel Building). Lines terminating within the secondary containment are not potential throughline leakage path. E = Environmental, beyond secondary containment. Such lines (. either pass directly through the secondary containment to the environment, or are connected to branch lines which pass through the secondary containment to the environment. For either case, potential throughline leakage is pre-cluded by a combination of leakage barrier. (2) Bypass Leakage Barriers Cw dedundant Primary Containment Isolation Valves A = Redundant Secondary Containment Isolation Valves L = Water Lag Seal i V = Vented to Secondary Containment with CLOC (Closed Loop outside Containment, see Subsection 6.5.3.2.1) e-r "P u.v.s e T seur.lo W us t-u meu Am. 6 N roi (3) ^^a*-i rr:nt C;;l 10-'=0= rnn+ val Cy;t==4 rovided. (3rt Skuur.Ja E P a.s. .2.i,s.c.s.3.u. 6.,y .p (4) Third Isolation Valve- (Remote Manual) Provided. (5) The system generally operates in a closed-loop mode, within the secondary containment. However, there are several lines such as flushing water, etc, which penetrate the secondary containment and offer a potential leakage path from the pri-mary containment to environment. For such case, however, throughline leakage and bypass of the secondary containment is precluded by the following: e( If the line provides a source of makeup water to the RPV, no isolation is necessary. b. If the line does not provide makeup to the RPV, isolation is provided by redundant valves at the secondary contain-w nt or a single valve with redundant solenoids. (6) Secondary containment. leakage control is provided. Type of protection is shown in Figure 6.2-52 for each individual case. s 6.2-192 1 l
480.33' Tha reactor C:re isolatica Cooling (RCIC) head spray line (penetration 48C) ( $
) AY Contains a s15pl3 Check valve outside Containment which does not meet th2 i
i requirements of GDC 55. Provide a Commitment to meet the requirentnts of GDC 55 or justify the deviation. 62 E S 4 6NJ& The 9. crc w s9%y t.swe wers see.o eenovreo iuTo he. [-su7s(t. Lim EU r-The Rag. 5 Huioowvs cootw. r FurJLhw s,Tiu exWn V' A THE t iwe ia MsTtM. esotaTem Met 0 To HAve. Tuao ci+ecAc. vAs.VU Vmes A Chec.K VM g 150L A74 8/h. NS1OC AW A moTdr 09Gr/CEr3 GATE VAL.vt OOT5i DE IS 3 PDNTLy fbEIMG ReuteWO, he, gesocs c)F Twts 92Vstw. uw ee. puseereo av me_ x%uay e,
480,34 The fuel transfer penetratten. SC contains yta velves inside and outside (6,2.4) containment. Add this penetretten to Tabla 622-25 of your FSAR. Indicato whether this system is en ESF system. If not, the velves should be iso-lated on diverse isolatten sipals and this fact should be added to Table 5.2-25, including the velve closure time. ( l. a?_ E.s P b RE. 1 1 h l l l ] t . Tat FvE.C._wsese_svet&rsti_Fr9ttwerug 4EC: is v a AdC i hG i *mers.ca wtrCA_ts 3M9__ LAB 4E .aca.S_To EN $_s.7f4C.CMr4uh*. ETAT 14ep oM. "T ati_. M a.a.weg,ge,W M T tt .D s b i t u G E.m. r>i L.f m G L S 1 4 4. __ m.F. A E*2 5 M 2 ' n ' 65Ec.cn e_hi Sl_r.S. b.L_lCL 5 %.t1 3. n 2 r is v.,G e f 9 TP_ 5 H r d.__Tr u s _ i KWI) 1:L A Q'a E. Tct_T R.SiLEE,_.LS.0MCLG_f _Nor___\\/wgr < F('2.-T-Q9 7 a.OP V42-Fooet, 'k i i il ii lil i i l i i l I ! ; I t i # i i i iie i.i i i i i i e it i
m M F W esto that the air a:A cater R ~ positive leaksip control systems are manually ccteated and Gill be j titiated cith' a about to minutes f;11owtg a postulated design basis LOCA, provide a detailed discussion of the operator cctions required, facluding valve posittening and the equipment line-up required to initiate the leakage control systems relied upor. to preclude bypass leakage. Indicate the time foi
==t ion of a s stene, ^fpce_
the become fullr operations Discuss the exten to it a estulded agie sc6iA Nea,%.as failure af an isolation valve to close, can be accommodated. / &E&PbNsq l t i i I I I Il l l! ll Il } } l !i LaierAw 4.G3 is ! cysvaro to scwc)E. 'I SEfe.is,Afriou ' ow e>mGr. ants 3mgejng An91TMittV WAVRhk..dWd.hEfte.li-GleL4 f M62211MtiL%tLce.VL*r4i ALL hmMs%D_.Strets i Acea9E F%orea V' : s,e See Averzwngum i! i i e ti ! i + i,;,i!i i t i ! i. > > ..e i. ,,,.4 ., i ( \\ I l
atssuwnr - r v~ 238 NUCLEAR ISLAND Rsv. 0 6.5.3.2 Secondary containment (Continued) f' -( SGTS prior to release to the environment.' Refer to Subsec-tion 6.2.3 for a description of the secondary containment boundary and Subsection 6.5.1 for the SGTS description. The shield annulus return / exhaust system described in Subsec-tion 9.4.5 provides for mixing in the annulus. The Standby Gas (J, Treatment System (SGTS) described in Subsection 6.5.1 provides exhaust filtration to remove iodines and also maintain the secon-dary containment at a negative pressure. The secondary containment provides its fission product control l function when the (-)l/4-in, w.g. pressure in established post-accident. Refer to Subsection 6.5.1 for specific times. 6.5.3.3 Isolation valve Leakage control System n The Isolation valve Leakage Control System limits the throughline leakage (past isolation valve) of the containment atmosphere to secondary containment and prevents such leakage to the environs. These systems consist of the Main Steam Isolation valve Air Posi-tion Leakage Control System, the Air Positive Seal-Isolation Valve Leakage Control System, the Water Positive Seal-Isolation Valve Leakage Control System, the Closed-Loop outside Containment Sea'. / System and the Isolation Valve Water Leg Leakage Control System. \\- hJI g sg C) opeo~w % k ei.xit*xs Auv Tm ws-The following is a discussion of each of these systems, except for ()f the Main Steam Isolation Air Positive Leakage Control System, which (Nel is described in Section 6.7. ppg $b The identification of the method used to provide containment isola-tion valve leakage control for the various penetrating containment is given in Table 6.2-24. 6.5-19 u A <r ~ ~
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gs.rFJ at prictEil, as required by EDC 55, 56, and 57. _..._ _ _ _ _....... - wn a nment However, it appears that many of the outermost containment isolation valves listed in Table 5,2-25 cf your FSAR are guite fcr from the containment boundary. One lina (penetration 39c) has a valve which is 44 feet from, tha contain-Discuss the extent to which containment / solation valves have ment. been located as close at practical to the containment boundary. Provie: unre details for each penetration, including the actual distance from the outer valve to the containment barrier. / odEEF4NJL l Isolation valves are located as'close to the containment penetration as practical. This consideration is made in all cases however, the following kinds of influences require some valves to have a considerable amount of piping between the containment fitting and the firi,t valve outside containment. The location of relief valves discharging into the suppression pool is dictated by design considerations to vertically mount the valves at a high point of the system being relieved. Vent and drain lines to the suppression pool have similar considerations to be located at the high or low point in the piping. These lines as well as minimum flow bypass and test lines are joined together where practical in order to reduce the number of pana+rneions through containment. Furthermore, tha consequences of a break in these lines result in no sign'ificant safety consideration as all terminate below the minimum drawdown level in the suppression pool. See Sections 6.2.4.3.2.2.1.1 and 6.2.4.3.2.2.1.3 for additional discussion. The attached Table 6.2-25 is marked up with the necessary l
.dditi.:..: and revisions.
).e p e..l i .R R Table Ee2-25 CONTAr9 MENT ISOLATION VALVE 2NFORMATION 1 .: = S 3: =ies 4 e 8 aet -u-I 3 hr - wi. ~e mies.eesus. I g j 5l E 38 i i 3 i? 3 a p [ 14 p 3 [ } l el ! "] y a 33
s:
3,
5 n e-I n
j p: = r3 m = 2 h: _ .s- ._c _L:. E., -_s. _E L a s a ~3 e l 15 I: Is p e _ i_s
ac orete
c _a s i a cu .e t 1: = ,e SS no .te une. .te.m a veo i.m .u-== 1 = .ie.e en.um. Piot. = = we. not. hot.et 2-S 2
t...et..
.te. a no R-iS...n-n=. =. .ie.e pnee. Pieten = s... e... .o. .et at.et et S-S i to let SS main steemitne 9 Steen 26 Tee R-1024 321-F0229 I We = Clebe pneen. Platee = Oree Shot Shut ehet 3-S 2 4 C.D.E.F.W. s Steam M Tee R-1344.8 921-r0290 0 We O Slobe pneum. Platen = J.m.tn.get Spee Shot shot meet 8 es lle $$ Rota Steamitne A Steam M Tee R 1024 B21*F0224 I no Cle%e Pneum. Pieten = Opee Shot What that $=$ 3 se C.S.E.F.g. q,3 = s Steam M Tee m.l56A.3 321-F02s4 o no e State pneues. Plates = Spee Shot Shot Rhet >S 1 to J m.ne. ass g 12C SS noin Steaaline C Steam M Tee R 102A 921*F022C f les = Globe Pneum. Pietem
Opee Shot Shot Shot FS f
to C.s.t.P m. 5 e Steam 26 Too R.1364.3 B21-F020C O No O Clebe pneum. Pieten J.s.ga,pge %e Wen Shot Shot Shut 3-S 1 g O = h,) 9C SS sust crain a water 1 1/2 Tee R.1024.8 521-r06?O O I he 4 Cate fester Elect. Man. Open shot shot shot 7.5 l n C.a.t.P.u, pm R.156s .... hr.e. 3p q) loc SS Inst Drate B Water 1-1/2 fee B21-F0679 0 -> me 6 l Sete fester Eteet. Non. Speg Bhut Shot Shot T.S 1 eg y W 11C SS MSL Drain A Water 1-1/3 Tee 323E0674 0 ' as S 'g Sete pester 33eet, seen. twen Shot Shot Shot T.S 1 O b4 12C $$ sest Drain C Water !=I/2 Too 321*F06?C 0 no 6 , este gester 24ect, geon. epen Shut Shot Shot 1.5 _1 ee 12C SS PSt. Drean Water 3 Tee R 1029 528*F016 I see = Sete fester Elect. seen. Open shut Rhet Rhet il 3 se C.D.C.F.R. 4 J.N.pt.tst I teater 3 R-1544.9 33t+F019 0 0 Sete fester RIeet. seen. epee ghet shot shut 35 1 o U treter 3 321-roes 0 le cate meter Elect. seen. cyon ehet shot shot LS 3 sp IOC SS Feed Water Llee A Water 20 Tee R-1920 921*F0104 I no - Cheet Seit last.' = t. = epee Shut shot = Water 20 4 921*FOl34 0 0 Cheet P.ea um. Self = Open Shut Shot = test. = 3 R-150 = 3,e SS Fee..ater u...eter = no R-i.= .n-,0= i

cf.ee. .eir Ope. not et = = Water 20 6 921*F0323 O O Cheet Pneum. Self = j Open Shot Shot = Inst. 3 = R-ISS = 2SC 55 contret med Ortoe water 2 R-104A.C Cll*F003 0 lee 4 Sete tester sleet, noe open shut shot As to to 1 g set water 2 S T11-FF215 1 Cate footer tieet. man. Open Shot Shot As le SG 3 T He = Water 2 R-IS7 Cit *F122 3 rheet Self = =
1 O en shot Shot P
= Inst = 3tc SS LpC3 A water le Tee R*1070 212-F027A 0 No 3 Cete fester tiect. pese. Open open ne go gg 1 3 R.gg.WI = 12 E12-F0424 1 Gate I**t er tiett. Shot shot oped Ao le 24 1 1 C.R.L.V.WI e 32 Elt r0414 I = Cheet Pneum. Self Shut Shut CMn = = Inst = I 33 E!).F02RA I Cate looter Eleft. = Shot Shut = As to 60 1 3 C.R.WI 14 C12*F037A I Clobe Meter Eleet. = = Shot shot Shot As to 60 1 I Itt 20C 36 FR8l/IJCf A water 24 Tee R-1079 212-F0044 0 ges O Cate peoter Elect. Man. Open Shot open As to wC 1 0 RM e 4 e I I g
A e e A Tablo 6.2-25 CCMINMENT ISOIX ION VALVE INFORMA 5 ION (Continued) j !E m .ies 1 e h e.t !c; g. j j.,. .ies .e.4e .ies
siue, r
8 y 3 e s N} I 3= 2e g li
j },
s: 3 eI I :a g.
6:
3 rr
3 r8
-l g:L a s-e h: se: u s ea a 3: 38 h as E z v c - e ~ z ?aa e ix:
:a e
= 1 n' ,,.te. = a 74,. ,s A,e Secu.n .eter to veo R-ie,. n 2-re 2 C.t. et., neo. ,een. et ht Ae no n i A.,... 20 ut-r00, i = Cato amt.r neo. = S, et S, A. Io n 2 A.... 33C S4 fore & Puup Test Water 14 Teo R-3074 212-r0244 0 no IS Cate senter Sleet. Sten. Shot Shot Shot he to 40 1 8 C.E.gg.48-manteen Fle. 4 t12-tentA e lebe pester Elect. Men. Shot Shot Shot As to 40 1 i C.R.co.858 J2 h -tate 4 E12-r0644 O Motor ricet. Mao. Opes Ao le T 1 8 gue = = pc n me A a.itet veter 4 Tee R-lats E 2-restA o 40 tief self shot shot shot 1 M Inst Line 2 t12-r001 0 25 settet Sett shot shot Shot 3 W = Inst = I 812-rF2 34 O $8 pelief Self Shot ht Shot = 8 E Inst = = 20C SS LPCf 3 Water 14 Tee R-1979 EI2=F027e O se 2 Cate faster Elect, men. Open Open As to te 2 i R.gr se esp = 32 2:2-r042e I c.te neeer rieet. shot shn even As to 2 2 8 C.e.i..v.ne .Eo = m 12 212-F0418 I Cheet Pneum. Self Shot Shot Open a Inst 5 = e 12 312-r028e t = gg Cato Isoter Eleet. = Shot ht As to te 2 3 C.R gg y 14 E12-r03?e i j g Ctehe Isotor Eteet. = Shot, Shot that As te 40 2 1 mg = I Slt S4 sursVIJCS S Water 24 Tee R-18?B E12-r004e O see O Cate Motor tiect. fien. Open shot Open As to to 3 0 = 34C se ma a top Test water 14 Tee t-107C t:2-r024e o l l se 2e cate smoter tseet, seen, shot Shot he Ae se se 2 C.R es. set NN g Manteum rio. 6 E12-rette O l 82 , Clobe fester tiect, seen. Shut Shot Shot Ae to te 2 1 C.E. set 26 h 4 112-F044e O . ate peoter Elect. Man. Open gg As to T 2 I he = = 43C '4 WR 9 Dettet water 2 Teo gg R 10?s.C tit-rotte O l jPellef 43 o lief self e shot thet stiet = teet t = Line 2 t12-r020 o Tee 64 = Self Shot shut thet = Inst t 40C St Pjes A/T peltet Cend. e fee R*101C 212*r036 0 l No 2 pelief Self Shot shut ht = = Inst I = co.dene.te O 3?C Se m e A 8ettet 18eter 12 Too R*1974 212*r0154 { p no 45 'jeelief self t' = Shot Shut shoe seet = e w s. 12 Ett-r0%C 4S l hellef @ g=,
p.,,, g g 1
2:2-r07:4 .44-tehe aster steet. noe. Shot shut shut A. Se as 1-I set Self Shot Shot Shot e = snet t = M % tobe senter Eteet. Man. that shot shut as to 15 1 8 tot 1 RI2-r0144 0-39C 16 we a peller w ter 12 Tee R-107C tl2-r0150 0 se 45 ellef a Self = Shot Shot 3%se 3 eg I 12 212-r0550 0 45 tief = self = Shot ht = t.et 3 = i 1 t!2-r013e. 1 .44. Cinbe Bester tieet.
pan, he shot ehet ne to 1g 3
3 get C* h b N 4 1 ES2.r074e o .j4-3 0 : who senter tieet. eien. Shot shut Shut A. to as - 2 3 see 29C M LPCI C afster 32 Tee R-1979 E12-r041C 1 10 0 % Pet Pneum. Self '4 met Shot Nt Opes = 3 = 14 812-r042C O O to fester Eleet. faen. Shot Shot Open As is 24 3 t C.R.t C 33 2SC 56 fuen C Pus, Test treter le Tee R-30?C 212-re25 e so 20 Ctehe soster Eseet, eten. Shot shut ht As.go to 2 C.a.m pinteen rise 4 C12-ros.4C l O 59 % Cate Stater CIect. Been. Open As to ? 2 .I r.Its = 42C 54 m n C pelief Water 2 Tes R-1979 E13-F02SC > su pelief Self ht ht Shot taet e i = tLee 1-1/2 E12*r101 0 $0 Petter Self Shot Shot Shot p Inst e t 32C 16 mm/L9CI C h *er 24 Tee R lete 212 #105 0 pe 4 Cate 8eoter tiect, men. open shot open As le 60 ts SM M te 56 Feel Transfer wn.. R-114 F42-r002 I Cate Pieten p d. peseg Shut ht At le N >N y e 8 F42-roed 0 0 Cate eseten syd. shot ht Shut As to 9 44
O W
O e e m% 4 I
j m A rag Table 6.2-25 C01 ?AINMENT ISOL. TION VALV.I INFORM \\ TION (Continued) -= , mise = y, ut gm 3 ~ g esta neee votes peestaan -{ g Ic, 2 s 3 fp 3, e: e - 3= i
- n:
e-e !e e a -1 s 3 e
3..,.
H a i h:
ec
.e 3 r p: 3
In e
3 e o i: eac .,ete. c s e e.s a ne,, .eie Steam ete-1 vee uc. m-=> = to .nete, noci. m. t so.et n to i. 2 Se,,i, 1 tu-ra. te sea. neo. ,,s. ~t n As.. i. 3 .....9.,.. RSt-r076 0 3 Clebe seeter Elect. seen. What feet inhet Am In S 2 se.V. sue Sec SS acte furtine Cond. 16 Tee B-1104 211-rece o too e cats septer flows, seen, spoe .ge, eg,e A. to em 3 8 ese Enhavet 16 E31-F040 0 6 Cheet Seir Shut sIhet 89e9 Mhet fnee = 4 160C 56 pCic feettne Air 3 E-1104 251-rote e o sete sentee tiect. nem stee egen sh=t As se ?.9 / e F.st.ast M tahavet Waesen ESt-re?? 0 2 Sete leeter t!eet. seen. Otee.en he Ao la ?.4 3 8 R.ee.ase (at. L8me W 4ec $$. ACI" poed Spray Water 6 Tee R-1104 211-7064 2 Cheet Pneem. Seir flhet What elpen e test e i = 6 251-r065 0 De O Cheet Pneen. Seat Shut shut egen. enet = t 0% Asc $6 actc remy water e Tee R-lien 258-7031 0 tie O Gate senter tieet. Ison. Shut shot eten As to 89 0 8' W.II t.s.v 9, f) e Sectlen N V.w,R.est 5.C S4 pCIC Mintoum unter 2 Tee R-180m ESI-r019 O No 2 State Noter Eleet. seen. SP A Shut shut Am to 9 9 9 T.T.ese elee tD 13C SS LPet Injeetten unter 32 Tes R-100 221 reef 9 5 Sete senter 23ect, seen. Seest Shot tgen As to 24 9 3 (*,r.",g,gpg ut 12 221-r006 1 Cheet Pneum. Selt S'est sIhet
gsref Inst

t
= g.g y $4C S4 IMS Fwp Seetten teater 24 Tee R-100 221*F005 0 Ise e sete sentee fleet. seen. Cguen eyese.al.g As to te 3 ee see MM g ,,C ,6 .mn.o_,,_ t,2, r.,2 le.e S,s .et t., 12 e R=,. 2 e s.st.r meet. se.n. t. .het A, to 6 t t..e., ,081 i. .te ..or meet. seen. S,.se ,.a ,se.t ~,. t t .,s. e See $6 LPCs setter water 2 Tee R-100 R21-rele O No 2 Deller Sett Seset R%et %t
Inst =
t U = = 2 * . til-roll O 49 pellef Seit bhet neemt shut In t. t = Sec SS NPCS Injeetten water 12 Tee R-1998 222-P004 tio 14 Sete gester steet, seen. shut ashot eg,ee Ae se 1 3 q'.r.est 12 222-F001 Cheet Pneum. Sett Shot Shot Wen-Inst
I
= 60C 56 NPCS res, test water 12 fee R-leen 222-7022 )0 Iso tote gheter tjeet, seen. Shot Shut Shot Ae to 1 3 9.W.001 b lobe faster Eleet. Men. Shot shut 39eet As to 3 4 Ice 10 t 822-rolei O 10 R-109s 222-rett Io 80 @ : tee,e sieter taeet, pen. Shot Shot Shut Ae to 3 es age g _ C1;telbl 0 10 ^ to sester Elect. Man. Shut Shot Shot Ae t e - 2 I hse FOM E22-% o 12 seller a Seit mt Shot shut ,teet I = ) 222-role O to 4{0 O h ;ellet Sett ht Shot Shot Inst = = 1 I 10 E-324A P46 open open %t Shot 2 O' me i tete rneen. Air = SSC 56 srPCS Sortion teater 24 fee E-tove E22-roll 0 0** d cate easter fleet, fean. Shot shut shot ne to 3 es v.v.We nic 55 PMCV Pony Water S see R*ll24 C33*r001 4 tio e. Cate fester tject. Han. Cpen Open Shut As le 11 2 0 p.s.CC.tst Settlen V.l*. 884 4 No C33 r004 0 2 Cate fester Eleet. Men. Open Open ht As to SS 1 to n.a.cC.te:. v.r.spe 64C $6 pescu petsen to Water 6 sie R-II29 C33-r0 39 0 0 Cate fester Sleet. Isen. Open Open shot As to 19 1 e p.e.CC.tt. M eeedvetor Line i W.l*.hPI N I%I' s on-r040 : cete wter 21.et. n... cpen Ope. Siist Ao le is 2 o s.t.iv.u, fD # v.r.m <: a e Q O a s agh na
t i e p., ^ o i-A /*N s g r L E Table 6.2-25 = e CONTAINMENT ISOLATION VALVE INFORMATION (Continued) =. e E
5 volse 3
jg Aeteetten. 7 e e e s ,else we elee o.elues t r = g g g h3 I l
f. d.
) I e s j j 1 j
,j j
j r 3 rf s'( e _H 1 r3 3_;g _15 h _h a- -e [o 1 a,,E 3 p-j] i !? ij ) h ) } 3 3 i r .,et. e
== 70C 56 Centainment vee. Air 24 Teo R-368 T41 Pr0334 0 Tee II 9' fly Pnese. Air Shot ehet Shot Ithat 19 l t Bel = mettet 24 T44-Fr034A I Check Self = shut shot shot snest poet = t = phet phet Rhet phet le t 0 get 72C 56 Centatnpent vee. Alt 24 fee R-160 ' Tel-rr032s 0 Tee 12 S' fly Pnese Alt = f Relief 24 T41-rf0348 I Cheet = Self = Nhet 8%et what ahet taet 3 = = b Il4C 56 Drywell 6 Catet. Water 3 No R 11GS PSS-Froll 1 De - Cate Mnter fleet. Sten. teen tapen Rhet As to SS 2 es p.t.get Q cate noter Eleet. men. open open shot As to IS I .e m.m.see M Egetp. Orann ICse) 3 6 PSS-Fro!! O Seep to Clean E-IS? Lef madweste g IISC $6 Orywell 6 Catst. Water 3 ses x-IlgC PS6-Pf036 I po Cate saetor tieet. nom. open apen the: As to l$ 3 es 9.s.se g 4 Ol Fleer Owain supp (Dawl 3 PS6-Fr039 0 0 Cate motor tiect. seen, open open shot As to IS I ee , D.R. Bet f e to Dirty Sadwaste R-IS? U N 164C S4 Orywell steed Alt 2 No K-160 T41-Pf03S t Tee Sete Meter Elect. seen. Shot Shot se,et Ae to le 2 es 9.E.lWI = E I 2 T48-tr036 0 2 Cate ptoter Elect. Man. Shot shut shot As to le I to P.E.les ( M 2 T41-rr0$0 t Cate Motor Elect. P.an. shut shot Rhet Ae to 3
e D.r.lft yi
= 2 Tel-frcSt 0 4 Cne Ran. Man. = shot shot shut Ao le E C. = ee y F 120C ST CCW supply water IS Teo K-120A P42-tr010 0 De 3 date Motor cleet man. Cppe Open thet Ao f g 50 1 I W.R.M 6 NI E 10 E=lSt P42-rr13S I Cate Motor Cleet geen. Open ripen Shot As te SO 2 1 9.r.NI UI{ = h; 121C St Ccw pee =rn water le fee R-120A P42-tro n
e =
Cate sister cleet Men. open epen shot Ao se So 3 1 p.t. net . a, P42-Pro,, 0 i f. gl 10 3 C.te eter flee,. n. n - 1 t 12SC St 8t Ch111 water water 6-no K-12SC P39-rr049 0 po O Cate pneum. Alt open open Shot shot 6 I t e.R tse = g supply 6 P39-87050 1 Cheek Self ctwo open shut that test 3 = = = 1*6C 57 at Chllt water Water 6 We R*12SC P39-rf058 3 Its
Cate Pneem. Air gaan. Open rJpen shut shut 6 3
f S.R.3e0 t .etern 6 0 6 R-!S7 P39-Pr052 0 0 Cate pneue. Air = open ep;n shot shot e I t p,s.sse 12?C 56 Condensate water water 6 see K-1246 P46-Fro 63 0 les 14 e Cate Pneen. Alt c en shot Shut 6 1 I S.R.fot Open r = 5 6 4 E-IS? P46-Fr193 1 = Cate Pneen. Air Open Open shot shot S 3 3 3.E set = 129C 16 Serette Air Air 4 3'o f*138 PSt-Fr000 0 De O Cate Motor tieet. Stan. Open Open shot Ae to 29 3 .I 3.u.pn l { 4 E-150 P61-rrol0 t = Cete Meter Eleet. seen. open open shut As le 3g, 3 3 3.R.1st 130C S6 teatrwient Alr Alr 3 see K-1308 PS2-Fr038 0 se O Cate soeter tiert. men. Open open shut As to 15 3 I s.t. net 6 3 E-ISS PS2-Pr040 I Cato noter Elect. Man. Open open shut As to IS 2 1 8.8.8e8 i r m< s_ e h e e O I k
l>Lw. O Table 6.2-25 ( CONTAINMENT ISOLATION VALVE INFORMATION (Continued) ( ll) Isolation Signal Codes (Continued) Signal Description T High pr' essure RCIC turbine exhaust diaphragm P High reactor vessel pressure - close R11R - shut-down cooling valves and head cooling valves V Close-through electrical interlocks with other p{ valves or pump motors W Condensate storage tank low level X Suppression pool low level Y RCIC 2 LPCS injection valve pressure AA. LPCS low flow BB Low main steamline pressure at inlet tio turbine (RUN mode only) . Line break in reactor water cleanup system - (high CC space temperature) DD Containment Pressure 2 High differential flow in the Reactor Water Cleanup System TT Annulus to Containment high differential pressure RM Remote manual switch from control room (All auto-i ina' tic initiated isolation valves are capable of remote manual operation from the control room.) (2) See Table 6.2-24 A W-(( W i s vu.va. g A s Aoo: Tron At. :: ^. ' vat.va sETwEEA t's uu,voa t courAismec [.' ' l ~ ( i tU 1 6.2-200 l l l
L3 U KUuMU/ numerous reports on the unsatisfactorfinirformance of' g u2Rb (6.2.4) the resilient seats for butt:rfly-type is'slation valves in th] contain-ment purge and went lines (refcr to 01E Circular 77-11, dated September 6,1977), we established Generit, Issua B-20, " Containment Leakage Due to Seal Deterioration," to evaluate the problem and to specify testing frequency for the isolation velves. Excessive leakage past the resi-lient seats of isolation velves in purge / vent lines is typically caused by severe environmental conditions and/or wear due to frequent use. Consequently the leakage for these valves should be dependent upon the occurrenc,e of severe environmental conditions and the use of the valves, 'rather than the current requirements of Appendix J to 10 CFR part 50. In this mgard, we reconsnand that the following provisions be added to the Technical Specifications for the leak testing of purge / vent line isolation valves: ment isolation valves with resilient material seals in: ) j (a) active purge / vent systems (i.e., those which may be ( operated during plant operating Modes 1 through 4) at least once every three months; and (b) passive purge systems (i.e., those which must be administrative 1y l controlled closed during reactor operating Modes 1 through 4) at least once every six months." In light of Ofvide a commitment to test these butterfly valves in this manner. g E. S P o N S E. ~.. b Yo Lruok.ong-u & ust, kst M ' d.1 exi& J>e. nAdA .h Lhklu.%&2oxik;td k%.aM&t _ Tial, Ao a.,M~.n2 Ulf.dh.m le a&: lit. 'l' P3N LC ? k* c[. 4 & PsAA V
' ~ TVN3 ) GESSAR II 238 NUCLEAR ISLAND 22A7007 Rnv. 3 / 6.2.4.4 Tests and Inspections (C'intinued) ( A discussion of testing and inspection of isolation. valves is provided in Subsection 6.2.1.6.ar.d Ch gi.m IG. Table 6.2-25 lists all isolati6n valves. Instruments are periodically tested and 4wenected.,fest and/or calibration points are supplied with each g instru W t f ^- b '? E$ kE- - ~= ~ ~, _ _ g f, , ~ / ~ ~ _ _ a h m, \\ / t. s. t wh -; u A - - + 4A j 3 9 f & Q,.'_ _ - _ k a l __1. +, 2 [ . y _n. ; __
n. fL ga,,0.1
] ) Q ,k J 0-t w &: - y&>_ ( g _D.-k I I (a. / t x = J r uve, a s y - w 3..... 4 purge the containment-drywell space to the annulus. The annulus' cxhaust is then processed by the Standby Gas Treatment System ( CC"'S ). Since there is no design requirement for the Combustible Gas Con-trol System in the absence of a LOCA, the following discussion of the requirements for, and the performance of, a Combustible Gas control System presumes that a LOCA has occurred. .y 6.2.5.1 Design Bases Following are generalized criteria that serve as the bases for . design of the Combustible Gas Control System: (1) The system is designed to control combustible gas concentration in containment to nonexplosive levels when k, ' the generation rate of hydrogen is calculated in 6.2-118
...is c.s-sz cv youv Lt2Z.4), FSAR s's havtg two 15;lation volves outside containment. However, this system is not classified as an ESF system, an ESF-related system or as being necessary for the safa shutdown of the plant. Accordirgly, justify the location of both volves outside containment 12 this line since only the types of systems cited above are allowed this configuration. dP_ E J F o W J E. I ! 1 I I f ' f 't ? t ' I e i - r t. l l l 4 I I e i i o I i ! i i ! i: i I M GJMClo d 2 G_C o_W 3 S bb-Ei._f 2 fE T u mare.s_ n.m.neceaw_As_.g' Am W.g_mM I A r, Ac4tmc2AT_E l : ; i i
i I
iiiiii,;i! :. e o D s-
l O .= ~ 1 m Table 6e2-25 s [ CONTAINME'IT ISOLATION VAIVE INFORMATION l \\ i 1B E jO vetoe Aet ett-I a a S I }r g"3 l e _.e t. _e w ee. colt o r l, ':,s j_ i o i 35 3 [ !i a 3 y ] l m1 ci 1: g .i e r l a b,1 _n, } 715 _. ! _I _a-. .= _= Il 12 143-e .Fot-5 3 ,e _ a n I B_i e_3 -1 V s s 3: = ,e SS in.t. u... tea. = T.e E.i.= .n-F0n. i . = .iehe .,ieten t.ee.et ut e 3-5 2 t, e ...F. = .te = Y.e E iS....n-F.2. O - O .ie.o -,i et. = t t t I t, ice SS sin. tee.u .to = T.e E-ion .u - F.2 = i .no .,ieten 0,s., met out e >-5 2 C = a.... . tea.. Tee E-il....n-F0a. O - O .ie.e ,iot. = o. t t. ut S i Inc SS feela Steamline A Steam 26 fee E*l02A B21-F012A 8 me Giapio Pneum. Pietem * = l .a Tee E=IS64.0 921-F020h O IIs O GInto Pneese. Plates epee 90 set Nenet tienet 3-5 2 se C.W.E.F.E. Steam 8.N.8uB.IEB Spee Seat shut that 8-S l se lac SS nota,eamline C Steam 26 Tee E 102A 928-t422C 3 IIe = Globe Pneten. Platen Spee SBout Mtset
Shot 35 2
es t*.S.it.F.3, = f' 6 i J.W.te.IBS st.o. = Tee E-IleA.a on-sekee O me.= Glene o.e. eletem
c, shot sh.t sh.t 35 3
g eo se SS ass on.a*
ust.r
-l/2 T.. E-10a.e on-reevo O no e Gate noter' st.et. men. two. %t se ee t 3.5 I C e.e.F.m. g D gas E-IS4S J.W. gen.We h gg) 10c SS nsL Drain a teater,4-l/2 Tee ett-Fe6ft O Ice e Gete suster elect, seen, open Shot WIset shot ?.S 8 te W lic SS stSL Drain A GBeter 3-l/2 tes 828-F06?A O ses 4 Cate gester Elect, eten. fWen SItest Shut shut T Ja l g N 3 82C SS 8 EEL Drean C Water 1-l/2 Tee 328-F067C O Its & Gate sister tiect. them, ftpon Seset phet phet 1.5 8 ee 13C SS sesL Drain aseter 3 Tee E-8029 328-F016 I IIe Cate gester Elect. soon, open glemt genet geset 15 2 se C. Bet.F.W. = 6 J.3,ps,gge 3 teater i E-1544.3 321-F019 O O Gate egetor Elect. pass. Open Shot steet M RS I to Q tseter 3 Fil-Fool O le Gate santor Elect. Isen. (Dee Shut Shot Slust IS 2 es 16C 55 Feed esoter Line A Water 20 Tee E 1020 f*t-F0B04 8 IIe Clemet = Self = Open Shot shot = Inst. = 3 = Water 20 923-F0324 0 0 Cheet P.et imi. Self = Open Shut Shut best.
I E-150 I?C SS Feed mater Line a efeter 20 Tee K-1020 021-F010e t IIe Closch self Open gemet Stunt
= = Inst. 1 = = Water 20 S28-Fol2e O O Cheet pnesse. Self Open Shut $het Inst. 3 = F-SSG I 2SC SS Centrol med Drive water 2 E-1044.C CII-F003 O Ian 6 Cete asetor tiect. Inon Open Shut shot Ao le le 1 3 55B esat er 2 6 Cll-FF255 3 = Gate teater Elect. Man. open shut Shot As le SG 2 8 888 Water 2 E-IS? Cll-Fl22 3 = rbot Self = Open Shut shut = Inst - I L 27C SS LPCI A 18ater 84 Tee K-107D E12-F027t O pas 2 Cate stater Elect. seen. Open Open As le 40 t t E.90.lat = 12 EI2 F0424 3 Gate 8 ester tiect. = Shot Shut open Ao le 24 I I C.E.L.U NB = 12 312-F0434 8 = Check Pneeen. Self Shut Shut Open = Inst = I 12 El2-F02A4 3 Gete Motor Elect. sheet shug As is 60 3 C.E.IWS = 14 El2-tol?A I CImbe Motor Elect. Shut Shot Shest Ao le 60 1 Ost 100 S6 IstA/tJCT 4 teater 24 Yes E-10?S E!2-F004A O Its O Gate Isotor Elect. Itan. Open attest Open As le te i O see ft ga gi. ww.gseet. meeres 14 64 sc.lg( sue 8er 1 de St u $@eis sinist seeise tlwir F+tl - = rest ) a a g a
m m I I s A _.Q Table 6e2-25 CONTAINMENT ISOLATION VALVE INFORMATION (Continued) L s 5E ai-e A Ic;
,, y b,
-.e . Pe.iu-I g g 3 -r - l .1 5 cI Lg 1= l 1 a.
y 5, n
e .n. 3i ns: er s-1 1y h: 3 :, he I 33 5 1 3 r8 -] e: h: I:e a. 3: bc . t o. a a a
:c i6, s6
- l .,,e t., .-i u. .>,-,0s, es. c te ,.,t.t . leo. .e.n. .,_n a A. to i, ......Ce..
t.,li,er
...r Ci,-ros. i c t. .ter uso. on., o,en t A. le i. ..E.. e.. ,,c ,6 m... r,e .er 4 .. m. C3> .se C.te ,s.t or ueo. on. t Ao le n 2 ....r...... s. tn c .ser u o n-r03. o o C.te . cot.r eien. ,e.n. ht As i. n i w ai 1191; 56 euCU Seckwash Water 2 No R-IS7 C14-FF23S S No Clette Pneum. Air shut Sluut Shot Shot 2 2 O 150 neeen 2 G36-fF218 0 1 Globe Peeum. Air thet shut Scout ht 2 3 O IWS E 2 C36-FF203 0 2 Clohe pneum. Air = Ehet Shot Shut Shut 2 i O IMO C gS) O it-56 es Paal Silves,r Watee le sto E-IISC C41-F029 0 uo 2 E.at e me.e Elect. seen, open tyesi Stout As to 29 e o S.E.V.gue g
w etn t. reet 30 G41-Pn44 3
Cate esot or Llost. seen. cy.n open shnt Ao le 39 2 o 3.E v. ape i T'at s=.eneratlred water to th R-ll%C C41-tv2e C 09 2 Clebe st.=t er Elect. seen, opes open shut Ao le 29 8 3 3.E.V.gue 'te y P
==
t. 9 i jo :n.
tD
3 33 10 C48-1040 8
heck
= Seit t).en opost Shot ht test O tvo ' s. PA Imol Drain t. ma*vr 12 Iso R-IISC C45 89884 I Tee - Gate faen. Nam.
3. hot ht shut shut L r.
O NN esa an. ne.de nser 32 C81-ett97 0 0 Cate
seen, stan.
= ht Shot Shut Shut L.C. O DN stM * ? f=t Pump suction water ,I,I 2 S. hut Ibn E-172 Gle-Frno) O see Cate peotos tiett. seen, open Open As le te 2 0 5. 8.858 G F.%' a* *a ). .at. h.t f.8 t. an. ..n o,.n t As... I 2 320C. $8. hMD Water f.eturn wat r e elo E-172 les 4 (tiect Self 3 U (ten upon 3het Shot Inst g Gin-e s. v *> 0 C.ete neutor Elett. tigen Open shut Ao le 40 t t 9.E.Ise yr $t, e>e tne r el s eed fA m a = E-324A P46-Filft2 1 Nr Gat e Poets. Air nye-a Open ht Shot 2 2 I" p.E ges Wates to Fem.V 4
omp te Panel water 2 R-IS7 P46-StOSS O I
Cate Pneus. Air open Op.ase Sher Shot 2 I I 9.E.gge 1 57C %S Aar ravitswe Att '/4 Tee E-ISE Pfal*Freet 0 18e O Cate sole. Elect. shot shut open shut gnet 2 3 ese pold 1/4 P68-Sto47 1 Chect Self sheet phet opeu open ggieg gnet I 8 list 5% Aer *rseltsve feel Air 3/4 see R-f14 P68-FF034 s sao t'hech Self senast ht Open Shut teet I to fit'tC.tean 3/4 P68-rt03SD o P 2 Glot e ,$.le-Elect. 9 shot shut it.es Ehet inet 2 S IBl '.'alve LSI-9001 8' eld d 3/4 P68-rF**003 O Clobe S.st e. Elect. ht Shut Upon Shut inet 2 8 note eE $6 fone s t nsace.t Att 42 No F-165A T41-l'9 014 9 Iso u t' fly Posesse. Air Open ripen 5%ist Shut 20 1 t B.E.99 Purge Eeg.gtv 42 R-ISS T414 rolS I D* fly Pneva. Air clen Open mhet Shut 20 2 i B.R.4ue M c= ' **. f.w t a ge o-o.L Ast 41 No E-!6SA T41-DFOOS I Hu = B' fly Pne.ne. Als = '4n n egen Shut shut 6 2 0 0,R.8st gM
weage I sh4.est 6
g pl 42 E-ISS Tel-rF0(t6 0 D* fly Peseuse. Air = Upen f1 peg ht Shut 6 1 t3 3.E. Pts 4 ql O: g .th J ' g e P-e a.
f 480.39,, In Section 6.2.5.2.2 cf your FSAR, you indicata that ta the drywell, (6,2.5)' the hydrogen monit';rtr.g points are located at various c1cvations along the inside cf the drywell wall. Indicate via drawings the location, including the elevation, of the hydrogen sampling points in the drywell and the containment. Discuss the instrumentation available to monitor the combustible gas control system under notsel and acci knt conditions. g E J t* o d I C THE COhITAIN HGNT ATHOS PH ERic f10MI TO RINC. S YS T & M (C A HS) DESIC.N, IMCLUDING THE H YO Ro&EM.5 A rlPLING POINTS,15 SY M THE APPLI CAN7. SGG S ECTIO N 7. S. I. l. 7 FOR DIS C US GIO N O F THE JMSTR U f*1CN TATioN AVAIL Ab LE To HONIToK THE CONS USTlDLS G AS CO N TR.CL SVS TEM \\ UNDER NORMAL lAND ACCIDENT CON DIT10NS. TNG APPLICAN T l<)lLL b2 REQUIRED TO COMPL Y, h d ^ E SEC434^1 hfP G5.1JC Q gll eq mgd gd 0 +o k c4 & C Cams oppbu4s nep~,Adh is
480.40 (6.2.51ig You state in Section 5.2.5.3.1 of your FSAR tha; there will be no hydre-gen generation res31ti::g from spray water contact c:ith aluminua or zinc cosponents becausa the containment spray watcr contains no additives. We have determined that Iqydrogen release resulting from corrosion of zinc and aluminum following a postulated LOCA should be ' considered in the analysis of hydrogen production and accumulation within the contain-ment (Refer to Regulatory Guide 1.7.)' Accordingly, provide the following additional information: The corrosion rate as a function of tesperature for all materials a. in.the containment which could become a source of hydrogen due to - 2 corrosion. We have endorsed the following equations for estimating hydrogen generation as a function of temperature due to the inter-action of peutral and alkaline water spray with zinc: Neutral Water (pH=7) e 1450 e H = 4.6x105 i 2 EXP (-E scf/ft hr Alkaline Water (pH = 9.5) f' 1450 I H = 1.3x105 2 EXP (- RT ) scf/ft -hr ( where: R= 1.986 cal /*K-gram T= degrees Kelvin b. Describe how you established the corrosion rates for the various materials in the containment. In responding, identify the experi-mental data used as your basis, including appropriate references. Discuss the conservatism in using this data in view of the environmental conditions which are expected foJ1owing a LOCA. Graphically show the hydrown concentration inside the con-c. tainnent(Es a function of RImIe) with no recombiners operating, with one PRDutt ? operating (minimum engineered safety features) and wi both recombiners operating.,,, 7 d. Graphically show the contribution of each sollrce to the total hydrogen generated as a function of time. Provide the mass and surface area of zine paint and galvanized. steel and other corrodible material in both the drywell and the wetwell for our confirmatory analysis. / p
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wwwe-svc.xi F GESSAR II 22A7007 dF 238 NUCLEAR ISLAND Rcv. 0 j 6.2.5.3.1.2 Long-Term Hydrogen Generation (Continued) (( within a radiation field, temperature of the water and concentration of the fission product gas as well as the nature of the radiation. The analysis is based upon Regulatory Guide 1.7, Rev. 2, hydrogen yield rate of 0.5 molecules per 100 eV deposited. 1 t g +t 6.2.5.3.2 Analysis A post-LOCA hydrogen concentration analysis has been performed. '1 The following is a brief description of the post-LOCA conditions and explains the basis for hydrogen generation ahd control. A complete description of the post-LOCA condition is found in, Section 6.3. (1) Double-ended break of a recirculation line coupled with the worst single failure Two RHR pumps, one RHR heat exchanger and HPCS are still in operation, g4 Drywell Pressurization causes non condensibles in the drywell to be purged through the suppression pool vents and released to the containment atmosphere. (3) Following release of primary coolant, the maximum postu-lated zirconium-water reaction produces hydrogen until a total of 0.72% of the core cladding has reacted in 2 min. rission products postulated to be released from the per-forated fuel rods are assumed to be homogeneously dis-( tributed in the primary coolant and suppression pool water. 6.2-131
GEOSAR II 22A7007 ,238 NUCLEAR ISLAND rov. 0 { 6.2.5.3.2 Analysis (Continued) r- +r he tu (5) Aluminum corrosion rate er 0. (6) Zinccorrosionrateh0. (7) Mass of zircaloy fuel cladding = 82,716 lb. {. (8) Reactor core thermal power rating = 3,651 MW. (9) Operating history of core = 2 years. (10) Reactor decay heat profiles, fission product release ed rates and the fractions of fission product energies ~ absorbed the the coolant are consistent with ANS 5.1 draft standard and Regulatory Guide 1.7 (See Subsec-tion 6.2.5.3.1.2 for discussion.) No rate curves are provided, since no deviation has been made from the above references. As shown in Figures 6.2-63 and 6.2-64, the hydrogen concentrations in both cases can be maintained below the 4% limit. Therefore, the mixer parameters and the assumed hydrogen recombiner operating parameters are adequate in limiting the hydrogen concentration below the flm=mability limit. 3
{L 6.2.5.3.3 Controlled Purge Site Dose
+ In the event of a design basis LOCA, the redundant recombiner sub-systems of the Combustible Gas Control System will operate to con-trol combustible gas concentration within the containment so that backup purge subsystem operation will not be required. No person-( nel containment entry is required to service any of the equipment required for long-term shutdown. 6.2-133
'1 !i a 0sn uw>$3 u$znMsu98 wo#. o g oo KN o t g g e n \\ eg \\ or \\ d A y) / l e f m / ti nT 0 / I 0 e t mn sD / no ii at / ta ni ot / Cin 9 1I / em su / am Ci o n /I ti t nM e( / isr nh )r h a ( r1 E T MI = T noe i m ti aT / r I i tnno ei ct na oi T Ct N i EM nn N eI l l t E I g A f' W T or Y N re R O dx D C yi W l 1 M I I o 3 6 2 6 e r ug i i 5 4 3 c P o o
5$E*
l e m$* l Ill ll tlll lI
/* ~ /h A, 5 0 V 40 - ORYWELL - - CONTAINMENT Y 3.C k / \\ c M 20 e / / \\ 8e ~~ L / f;a O 0 1.0 MM HH / 8 __ s " / 00 tl e ,,,l j -1.0 t I f f It , e 8 I 8 8 I ffI e e ee,ae 0 01 0,3 g 80 100 soon TIME thel M
o u Figure 6.2-64.
e> Ilydrogen Concentration Transient Case 2 Containment flydrogen Mixer Initiation Time = 9.45 hr (Maximum Initiation Time) o ,1 W
e .. ~ ' '.. r s... GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 i 6.2.8 References (Continued) ,6.. 6. W. J. Bilanin', "The General Electric Mark III Pressure Suppressi.on Containment Analytical Model", June 1974, (NEDO-20533). 7.
d.
J. Bilanin, "The General Electric Mark III Pressure Suppression Containment Analytical Model", Supplement 1, September 1975 (NEDO-20533-1). 8. E. H. Van Zylstra, "Drywell Integrity Study: Investigation of Potential Cracking for BWR 6/ Mark III Containment", /f August 1973 (NEDO-10977). . s. 9, M.64. O s g -. ?vsswwww.%s. Gao.oeu oc4 AuoNoK., p. 42M. C17, J. Ms.ur.1 t L4 ; 5 LO, S. M h% M R.Noo na t hr.,vod.. " Cocs.s.es.ged 17.ac.5.g srMg op MU^ b* Ad> AMM y. IFb 16 9, 99%gp1E.cr -7.stewo futwWu Wg Certe. l 'W 6 ( 1 i s t t 6.2-160
480 41 The _F_ederal Register, dated Decoster 23, 1981 contaics a pro 2 (0.2.5) cntitled. " Int;rta Requirements Related to Hydrogen Control.gosed rula This proposed rule contains en addition to Section 50.44 of 10 CFR Part 50 which pertains to the Mark III containment. The proposed rule would require th6 Mark III containment in your proposed nuclear island to have h drogen control systems capable of accoasnodating an enount of hydrogen equivalent to that generated from the reaction of 75 percent of the fuel cladding with water, without the loss of containment integrity. The progen ignition system (HIS) is an exagle of a system which has been ' proposed by Mark III owners to satisfy the provisions of this proposed rule. Discuss'the provisions you propose to satisfy this ' Interim Requirements" rule for degraded core accidents. Compare your proposed provisions for the GESSAR-II 238 Nuclear Island with the cogarable provisions for the Grand Gulf. Perry, and Clinton facilities. 3c C. S P6 d E E-4C jlo & lcov (ro.btf Of L e } FA N cm emt u, He G 6 s w '1 G ~"^ h*d E ' E! . J., l Ok/ 4 m c/ iMr /h 4erMd atbtlik of & prenure . r.< p.-e s.r m p o o / is sceu 6 ea (Erreo, p,o A e.f.r- & <e h e e J f-o e core g.s no d droy % i) N 'b * ^ I b/ )'t
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$46 ili o h 480.42 Provide an analysis which shows f. hat the hydrogen mixing system will, (6.2.5) within the containment or within a containment subcompartm This analysis should also show that combustible gases will not accumulate within any cocpartment or cubicle to form a combustible mixtur~e and that the containment internal structures have been designed to promote free circulation of the containment atmosphere. to permit us to determine your conformance with Sections 50.44 and 50.4 .of 10 CFR Part 50. ~~% Bw2 6 flas h z5r .ry sk ir oGi}~g -/o s<,Ce/p d e -z u <. -/xe 99m ye.r w~vn;9 c%&n 4w 's acac4-Q:.r. /de sn /y .En. -- u-c%Q, Ava &d%/ w/ <cw% is yy-ecau'e r 4 J.,y m 9 i.s
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n 480,43 (s.L6) j Provide the following additional information regarding the venttg and { drating for the Type A containment leak rat] test:, For each system penetrating containment, discuss your proposed a. venting and draining provisions. Systems which are not designed to remain intact following a LOCA should have the isolation valves exposed to the containment atmosphere to permit the test differen-tial pressure to be applied across them; 1.e., the system should be vented and drained both upstream and downstream of the isolation valves. Provide justification for each system penetrating contain-ment which is not vented and drained. b. Identify any gas-filled lines which will not be vented for the Type A test. Provide justification for not doing so. Q E. S P 6 d S E. - 1'T Revised FSAR Table 6.2,G includes an itemized listing of systems penetrating the containment. Provisions for venting and draining during Type A testing are discussed in rri;4 subsection 6.2.6.14. Those systems penetrating the containment that are not vented and drained during Type A testing are discussed below. All other systems will be vcnt:" and drained. RcMap Teesu 4.2.-2.4 is W CME.P Systems penetrating the containment that are not designed a. to remain intact following a LOCA, but are required to-pnin+ M e-punt 4 2 r2 % c.hutdcxn-condit-i1m during 1.he Type A test. These systems will be Type C tested and the results added to the upper confidence limit (UCL) of the Type A test leakage rate (Lam). The systerav ia-listed below: (* 5 'p>y ' ................, i i i i .. D. "M M_ =a=rChilled Water Systems ( 'Dwtwru C4wac N'1 T "- t 7. v-i ' Thos'e systems associated with note 4 do not have to i I be vented and drained for the Type A test g WW ~ & t - they meet the following requirements: ! 1 I _1 i 1) The system is protected against missiles and pipe whip. ' 2) The system is designated seismic Category I. ~ ~ lg j j 3) The system is classified Safety Class 2. i,., I ' !.) The system pressurc is grec.ter than the contain-2 I ' ment pressure at all times during the course of [;[j;g the accident. j,,5) The system will remain full of water for 30 days. l 7.- ' 6') Both items 4 and 5 will be maintained when a ' I ' ' single active failure is assumed in the system. I M, ! MLI._i,, e4ASiF.n LCO._.!M9.n7tu:bN1L L_Lt3E' DrMnQ _Ae4 i i e i ! I ' f I I ! I I I f,_ij !
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480244 Section III.C.1 of Appendix J to 10 CFR Part 50 prescribed methods fcr (5.2.5) conducttg the containment is;1ation valve leak rato t:sts. Th:so req 11rements state that containment isslation valves should be leak l tested with the test pressure applied in the same direction the valve must function to preclude leakage in an accident condition. In Table 5.2 29 of the FSAR. you list butterfly valves such as the containment vacuum relief outlet velves, which are tested in the reverse direction. Justify that testing in this manner will result in equivalent or more conservative results compared to those resulting when the test pressure i is applied in the direction of potential flow following an accident. (Table 6.2-29 lists a " note (4)* which indicates certain globe valves are tested in the reverse direction. However, no globe valves are listed in the table with " note (4)." Clarify this apparent inconsistency). (L E. S /* b N S E.- .ii..iie i ',- i:,. ! M oTICll_.Lp.55___ 6,ft L4 _ild."Lf_1T_ S Mo %D_.B 8-- l ! 1? ELLE.TSlUP. 9t4c til Abb._WWP 5.E.dc E J~__ 'W1hPo).Z. At,t.hAQ._0%And. 51~eMi L.L4_6 O_ I!l!M B i ! !I ' AN o Eff-Fo P___E.n! _EO_7(.IF.c.ic._5 r5Le_m_.>_. ! i i i i i'I i! i I I I i 1 I I ! I ~ I. 4tio tut tu.. : qs,,e na We i i l III I 'I I l 5_,JLdt=' careec rte +4_o ey ; n11r,_y.a.w.g > 2 m r. r i T-vds.no.4 -m Wettaxwh_Lnsya.f_W._r.y-- ! !i i I.I I I I I I ! h e _TM E C ct4TAl t4 t=%t.t 4*C.A %ts E_6.1.-- 2,3 - l l8 L' I i "3 4W%e.wp Sf,,E__W_W.mc.Q _Ja_.wowi D4 AT - i j N m g,,._$ : M *t'Ltfc 5 : To_.32.1-FoJZ A r5/.,p,,_ K'! I i i MA tel 5"temAdovat30M YAYEf _At49_ _f dv 'tOof !?M_cKtc.a.L.! f w__.2 d sr cru.c_.5s _vALvf f ' F 9 - W QAgq_f.Ti.Eq74.LEMC, W,.,,/GS. _ '_h t re:c_ m 3.e_n1.o C ! ALL_A AEL4bQMW.__hME Suji!Xc !E.75~ SfL-EC_Q ' thfEth'Cle hP ft ' Mutrat_15t._A_ gar _1EL_ vhME Fog:_Qt eW. ! 4C4__.th4040fa__erf _VA1.d3ES ' %L9 E.Sj "T'8Mftt& _.thL"ChE_V.G'_nUzh 6 ' ' A c'- 'ac A:rt.pJE9Th.EA% Q\\tzEc.Jrinct 9tCubp_.it4 21C.A'GE:.e_.tacrJ.E. AtgSAlf_EfoVNrWF _Ff.541.f!J Krarg,._ h,.vuy -rd. t I I c.off,EmfATyJ_CJfj Ltrn g, u t /eet'tc LDxx 4,Li W.r v G M V 1 t73a A L,s.1_1.e4 W.fX9 Q F N t- @ P '-'.; F *_ 3hvF a t N viArt.i\\ e,S m-sr; ! I i Jl i !I rG Aqs Tb bit:r vw Plot.' l ! I i mesa st:me. i Vi ! I i 2 ! iil I ' I I I ' l i i I i ! i i I 5 i i i i ! ! % s- 'etm< gp: tugwe. b. I : I i ! ! I i ! I I
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6 i i i i l l j j { ~, ~ ' ~ ' l ' '.9 ' i - i i cb4 Ed5 ICn5E TO yh.)C.5"Tt,vh 480.% F0(L ocR TestTtdh DJTETLPRETNT10*J AM'.) ;J:" W.^T op: TYFG C TESTm M APARO 'To rw i.C LEAKosE ConTv'tn Tg5TGEM S..
-.~. r f*[ p r Tcble 6.2-29 CONTAINMENT PENETRATION AND CONTAINMENT XSOLATION VALVE LEAKAGE RATE TEST LIST (Continued) l l Inboard Isolation Outboard Isolation Barrier Barrier Penetration Bellows Test Barrier Description / Barrier Description / d Number Description Seal Type Valve Numoer Notes Valve Number Notes 2c Equipnent flatch No B Double O-Ring 1 3c Personnel !@ck - Iower Inner Door No B Double Gasket 1 ~ Outer Door No B Double Gasket 1 Barrel No B Inner Door 2 Outer Door 2 u t.s 4c Personnel Lock - Upper co Inner Door No B Double Gasket 1 2 Outer Door No B 1 Double Gasket m Barrel No B Inner Door 2 Outer Door 2 Q a l> o" 8c Fuel Transfer Tube No B Double 0-Ring 1 W 9c Main Steamline D Yes C B21F022D 3,4 B21F028D 3 g~ Yes C B21F067D 3 Q Yes B21F086D 3 0 10c Main Steamline B Yes C B21F022B 3,$ B21F028B 3 k Ye8 C B21F067B 3 Yes C B21F006 3 lle Main Steamlina A Yes C B21F022A 3i B?lP028A 3 j B21PO67A 3 B21F086 3 12c Main Steamline C Yes C B21F022C 3$ B21F028C 3 j B21F067C 3 g B21FO86 3 e O ( "T*g6 k DI At.SO TLEJX?%7iG-U
I \\ ig gy .) e Table 6.2-29 CONTAINMI:NT Pl: aETRlYT N AND CONTAltP4ENT ISOLATION VAIVE LEAKAGE RATE TEST,IST (Continued) Inboard Isilation Outboard Isolation Barrior Barrier Penetration Bellows Test Barrier Description / Barrier Description / Number Description Seal Typet Valve Ntaber Notes Valve Ntaber Notes + 13c Steam Isolation Valve Yes C B21F016 3 B21F019 3 Drain B21F005 3 16c Feedwater Line Yes C B21F010A SjL B21F032A S lb J j B21F065A 5] ku 5 B21F102A ou ca 17c Feedwater Line Yes C B21F010B S lG B21F0325 5 )g j B21F065B 5 C 2 O B21FF1028 5 't T 4 25c CRD Pump Discharge Yes C CllFF215 6 C11F003 6 Y u' C11F122 6 tn 27c RilR System (LPCI Mode) Yes CM E12F042A j f 3;G,1G E12F027A p,1 G l Line A Division 1 J o 28c RllR System (LPCI Mode) Yes C 44-E12F042B JM E12F0278
~~30 i~~
Line B Division 2 29c RifR System (LPCI Hode) Yes C N E12F041C [274 E12F042C j 43 Line C Division 2 30c RilR A Pump Suction No C, M' Closed System R E12F004A ! 6-Division 1 l 31c R11R D Pump Suction No C ff Closed System T. E12F004B fA N Division 2 e <J 32c RilR C Pump Suction No C ** Closed System $J E12F105 I4 1 r O Division 2 og 4 Tf FC A Mar A L5o 1F.EQuin.gp t >dbE% McE9 c>mf.tr Wg h f7 tMm u.a. W =-w rr.va.. -a n.~.... 'I'. e---- A
m. f e. - g g, Tablo 6 2-29 CONTAINMENT PENETRATION AND CONTAINMENT ISOLATION, VALVE LEAKAGE RATE TEST LIST (Continued) Inboard Isolation outboard Isolation Barrier Barrier Penetration Bellows Test Barrier Description / Barrier Description / Number Description Seal Typet Valve Number Notes Valve Ntsaber Notes 33c RIIR Pump A Discharge No C eg Closed System 'r If* E12F024A y IG' I to Suppression Pool E12F011A E12F064A I \\ b 34c RilR Pump B Discharge No Q U Closed System ' (C E12F0245 h to Suppression Pool E12F0llB. j E12F0648
~~U co 35c RIIR Pump C Discharge No OM Closed System~~
' ' lGP E12F021 g2 to Suppression Pool Closed System 1 ik E12F064C g 36c Demin Water to G33-2020 No C/ P46FF182 6t P46FF055 6 j Om RWCU Sample Panel N I (A 37c RilR SRV FOSSA & FOSSC Yes Q Mt Closed System 1 16' E12F055A $ 11 16 y to Suppression Pool Closed System ~ E12F055C l' 1 1 2: D ( l 39c RilR SRV F055B & F055D Yes C 4 4-Closed System 9 E12F0558 il to Suppression Pool Closed System 1r E12F055D 1 1 40c RilR SRV F036 No p 5 (- Closed System ~ E12F036 l 1 to Suppression Pool 4u 41c RilR SRV F005, 5'017A Yes C $r $- Closed System 7 E12F005 Il and FF236 to E12F0177. 11 Suppression Pool E12F236 s i l. 42c Ri!R SRV FlOl, F025C Yes C N I' Closed System 'F.4 E12F101 I w to Suppression Pool E12F025C h1 v. sg , i g 4 TvFC A 1rtc+r At 5o e-coru.a.ee oe o s t.c 55, N W 9 NEA W:4-Walr hair 43, WTEfty FwcEP Durtir% h pc., Tco-- V.j J
..w , m w, m gm.3 pT Table 6.2-29 CONTAINMENT PENETRATION AND CONTAINMENT ISOLATION VAINE LEAKAGE RATE TEST I,IST (Continued) Inboard Isolation Outboard Isolation Barrier Barrier Penetration Bellows Test Barrier Description / Barrier Description / 4 Number Description Seal Type Valve Number
~~Notes Valve Number Notes U uppression Pool I,15 E12F017B 9~~
43c RilR SRV F0$7B, F030 Yes C S to Suppression Pool E12F030 ,13 44c R!lR Suction from Recirc Yes C 4'4' E12F009 h E12F008 y 47c Steam to RCIC Turbine Yes C E51F063 5,4 E51F064 5 Est Fo76 Sj4 0 40c RCIC Pump Discharge Yes C tt-Closed System 1,44 E51F065 f/ il, /4 -v "' Ilead Spray Closed System 'l E51'Ol3 lj i .y 49c RCIC Pump Suction No C 40~ 40 M - Closed System 7 E51F031 From Suppression Pool ,1i V $' { 50c Turbine Discharge to Yes C Closed System E51F068 1 i Suppression Pool j $U 52c RCIC Pump Ninimum No C 4$ Closed System d E51F019 i l U Flow Dypass / ( e 53c LPCS Pump Discharge Yes C %* E12F006 ki l E21F005 l l 54c LPCS Pump Suction No C d 4' Closed System ~ jf E21F001 1 55c LPCS Pump Test No C n'4 Closed System E21F012 / E Closed System E21F0ll fI 56c LPCS SPV Discharge to No C A8i Closed System E21F008 tt Suppression Pool Closed System 1 E21F031 11
~~n U ew 57c Air Positive imal to No C~~
~~Closed System f-- P61FF046 <J if~~
~~C P51FF010 & P51FF040~~
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'f.,. g n ', ^ Table 6.2-29 CONTAINMENT PENETRATION AND CONTAINMENT ISOLATION VALVE LEAKAGE RATE TEST LIST (Continued) Inboard IsoleM on Outboard Isolation Barrier Barrier Penetration Bellows Test Barrier Description / Barrier Description / Number Description Seal Typed Valve Humber Notes Valve Nusber Notes fp 58c IIPCS Pump Discharge Yes (, $ $ E22F005 bl E22F004 59c IIPCS Pump Section No C, O Closed System 'l4 E22F015 A 60c IIPCS SRV Discharge to No C (,4 Closed System F E22F023, E22F012 4 Suppression Pool No qg Closed System '.J C E22F014, E22F035 y ] ta h9 63c RWCU Pump Suction Yes C G33F001 G33F004 from Recirc g m O 64c RWCU Return to Yes C G33F040 5,10 G33F039 5,10 N 4 Feedwater I,ine g 65c RWCU Discharge to Yes C G33F028 6 G33F034 6 Main Condenser p Z U 68c Containment Supply No T41r'F015 5 T41FF014 5 Purge IIVAC 69c Containment Exhaust No O T41FF005 5 T41FF006 5 IIVAC f,42 70c Containment Vacuum No ,C T41FF034A [ T41FF033A Relief Outlet ((* 72c Containment Vacuum No ,C T41FF034B f* T41FF033B Relief Outlet 78c Skimmer Drain No C G41F044 6 G41F029 6 to PPCC W he't! A TEF AL M t*4Aump o 4 uru.ess H.vce o m a.w s,c. p e FCwe > w m e-n u. o osci % T m c.4 w y p L______
g f.jp ^ i Table ( 2-29 COtiTA I Nr 't. T 11:
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! AND CD TAIT, Fir.' ISOLl ION VM E LEAKAGE RATE 'I EST IST (Con tintu d ) s.nboard I. elation Cotboo Isolation , Bai r..ir i rrier Pen itrat ion: Belloas Test Barrier Description / Barrier description / thmbe t Descrap. ion Seal Type
~~Valve Number Notes Valve Number Notes 79c Demineralizpr to FPCC No C~~
G41F040 12 G41F028 12 83c 24-in. Pipe Spare No Capped 84cy Instrument Line Yes Capped u 84c2 Spare Yes Capped ~ y hc 84c3 Spare Yes Capped m O tt M 84c4 Spare Yes Capped N m Ash M ll4c Drywell CRW Sump to No C P55FF0ll 6 P55FF012 6 gg Clean Radwaste in H E 115c Drywell CRW Sump to No C, P56FF036 6 P56FF038 6 g Dirty Radwaste ll6c 12-in. Pipe Spare No Capped 117c 12-in. Pipe Spare No Capped Il8c 24-in. Pipe Spare No Capped Il9c RWCU Backwash Drain No C G36FF235 6 G36FF238, G36FF203 6 120c CCW to Containment No C P42FF135 6 P42FF010 6 u
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121c CCW Return from No C P42FF042 6 P42FF021 6 <4 Containment ~ $ T4VE. A-rw ess 7o stee.,sase
f ~ w ~ rQ ^ Table 6.2-29 CONTAINMENT PENETRATION AND CONTAINMENT ISOLATION-VALVE LEAKAGE RATE TEST LIST (C:ntinu;d) Inboard Isolation Outboard Isolation Enrrjor Barrier Penetration Bellows Test Barrier Description / Barrier Description / Number Description Seal Typen Valve Number Notes Valve Number Notes 124c 12-in. Pipe Spare No Capped 125c RI Chilled Water to No C (,& P39FF050 12 P39FF049 12 I containment M 126c RI Chilled Nater from No C#4 P39FF051 6 P39FF052 6 w Containment 127c Condensate Dist. to No C P46FF183 fW P46FF062 p = Containment C 3 M b 128c 3/4-in. Pressure-Yes B Double O-Ring 1 4 Sensing Line y tn 128c Spare Yes Capped 2 129c Service Air Dist. No C P51FF010 5 P51FFOO8 5 0 130c Instrument Air Dist. No C PS2FF040 5 PS2FF038 5 131c ADS Pneumatic Supply No C P53FF017B 3l3 P53FF015B M Division 2 135c 2-in. Spare No Capped 136c 2-in. Spare No Capped 137c 24-in. Spare No Capped 142c Chilled Nater to No C, $g P44FF005 12 P44FF004 12 h Drywell Coolers
~~TMPE A ren Auo stravige h~~
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Ps: f\\} Tchle (. 2-2 9 CONTAINMisNT PE ETlWr10N AND CONTAltr4ENT ISOl> TION VAI VE LEAKAGE RATE TEST.IST (Continued) Inboard I:olation Outboa*d Isolation Barracr llarrier Penetration Dellows Test Barrier Description / Barrier Description / tiumber Description Seal Type + Valve Number Notes Valve Number Notes 143c Chilled Water from No ( MW I4 NEON
~~744 f%nT Drywell Coilers~~
=
+ro ti 145c ESW Supply to li y g Yes C P41FF169 h Mixing. Blower (Div 1) 145c ESW Return from If Yes C P41FF172 P41FF171 [ y 2 2 u Mixing Blower (Div 1) f 146c 24-in. Pipe Spare No Capped b O 147c 12-in. Pipe Spare No Capped 140c ADS Pneumatic Supply No P53FFOl7A Division 1 r(*2;B PS3FF015A bdL $y 156c Spare No 2: U 157c 150c ESW Supply to Il No C P41FFil5 P41FF114 g Mixing Blower (31,23 160c Air to RCIC Turbine No Exhaust Line 5 E51F078, E51F077 5 164c RWCU Pump to Filter Yes C C35F053
{(
G33F054 jg Demineralizer g 165c ESW Return from No C P41FFil7 Mixing Dlower (Div 2) P41FFil6 [ O AMF6 M ten Ab>O ###8"\\K89 v m> norse oma~use t g t. TRE.etAtrts \\O6Mit-F t ( L.E p Dus2 ggg Typg,4, 3 gsy, I,.
A
~ g. y^ Tablo 6.2-29 CONTAINMENT PENETRATION AND CONTAINMENT ISOLATION VALVE LEAKAGE RATE TEST LIST (Continued) Inboard Isolation Outboard Isolation Barrier Barrier Penetration Bellows Test Barrier Description / Barrier Description / I Humber Description Seal Type Valve Number Notes Valve Number Notes 166c Drywell Pressure Yes C T41FF035 T41PPO36 Bleedoff C T41FF050 T41FF051 w 168c Upper Containment Pool No G41FF186 12 C41FF187 12 to Main Condenser M h 178c Air Positive Seal to No Closed System P61FFOO6B so E51F063 5 189c Annulus Pressure-No B Double 0-Ring 1 M Sensing Instrument b Line U H 190c Annulis Pressure-No B Double 0-Ring 1 Sensing Instrument Line ts 195c Containment Test Connection 197c1 Spare Yes Capped 197c Spare Yes Capped y 197c3 Spare Yet Capped 197c Spare Yes Capped 4 Wn 190c Spare Yes Capped 3 O 198c2 Spare Yes Capped 4 hFC d W+T As.so 1rt s'm o.sce p 8
m t%, ), , *^ Table 6.2-29 ~ CONTAINMENT PENETRATION AND CONTAINMENT ISOI.ATI'ON VALVE I.I'.AKAGE RATE TEST LIST (Continued) Inboard Isolation Outhorird Isolation Barrier Barrier Penetration Bellows Test Barrier Description / Barrier Description / T Number Description Seal Type Valve Number Notes Valve Nasaber Notes 198c Spare (Div42 Only) Yes Capped 3 198c4 Spare (Div 2 Only) Yes Capped 200c Hondiv Medium Voltage No B Double 0-Rings 14 Power M Lap 201c Nondiv Medium Voltage No B Double 0-Rings 14 Power g 202c Nondivisional Control No B Double O-Rings 14 [ 203c Div 1 Instrumentation No B Double O-Rings 14 M 204c Div 1 control No B Double 0-Rings 14 y 2 205c Div 1 Containment-No B Double O-Rings 14 Scram Solenoid 207c Spare No Capped 200c Spare No Capped 209c Sparo No Capped 210c Div 2 Neutron Monitor No B Double O-Rings 14 211c Div 2 Instrumentation No B Double O-Rings 14 Afn4 212c Div 2 Instrumentation No B Double O-Rings 14 o. ye A Tew At.so moino
n
~~J'88 't gg.~~
Table 6.2-29 I CONTAINMENT PENETRATION AND CONTAINMENT ISOLATION. VALVE LEAKAGE RATE TEST LIST (Continued) Inboard Isolation Outboard Isolation Barrier Barrier Penetration Bellows Test Barrier Description / Barrier Description / Y Number Description Seal Typo Valve Number Notes Valve Number Notes 213c Div 2 Control No B Double O-Rings 14 214c Div 2 Contginment-No D Double O-Rings 14 Scram Solenoid 215c ILRT Nondiv Instrument No B Double O-Rings 14 w 216c Div 4 Instrument, No B Double O-Itings 14 Scram Solenoid Control g <n O y 217c Spare No Capped N 210c Div 1 Neutron No 8 Double O-Rings 14 Monitoring N 219c Div 1 Neutron No B Double O-Rings 14 Z Monitoring U 220c Div 1 Neutron No B Double O-Rings 14 Monitoring 221c Div 3 Neutron No B Double.O-Rings 14 Monitoring 222c Div 3 Instrument No B Double 0-Rings 14 Scram Solenoid 223c Div 2 Low-Voltage No B Double O-Rings 14 g Power b E. TG W AbM N.AeLEo
,3 s ,fM. ?% (i. f Table 1.2-29 CONTA 1 '4 it:NT I 1ETL )N AND '!aNTA:!M!!NT ISO,iTION V \\LVE LEAKAGE RATI: TES' LIST (Con t i nnerl' Inboari solat.fon Outh..rd Inolation Baialer . tarrier Ponetrati< i Bellows Test Harrier Description / Barrier Description / Numbet Deu.rit tion . Seal 3 pet Valve Number Notes Valve Number Notes 224c Div 2 Low-yoltage No B Double O-Rings 14 Power 225c Spare No Capped 226c Div 4 Neutron No B Double O-Rings 14 Monitoring 227c Nondiv Control No B Double 0-Rings 14 h> m o M 228c Nondiv Low Voltage No B Double 0-Rings 14 i w Power V i tn 229c Nandiv Instrumentation No B Double O-Rings 14 H un E 230c Nondiv Control No B Double O-Rings 14 g 231c Nondiv Control No B Double 0-Rings 14 232c Nondiv Low-Voltage No B Double O-Rings 14 Power 233c Nondiv Low Voltage No B Double O-Rings 14 Power 234c Nondiv Low-Voltago No B Double O-Rings 14 Power gb bouble O-Rings 14 237c Division No B o, Instrumentation o.
~~Ttec A Tor A'50 ftEavin o~~
U% ^ A3 Table 6.2-29 j CONTAINMENT PENETRATION AND CONTAINMENT ISOLATION VALVE LEAKAGE RATE TEST LIST (Contin P l Inboard Isolation Outboard Isolation { Barrier Barrier I Penetration Bellows Test Barrier Description / Barrier Description / Number-Description S:=al _ Type" Valve Number Notes Valve Ntsaber Notes l .,2190 Nondiv~ Instrumentation No B Double O-Rings 14 240c' Nondiv Ici@ltige No B Double O-Rings 14 Power '/ ) 241c Nondiv Low-VoltJ@d N7 B Double O-Rings 14 Power /< ' ) /- u M 243c Hondiv Communication Wo B DoubleO-Rinh4 14 g2 j m j g 244c Nondiv InstYumentation-No-/ B Double 0-Rings 14 N u %o, P 319c Suction Line t-,SPCU N C Closed Systra 4 G38M 003 Jr i
~~h. y pwny -~~
~~/ r -320c Stc.t Raturn;Eh,,[ r No C Closed System,' I G38NOO2 y ~' i SuppternioA Pool ', e / i s 4 A M g k TEST AbSD 94ith.as M f) e ,f~~
~~ar-4~~
= i vf g i ~ r e T a { p t s l ', %M + e 3* 1 4 cd] 's 1. f j ': o e o l >+ oa s f: e f .*/ f / l l ,/ T ( b h
F; 238 NUCLEAR ISLAND .m dMMM ) Rav. O i m Table 6.2-29 ( CONTAINMENT PENETRATION AND CONTAINMENT ISOLATION VALVE LEAKAGE r RATE TEST LIST (Continued) { ~ NOTES: (1) Penetration is sealed by a blind flange or door with double 0-ring or double gasket seals. The seals are leakage rate Lested by pressurizing between the 0-rings or gaskets. (2) The personnel air lock volume is pressurized to primary con-tainment peak accident pressure and tested periodically as described in Chapter 16. During the airlock test, tiedowns ,(_ are installed on the inner door, since normal locking mech-anism are not designed to withstand a differential pressure across the door in the reverse direction in excess of 5 psig. Pressurizing the lock barrel also tests the lock mechanical and electrical penetrations. (3) W'/F C tc S ' r - et rc q* d due.- te pro::scior. --' the 4To.% p W ": M are N % elrt:. Type A testing is performed .m a conventional manner with positive leakage control systems deactivated to prevent seal media in leakage from marring the test resu g,,, %, ,...u,,,,,,,,g,, .. u c r..., ,- ~ _g . jcm.sc v44.v ts *Asnes ew seas ev,s a g,.u.c res e p< assn 4 nwas ro wa r Au s 8'***1W (4) ':t ' ' = ~1ssHrad!-~"m er_ [_- (5 ) Qes-wn.uves.> 4.rus s.,rvro m t 4 E TF-E V E WC##8 The isolation valve is sealed with Air Positive Leakage Con-trol System. Th: -2 z i m ca s '=n-Tg ; ? Mrt. 4to.q, (6) The isolation valve is sealed with Water Positive Leakage ' y, control System. h 9 mv? mt 4re.1 Typc C t; z. see. R est 3. g Tvt. C itsT WAu Ot. vwneco vhm yherhter_ of M5wt. : 1,)c q p (7) se 1 p o ts' e nta;'t. t f(CL C) kreexef.pt komp
~~'--9 0,~~
e 1 J la. .ve yp Ctes osc .gi. sm' ai Itd ver'fn piac pir ocic T e st ad in sys em o ratgon tg tii .i (8) T is valve is exemp from OCRF50 Type C t t, rei is n 4 1, a e.losed-op outs'd pri ry . ta ' nt d e ina+e .s idside gounda y f s c nd r3 co t i er: E I N ho N - ...p s on ists o ali r p C as.a iren io ) aM he e f I 5 is.. c Ca tegor' I ipi g, ic L s all e e ularly examinedj a s.ecifj ed i the' AS E o e, e ti n XI -SepvieIasecpf
~~biu.~~
T1 h ystem 4 fu tion =1 y sted/ e ular Q sis. na i)/) i p th a a \\sdp h 4 (10) The RWCU return to *.e feedwater line is inbound of the feed-s s O water valves sealed by the Air Positive Leakage Control System. 6.2-217 s
esb1Rrlof 22A7007 238 NUCLEAR ISLAND Rov. 0 l Table 6.2-29 k CONTAINMENT PENETRATION AND CONTAINMENT ISOLATION VALVE LEAKAGE r RATE TEST LIST (Continued) ((11) htE.T1E.D h ttc :1 _s v~ 7
~~rted i_ re'7rrre dircctic..~~
~~4to.q l W Ays.n-.- 21 r:% 5e.As. W FtSeC7E#. %gC, tvst' 5.4m oe L. (12) Th1- _ -- 7 t freen -~~
~~.c s -_ uur -- -.~~
<- >_m;cn 44o, 4 cf .Ler ic; seel: Senromtv Wm1 1NWss E %gvag: 1 lc R 6" ** h 44o.4T (13) This valve is located on a non-Safety Grade piping system which is not a CLOC. 7'(14) The primary seal for each of the electrical penetrations con- .t sist of two concentric O-rings with a test connection to permit leak testing the space between the 0-ring seals. Test volume is pressurized with dry nitrogen. (15) Influent lines terminating in the suppression pool are dis-cussed in Subsection 6.2.4.3.2.2.1.1. mp (jQ %gse, Ms,ms, penetrate the containm :nt tit:ac are designed to remain intact following a LOCA.'---- i- - t M _._,;:.ne 1-;. age - t ::7_wre not specifically . Ad vented to the containment atmosphere or to the outside atmosphere and may remain water illed during Type A + test. ing. g %E4 s. i 6.2-218
^^ 480.45 la section 6.2.6.2.3 cf your FSAR you stato that the ttegrity of th) (6.2.6) inflatabla seals of the personnel access air 1:cks is t:st:d 'at sel cted intervals" act to exceed 72 hours. Indicate whether this means compliance with the requirements of Section III.D.2(b)(111) of Appendix J to 10 CFR Part 50 that air locks shall be tested within three days after being opened during periods when containment integrity is required and will be tested at least once every three days during periods of freguent opening. -~ If not, discuss the differences. eS.SPDd3& THG REO UIRG h5Nrs OF SECTioN LI1. D. 2. (b)(III) OF h PPEublX ) TO 10cF?, PART 50 THAT ADA L OC.l< S SilhLL 6E YESTG b talrHIN THRSE DA VS AFTER SEIN G 0PENGO DURINC, PERIODS WHEAL CON TAIU MEMY INTEGRITY IS RCQUIRC A M D S H ALL B E TESTED AT L EA ST O NC G E v'f RY TH R E2 D A VS CU Rs NG P5g to OS OF FREQUGN T OPEN EN G,. ARE. CO!1PLYED. WITri SY HAVIN G THE INTEGRITY OF THG ADR LOCl4 DOOR SEALS TESTED AT SELECTE D INTER \\fAl.S NOT TO GXCEEb 72 HOURS,WHicH 15 THREE DAVS, AS S TA T & b IN.SEC. TION c,,2, s,2,5 f r l l i l i i O
.. m m.n u m m.m ,w v. v 4, 6.2.6.2 Containment Penetation Leakage Rate Test (Type B) i s l[,j 6.2.6.2.1 Local Leak Detection Tests Containment penetrations whose designs incorporate resilient seals, gaskets or seal'ing compounds, piping penetrations fitted with cxpansion bellows serving as the containment boundary, air-lock door seals, equipment and access doors with resilient seals, end other testable penetrations are leak tested at periodic inter-vals during the lifetime of the unit in accordance with Appendix J of 10CFR50. The leak tests are performed to ensure the penetra-tions continuing integrity. To facilitate local leak testing, a permanently installed system is provided, consisting of a pressurized gas source (nitrogen or air) 'cnd the manifolding and valving necessary to subdivide the testable penetrations into groups of two to five. Each group is then pressurized, and if any leakage is detected (by pressure decay or flow meter), individual penetrations can be isolated and tested until the source and nature of the leak is determined. 6.2.6.2.2 Acceptance Criteria The combined leakage rate of all components subject to Type B and Type C (Subsection 6.2.6.3) tests shall not exceed 60% of L (cfm). 7 If r2 pairs are required to meet this limit, the results shall be >G reported in a separate summary to the NRC. The summary shall include the structural conditions of the components which con-tributed to the failure. 6.2.6.2.3 Retest Frequency In tompleones wths be resuurement of section !!l.b 2 (a)of aprendos ] 40 10 CF A ra rt $4 ftype B tests (except for air locks) 6hWlixbb performed during each are reactor shutdown for major fuel reloading, but in no case at Qs7ffer curveeiruf u fa, 6.2-143 l
~ 238 NUCLEAR ISLAND R v. 4Fl? ~~ l ~6.2.6.2.3 Retest Frequency (Continued) j g stE tuo& I
((
intervals greater than two years. t#he integrity of the inflatable seals of the personnel access air locks is tested at selected i (3 day 0 i intervals not to exceed 72 htt' Testing is automatically initiated at tha and of each interval by the seal test instrumentation [ oanual override of the automated sequence is provided I b3 3 ca... for la the associated logic. Testing involves the injection of [ air undet pressure (15 psig) into the space between the two I redundant seals in each door of the air lock. The leakdown t rate is measured by sensing the pressure drop and/or flow rate [ necessary to maintain the pressure. Main control room readout { of time to next test, test completion and test results is pro-l vi'ded. 7m alarm sounds if the specified interval passes without l a test beinq effected. No direct, safety-related function is l served by the seal test instrunentation sys' tem. I t t 6.2.C.2.'. Decign Provisions for Periodic Pressurization t In order to assure the capability of the containment to withstand the application of peak accident pressure at any time during plant } liiu Z_ J.m parpose of performing ILRTs, close attention is given to certain design and maintenance provisions. Specifically, the l cilccts of corrosion on the structural integrity of the containment are compensated for by the inclusion of a 40-yr service life { corrosion allowance, where applicable. Other design features that haua +ka na+antial to deteriorate with age, such as flexible seals, I l inspected and tested as outlined in Subsec-l Liva 6.2.G.i.2. In this manner, the structural and leakage (
~~ht containment remains essentially the same as~~
~ ~ r :s?.:'- :crepted. . :....:e w th the re9,vsre. ment af seetion 1i!. D.2 (t>)Oli) o f appench s p.- .,, u..., ] to iG Criz parT S*0, i ^ l 6.2-144
480.46 You stata ia Iloto 12 to Table 622-29 cf your FSAR that valves such as (5,2.6) those on the " chilled water 13 drywell cootrs" line are caempt from a Type C test since you provide a watcr tg seal. It is our position that no valve is exegt from Type C testing. In the case of valves which are sealed with a water seal system, Iqydrostatic testing (as opposed to air pressure testing) is permissible provided the line is not a potential ( containment atmosphere leak path. Additionally, the testing procedures ( require that the line be pressurized with water to a pressure not less ' than 1.10 P. Accordingly, revise your proposed design with respect to the hydrosfitic Type C testing of all valves sealed by a water seal system. Discuss how the water seal is established and maintained using safety-grade pipes and components, including the assumption of a single failure of active cog onents. ~ g t= S P D A S E-2eG.6de ( 9G, ST4les Cr Staff analyses of the contribution of main steam hoistion valve leakage to total calculated offsite doses in postulated * 'f ' loss of coolant accidents made with conservative allowances for transport delay j effects show that the twohour alte boundary dose is_ act a#ected by the subject leskage. The long term dose (Rg 14 Mi% in the low population zone, however, is affected for uncontrolled isolation valve leakage rates typical of current technical specification values. Thus, the stan has concluded that a fully automatic quick 4cting leakage control system is not required to meet the system objectives. A rnanually initiated leakage control system capable of being actuated within about 20 rninutes of an accident requiring use of the system would be acceptable. /' N A A % StJ A6yee cap 7pf 7)y,} g(pg(u3,(g gy,Q 'W COMMl- [yJYGs s t S/Atcc. h try?W W WE &ff4t;' 11 1114we f*W Andukacat M&, j%vs 17 t.c outL Pos> 7tr>t-D M 7 & c a tvec w e gbiHrs NAkAGe e p tg e @f 7/Mo /Jol471!h t% vt*5 SNOU N N 8{'N16 lA' p Cruruucrim wtw ke JmkAce twirrol Sf'** /$ Ollr fD%t7/M Jh7 }))is im WfrV2c/ TPJ7sd Lbed-h li'M c 7 es T. Cem)6rm 70 \\/e M17eva7 e-f / f sa MdM N 'I" " " l
GESSAR II 22A7007 I 238 NUCLEAR ISLAND R3v. 0 ) 6.2.1.6.1.2 Post-Construction Component Test Phase . (- After installation and immediately preceding the initial ILRT, local component leakage tests will be conducted to ensure that any leakage is detected, measured and minimized. The leak tests, in g emer a.L, follow the Criteria established for Type B and C tests of 10CFR50, Appendix J and include the testing of: (, (1) mechanical and electrical containment vessel penetration sleeve welds; (2) all resilient seals in personnel air locks, equipment hatches and fuel transfer tubes; (3) all isolation valves - operability of valves (prior to leak test) must be demonstrated by closure utilizing .he normal mode of operation; (4) air locks, by pressurization between the doors; (5) equipment hatch, by pressurization of the space between the seals; and (6) guard pipe and fuel transfer pipe bellows. euct T-ere C vs r; a *ruz-Su w ex no w.-4 vwed [ All tests will be performed by local pneumatic pressurization of 3 --- :: c; containment components, either individually or in groups . pr :.;ur: P. Leak detection will be by pressure decay, flow a rate measurement or equivalent means. T..., < v v a.v u s s v s m w % W rz. A J A 1 Faba W W6V4 %I wrw w w a sr. w 4. sysy,wra.s q, g.g p,,,3e n,_ gg p, The acceptance criteria for the combined leakage rate of all com-
~~ent:. Oc.nnot exceed 60% of La.~~
The specifics of the acceptance criteria are defined in 10CFR50 Appendix J for type "C" testing. s 6.2-68
CESSAR II 22A7007 238 NUCLEAR ISLAND R:v. 0 6.'2.6.2.5 Penetration Types and Leakage Rate Test ( 6.2.6.2.5.1 General The local leakage rate Type B and Type C tests,---f ; -- rr;;;;d aszEggadiEI!EBJun, must be completed prior to attempting the Type A test, and the sum of all leakage rates must be less than 288 scfh. 6.2.6.2.5.2 Personnel Locks g The personnel locks are located at El 11 ft 0 in., azimuth 120* and El 84 ft 7 in., azimuth 119*. The supplier's instruction manual should be consulted for proper setup and operation of the locks prior to the local leakage rate test. The local test should consist of a pressure test of the lock itself (pressurizing between the doors) and a test of the double inflatable seals on each door. () 6.2.6.2.5.3 Equipment Hatch The equipment hatch is located at El (-)5 ft 3 in., azimuth 220*. The double seals on the hatch cover are provided with a test con-nection to leak test the space between the 0-ring seals. This test must be performed in accordance with the supplier's instruction manual. ( 6.2.6.2.5.4 Fuel Transfer Tube Blind Flange Seals The fuel transfer tube is located below the fuel transfer pool at approximately El 38 ft. A local leakage rate test should be per-formed on the blind flange to ensure leaktightness prior to the containment test. Provisions for leak testing the flange-to-transfer tube seal are provided for on the transfer tube via an air pressure connection to a pair of 0-rings on the lower surface l' of the blind flange. O 6.2-145 l
11n3 Ems tRD]O@Y ^ 238 NUCLEAR ISIAND R;v. O l 6.2.6.2.5.5 Electrical and Bellows Penetrations Electrical and bellows penetrations are mechanical penetratio.ns with allowable leak rates for testing. The primary seal for the electrical penetration consists of two concentric O-rings, with a test connection to permit leak testing the space between the O-ring seals. 6 There is also a total of 43 mechanical penetrations which utilize bellows. These are testable connections with a test fitting to ellow for pressurizing between the two laminations of bellows. The leakage rates are included in the sum of Type B and C tests for limits in accordance with 10CFR50, Appendix J. Individual penetrations should be tested in sequence, using a check list to assure complete coverage, based upon Table 6.2-29. 6.2.6.3 Containment Isolation Valve Leakage Rate Tests (Type C) 6.2.6.3.1 General TwsCvw wmcTe575 A m E. R E e.* *-4.9 oc *u. thM*4 VAN 1Eu-cw Tet. VHW4 W e asA g,g w a g y, 'rhe.rbe r o * '""y c t er ts-re;"4 v=A h=r been redu::d t.c : ainicas excugh-the nee n* e4* =rM
~~tc. gritifc la= % c rclay~r prezid.2 that:~~
h DE (1) The valves h:Je h::a demonstrated to exhibit leakage rates that do not exceed those specified in the technical specifications, and (2) the isolation valve seal water system inventory (or air-flow rate of the compressor) is sufficient to provide sealing for at least 30 days at a pressure of 1.10 P,. See Table 6.2-29 for a tabulation of the valves (APS, WPS and MSPLCS) thus equippe6. These valves are also illustrated in Figure 6.7-1. 6.2-146
m uLNamu j 6.2.6.3.1 General (Continued) The containment isolation valves are listed in Table 6.2-29. Type C tests are performed by local pressurization ' sing either u the pressure decay or flow meter method. The ~.est pressure is applied in the same direction as that when the valve is required to perform its safety function, unless it can be shown that results from tests with pressure applied in a different direction ( are equivalent or conservative. For the prensure decay method, the test volume is pressurized with air or r.itrogen to at least P. The rate of decay of pressure of the known test volume is a monitored to calculate che leakage rate. For the flow meter method, the required test pressure is maintained in the test volume by making up air, nitrogen or water (if applicable) through a calibrated flow meter. The flow meter fluid flow rate is the .j isolation valve (or Type B test volume) leakage rate. , ~., \\. / 6.2.6.3.2 Tests Tvre c. (ww-rta nr.5r> a.ra.ca Fbw.mcr.a most w t. sww.c YM de+ ef M e t~ A 5 P sc.w w s e Oswo c-wg(_. rw.< qq wm M Meim of the valves listed in Table 6.2-29 are Ox;;ptr_ ,r;; t.m WAvsa we by virtue of a water leg seal, a pressurized w seal, - ;p _ -m 27 -r Tu watwa. a== tr. The M valves ^e 1ss. vwW4r@tre licted ic1;w.
1
~.. in 2 _E'_OO. d et = ' ' ^ ~ The t.f WW bs5r ul amed 12-b gae, chen *=eted "ith mir er nitregen, are La;;d-en z:1: -4'a. tvoe and expar-+-d ur:7 : T.#35EeX' (1) RHR system relief valve E12-F030 (one 1-in. relief valve); (2) drywell blee nt valve (two allel sets of two 2-in. gate valves se s); (3) containment cuum relief v es (one 24-in. check valve in s .s with one 24-in. butter alve in each line, o lines); and 1 d W.... _ 6.2-147
1 >W ,$4ae d 2.-/4.L7 Ms llCRD 8-w A_+ M (s a) 2.$nt., L, k Mbt! A G$5-zozo (z.,,s-A 3 R2at bac R W - ' M -- ~ ( s u = L u ) +- +. Ssf -.., A '.a k FF&c (z.v 1-m) ..~.. A Cral % - - 2% FA -,- DELA (< M b?hl!,nis;D2WL L C Ek #16 # (1.vdm) 7. D --n Cs rAA c P&d te A) _c f P W at /9 b A 4 & --u (1- &) 4 cc.uj f M :--- J (v 4 ) 10 ccw >6& K L &:=~.-r (z w1m) RE dedD $dC+ L L iev'= urJ (2 & & H. c O kAC. #WJ (* 2 es=4w)
~~u..~~
~~/s. c4 h ) 44 C, K 'A na >t O w A d sa.~~
~~o. r... s te d w # P J s x a w --,~ n w ]~~
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m 7RMWAW 238 NUCLEAR ISLAND R;v. O ~ .2 Tests (Continued) / '/~. (4) ESW ' e to Hydrogen Mixing System (four va s, divisiona Measurements from these valve re incl d in the 10CFR50, Appendix J limit on the sum of Typ and C tests. In terms of radiological release limits e RHR rel valve leakage goes to en area processed by SGTS, but not mixed, is therefore, i included in the l' s for those areas. The leakag of the dry-well bleed an rimary containment vacuum relief valves to the Shield B ' ding annulus and are included in the limit for lea e mixe d processed by the SGTS. 6.2.6.3.3 Acceptance Criteria The combined leakage rate of all components subject to Type B and Type C (Subsection 6.2.6.3) tests shall not exceed 60% of L,. If rcpairs are required to meet this limit, the results shall be rcported in a separate summary to the NRC,'to include the structural conditions of the components which contributed to the failure. 6.2.6.4 Scheduling and Reporting of Periodic Tests The periodic leakage rate test schedules for Type A, B and C tests are described in Chapter 16. Type B and C tests may be conducted at any time during normal plant operations or during shutdown periods, as long as the time interval between tests for any individual Type B or C test does not exceed the maximum allowable interval specified in Chapter 16. Each time a Type B or C test is completed, the overall total leakage rate for all required Type B and C tests is updated to reflect the most recent test results. Type A, B and C test ) 6.2-148
hem _drawM s. -howing the routing and fil6ation of the pip:ng should be ~ ( 480.47 (6.2.6) tsed to show the existence of a wat r seal. When the operation of a system is needed to maintata a wat r seal (2.g., the ECCS system) show that th] water seal will be maintained for 30 drys if the system, is removed from operation. Limits on liquid leakage rAould be assigned to these valves and included in the plant's Technical Specifications. r. f bfb I I I, e e i ! I e I t t t t t I t t ? 9 I f ' ' 'l*MI b3'Cr bl% _ NEON \\ W U C i i I %vo - c st e s m +c m.35._.s.cv%s. i I l I i !l l ! s I t ; i ( i i 1 i i t s ; ,,,.,i . i i j i, Eq% + ! 4*yl i x t<6 tT. W 6.y e w i r_.6
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480.48 In Section 6.2.'5.3.2 of your FSAR, you state that most of the valves (6.2.6) listed in Table '6.2-29 are enestfrom Type t tests by virtue of a water leg seal, a pressurized air seal or locationln a closed loop outside containment. As has been discussed in Question 480.46 concerning water leg seals, no valve is exegt from Type C testing. The use of an air or water positive leak?ge control system does not in any way eliminate the need for Type C test 1 rig. Accordingly, revise your proposed design to j reflect our position on this matter. dt. E, S ed rd S L dspawse.ko kon 4 80.96 S ea. c
~C80.09 ~ Air and watcr positive leakage control systems must be deactivated during (6,2.6) the conduct of the Type A containment integrated }eak rate. If positive kakage control systems will be used 12 your design such as on lines associated t:ith penetrations 12gc and 130c state whether these systems ' will be deactivated to insure that they do not provide a source of in-leakage into the containment and consequently distort the calculated leakage rate. g L 3 P O T4 ~S & 3s4 % 3 y,ll .' a l e H <. cow 4c.; 3 A %smw. se uwekwent % 4 T3 gr ~ Su no4e. 3 ma#W paq'. ow s.2 zn ( A779cumsN r g),
480.50 The standby gas treatment (SSTS) is an ESF system whose effectiveness (6,2.6) must be periodically verified as required by Appendia J.to 10 CFR Part 50 Ia so doing, the leaktge limit of the secondary containeer,t is measured t and till be found acceptable if it agrees with the limit used in the analysis of the secondary containment depressurization time. These tests e should be conducted at each refueling or at intervals not exceeding 18 months. The test limit should be consistent with the limit used for direct I leakage in the analysis of the radiological consequences by the Accident Evaluationtranch(AEB). Indicate the proposed tests which w111 be performed on the SSTS including: (1) their scheduled frequency; (2) a description of the test itself; (3) the depressurization time for the volumes listed in Table 6.2-31 of the FSAR; (4) the method used to measure them; and (5) the means by which the effect of open doors or hatches is included in the test program. State the design leakage rate and the SGTS fan capacity. (it 6 J P o N.I F_, _ Puahx. S 6; Ts roO d no a4 :4:.- 0Rcc.h '-c... ~ 6% k thud =.~ LnLL k nH-L -hdi$ "MLl%L- - E1%.sk!;t;.w ]~ J nnL.&.m -&d 1,' h -duII wrtr..f % h h4mz6 e d
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- h um!m&Lwh.4 mLh.em..ap. n +!, t an Mt.lt e,, -Ir;. Mcs.. h.L w..w. ' ~ M tt A m M ~4 7 d, Y d ri f.- A.2Erik N d.:..i .iba. L e w/.io MTiA 731% h,2.,z / h.+ i - FSA E v) k %.tLA wAdram -)h. 1 tLJ. H1:e usa 4 in uLA h oltad n n l ~ dess? w $0$tlJs UN W. ts.a.lshdk h.hl e 'W-b 'M ~y o.us. -71c. ss rs.w.- wu,;&J. /w o.- c.,m.a+:.' iOm cm M- & C. 1. 7. z. tL M j e y -A-- - h_ - N - lo..r- } e-g 4J~ Jc rs 1 P
es b a ssenYt a$qhYf C L 480.52 In Table 1AC.2-1 ef your FSAR, you list the conde: sate water system as (TMI)g, being both essential and non-essential. Clarify this apparent misstatement. be. E. s P a N s & TO +60,5 t MO sy3o, St %:6 Table 1AC.2-1 shows the information requested. 5hese changes l will l>e incorporated into a future GESAR revision. e e
Tablo 1AC.2-1 ESSENTIAL / NONESSENTIAL EQUIPMENT,' (Continued) few14line a-p System j Essential Comments o 35. Fuel Pool Cooling 7 F, 99 No Boiling is acceptable, but make-up is necessary. Heat exchanges cooled by RBCC system. 36. Drywell Bleed.e >on ' Yes Pressure control vent. Back-up to hydro-vu.,- uc..re /u gen control. 37. Positive Seal System 57, i,# Yes Insure that highly radioactive fluids are g o confined in the reactor building. ~ sa cau nw /ou+sms 38. Traversing In-Core Probe (TIP) No Not required for reactor shutdown coolins 39. Fire Protection d Yes Availability is essential, as the "acci-dent" may be the result of a fire. so enem -w N~r-uses,a 40. Mtske-up Water Treatment
~~M No Serves no purposes during and immediate13 after accident.~~
Longer-term availabilit) w. Jy/m-p.I cA.s r* y, pay, p e> necessary. r= r~ 49 "I'd & " 4*/~ r. 16J6 h A)?'/WAq-fu fROWPG * ,_ a
Table.ac.2-1 .4 ESSENTIAL / NONESSENTIAL EQUIPMENT (Continued) o N As./i-System j, Essential Comments No,t safety systemj sah err-e 27. RHR Vessel, Head Spray up No 28. RHR Containment Spray de Yes Necessa % to control nressure. m.res rr fewer.e.ser..s 29. RHR - LPCI Function ['[,#'(( Yes Safety function. 30. RHR - Steam Condensing p// No Not required as safety equipment. ~ wo up,* r s.o ra x-rren." - r Function b 31. Waste Collector and Surge #44-No Not required for shutdown. Tank No Used only in normal operation. Desirable 32. Drywell Cooling m,,yy to keep running. 33. Demineralized Water f,2/ No Not assumed available in ECCS analysis, ua can~-a e A ar ra n..a 34. Condensate Water "A No Not assumed available in ECCS analysis. GEIC5 8 i Q
s ESSENTIAL / NONESSENTIAL EQUIPMENT (Ccntinund) e />nfrak
~~p System No. _ Essential _~~
Comments refu,oer e snoJ~W m*nr**J JO' /r****. pe 20. Reactor Water Sample su No Not required for shutdown; but Ser M 50 ^ necessary for post-acci$ent assessment. Post-accident sample'is a separate issue. 21. Control Rod Drive Cooling af Yes No credit taken for reflood, but is desirable. ,k 22. Reactor Water Cleanup 3 6, t.3, No Not required during and imunediately l 0 ' ' 5 " ' following an accident. Necessary in o osy long-term recovery. j 23. Radwaste Collection No Not required for shutdown. Inr ere-* <V wd. m ta mmr*- Ar.er/ser*4~ 24. Recirculation System No Not required for jet pump plants because core can be cooled by natural circulation. stese JWe*~e f.~* c ** sos sce r e sz. Main heat sink durin;g isolation. Yes 25. RHR Heat Exchangers No Not essential but desirable to use if i 26. RHR Shutdown Cooling g available. Not redundant, but sr.fety i grade. i 16J4 u i
Tabl ..C.2-1 ES: IN:'IAL/NO' SSSENTI.'1L EQUIPMENT (:antinuer) f,.6,nn .c stmrr ,,. Essential Comments 14. Drywell and Containmerit nu n[ No Not necessary for core cooldown. j Floor, Drains *
~~e ce.en e 15.~~
Emergency Service Water f vf, isp Yes Necessary to remove heat following System /6 ( acc Ment. Includes the ultimate heat u sink. w F 5 g 16. Instrument Air f3, Yes Regarded as essential because this system o supports safety equipment. Back-up accumulators are available for the safety equipment should the system fail. l 17. Service Air i19 No Serves no safety or shutdown function. 18. Main Steam Line y, io,,,,,3. No Not required for shutdown. a s ora 19. Feedwater Line /4, e v No Not required for shutdown.- Portion that is Class I is essential. ~ i N GEIC3 j i
I { ESSENTIAL / NONESSENTIAL EQUIPMENT (Continued) l O M x t .D System k^###* Essential Comments N0. Scr srr*'* Le 8. Containment Spray Cooling ep Yes !!c c c r e a rj t o s v u L v i s ell /centek ini"- p i'e S e n e. 9. Automatic Depressurization/g ni Yes Safety system; control of RPV pressure. System 5 ? 10. Standby Gas Treatment g/,4 7 Yes Necessary to control emissions f.o environment. o sv. e, e.* w a + neum..o 11. Auxiliary Building / Fuel ,A Yes Neesssar M e-ceel rafety cy t:a p p: :d Building Emergency Cooling meters. 12. Reactor Core Isolation g,yp,y, Yes Nccessary for core cooldown following Cooling 5 e s2 /'a isolation from the turbine condenser and i feedwater makeup. >= caower festruenm 13. Auxiliary Building / Fuel p///- Yes/NO If drain is required, the equipment is Building Equipment Drain probably out-of-scrvice; check for indepen-dent isolation; drain should not back up and flood essential equipment.
~~GEIC2, c~~
i
Tablo iAC.2-1 ESJENTIAL/bCNESSENTIAL EQUIPMENT a m System /Puercde Essential Comments ,u d. 1. Reactor Head cooling v/# No Not a, safety systemj Aa*
*'8 MW Hs we enea,s r w m ero nea 2.
Standby Liquid control .Velli Should L. avaitabler. inch p te ;n - i rycter.t ~/+ 3. Low Pressure Coolant w/p -Ves E f0ty ;yetee. N / r,--. 2.9 Injection aw n. t y 4. Ocparete Suppression Yes Main heat sink during isoladon. 33,3y l b Pool Cooling w 5. Core Spray (High-Low Yes safety systems. r3pge,. Pressure) f 4 say, s. 6. Closed Cooling Water n.,,,,, No Used for normal operation only. Not ce-ers,ce.A/-c-) np e'd required for DBA btg is necessary for the recire, cleanup system operation, and fue'l' pool heat exchangers. 1 se e n n,,, e M r*.*n a 7. Containment Atmospheric A Yes combustible-ga,s-cont-rel fr mtien-sweeesery l Control t0 cli&II.;t: hyd~0-=,'^ y;;;r. cc.1 etible artacephne. 16J1 1
w suru w r wrv (TMI)A should be performed for the purge valves. In your response 13 TH1 Action 4 Item II.E.4.2. you propose that no leakage be measured for th:se valves since they are covered by an air positive 1:akage contr;l system, However, the air positive leakage control does not preclude the need to test the purge valves in accordance with the Appendix J requirements. Accordingly, your position is not acceptable on this actter. Discuss your proposed test procedures for the purge valves in light of this, _6?_ E S P O N J & See EOS%5e. To 4-so.4c. l I l
ATTACHMENT NO. 3 FINAL DRAFT RESPONSES TO QUALITY ASSURANCES BRANCH QUESTIONS l
j-pm,d \\ c _., l - e., c 3 7 e,3 s 3 + g i s -o
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+ 260.01 In Section 17.2 of your FSAD., you provide a 'cdn61tment to J, apply the previously approved General E?ectgic QM topical I , repo'rt, NED0-11209-04A, to thedesign pdu' dot struction of i t ..y the nuclear island. Jowever, since thif appm val was granted, Section 17.1 of the Standard Reuteil?la6(SFP),has been updated to, reflect current revisions [thrigulatory guio6s and ANSI ' = standards and to incorporate gnhancemnts of certain QA controls. Wesbelieve that you shoeld corisider incorporating these upgraded control s. Accordingly, n;odify, NED0-11209 to reflect. the latest revision (Revision 2) to Sectiod 17.'1lof the-?SRP. Alternatively, ' provide a supplement to Chapter.17 of..your FSAR specifying the ' additional controls contained.in this revised section of the SRP uith which you can comply._, ~ t ,g s t
~~RESPONSE~~
Domestic BWR Licensees reference, the URC approved General Electric Topical Report NEDO-11209-04A in their', Safety Analysis, Reports.- The General Electrie quality assurance prorfe4 described .'in NEDO-il209-04A fully cor.1 plies with Appendix 3 of 10CFR50, and currer.cly applicable Regulatory Guidos and ANSI St anda rds, as evidehced by NRG acceptance letter dated December 6, 1982. Since it is expected that revisions to the NRC's quality-related requirements and R2gulatory Guides ;will occur in the future,- General ' Electric will comply with thediatest Japproved version of yNEDO-il209. This document will be revised ts:' re flect, la't e r requiremedts as determined necessary by the IRC and so cearnunicated through the applicant.at the time of docketing an application. 3 ,s .: \\ t o ') \\ t 'l i .\\ i l s g = 1 l 3 l \\ ,I i s k ( 2' '[ ' o' .\\ is ~' ,f ' z. a 3, a g \\ ' i, e. ' Y 'l\\ m' y y g 'N $5 ['] T 3, \\ "\\ a? \\.: n\\ l ,'i 6 l \\l s f 3, 1 ..\\ s ' I '( f [ N ssTX i i 't l l .3. 'T ~ i (y\\
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lb 260.02 In ' Chapter 1.4 of your FSAR, you identify C.F. Braun & Co. as a principal contractor performing engineering services for the design of the nuclear isla.nd. Describe C.F. Braun's QA program with emphasis on how it ;omplies with the latest revision to Section 17.1 of the SRP. Additionally, describe the QA program of othe.r principal contractors, if ar.y are identified at this time, who will be engaged in construction management activities on the nuclear island.
~~RESPONSE~~
General Electric will be responsible to specify, in its procurement documents, the QA program requirements applicable to the engineering services work performed by C.F. Braun and Co. The requirements in the General Electric procurement documents will be fully consistent with the applicable requirements specified in the procurement documents received from the applicant. The General Electric QA Program for performing this work is described in NED0-11209-04A. In particular, NED0-11209-04A Sections 14 (Procurement Document Control) and 7 (Control of Purchased Material, Equipment and Services) describe how GE will assure contractor compliance with the applicable requirements, including the applicable requirements of Regulations, Regulatory l Guides, and ANSI Standards. 1 i I 4 4 4 1 ,\\ i i3 \\ -. _, _.. _.. _. _ _. _ _...... -,., _ _,... -... - _,}}

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