ML20082Q152
| ML20082Q152 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 11/17/1983 |
| From: | Rubenstein L Office of Nuclear Reactor Regulation |
| To: | Novak T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20079F427 | List:
|
| References | |
| FOIA-84-35 NUDOCS 8312120002 | |
| Download: ML20082Q152 (88) | |
Text
{{#Wiki_filter:a... g8p' o 5 g UNITED STATES 8 Q( g NUCLEAR REGULATORY COMMISSION ,4, WASHINGTON, D C. 20555 %' ".... / 'NOV 17 IS83 MEMORANDUM FOR: Thomas M. Novak, Assistant Director for Licensing, Division of Licensing FROM: L. S. Rubenstein, Assistant Director for Core and Plant Systems, Division of Systems Integration CUBJECT: SHEARON HARRIS SER UPDATE (DOCKET NOS. 50-400/401)
Reference:
Letters from M. A. McDuffie (CP&L) to Harold R. Denton (USNRC) dated October 11, 26, 27, 28 and November 4, 1983 Based upon new infomation (Reference), we have closed out the open items in sections of our SER forwarded to you by memorandum of September 14,1983 in which the Auxiliary Systems Branch has the primary review respnnsibility. The sections which have been closed are: 5.2.5 - Reacter Coolant Pressure Boundary Leakage Detection 5.4.11 - Pressurizer Relief Tank 9.1.3 - Spent Fual Pool Cooling and Cleanup System 9.1.5 - Heavy Load Handling System 9.2.2 - Reactor Auxiliary Cooling Water System 9.2.8 - Essential Services Chilled Water System 9.2.10 - Waste Processing Building Cooling Water System 9.4.1 - Control Room Area Ventilation System 10.4.9 - Auxiliary Feedwater System Copies of these updated sections are enclosed. The only sections in our SER remaining with open items are: 1. 3.3.1.1, " Internally Generated Missiles (Outside Containment)" and 2. 9.3.1, " Compressed Air Systems."
Contact:
N. Waaner PY. Ras Been Sent to PDR lJl Odd S 1
\\ a a (I .M 'NOV 171E33 Further infonnation is required to close these sections. After this infonna-tion is received, and reviewed, the rescits of that review will be reported in a supplement to the SER. /7 ITN b/ v61.Auc U Q ~ L. S. Rubenstein, Assistant Director for Core and Plant Systems Division of Systems Integration
Enclosure:
As Stated cc w/o enclosure: R. Mattson D. Eisenhut L. Rubenstein
- 0. Parr G. Knighton cc w/ enclosure:
B. Buckley J. Holonich(2) R. Lobel N. Wagner s t
,i Distribution Docket File ASB Rdg. File ASB Members NOV 171983 LRubenstein EMORANDUM FOR: Thomas M. Novak, Assistant Director for Licensing Division of Licensing FROM: L. S. Rubenstein, Assistant Director for Core and Plant Systems, Division of Systems Integration
SUBJECT:
SHEAR 0N HARRIS SER UPDATE (DOCKET N05. 50-400/401)
Reference:
Letters from M. A. McDuffie (CP&L) to Harold R. Denton (USNRC) dated October 11, 26, 27, 28 and November 4, 1983 Based upon new infomation (Reference), we have closed out the open items in sections of our SER forwarded to you by memorandum of September 14, 1983 in which the Auxiliary Systems Branch has the primary review responsibility. The sections which have been closed are: 5.2.5 - Reactor Coolant Pressure Boundary Leakage Detection j 5.4.11 - Pressurizer Relief Tank 9.1.3 - Spent Fuel Pool Cooling and Cleanup System 9.1.5 - Heavy Load Handling System 9.2.2 - Reactor Auxiliary Csoling Water System 9.2.8 - Essential Services Chilled Water System e 9.2.10 - Waste Processing Building Cooling Water System 9.4.1 - Control Room Area Ventilation System 10.4.9 - Auxiliary Feedwater System Copies of these updated sections are enclosed. The only sections in our SER renaining with open items are: i 1 1. 3.5.1.1, " Internally Generated Missiles (Outside Containment)" and 2. 9.3.1, " Compressed Air Systems."
Contact:
N. Wagner XE9467 l ,,,,,,,,,,,,,,,,,,,,,,,l,,,,,,,,,,,,,,,,,,,,,,,, oss ca su:==s q,.,~............ , l..,,,,,,,,,,,,,,,,,,,,,, emy........,....................,,,,,,,,,,,, ,,,,,,,,,,,,,,,,,,,,,,,,j,,,,,,,,,,,,,,,,,,,,,,,, we noau va oo.aomcw oaao OFFICIAL RECORD COPY m,,,, .. _ - =
., l, .s -=4 ~. 5.2.5 Reactor Coolant Pressure eovndary Leakaoe Detectien The reactor coolant pressure boundary Leakage detection systeus were reviewed in accordance with "Section 5.2'5 of'NURIG-0800 (SRP). An audit review of each of the areas Listed in the " Areas of Rgview" portionfoftheSAPsectionwasperformedaccording to the guidelines,provided in the " Review" Procedures" portion of the SRP section. Conformance with the acceptance criteria. - -..'._ _ formed the basis for our evaluation of the reactor coolant pressure boundary leakage detection systems with to the applicable regulations of 1D,**CFR 50. respect A Limited amount of Leakage is to be expected,from components forming the reactor coolant pressure boundary (RCP9). Means are provided for detecting and identifying this Leakage in accordance with the requirements of Gen +ral Design criterion 30, " Quality of Reactor Coolant Pressure Boundary." Leakage is classified into two typejMhidentified and unidenti-fied. Components such a,s valve stem packing, pump shaft seats, and flanges are not completely leak- { tight. Since this Leakage is expected, it i s con-sidered identified Leakage and is monitored, limited,*~ e ,,w.-- ,,-w--r-.g.m.,-_ -,.-#,.,a,-...7 .-,y ,,e,y-. ,ymw-,*w..wy-wv.4.. e. ,,,-,,,e g ww_y-m,,,,,,_m,.-,.,.. - -,,, ,_m w-,,,_,m.,%-,,w
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and separated from other Leakage (unidentified) by directing it to 4tosed systems in compliance with the, guidelines of Position C.1.a of Regulatory Guide 1.45, " Reactor Coolant Pressure Soundary Leakage Detection Sy::tems." Within the containment building, identified Leakage is kept within a closed system by being directed to ,~ the reactor coolant drain tank (RCDT) or pressurizer relief. tank (PRT). Pumpy seal or valve packing Leakage is directed to the reactor coolant drain tank. % y a monitored y*=4 c-p,e.. or ?.- pressure, temperature, t ev e l ' ': : 1,; c.:.. :.., and f low i ru.t rument a t. ion on y the reactor coolant drain tank discharge Lines. Leakage past the pressurizer safety valves or power operated relief valves is directed to the pressurizer relief tank. This Leakage is monitored by gemperature QS wcu tR.S instrumentation in't'he piping system a b tank pressure, temperature and Level instrumentation. Leakage collected in the pressurizer relief tank is ~ directed to the reactor coolant drain tank for sub-sequent discharge. 4 4 e n-en ,-,arn,- w-,-.. ,n,m-.- .nv-vn--
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= .! g: 4' a..; ~n. ...a how t measurement /for The applica.t identifie Leakage TLow to the RCDT and PRT re ""Ma* kr*L'%"%"d"UL* %" M" h nA al j Q,10 gp above the bactgroun teakage in der to / M { comp with Position c.1 of Regulat y Guide,1.f5 for identified Leakage. Unidentified Leakage is monitored in several ways. e44 ~~ '"-'t.;:'
- '-_t Leakage to the containment is monitored by airborne particulate and noble gas monitors, by containment air coolers, by condensate tev'eI*and flow monitors, and by containment sump flow monitors, thus complying with Positiott C.3 of Regulatory Guide 1.45.
In normat operation, these i primary monitors show a background Level which.is indicative of the normal Level of unidentified Leakage inside the containment. Increases either in l airborne radioactivity or in containment sump flow or in humidity of the containment atmosphere above the normal background Level signify an increase in unidentified Le~akage and, provide notic,e to the plant operators that corrective action may be required. Unidentified Leakage L'h..
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.o i '%(~ k ~}8'a - Insert 1 par + o:A :%f &skna 44tcle xs The applicant noted that the PRT and RCDThsed to monitor the' Mr; inws by bak*CR, e reactor coolant system invent ory balance,w,1+eh ia-required by th e ' Standard Technical Specifications a N is an integral part of ) the Shearon Harris Technical Specifications. The inventory balancer E.;? -6 required within each 72 hour period, wiLL aid in estab-Lishing the identified Leakage rater thus complying with Position C.1.b of Regulatory Guide 1.45. ~ ~.r'.~ O e e , +, t ,n-e
s s.yp_ Qhe
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[.... :;..,$ result s in a aicka-a-flow of water'to the contairment sump,F ;;:: '- :: -et% Fw...;w. '.7.,vi S'. .n;;;d ; ;
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The water represents $mch fra$ jag 10em conoporren.s of'teakage-fromunidentifiedsource%=aswater.*'Y A'n,. / ttCe25 r.*,1/9 :-:r ow. p t* Ve In tmwn.., ry
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.. n u.. w : wkick 'kn$ }eCh Ce den & bV L. .... wi w w.. w..... ;. '-d a.>inment fan cooters 1 and fan coit unius. The water is routed through a si bfinch single diameter pipe to a measurement tank Located inside the containment sump. The floor of the reactor cavity is sloped to the sump so ttrae drainage ~ on that floor is also collected and monitored in the measurement tank. A flow switch is employed in a drain Line on the botton of the' measuring tank. Water flow past the flow switch wiLL actuate an algr,a in the control room when the. set point i s reached or exceeded. The swit ch set point is dete'rsined by first establishing the normal maximum and then cetting the flow switch to alarm Leakage rate at 1 gpa flow above the expected normal maximum Leakage c.2. rater thus complying with Position M of Regulatory ~ s,enib Guide 1.45 with rega.-d to l ' ;fllftCd f /s n! I'd Leakage :nd 'ii as46 O S g D ea- - -., -- -,,.a . -, -, + _ - -,,, -, - - -,, _. - - - -, - - --,,w,.-,,_.,,-~-----,,---------e------,--e ---,.---.--,,---.---en-
I. ~ l 1 A Level transmitter is installed in the containment sump to indicate water Level on the main, control board. ng 1 Level and flow monitoring equipmentgGh #Iesigned Sump "to'r'emain functionaC efter an OBE. The main control board has a seismically qualified monitor Light to indicate sump high-high level. Also, an' alarm wilL Me. sound on sump high-high Level and/or whendleakage rate exceeds 1 gpa above the maximum normal Leakage. The airbcene particulate and gaseous radioactivity monitoring systems are designed to remain functional when subjected to the SSE. Inaddition,t'h,'contain-ment temperature and pressure monitors, which serve as backups gross Leakage indicatorse are also quali-fied for the SSE. The containment sump flow monitor is capable of performing its function following' seismic events not requiring plant shutdown. The time to detect a 1 gpm L'eak by radioactive coolant monitoring depends on coolant activity and previous Leakage. ' d - " ' ' ' ' '. : ;
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'::---- c' '.:" ...e The sump Level is monitored by a Level. transmitter and is. indicated in percentage of sump 1 e ..v :. .n ,,,-n.,, ,,,-.m,, ...n,, , ~ _,,., _ _ _ - -. - ,ew-------=-_,e-
s. .'~3% 4 Level in the control room. The flow rate from the-drains is detected by a flow' switch and alarmed in the control room whenever the Leakage rate exceeds the normal maximum Leakage rate by 1 gps.' Thus the guidelines of Regulatory Guide 1.45, Position 2, C.5, C.6 and C.7, with respect to sensitivity 4' response time, airborne particulate monitoring systems.' seismic' qualification, and control room indication and alarm are met. The containment airborne radioactivity monitors are seismic Category I and are located in flood-and t ornado protected st ructures, thus, meeting--the-require-ments of General Design Criterion 2, " Design Bases for Protection Against Natural Phenomenas" and the guidelines of Regulatory Guide 1.29, " Seismic Design classification," Positions C.1 and C.2. They are also testable in accorda$ce with the guidelines of Position C.8 of Regulatory Guide 1.45. The applicant has pro-N posed Technical Specifications which include the Limiting conditions for operation with identified and ~ unidentified lea'< age and also e initing conditions for operation when one or morefleakage detection s y s t eris are unavailable, in accordance with the guidelines of.' of Regulatory Guide 1.45. /YOWeVC/ N' Position C. / / oposed o a s oeuWis;+on 75 deal' w!6 /4aor /as p%nh teacfor,eso a, th /aahye r sfemfressore. 7e~c4-j.so /ahan ya/ves.a s' can?ainea'in Ne -o ica/ s ceificenocc f" Wes;%)deass Pressorn e o r C A c.A./ W O A A Yy ~ d ! C h i^,'E N " Q A S h0W hN.
- 96.
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- ,fp Leakage h auxiliary systems cohnected to the RCP is
- ' detected by increase in Level, temperature and/or pressure in the auxiliary system.
The auxiliary t. systems for*which suth methods are svailable are the accumulators and centrifugal charging pump subs,ystem in the emergency core' cooling system and the suction' kLeakage side of the residual heat removal (RHR) syste fiers d.4CPb into the discharge side of the RHR syste%wiLLresult;
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in pressurizing the disc ar Wen tje.pressun reubarge side of the RHRA nket &n s*sfste$'"C*> fr He, s/le)'(rkNe.s"5777//$)'?/N$$$thYnj'&AgV inYo lk: c "'f? ....uv w. .. '. ? : ' : :.;:: yr':- p;' + -.. t44 u. ' EMY;,,, 2 si I/l? S t. ..'. :, borona recycle holdup tank g :;;. '. 6 6.nw....- dh/ w//[rssa// /b indication and alarms in the control room. Leakage from the RCPB into the component coolin.g ) 1 water (CCW) system through any of the folLowing paths wiLL be detectectby. radiation monitors in the CCW i system: 1. RHR system heat exchangerse 2. Re actor c oo*Lant pump thermal barrier, i 3. RHR pumps, y* 4. Letdown heat exchanger 5. Seat water heat e x c h a'n g e r, L 6. " Excess Letdown heat exchanger u) g TL ....,getua Ly part of the CvCs but.f... oftheRCPkwh'en i n o p e r. : -; ;,.r.. part ~.w-- g~ >y- ~. - -,,, e o re ,--..m.n_ -,,,,. ..,-,,n,----,w.-, ,a-an ,,__w,,,___,,
- We find therefore that the applicant has made provisions to detect intersy{ fan Leakage in accordance with the provisions of Position ~ c.4 of Regulatory Guide 1.45. However, the applicant has proposed no Technica*. Specification to deal with such intersystem Leakage through valves and other components which compose the interfaces between the reactor coolant system and these auxiliary systems. It must be noted that-determination of intersysten Leakages as required by SRP Section 5.2.5, devolver upon protection of safety-related systems contiguous to the reactor coolant system (RCS) against damage as a result of Leakage from the RCS by continuous surveillance. Therefores the applicant wilL be required to include a Limit for Leakage into systems contiguous to the RCS consistent with the requirements of the Standard Technical Speci-fications for Westinghouse Pressuriad Water Reactors (NUREG-0452) in order to comply with the guidelines of Position C.9 of Regulatory Guide 1.45. We wilL review the Shearon Harris Technical Specifi-thisrequirementjghk# cations in order to assure that is co'ntained therein and that the Shearon Harris Technical Specifications for RCPB Leakage are consistent with the requirements of NUREG-0452 Af fkodefsM#5'd in order to " comply with, Position C.9 of Regulatory Guide 1.45. O s N = = = = = = -
1 nj v Leaks through steam generator tubes from the primary system into the secondary (main steam) system are detected by radiation monitors in the steam generator ~~ blowdown Lin'es. In this way' detection of Leaks from the RCPS into auxiliary systesse and into the s, team and component cooling water systemscomplies with the guidelines of Position C.4 of Regulatory Guide 1.45. '~ Based on the abover we conclude that the reactor coolant pressure boundary Leakage detection systems are diverse and provide reasonable assurance that primary system Leakage (both ident ified and unident i.fied)".IeTL L be ~ ' detected.and meet the requirements of General Design Criteria 2 and 30 with respect to protection against natural phenomena and peovisions for reactor coolant pressure boundary Leak detection and identifications ~ and the guidelines of Regulatory Guides 1.29, Positions 8 C.1 and C.2, and 1.45, Positions C.1 through C.$r with l respect to seismic classification and reactor coolant l pressure boundary l e a k a g e d e t e c t i o n s y s t e m d e s i g n,e$ vsp o_ __2 1 =___; --,,,. ,,,r4,...+e. 4. ___.12_ m-e._,,,.. 1_ _&_ _ _ , _ _. - - C%dreloWp & we, conc.ude that, O cae&raf' /NW' the dts en of the facility for reactor, boundary leakage detee a C... 2 meets the acceptance crite'ria of SRP M n* LLl,Pmv de re stu+sn & w w a ssars yppyius Section d.2.5., y,,,,,'_ zz; /- -- =
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-m I-5.4.11 _ - seurizer Reti.f Taak (prassueir.e Rati f 04eeks--. p Svstem) The pressurizer relief se=%$was reviawed in accordance (S A P ). with Section 5.4.11 of NUREG-080g 6944. An audit review of each of the areas Listed in the " Areas of Review" portion of the SRP section was performed accord-ing to the guidelines provided in the " Review Protedures" portion of the SRP section. Conformance with the acceptance criteri.a formed the basis for our evaluation of the pressurizer relief discharge system with respect a=~ to the applicable renulations of 10 CFR 50. The pressurizer relief discharge system consists of the pressurizer relief tank, the discharge ' piping from the pressurizeNHLIFf inif sa f ety valves, the internal spray headers the nitrogen supplys the vent to containments and the drain to the waste processing system. The system is nonsafety-related (Quality Group De nonseismic Category I) and is not part of the reactor coolant pressure boundary since aLL of'"its components are downstream of t, the reac%or coolant systeds# safety and relief valves. Thereforer its failure woulo not affect the integrity of the reactor coolant prossure boundary. M b .,n --y,-- a
99-The pressurizer relief tank is sized.to absorb.the enetgy content of 110% of the full power pressurizer steam volume through t,he gejmar,y relief and safety valves. Other relief, vajv h...dischargeStothepressurizer LWG/40s 1'40GL re lie f t ank wat4f rom t he residual heat removat system and from the chemical and volume control system., Releases from these sources are Less;than the design basis release from the pressurizer. T r. e internal spray and tsttom drain on the pressurizer relief tank are used to cool
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the water within the tank. A nitrogen blanket is also provided in the tank to permit expansion of entering steam and to control the tank internst atmosphere. If adischargeexceedingthedesignbasisshould[,6ccure the rupture discs on the tank would fait, allowing the dis-charge to pass in:- a containment. The contents of the tank can be drained to the waste ' holdup tank in the waste processing system or the recycle holdup tank is.the boron cycle system via the reactor coolant drain tank pumps. The rupture discs on the pressurizer relief tank have a capacity equal to t h e c omb i n e d c a p a(it y o f t h e g&4!SC(Ar, 2 E/'f e t y v a l ve s. A 'sa .....r The tank and the rupture disc holders are designed for full vacuum to prevent collapse if the contents cool following a discharge without nitrogen being added. Thet c pressurizef relief tank is provided with instrumentation M Oc cordTt3l r comers ctbring-met I indicate and alarg pressures temperature and htgh and to Low % nk Level. ;.; ...e c ;.. ~.. s s..:.- g 8
t = a4g. F a i. L u r e o f t h e p r e ss s u r i z e r r e l i'e f if is c'h a r g'e s'y s i: e m I n a seismic event wiLL not affect the capability of the plint to shutdown, thus complying with the guidelines of Sesenso hsign dqssig,efecy F Position C.2 of Regulatory Guide 1.29g Interfaces between this system and the pressurizer are formed by valves designed to seismic Category I standards, tKus complying writh the guidelines of Position C.3 of Reguta-tory Guide 1.29. l,,", s - " * ' '~ w= a-. x r. .~ eiv.es ara, W i 7-7.,.,.':__ _ 'ra en=* -, e r n se
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gir" : i. = ..i.urv w1LL not acversety a nect sa. ty- _kediscussionofenviornmentalquali-c-tered : r'
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.,6.. fication of safety-related equipment as a' resu0t of Q,,p f ailure of the pressurizer,,ehe.F +ctnN r 2 see Section 3.11 of this SER).j N ^ ' :. :' a. the frIsidnz.ev tytMYJ-p Th e r e f o r e, = 4: ^t ' t ': ^ '^'^^ ^' /Y'II M system meets the requirements of General Design Criterion 4, " Environmental and Missile Design Bases."1hte L 4 Based on our review, we** conclude that the pressurizer relief discharge system meets the requirements of General Design Criteria 2 with respect to the need for protection of safety-related systems in the event of failure of the pressurizer relief discharge system by meeting the guide-. 1.ines of Regulatory Guide 1.29, Positions Ce2 and C.3, e ,m.,,,__,---% ..r+-- - - - - - - - - - - - - ^ * ~ ' ^
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==*((, ' -j,,,.i f _ *. 7 concerning its seismic classification. nm u...g = =-3.i.-_-. ..s: S 4 u- = y ... sus,,- t--, ...w u - r
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4-th the requirements ~ =..
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of Generat Design Criterion 4. Therefore he pressurtzer 3 relief discharge system e m mees the acceptance criteria of SRP Section 5.4.11.
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9.1.3 Fuel Pool Cooline and Cleanu'o S y' s t e'm J S e e n't Fuel' Poet
- Coolino and cleanuo Svstem)
The fuel pool cooling and cleanup system was reviewed in accordance with Section 9.1.3 of NUREG-0800 (SRP). ~ An audit review of each of the areas Listed in the " Areas of Review" portion of the SRP section war per. formed according to the guidelines provided in the " Review Procedures" portion of the SRP section. Con-formance with the acceptance criteria, except as noted below, formed the basis for our evaluation of the fuel pool cooling and cleanup system with respect to the applicable regulations of 10 CFR 50. The acceptance criteria for the cool.ing portion of the fuel pool cooling and cleanup system (FPCCS) consists of meeting the guidelines of Regulatory Guide 1.52, " Design, Testing,'and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtratio~n*and Adsorption Units of Light-l Water-Cooled Nuclear Power PL&nts," in the event the cooling portion of,the system does not comply with the requirements o.* General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena." This does not apply to the Shearon Harris FPCCS, since the ~ l l - - - - - + .----e-
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~ y , fuel pools cooling portion of the FPCCS do'es~ meet the* requirements of General Design Criterion 2, as a below. ~ The two new fuel storage pools' Cone for each unit) and two spent fuel storage pools (one for each unit > are housed.in the uelfandling uilding (FHB). The FPCCS consistsoftwofuelpooljcooling systems (one for each unit); the fuel pool cleanup systems for both units are separate from the cooling systems. Each fuel pool cooling system services the new and spent fuel storage pools associated with that unit. Each FPCCS contains two fuel pools cooling trains, each train containing a heat exchanger, strainer and fuel poolfcoolingpumpwitheachpump capable of being manually Loaded onto a separate emergency power supply, in case of Loss of offsite pnwer. Each cooling train is a 100% subsystem, servicing both the new and spent ~ f uel storage pool id 'that unit. t 1 The cleanup system.s contain two fuel pootg scimmer pumps, 4 two domineralizers, filters and skimmers for alL the pools (including the cask loading pool in Unit 1), skimmers for the transfer canals and two fuel poo(g and ' water purification pumps. Tha.mainfueltransferaanah, 1 which runs almost'the length o'f the FHB, interconnects the two spent fuel pools.
Mb-The neu fuel storage pools, at opposite ends of the i - /#8 ~ the are each connected to the main fuel transfer canal via fuel transfer canals; each spent fuel i pool is also connected to the fuel transfer canat in its unit. ALL pools and canals are kept separate \\ by means of removable gates. Makeup to the pools may be provided feca a seismic Category I source, t the refueling water storage tank ' y means of the s fuel pact cooling pumos. This cesplies with guide-Line C.8 of Regulatory Guide 1.I*3, " Spent Fuel Storage Design Basis " . i.. ThefueL.pooljhcoolingportionoftheFPCChis designed to seismic Category I, q'u'ality group C standards, while the fuel pool $ cleanup systems portion of the FPCCS i s designed to nonseismic Category I, quality group 6 standards. The FPCCS . is designed to remove the decay heat from the spent fuel assemblies stored in the pools and to maintain the clarity of the water in the pools. The essential porti6ns of the system are housed in the seismic Category Is flood and tornado protected .e fuelyhandling building (refer to Sections 3.4.1 and 4 e:- -+-- + --e, r--e.,~- ,-,-%-.,-.-w,- ,..-e-.,.,-.r--- .,,-__-._.,-,_,,-,..--ew,,------.
. 3.5.2 of this SER).. As noted 'abovt, ?the FPCCS} With the exception of the cleanup portion, is designed to Quality Group C and seismic Category I requirements. Failure of the nonseismic Category I, Quality Group D cleanup portion ~ wiLL not affect operation of the cooling train as iso-Lation capability of that portion of the piping system is provided, and therefore, no adverse effect on safety"- related equipment would result from such a failure. -~ The applicant reported the elevations at which the FPCCS pipelines entered and Left the fuel pools. The applicant also noted that skimmer hose Length was Limited so that only five (5) feet could be submerged. Based an this information*, we conclude that the minimum level to which the water in the pool could be drawn down to, assuming U SyF *p Oh-k.A N a pipe failure, is about 274-1/2 feet, while the Level of the top of the spent fuel would be 260 feet, leaving at least 14-1/2 feet of water above the top of the fuel. 3 u.7 d..- h: re t : J. n. w^ g. w.... Therefore, the .....s, i design satisfies the guidelines of Positions C.1 and C.2 of Regulatory Guide.1.29, " Seismic Design Classification," and the guidelines of Positions C.1, C.2, and C.6 af 1 g Regulatory Guide 1.13, the guidelines of Position C.2 of t \\ Regulatory Guide 1.26, " Quality Group Classifications and ' Standards for Water-Steam, and Radioactive-Waste- .l s j N.,,,.., t ~- 4. l i i
"""G E4 u w $ u w I. ; f, "I. '. : C .- U p ".. ! ^. '.... s. 2 v 1 n e s u s.: s o r y e,...,>> ..scr-ate.. . ;. i ,;;'...:.. ": t:-Containing Components of Nuclear Power Plants," thus meeting the requirements of General Design Criterion 2, " Design Bases for Pro-tection Against Natural Phenomena." The various components of the system are Located in separate missile shielded cubicles within the tornado-mis' site protected fuel handling building and are separated from other moderate and high energy piping systems'.(refer to Sections 3.5.1.1 and 3.6.1 of this SER). Thus, the requirement's of G neral Design ' Criterion 4, "Environmenta4.and Missile Design Bases," and the guidelines of Regulatory Guide 1.13, Position C.2, are satisfied. The applicant had originally intended CFSAR Section
- 1. 2. 3 c, p a g e 1. 2. 3'-1 ) to make the unit 1 FPCCS
~ operative with one cooling train. However, in response to our question regarding possible failure W of the FPCCS the applicant committed to install two (2) fuel pool cooling pumps and heat exchangers A s madW for operation of Unit [. The applicant r. ; _ ' _, : -- tdri @ a ar oma d-- an. 2. ./ L.s.. J ,./ =.. 1 a L'L 7 ' *::";1.,,/ jr -"I'Tj g$ ~G y v'ry-ny'Y'" W IW T C - ' Y ~T 4, _fL e.,,~. yyjuuSee,s / I D ---r- .-,----y -o. ..-.-.--v----- ,,,,,,.,,,,,-.<.-w-------,%--,.-.,,,c.,,-.-.w--,....-.-w.---------.--e.- - - = =
ss ~ The fuel stora~ge pools and FPCCS have been designed to permit transfer of fuel assemblies between units; this does not affect operation of the FPCCS since it is intended to be usej to rem ve thJ decay heaJ h wiYA e fiw// Dr&IT d+ SfenT fron4 fuel poots & ;'....",.". 6.. . 6u w.6.ying l fuel, assemblies. Therefore, the FPCCS meets the requir'ements of General oesign criterion Se " Sharing of Structuress Systems, and Components." The applicant has designed the fuel pool cooling system to maintain the Unit 2 spent fuel pool (the pool with the greatest estimated maxjg,u.m heat I Load) temperature at 124*F with both trains of cooling in operation and assuming a futt inventory of fuel assemblies including PWR and BWR spent fuel from the H. B. Robinson and Brunswick plants. Under.these same conditions with only one train in operations the pool temperature would reach 142*F. We b4/s Con fim & ": ^;..: 6 1 _,,. 7.; the, applicant's 1 estimate r thegcooling heat load c.. wi+4 a H c.oinp/ein as/ ef rJed:~Kw..e/,- ek'
- ~~'
- , i '
0 $ $d/n}//kf C'elekk. k.
- l. n gMdW/'fa a.c n 1-c.
6n-and t. Wi
- .2
- --- '
- :
t--- ~c-.:... .u. n.. A,_f*P A.sa.. 9 L- .2 __ g. m.... 3,_ [' dura: j '
- -J-a-.
uf*h 9T9 o-S . ',,,.. 6ais 6 A g [ m l-l*" ..-.,.n-..---,-,,,n ..--n,---.,_ ,a.,,
M4 ....vsmation n&s Des.. A C '.... w r... w 6.
- 6. i e w s w w s e s.in t n e g
M.. .w.sw.wi .. ' ^. P.. ; ;. '. :... '.. ' ; '. : n ".....w... w..... k; "k k W e. ' : " - ' - - - - ' r i r '... " ne.. T c;.. 't':r...."A m conclude that the FPCCS has been designed to meet the requirements of General Ye doebh] C,4f dllk,Yl Cooling Water " W/ k f8f' Design Criterion 44 Ta,,,,, -,,,4 we...
- 7, s.,4-..s
.w.. .s. e m ';;; ::r-7 area canaa* $: d r :.. wwwn ow v t... '. '.... ....o .ppi u is.. 6, .=. .w... 6ae 6m. ,r'
- :...~ w. s The safety-related com-ponent cooling water system provides cooli g water to the ' fuel pool heat exchanger a.pd transfers its heat to the ultimate heat sink (refer to Section 9.2.2 and 9.2.5 of this SER).
The spent fuel pool pumps can be powered from the emergency (Class OE) h(,3% ~ power sources) ..,,... e the design meet % .....e the requirements of' General Design Criterion 44. Normally, makeup. water 6 :to the;fuet pool.is fWO i supplied fromidue., seism}cCat gory I refueling water "AI
- A
,$t* Qm P w Sey* gc,y y l-storage tanks The applicant state (that I ^ % (y system is available to fill the fuel pool through 4 l r ,---,--,--,n
j .backtsy a s at valved and flanged emergency connections ^ ' ' M -$eismicCatygoryIwater h un// Anv'ib = = A""::!: - ~~ ~ W zi :&f u'a ra,,,j ". ^.^,y c e O A "%',*.E7. sou 'L C/ Th; ; B nt ;h ;;',/. ;h ;. i;; Gi; ;;..e:tica i;. -/
==d
- ... Cong it wousu 6...
- . ;;.;;;; th
- i-d u-s9"**!
2-d ?-- -- -- il L- ^*'- - th-e.... i v i..aing'e ::tf: f ailu. e -;.. i ;., w+te- - ' ' ; ; ' ;' f ,e,, f- /, _ > fsw 5 ]e, i, g, a y s yu4 6ne cou tea w th Loss of u off;ite,e-e.,jin e..r to snow comptience = l w ;. ;F'- W9ut M *th h#4T'i.+ a.esseet of General Design Criter'an 44. 4 The design of the spent fuel pool cooling system and its accessible location are such tha.t p e r N,'d'i c testing.*and inservice inspection of the system can be accomplished. The active components cf the spent fuel pool cooling system are either in continuous or intermittent operation during aLL plant operating conditions. Thus, the requirements of General Design Criteria 45, "Insp.e.ction of Cooling Water Sys'teme" and 46, " Testing of Cooling Water Systems" are satisfied. e The system incorporates control room alarmed pool water Levels temperature and building radiation level wenitoring systems. Th e pecH. L i n e r s ame S/" T-. haw 4a/sAnj. .ak ad spad faa/ shaye pso/s, whik c/a ri4W os nanuckse safefy, are. l \\ ,-,..,---..----..-,,....n---- ..n--. --,-...---.---,--u---- e.- a s.
6,jir /Yst/r.5-0l0 .u<a2" 2 p 4%g Q fd g i r W w l a f e a K. e + w j o r g,u co u a p ~ ~ ' pfeuf p/ps/.skp q a uacu+nws u.y ases wfucb b d ik.f p af i ht/3/ a _&& v g ? p y / h? %,E s a of a &,eaue' Md 199'G k' iw ui.ef A s f I h i9 ~ 4 p s Q
- m L,eas - m fd $HP wfs ate Qaw cy
~ 4/'t ik&E ifoguldde af(1%m / $d ,tw &qA ,# n ,m A W$ %'W e /ok.atA &# hRe &Stg5 * % $.Th G w % d$s
L designed te seismic Category I. h. A low flow-A alarm is provided t'o warn the operator of any interruption or cessation of flow in each fuel pool cooling train. Thus, the requirements of General Design Criterion 63, " Monitoring Fuel and.Was_te Storage," are satisfied. Based on our review, we conclude tha the fuel pool cooling and cleanup system is in conformance with tne requirements o f G e n e r a l D e s i g n C r i t e r i a 2, 4, 5,'PIs 45, 46, 61, and 63 as A they relate to protection against natural phenomena,
- ~~
mi s si les and envi ronmenta l e f f e ct s,f of St #nc$,Codt uSt '4/DN/0h s(,1ertn inservice inspection, g ca e///r functional testing, fuel c f f ~ and radiation protection, $".IN' and monitoring provisions, and the........ of.. Regulatory Guides 1.13,.1.26, and 1.29, relating to the system's ~' design, and quality and seismic group ~c'Lassification. 94ma' wH.,.. ~- _,..,,....... ou... 6 ,,,,m-y ui my; nowever, saw .m ~ ms y
- 7 " ~ 7 " Q ? u ; l e - _ ' - ! :.,..... m
- c. :.
ti. :7 ^. ..,,_,,r. .. -~.. e. ~ h... = __ = . Cased ,,,meru epue n m e t '7 7 > : 2 There-d fore, we m find the FPCCS to be acceptable. The i F P C CS dewsmunes s e c 1;( t'h e a c c e p t a n : e criteria of SRP Section - Mr5Seng-- ~ p 9.1.3. eana** f 5: - 4 c= g s'-'....... ..,.......T.. l Tp gpp.ccus+ tJ M Qt of ecct-t on c4-the neto a ci sprit he] s+6eocyt. Foola= For Orsr+
- Willnot te.,n cr.m+.s w r+h i ess fco l C.coIins puf'hf 5 Cretl. %.cd o heec4 cxc.ho ngo s c.p era b le, 9
~ -,._, m.,-, -..e ,.--.r ,_.-.......,----..--e
..s, -- ;s. f 9.1.5 Fuel Handlino System (Portion Related to Overhead Hesvv Lead Handline System The portion of the fuel system related to overhead heavy Loads, was reviewed in accordance with Section 9.1.5 of NUREG-0800 (SRP). An audit review of each of the areas Listed in the " Areas of Review" portion of the SRP section was performed according to the guidelines provided in the " Review Pro- - -~ cedures" portion of the SRP section. Conformance with the acceptance criteria,.except as noted below, formed the basis for our evalua*. ion of the portion of the fuel handling system related to overheat heavy Loads with respect _,to the applicable
- regulations of 10 CFR 50.
The atteptance criteria for the overhead heavy Load handling portion of the fuel handling system include meeting the guideL*i'nes of ANS 57.1, " Design Require-ments for Light Water Reactor Fuel Handling Systems," and ANS 57.2, "D,esign Objectives for Light Water ~ Reactor Spent Fuel Storage Facilities at Nuclear Power Plants." The guidelines contained in the " Review Procedures" and NUREG-0612, " Control of Heavy Loads At Nuclear Power Plants," were used in ~ Lieu of ANS 57.1 and ANS 57.2. l: .... _., _ ~ _ _ _ - -.,-_ _ _ _ _, _.
~ "M The overhead heavy load hahdLitig 'COHLH) ' portion of* the fuel handling system consists of equipment necessary for the safe handling of the spent fuet cask and removable protective barriers, and for ~ safe disassemblys handling,'and reassembly of the reactor vessel, heads and internals during refueling operations. The containment circular bridge crane is used for handling heavy Loads in containment and the spent fuel cask handling crane and auxiliary crane are used for handling heavy loads in the fuel storage. building. The ent is housed within the fuel handling rLsr. ~g em 5'ed buildi pano, and reactor building (containment) which are seismic Category 1, flood-and tornado-protected structures (rufer to Sections 3.4.1 and 3.5.2 of this SER). Although uel handling system components are not required to function folto n SSEr critical comp"ohents of the fuel handling system l are des!cned so that they yiLL not fail in a manner s o o s.t.o..,CQt4ST. C634NF., i l tet 4-h ;- - - - unne,--_. 77-- a 4,,.. g....,. a,.,;. -1.... ..;,_y.. l The 250-ton circular bridge crane is capable of retaining a 175-ton lifted load (weight of the integrated reactor vessel head with lifting rige L 5 ' M ,-n a 3 e % c a G e t r ceIced da s ++wm ta cc,,,for:ents.
i - "1Tr-which is the h'eaviest component
- to be lifted during i
refueling cperations) during an OBE or SSE, although l the crane may not be operable after the seismic event. The bridge and t.oLLey are prevented from leaving their runways with or without the 175-ton lifted Load during operation or under either seismic event. Both the 150-ton spent fuel cask handling crane and the 12-ton auxiliary crane ar e capable of retaining their maximum Loads during an SSE .-~ ~ although the cranes may not be operable after the LJw.. ".*4 seism ~ic event.
- 77..
..99 ...J '. '. ;, :'t-*}-ra_ p - -.. s y g _ ...n. 4,. 1 -' ..,4.u-..'. '--c"e,ad
- t.., 4. -
- 5;e -
-..,0 v. u4,6,, gu 4, 1-.... -f ,,,yot...s 4.... The design thus satisfies the requirements memana. of General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena," by compliance with the guidelines of Regulatory Guides 1.13, l " Spent Fuel Storage Facility Design Basis," Position C.1, and 1.29, " Seismic Design C la s s i f i c a t i on e Positions C.1 a,nd C.2. Each unit of the plant has its own containment buildingcircularbridgecranej{ The spent fuel 4 e
e 9% cask handling and auxiliary cranes are shared ~ between both units in the fuel handling buildino. 64A For unit 1, prior to completion of unl:it 2, the fuel handling and unloading area and the spent fuel cask area wiLL be completely constructed and a,vait-able. Sharing of the OHLH system between units does not affect the capability of the plant to shut-down since the OHLH system has no safety function. Thus, the requirements of General Design Criterion 5, " Sharing of Structures, Systems and components," are not applicable. The spegt fuel cask handling and auxiliary cranes are carried on common rail supports which travel the length cf the fuel handling building. Movement of the cask crane is restricted to the cask haiidling area by mechanical stops, limit switches, and er hd's;Q h i administrative con.trols. he mechanical stops a l
- C ne ensk.cr-or14 l
prevent longitudinal movement,,T portion of the fuet-handling building areas containin % t. el\\OSanbo W trs t h _ 0 fuelastoragepools}gn:Qthenew the spent fuel storag f00($. ,..<;--...~-_>;.m. 1,, : _ eu4*ekst ereuma+
- he scent fet a l esek bandt4me eraam 2---
r 2-a m.y - .k. nau
- 4... t 7;'
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~
- I.'O -
- nd.' U:
_ The spent fuel cask handling -a crans is also equipped with stops which limit main hook vertical travel to ensure the shipping cask could never fall more than 30 ft through air to any J h34O Load-bearing surfacei W iLL not be raised more 6 s echtS than ?? ' above the operating floor. 4 ty e ted eauipment within the administratively n controlled area of main hook travel, eitner on the operating floor Level or on floors beneath. The floors of the cask decontamination areas cask head storage areas cask loading pools railroad car un-Loading bays and the operating deck can wJ.thstand a postulated drop of the spent fuel cask f on the maximum height to which it can be' raised. There is a removable barrier between the cask pool and new fuel storage pool at the north end to prevent damage to any spent fuel stored in that fuel pool in the event e s a cask drop. l l The auxiliary crane is used for handling gates i +be % e. between,transfpr canals and fuel pools and for 3 handling fuel racks and the removable barrier dis-cussed above. This crane has access to the entire fuel handling building and is atLowed to carry Loads. .D Q@I'M 55wl9dMYh ~ up to 10 tons Me auxiliary crane is designed to b$6 - p, appl,ca,4 r-e.ed h I'm Y g g g c.g g e s o at A9COSk00'II
I I Ni i the requtrements of ANSI 93d.2,'" Safety meet Standards for Overhead and Gantry Cranes," and 1 CMAA Specification No. 70, " Specifications for Electric Overhead Traveling Crines," and meets the i Ga.ses fav Aloc.bar lbMO' hekh. 4 recommenciations of NUREG-0554, "S nele-Failure-Proo t / , m,,..,, c. i > u - W m 't:d4Lg 3 dl3fgpl e$rg+ NM g __ 1 =A "' C nsjfe s f o r Nu c l e a r P op e r P l a n t s."
- t. 7 ; 3T ' T - ~ ~_ ~ ~ - -
' kS$'hhN , ?. 5 feh....,,Y.W bW Y $h _ , g s..., / 7 @_ -. W... h,,. - - N,.. f. W-. N..t/ E. F_" M. /.4,k,., ffS l / MMMMdO Qd hM,P @Mr,gM{ ~, biU.N:kidhrvi.$0:ddequeSc.g5 lo y +.M #tqm ~~ l ,/ .cra nen 6!, ./ /.> h /
- n a,. c
/ i
- !%ef-in ord er t o c om p ly wi th t h e gu i d e tA n c s.. o f g
/ Position cjS of Regulatory Guide 1.1' i theVeby l w.ee-4v.a.,.t j j / c.L" ; with the requirements oY eneral Des.fgn I / CriteHyn 4, " Environmental and ssileDesig/ v n f Bases / NJ -h & &[Mbh, M W. /bD 0 ,sts - 3: The applicant h provided a response to our
- )
question regarding analyses of dropping heavy Loads other than the spent fuel cask. Th.. '. '--a* -~ M t"bsJtCdQd - - " ' +-
- ' a n a n a l y s i sf25 f t c h
- id;; G.;
n the conse-e,uences of dropping the' integrated re' actor vessel C' we6.c. i,.... .7^ '.....m n. jg e. k n 11 fue J _ - E 4 * ^^^" --..e
- e. d 6(
- <# I
. I ine e. e ~ .-..-.-w, ..-.,..-v--.
Ah ~ OEC3el beed N h3rhen$ h #49,3 C.the"EK4% eMIM M I4 ' I >-. a a. head (IRVH) wa s d-h~ The applicant concluded 4 l thats as a result oftheanalyshse the integrity of g -- the reactor vessel we's not ieopardized meewams coolin i ~~ ~. . TE e h e. capability,nw makeup to the reactor vesselk n.;... LA fWoo erS4 @'. fed rated" f"S.5u ff".L.E. A bNeMr 2- - ' a"' d t;. c... .:.f- u. ;. ; - - - - as the applicang .k ...;d. automaticacJuationofcontainment isolation \\ Nhighradiation*L1mit% releases ad hnyf-so that doses P "wel' below" 1/4 of the Limits required by 10 CFR 100 w,e +f i; ....,_-*--=d 4- 'm
- _;.t The applicant also stated that proarction for required safe shutdown equipment is afforded by redundancy since the dropped IRVH cout on damage the. Lim ted directlylbelowit.
Ti s GCCffht In tt amount of* equipment -.w'ti... ~ u 4 _ _---p m -1.gr s;57 y 3gpyy j_ z mine th WylPW Ne deffects of dropping re peevU ! Ykt FP4uk 0.intjenals. M/' de i cto vessel i &C Q kd fkGf ffMW M u t :- "*e y=- n vet -tee e-c'e :ra a 4pg hroxt i h 0f C N( 4-n. 'y--- e< e- ::
- 8
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g'Of d c;;' ; th: '..; w.*J :... d L ; e ; k ; ; f 5 0 t '- t5: : : r. : - i r. - c ( l b e ' d i..y g u i. c gr nc Jud.,...; f;;t :::k 5 0. d L *.3 1.. C
- r. :.-
- -d
- T. f G i n,c r Me'4Vy L04Gs w A s ' i. is a4y Lw mi.,wrd s
n c '.' O." --th: " ; ; ; i ; a- .~s e e i l = I. i ; h ;., u l d d e us a g e 6 in g sort (
- ett rate m of - seise: -
.g Msur3 en 3 * % ertoun%M snSe 92Y5h h e F an riw f cjrcP 4>0uld bt e w,,- w --m----e -p. --,---,.--.--e -y e.-~~,v- ,m.-,,-c,w,-m-,----,,,,._------..--,--,--r.r--,-,,--a--,..-.,---we,,, - - - - - - - -
e t l "*595== i The applicant '.....J L, O.. .P':
- -+4aal -
has kf.O N W ith the portion of the overhead heavy a Load haddling system dealing with fuel handling. e vnva w w ,The applicant must d il c';t
- ..f heavy loads wh i ch could damage safety-related equipment required f,o r safe shutdown and/or removal of decay heat in addition to a heavy Load damaging the core or spent fuel, tr.ereby releasing radioactivity (note that a heavy Load herein refers to any load which could cause such he.c4 4 damage).
TF--=*aa--
- .-'4rma*
r:
- L
....w., o'=
- I c' th..;....'.
7 C.. 5 : - " '/ '**f ; r e s -., (?--** ?.*.* .: m J., - 0 4 t?4 " * : n t r ; '. :' " *: 1., _ -d_ a-u,.... -,_,as i;2-.. 1 ui %= 4d 'O...:.. 3.4 4 unrougn i S 1.5 a# """ : 00 ^ 21 aust ce campic6.w wy cae cime i t'- ':: :: ;L:::d i; ::
- terf_ -
) fM Y Based on our reviews,.we conclude that the everhead l g heavy Load handling system is in conformance with the cequirements of General Design Criteria 2 fnd 61 as ) they relate to its protection against natural phenomena, -hdA 4ud and safe handling of the spent fuel cask, and the 4 guidelines of Regulatory Guides 1.13 and 1.29 with respect to overhead crane interlocks and maintaining 6 6e e .___,.,_.,,...g-,,_ y _-,_y .w, ,,-w,--.- -,m.-.---_---n- ,,-,,,-,e.
+. YS Y 4
- fMk ?
bY g M 9 Such onalysos are required as part of the applicant's response to NUREG-0612, which was transmitted to the applicant for action by Generic NRC L,etters dated December 12, 1980, and February 3, '1981, and will be included in the staf f's review of the applicant's " ' * * " * *
- xGr ap cut-Ins cohrmtte$ $ yy~
fe:uf f,$a44r Mah -0&& 41 Q the1 war czeu & M)t ( & cl % 4 /*f/ ' Allt flee 1 W fj x &,s /,.1 m.c d f b,. i,1a p g )[ &e J& - a; a c-(f1 4 'ac g". 8,% \\ .~ ^,i,l ^ h 04- _. Ag ~nd , L,8 ~~riby' m / 1 N $Y ,w ( Maug-es/v.
k 6 . ) 1 plant safety in a seismic event. fecoet+e f..). hi&&.42 b.t RI.9Y i.dsd,ta.. shoe-4ht;
- k-dr:;;;d :: ' - - - - -
ss - :n :_
- ...- /. :
. : ::; 's.w i t h G en e r a l.4estin Tr1T(fT6'n't.(- etee % w.. roc-4 Th :;;;'oest aust resolve the WUREG-0612 m.e.-r iesues.A Thereforer we f: -, ; fi n d x ;a-
- ; ; ' '.. the portion 'of the fuel handling system reta J.s dcy?f bg,ted to handling of overhead heavy Loads Therefore, the a
fuel handling syste-i d::- - :: seetSthe requirements of SRP Section 9.1.5 g rf M oluhlCA M 4/# % vylelt ",. :;tYdYds;'g " Twat shh rnNf -fe &S ~ SGR. M 71 fort d~bE ~5FMp-kp-P', i, g w,i. [ew m., tA t e-r a _,., o. $, @s'E, h!2'2,.ig2 ~. tw ar,s -me afphcaJ hes cemmihW 6 tomftk A'a a,aq oedn a w+ke wr accepfrdie to -Me deff pmr& ~Ae/ <u.cedy4 diuj ax tte q pOcas]1 ]sfrsq W Arst wfsad /ro.cg/gT wShaald te caeri d Me 1 /rioe /o o# , z z z+ - Y CM l du 4.f or,lvd,t Cart >Ftf 'k%Vfra ffokY' G do.R' W=: -2 ew-b 'l4' m a +- ~n - w
- renn,
+v fa & fee cesphauer wrt4 the Mmwenda1wnc of du. tran.. ,y.. ,-,.-,,_n,,., - - - - -,,,,, -, -, --,--....-,,.-..n
- ' ~ ~ ~
o _ g s-9.2.2 Comoonent Coolina Water System (Reactor Auxiliary Co.'.ine Water Systems) The component cooling water (CCW) system was reviewed in accordance with Section 9.2.2 of the Standard Review Plan (NUREG-0800). An audit review of each area Listed in the " Areas of Review" portion of the SRP section was performed according to the guidelines provided in the " Review Procedures" ( - -~ portion of the SRP section. Conformance with the acceptance criteria, except as noted belows formed the basis for ars evaluation of the component cooling water system with respect to the applicabte regu-L&tions' of 10 CFR 50. The component cooling water (CCW) system serves as an i ntermediate closed cooling water system between radioactive or potentially radicactive systems and the nonradioactive service water system. The CCW system rejects its heat via the station service water system (refer to Section 9.2.1 of this SER) to the auxiliary and main reservoirs 'and cooling towers comprising the ultimate heat sink (discussed in Section 9.2.5 of this EER). This arrangement E M e + 9 y +- m- ,a--w., ,-,,,,-w-,.,, - - -. -. - - -, ,,_,-w.,-e- --,.-+%v e,.--w - -...,ww-m -_._-,,,----._.-3,e-- --.y.,
. - ~ ~ 4 minimizes the possibility of Leakage of radioactive material into the environment. Each unit has its own CCW system; thus, General Design Criterion 5, " Sharing of Structures, Systems, and Components, does not apply. The Component Cooling Water System provides cooling for t?.e following heat sources: a. reactor coolant pump motor bearing oil cooler; b. reactor coolant pump thermal barrier-c. Letdown heat exchanger mical and c.L.une Control. s ); .s Syrt?m k 'mical and Volume k d. seal water .: exchanger LCne k k Control Systsm); e. excess etdown eat exchanger (Chemical and M Volume ntrol.,_ -';.Sp h y heat.r.emoval pumps ( M f. residual Residual Heat ) g oval M tem); g. residual heat exchangers idual at emoval -4 ^ b h. recycle evaporator package (Soron Recycle System); OC. i. reactor coolant drain tank neat exchanger (Waste ocessing tem); m
i ~9r-g J. spent fuel pool heat exchanger (Spent el Pool A S Cooling and Cleanup $A System); k. sample heat exchangers (Process Fampling sy~stes); and k L. oss failed fuel detector (Process Sampling S y,s t e m). The applicant noted that the safety-related heat loads included only the RHR pumps $3 tem c) RHR heat "~~ exchangers (((temg) and spent fuel pool heat exchangers Z+e[ e 6.. ~ The CCW. system con,tains two separate trains each containing a component cooling heat exchanger. There are three component cooling pumps for the two trains. Two pumps are operated during cooldown with each pump supplying hatf of the total CCW flow. Normal operation &&Ly requi.'es one pump with another I on standby; in the event of a LOCAr only one pump is required but two CCW pumps starts to assure cooling flow to the safeguards pumps in the eyent of a single failure. The motor-operated valves and pump: in one train are connected to one emergency power on % Sr* (s supply while thevalvesandpumpfa~re connected to
t NGw another power supply. The third pump has a dual i breaker arrangement so that if it is needed to replac1 a failed pumps it can be connected to the same power supply a4 the pump it replaces or to the same power supply as the operating pump. Interlocks are provided so that the third pump cannot be connected to both redundant circuits, simultaneously. Thus, cooldown can be achieved with two pumps even with a single active failure. The CCW system contains a surge tank on the suction side of the pumps. This tank sitL accomeadate an increas,ed system water volume resulting from thermal ~~ expansion and/or inteakage from components being cooled and to provide CCW system makeup in the event of smaLL system Leaks. The surge tank is separated into two parts by a baffle to permit operation of one train in the e,v,ent of a failure of part of the surge tank. Inteakage may be detected by an increase in the surge tank level or as an increase in radiation level (if from.a radioactive source). The two trains are connected downstream by a crossover with four l motor operated valves; these are closed wher. i llT. nS A necessary to maintain train separation. .l. a i l ^ - ~ ' ~ ~ ~ ~ ~ - ~ -
t i In'sortfh ~ ~ Thus, the two trains are isolated and the CCW system wilL have one operable train (in the event of a failure in the other train) or i two operable CCW pumps (in the event of a pump failure). This wiLL be sufficient to provide cooling to a single train of the safety-related heat loads requiring cooling by the CCW. This complies e**** e e 8,e e s g e 9 a ee O e G e 4 7. s
-w f;.;;' ; : :_: ....s.$,
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- .-n: :- a with the criteria of General Design Criterion 44s " Cooling Water."
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f;:'---' ^ ~~" 2 : : ', ...',c t.:' ..b., ...~f: .y. _:,.su.. s. . - ~ h the CCW system'is used to cool the reactor m.w. coolant pump (RCP) thermal barrier and RCP upper bearing and low bearing oil cooters.' Int $eevent of a ph'ase B containment i s o la t i ca s i gna l 4P+r the at*w cooling water headers to the RCPs 44 isolated wijh 3 NM be t o t h e p u m pjf H o w e v e r, ' men O M Otp"rWal NtT1e no CCW sen injection water is prcvided to the RCP seats to prevent overheating via the chemical and volume control system but no water is provided"t'o cool the RCP motor bearing =. =t99 The applicant stated that loss of CCW to the RCP bearings wiLL be detected by flow instrumentation with low flow alarms in the control room to be designed in accordance with IEEE 279. Furthers the applicant stated that procedures must be provided to trip the reactor and reactor coolant pumps if I p, g,e-__ .,,,,,--.-,,---a .--,u,, a,-,,,w me ,,, - ~,, -,,,, - - --,.-m -,...a, ,n
%= CCW to the RCP's cannot be restored within 10 minutes. bib r? l " f 'l,' 5 4 l k 01 N "_"~M ? % 'E 6*pe Des,2u minutessecms an u n. Least initiate y ant protection in the event of loss o coolant o the pump bearings in order to comply fully with he requirements o,f General Design Criteria-4. In additions the appli- ~ cant must provide assura.'ce that the reactor coolant pumps wilL not be requ red to operate to mitigate a smaLL break LOCA et erwise the applicant wiLL have to make provis ns to kee the RCPs operation in the event of a mall break L A in order to comply with the requirements of G nical Design Criteri'on 44. The safety-related portions of the CCW system are designed in conformance with seismic Category I and safety class C standardsr thus complying with the guidelines of Position C.1 of Regulatory Guide 1.29s " Seismic Design Classification." The applicant 54cd g 44444 that Lines to and from the excess Letdown and reactor coolant drain tank heat excha'ngers are-isolated on a phase A jj2 containment isolation signals while the lines to and from the reactor L 'i 4
~100 ~ up. n coolant pumps close automatically en a phase 8 A i /Afsl&~ 2 - containment isolation signal. '! ; c h e;;g ; f ) HTS ~/'iO N PMTM .v g t,,- w.-.... a t twet e s s a r y"* (ggps a f at.y-24 La ted) m e e i to t sTe d 'i n t he eve n t- = pg_e - ,t_.. ~. _ _, p ^. @) w.-e. V.7' -signeLs,.44.Jequirede 4n order *to - - ~ w - m rs essemppmumess+i t h e gu i d e L i n e s o f P o s i t i on C. 2 o f hasaf'2'fv.M Regulatory Guide 1.29. The applicant c_:: ; '.. ; A that the safety-related portions of the CCW system ..~ are protected agains: natural phenomena such au sN fLoodsi hurricanes & 0 45 f% ternt"\\% ~ tornadoese r ir ...c g 4 erthe requirements of General Design Criterion 2r " Design Bases for Protection Against Natur&L Phenome'na." M CA [ - g hah Nc.cA b hThesafety-relatedportionoftheCCWsystem i s designed and Located to permit preservice 2nd in-service inspections in compliance with the require-l ments of General De' sign Criterion 45, " Inspection of I Cooling Water System." i l to 4s E cart. f t C3pll2be l site The CCW system is testable -':_... /* eJanat e.euenca Jncluding reactor shutdown and srbFN b d[ a accidensi - "= ara In additions the CCW system is d e ,m-., ,y..m,
~ -to f-MSMr v An S signal (safety injection actuation signal) wilL cause the nonsafety-related heat loads, the sample heat exchangers and grossfa;1{edfuel detectors (Items k and L)r to be isolated from the CCW system. No further action is required until a residual heat removal (RHR) heat exchanger is required to cool the reactor coolant in order to establish the post-LOCA recirculation phase. In that events the remaining nonessential safety-related heat toads would be isolated manually. However, we noted that this manual isolation would include the isolation of the spent fuel pool heat exchanger. In response to our concern regarding the spent fuel pool the applicant noted that the fuel pool load would 0 6 be reduced from 38.3 x 10 to only 18.2 x 10 BTU /HR because a LOCA in Unit 1 would not be expected to occur simultaneously with the need to completely unload the Unit 1 core. In that cases the applicant states that 5600 gpm would be required for the RHR heat exchanger. The flow remaining from one operating CCW train (3550 gas) would be sufficient to keep the Unit 1 spent fuel pool at a temperature of 150*F or less. The applicant noted that, after isolating the spent fuel pool heat exchanger, the operator would have approximately 4.5 hours time in which to disconnect the nonsafety-related loads on the same header as the spent fuel pool before the water in the spent fuel pool temperature reached 150 F (12 hours before the water started boiling). N h M b5 Sofisfies
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'lof, constantly in"use t'o cool thosi heat loads required Inssc4 9 3. 9x Gu*/W the CCW system _l* W during normal operationg
- e 3
requirements of General Design t t Criterion 46, " Testing of Cooling Water System."
- 2) b; Whd yo The CCW system complies
,ith the requirements of 5> ,z- --D,Qga/ AA w_ura Genera'l Design Criteria ] 5 '"_- g '*-.- a - 'a= ,g ef/ggff; %. w.. w r inw. -.;;'Or t; d ; t e..- i... ..Wivr the CCW ^ r system complies with the requirements of Regulatory
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Gu i e e 1. 29, = e n r. a r ;.
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<.<r. - o e s-r..A m e.ets e rkr#4ry e CCw system e g;..... .. - - the requirements of rs.,,9,w,,.,*b.ecti on 9.2.2. Q tas all,"tMPe%9.I. SRP S <= u< m a s u,eeie m e,a- % % Seg. -emm8 A8m - .r. e <4 g *Iq m -. -.. M _ __ se 3 - -,. ygg, g c11. 6 w mM=g, ggp w nme l i I =4 e. -,p, - - - - -. ~,.,,-- ,--,r.--
ti3 Insert # 3 7.2.2 %ST $@VidA The applicant hoe r: iddTechnical Specifications ci:' f: 'td b include periodic testing of aLL of the safety-related portions of .the CCW system as a unit with al,L valves operating to isolate nonsafety-related Loads and having two component cooling water ~ pump trains start to serve the safety-related Loads upon a suitaole initiating signal. The applica.it should 'opose such a surveillance test requirement 16 the Technical Specifs ions to be shutdown. hg & I CEAMAAU $pe r f o rmedN& }M every 18 montas, ducir.g b A gua& pwdu e. G O l l e 1 e
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'- l a t -. 9.2.8 Esrential Services _ Chilled Water System The essential services chitled wqter system (ESCWS) was reviewed in accordance with Section 9.2.2 of NUREG-0800 (SRP). An audit review of each of the areas Listed in ~ the "Araas of Review" portion of the SRP section.was performed,&ccording to the guidelines provided in the i' Review Procedures", portion of the SRP section. Con-formance with the acq[)ptance criteria formed the basis ~ ~ ~ for our evaluation of the essential services chitled water system with respect to the applicable regulations of 10 CFR 50. The ESCWS Eonsists of two redundant trains, A and 5, for each unit. Each train consists of a package water chiller, an expansion tank, a chemical addition tank, and a chitled water pump together with associated piping. The servica water system (see Section 9.2.1 of the SER f or di s cus siem of the service water system) suoplies water for the condenser section of the water chiller during normal and emergency conditions. The expansion tank a,c'ommodates system volume changes, maintains positive pressure on the system since it is located abeve the chitled water pump ano is used to 4 ;,-. ll l$ C 0 }f - Ef f. '
- l a3. ~ add water to the system. The chemical addition tank conq9$ns chemicals designed to prevent corrosion and scale buildup; chemical addition is made manually, as indicated by periodic water analysis. The ESCWS serves both safety-related and nonsafety-related systems which include the control room air conditioning system, CRACS, (see Section 9.4.1 of this SER for a discussion of the CRACS) Spent FWQg Pool Pump 4'," 4 Ventilation System, FPPRVS (see Section 9.4.2 of this SER for a discussion of the FPPRVS), and Reactor Auxiliary Building (RAS) subsystems. The R ubsystems include th'e,followigg:_ _,_3. ..,.,.,v,..,. - ~, y 1) RAB NN,,- Ventilation System (see Section 9.4.3 of this SER), 2) RAa ESF Equipment Cooling System (see section 9.4.5 of this SER), 3) RAS Switchgear Roo'ms Ventilation System (see Section ~ 9.4.5 of this SER), and 4) RAS Electrical. Equipment Protection Rooms Venti-Lation System (see Section 9.4.5 of this SER). The system is designed in compliance with OP:
- r'.e.
i. sentivo-o (=> seismic Category I requirements, thus cc ;'_23: with
Ins-Position C.1 of Regulatory Guide 1.29, " Seismic Design Classification." Part of the system is designed to operate during accidents; this safety-related portion of the ESCWS is automatically i.solated from the nonsafety-related portion, thus complying with Position C.2 of Regulatory Guide 1.29. In this way, the system complies with the seismic portion of General Design Criterion 2, " Design Bases for Protection Agains: Natural Phenomena." The ESCWS design requires a separate ESCUS (two chillers) per unit. While Unit 2 is under construction,..:ne applicant intends to supply temporary piping between the chilled piping systems of Unit 1 and 2 to provide chilled water for the Unit 2 control room; the ten-porary piping wi L L be removed when Unit 2 becomes operational and each unit wiLL have its own separate ESCWS. The applicant stated that the coolant f lo'w to Uni: 2 wilL be shut off, automatically, during post-accident operation, throughou: the Uni 2 construction period. We find,th'is acceptable and in compliance with the requirements of General Design criterion 5, " Sharing of Structures, Systems and Comconents." e ?.
== 1 - ygg q, The applicant has aume shown that the sa'fety-related portion of the ESCWS is capable of cooling the safety-related components after ar. accident under the most f WluCA1 foss of OMr hst!M) adverse conditions (includ*n*s * -4^;'-
- s i'u-M in compliance with General Design Criterion 44, " Coo
- Ling Water."
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Therefore, the appli-he.s As /- cantanne.cproposediuitableperiodictesting for this ow,L..... system in the Technical Specifications r v;!!A27 2 9 t h a t ths E'SCWS has been designed to. permit appropriate tinu % Nate/ t u.ff_ : pressure and functional testing i the requirements of General-Design Criterion 46, " Testing of Cooling Water System." The applicant stated't' hat the ESCWS has been designed 1 to provide accessibility for inspection. We find this acceptable and in compliance with General Design Criterion 45, " Inspection of Cooling Water System." l OS* M 'Y# Y UCk*10CA.$ O f Si$0ll$N wrs. QCCOVCh1Acf MfA Y Nf'YD*'fM hY W&nt hl/S/ kt*ESt.l}*/.CfY Y, YHW ~ON)~3.** h th O E 'S N f N/$ Y Ni j ]Q f W Vl W Y hte h G Y$ Q rtoac.c. &<rveil/once 7 kith krosi asella(14;fu !cw urc6/s, &,Cieht & Nwf-0W2..) o o & ^^= ""'* Me P0V!- a g, escues n n to cdp e*de H Cl* ycgidistornohah * *no w no.t 14e Escws 2 anke te Qf)Wgth PWC5t/W d entf envntef M rfa ap/yovtsu w +as Aca a1 yas Accdea vatn cu<.+.6f. *th! EGcnKa lsupphe$}treH,4H ove Act Hit'fe'r s Seure9, wan wkie4 ja enfuem the tr>furnuYMw m&..am Citahti ledev g,4p tk*.o, Avg nere.usNxg dpVro.s1Sie $wf bck t'Ayf, s wtutu is M ct 4 a ' - --.....---..--.z.
i p ~m y ---X, ~~ -as-Based on the foregoing we ememec conclude that the g essential services ch.Lled water system complies W with the provisions of-General Design C r i t e r i a,1," 43~r 2 ,44 and 46. C. _. ; ; meets.the acceptance criteria of 4 SRP Section 9.2.2. ^.- - pp- ' 3.. -.--; --
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v IA % 9.2.10 Waste Processing Building Cooling Water System The waste processir.g building cooling water system e - CD r"dbQ* 03 A RC (WPSCWS) was. reviewed in with Section A 9.2.2 of NUREG-0800 CSRP). An audit review of each of the areas Listed in the " Areas of Review" portion of the SRP section was performed according to the guidelines provided in the " Review Precedures" portion of the SRP section. Conformance with the 5 acceptance criteriar ;-.;;; -- :d 5.'.... -formt1 the basis for our evaluation of the waste processing building cooling water system with respect +to.the applicabte c.,_ sui UIOr.5 t* -t .;;i;... of 10 CFR Part 50. W%C G0S The 17355 is used to cool various cocoonents in the A waste procissing building. The systen serves as an Acr-rstr-i n t e rm erM a t a cetween the waste processing system (WPS) d and service water. system in order to assure the e=m' fthenSso n oC-t;:----- -f radioactive water within the plant. A The WP9CWS consists of we coe ing water heat exchang-erse two cooling water p mp a surge tank and associa-ted piping and instrument ion coolant from the pumos flows through the shelL i e of the heat exchanger
"i30= int waste processing building is shared by Units 1 and 2 and is served by the WPBCWS. The WPSCWS is not required for mitigation of transients or accidents or for plant shutdown; n6 credit is taRen 16r operation of this system under those c6nditions, thus complying with the requirements of General Design Criteria 5, " Sharing of St ructures, Sysi, ems and Components." Therefores tRe requirements of General Design Criterion 44, " Cooling Waters: General Design Criterion 45, " Inspection of ^ Cooling Water System," and General Design Criterion 46, " Testing of Cooling Water System" do not appty. The applicant has not shown how radioactive fluids are prevented from entering the service water systam in the event,o,f a Leak into the WPSCWS as.a result of a seismic event in or' der to ~ comply wit 5 the criteria of Position C.2 of-GDC 1.29, " Seismic Design Classifications" thus meeting the requirements of General Design Criterion 2, " Design Bases fo g. rotection Against Natural A*d g P h e n om e n a. " "tks s Cow.frst krill Yola t l y; n. u.. _ a" ""' Wf a &+Mk#--T&w-SnM m $ nut a s <ppGownt 6 Mi.s sd in Sec.ttCH ll b-Thereforer we menem$ conclude that the WP9CWS meets th* acceptance -~A-e- criteria of SRP Section 9.2.2. W. _ ' ' I,... ff :
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'I % 9.4.1 Control Room ar,a v.ntitatien Svst=m The control roo1 ar,ea ventilat4'on sy,st'em was* reviewed i n* accordance with Section 9.4.1 of NUREG-0300 (SRP). An audit review of each of the areas Listed in the " Areas of Review" portion of the SRP section was performed according ~ to the guidelines provided in th* " Review Procedures" por-tion of the SRP section. Conformance with the acceptance 0-t ce.P V O t 5& be{OQ criteria forned tnr asis for our ? valuation of the control g room area ventilation system with respect to the applicable regulations of 10 CFR Part 50. system, des emhOd.heaffh The control room area ventilation -- --as o control room air conditioning system (CRACS) is com osed of supply, exhausti purgesand emergency filtratlen systems. In addition *to the control room itself,, the CRACS serves the office areas relay and termination cabinet rooms, kitchen and sanitary facilities and the component cooling i water surge tank room. Units 1 and 2 share c ccamen ::e n trol room and the CRACS is designed as two identical and %c e n G3 redundant h ; ;.' ,......ni; each with sufficient capacity for [ the total control room. ...M Lc ::. ;.;- t ...u.. l
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- 1.r.n _ The supply subsystem mixes a smaLL amount of outside air l l a r g e r c a p a c,i t y n o r m a l r e c i r' c,u l a t i n g system.- with a 1 System balance is maintained by exhausting a flowrate equal ) to the outside intake aire less exfiltrat'en. The purge subsystem feeds a large capacity of outside air into the ~ recirculating portion of the normal subsystem and dis-charges a like amount of air to the atmosphere thro. ugh the unit stacks. The emergency subsystem takes a smalL amount of filtered outside air as makeup to the normal recirculat-ing systeep
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y-4 F AA,off,nn four 5{4;,3_ ,ee The normal supply ':- -'- subsystem consists of f C.~... ; ---- :'- .. ~ : d ^'. air handling ~ units with oamperse isolation valvesi filtersi cooling ano heating coilse fans and ductwork. Components are designed to comply with Quality Group Ci seismic Category 5+taded3s Ig except for a tornado protected intet check valve. The exhaust for the normal subsystem consists of two parattel I l t rains of a dam;:er and *f an which are nonseismic Category I. This small normal exhaust is not needed for emergency or safe shutdown operation. However, two seismic Category I l isolation valves, in seriese are provided.after the two trains unite to fo.m an outside exhaust. N ;. : : t* -'= .%,o- < t
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-m The purge system consists of two identical subsyst ems, one - for each half of the c'comon control r o o'm. Two butterfly valves, in parattel, and two smokw exhaust fans, in paraLLet, w',th a single duct connecting the valves and fanse serve as one subsystem - the valves are designed to seismic Category I, Quality Group C standards while the fans are designated non-Cz's safety angnot designed to seismic Category I or to Quality Group A, 9 or C standards. Smoke detectors in the control room wilL actuate an alarm so that the cont rol room 6perators ..can initiate the purge system. 'I'M cfPI tCM S Ct tg Yh,;{ .c= The emergency filtration system is initiated automatically Apek a kejfh ^ CCitAClt Rf*4G494(W.5) kduahon Mfagf>upon a safety injection signal or high c h l gr'i n e n A ^ concentration s $gnal. - L-- - re-P{p : I-* kmdtv" E 14taE# cadsfidhf M all isolatiun valves at the normit outside air intakes wilL close as weLL as the discharge valves in the purge and normal exhaust subsystems. Thus, air wiLL only be.taken in through infiltration until the operator opens one of the two air intakes in the emergency filtration system. In the event of a chlorine signali the operator wilL not open any air intaker.and if one is open, the operator wilL close it in ordec 'o completely isolate the control t room to prevent entrance of chloeine. In the event of signals e@s c audible alarm wilL indicate the need a chtorine CP93CCdtAS ftr the use of breathing e.......-+-cy the operators. A m grm e
\\ e i ) i rm .f;..J c ;.'.. :.. ; ;... '. :.. ---~it i l**'.... 'd: s, -+d' I I b8;n;'. '-'-: :-f : : - s r. I There are radiation monitors in each normal air intake and in 1 l each air inteke associated with the emergency filtration The cggy ) h ighcztid ;. C y d fC N Y $ N M. MT d system. A radiation signal in a normal air ntaxe wilL 4A isolati nJhe4 Intake and wiL L st art both trains of the emer- ~ gency filtratfon system and placY the CRACS into the recir-cutation mode.
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...::"; / :. r . :.g..., .i, ^ 77 '-'d e...essyi ey Ene apcr e. 7 .i... 'e t .,,,, s. e.. v.. e e * - a .u_ ) - ~_ m .u _e ? .6,. I C- :... '#' ' "* 8: .;;.d .d5;
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...aa. .aw.a....g '. h - .0^*d: 2 ^ '^t... I .u. u n.6 .,a,..,.. _r.,__... __s ..._.r__ e,.. -- t 4.. . e.. s ...r<; .6... 4, 4..: u ~ The emergency filtration systems consist of two identicale l parattet trains. Each train cont 4 ins two subtrains, also operating in parattel; the subtrain consi,sts'of a motor operated (butterfly) supply valve, a 100% centrifugal fan, and a motor operated (butterfly) discharge valve. One e v-r-w- ,w-r r ....~-m- .-r.,._,--....w e,.e
K, "_I[O-w, s. Insert 'I_. . ~ \\ .. / Howeve.ri in Section 6.4.3.'of the F S A R J t h'e, a p pl i can1: s'tatis thit 'i i
- f. both outside normal intakes wiLL be isolatcJ on a high radiation
~ signatpr or SIAS and that the emergency filtration system wiLL then f be put into operation. The applicant should clarify the operation i ~ j sequences used for each signal, i.e.r SIAS, chlorine or high radiation, and should justify each sequence on the basis of the , operator's safety and ability to continue to maintain the plant n a, safe condition. ._,. ~ __ s ,/ /. 3 /)* ^ / v t 'i
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~ 8 91-e subtriin duct in each train then unites with a subtrain single $ duct in the other train to form a singte duct; t.h a t duct contains a charcoal filtration unit consisting of a 'demister, two electric heating coils, a HEPA filterr a e charedal adsorber and a HEPA filters in series, Down- ~ stream of the Last filter there are crossovers between the two single ducts so that one charcoal filtration train and one. centrifugal fan can sumpty air to both halves of the common control rooms thus ccmplying with the 4-. e e 8 to ee e 4 4 9 e e 4 9 e e e e e e e
'ISA requirements of Position C.2.a of Regulatory Guide 1.52, "Designf Testing and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Units et Light-Water-Cooled Nuclear Power Plants." The emergency filtration systems are designed to seismic Category.I standards, tnus complying with Position C.2.c of Regulatory Guide 1.52. The charcoal filtration train in each emergency filtration train is designed to permit a vcLume flow rate of 4000 c fs, thus complying with Position C.2.f t.0 hic.h T"SCt4LF'4% *tY'Mt I'-low mit, h b 4 !%3:o f Regula tory..Guice 1.5 3 OQn 30 CCD CW. 3 The normal outside air int a ke s a re h usMiriitd lo
- .;
- . with wiremesh a
thus comptying with Position C.2.k of Regulatory
- screen, Guide 1.52.
The charcoat filter unit is designed to be removed and reptaced in two segments in compliance with Position C.2.j of Regulatory Guide 1.52. The applicant states ^ that the emergenpy filtration system conplies with the i. .V . require'ments :.t rhsition C.2.1 of Regulatory Guide 1.52; f Pcs,ition C d,.1 requir6s that the charcoal filtration unit '! ? T ' s leakage in accordance with ' shtuld show pressu.revboundary n >3ec$ionL.12$of[ ANSI,509-1976and ~ that Leakage tests be s la \\, 't .; '/ perform 2d in accord 37ct 4 th the provisions of Section 6 [g 'v. tc? ANS,I' 510-1973. + 9 s \\ 3 + t ?w t'. ,( rd' t ~' .-u'h & ~
- - - ~ -
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^ /pp The CRACS has redundant chlorine dett: tors'near the chlorine p55 steroge sito in addition to detectors in each normal and rw apiscos f Ao.c forridtd emergency air intake.4"'.... -_ two isolation valves in (4ch & ea ci, >0***"E 0 'Y '*' M* ' OC M *'! # "P"" duct-2ndak<desh,.s 2y 4 perroit ede ency p&,rer app 1' 7 w swe <ppga 9 l0 $ fVSKY Df Q .MM port juse u,aw g m
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^ auLevoMi p a M %ppqp,auzt thpdad of '=/ i Position C.14 of Regulatory Guide 1.78, "Assumptiens for Evaluating the Habitability of a Nuclear Power Plant Control Rocn Dur.ing a Postulated R ele a s eg"fd, kg g$lO CMd G Hazardous Chemicat ^ % wto%rg G u ide t.%, " Pre +ec. tion cA Aclev-Pou.xz e fbos Cchl (2eoen Opera 4ces Q 0sns 9 Gn. O c c.ie!c nh! C.' ,ne la {gagg. ri,e opma*r />rNy y 34 yst'.sle&J AlWS
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7 Mrp(xAarai9) '^ ,A a w s @ 9 P W IQD P' ^ Guie,e & 24 ^$ g-pctV g d1A &4 Oj ? & c 4 4 ($Q / VC - - ~ -- . : :... - -....... *h ap pl i ca nt (-ds e-n :- cocis 3 a er ed -H se re ons n e,eo c om p.1, a s,. og i1-iexfe,Wpn,Wjg tu 90 S i+ ion O S O 4 e-Sn ict+Cws, G u n cle ps3 w rti t.n ?- gg( 4 Saiz,$ftseMAS M !< M f 5.' ________________,.__..--.-,.._.---_-,.----..,--_----.._r-.-_,,-_.,--..,--..-c,.-,,,, --,-,..__---[-I.
C#XT ALL essential portions of the v7. e c; are designed in accor-dance with the standards for seismic Category I, Quality nod Group C standardsgere are Located in the control building which is a seismic Category I, flood-and tornado protected structurer which complies with the guidelines of Regulatory Guide 1.29, " Seismic Design Classification," Positipns C.1 and C.2r thus meeting the requirements of General Design criterion 2, " Design Bases for Protection Against Natural Phenomena." The CRACS is physically separated from high-energy systems and is thus protected from the mechanical and environ-mental effects of postulated pipe failures in h[g'h energy systems, the* redundant equipment is located in separate missite protected rooms provided with suitable drainage for protection against flooding due to mdderate energy . piping system failure and failure in nonseismic water systems. AlL CRACS air intakes are provided with tornade missile barriers. The*CRACS is designed to remove the heat generated in the areas sersed and maintain the environ-mental cor.ditions within the limitations of the equipment invcLved durin. aLL' operating modes including LOCA condi-t i o n s.$ns tr' .3N& $4.Y 4 ~T/M.fA ktlRYf 9 A Memnummmume Th e a p pl i c an t has d o f f......, y_, infma--_ ,nnw; w.., om;; ; ', ', .;;-==r; o sngw rns-5.; b Y Q &)fff $V h D $N $Ul N fkWbff O & C44cs Ufrtin Aelf1(At .Gr bcftdie 4tny O t h e d 4 x f ~ 14k lb 4.1GW Q 4tl0hk %) #& M6ti ad Nak+da f'& QU W 5 tb Anp/ aib 6]. A Q$ @45 e
-Is s Insert } Thereforer the CRACS satisfies the reduir'esents bf Pos'itidns C.'2.6 of Regulatory Guide 1.52 and the requirements of General Design Criterion 4, " Environmental and Missile Design Sc:es" (refer to Sections 3.5.1.1, 3.6.1 and 9.3.3 for protection of essential con-trol room area ventilation system components from internally generated missilese postulated failures i n piping systems and f internal floodings respectivel. S.9e ee 9 4 0 9 .O e e a S 4 v -.y--- --,----,-4.-, -ww, ,--,v-, e --,,,r----- -v, m,-,. - - - - - - - -..., ---w
- f S t.-
- - pyl-4ceset -
cHr t4W ~T+rr t~ 'ttr, CRACS ffas- / s ufficient capabi ity to cool o eat the conte com / / a ces ti maintairy the desired mb.ent conditions forsper-i / /\\ ~ \\ lonnel nd equ,f'pment under the'mo't adverse, condition,sr k r o 1 + ncludin'c a malfunctionine ha**ing - tem e (g. g. rp,..... ; rhw s u-n Mf5 30.4 P S h O _AM Como h e extra n e a t in.sgi n the summer orgteoling is su pp), Q s in the winter), a.. ....'.;e'-~
- ,.. -+
- 9 d'e-- w y -
ni-lags o ggg m $
_r.oes 1-.= - The appticant s': '.J ..;o c..Wr the effect of a pipe break (especially one con- ~~~ taining radioactive fluid) i acent isGS Cone.lude$ fintf +h/ fo >tfroh ps adj& & >Mt. area or build y 8
- rveM, i-- 3:_.
c... r, z -at----- e u ff i c i e n t l y af,.fcZid. tg' atuf. :cuC4 Jua/u. & we are.luds thahlp, me _eu d =: ae :.e e < ~ & conte =t n .oom ..ea. u -e
- u ---
>:<e-g. g... ; .J, s - . r -.,. -.. e. n.,4,, es..-g 4 ~ -.>---.; :c --e ->:.. u. 4:n a....," .-.-41 ~ ^ ......,.... ;1 Id." ~.$.5 .. ~
- '?
^ f.- -. ^.. ; ^. ' ; -.......... 7. ' r.. ^ - A. ff"' as=a .s... .a. ,.. -,; g--- ,,7 ...-...a .;g ..... -~. s s'_ .i e..,,.-- 2-~---- a'--a-N $20.L U~ .;.... p applicant z u.a.t. provide [ h r informa-tion to show that the CRAC3 i s capable of maintaining the f.nw te-o n mC5+ desired control room area e-
- .- :-- 'nder the most adverse conditionsfd WPf$
b
- -d;- *- -
'~~~C the requirements * ' Nn o f G en e r a l D e s i gn C r i t e r i o n 4 x Pd/d18tf S [nv/rt/ htFifb.[ f $&fhrG #f fff!Of*MrG *t h O'A f/kfk WctNiiS % 4 QU & Ch$ / l y ,,,-.--,--v. ..y, -..--,,e,.,-
- w. -,,,,.,
- 06 7 Tk' "' '. ? " e's-ada' - ;... 'J c ' ' * *,- s * ? m. 's---i
- 7 ;,,,,
e4 -.., -,,..-... -.,.,.....;.:.;.;. ;_ _zur Interconnections between the ductwork are provided with motor operated valves or Locked closed daspers. Duririg the construction of Unit 2 with Unit 1 in operation, a temporary wall wiLL be erected to isolate the uncompleted Unit 2 control room portion from Unit 1. The Unit 1 por-tion of the essential emergency subsystem of the CRACS has 4 sufficient capacity to perform its safety function, with redundancyr for the control room during the Unit 2 con-struction. Thus the design conforms to the requirements of General Design Criterion 5, " Sharing of Structures, Systems, and Components," and is acceptabLt. CRACS components are separated and redundant so that a single failure of an active component in a shared system or the single failure of an emergency power supply i r. ,each unit coincident with loss of offsite power cannot result in loss of the s~ystem's functional performance capability. During loss of offsite powere alL active ofm e1Pe r ccmponents, such as valve and eea4+a coerators, fan i 4 motorse control and instrumentation witt be served by their etrec;.rw redundant emergency power sources. o 9 ~.--,,an._ n., ,-.-7.- .,.,,,,,n e
~ t ss --- ~ u;.
- a-
- r
- __
- a-
.......... -.;; ::: : 5y --- : c' .... '. : : '. : n - Tsased on the foregoing, we conclude that the CRACS complies with the positions of Regulatory Guide 1.29 and criteria of General Design criteria 2,4 and 5. u-"*"-'- the applicant Itas I mu.s.1 p r o v i d ed ' " ' : r information to show that the CRACS co ret P h t s d::: -- ;'._with Positions C.7 and C.14 of Regulatory Guide 1.78 relating to isolation of the control room a n d -( Iw,, .Tccer-T8t' 4es al.So single failure criterion) W 7he applicant mee.t provided A homew information to show compliance of the CRACS with C*4.4 I P o s i t i o n s t':+ a nd C. 4.d o f R e gu l a t o r y Guide 1.SS..relatin / GM h nq QfbN to exfiltration rate and initiation of operatio of tha* 'futnfffht$.. f charcoal filtration unit .-:-d:r the requireoents ) of General Design Criterion 19, " Control Room." A s .The acelican-u:: ..sw tnat th S Ts acabte of mai - I taining s u i t a b t. e ambie t ohd - nder the most / / adverse circumstin i, rder to empp fully with the requirements General 6'esign riterion 4, "Environnectul T- . y-- and fiissi e De Bases."_\\ N 9' u.m ,J_1 t -_r-e...,y. w. -.s . i p, if _., r. e i r '-;% 1. v % - a .C 9~t
- ,+ c,,aL*s y 4 x %L,'bf - *' ~Q.~
_-..__.n s-Ar e..- .m ~'y'," 4 - _Q y ; f" am. . / -- of the CRA coaTrLies wit .. ~.. - T e enarpo.T.., h.f i l t r' t ion / un i t a / ,,,,p e quirements of,R[siyions C'. ,C I bi C.2.c, C .f.. .ar / j s , ser +,The dh 4-k, shy ff A> E 2 \\ d as &
- n. c.5,re t cm
.2 n re leo b o, edoc chsen
- a. y p
-? W 'N &T~u E m .-.. c,s ..,e
Insert / N \\ The a plicant au show that the CRACS is ca able f maintaining suitabl ambient con 'tions under the most ad erse dircumstances in order t comply fully +3 th the r e q u i r e m e n t s 'o, f G e n e r, a l Dhsign C riter on 4, "Environme tal and Missile Design Ehes." r. The applicant must clarify the ope. onal sequences for the CRACS for each sig and should justif each sequence,on the \\ basis of the operator's afety and ability.c maintain the plant \\ in a safe condition. _ l Insert 4 The applicant has s'hown compliance with Position C.3 of Regulatory Guide 1.78 relating to release of hazardous chemicals. e e e 1 I
~ -Is b_ Y./ CA W Y 'M1 W Y / Q C 2.Q, C. 2./y c. 1.c, c. g,f,, C.2.j, C.2.k, and C.2.,1 of Regulatony Guice 1.52. The charcoal filtration unit is used only for emergenciesy thus)(,is not required to comply with the requirements of Regulatory Guide 1.140, " Design, Testing and Maintenance Criteria for Nor=al Ventilation Exhaust System Air Filtra- ~ tion and Adsorption Units of Light-Wate.--Cooted NueLear Power Plants." Therefore, the CRACS meets the requirements of General Design criterior 60, " Control of Releases of 'I'h2 C<A cs a Radioactive Materials to the Environment." T ;.. ;;;'2 :- n t erst ;: 'd;~.;. . :. a '.- 4 ; : : ' ' ' : : - * - - - 'c e-p# .. '. ; ' : n / T:T:7'
- "a
--"--at aa-*
- -d--
t. .,.t., m e a a r. A n':. ? .iew sa.+.,
- "t t: ;.;.
- o '_
The CR,ACS de:; ._r meetsthe acceptance criterian d ,_,, /f s. _ f.,, _ ; OI of SRP Section 9.4.1. w, s ~ s-F vu w w i w.= - ^ < - ri r wue in o c une e ~.%. l. ' z +L.e c: n o w w -Ii46' = .., sq \\, gy6 m y w, e .e o I
3. '-;1:2 f., 10.4.9 Aux 4Liary Feedwater.svs. tem _ ~ The auxiliary feedwater system was reviewed in accorcance with Section 10.4.9 o f NUREG-0300 (SRP). An audit review of each of the areas listed in tne " Areas of Review" portien of the SRP section was performed according to.the guidelines provided in the "Rev,iew Precedures" portion of the SRP section. Conformance with the acceptance criteria formed the basis fer our evaluation of the auxiliary feedwa'te. system with respect to the applicable regulations of 10 CFR 50. The design f the a xiliary feedwater syste. was reNiewed in acc dance with S. 0.4.9 (NURTGP 00). Confo. a n,c'e ith th acce tance cri ria ormed t. basis for -he staff e untion of the sign of t e auxi iacy fee at e system ith re pect to -h appli-ble regulat'ons f 10 CFR Pa. 50. The staff reviewed t,b.e auxiliary feedwater system ( A F'J S ) against the specific acceptance criteria of SRP 10.4.9 as folLows: (1) GDC 2 as re ated to structures housing the system and the system itself being capable of w1thstanding the effects of earthquakes. Acceptability is base.d, on meeting Position C.1 of RG 1.29 for safety-related portions and position C.2 for nonsafety-6 l
' - D g, ~ related portions. (2) GDC 4 with respec': to st uctures housing the system and the system itself being capable of withstanding the ef fects of external missiles and internally generated missilis, pipe whip, and j et impingement A. forces associg:ed with pipe breaks. The basis for acceptance for this criterion is set forth in SRP A YfdSW 3.5 and 3.6. (3) GDC 5 as related to the capacility of shared systems and components important to safety to perform required sarety functions. (4) GDC 19 as related to the cesign capability of system instrumentation and con:rets f o'r','p r o m p t hot shut. gown of the reactor and potential capability for subsequent cold shutdown. Acceptance is based on meeting BTP RSS 5-1, with regard to cold shut-down from the control room using only safety-related ecuipment. (5) GDC34and44k.toensurethe capability to transfer / heat loads from the reactor system to a heat sink unoer both normal operatirg and accident cenditiens; redundancy of components so that under accident conditions the safety function can be performed assuming a singlcactive component failure (this may, be coincident with the loss of offsite power for T. e + .,-.,,-,.,a
4 s, certain events); and th'e capab'iLity t'o isoL'te a components, subsystems, or piping if required so that the system safety function wiLL be maintained. (6) GDC 45 as related to design provisions made to permit periodic inservice inspection of system components and equipment (7) GDC 46 as related to design provisions made to permit apprcpriate functional testing of the system and components to ensure structural integrity and-Leaktightness, operability and per-fermance of active components, and capability of the integrated system to function as i ntended during normal, shutdown, and accident conditions. The following evaluation discusses the implementation of these acceptance criteria and foilows the order of SRP 10.4.9 (NUREG-0800). This evaluation also incer-perates the staff review of the apolicant's respense to NUREG-0737 Item II.f.'1.1. This includes (1) ao evaluation against the deterministic criteria of t h e *dEE Man $ As d eVi e.w/ IAU (2) an evaluation against the generic recommendations of NUKEG-0611 (3) the evaluation of system reliability based on the applicant's reliability study ( ,,.,.9.# ,..e., .s .m,- _.,,.mm.- ,m
.~ 4 9. % (4) an evaluation of the design-basis for the flow caoability for the system. The auxiliary feedwater (AFW) system is designed to e. supply fesdwater to the secondary side of the steam generators when the normat feedwater system is det availabl'e, thereby maintaining the heat sink capabilities of the steam generator. The system is also used during start up, het standby and cooldown and is considered to be an engineered safeguards system. As such, it is directly relied upon to aid in preventing core damage in the event of transients such as Loss of nermal feed-water or a secondary system pipe rupture. ~'e system
- p ew s, A.v M e w towfe tN t M ol#MY consists of twoTaotor-driven puinps =J m
A*;*4fff fie/ mrv Ndnf CodetML det!NU'&:c -:nwry pApp. inc4 frefg Gedent.fu t : n - ~... associated valves, piping, c o n t r o l s y. and instrumentation. The two motor-driven pumps are powered from two separate buses of emergency onsite electrical power and discharge into a common header '~ with three independent Lines each leading to one of the three steam generators. Each of thess Lines centains check valves, motor-cperated isolation valves, and flow controt valves. The steam turbine-driv'en pump suoplies a d/W[#r Lri fh three separate Lines, one for each steam generator. checkvalves, motor-[' Each of these Lines also contains operated isolation valves, and flow control valves. .=
.'" EL;> 5. The steam for the turbine is'supptfed fros two s' team Pak generators (B and C); it is extracted f rom ame main Line upstream of tha main isclation valve)(. steam The control system wiLL automatically isolate any steam generator affected by a steam Line or feedwater Line break. Any pu=p wilL be able to feed the remaining two steam generators. Each of the two units has its own AFWS with the water supply provided by the condensate storage tank (seismic Category I, Quality Group C) through a seismic Category C I, Guality GroupAsucticn Line that contains check valves and' Locked open manually operated isolation valves. lince.there is a separate.AFW system for each unit, Genera L Design criterion 5, " Sharing of Structures, Systems and Components" is not applicable. 25 The AFWS is designed ;<f seismic Category I, Quality Group C f rom the CSI'un to but not including the motor-cperated isolation valves. The motor-operated isolation valves and the pining and valves frem the motor-operated valves to the steam generator are designed to seismic Category I, Quality Group S standards. Thus, the AFWS - complies with the guidelines of Position C. of Regu-Latory Guide 1.29, Seismic Design Classification." .---,___-,,..-wew-,_ ,,----_,.-r P-**.- ~ ~ * - - - - -y w---,. .-m--,
.\\ %gg . ; J.'{ i}. .i "~u.; , sl y' ' V" J There.we separate cubicleX.for each AFW cump in orde T-to prevent possible internally =ner ted missiles from he rvrVir,& c.ube cHg dets not cou,/gle &rrenW4$ UMP. damaging more than one pump. The AF S is located within }A the reactor auxiliary huilding and containment building and is thus protected against the effects of natural ~ phenomena and tornado missiles (the CST is located outside'cf the buildings but is protected against the effects of hurrica.es, tornadoes and missiles by a concrete enclosure). Thus, the AFW system meets the requirements of General Design Criteria 2, " Design Basis for Protection Against Natural ?henomena," and f 4, "Envi rc' menta l and Mis sile Desi gn Ea e s",,(see Sections 3,.5.1.1, 3.5.1.'2, and 3.6.1 for proi:ection against missiles outside and inside of containment and for protection against high and modsrate energy Line breaks, respectively). t / ^ z'.'r',.- ( hours,with I, The AFW system can be ocerated -fer approximately eight 1 + ......s i n _ _. ---n,,#_ . - m-s eg V l the water (240,000 gallons) in the conden- . g sate storage tank. Thereafter, water from the auxiliary dan i s available via the emergency service water system. .\\ c ', f 2,# T... a .2 f:- t': .y ert:.J.
- -dad a'
The energency service water L I system (see Section 9.2.1) issafety p M, single-IC.[4.Tod f f 1 .. -. -. -- --- --. - ------. - -- - --- ~~'-
i ~ % M1] w failure proo,f and can be operated,fr,o,m the controf ,-..-w,- ,...i m rooms ::.x.;;;.7,. Therefore, the AFW system coeplies with the guidelines of BTP RSS 5-1 and the requirements of General Design Criterion 19, " Control Room." . g is now itear from the PSAR that the AFWS ha s "th e; apability : transfer de ay heat leads from the i t i sactor to he secondaryf(steam) systeg'under aLL con-L itions. .f a b L e 10.4.9-1 shows that th'e motor drive'n .-~ i
- lumps are rated at 450 gpm while the turbine driven!
dump is rated at 900 gem. Yet, in Section 10.4.9-1.of the FSAR, the applicant states that the mot.o.r-driver dumps are capable of delivering 475 som and'tha: 47 i i-igpm i s r e'q ui r e d to mitigate a loss of main feedwater 4 .ransient. It is also noted that the flow rate shoin t Includes some, water for recirculation. The applican t 2 ..rG
==it+... ...;; th:: ; b ATL _- fgg!(gf > s y s t em comoties with General Design Criteria 34, " Decay j ~ Heat Removal" and 44, " Cooling Water." The AFW system has been designed to permit periodic tasting. In addition, the applicant ha's propcsed periodic tests in conformance with the Standard Technical Specifications for Westinghouse Pressurized
jq y The applicant calculated that the greatest demand upon the AFWS was the loss of main feedwater event (LOMF) which required that the AFWS supply 475 gpm to the steam generator (s) in order to mitigate the LOMF event. The applicant assumed that two of the three AFW pumps, the two motor driven pumps, started and delivered water to the steam generators within two minutes (in about one minute after the reactor tripped and the AFW system pumps received a signal to start on steam generator low-low Level), thus mit4-gating the transient in a timely manner. It - h.. '. d L. nasca ca.c applicant provided an AFW pump curve which i s based upon tests . A which showed that one motor driven pump could supply enough j nntesi water (490 gpm) to mitigate the LOMF transient. A is the recircula U$0 W S powevere pet when the LOMF incident is initiated, -{%Lt a V40.VP l g tion Lines are opens to protect.the pumps in the event S n\\ AP f tqP % Ke 5 9 PW/1 h 'Ast* MW )!YM GMFkk _ _ - - -
- __g__,
- r y y g____.__
G In that caser approximately 50 gom flows back hv tha *"" to the CST, leaving 440 gpm for the steam generators from one pump. H= : ; : ~- i_f one of the motor driven pumps becomes unavail-6 - able because of loss of power ;; '..., : ) the recirculation Line for j the other pump closes autonaticalLy and 490 gom is fed to the steam generators. In addition, the recirculation Line valves can be closed from the control room or from the auxiliary control panet. Therefores the AFW s er w,w* ww
==ew-----w
v-w-e-w--
-m w --w-- -w
A Y, g Water Reactors, flUREG-0452. ' This Eeet5 t.$e re:;uirements "of General Design Criterion 46, " Testing of Cooling Water System." The AFW system has been designed to permit inservice t inspection and periodic inspection o7" valves and cumos, thus meeting the requirements of General Design Cri-terion 45, " Inspection of Cooling Water System." c. The AFW sysupi has two power sources which consist of offsite or onsite (Class 1E) AC power for the motor driven pumps and steam for the turbine driven pump. There are,no auxiliaries in the train'for tHe turbine-driven pump which require AC power to maintain operation of the train. This meets the guidelines of STP ASS 10-1. I 4 [h t_ a p u y _so designed that the turbine driven - am is $ # 2 pump portioil of the systee 5 ~ ~ I Lated from the MP$9D W PQC]s ['ft0Jf xg b .ty t ff t istk f >) CHf f & tfa & W h $codtainin-4RK5&y,drMMfC/t Nt 6 lt Elbr/ fotrw f4/c;lh Tp* ocrtion
- f d
'"2" cen1s. in udition I ea-:h portlen (motor dr1ven pumps or turbine driven = ~ q punp) supplies water to a separate header which, in l turn, supplies all three steam genera
- However, 4.
tne applicant has not provided inTormation to show tha't i. / 1-3- m ~lE
go Insert 2 1
- Mr Ir?
T'.1 e A FW S h a s b e e n designed so that the piping of the ;;r;;vn con-taining the motor driven (MD) pumps is separate from the piping treat of the ;::::M... containing the turbine driven (TD) pump, except c/M977 'W ~1L'H31n Line from the TD Y1@ly rkin one line from the MD ; -t':3 merges with a e na**8-- to form a single Line to inject fluid into the normal feedwater Line and then into the steam generator. This common valvex in e$dt Em/ d M d:ScM4r fid L i n e.,"--we i s d o wn s t r e a m of check s ach AFWS
- e-**
- -
g so that a failure of any part will only affect one steam generator. In the case of a broken feedwater or steam line, the affected steam generator wilL be isolated and alL the operating pumps in , the AFWS wiLL be able to pump water into the unaffected steam generators. The MD pumps feed into a common header which then separates into three branch liness one for each steam generator. The TD pump feeds into a separate header'also with three branch lines. Each of these six lines is provided with a normalLy-v. open r: --'ed isolation valve i?r..:::,w with power from one emer-gency bus and a flow control valve powered from another emergency bus. Each of these lines also contains a check valve (asnotedAbcVN) +144Wem so that a break in a feedwater Line or steam Line wilL not prevent the AFWS from feeding the unaffected steam generators. Thuss an internal breaks a break in any line between a pump and 9 the check valve wilL cause that portion to become inAffectual since most of the pumped water wilL spill through the break, leaving the other portion (either the TD or MD pump portion) with the flow capability almost 100% in excess of the flow necessary l
M
==< ty$t/O" L, G9A cLu2gD, l to mitigate a transient or accident $ i.e., each portion can provide almost 200% of the needed flow without operator intervention. In the event of a break in either portion of the AFWSs the opera-tor would have at least two hours to discover and isolate en internal break before running out of water in the primary water level,ffhetime supptys the CST, with the CST at the minimum effMDefheh*']f would be extended to m, five hours with the CST full and with the portion of the system providing water to the steam generators controlled to provide the minimum required flow ofwatejh, 3 s l l i e ,,a,, v.~-, n _...-,-..-------cn_. ..,-a.,_,,----,,-,m, ~,,, _,, -, _..., -..,..,,,.,, - -.., - -,
p Ian epe tor can d c: and can i . ate an AFW ine O be ak downstr a of the ch valve in ficient tim w to preve daa: age to e plant. J ~/ / ha1 vivt$& The applicant - _ i d y ' f, that a rupture of the steam Line to the turbine of the AFW turbine-driven pump, (frola or* Of 4 konee~ dt*V!" P% coupled with the most adverse single failure"would stilL reewasm=p. bder kW>t pkhe m Tier permit the AFW system to provide the necessary flow of McC& W C";L,ar;l eVr^rf6fgtotheun,dacage,d_steamgeneratorM (; cool t 9 mf 9W/WMr faautecenow granc,ey 4.. 22: ;-- ae - i.. w - - g ;,- 'i,,, ;,,,,,g ; is any i 6 w.......; L & n! fel.h & ' l >r a// hi/ Jo.2<,psear9t x, 4 a ^ h/ C>0V!$fl uk>rfy!Wb / Pf' J .ha - --Tiremp'pti c-a rrt ha s eva lua t e d t h e,A FV s y s t em a ga i n s t the short and long term recommendations of Nt! REG-06M, " Generic Evaluation of Feedwater Transients and Sma L L Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plar.ts." We have reviewed the apclicant's evaluation. The*results of our review are discussed below: TheTechnica(Soecificationsproposedby the applicant for the AFW syste= do not include a specification for cutage times for two (2) and three (3) inaperable AF'# untlllM9W)wo*W fo pumps. The apolicant C :... include such outage tipes W A w f &s[g t l~ k N A t t A l* rfe 4pf} 5 ~NYh*** /es4tf 3 1 Mukal 6 koteafAw;j&' tW h,.
- v e e v &e H e o j i c o t
r v n tn+6,.n u,yww.<my a$5 & 4t*k* rw 68 &ke Af w maior &r w>, 9u* tay envisear-.
- f6 <c' b
e m n fsw Hent bm- /fd'n hs
- ua[d,','9 "ag,,,
vn ,ea ey
. _ _. _... _. _. 1 00 .h (( in the Shearon liarris Technical Specifications; these ' should be consistent with the Standard Technical Speci ~ ficationsforWestinghousePWRplants,NUREG-0452.h Y f g MGP0H kFV/S NelfN ? ft te h e k f.h M Sctpt* H ff/te g g g Q f ln C$u df do Y ^:^Fie
- Y@JUWNW W g er The applicant has proposed Technical Specifications for b
EN surveillance of Locked valves consistant with those in the Standard Technical Specifications. 1 l The appli cant stated that the AFW system is designed to supply water to the steam generators without flew throttling, thus avoiding throttling as a petential t scurce of' water hammer. Water hammer is prevented by f-Lines futt of water prior to AFW system startup. The 6 \\ system depends upon the trickle flow from the m'ain feedwater system through the AFW noz: Le on the steam genera: ors to keep this portion of pipe full of water. The applicant stated that the pressure of the CST wiLL keep the suction side of the AFW system full be:ause the CST is at a higher elevation; leaks through pumas and valves wiLL maintain the pipelines on the discharge of the AFW pumps futt of water. F -"-, the applicant ,= ..e;..._..._.- t'!t ' t...ausissu n '. 3..,, ; '. t -
- ....ww 6 rap alr wu M C S " '. d a '. 0
_It** b**" e e n AFW
==omm
- t : t_a p.
p =
- 4 Akt,pdp puti an opeuf i,9
~ L 4 % M,W A t mgp as .._...,.._m,_.,_-.. .we,
~ ' 7 y The applicant stated that safety gr'ade'indicatio'ns an'd - alarms are provided for low AFW pump suctionpressurg (Od there are control switches which can be operated from the control room to switch the pump suction to the ~ alternate safety-related source of water, the erergency service water system (ESWS). In addition, the sopti-cant stated that operating scocedures include alignment of the AFW system suction to the ESWS and that pro-cedures wilL identify the steps to assure AFW system operability after transfer. Mc_ ; ; -. -- - t -'s- -4
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Operating and maintenance procedures wilL require the operator to check proper alignment of valves; a second,
- puali fied individual wi L L verify the valve alignment.
i h4% rse.s.s.acsurefusthat l The apolicant post-test procedures recuire such valve alignment checks. The a:plicant must provide Technical Specifications te, require a flow test of the AFW system, prior to startup after any extend cold shutdewn, to verify the normal M AFW W g f3 fg systemg ath from the CST to the steam generators. NWf&Y h $14 Art Ydi sc4 e & Udk A2farve.Ami 4 A S f41) JMtM6 7$ 2dulW NAlak) A QM 4. k s u l Md &Pfw% - . =
1 $3 w The periodic, 18 month test, is not sufficient for such verification. h/ M I c The primary source of water for the AFW system, the CST, has low Level alarms to permit the operator to shift M to an alternate water supply. There are safety-related redundant Level incicators with readouts in the control lot fit PGCh Ind CQ$0V ff.CGIV/*g 4 Q ^" "'"2 roong TM *- '4-=~* m f0Vttif ':: Y. gVtt$f $fVfhc.,fVQJrD99i W Y e k 32 f M(,( $~ YC he v im.... wot Sf ': m...n; .nc ; 2 en. ~.<. m o., Q G f (t.33 fVirt H A ev'CKlin B). Thp ffatn s n<ttf ..,-,.,.m % m e a m n er'osstVQinI d~fobf $$dPn3 4Y~ d/ Wt9M10Y.bFW kitW kffkJ :(!) leW-wty - 28 g$M, .a . m,. n x e seu-e before causing d= e;e te t5: '~;=s. $) & ~h-2A C OtFC f /YWN/HI%ng kV* WhGkW d ypcJ,mcaf yecikak,ov, suats) En, hf 'irsw peXM CMPi"td p',,. @ The aopticant has conmitted to concuct a 48 hour AFW reNF* pump test. The applicant should conduct this test on Q S 20 all three AFW pumps and should provide us with the /D SM//C/1 test information specified in Section 3.1.4.2 of NUREG-f5 4 kJ gpg 0611. We find this acceptable, pending results of the tests. Indication of flew and automatic initiation of the AFW system are discussed in Section 7.3.3 of this SER. l' N @A applicant : 'd prevent damage to the AFW pumps in the event of failure of the Locked meen manually operatec 15ygQQ [$f & glQ/tp&gggfghQ gh /gfy 4 ,2wel tharfJ sy afe%4 s%$ " n d7 &M 0 aq,'/to$ci ufdb be d H M Ti Y . 4,acQL & GlaM4 h r2h - n g =.
~ o99 p o-p ~4 9 W Lin / rom t. CST. S+ee-gate valve in the single f ibm 8 & h dl$h4ttAf N Mt2%'& (Ms ' kW!!'L e r o b t e a_a cm ' a r a " 3" Af r
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-a fnDe Mugl9gAnd(Nxcy) e av Wpd.' fa af f a k d,cn b 1Asi,$ g w w A y & p aa p # MNh' '%Ca' }} &,4'x h 7 $jp } 1-7 <tp'L t % w . r e c u n c a rrr _ vark-d :;; c, ..L,, I in parstlet with the present valve, 2 Provide for automati opening of the valves from ? the alternate ' water supply, as necessary, f h' Low suction trip of aLL AFW pumps if it can be i j shown that this wiLL prevent pump damage, '[' '4. Removal of internals of thesinglevalveinthj. 'n. w single flow path, or ~ 5.( Locking open all valves from the secondary wate r g.- %s u p p ly_,* c.A L4-A FW vus#5. ~ In the event of loss of atL AC power the turbine-driven AFW system pump wilL start automatically to feed water to the three steam generatort and wiLL centinue for at least two hours. There is no auxiliary system required fine this /gp to maintain operation of this cump train. We acceptable.
- p. view of the foregoing we find the appOi:7nt 't*ees ws.t.
'y H t h- ;he i-ctt0trfny"re commendaMuns 'et Mt! REG-061 th. w ._m. l =
~- i -g n ( 1. Technical Specifications f r-outage times with 2 - /, / and 3 pumps inoperable, )( 2. The possibility of water hammer bgeause of trapped air h high points, / 3. Procedur'6 for startup in the event t CST is unavailable \\, N Provioe for \\ \\ vatve alignment checks after tehting, . 4 \\.. / 5. Technical Specif,ations for flow testing af ek e i extended ccLd shutdh n to verify the normal flod x
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/ N d V' [ $. Ashprance that there are redundant alarms to indi-CST; the operator.us: have ^ %levelinthe ~ cate at least 20 'nutes, thereafter, to swit'ch to a I s I new supply, $ ;7f',The applican; shoqld provide sooe means for pro-N / tecting the AFW p um;iiii, i n t h e ev ent of failure of // the single valve in the ingle sucoly/Line\\from s / l Qe CST. Y I The applicant reviewed the fotLowing events requiring AFW initiation to determine the basis for AfW flow i requirements: i 1. Loss of main feedwater transient f . i a. with off-site power available \\ 1 j b. without off-site power available
~ f ]O e S. 2. Secondary system pice ruptures a. feedline rupture b. steamline rupture 3. Loss of coolant accident (LOCA) 4. Loss of atL AC power 5. Cooldown The applicant determined that the event r e q u i r i e.g the maximum input from the AFW system is the loss-of- ..~ ~ feedwater (LOFW); this requires that the plant provide a minimum flow of 475 gom. We have asked the applicant,
- k? 4fSbCQ n Y noted above, se provideg /bt
[4 assurance that t.h.is require-4 ment would be met. 1 'The appli Ant, in re y nse to our cc ern, revised the AFW sys e's figure (PSID 10.1.0-3) to ' include ghe proper ~ / l valving. However, t,f. e a p p l i c a n t has not explained the / symbols used for the valve ocerators, as noted above, j= l which do not aopear on the flow (jagra= legend (PSIDs 1.1.1-1 and 1.1.p-2). In view of the foregoing, we find that'the AFW system complies with the recuirements of Regulatory Guide 1.29 and General Design Criteria 2, 4, 5, 19, 45, and ~. wm-
y A ~., m-70 ~ 46 with regard to seismic crit eria p rot e ction a ssinst-natural phenomena and missiles, AFW system sharing { i between units, capability to aid in shutdown with system operated from the control room, inspection and W "fi$ A CArtAS4 /Al'O<lftb Th G testing of the AFW sy' stem. -,-fmmener{infor-pppso? W N 4 +krf+kPAfw3 tTis $ mation ar* ':- --- 'd:d to show c c ; '. :. : :
- 5 General j
i= a-d" f^ I +kus, ccw Design criteria 34 and 44 y with BTP In addition,,/4t*4kf A>: 14t* Qf J/cw",, & ASB 10-1. the AFW systen co y 6c r a with the reccamendations of NURSG-C611, "': : ' " ~ p Lt T-~
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NUREG-0737, "C la ri fi ca ti on o f T*4I A c ti on P lan R equi re-4 ments," Item II.E.1.1 " Auxiliary F e e d w a t e r 'S,'y's't e m EvaluatioE" requires a reliability analysis of the AFWS. The applicant submitted this analysis for staff review. Cur review is not,e completed. e4 In view of the foregoing, we conclude that the AFW system meets the acceotance criteria of SRP Section R f VGStrif$ Of NG Jt'Vlfte 0 W1/ Mbe A't/t& t 10.4.9. We wiLL report
- : ; '. ; i c.- cf : 5::c i : :;.. ; -44.,
U pr yg /pf a suoclement to tnis SER. QU \\ .j g) f 1 e. e O -}}