ML20083D167
| ML20083D167 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 09/20/1991 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20083D165 | List: |
| References | |
| NUDOCS 9109300071 | |
| Download: ML20083D167 (72) | |
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GE NUCLEAR ENERGY TIER 1 DESIGN CERTIFICATION MATERIAL PILOT ITAAC EXAMPLES FOR THE GE ABWR DESIGN SEPTEMBER 20,1991 9109300071 910920
((DR ADOCK 05000605 PDR
ABWR Design Document TABLE OF CONTENTS
- 1. Introcluction
- 2. Pilot Examples of Tier 1.\taterial 2.1.1 Reactor Vessel anci Internals Systern 2.2.4 Stan(thy 1.iquici Control System 2.2.7 Reactor Protet tion System 2.4.1 Resiclual IIcat Removal System 2.6.1 Reactor Water Cleanup System 2.10.23 Circulating Water System 2.11.3 Reactor Cooling Water Systern System 2.12.13 Emergency Diesel Generator System (Stantiby AC Power Supply) 2.15.12 Contr ol lluilcling
ABWR Design Decument
1.0 INTRODUCTION
l he p uin,se of clus nn inoraniinin is to pinent unic pilot nainpin ni jaltelHssrtl 'ller i drugli ( t flil'u aluni stiatollai Ills thf* ('l Ald 'N. l' a( li lallt it e
starts with a proposed ' lier i dnign <tru ription atul then pri snin suggnied nopr< tions, ents, anab sn, .nni at e rplatur c riteria (II ATC) nased on the drugn deu ription c ontenis. t he u npr of thne pdots n based on the nn< trar ituluttn's c ur rent untientatohng of ihr guiding penn ipin g<nettung u ope aint i ontent of lier I inatorial. ~l hese pa nnipin are pri soliteil in the NL'.\lARC report of the lask l'on e on Inspec tions, Ints. Anabsn avul At t eptanic Criersia. NL'\l ARC 9015. Det rinher 1990.
It is intrinted that this pilot set ter datobuted to the NRC anit other ititritated parlars Itir f rsirw alul < <sttillioni. llasral oli It e tlliat k leistti tlits truen, Gl' will (il oca ruan) inodih the u ope and inntent guulehnn and will then niitiate perparatnin of a full set of AllWR dnign dru ripuons and
!! AAC for unhinittal to NRC as part of the AllWR drugn v eredh ation a ppla ation.
i l
1 Intro 1 9/20/91
ABWR ocsign occument 2.0 PILOT EXAMPLES OF TIER 1 MATERIAL
~1 his sn tion proudes inlot cumples of lier i design dru siptions and their asua ietni l'1 A \C. I hesc inlots base been < hosen with the objn tne of I osoliflg f rpf ru'sttalts-c CXatnjilCs id fitn l Car Islalld IllC< balilt al M sf rills, turbine island inn hanic al ss strun, strui tuaal/<iul itenn, c in un al o stenn, c ontrol and untiumentation osiena. lhe u < non nutuhrinig usicm is based on the AllWR produc t strut ture.
FIGURE For a number of the !! AAC, simphlied figures base been ituludni to INFORMATION help f.u ihiate the design deuription. 't he figute* c ontain ini'ormation that uses ihr following c ornentiorn:
1.ille < lassklic allof t ASME Code Class 1 ----------
ASME Code Class 2 - - - - - - - - - -
ASME Code Class 3 Non ASME code - - - - - - - - -
lustruinentation Flow riement FI'.
Restric ting orific c Rti Irmperature element ' I F.
Radiation element RI:
1.esel indic ator I, Picssure indicator P 1:<[uipinent Gate valve >(
Globe valve $$(
Check valve [/T Valve type not specified B Relief valve Open cirucit breaker I i I
Closed circuit breaker M I
Vahes are shown on the figures in their normal postition.
I l
t l lMfo 2- 9/20/91 l
L
ABWR D:sion Document 1
2.1,* Fleactor Pressure Vessel System Design t he tra< los pe ruuto seuil mtrin (RPV'u a olants of (l) the trat tot pic uurr Description stuel and ih apput tonaiu tt suppot h aini unulation. .unt m ihr t rac toi inteluals rtu losed by the truel, est ludung the c iere, ins ide flut trat untt uttarntatnui, tr.u tor itFot nal putnln and c onttol tod de not t he trat for t oolant lHeuule boundan (Rt:l'll) portton of the RI'VN triann Itttrgrth ladn t.H lo r tilairrial bJt fit i during rupttfial (sl4 ratnui aint f ollowing abnotinal opetational tr arnarnh arni design h,nn at i nletits l
1
('.o r t a ill t r.<( t a il lH ouurr stuel (NpV) slitrif tals utpl Hill the (sito, t'aNul the c ore dutilig a drugte basis at < nletit, atul sul lv<il niste uenclatainers utilitrd i dusing a ih sign b.nis at < ident. ()thr RPV inter tials diert t t oolant flow, j srparate strain, hold tnatorial suncillain e spet ituren aint suppot t instr unirfilation utih/rd lot not tual oprf allon.
'l hr to,u tor piruurr gruel sutett* proudrs guidarn o atid suppor t los the (ontr ol tod dr evet It aho adtnits atol dntributes the sodiurn protabotate ,
f roen Ihr statalhv hquid c ontrol mtiin. '
t he RPV svuren erstrairn the CRI) in order to preseni the r;rc tion of thC (ontrol tod (ontarited with the (:RD in the esrrit of a pintulatrd f ailure of the t rat tot c oolant bouild.u) auotiated with the CRI) honnifig. A trstrailit is also prosidt d for the trat tot nitrinal puinp (RlP) in utder to present it lioin bet onung a iniude in c ase of a pouulated hulute of the tratlot t oolant boundar) auot tated with the reac tor intrenal puthp.
Heartor Preneurr Yrmmel, Appurtenatu re, Supportn and innulad n
'l he traitor ptruute sourt (RI'Y), .n shown u hrinatu alh I'ig u i r , l .1,
< onshis of a urtic al, < 3 hndric al preuurr gruel of welded c otnton tion, retnosable top head and head t knure bohing and irait the scurl unludes the ohndric al shell, llat gr, bottorn head, trac tor internal putnp (RIP) c.nings penritatiotn, bl.u kets, noriles sentuli shaped flow restric ton in the strain outlet en>/ilet and the shroud support whi< h nuludes the putnp dec L fortning the partition between the RIP un tion and diu harge. I hr sluond suppor t is an antinbly c onsisting of a set ti< al c in ular plate, a hoeirontal annular purnp dr( L plate atul serin al stilt !cgt "I hk support < ar ties the weight of penpheral f uel rlctnents neunon soun ts c ore plair, top gunir, shroud and shroud brad with uram arpatatof t it abo suppot's lairralh ihr f uel aortnblics aint the poinp ddhnert 'l he shroud support aho untains the differential peruuret
't he (ontrol tod driws are inounted into the (ontrol rod drivr housingt Nuhutn pentahorate solution trotn the uandbs hipud i ontrol wuren entrn the worl via one of the two high piruute t ote flooding hues and a dktrihrted through the sparger c onor< ted to the hnr.
1 he c ontrol rod dtko housings arr unerird ihtough and c onnc< ted to the o attol rod drhe penetiations blub tubro in the trat tio wart bottom head.
~ br in< oto neutron flux anonitor houungs are untited through and c onnr< ted to the bottutn head.
211 1 9/?W91
ABWR ossion occument a nangra in uir is prmiaca in oir ion heaa io, hoinng oi ihr nange auot uted with the inso urnenlanon los uhr ation n u ol inirinals.
'l he integral trai tor sruel sLui suppos a the trurl on the RPV pedestal.
Strel aan hon ludts estrial thstoigh the pr tesial aiul sn inre the flange e.f the sList to the prorstal. RPV stabilvers are prmuled in the unpre pothon of the RPV to erskt hornton 'loadt I. atrial supports aniong th CRll houungs arnt ins oir houungs arc lnosnird hv seuraints whn h at ene peripia ry air suppoited oil the CRI) housing ersti unt heanut A irstraint s onsisung of a pair e energs ahmthing rods k prosnied to prew ut the RIP horn bring a un.ule in c ase of a failuir in the c asing wehl with the bottorn head penetration. Ihr erstraint n c onnet ied to lugs on the RPV luittorts head atul the RIP tinotor (mor.
't he tr4< tor piruute vrocl insulation is supported horn the toologh al shirbl wall surrourninig the tourl. Insulation for ihr upper brad arnt flange n supported by a steel f raisir innlepriniriit ed the sesse i and i dpisig, leistilattoit ac t eu panel atul insulation around penetiaisoin are drugned for rase of lustallatiori arnt st unnal bit sessel inselTu r ins lPr< tildt Jnd sliailltellain e ope ationi.
'i hr enatoriah ined m the sca< tor < ootant pernure ,undary (RCPil) portion of ihr RPY and appurtenain es are as hsted he re an 'brse on their njuivalents will be med: ASA1E SA 533, 'Ispe it, Clau I (plate); SA 50H, Clau 3 (forging): SA 508, Class I (lovging); Sil 166, hpc 60 (forging); SA IN2, 13161. or 17316N (forging): and SA&lu, Gnule 1123 or 1521 (hohang).
A staintru storl wrhi merlas is apphrd to the intern.r of the < 3lindeh al shril and the sicatn outlet nonic. ()ther nonics and the RIP inoto c asing do not base < tadding. 'l he honom head is c lad with Ni Cr.I'r allo). the RIP penetratiom air < tad with Ni Cr lic alloy or uandru steel altrenatisch, l'hr seat tor c oolant peruurc boundary portion of the RPY and appurtenatu rs and the supporn (RPY dirt, uabiliter and CRi> housing /ioq ore housing reurainn and hearns) air c lauified as Quality (.roup A. Scistiiir (;ategor,' l.
'l he driign, anaterials, iriariuf ac turisig, f abric atiori, tr$titig, esattiisiaisissi, and innpn tion used in the c omts tu tion of thesc < oinponents niert acquirernents of ANME Code Clau i sturl and supports, respnliveh. Ihr shroud support is clauified as Quahty Group C. Sciunic Category 1, and designed arul labth ated to ASME C<nte Clau CS, c ore support struc turet lipirostatic tru of the RPV n performed in atiordane r with the requirernents for ASME Code Clau I scuch. Ihr drugn pernuar and temperature of the RPV are N7.9 kg/c m?g eini 302^C, irspt-(tarly. Ihe c oinponents arr c ode stamped an ording to their (ode (lau.
'lhe inatrriah of the low alloy plates and forging med in comtruoiou of the RPV air melted to fine grain prac th e ami air supplied in queru hrd and tempered c ondition. Va< uum deganing is pnlorrned to lowr the hultogen irsel and innprme the cleanlincu of the low alloy urch.
Submerged ar< and manual sth L rintrode weldmg proicurs are rinpimed.
Ein timlag wrkhng h not applied los strm tural weldt Ptrheat and interpan temperatures eniployni los welding of low allov urri mort or ext red the values given in ASMF Sn tion 111, Appendis 1). Pou wrld heat treatrnevit at 59Y(; nissiistiurn n apl died tre all low-alho sicrl wehh.
?11 2 W20/91
ABWR D: sign occument
~lhe ha< ture toughnew tnis of preuure boundan anaierials weld un tal and heat alin ted tone (llAZ) are periorined in a< <oidanc e with the acquircinents for ASMI: Code Clau I sencl. Iloth longitudinal and tratisserse spn itiiens are used to deternune the ininimutn upper shell energs inel < l the core belthne inaterialt heparate, unitradiated baseliar spniniens are used to deterrnine the transition teinperature t une of Ilic iore beltline base snaterials, weld nietal and ilAZ.
Prewuse boundary welds are gisen an ultrasonic esamination in eduition lo the radiographic exarnination performed during f abric ation. t he ultrasonic examination snethod, including calibration, mstrumentation *< atinisig sensitisits, and (userage, is based on the acquirements imposed bs A5Ml:.
Sc< tion XI. Appendix 1. Aucptan,c standasds are equivalent or more entric tive than required bs ASME, Sec tion XI.
For the venel material surveillante program, spetirnens are manufac tured from the material actually used in the reac tor belthne region and wehl typic al of those in the beltline region, thus representmg haic metal, weld material, arid tlic weld heat affec ted to ic anaterial. 'lin plate arid weld specimens are heat treated in a raiatiner whitti sirriutairs the at tual heat treatinent performed on the core region shcIl plates of the coinpleted sewel.
Eac h in reactor suncil!am e capsule c ontains Charpy V-notc h spnimem of base metal, weld metal, and heat affn led et oe inaterial, arid tetisite spnimens from base rnetal and weld metal. lit u tets are welded to the vessel (ladding in the < ore belt region for retention of the detat hable holders, rat h of which contains a nuinber of the spnimen capsuln. Neutron dosimeters and temperature inonitors are located within the (apsuln.
Au cu for examinations of the inntalled RPV is im orporated into the design of the vessel, biologic al shield wall and sessel insulation.
Reactor Preuure Venel System Internals The inajor reador internal c omponents that are ituluded iri tiie RPV5 are:
- a. Core Support Struc turn:
Shroud; shroud support (integral to the RPV arid including the internial pump deck);(ore plate (and core plate hardware); top guide; f url supports (orificed fuel supports and peripheral fuel support.s); control rod guide tubes; and
- b. Other Reac tor Internals:
Control rods; feedwater spargers; RilR/ECCS low preuure flooding spargers; ECCS high pressure core floodirig spargers and c oupinig; In-c ore guide tubes and stabiliiers; core plate dif ferential pecuure lines:
surveillame sample holders; shroud head and steam separators auembiv; and steam doer anembly.
A general auembly drawing of the important icac tor romponents is shown in f igure '.'.l.1. The c ore support structures locate and scpport tiic fuel assemblies, form partitions waln the reactor seuel to sustain preuure ddlerentials d(row the partitions, and diin t the flow of the usolant water.
, The shroud support, shroud, and top guide make up a stainlew steel l glindrical auembly that provides a partition to separate the upward flow of i
21 1 4 T20/91
1 ABWR ossign occum:nt l
l l
t otilarit thrasingh the t orr f roin ihr dowithald tra itt ulatnos flow. I htn I
, 9as tatsota wrpatalrh the t ore legion lioin the thruntiqnrt atinulus. )
l he a ole plate moiskts of a titt ular staintru i,terl plate with routul openings and 16 htilletW ..th a fitti and bratti sitin tute. l he t eere platr ptosnles ,
lateral sup[n, ated gundatu r for the (ontrol tent guide tubr% in< ole llus innuitor guide tuhrt prfiphetal lurl suppierit and startop orutton stulit es
'lhe last two iirins air aho supported sollitalls hs ilir 4 ore plate.
'l he top gunlc t eenshts of a sitt ular plate with s<guage espettitigt het luri widt a olindan al side for ening an uppri whf oud extrusion. Eac h opening prosides lateral suppoti and guidant e for four f uel awetnhlics or, a the < ase of
,>copheral fuel, len than four f uel aurtnbhrn. Ilotes are prosnied in the bottoin of the tuppoll infetter lions to atu hot the ins otr anstr utnentatuiti delet torn atul star tup neutrots souls es.
Ihe fuel supporth ate ol' two t) pet 'lhe prtiphetal lurt suppotts air lin ated at the outer edge of the at live ( ote arul are ttol .nljat ent to c ontfiel todt. lla< h peripheral fuel support supports our preiphrtal fuel auctubh atul a otilaisis an orilit.c to proside c oolatit flow to the lurl auctuhly. lw h orilis ed luri suppor t supports f our f uel durtilbhrt settit ally upward a:HI horiteitstalh atul c ontaitis four orilit es to proside (oolarit flow dai,tributiott to rat h fuel asseridd). 't he utilit ed f uel Support.s srst ott the top of ihr < otattol tod guide tubos whic h ate suppor trd latrially by the (ote piale. 'I he ( otitsol raHh pau theough t f ut iloten opettitigs ist the t ruler of the otilit rd f uel support.
'Ihr 4 otittol rod guide tuhrt hw ated itiside the sensel eslotul Iroin the top of the c ontrol rod drhr housings up thiough holes in the c ore plate. I at h guide tuhr is designed as the guide for the lower rfni of a c ontrol rod atul as the support for an orilh ed leri support. 't his haates the four f uel aurtnhhes surtoutading the t oliitol rod. The lower rint of the of the guide tuhr is supported by the c ontrol tod drive housstig, whi(h in l'.tret tratistluts the wright of the guide tuhr, f uel supporth, atul lurl awrinblics to the trat1 se vcwel bottoni bred. 't he (ontrol rod guide tubes ako e ontain holes, near the top of the (ofitrol tod guide Itahr atad holow the ( oro plair, for ( oolant llow to the utilit ed lurl supports Ihr (ontrol Ital guide tuhr base b prosided with a desit t los e oupittig conttol rod drive (Cit!)) with it. The Citi) in testrained trotn cir< tion, in the iair of a stub tuhr weld f ailure, by the c ouphng of the CRI) with the c ontf oi f od guide tube base; sti this ovesit, the flatige at the top sif the guide tuhr will c ontat t the t ore plate and restraitt the rjet tiott. 'l he t ouphug will also presrnt ejet tiori if the housing f ailt at the sf uh tuhr wohl; in this esent, the pu' W ac rosnaitis supportrd uti the intaa t upper housing.
li.c t otitrol rods air ( tuc iloitti shaped neutron absorbing starttibets thal c an he itiserted or withdrawin froin the t. ,o by the c ontrol tod drhet to (toitrol trat tmt) dnd trartor powel.
Eac h of the two fredwater hues n tonnet trd to thite spargets sia tiirre illT nonien. Ihr f redwater spaigris, whk h also f urn iion as 1-3 O high of low prruute flooding spargers deproding upon thrit t onnec tion to the line designated to f r< rise high pienure of low prenuro c oolant ihanhng supph, respra tisely, att staildeu storl hradern loc ated in lhe snixitig plenutu ahose the downminer annulut Lac h sparger in two halvr% with a ter c olitire led in the iniddle, h fitted to rat h fredwatre nouir with th< ter. 'Ihr spaiger 21.1 4 9/20/91
ABWR Dosion Document n e i,do s uinnn inne n,r RPv no,,a ,.ar , na b, a ao dde din manin.
ar r angemen t. l'rrdwatre flow onn e. the t enter of the spaigns aini n diu haeged radialb mward to umt the < <niles h edwatre with the dimin omes flow f oun ihr strain u paratoe s and strain dner bef oir it c ontai n the s curl wall 1 lhe design featuer on the two ersidual heat vemosal (l(llit) shutdown cin hng j mtrm spargen, whit h aho lum hon as 1 t 1 A low pn uure floodmg (1 Pill spargess, a sinnlai to that of tiu fredwain spaign t two lines of ItllR shutdown t oohng mirin enter the trae f or gruel through the two diagonalh esppositt no/iles and a onnn i to the spange is, l he sparger I r illiti is
< onnn trd to the itPV noisir safe end in a thersnal slene assangonent.
'lhe two 11LS lugh peruurr aore thunhng (llPCI) spargers and c ouphngs asc the means los dun tmg Ingh peruurr 11 CS flow to the upper end of the l' 4 ore. 1;st h of the two !!PCI lines entras ihr tras tot truel through a diagonath oppouir norile with a thermal stone , orange rnent. 'i hr t uned j sparger imludmg the c onnn ting tre a loc aird around the nnute ed and n supported in the e stuulru al [nahon of the top gmdr. I hr sparge tre n i connn led to the thesinal slene hv the llPL12 touphng l In4 ore guide tubes pinto i the ins ure ilus anonitormg natrumentation f rom flow of water in ihr bottom head plenutn, lhe inq one guide tubes entrnd f roin the top of ihr in4 ore housing lo the inp of the our plate. Ihr kn al power range monitoring (1.PI(N1) denn ton los the power range neutron numitoring (PitNN1) spiern and the dninion los ihr ihr stattup range l neutron rnoninermg ($RN%1) sotom are insrated through the guide tubes. 1
'lwo Irveh of stainicu steel stalnhirl latti< rwork of s tamps, lic hars, and spa < rrs gne lairaal support and rigidity to the guide tuhn 't he stahihirn arc iormn trd in the shroud and shroud support.
1 he < oir plate ddiverntial pronuar (1)P) hnen entes the tra< im truel through trai tm bottom hrad priirirationis. I~otar liairs of the c orr ;date 1)P Imrs enter the head in four quadrants through lous penetrahons and terminate immediately above and below the une plate to sense the perums in the region outside the bottorn ol the furt aurinhhn arnt below the c ore plate during norrnal opriatiori, hur' Cillatu r slin ifnen t a[Muln, W.In it att hrld ill cajalilr hiddres inentioned earlier, are lo(ated at thier afiriiuths at a (- ininori rinatiosi isi thr (orr lirlillnr f rgniel, lite (alnuir lloiders air nofl4atrh erlaird in ter na h. 'lhr capsuir hohlen are inn hanic alh retained by t apsule holder brac ken welded to the stuel clarlding in order to allow their remosal aiid r eattat hinreit.
Ihr shroud head and stram srparators aurinidy lorna the hip of the t of t diu harge nosture plenum together with the separaton and thnt
< ormn ting standpipn. 1he stearn dner aurmbly scimnes mohtuir hom the wet stram leaving the strain separaton. Ihr extrac ted monture flows down the dr)ct vanr4 to the t ollN ittig trougin, thrri llows through inhn mt<>
the dowm omer annulus. Ihr shroud brad ami steam separator aneinhh and the stram dner anembly are non4airly related intrenah.
~l he < ore support strm turn air c lawilied ,n Quahty Group C, Sciunk Categon
- 1. Ihr dnign, snatreiah, snanuf a< turmg, labr u abon. exaininahon and irnpra tion used in the < ointrm tion of the ore support sti a< turn meet 211 6 900/91
ABWR 0: sign Document is t{uit t ettents of A%hll' (',t nir (;lau (.h str us lures. I hrse sit tia tut e a der t ante-staitipnl at s otalitigh , ()thri trac for liitet tials air tirsagaint pot the guiilrknes of AN\ll'. Co<le N(adOno anil air i onstsuc Iril so as not to aihi rsch atlet t tric intogtits ol' the (otr sup}neti sit tu tuits as trajustnl hs N(;-11V2 Inspection, l able 2.1.1 prosuln a airlinition ol ihr insit tu tiom, inis, anal, or analoes Test, Analyses togethes with anos iatnl au e piant o i ntena whn h will be uvulretalen for the and Acceptance ira < tos preuurr scurl wstein.
Criteria
) 6 4.
Table 2.1.1 REACTOR PRESSURE VESSEL SYSTEM inspections, Tests, Analyses and Acceptance Criteria
(
inspections, Tests. Analyses Acceptance Criteria {
Certified Design Commitment Visual field inspntions will be conducted I. The installed omliguration of the RITS
- l. System configuration of the reactor I.
wiri be o,nsidered aueptabic if it pressure vessel system (RPVS) as described of the installed RPVS key compments in Section 2.1.1.1 is shown on Figure identified in Sntion 2.1.1 and Frgure comphes with Figure 2.1.1 and Susion 2.1.1. 2.1.1.
2.1.1.
- 2. The reactor coolant pressure boundary l Oxle required documents and the Oxic required dmuments and the owie stamps (RCPB) portion of the RPV and stamp on the components. on the components confirm that the appurtenances and their supports are o>mpments in the RCPR of the RPV, its l l classified as Quality Group A, Seismic suppwts and the aire supp>rt struoures are Category I. 'Ihese components are designed, fabricated and examined as designed, fabricated and examined in ASME Code Clau I and CS resputively.
accordance with the rules of ASME Code Class I vessel or component support, and are osde stamped accordingly. The core support strue tures .re Quality Group C.
Seismic Category a, and are designed, fabricated and examined in accordance with the rules of ASME Code Class CS structures, and are code-stamped according!y.
- 3. A hydrostatic test of the RCPB will be 3. The results of the hydrostatic test must
- 3. The RCPB of the RPVS retains its confi>rm with the requirements in the integrity under internal pressure that conducted in accordance with the ASME Code requirements. ASME Oxle.
will be experienced during the service.
- 4. Inspection will be conducted of the records 4. Rnords c' the materials and prmesses
- 4. The materials used for RCPB p>rtion of of materials, fabrication, and must contirm that the requirements the RPV and appurtenances are certain specified for the RCPB in Section 2.1.1 are proven low and high alh>y stects with examination um! in umstruction of the RCPB and austcnitic stainiew steel reactor satisfied and that the manufacture and certain additional requirements for construction, as identified in Sution 2.1.1. internals. fabrication of the RPVS internals made of austenitic stainless secci avmd p>tential for Special controls are exercised when cracking in serske.
austenitic stainless steel is used for construction of RPVS internals in order to avoid cracking during scrwice.
- Table 2.1.1
- REACTOR PRESSURL VESSEL SYSTEM (Continued)
Inspections, Tests, Analyses and Acceptance Criteria ,
Certified Design Commitment inspections, Tests. Analyses Acceptance Criteria +
- 4. (Continued) r Validation Attributes: i j
The following special controls are exercised .
- when austenitic stainless steel is used in '
manufacture and fabrication of RPVS ..
internals. ' Where stainless steel surfaces are exposed to water at temperatures above 93*C, .!
- low carbon (C(0.03%) grade materials are 1
used; for high strength applications, nuclear r grade materials (C<0.02% with nitrogen added) are used. All materials are supplied '
4 in the :olution heat treated condition. i Sensitization tests are applied to assure that
- the material is in the annealed condition.
During fabs~ cation, any headng operation (except welding) lxtween 427'-932*C is 3 avoided, unless followed by solution heat ,
treatmen t. During welding, heat input is - 'I controlled. Weld filler material is Type t
- 308L/316L/309L or equivalent. All weld .
4 filler materials have a minimum of 8 FM l
(ferrite number) deternuned on undiluted weld pads by magnetic measuring ;
- instruments. During fabrication, cold work '
is controlled by applying lim ts in hardness, i bend radii and surface finish on ground .
j surfaces. Pro (ess controls are exercised
, during all stages of component manufacturing, fabrication and installation to minimize contaminants. Surface contaminants are removed prior 13 any j heating operations. ;
Y F
L
Ygbla 21.1 REACTOR PRESSURE VESSEL SYSTEM (Continued) 1 i
inspections, Tests, Analyses and Acceptance Criteria -
Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 5. The ferritic materials used in RCPil 5. Fracture toughness tests will be condut ted 5. Reo>rds of the fra< ture toughness data portion of the RPV and appur;enantes are in aco>rdarice widi th( sequire-inents for snust coriferr.i that the RCPit will riot he not susceptable to brittle fracture under ASMI: Class I (omimnents of the ferritic susc eptible to oristle fratture during the pressure during the senite. rnaterials, weld and heat affected inne service.
(IIA 7.).
Validation Attributes:
- a. The minimum upper-shelf energy '
lesel for base and weld metal inust be IDA kg-m.
- b. The predi< ted minimum upper-shelf . !
ener;v 'evel for base and weld metal at l ern oflife must be 6.9 kg-m. !
- c. Th predicted value of adjusted RTNDT of rase and weld metal at end of life I must be 95#G or less. I 5
I e
C
- 6. Specimens for the surveillance program 6. Inspecti >n will be conducted of the rea>rds 6. Records of the specimens with respect to
- are selected from the v-ssel base metal of the sperirr eus sele (ted from the reactor location and orientation, types (tensile or i
! and weld metal. beltline rer, ion. Chaspy V-notch), and quantities mus meet the requirements of ASTM E-185. '
- 7. Design and construction of the RPV 7. A vibration ass ssment will be conducted 7. Reactor vessel internals vibration is !
internals assures that they can, withstand of the reactor internals to verify the considered acceptable when results of the
- the effc(ts of flow induced vibration (FIV). adequaq of the int -nals design, vibration measurement testing and ;
manufacture, and assembly with respect to inspection of the internals irdicate no !
the potential effects of FIV. The first+f-a- sign of damage, loose parts, or excessive t k'nd prototype internals will be flow - wear in. the prototype test. The vibration tested by vibration instrumentation of reactor internals in subsequent plants is i followed by inspettion f ,r damage. The omsidered acceptable when inspection of i infernals in subsequent plants will bc the internals indicate no sign of damage. !
llow tested, but without vibration loose parts, or excessive wear.
-instrumentation, follwed by inspection j
for damage. '
D i
! Table ' 2.1.1 REACTOR PRESSURE VESSEL SYSTEM (Continued) :
1 A >
l . inspections, Tests, Analyses and Acceptance Criteria 1
- ' Certified Design Commitment' Inspections. . Test, Analyses ' Acceptance Criteria '
t.
- 8. Auess- for examinations of the RPV 'is .
- 8. - Visual inspet tion will be conducted of' 8. Providons for au ess in the <lesign of the [
incorporated into the design of the vessel, auessibility for examinations of the vessel sessel, biological shirl l wall, and rewl i biological shield wall and vessel and welds. insulation shalf be, in the minimum, as t l insulation, follows: l j The shield wall and seswl insulation i j behind the shield wall must be spaced I
- away from the RPV outside surface. Aucss ' ;
for the inwrtion of. automated dernes must be prusided through removable ,
insulation pancis at the top of the shield '
wall and at auess ports at reactor vessel nonles. Auess to the reattor pressure v-ssel welds alweve the top of the biological shield wall must be pronded by
! remosable insulation panels. The chisure :
head must have removable insulation to !
provide access for manual ultrasonic examinations of its welds. Aucss to the .
bottom head to shcIl weld must be provided through openings in the RPV ;
support pedestal and removable insulation !
panels around the glindrkal lower !
}mrtion of the sessel. Acc ess must be prmided to partial penetration nonle welds, i.e... CRD penetrations,
- ' instrumentation noules and retirculation internal pump penetration' welds, for ,
performance of the visual examinations.
- Access must be provided for examination I of the ' attachment weld b<-tween the !
support skirt knu(Lle (forged integrally [
on the shell ring) and the RPV support :
skirt. Aucss must be prmided to the balance of the support skirt for perforrnance of visual examination. ;
4 i
VIBRATION _~ ~7,s=nji o'h INSTRUMENTATION F (
I CLOSURE HEAD * '/ I STEAM FLOW RESTRICTOR i
STEAM DRYERN -
t_-- l, -El N "
/
STEAM OUTL2T s
p 4 -
1 -
n I RPV STABILIZER a ._ _ _ ___
l-d SHROUD HEAD LOW PRESSURE FLOODING SPARGER k < '[ FEEDWATER SPAR 000 y ) _ y 4i .-
INL T SHUTDOWN COOLING INLcT Ta w [ -H HPCF COUPLING HIGH PRESSURE CORE [_ ITP"-~ ~~
!! I m CONTROL ROD
/ i 11 i
FLOODING / CORE PLATE TOP GUIDE SPARGER[
f
[ ll /
SURVEILLANCE SPECIMEN HOLDER f} '
l ii , 'y SHROUD i li .- PERIPHERAL FUEL ORlFICED FUEL SUPPORT i i g SUP.'..T
.a <
]n -
,l 3
' F '
/ s i _ _. __ _
RPV SUPPORT SKIRT c=:g4 .,
-. ! 'N IN-CORE GUIDE TUBE AND STABILEERS CONTROL ROD GUIDE TUBE I I m' ! .c. J. "P t
\ N N SHROUD SUPPORT INSULATION %;. jg .
l l l REACTOR INTERNAL i i , i : PubP CASING CRD HOUSING 2- - ,- - .
.a IN. CORE HOUSING CRD RESTRAINT BEAM #
Figure 2.1.1 REACTOR PRESSURE VESSEL SYSTEM KEY FEATURES
ABWR ocsign Document 2.2.4 Standby Liquid Control System Design The standby liquid control system (SLCS) is designed to inject neutron absorbing poimn using a boron solution into the reactor and thus provide back-up reactor shutdown capability independent of the normal reactivity control system based on insertion or control blades into the core. The system is capable of operation over a wide range of reactor pressure conditions up to and including the elevated pressures associated with an anticipated plant transient coupled witn a failure to scram (ATWS).
The standby liquid control system (SLCS) is designed to provide e capability of bringing th. .cactor, at any time in a cycle, from fun power and minimum control rod inventory (which is defined to be at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive xenon free state without control rod movernent.
The SLCS consists of S boron wiution storage tank, two positive displacement pumps, two motor operated injection valves which are provided in parallel for redundsnq and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The horated solution is discharged through the 'IP high pressure core flooder (!!PCF) subsystem sparger. Figure 2.2.4 shows major system comjonents. Key equipment performance requirements are:
- a. Pump flow 50 gpm per pump
- b. Maximum reactor pressure 1250 psig
((or injection)
- r. Pumpable volume in 6100 U.S. gal storage tank (minimum)
The required volume of solution contained in the storage tank is dependent upon the solution concentration and this concentration can vary during reactor operations. A required boron solution volume / concentration relationship is used to define acceptable SLCS storage tank condirons during plant operation.
The SLCS is manually initiated from the main contro. room. The SLCS provides borated water to the rearsor core to compensate for the various reactivity elTects during the required conditions. These effects inclune xenon decay, elimination of steam voids, changing water density due to the reduction in water temperature, Doppler effect in uranium, changes in neutron leakage and changes in control rod worth as baron affects neutran migration length. To meet this objective, it is necessary to inject a quantity of boron which producer a minimum concentration of 850 ppm of natural boron in the reactor cora at 70*F. To allow for notential leakage and imperfect mixing in the teactor systern, an r.ddiuc 4 25% (220 ppm, is added to the above requirement. The required conceneation is ru hieved accounting for dilution in the RPV with normal water level and including the volume in the residual heat remov;.i shutdown cooling piping. This quantity of boron solution is the arnount which is above the pump suction shutoff level in the tank thus allowing for the portion of the tank volume which can:mt be injected.
2.2.4 9f20/9I
1 1
ABWR Design Document l
l The pumps are capable of producing discharge pressure to inject the soluuan into the reactor when the reactor is at high pressure conditions corresponding to the system relief valve actuation. Signals indicating storage tank liquid level, tank outlet valve position, pump discharge pressure and injection valve position are available in the control room.
I The SLCS uses a diuolved solution of sodium pentaborate as the neutron- ;
absorbing poison. This solution is held in a storage tank which has a heater to maintain solution temperature above the saturation temperature.
The SI.CS solution tank, a test water tank, the two positive displacement pumps, and associated vahinFi s located in the secondan containment on the floor ele y;<.n below the operating floor. This is a Seismic Categon I structure, and the SLCS equipment is protected from phenomena such as earthquakes, tornados, hurricanes and floods as well as rom internal postulated accident phenomena. In this area, the SLCS is not subject to conditions such as missiles, pipe whip, and discharging fluids.
The pumps, heater, valves and controls are powered from the standby power supply or normal ofTsite power. The pumps and valves are powered and controlled from separate buses and circuits so that a single active failure will not prevent system operation. The power supplied to one motor operated injection valve, storage tank discharge valve, and injection pump is powered from Division I, 480 VAC. The power supply to the other motor-operated injection valve, storage tank outlet valve, and injection pump is powered from Division 11,480 VAC. The power supply to the tank heaters and heater controls is connectable to a standby power source. The standby power source is Class IE from an on-site source and is independent of the ofTsite power.
All components of the system which are required for injection of the neutron absorber into the reactor are classified Seismic Category 1. All major mechanical components are designed to meet ASME Code requirements as shown below.
ASME Design Conditiona Comnonent Code Cla.w Piewure Temperature Storage Tank 2 Static IIead 150*F Pump / Motor 2 1560 psig 150 F Injection Valves 1 1560 psig 150 F Piping Inboard of 1 1250 psig 575 F Injection Valves Design provisions to permit system testing include a test tank and associated piping and valves. The tank can be supplied with demineralized water which can be pumped in a closed loop through either pump or injected into the reactor.
Inspections, Table 2.2.4 provides a definition of the inspections, tests, and/or analyses Tests, Analyses together with associated acceptance criteria which will be undertaken for the and Acceptance SLCS.
Criteria 2.2 4 9/20/91
i Table 2.2.4 STANDBY LIQUID CONTROL SYSTEM i
i ,
- inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses - Acceptance Criteria l 1. The minimum ave
- 3ge poison 1. Construction records reviews and plant 1. It must be shown the SLCS can achieve a 1
concentration in the reactor after visual examinations will be underta en *
- poison concentration os850 ppm or ,
operati< n of the SLCS shall be equal to or to assess the a4uilt parameters listed greater assuming a 25% dilution due to i
. greate' than 850 ppm. below for compatibility with SLCS design non-uniform mixing in the reactor and calculations. If necessary, an aduilt 4 '
accounting for dilution in the RIIR -
SLCS analysis will be conducted to shutdown cooling systems. This
- demonstrate de scceptance iteria is concentration must be achieved under me t. system design basis ronditions.
{ Critical Parameters: VaMstion Attributes:
i a. Storage tank pumpable volume Storage tant pumpabic volume range
.I 6100 - 6800 gal.
, b. RPV water inventory at 70*F RPV water inventory s 1.00 x 101b 6 4
- c. RIIR shutdown cooling system water inventory at 70 F RIIR shutdown cooling system inventory 1.287 x 106lb
- 2. The system configuration is in
- 2. A simplified system configuration is 2. Inspections of .mstallation records shown in Figure 2.2.4. accordance with Figure 2.2.4 .
together w,th plant walkdowns will be i
conducted to confirm tha. the installed equipment is in comphance with the
, design configuration defined in Figure 2.2.4.
s t
b
_i _ _ _ __ 6__ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ ______
k Table 2.2.4 '
STANDBY LIQUID CONTROL (Continued)
Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses . Acceptance Criteria
- 3. Each SLCS pump shall be capable of 3. System preoperation. tests will be 3. It must be shown that the SLCS an delivering 50 gpm of solution against conducted to demonstrate acceptable inject 50 gpm (each pump) against a the elevated pressure conditions which pump and system performance. These reactor pressure of 1250 psig..
- can exist in the reactor during events . tests will involve establishing test invohing SLCS initiation. conditions that simulate conditions which will exist during an SLCS design basis event.
l 4. The system is designed to permit in- 4. Field tests will be conducted after system 4. Using normally installed controls, service functional testing of SLCS. installation to confirm, in-service power supplies and other auxiliaries, system testing can be performed. the system has the capability to perform:
- a. Pump tests in a dosed loop on the test tank and
- b. Reactor pressure wssel injection tests usmg demineralized water from the test tank.
- 5. The pumps, heater, valves and controls 5. System tests will be conducted after 5. The installed e guipment can be powered can be powered from the standby AC installation to confirm that the from the standby AC power supply.
power supply as described in Section electrical power supply configurations 2.2.4.* are in compliance with design commitments.
- f
_ i -
VENT PRIMARY CONTAINMENT STORAGE TANK HEATER I I
,-- -g i 1
--A-- - --- -- H PCF 'B' l '
SUCTION VALVES (WITH POSITION i
, , INDICATION) i_ _ . M. _ __ ____M7 ASME F---------------~~ _______
CODE CLASS r- % [- 1 1
--- I n ! : !
j . . I I PUMPS i INJECTION I I I VALVES I I P ' '
(WITH POSITION
__: - -lNDICATION) ,
l---
L______ ____________I
--l l I I 8 s t V i l L_____p_____'I g iL_____i :
l CODE CLASS 2 l l
- I I
- - - 1 TEST '
TANK l
_a Figure 2.2.4 STANDBY LIQUID CONTROL SYSTEM (STANDBY MODE)
ABWR D: sign Document 2.2.7 Reactor Protection System Design 'l he reactor protec tion system (RPS) for the Advam ed Boiling Water Reac tor Description ( ABWR) is a warning and trip sutem where initial warning and trip decisions are implemented with software logic installed in mic ropro(essors. l The primati functions of this system are to: (1) make the logic dec isions !
related to warning and trip conditions of the indhidual instrument t hannels, and (2) make the decision for system trip (emergemy reactor shutdown) based on coincidence of instrument t hannel trip conditions. l I
T he RPS is classified as a saf ety protection sptem (i e., as dif fering from a j reactor control ssstem or a power generation system). Fun (tions of the RPS l and of the components of the system are safety-related. Ihe RPS and the l clectrical equipment of the system are also clawified as Safety Class 3, Qualits Group C, Seismic Category I and as IEEE electrical category Class IE.
Basic system parameters are: I
- a. Number of independent divisions of equipment 4 i
- h. Niinimum number of sensors per trip sariable 1 (at least one per division) 4 1
- c. Nurnber of automatic trip sutems (one per division) 4 l
- d. Automatic trip logic used for plant sensor inputs (per division.) 2-o u t -o'-l e Separate aun,matic trip logic used for division trip outputs 2-out-of-l
- f. Number of separate manual trip sv. stems 2
- g. N!anual trip logic 2-out-of 2 The RPS consists of instrument channels, trip log;rs, trip actuators, manual controls and scram logic circuitry that initiates rapid insertion of control rods (scram) to shut down the reactor for situations that (ouhl result m unsafe reactor operating conditions. The RPS also establishes the required trip conditions that are appropriate for the different reactor operating modes and provides status and (ontrol signals to other sutems and annunciators. The RPS related equipment includes detectors, switches, microprocessors, solid-state logic circuits, relay type contactors, relays, solid stat- load drhers, lamps, displays, signal transmission routes, circuits and other equipment which are required to execute the functions of the system. To ars ompikh its overall function, the RPS utiliics the functions of the essential multiplexing
<vstem (EA!S) and of portions of the safety system h>gic and control (SSLC) system.
l As shown in Figure 2.2.7a, the RPS interfaces with the neutron monitoring system (NNIS), the process radiation monitoring (PRRN1) ssstem, the nuclear boiler system (NBS), the control rod drhe (CRD) system, the rod control and information system (RC&lS), the reactor recirculation flow control system (RFC), the process computer sptem and with other plant systems and equipment. RPS components and equipment are separated or segregated from process control system sensors, circuits and functions to minimize control and protection system interactions.
The RPS is a fom division system which is redundantly designed tc proside rehable singleai:m e-proof capability to automatic ally or inanually initiate a reactor scram whde maintaining protection agairnt unnecessary scrams 227 9/20/91
ABWR Design Document
.nunn,g i.e,n ungic bido,m in ih, ReS. n,c RPS n.,nains angicaadu,e proof nrn when one dividon of < hann< l $rnuin k bspaned aanl/or when
) one of the four autoinatic RPS tiip logn mirins h out oI4rnu e. 'the mirm ahe imiudn nip bsp_m and houted outpun le, thsia.<v. nnum iatnin e, performam e monitoring. 'lhr RPS trlated equipment n duided into lour ordundant dahions of sensor (itntrument) t hanuch, trip logi< s and trip
.u tuaton, and two dividons 01 inanual staan. < outroh and u ram loga rin uitn. 't he autornaii< and manual u sam initiation h,gii mienn are independent of cat h othr and use dnene methods and r<pnpment to initiate a rea< tor u ram.
Figme 2.2.7b shows the RPS dhidonal scpatation aspet h and the dgnal flow paths from senson 10 u ram pilot sabe solenoids. lapiipment within a ItPS related scenor (hannel c ondsts of semors (tranulm a n or swm bn),
multiplexen and digital trip moduln (1)l his), 't he setnon within ca( h (hannel monitor hir abnoimal operating c onditiom and send rither diu rete bi, table (trip /- trip) or analog Ugnah dien tly to the RPS telated 111 N1 or che send anah,y, output ugnah to the RPS related I)'I M by incans of the irmote multiplexer unit (lulu) within the asso<iated dnidon of curntial multiplexing mtcm (DIS). The RPS relaird bktable switc h upr senuits or, in the case of analog c hannth, the RPS unique vdtware logic ,
will initiate reartor trip - .ials within the individual sensor < hannett when any one or more of the c onditioin liurd below exist within the plant during dillerrut c onditiorn of trai tor operation, and will initiate t rat tor u sam if coituidrin e lugn h satklied,
- a. Tushine stop vahn (losure (at high power irsch) lRPS)
- b. 'l urbine c ontrol vahn last < lmuir (at high power inch) { RPSI
- c. NNIS monitored SRNM and APRM ionditiom ru ced au rpiable linuis (NMS)
- d. Iligh main sh am line radia"on IPRRM l
- c. liigh reactor preuure (Nils)
- 1. l.ow reactor wat:r level (1.nci 3) INBS)
- g. liigh drywell preuure (Nilsl
- h. Main sicam hors isolation (MSt.1) (run inode ontv) (NBSI
- i. l.ow control rod drive an umulator c hanging header prenure ICRI)l
- j. Operator-initiated manual u ram (RPSI -
The system inonitoring the proccu condition is iinntated in brac kets in the list above. The RPS outputs, the NMS outputs, the PRRM Spirm outputs and the MSI.! aiul inanual u ram outputs are provuled dirnily to the RPS by hard- ,
wired or liber-optic signah. The NHS and the CRl) Sutem provide other semor outputs through the DIS. Analog to digital < omersion of these latter semor output salues h donc by DlS equipment. The 1)TM in eat h daision mes either the diu rete Instable input dgnah, or comparn the current values of the indhidual monitored analog variables with their trip setpoint values, and for cach variable sends a separate, diu rcic bhtable (trip /no top) output signal to the trip logic units (T1.Us) in all four divkiom of trip logics. 't he 1)l Ms and T1.Us unlued by the RPS are mic ropioicuor < omponents within the SSI.C spicm.
RPS related equipinent within a RPS dhkion of trip logit ronysts of inanual routrol swin hn, bypan units (ILPUS), trip logi< units (TI,lk) and output logit unin (OI.lN 't he n,anual i ontrol swin hn and the llPlN, TI.lN arul Ol.Us 227 7 9/20/91
ABWR Design Document arc c oinponents of the RPS portions of the SSif S3 stein. The sarious inanual j switc hes prmide the operator incans to inoddy the RPS trip logn for special l operation, inaintenanc e, testing and splein reset. The b>pau units perforrn bypaw and interiork logic for the single division of t hannel sensors bypaw function and for the single dhision T1.C bypass function. The Tl.Us perforin the automatic s< rain initiation logic, nor mally c hec king foi two-out of lour coincidente of trip conditions in any set of instrument c hannel signals coming f roin the four division DTW or from holated bntable liiputs froin l all [our divisiolis ol bMS equipinent, and outputtkng a trip skgnal kl any olle of the two-out of-four (oin< idente < he< b s is sathfied. TI.U trip decision logir in all four RPS Tlfs becomes a < he(L for two out-ol.thice (oincidence of trip conditions if any one livkion of c hannel sensors has been hypawed. 'l he 01.0s peiform the divnion trip, scal-in, teset and trip test lurn tions. Trip signah f rom the OI.Us within a single division are used in trip the trip actuators, which are fast respoine, bhtabic, solid 4 tate load drisers for automatic suam initiation, and are trip relays for air header dump (ba L up stram) initiation. l.oad diner outputs toggled by a divkion 01.C interconnect with load driver outputs toggled by other division Olfs into two separate arrangements whi< h results in two-out-of four scram logic, i.e.,
reactor uram will orrur il load drivers anoriated with any two or more divhions re(cive trip signah.
The isolated ar load drivers are fast response time, histalle, solid state, high current interrupting devices. 'I he operation of the load drivers is sm h that a trip signal (logir "0" voltage level) on the input side will create a high impedance, current interrupting condition on the output side. The outpm side of each load driver is ciertrically i olated from its input signal. 't he load driver outpms are arranged in the w rain logir circuitry, betweeti the scram pilot vahes' solenoids and the solenoids ar power sour < c, sm h that when in a t ipped state the load drivers will cause deenergitation of the scram pilot valve sol noids (stram initiation). Normally closed relay contac ts are arranged in the two back "p strain logic circuits, between the ali header dump vahe solenoid and air header dump valve solenoid d< power sourc e, such that when in a tripped state (toit drenergired) the relavs will cause energi/ation of the air header dump valve solenoids (air header dump initiation). Associated dc voltage relay logic is aho utilized to ellett scram reset permiuhes and scram follow (control tod run-in) initiation.
The RPS design for the AllWR is testable for correct response and performam t, in over-lapping stages, either on line or off-line (to nunimite potential of unwanted trips). Arren to bypass rapabilities of trip funruons, instrument channeh or a trip sptem and arten ts. setpoints, calTration controh and test points are designed to be under adminntrative control.
Tests, Table 2.2.7 provides a definition of the vnual isnpertioin, tests and/or inspections, analyses, togsthen with ano<iated arreptan(c criteria, which will be used for Analyses and the RPS.
Acceptance Criterin 227 9/20/91
l Table 2.2.7 1
. REACTOR PROTECTION SYSTEM l-
! Inspections, Tests, Analyses and Acceptance Criteria ,
j_ l t
Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria *
- 1. The four reduncimt divisions of RPS 1. Inspections of fabri(ation and I. Imtalled RPS equipment will be equipment and the four automatic installation records or visual field determined to conform to the j trip systems are independent from inspections of the installed RPS documented devription of the each other except in the area of the equipment will be used to confirm the design as depicted in Figure 2.2.7b.
l requiwd coincidence of trip hygic quadruple redundancy of the RPS and _;
decisions and are both electrically the electrical and physical separation and physically segurated from each aspects of the RPS instrur nt other. Similarly, the two manual channels and the four automatic trip trip systems are separate and systems as well as their diversity and
- independent of each other and of independesue from the two manual the four automa ;c trip systems. trip systems.
- RPS software logic for making the
- 2. 2. The Verification and Validation (V 2. The V&V program has Imen .
individual sensor and coimidence
& V) program for the RIN softwar e completed and do(umented..
trip decisions (as implemented by that ic utiliicd in the DTM and TLU firmware) is verified and validated. micropro(essors will be performed .
per V & V procedures and will be !
i documented.
- 3. Provisions exist to limit access to trip 3. Visual field inspections of the installed 3. The RPS hardware /tirmware will i serpoints, calibration controls and RPS equipment will be uwd to confirm be considered acceptable if
, test points. the existence of appropriate appropriate methods exist to enforce i administrative controls. administrative control for auess to ;
sensitisc areas.
i f
i r
i k
- _ m .__ _ _
Table 2.2.7 REACTOR PROTECTION SYSTEM (Continued).
Inspections, Tests, Analyses and Acceptance Criteria r
Certified Design Commitments inspections, Tests, Analyses - Acceptance Criteria
- 4. It is gunsible to conduct verification of 4. Field tests will be conducted to cor,'irm 4. The installed reac tor proscetion RPS operations, both on-line and off- that system testing suc h as channel system c onfiguration, <ontrols, line by means of individual (b-< ks, channel functional tests, guswer sources and installations of
- instrument channel function tests, channel calibrations, coincident logic interfacing systems supports the RPS trip sytems function tests and total tests and paired wntrol mds scram higic system functional testing and system functi mal tests. tests can be performed. 'l hese tests will the operability verification of design
- involve simulation of RPS testing as follows
- !
modes of operation. Interloc ks associated with the reactor mode a. Installed RPS initieres trip . i switch positions, and with other mnditions in all tour RPS operational and maintenance bypasses automatic trip systems upon ,
or test switches will be tested and mincidence of trip conditions in annun- ciation, display and logging two or more instrument functions will be confirmed. t hannels associated with the same trip variable (s). 'l l
- b. Installed system initiates full ;
reactor trip and emergency I shutdown (i.e., deenergiration l of both solenoids associated with l all sc ram pilot salves) ugam l coincidence of trip conditions in >
l two or rnore RPS automatic trip systems.
- c. Installed system initiates trip ,
l condition in both RPS manual l trip systems if both manual trip swiuhes are operated or if the -[
l reactor mode swit<.h is placed in [
l the shutdown" position. ;
1
Table 2.2.7 REACTOR PROTECTION SYSTEM (Continued)
Inspections, Tests, Analyses and Acceptance Criteria inspections. Tests. Analyses Acceptance Criteria Certified Design Commitments
- 4. (< ontinued) ,
- d. Trip sprem (automatic and manual) trip c onditions scal-in and prote< the u tions go to c ompletion. Trip rewt (atter delav for trip < ompletion) requires deliberate operatur action.
e alled wstem energires both neader dump (ba:Lup i
x ram) v.*hes of the CRD j h3drauhc system and initiates CRD motor run-in < onc urrent only with a full sc ram condition.
- f. When not bypassed, trips sesult upe.n low or dixonnecion of portions of the system. When inpaswd inappropriate trips do not result.
- g. in .alled sptem provides status and control 4gnals to data logging. displav and annunuator sutems.
i
, Table 2.2,7 -
e
( REACTOR PROTECTION SYSTEM (Continued) j Inspections, Tests, Analyses and Acceptance Criteria
{
' Certified Design Commitments inspections. Tests, Analyses Acceptance Criteria ;
- 4. (<ontinued) t
- h. The installed system f demonstrates operational interhxis (i.e trip inhibits or ,
permissive) required for '
i different (onditions of reactor
{
j operation.
- 5. The RPS design provides prompt 5. Field tests will be (onducted to measure 5. The RPS hardware /linnware protection against the on et and the' RPS and supporting systems response to initiale scattor scram consequences of events or c onditions response times to: (1) monitor the will be c onsidered aueptable if such that threaten the integrity of the fuel variation of the selected processes; (2) response is demonstrated to be
- ba r rie r. detect when trip setgx>ints have been sufficient to assure that the specified exceeded; and (3) execute the acceptable fuel design limits are not .
subsequent protection actions when exceeded.
coinciden(e of trip conditions exist. '[
Validation Attributes: >
Total trip system response, frori I time when sensor input is beyond -
, . serpoint to time of scram pilot valve- i solenoids deenergiication: ?
- NMS APRM s 0.090 sec. ,
- Rearsur pressure . s 055 sec..
- Rear tor water level s 1.05 sec.
6
Table 2.2.7 REACTOR PROTECTION SYSTEM (Continued) inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitments - Inspections, Tests, Analyses Acceptance Criteria Validati<ni Attril>uirs: (t ou tin ued)
- Turisitic stop-valve
. closure s 0.060 we.
- Turt>inc < enitrol vahe :
fast (losure s 0.080 ser.
.\ lain stearn line isolation s Oh60 w(.
I r
i
APau namn
- u ts.c== MAIN CONTROL ROOM LOCAL AREA '"""*'"***'"'** RPS LOGIC & CONTROL Cs1 PLANT SENSORS NMS Core sw Rape cocn.ase y CT1 C71 C71 C71 Cri Cn Bypassed (not F gsmsc: RP*. RPS RPS SRNM PW Fbs Upsca;e RPS RPS RPS Inp Reac5cc S*ior1 Pemd Manuel Dev Tre MAtJUAL MANUAL Parec Horh Heac1tF SCRAM Scam Tes-Mace BMS oulo:ts INOP erm SCRAM B Saric es s, ten et Hit Bypassec (nor egynx4 , ses e one rw A Jk da MPF '- T
$d ] 99 PRRM MSL Raomon -
to N1 puts c71 m Hti Rps qp q, MaMM4'
- g" r
S$tC RPS LOGIC PERFORMS: , r
-> AS S - C - Tm o.osen
- -System Concoena Teo Tbcrsea e'7 c2 3S u am.,en-.tw
- Manua; Desen Ynp -g CHO rC12 CR0 HCU Accumew
- E 2L S- e-s
. h 3 Maimenwa Pypass
.b d CRD Cha png tiender Pressora " " nh H23 M 7 C .Ca m .ss oisa w p s,am m ,i vane i ory.eii Pressure 3 - + Sderwat I ow thwm **
B21 Raaco Ww led C71 RPS LOGIC <ix, n. m Ac r.,*ny NBS N"*'*_*_"'#_P essee_____; _
DTM MSIV Posnm Seicnes ;
REACTOR PROTECTON TLU Actuaws he stam As
- + iteak, thenp vanes - -
Tm Stop i'w.e Status SYSTEM OLU
- (IA+. II. In tr hmn)
C71 Turtme HTS Od Pressure Nrwetow : 4k d64i RPS Twtzna tst Stage Press <a koc SSLC Log.c S85N h4 orocesseg F-ocewng q, y -m elupa4'r't go, ot ,
C11 Wtee Sea-' i om=q loca1ed Satay Systwns m Hti RC&lS (Cont w Rod Ran 4-)
! GLOSSAM WM - Amgo Power Range hw .
INTERDistS80NAL SIGNAL TRANSTER N -Cone % % !
F OR 2 outc'd COfNCIDENCE LOGIC '
EOC - Endo Cy1cie 1 9
EMS - Essental MJ1pleseg Sys5em i f vnac aC Pourer SUPPiv HCU - Hyormule Control tht C11 RC2S Rod Weideawal B!och Viheft Bypsuss S*4ch For C81D E=w E SM vac PS)
HTS - Hydrade Trp System p w g p,ess,se im m e " Bypass" Pm R46 m, MCRP - Mam Control Room Peres Recret;.amt Fee Cor vol System - EOC RPT Command ses toec === wm CS1 RFC MSfV - Man Steam isonanort Vahe , ,,
MSL - Man Swam tra C91 Prmass Corapeter - Operamr Orspiays. on a Log 7ng g Contred Ceri pies - Oprator Despiays. Arnun cians c *E v2s cci H11 MCRP 1
[ R42 #s Base $w A-w PRRM - P omss Radam Morw1orng Syse m , u %w RCats - Rad Comas & InMatm Sysierm o-, ~
Rrc - Recretaanan s == Conwa Syst c- (
Heac1or Proww System RPS RPT - Recacuawn e mo im .
- Diag.am is Typical For One Of Four Dinsions' SRNM - Stan.e Ra ve n eon uonaor SSLC - Sa*e'y Systam Logc & Corvo6 UPS - LWrt.arruptoie Power Lopy _ _ _
FIGURE 2.2.7a deactor Protection System
Division 1 Division 2 Division 3 Division 4 Jensor Sensor 'enso 'ensor A B C D I ,
I I I I I I ._
,J Divison 1 gRaceway Divison 2 Raceway Divison 3 Raceway Division 4 riacoway DTM l DTM DTM DTM A B C D
% % s. q~.
.) k k.^
- . J,) % . J. .-, ',)
-p s x.
s-
.x.
s
./ * , ' ,' %
~.t g '~~
g .
, / ,. f, . <
/ s '. < ' %
/)s '
~~ ,*
's %
%.} s: s% , %
~~.
P l I I I I I I Divu. ion 1Flace way Divn ion 2 Hacoway Dive on 3 Hacoway Divit .on 4f acoway TLU A TLU B TLU C TLU D i i i t l OLU A l l OLU B l l OLU C l l OLU D l I i
~ ~ ~ -
Eg PL*. _ .
I 4 _ _ _ _'_ - - - -
1 I II I ___I I RPS G2 I RPS-G3 RPS G4 I l RPS-G1, I i i l l l i
l Raceways I l l l l l l ll ! , I I I I I Output Output I same I same I same l l Circuit Circuit l as G1 l as G1 l as G1 l g Group 1 (A) Grup 1 (D) l l l l l Div. 2 Div. 3 l l l l .
, I I I I I I I I Solenad A Solenoid B oas '
tas l lg Exhaust e
l (Manual trip of test logic inter
- aces not shown)
Inst. Air M -
h To scc m ExhaustM Y!- a -
valve Figure 2.2.7b: REAOTOR PROTECTION SYSTEM l
- - p
ABWR D: sign Documont 2.4.1 Residual Heat Removal (RHR) System Desigr) The Rllis system is comprised of three divisionally separate subsystems that Description PC'fo'!n a varictv of runctions utihiing tw fonowing six basic modes of operatmm (1) shutdown cooling, (2) supprossion pool cooling. (3) wetwell and drwell spray cooling, (4) low pressure core flooder (1.PFL), (5) fuel pool cooling, and (6) ac independent water addition. The configuration of each loop is shown on its P&lD in figure 2.4.1 (aligned in the standby mode). The major functions of the various modes of operation include containment heat removal cactor decay heat icinoval, einergency reactor vessel level makeup and .ug, Tented fuel pool cooling. In line with its gisen functions, portione, of th system are a part of the ECCS network and the containment coe mg system. Additionally, portions of the RilR system are considered a part of the reactor coolant pressure boundary (RCPil).
The RilR sptem is designed to safety related standards although it performs some non-.afety functions. The safety related modes of operation include low pressure Hooding, suppression pool cooling, wetwell spray c ooling and shetdown cooling. Non-safety related modes of operation include dyrwell spray cooling, AC independent water addition and augmented fuel pool cooling. Iloweser, RilR also provides a back-up, safety-related fuel pool make-up capability. Ancillary modes of operation include minimum flow bypass and full flow testing.
The RilR systern will maintain the capability to perform its intended safety related functions either following a Safe Shutdown Earthquake or during the environmental conditions imposed by a LOCA, and in each case assuming the worst case single failure. The system will also accommodate calculated movement and thermal stresses. The system is designed so that thc pumps have suf6cient NPSil available. The system can he powered from either l normal olf site sources or by the emergency diesel generators. The RilR
, system is Seismic Category I and is housed in the Seismic Category I reactor l building to provide protection against tornados, (bods, and other natural phenomena. RilR system equipment is constructed in accordance with
! ASME Code, Section !!!. Class 1,2 or 3 requirements (depending on its relation to the RCPil).
The ECCS function of the RilR system is performed by the LPFL mode.
Following a LOCA the RilR system opuates in the LPFL mode, in conjunction with the remainder of the ECCS network, to provide emergency makeup to the reactor vessel in order to keep the reactor core cooled.
i The containment heat removal function in the AllWR is performed by the l containment cooling system which is comprised of the low pressure core flooder (LPFL), suppression pool cooling, and wetw-ll and drpell spray cooling modes of the RilR system. Following a LOCA the energy present within the reactor primary system is dumped either directly to the suppression pool, via the SRVs, or indirectly via the drywell and connecting ven ts. Subsequently, Ossion product decay heat continues to add energy to the pool. The ontainment cooling system is designed to limit the long-term bulk temperature of the suppression pool, and thus limit the long term peak temperatures and pressures within the wetwell and drywell regions of he containment to within their analyzed design limits (see Contaiament Section). The requirements of the containment cooling function establish the n(cessary RllR heat exchanger heat removal capacity.
2.4.1 9/10/91
ABWR Dasign Document i The RilR system aho provides the required safety related capability for bringing the reactor to the cold shutdown condition by operating in the shutdown cooling mode. The augmented fuel pool cooling mode of RilR supplements / replaces the normal fuel pool cooling system during infrequent conditions of high heat load and provides a safety sclated fuel pool make-up capability in emergency situations. The AG independent water addition mode of RIIR provides a flowpath from the plant fire water system to both the reactor and the drywell spray header, in the absence of the normal ECCS network and independent of the normal essential AC power distribution n e two r k.
The divisional subsystems of the Rl!R system are separated both mechanically and electrically as well as being physically located in different areas of die plant to address requirements pertaining to fire protection an.' other separation criteria. The Rl!R system includes prosisions for containment isolation and RCPil pressure isolation.
The Rl!R system is comprised of three separate loops or subsystems, ca< h of which includes a pump and a heat exchanger, takes suction from either the RPV or the suppression pool, and directs water back to either the RPV or the suppression pool. Two of the three h> ops can divert a portion of the suppression pool return flow to a common wetwell spray sparger or direct the endre flow to a conunon dr3well spray sparger. Each of the three subsystems is povered from a separate divisional power distribution bus that can be supplied from either an on-site or off site source. Cooling water to each division of RIIR equipment (heat exchanger as well as pump and motor coolers) is supplied by the respective division of the reactor cooling water (RCW) system. .
The RIIR pumps are motor-driven centrifugal pumps capable of supplying at least 4200 gpm at 40 psid (drywell to RPV). The pumps are ASME Code Class 2 components with a design presuie of 500 psig and a design temperature of 360 F. The pumps are interlocked from starting without an open suction path. The Ri!R heat exchangers are horizontal U-tube /shell type sired to provide a minimum effective heat removal capacity (K coefficient) of 195 Iltu/sec F. The primary and secondary sides of the heat exchangers are ASME Code Class 2 and 3, respectively. The primary side design temperature and pressure are 500 psig and 360 F, respectively. The secondary side design temperature and prescre are consistent with that of the RCW system. Each loop of RilR has its ownjockey pump to act as a keep till system for that loop's pump discharge piping. The jockey pumps are ASME Code Class 2. The RilR system piping and valvo are ASME Code-Class 1, 2 or 3 as shown on the P&lD (Figures 2.4.1). The design pressure and temperature of piping and valves varies across the system. For that piping attached to the RPV, from the RPV out to and including the outboard containment isolation valves, the design pressure and temperature are 1250 psig and 575 'F respectively. For other piping open to the containment atmosphere, out to and including the outboard containment isolation valves, the design pressure and temperature are 45 psig and 219 'F, respectively. For piping and valves outside the containment isolation valves, the design pressure and temperature depends on whether it is located on the suction or discarge side of the main pump. Those portions on the suction side are rated at 200 psig and 360 F, while those portions on the discharge side are rated at 500 psig and 360 F, respectively. The low pressure portions of the shutdown cooling piping are protected from full reactor pressure by automatic pressure isolation valves that are independently and diversely interlocked with reactor pressure. Relief valves are also provided.
2.4.1 9/16!91
ABWR D: sign Document The norinal operational rnode of the RHR systein is in the shutdown oneling enode of operation which is uwd to rernove decay heat frorn the trac tor c ore 1o a< hirse and inaintain a (old shutdown condition. Shutdown s oohng is initiated inanually onc e the RPV has hern deparnurised below the low preoure permiuive. In this inode rat h loop takes sut tion trorn the RPV via its deditated sui tion kne, pornps the water through vs respc< tive heat exchanger and returm ihr c ooled water to the RPV. Two loops (11 & C) diuharge water back to the R!V via dedicated spargers while the third loop (A) utilizes the veswl spargers of one of the two feedwater lines (FW-A).
The heat ternmed in the RilR heat exchangers is trarisported to the ultiinate heat sink via the respective division of tractor cooling water and scrxi(e water. Each shutdown (onhng suction valve it interloc ked with that loops suppression pool suc tion and diu harge valves and wetwell spray valve to prevent draining of the reactor vruel to the supptruion pool. Also, each shutdown rooling suction valve is interlocked with and autornatically rimes on low reu water lesel.
The only autornatically initiated inode of R!lR is the I.PFl. rnode whic h comprises a portion of the ECCS network. The I.PFl. mode is ac complished by all 3 loops of the RilR systern by transferring water from the suppreuion pool to the RPV, via the RilR heat exc hangers. This mode of RilR is initiated automatically upon receipt of a I.OCA signal (low reactor water level or high drywell prenure). The system will also automatic ally revert to the I.PFl. rnode of operation from any other operating unwir upon receipt of a I.OCA signal.
The supprenion pool :ooling mode is ac cornplished by all 3 loop. of the Ri!R system by circulating suppreuion pool water through the RIIR heat exchangers and then directly hark to the supprenion pool. This mode of RIIR is initiated manually.
I
't he wetwell and drywell spray modes of RilR are cac h ac complish,d by only 2 of the 3 subsysteins (loops 11 & C). This is performed by drawmg water from the suppression pool and delivering it to a c ommon wctwell sprav header and/or a common drywell spray header, both via the associated RllR l heat exchanger (s). These containment spray modes of the RilR system are initiated manually. Ilowc/cr, the drywell spray inlet valves can only he opened if there exists high drywell pressure and the RPV injettion valves
[ are fully closed. Wetwell spray serves to mitigate the ronwquences of stram j bypassing the suppression pool.
The augmented fuel pool cooling mode of RIIR is arromplished manually in one of two ways. When the reactor vessel head is rernoved, the cavity flooded and the fuel pool gates removed, the RilR system c ools the fuel pool in the normal shutdown rooling mode. When the fuel pool is otherwise isolated t
f rom the reactor cavity, two loops (11 &C) of the RilR system can directly tool i the pool by taking suction from and discharging back to the normal fuel pool l cooling system. This connection alw provides a safety related make-up path to the fuel pool from the supprenion imol.
One loop (C) of the Ri!R system also f unctions in an at independent water addition mode. This anode provides a crms connection from the reattor l huilding fire protection system header to the RilR system, st outside containment. The connection is made by manually opening two salves just upstream of the tire water tie-in. Fire water can he dircord to either the RPV or the drvwell spray sparger by manual opening of !! c re3pri tise Rili.
2.4.1 9/16/91 l
1
ABWR Design Documsnt injection valve. The fire water is supplied via the reactor building header by either the direct diesel drisen tire purnp or trorn an external souric utilising a dedicated connec tion just outside the reactos building.
Each loop of Rl!R also has both a ininimurn flow mode and a full flow test mod e. The minirnutu flow mode assures that there is pump flow sullhient to keep the pump cool by opening a minimum th>w valve that directs flow bac k to the suppression potil anytime the pump is running and the main discharge valve is (losed. Upon sensing that there is adequate flow in the purnp main disc harge line, the minimum flow valve is autornatically closed.
In the full flow test inode the system is essentially operated in the suppression pool cooling inode, drawing suction frorn and disc harging back to the suppression pool.
Inspections, Tests, This nection provides a definition of the inspections, tests and/or analyses Analysett and together with associated acceptante criteria which will he undertaken for Acceptance Criteria the RllR system.
l t
1 l
2.4.1 -4 9/16/91 l
l
i Tabla 2.4.1-1 RESIDUAL HEAT REMOVAL SYSTEM Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment ' Inspections, Tests, Analyses . Accepsmucecriaanis
- 1. The configuration of the RilR system is 1. Inspections of the as-built RIIR 1. Actual RIIR system configuration, for shown in Figures 2.4.1.a. b andc. configuration shall be performed. those <:omgmnents shown, conforms with Figures 2.4.1.a. b and ci '
I
- 2. The RIIR system operates in the LPFL 2. The ECCS LOCA performance analysis for 2. RIIR system actuation and operation shall mode as part of the overall ECCS network. assuring core cooling shall be validated by be consistent with the ECCS performane RIIR system functional testing, including anaylsis that demonstrates that the demonstration that the LPFL mode (of calcualted fuel PCI' remains below 2200*F. !'
each RIIR loop) is capable of automatically initiating and operating in response to a Validation Attributes LOCA signal.
RI1R Pump Flow (at 40 psid) >4200 gpm Time to Rated How <3ti seconds
- 3. The RilR system operates in the 3. The primary containment performance 3. RilR heat exchanger shall be consistent i suppression pool cooling mode to limit analysis for long term peak pressure and with the containment cooling system
{ the long term temperature and pressure of temperature shall be validated by RIIR performance analysis that demonstrates the containment under post-LOCA system functional testing demonstrating the capability to prevent the bulk conditions. - the flow and potential. heat removal suppression pool temperature from I capabilities of the RIIR system heat exceeding 207 F.
exchangers.
Validation Attributes ,
' t RIIR lleat Excganger effective heat l removal (.apability (K c oefficient) l i >l95 litu/sec#F. [
RIIR lleat Exchanger tube side flow I
>4200 gpm
- 4. A portion of the RilR system return ih>w 4. RIIR system f unr.tional tests shall be 4. RIIR , loops 11 & C cach separately shall be (in hops 11 & C) can be diverted to 'the performed to efemonstrate wetwell spray capable of prmiding wetwell spray thiw of wetwell spray header. flow capaleility. . 2 500 glim.
- i
Tr.bla 2.4.1-1 .
RESIDUAL HEAT REMOVAL SYSTEM (Continued)
Inspections, Tests, Analyses and Acceptance Criteria 3
- +
Certified Design Commitment inspections, Test, Analyses Acceptance Criteria i
. 5. The RIIR systern operates in the shutdown 5. RIIR system functional tests shall be i
' 5. RiiR system (all loops) shall be capable of.. ;
cooling mode to rermne reactor core decay performed to demonstrate operation in the taking suction from and discharging back l heat and bring the reactor to cold shutdown cooling mode of operation. to the reactor pressure sessel. '(Ileat !
i shutdown conditions. exchanger heat removal capability in this-mode is bounded in containment cooling requireemnts - ITAAC # 2)
]
- 6. The RilR system (loops B & C) operates in 6. RIIR sptem functional tests shall be 6. RIIR sistem (loops B & C) shall be capable I the augmented fuel pool cooling mode to - performed to demonstrate the capability of takmg suClion from and discharging supply supplemental or replacement for augmented fuel pool cooling. back to the normal fuel pool cooling
- cooling to the spent fuel storage pool under system.
abnormal conditions.
- 7. The RilR sprem (loop C) provides an AC 7. RIIR systems functional testing shall be 7. Flow capability exists for directing water independent water addition function. performed to demonstrate operation in the from the fire protection system to the RPV t ac independent water addition mode of and drywell spray sparger, via the RIIR' operation. system (loop C), without power being available from the essental AC '
, distribution system.
T
- 8. The RIIR system operates when powered 8. RIIR system functional tests shall be 8. RIIR system shall be capable of operating i
from both normal off. site and emergency performed to demonstrate operation when when supplied by either power source. .;
['
onsite sources. supplied by either normal off-site power i or the emergency diesel generator (s). ,
d i"
$. ~
j ..
?
4
[
! 'I
Tcbla 2.4.1-1 RESIDUAL HEAT REMOVAL SYSTEM (Continued)
Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Crieria
- 9. If already operating in any other mode, 9. Using simulated inputs, logic and 9. Autonutic 1.PFL mode recmfiguration the RilR system autonutically reverts to funcdonal testing shall be performed to logic shall function.
the LPFL mode in response to a LOCA demonstrate the RIIR systems ability to signal, automatically revert to the LPFL mode from any other mode.
- 10. Pressure isoladon vahes are provided to 10. Using simulated inputs, logic and 10. Automatic isoladon and interlock protect low pressure RilR piping from functional tesdng shall be performed to features shall func6on.
being subjected to excesshely high demonstrate opera 6on of automatic reactor pressure. isolation and interlock functions of pressure isolation valves.
I1. Each Ri!R ioop operates automa6cally in a 11. Logic and functional testing shall be i1. .\finimum flow logic shall function to minimum flow mode to protect the pump performed to demonstrate operation of pmvide a minimum flow path.
from overheating. the minimum flow mode for each loop.
- 12. The RIIR system automatically isolates 12. Using simulated inputs, logic and vahe 12. The shutdown cooling suction isolanon shutdown cochng sucdon v4hes to prevent functional testing shall be conducted to vahes slull automatically isolate on a low draining of the reactor sessel. demonstrate operation of the shutdown reactor water Icvel signal.
cooling mode isoladon function-13 RilR system valve interlocks present 13. Using simulated inputs, logic and 13. RIIR system vahe interlock logic shall establishment of a drainage path from func6on.J testirg shall be conducted to funct on.
the reactor sessel to the suppression pool. demonstrate operanon of interlocking between RPV suction vahes and other vahes prosiding potential flow paths to the suppression pool
- 14. The drvwell spray inlet vehes can only 14. Using sirnulated inputs, Irgic and 14. RliR drwell spray inlet vahe permisshe be opened if there exists high drvwell functional testing shall be conducted to hgic shall funcuon.
pressure and the RPV injection vahes are demonstrate operation of drywell spray fully closed. permissive logic.
1
' Tchla 2.4.1-1 RESIDUAL HEAT REMOVAL SYSTEM (Continued)
- inspections, Tests, Analyses and Acceptance Criteria l Certified Design Commitment Inspections, Tests, Analyses Acceptance Criteria l
- 15. The RIIR pumps are interlocked from 15. Ingic tests shall be conducted to 15. An RilR pump start signal is not starting without an open suction path, demonstrate that the RilR pumps will not generated in the absence of an open l
start without an open suction path being - suction path.
available.
l
- 16. The RilR system utilizesjockey pumps (1 .16. Functional tests will be performed to 16. Each jo< key pump performs its keep fill-in each loop) to keep the pump discharge demonstrate the ability of the jockey function.
lines filled. pump (in each loop) to keep its respective RilR pump discharge line full while m .
the standby mode. !
l
- 17. The RIIR system full flow test mmle 17. Functional tests will be performed to 17. Each RIIR subsystem shall demoastrate l allows periodic demunstration of RilR demonstrate operation in the full flow full flow funt tional capability while capability during normal power test mode. . approproximating actual vessel injection operation. conditions during operation in the full !
i flow test mode.
- 18. The RilR pumps have suflicient NPSIl 18. Actual system installation will be 18. Alinimum pump NPSII available exceeds l during all postulated operating inspected, and appropriate measurements as. procured p>:mp requirements.
conditions.
taken, to verify adequate pump NPSii.
t
- 19. RilR mechanical equipment is built in 19. Procurement records and actual 19. RilR equipment shall have appropriate j accordance with ASNIE Code, Sectior. III equipment shall be inspected to verify ASNIE,Section III, Class I,2 or 3 requirements. applicable RilR system components have certifications in accordance with its -
been manufactured per the relevant proper classirvation (as described in i ASN1E requirements. Section 2.4.1). '
- 20. Control room indications are provided for 20. Inspections will be performed.to 20. The designated instrumentation is ,
certain RIIR system parameters. verify. presence of control room present in the control room.
indication for the RIIR system as '
l described in 2.4.1.
i' i
-_ _ _-. _ _ _ _ _ _ _ _ . _ _ _ mm--__ __m_ __.._____.--_-___-
ASME ASME CODE CLASS 1 CODE CLASS 3 FEEDWATER *A*
w_____,___
1 I .
1r 2 A
JL 3
- - - - >4- ?4 g LY t
2 3Q 2
{Q h
p
'^^^ ^
1 r V n 3 4JOCKEY PUMP ,- 7 y N" sx MAIN PUMP TO FROM RCW 'A' RCW*A*
Figure 2.4.1a RESIDUAL HEAT REMOVAL (RHR-A) SYSTEM
' ,,=
- w& .. gr -l y - .- ,. . - . 1 , . ,. ,
~
- f, s s
^
Q -
[ + " e *g t
A - ,
, . _ g l
l ASME ASME CODE CLASS 3 CODE CLASS 2 l
l t TO FROM l
^ l
- : :s- >< f . _ _
u, v ,-- , . _)
I 1 3 1 :
I <
>4- 4J - - - >
y 3l1
- ._ w _ _ ____
4 M 3l2 O _,
i M >< W D
32-. RO I I t
^f^^-
- j 2 1r dk Y
- " -3 Q'
- JOCKEY PUMP s,y T_ _ _ , 3l2 Hx V 1? lj 9 -
n Y MAIN PUMP FROM TO RCW 'B' RCW *B*
Figure 2.4.1b RESIDUAL HEAT REMOVAL (RHR - B) SYSTEM
h HEACTOR BUILDING ASME ASME CdXTERNAL i CONNECTION {_p, ___
CODE CLASS 2 CODE CLASS 3
] ,
PROTECTION ------N ,
SYSTEM Q l 1r >< >< : : l I .I y,r
- - - LA-In, .-, $a 3
ila
- - - >4- -,
, , y q
2 3 N 0 V Q ><
! ! O #
i
_ _ _ _ .____________________ ;; 3 1 P A-no 213 :: Or 3 r 2 a s/ ___
vE II d k 3-
'~'
,, JOCKEY PUMP d-
' E 1r i / Nx MAIN PUMP 1P TO FROM l ROW *C- RCW *C' Figure 2.4.1C RESIDlJAL HEAT REMOVAL (RHR-C) SYSTEM i t
i
ABWR oasion Document 2.6.1 Reactor Water Cleanup System Design t he Lt*W mirin ernnnes paininiate aint am,lsed unpuiines lioni ihr Description s ca< t< >r < < >olarit les s c< irc ulaurig a e u e:In. til the scat oir e initatit thri> ugli a fil t el-d rinin t r a llio t , t he (.t'W M sit 11: n de sigurd its pit 4 0ss a flot1Hil d flow of M of rated f eedwater flow. I hr ( l'W n drugne d for A Lga in?g and w/ C
( l he ( AIW M stritt f ritn 60s cu rss < t Hdaill Ir t eill the loa ( f'st %strill dulatig l
hlalltip, shutdt'%Il alld Inst starldh% lhr C\c cu kJter is dirr(led lti thP lflalll t otsdraiser, radwasic (or supplentols lHu d. 'I hr (d'W ssstritt also prosidt s prtu rned water lti the lea ( tot llcad kl> ras littiilt haf NPY < (Hddtshn.
't he CW mirm ininiiniers RPV te inpriature gradients in inaintaienng tif t ulatioll ill the hollorft head ol the NPY dufing pt fitHlh hbon the f rat lof
, litteillal purlips are uriavailable.
I l bC ktn inHi linC tilrtfugIl thC P( Pil t olitaills tho Inottif (8pPlaled istilalnIn l ValVCs hhNil autolflath ally c lose ill t oslHeine its siglials lf ollt the leak d0 tC( titill % s t r itt , at tuatn>fl (d the bl A$S alnl fligh filter-delninerallier illiet j telnpe r anir e.
l
'lhe CUW ustein is clanified as a violisalett nsterni with a analter l>ortiori of l the systern located outsidt of the prunar3 c ontaitunent peruurr boundats j (P(IPil) and auton.atu alh isolatatile. Systrin l>ijutig alid ( fif tilHent'ils *vithili i the PCPil. in,luding the sm tion piping up to and includmg the outho od suc tion isolation tahr, and all c ontaininent isolatum abra ituluding inten onors ting piping, are ASNIE Se< tu - 111, Sciunk Category 1, Quahts Group A. '!he flow clernent used for Ct EN nstein leak detr< non incets Seninic Category I and Quality Group A requisen.cnis to inaintain stiut tural inirgrity dunng a faalted condition. All nonsafets equipinent is designed as Nonscisinir, Quahty Group C 1.ow prewure piping in the filter-desnincialiirr area, downstre in of the high prenure him L sahes, is designed to Qualil) Group D.
The CUW systeln is a single closed loop systein that takes sut tion liom the tra< tor sessel botto.n licad draiti linic <>r tiir sliutdi>wri < <ndhig sut titili hiic t onnec tion to RllR loop "in" CUW flow passrs through a regeneratise heat
-u hanger (RilX) and two parallel nonregenerative heat ru hangers (NRilF) to 'wo purnps in l . *allel 'l he flow is dn(harged to two filer-dernincraliiers and returnro, through the irgenerathe heal eu hanger to l' redwater lints "A" and "It" 1:ac h pump, NRilX and fiber dominerahn r n iapahle of 50% systesa capacity operatiosi. See l'igure 2.ti.1 for ssste tii arrangement.
I Eac h filter-dernineraliter tewel is isntalled in an indiudual shicided i nimpartrnerit widi l>rovisi<>iis for liaridlitig filter inaterial. Inlet, outlet, scrit, dra;n and other procen sahes ate located outside the filter deminesahier t ompartment in a separate sliielded area togetlier with die nec ewary ing>irig afid aMo( kat0d egilkp!nflit, Proteu equipenent and controls are arratiged so that iiorriial operattoris are condm ted ,t a panel f rom outside the sewel or vahe and purnp compartinent shiehling walls.
2.6.1 1 9/16/91 i
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ABWR ossign Documcnt l'enetrations through t onipartenent walls are desagried so tiiat tlics girc<listle du et i racliation sinne.
A restiote, stiaritsally <igierated sahe or tiie rettir ri litic to ilie lectiwater lities in the itcain tunnel provides long trani leakage c ontrol atul rescrse flotv isolation h provide d by a (he< L vahc ni the 01.'W p ping. !
Inspection, 141>le '.>..fi.1 l>r< >siiles a defiriitiosi ol flic irist r t < tioris, tests, arid /< >r Test, Analyses analpes togethrt witti asui< iatect atiel>tarit e c riteria wiiic li will lie and Aceptance undertaken for Cl*W.
Criteria
)
2.6.1 2 9!16/91
1 Table 2.6.1 - REACTOR WATER CLEANUP SYSTEM :
Inspection, Tests, Anadyses and Acceptance Criteris -
T i
Cerefied Doesgn Commitments W Test. Analyses Acceptance Criteres ,
I t
- 1. The configuratum of she CUW l. Insgwetion of the avhuilt CUW l. Ac tual CUW sntet.: < imfigurat u.n [
svstem is shown in Fegure 2.4.1. (onfiguration shall be performed. tonforms with Frgare 2f3.1. i
- 2. Suction line isolation valves 2. Field test wdl be conducted to 2. CUW isolates when Sifb s attuated autornatically isolate CUW upon confirm that CUW will isolate upon or leak detetta.n lim t is vnwd in t SLCS actuation, leak detecti<m, and SI.CS a< tuation and leak detettion in < lming the priman containment ;
high filter-demineralizer applying a simulated isolation signal prenure be%rdan t'wlathm uhes.
tem perat u re. to the isolation h.gic circui! ,
c
- 3. CUW equipment is proided with 3. Inspc< tion of the as4 uilt CUW 3. Ac tual locath.n of the CUW !
shielding. equipment location will be performed equipment comforms with the reactor !
to show that equipment is Imated in building arrangement drawings (sec j shielded areas. Figure (leser)). [
i I-i i
I
?
i I
_ _ _ _ _ _ _ _ _ _ _ _ __- - ~
1.
9 i
i' ASME '- ASME CODE CLASS 1 CODE CLASS 3 i I l -----N- -
?I g FROM RHR g i !
I I FEEDWATER i 4 i 3
3
.2l3 l s
RHR Loog o I._g /'
, RPV r----- I i
4.J.__ {
N
- (j N_Mm _X NRHX RHX j 1 3 x '
i -
l -4 i
CONTAINMENT (PCPB) RHX j TO MAIN CONDENSER j
- s l
__ FILTER __ I I
TO 3ADWASTE DEMINERALIZER C [
i 4 '
i l T i i l FILTER l i
TO SUPPRESSION POOL - DEMINERALIZER -
- < C I
i Figure 2.6.1 REACTOR WATER CLEANUP (CUW) SYSTEM P&lD 4 i
. 4 _,
l ABWR Design occument I 2.10.23 Circulating Water System Design The circulating water system (CWS) provides a continuous supply of cooling Description water to the maH condenser to remove the heat rejected by the turbine cwle and auxiliary systems.
The CWS does not serve or support any safety function and has no safety design basis.
To prevent flooding of the turbine building, the CWS is designed to automatically isolate in the event of gross system leakage. The circulating water pumps are tripped and the pump and condenser valves are closed in the event of a system isolation signal from the condenser area high-high level switches. A condenser area high level alarm is provided in the control room.
The CWS is designed and constructed in accordance with Quality Group D specifications.
The CWS consists of the following components:
- a. intake screens located in a screen house
- b. Pumps
- c. Condenser water boxes
- d. Piping and valves
- c. Tube-side of the rnain condenser
- f. Water-box fill and drain subsystem inspections, Table 2.10.2.1 provides a definition of the inspections, tests, and/or analpes, Tests, Analyses together with associated acceptance criteria which will be undertaken and Acceptance for the Control 11uilding.
Criteria 21023 1 9/20/91
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Table 2.10.23 !
CIRCULAflNG WATER SYSTEM (N71) i inspections, Tests, Analyses and Acceptance Criteria i i
I e
Certified Design Commitment inspections Tests, Analyses Acceptance Criteris "
i
)
- 1. Flooding of the turbine building will 1. Visual inspection of the installed 1. System isolates.
be prevented by CWS isolation in the equipment coupled with the analyws of event of gross system leakage. the leekage/ flooding charact ristics of the as-built CWS will be performed i using simulated signals to verify syst, rn .'solates on high level. [
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e + s - , r m: - , - ., ..
ABWR Design Documcnt I 2.11.3 Reactor Building Cooling Water (RCW) System Desion Description Tin. reactor building cooling water (RCW) systern aistribute, cooling water during various plant operating modes, as well as during shutdown, and during post LOCA operation of the various safety systems. The systein reinovcs heat flotti plant auxilir. ries and transfers it to the ultimalt heatsink (UllS) via the trattor serTite water (RSW) systran. 'Ihr RCW removes heat from the LCCS equipment imluding the einergency diesel generatort during a safe reactor shutdown cooling futution.
The RCW sy tem is designed to perform its required safe .cactor shutdown cooling furKriori foHravi'ig .1 postulated loss (,i toolant ac cident/ lou of o!! site power (LOCA/l.OOP), assuming a single active failure in any mechanical or electrical RCW subsystem or HCW support system. In case of a failure which disables any one of the three RCW divisions, the other two divisions incet plant safe shutdown requirements, includirig a LOCA or a loss of offsite power, or both.
Redundant isolation valves are able to separate the essen al portions of the RCW tooled components frorn the nonsafety-related RCW cooled components during a LOCA, to assure the integrity and safety functions of the safety related parts of the system. The isolation valves to the nonessential RCW system are automatically or remote rnanually operated and their positions are indicated in the main control room.
Each RCW division inclu.!cs two pumps which circulate RCW through the various equipment cooled by RCW and through three heat exchangers which tramlers the RCW heat to the UllS via the RSW.
The three RCW train configurations are shown on Figure.2.ll..t The RCW system provides three similar complete trains, A,11 and C which are mec hanically and electrica'ly separated. The RCW pumps and valves for ca< h RCW division is supplied electrical power frorn a different dnision of the LSF power system.
During various plant operating modes, one RCW water pump and two heat exchangers are normally operating in each division. Flow balancing provisions are included within each RCW division.
Pump design parameters are:
l RCW A/11 HCW C Design pressure (psig) 200 200 Design temperature (*F) 158 158 Discharge flow rate (gpm/ pump) 2 5,700 2 4.500 Pump total head (psig) 2 80 25 7 6
Ileat exchanger capacities are: 2 60E litu/h 2 56E6 Iltu/h The RCW pumps and heat exchangers are located in the lower ihmrs of the control building. The equipment cooled by the RCW divisions are located in the reactor building, turbine building, and radwaste building.
l 2 11.3 4 9/20/51 i
ABWR Design occument I)uring norinal plant operation, RCW llows through equipenent whic h is norinally operating and requires rooling and all ECCS equipinent except RilR heat exr hangers and einergency diesel generators as shown by open or (losed vahes in Figure 2.11.5.
If a 1.OCA ouurs, a second RCW purnp and the third heat exchanger in each loop are placed in service and autcunatic isolation vahes c an separate the RCW for the 1.OCA required safety equipenent from the nonsafety-related equipinent.
After a I.OCA, the following sequence will be followed:
(1) If the nonsafety portion of RCW is available to the instrurnent air /senice air (IA/SA) cornprenors, the CRD purnps and CUW purnps, RCW flow to these nonsafety components shown on Figure 2.11.5 is inaintained. Flow is autornatically shutoff to other non-essential equipment after the 1.OCA.
(2) If the operator deterinines af ter the I.OCA, frorn essential RCW instruinentatio.1, that the integrity of the su ri-safety RCW systern to the above inentioned cornpressors and purnps has been lost he can shut the reinote operated nonessential isolation valves shown in Figure 2.11.3.
A separare surge tank is provided for each RCW division. Nortral snakeup watar source to the surge tank is the rnakeup dernineralised water system.
For I.OCA conditions, the suppression pool cleanup system (SPCU) provides a hatkup surge tank water supply.
If the snrge tank water level reaches a low level, indicating loss of water out of the RCW system, isolation valves in the supply and return piping to the nonessential equipment including the cornprenors and pumps mentioned above will autornatically close.
The RCW pumps have the flow capacity to deliver required flow to the 1:CCS equipmcnt in each division and the above snentioned comprenors and pumps if the isolation valves cannot he closed.
After a LOOP, the RCW pumps are autornatically powered by the emergency diesel generators.
Connections to a radiation anonitor are provided in eac h division to detect radioactive contarnination resulting from a tube leak in the RilR exchangers, fuel pool exchar.gers, or other.exchangers.
Inspections. Tahic 2.11.3 provides a definition of the inspections, tests, and/or Tests, analyses together with anociated acceptance criteria which will he Analyses and undertaken for the RCW.
Acceptanco Criteria 2113 9/20/91 i
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- Table 2.11.3 i
REACTOR COOLING WATER (RCW) SYSTEM inspections , Tests, Analyses, and Acceptance Criteria l
l l
Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria l
i 1. System < onfiguration including key 1. Intgr< tion of soustruc tion records will lac 1. The systern <onliguration onnforms
,: < omponents and flow paths are shown performed. Visual insaws tion (VI) of the with Figure 2.11.3.
l in Figure 2.11.3. installed RCW system will be performed hawd on Figine 2.11.3. !
- 2. Three RCW trains are mechanically 2. Plant tests arid VI c onfirin 3, roger and electrically independent. 2 Tests and VI of the three indegwndent indegwndence of three RCW dividons.
trains will be umducted which will !
irclude independent and toincident ogeration of the three trair:s to demonstrate onuplete di.idonal ;
separation.
- 3. During various nu> des of operation, 3. The results c onfirm the RCW has the !
RCW has adequate hydraulic capability 3. I.imited system hydraulic tests will l>c water flow capabihtv siwcified by the for o.oling the plant auxiliaries and the omducted according to available < crailic .I design ummitment.
primarv (untainment required for safe nonnuclear heat plant omditions.
shutdown following a design accident or transient.
- 4. Isolation valves as shown in Figure 4. Isolation valves are properly located as 2.11.3 can automatically or remote 4. VI of the installed RCW system and shown in Figure 2.11.3 and are ;
manually separate the RCW for the RCW preoperational tests as follows dennonstrated to operate autoenatically or essential equipinent from the RCW for will be onnpleted. remote narmally to isolate RCW for i the ruineswntial equipment. nonessential from RCW for ewential
- a. Remote-raanual operation of the equipment oncled liv RCW.
isolation valves from the main
. (ontrol room.
- b. During simelated 1.OCA omditions, a simulated I.OCA omdition will be onnbine 1 with a simulated RCW surge tank water level signal to automatically close the isolation valves.
4
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Table 2.11.3 i
i: REACTOR COOUNG WATER (RCW) SYSTEM (Continued) l inspections, Tests, Analyses and Acceptance Criteria i Certified Design Comtr.itment inspections. Tests, Analyses Acceptance Criteria
, 4. (coratinued) t
- . c. A LOCA signal will shut RCW isolation valves which will shut off i ,-
RCW flow to all nonessential j equipment except the IA/SA compressors, CRD pumps and CUW pumps.
i 5. A LOCA will result in the automatic 5. Tests simulating LOCA/ LOOP 5. LOCA/ LOOP signal successfully starts l start of the second RCW pump in each conditions will be conducted for the second RCW pump and initiates j division and start flow through the RCW sfstem whkh confirm the RCW RCW/RSW Ils flow in each division
! third RCW/RSW lit in each division. and it's supp>rt systems will perform it's including the following l function under those conditions. Tests confirmations:
! During LOCA/ LOOP (loss of coolant will be conducted for the RCW, which l
accident /los. ? Mr4;te ,>ower; confirm that after the LOOP, each a. Regardless of which RCW pump conditions, RCW pura ,s and valves are division of RCW pumps and valves was operating during normal powered by the emtr;ency diesel operate with the same division of operation before the LOCA, after i j generators (D/G). emergency D/G power and associated the LOAC/ LOOP simulation DC control pcwer sources. occurs. the first and second RCW pump will start aatomatically, p>wered by the emergency dicsci generator.
l t i
- b. Regardless of which two l RCW/RSV. Ilx's were operating before the LOCA, arter the :
j: LOCA/ LOOP occurs, the RCW motor operated ulve on the third t
- l. Ilx discharge will open automatically. ,
i
, -m , . . . , ._ , , _ - - - , .-, ,- v
. . - . . _ _ - - . - - . - . - - - - - - . _ . - . _ - . . - . _ _ ~ . - . - - _ _ _ _ - - . . . - - .
RHR Hx mm SURGE TANK f' (Reactor Building) "'
EMERGENCY D/G m2 (Reactor Building) 3 y J L FPC HX (Reactor Building) y SPCU OTHERS (ESSENTIAL)
(Reactor and Control Buildirg)
CRD & CUW PUMPS g y7__ ,_I i (Reactor Building) 7 _ _ _ _ _ _ _;
I L_______)
OTHERS (NON ESSENTIAL) _,_____
(Reactor and Turbine Building) ,
I I l N IA/S A COMPRESSORS ,________
l (Turbine Building) I 1 1 I, u I.
L - -C><}- -M DRYWELL EQUIPMENT -[><}- - C>4- - 3 CONTAINMENT CONTAINM8ENT RCW Hx i (Control Buildino) Q RSW -M I
> RSW RCW PUMP 1
RCW Hx I " " ' " '" 9 '
N (Control Buildirg)
RCW Hx C (Control Buildina)
RSW DI I
> RSW RCW PUMP (Control Building)
Elgure 2.11.3a RCW DIVISION - A i
RHR Hx m2 SURGE TANK *
(Reactor Building)
EMERGENCY D'G mm 7'
(Reactor Building) 1 P J L FPC HX (Reactor Building) y SPCU OTHERS (ESSENTIAL)
(Reactor and Control Building)
X 1 CUW PUMP X X 4y _
, ,,_ I i (Reactor Building) i i I -
1 1
OTHERS (NON ESSENTIAL) I p _ ___ _ % _ __ _ (Reactor and Turbine Building)
. _ _ __ _ _ . _ _ _ q I- I ,
I .Ik 1 i l I u 1 1 L - ---C)<$- - M DRYWELL EQUIPMENT -X- - W J CONTAINMENT CONTAINMENT i
- y RCW Hx i
(Control Buildinn) g l RSW WI I
> RSW
- RCW PUMP
! ( ntrol Building)
RCW Hx l- X (Ctntrol Building)
- RCW Hx C
(Coatrol Bu%na) ,
'I I RCW PUMP RSW ' RSW (Contrul Bunng)
Figure 2.11,3b RCW DIVISION - B l
RHR Hx SURGETANK (Reactor Building)
EMERGENCY D/G mm F'
(Reactor Building) 1 P J k SPCU
(Reactor and Control Building)
CRD PUMP g - 4 7 .__ _ _
i (Reactor Building) _i i g _._ _ _ _ __ _ _i i
m______) I OTHERS (NON ESSENTIAL) ..-.__- _ _ _ _ _
( _ ._. _ n _ .__ _
(Reactor, Raduaste and Turbine Building)
I l 0 L.___,_______._ IA/S A COMPRESSORS ,,,_,,_ _ _ _ _ _ _
(Turbine Building)
U g RCW Hx (Control Buildina) e RSW W I I
> RSW RCW PUMP (Connol Bueng)
RCW Hx X (Controf Guilding)
RSW- > RSW t
I
RCW Hx C (Control Bueno)
RSW I
' RGW RCW PUMP (Control BuMing) l Figure 2.11.3c RCW DIVISION - C
_ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ __ _ . , _ _ . . ~ . _ _ . _ . _ _
ABWR D: sign Document 2,12,13 Emergency Diesel Generator System (Standby AC Power Supply)
Design 't he diesel generaion < oinprning ihr thsnnon I, li and 111 uandin Description a( power supphes are designed to rc+ tore power to thou respn ne (: taw IE dntr6 button sntem dhnitun. I'.u h of the thr ee dn ninen of the at. power ostem has its tiw ri diesel generator,
't he three dnisions arc t apable of supphing sot'lhirni at power to satnh Ihr lun< tions of the Emerkeno Core Coohng Sntem (I CCS) network, the IT(.S auxibart and support nstems, ar.d other loads requirett in the plant dedgn.
't he major loads c oinnt of the following setenn for all three dicnioin:
Reddual lleat Removal (RilR) Sptem. Heat for ihnhhng Coolmg Water
( RCW) System, llY,\C Emergen(v Cooling Water (IllLW) Sutem, and Reac tor Servh e Water (RSW) Sntem. In addition, linidorn 11 and til imlude the liigh Prenure Core Flooder (Ill'CI') Sutem loads. fl he 1)hiuon 1 RCIC nsiern is aho part of the 1 CC5 networt, but n steam-driven and therclore not dependent on the standin ac power supph nstem.)
Ea(h Clau IE standby power spiem dnhion, imiuding the diesel generator, its auxiliary i)4tems, arid the distrdnatiori of power to variota Class IE loads through the 6.9LV and 4kOV splenn, a segregated arid separated f rom the other dickions. No inten onnn tion n prinided between the Clau IE dhhiorn. Eac h diesel-generator set is operated independentiv of the other sets, and is mnnn ted to the utiliiv power ustem by manual c ontrol only durmg testieg >r f or htn Harnier. lhe diesel generators are Schemic Categorv 1, and are located within the Reaitor lluihiing (ser ,
Figure (later)). I l
Eac h diesel generator unit h rated at 6900V, so,60lic and k c apable of automati< ally startmg ac ecierating, and supid ying its loads in the sequein e I
and timing required by the plant design. In addition, eac h diesel generator is c apable of starting, au clerating and running its largest motor l l at any time aber the autoinatic loading sequeme is c ompleted, anuming that the motor had failed to start initially, ]
l The diesel generators start autornatic ally ori lou of hta voltage. Under-voltage sensors are used to start eac h dicsci engine in the ment of a untained drop in bus voltage below 70% of the Horninal 6.9kV rating of the bm. I ,0% -
water level scruois and drnsell high pressure senson in rat h dhision aie j aho used to initiate the respn the dicsci start under ac cident c onditioin.
j llowner, the diesch will remain on idle standby unlew the bus i undervolatge sensors trigger the need for bus transfer to the dinel supph.
l Manu.d start capability (without need of dc powei) is also prosided.
t k.at li dhe,se s suhpkked by its own independent luel rank, wbhb ks iorated in an area protn ted hom natural phenomena, and whic h h c apable of mpphing f uel for at least 7 dan of c ontinuota operation.
The standby at power systems are designed suc h that testing arid inspec trori of equipment h pooible during boiF normal and shutdown plant c onditiom.
l i
2.12.13 .I. 9/20/91
ABWR D: sign occument 1.ac h standin ai power supply is inintuned of a duce. phase uniluonous generator and cuiter the diesel engine, the a ngu:e ausiliaries, and a c ontrol panel, l'igure 2.12.13 shows the einergent) diesel generator n steni interfaces, ahir h includes the intestonneitions between the pieles!cd power supphes and the 1)nisions 1,11, and 111 diesel $cnerator stannhv at power supplies.
Ihr trarnier of eac h Class 11: hus to its standhv power , pply is autorniatic ,
should thn bemene nec eswy, on lou of its of fsite power. Altes the < ist uit breaker (onnec ting the bus to the preferred power supply is open, niajor loads are tripped Irorn the Clau IE bus, cuept for the Clan 11. 480V unit substation freders. ~1 hen the diesel-generator breaker is c losed when required generator voltage and frequency are established. 'lhe large anotor loads ase later re applied sequentialb and autoinatically to the bus alte
< losing of the diciel-generator breaker.
Eac h diesel generator is c apable of being started or stopped nianually from the snain control rooin. Start /stop control aiid bus tratisler (oritrol tria, be transferred to a lo(al control station in the diesel generator roorn.
Eac h dicsci generator, when operating other than in test inode, is independent of the preferred power supply. Additional interloc ks to the I.OCA and loswi power sensing tirc uits terininate parallel ope ationi tests and (ause the diesel generator to autoinatic ally resent azul reset to its standbv inode if eithet signial appears dining 4 sest. A loc kout or snaintenanic enode reinoses the diesel generator f rorn servi < c. 'l he inoperable status is indicated in the control roorn.
()ctit es inonitor the (onditioris of the die' 'l ger.crators, and elfert .s(tiori in ac cordance with one of the following categories: 1) Conditions to trip the diesel engine even und-r 1 OCA, 2) Conditions to trip the diesel engine except uinter I.OCA,3) Conditions to trip the generator brcaler but not the diesel, and 4) Conditions whic h are only atiniurniated, inspections. Table 2.12.13 provides a definition of the ienpet tions, tests, and/or arial>>es Tests, together with auociated au cptance c riteria whic h will be undertaken for the Analyses and einerge.ncy diesel generator systern..
Acceptance Criteria i
l l
i l 2.12.13 2 9/20/91 t
l Table 2.12.13 4
[ EMERGENCY D;ESEL GENERATOR SYSTEM i
i I
i inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The tNrce disisions are <apable of 1.a Confirmatory inspection will be I.a The nusimum inads expn te.1 to m<ur l supplyia g sulfuient AC power to perforrned to assure the maximum for eac h dhision (auerding to
- satisfy th- fumtions of the ECCS design huds expected to occur for nameplate rating 9 dull not ex< ecd
! network, the ECCS auxiliary and each divisi<m are within the ratings 90% of the rated power out;ms t l support systems, and other loads of the corresponding diesel (5N10Lw) ci the diesel generzior required by the plant design (see generator. (i.e. 4500kw). t l Figure 2.12.13).
l' l.b Testing will be condu< ted iy 1.b Each u. the thrn units dull pr.= luce
- synchroniring cath diesel generator rated power outgmt (5(Mieks) atoz fpf
} to the plant offsite 5xiwer system and for a pers wt of2 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
]- increasing its output gxmcr level to its .
(momentarv transients cuepted.)
j fully rated load condition. Fach unit will then experience full l load rejectiam by tripping the hud j and verifp.% the unit does not trip.
i i 2. Each diesel generator is rated at 6900 2. Perform a test of each dicsci 2. Each diesel generator attains a
- volts, three phase,60 Ilt; and is generator to ombrm its ability to voltage of 6210 to 7390 volts, and a capable of attaining rated frequency attain rated frequency and voltage. fiequency of 58 M to 61.2 Ili, within W and voltage within t') seconds after seconds after applicati.m of a start reteipt of a start sigrial. signal.
- 3. In the event of a loss of pe=cr. ca< h 3. The automatu nd manual start 3. Ea4 h of the three units starts (sia diesel generator unit is (apable of sequemes will be tested f< r ca. h ca< h enant.at and automatie sigtul), !
1 starting (both enanually and diesel generator unit. ac celerates, and sequences its loads. !
) automatically), accelerating, and The bud sequence begins at 2012 supplying its loads in the proper seconds and ends s 65 seconds sequence and titaing required by the Following appIkati.m of eas h 1.utt.
plant design. It is also capable of the bus voltage will rw>t drop more ;
recoscry following trip and restart of thzn 23% at its terminals.
its largest bud. }
}
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A Table 2.12.13 i EMERGENCY DIESEL GENERATOR SYSTEM (Continued)
Inspections, Tests. Analyses and Acceptance Criteria c
i Certified Design Commitr.ient inspections. Tests, Analyses Acceptance Criteria 4
1 1 (c ontinued)
)
Frequentv sin >uld 1.c testored to within M i of rusmitul. and si.itage should tw
! ressured to wtthin ItG of ineminal within
] tisri, of ca. h luad-sequen< c time interval.
j In addition. the umti largeu tuotor I..ad shall le orig.gwd and restarted after the unit has < mnpicted its wquente, and the lius w4tage will ret mer to fi9LV1 10'5 at tio 1 'f5 Ili within 10 set otuis l
4, Eac h diesel generator unit is caguhic of 4 Each unit will I c tested to auure its blac k- 4. Bla< L4 tart t apabdstr is dernonstrated l talantially startiftg without the need for start <Jpability is Itm(tional. ft.llipwing TIuttuJI stJrl, at t eIeratis tn, atHI 4
external electri(al gM)wer. hus energi/atton f~or rat h ed the thFre l units watItoest assist trotn any rxter sul elet tric prwer.
i 5. Interloc ks to the IDCA and hismf-p>wer 5. Interlocks for the standtn AC power 5. While in a parallel test mode, rat h umt J sensing circuits terminate parallel system will be tested. wdl resert ami rewt to its
- andin nu.de ope ra. ion tests and cause the dicsci following indisidual appli< ation of a le generator to autornatically revert and of-p wer sign d and a IUPP segnal.
reset to its stattdhy mode if either signal appears during a int.
1.
Table 2.12.13 1
EMERGENCY DIESEL GENERATOR SYSTEM (Continued)
Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria 4
(
- 6. Devices monitor the conditions of 6 Using simulated signals, protet tive 6. Successful (irc~it tesiing will f.e the diesel generators, and efIc(t interlm ks and annunniations will bc <'onfirmed for the indnidual drewi action in ac < ordan<e with one of the tested to assure they perform their generator p usective wnsor a< <ordmg ;
folkswing categories: 1) ce.nditions f unctions, in arc >rda.n c with the tu the following:
1 to trip the dicsci engine even under four categor; cal conditions described.
1.OCA,2) Conditions to trip the Category I sen-ors: An n u m ia tion s j diew! engine ex<ept under LOCA and diewi engine trip signals will ;
- 3) Conditic,ns to tip the generator be c onfirmed in o.mbination with a ,
breaker but not the diesel, and 4) LO(.i\ signal.
Condits is w bkh are on'v '
annur .a t e d. Categurv 2 acnwre_Annunoarions and dicwl engine trip srgnals udl l 1,e confirme<8 without a LOC.( but ,
inps shall be hvpaswd when I A Mi\ {
signal is prewns. !
Category 3 sensort Annu:wiations and Generator (ircuit breaker inp ;
signals will f.e confirmed.
Category 4 sensorm Annum ia: Leo signals will be e untirmed.
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, ABWR Design Document l
2,15,12 Control Building Design Description 't he c ontiol building (Cil) is ha ated between the scat to: buihlini., and the turbine building. l'or general arrangernc nt see l'igures 2.15.12a thtonth 2.15.12g.
l'he (:ll houses the cuential etc< trual. (o itrol anni tristr<stiientation equipment, the c ontrol room for th< scat tor and turbine buildings, the Cll heating, sentilating and air (onditioning (llVAC) equiptnent, rea(tor building (Ril) component conhng water purnps and heat ex(hangers, the cuential switchgear, cuentie i .tten rooms, and the stearn tunnel.
lhe Cll is a Seismit Categon I strutture that houses control equipint nt and operations personnel and is designed to prmide minde and tornado protc< to,n.
The Cllis constrm ted of reinfor(ed c ont rete with steel truu roof. 'Ihe Cll has two Hories abme the grade levei and four stories below. The buildmg shape is rectangle of approximately 56 m in the 90' direc tion, approximately 24 m in the 0* dirc< tion, and a height of approxistiately 31 ::i f rorti the top of the base mat.
^1 hc Cll is a shear wall structure designed to ac c oriiriiodate seistriic loads with itri perimetri walls. Therefore, traine members suc h as beann or (olurniis are designed to ac c otiirnodate defrotiiations of the walls in c ase of earthquake < ondition.
To protec t against flood daiiage, the followirig design features are prosided:
- a. exterior wall thic knen below grade level greater than 0.6m.,
- b. water stops provided ia all constru(tion joints below grade,
- c. water tight door and equinment hatc hes installed below Ilood lese:,
- d. waterproof coating on exterior walls,
- c. foundations and wall of structures below grade are designed with water stops at expansion and construction joints.
Water sertnes to the control building c omprise 150mm l' ire protec tioin hne, 150mm cooling water line to the chiller condenser, a 150.iun chilled water heater, 400mm closed tooling water lines, and 700 nun servic e wat r lines.
Smaller lines supply drinking and sanitiary water, avid inateup to; the chilled water systein. Areas with water pipe routed through are supplied with floor drains and prote(tion supplied to route leakage to the floor so that control or ' computer equipment is not sutdetted to water, in these areas where water infusion cannot be tole ated, the auess door sills are raised.
Maximum flooding may occur from leakage in a 700 nun service water line.
The failure of a service water line in the mechanical room of the building may result in an amount of water flooding that room; elevation dif f erences and divisional separation of the mec hanical functions from the remainder of the control building present propagation of the water to the control area.
Early detection by control room personnel will limit the extent of flooding.
During normal operation, the concrete surrounding the steamline mnnel provides shielding so that operator doses are below the value assonated with unc ontrolled, unlimited au ess. 'Ibc outer walls of the control building are designed to attenuate radiation from radioactise materials contained within the teattor building and from possible airborne radiation surrounding the 21512 4 9/20/91
ABWR Design occument
< ontrol building following a IDCA. I hr w th l> oude shichting to liinit the dun t4 hine exlnaure of c ontrol soorn priuinnel f ollownig a I A )CA shiehling for the outdoor au <leanup tilters aho is proviled to allow u rupivian ac t e ss lo the sini c hasist al nitisiirrirait arc a cif ilie ciiritrid buildieng following a IDCA, should at he artjuired.
'l he o,otrol hushhng a not a sented struiluer. 'lhe expou d estrinir re.oh arnl walk of the struc ture are designed for the er<juired pirwure drop.
Tornado darnpers air prosided on all it iniale and exhaust opennigs
~l hese d.itupers are designed to withsiaant the spra ilb d negain-e peruurt ,
inspections, lable 2.1112 prosides a definition of the inspot tions, irsa, and/or anahses, Tests, together with associa ed aurptan< r t riirna whic h will be under taken fri ihr Analyres and c ontrol building.
Acceptance Criteria 1
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!. t l- Table 2.15.12 l i t i e CONTROL BUILDING inspections, Tests, Analyses and Acceptance Criteria
! Certified Design Commitment Ir.spections. Tests Analyses Acceptance Criteria !
i j
- 1. Cor. trol building general arrangement is 1. Plant walk through to check and serify 1. Per Figure 2.15.124 through 2.15.lg..
shown in Figures 2.15.12a through requirements are met.
2.15.12g..
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- 2. Design features are prmided to protect 2. Review wnstruction ret'ords and perform 2. For flo.=hng: I
! against dnign lusis internal and external sisual inspettions of the 11om! control I l' tbxis. featu res. . a. Exterior wa!! thkknew below thed j lesel greater than 0. tim.
- b. Water stop
! c. Water tight door
- d. Water pr =>f r.uting on exterhw I s walls
, c. Foundations and walls of stru< tures below grade are designed with water j stops at expandon and construe tion !
- joints.
! f. Elevation dif ferentes and dividonal i separation of the met hankal t functions from the remainder of the i
! Cli. l l
l 3. The mntrol building is designed to have 3. Performed disuensional inspections of the 3. The concrete thiskness for the steam adequate radiation shiciding to protect Contro" Building walls, ceihng, floors, tunnel wall, thir and tenling shall be j- operating personnel during operatmn and and odwr structural features. greater than 1.6m. The steam tunnel i following a LOCA. interfa<c strutture and control building t wall beluw the , seam tunnel should I
have a uxnhined thi(Eness of Ifnn. ;
3
- 4. The CB is designed to protect against 4. Resiew construction reweds and perform 4. For tornado ;
- desgn basis tornado and tornado missiles. visual inspections of the tornado protection i
features. a. Roof and walls at=we grade de4gewd j greater thavn 0.5:n
- b. IIVAC dampers dedgiwd ihr defferential presure > 1.4ti p4.' !
c IIVAC dampers hase tornado misdic j ba r r ie rs. t d
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