ML20127B641

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Review of Sequences Following Release of Excessive Water in Elevation 8 of Reactor Bldg in Shoreham Nuclear Power Station, Ltr Rept
ML20127B641
Person / Time
Site: Shoreham File:Long Island Lighting Company icon.png
Issue date: 04/30/1984
From: Shiu K
BROOKHAVEN NATIONAL LABORATORY
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NRC
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FOIA-85-199 NUDOCS 8405210470
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{{#Wiki_filter:% ,.Q. e ' ENCLOSURE i LETTER REPORT ON THE REVIEW OF THE SEQUENCES FOLLOWING A RELEASE'0F EXCESSIVE WATER IN ELEVATIO.N 8'0F THE., REACTOR BUILDING IN THE, SHOREHAM NUCLEAR POWER STATION', ~ K. Shiu Y. Sun E. Anavim I. A. Papazoglou ~ Risk Evaluation Group Department of Nuclear. Energy Brookhaven National Laboratory. Upton, New York 11973., .i. April 1984, Prepared for - U.S.' Nuclear Regulatory Commission;- Washington, D.C. 20555 Contract No.DE-AC02-76CH00016 t T gt{g53l6'06 yg e o.a e. ep +w r - eeeeee e e s. ab.

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I o n ABSTRACT The core vulnerable risk resulted frora Reactor Building flooding events is addressed as a part of the SNPS PRA.(1) The analysis was reviewed and re-evaluated at.BNL and the results are presented in this report. The BNL ' review includes both qualitative and quantitative analyses of flooding initiato,rs, operator errors, and. accident, sequences which result in a-vulnerable' core state. An estindte of the uncertainty for, the core vulnerable risk is also included. 6 e 9 E 6 e e 8 O e e g 6 e e g e 8 8 e D- / e e g e e O 9 0 g 4 g e 9 e 4 g e e 6 s 9 a 4 e e 9 e e 0 ,e O e 4 o 4 e e b iii 4 0 e ee ee-e s ee.,... -.neee .me em . eso m e.e - - ap eemusee. ee e.

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a I x TABLE OF CONTENTS Page A BS TR A CT................................................................. i i i L I S T OF F I G UR E S.....................................................,..... v i L I S T OF TA B L E S........................................................... v i i i 1.0 INTR 000CTION....................................................... 1-1 2.0 SNPS HETH000 LOGY AND ANAL YSIS...................................... 2-1 2.1 0 v e r v i ew...............................'....................... 2 - 1

  • 2.2 SNPS-PRA Quantification of the Frequency of Flood, Initiators.. 2-3
2. 2.1, 'Ma i ntenance-Induced Fl ood Ini ti ato rs...............'....

2-3 2.2.2 Ru ptu r.e-induced.In i tia to rs............................. 2 '4 .2.,3 Ini ti a to r E ven t Tree s.....'.................................... 2-5. 3.'O B L ACCIDENT R5 VIEW AND SEQUENCE QUANTIFICATION........ h.':.....;... 3-1 , 3.1 Ficod, Precurso r Frequen cy.................... ;...... :......... 3-1~,

  • 3.1.1 Maintenance-Induces Fl ood Ini tiato rs...................

3-1 3.1.2 Rupture-Induced Flood. Initiators...........:....... 3-4 3.2 BNL Quantitative Review of the Initiator Event Tree........... 3-7 3.2.1 Review of Flooding Alaria Related Procedures............. 3-7. ~ 3.2.2 R eq ua n t i f i c a t i o n....................................... 3-9

3. 3. BNL Re vi ew o f Fu'n cti onal Even t. Tre e...........................

3-10

3. 3.1 ' Qua11 t ativ e R evi ew....... ;..............................' 3-10 3.3.2 BNL Ti rne Ph a se E ven t Tre e....................'..........

3-12 / 3.3.3 Qua nti tativ e An al ys i s.................................. 3-13 3.4 Un c e rt a i n ty Es't ima i:e s...........'..'........... '...... '........... 3-14 4. S U M MA R Y...................................... '....................... 4-1 R E FER E NC E S............. ;. '......... j.........................'........ c.. c. R -1 O g e 0 0 O g 6 e, 3 i e I J O g g 4 e O e e. 9 .? s O s V

I e. LIST OF FIGURES Figure No. Title Page 2.1.1 System event tree for manual. shutdowns with greater than 3'-10" of water in the Reactor Building (Source = CST).......................................... 2-8 2.3.1 Tpti: Initiator event tree for postulated flooding sequences initiated during RCIC maintenance.......... 2-9 2.3.2 TFL2. Initiator event tree for postulated floodi.ng sequences' initiated by an error during HPCI major. ma i nt e n a n c e.........'.................................. 2-10 2.3.3 TFL3: Initiator event tree for postulated flooding sequences initiated by an error during core spray - maj o r mai nt e na nce......................~.............. 2-11 TFL4 Initiator event tree fo'r postulated flooding 2.3.4 sequences initiated by an error during. LPCI major ma i nt e h a n c e.................... :...................... 2-12 2.3.5 TFLS:. Initiator event tree fo'r postulated flooding sequences initiated by an error during service wate,r major. maintenance (i.e., heat exchangers )............ 2-13 ,2. 3.' 6 Initiator event tree for postulated flooding sequences i ni ti at ed, by a HPCI di sc h a rge pi p e b reak ;........'.... 2-14 2.3.7 Initiator -event tree'for postulated flooding seque.nces initiated by a CS discharge pipe break............... 2-15 2.3.8 Initiator event trees for postulated flooding sequences. initiated by a LPCT discharge pipe break.,............ 2-16 2.3.9 Initiator event tree -for postulated flooding sequences i ni.ti at ed by a s ervi c e wat e r l i n e b re ak............... 2-17 2.3.10 Initiator, event tree for ' postulated flooding sequences i ni tiat ed by a WFPS break...........'...... ) ;......... 2-18 2.3.11 Initiator e' vent tree for -postulated floeding. sequences initiated 'by a maximum RCIC suction line ' break....... 2-19 2.3.12 Initiator event tree for postulated flooding sequences ~ in.itiated by a maximum HPCI suction line break....... 2-20 Initiator eve'nt. tree for postulated flooding sequences 2.3.13 initiated. by a large HPCI suction line break......... 2-21 2.3.14 Initiator event tree for postulated flooding sequences initiated by a maximum core spray sucti.on line. break.. 22 2.3.15' Initiator event tree for postulated flooding' sequences initiated by a large core' spray suction line failure. 2-23 2.3.16 Initiator event tree for postulated floo' ding sequences initiated by a maximum LPCI suction line break....... 2-24 2.3.17 Initiator; cvent. tree for postulated flooding. sequences ! initiated by a large LPCI suction line break......... 2-25 2.3.18 Comparison of the HEPs associat'ed with operator actions for singular events and coincident multiple events.......... ....................................'. 2-26 6 vi

E o LIST OF FIGURES (Cont.) Figure No. Title Page 3.1.1 State transition diagram for component-mainten-a nc e i nd uce d fl ood s.................................. 3-16 3.1.2 State transition diagram for rupture-induced floods... 3-17 3.2.1 Problem-solving human error probability vs. time sc ree ni ng va l ue s..................................... 3-18 3.3.1 Phase I of internal f.lood functional event tree....... 3-19 3.3.2 Phase II of internal flood functional. event tree...... 3-20 3.3.3 Phase III of internal flood functional event tree..... 3-21 3.3.4 Phase IV of internal flood func' ional event tr.ee........ 3-22 t 1 6 4 8 ag g O s 8 ~.... g vii f e e+ aume _ o weague e gn e -m

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.p. 1 4,. e t s \\. LIST OF TABLES ~ Table No. Title s Page' Summary of Potential ' Water Sources' and~ Types of 2.1.1 ,) Initiators which may Lead to Release-sf Excessive s' ' Water in the El evation 8 Compartment................. 2-27 2.1.2 Summary of Internal Flooding Initiator Types:ySource, Pathway, Flowrates, and Time to Critical 1 Flooding i 0epth.............................,................... 2-28 2.1.3 Sumary of System Event Tree Entry States by ' . Initiator Type.... 5.......'......................... 2-20 2.2.1 , LER Data for BWR Standby Pumps for the Period of Janua ry 1972 Through April 19784....................... 2-304 3 t 2.2.2 y Frequency'of Online Major Maintenance System in the Reacto r Bu il di ng..................................... 2-30 Sumary of Failure Rates for Major Components 9 2.2.3 ~ Involving Elternal Leak and External Rupture.....'.... 2-31", . - 2.2.4 Conditional ' Probability of Pipe Breald Size............. 2-31 1-2.2'.5 Initiat.ing Event' Frequency Est'imates' Involving Cmpo.nent Leak /Ru ptures...................?.......... 2-3 2 4 2.2.6 Calculated Frequencies for Initiating Events Re-sulting f rom System Ruptures (SNPS-PRA).............. 2-33 2.3.1 The. Probability that Flood Remains Unisolated for X Minutes After Automatic Plant Action, e.g'., Turbine Tri p o r MS IV Cl osure................................. 2-34 ~ 3.1.1 LER Data for BWR Standby Pumps for the Period.of January 1972 through September 1980.................. 3-23 '312 Frequency of Maintenance Induced Flood Precursor's... 3-23, .J, '3.1.3 Fl ood Precurso r Frequen cy,.........'.........'........... 3-24 Maj o r El evati on 8 Equi pment Li st. l.................'.... 3-25 3.2.1: 1 .' 3. 2. 2 - k Timess o Flood Depth of 3'-10"c 1'-10", and'l'-3" in s,- s Re a ct o r Bu il di ng............. '......................... ' 3-28 3.2.3 Muman Error' Probability: Screeni'ng Values........... 3-2g 3.2.4 YEP (Event A). Single Alarm Condition Manual Shutdown 4 (NUREG/CR-1278)....................................... 3-30 3.2.5 ' HEP (Event A), Multiple Alarm Co'ndition (Nomtral. Val ue, PR A Procedures Gui de )................~... e..... 3-31 Vital Equi pment Locations. atil evation 8........'... l. 3-32 3.3.1, 3.3.2 ~ Conditional Frequency of Core Vulnerable.......,....... 3-33 3.3.3 Co re Vul'n erabl e Freq ue ncy............................. 3-3 5 c. l 4.1 Summa ry of Core Vul nerabl e Frequency.................. 4-3 I, .ii i e r i i e 'xs l s g I viii ) g 9.-. 1 _,,.L, c._.

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1.0 INTRODUCTION

At the Shoreham Nuclear Power Station (SNPS) the majority of safety-related equipment are located in the Reactor Building (RB). The Shoreham Reactor Butiding is a cylindrical, building surrounding the MARK Il containment. structure. Water leakage from equipment in the reacter building will drain to Elevation 8 (the lowest level of the RB) via openings and . stairwells since t'here is no structural separation between s.afety systems.. ~ Floodi,ng of the Elevation 8 compartment may potentially disable al1 the ECCS because they are locat'ed in the Elevation 8 compartment. The SNPS-PRA(l') has includ,ed "flcoding as a' conmon-mode event which may disable the $CCS equipment. The'.SNPS PRA assumes that a critical flooding depth of 3',-10" from' the RB floor will disable all the ECCS equipment., Operator diagnosis and iso'lation of the floo' ding before it ieaches 3'-10" - depth is considered in SNPS-PRA. Because of the potentially si.gnificant impact, the'SNP5's evaluation of the core melt. risk due to RB flood' ng warrants a special' review. A ~fiel d ' trip i I I:0 the Shoreham plant has been made by BNL perso' nel for o'btaining detailed ' n informa' ion' on 'the equipmen't and power conti ol la,youts in the 'RB, especially -e ~ t in the' Elevation 8 ccmpartment. BNL has ' determined that there are 'three flooding depths (:1'.D"; l'-10", and 3'-10". ) that are. ritical, to.the c The initiator, event '. trees are thus; * . availability 1 of..va'rious ECCS equipment. revised accordingly. BNL also identified that the random failure of a equipment. protection. circuit br,eaker coinsiding. wit,h the RB flood. event may cause the pr,opagation - of failures to eqdipment powered-by separai;ed Motor Control Centers' (NCC). 'This potential common mode : failure event has also been modeled in.BNL event trees. Shoreham Plan,t Procedure Gui, des, relevant to the RB flooding have been re, viewed by BNL. BNL found that these proce' dure guides fail to require a sys-tematic check of system parameter indicators in the control roan following a RB Flooding Alarm annunciation. This may cause the operator to ignore an abnonnal system' parameter,,espect' ally under a multiple alarm situation (such as a turbine trip). e

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s e 1-2 BNL's revised event trees, quantitative evaluation of core vulnerable risk due to RB flooding events, and an uncertainty estimate for the core vulnerable risk are presented in this report. i l l The report is organized as follows: Section 2 summarizes the SNPS-PRA ap-proach.to the flood sequence identifications and quantification. Section 3 pr'esents the BNL revision both in the methodology and in the quantification.. Finally, Section 4.0 summarizes the' tesbits.. l' [ I 4 t l 8 g g e k-m 3 m ek= w.- e e wede - g> mm. pee m p. eenm =************e*em en'hou-

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[: i 2-1 ( 2.0 SNPS METHODOLOGY AND ANALYSIS ) 2.1 Overview The SNPS methodology for determining the contribution to the risk of the internal floods can be divided into,three steps. l 1. Identification of water sou'rces and pathways to Elevation 8 com-pa rtment. l

2.. Evaluation of operators responses and assessment.of likelihood, of ar-l-

resting' the flood.

3..'Ev'aluatio'n of system responses a'nd identif'icatidn of the sequqnces leading to a core. vulnerable state given a flood.

In the Shoreham PRA approach it was determir.ed that flooding at location.s other 'than Elevation 8 would.be bo'unded by the anal.ysis of flooding at the L lowest lev'el of the reactor. building Elevation 8, since the flood water will. drain and cascade down.to that' level through stairwells and openings. 'All the evaluations of flood are hence focused on equipment at the Elevation 8 level.. '. The volu'me of, water requi. red 'to flood the reactor building Elevation 8 la compartment, with aT1 equipment a' d ' iping insi:alled, is.estimat'ed to be . n p 41,600 gallons in SNPS-PRA for each foot of depth. The fol. losing ~ drainage systems are available to receive.the initial volume of flood water: - Reactor Building Floor ' Sumps - . - Reactor Buil' ding Equipment Sumps ,' - Reactor Building Porous Con' crete Sumps. These systems have total. sump capacity of 4,650. gallons, and total sump pump capacity of 640 gallons per minute, however, they are not included in the analysis. I ~ ~ The potential water. sources which may release excessive water in ' Ele-vation 8 are summarized in Table 2.1.1. For each of these sources, a pathway inve.stigation has been perfonned in the SNPS-PRA, to define the potential for -..---~.m._

e ~ 2-2 flood at Elevation 8. Table 2.1.2 summarizes the water sources as evaluated in the Shoreham PRA. For each water source-the largest possible flow rate has been determined and the time required.for,'the flood to reach the 3'-10" level in Elevation. 8,,have been estimated. These times are also given in Table 2.1.2. These times provide the bag'is for estimating the probability of successful prevention of' flood at the 3'-10" level by ope'rator actions.. A survey of.all, vital equipment by Shoreham'ident'ified a number of ~ , components for the various acciderit mitigatior) systems which could potentially be submerged in the event' of an internal, flood. Ba' sed on this informa. tion, the critical. height of 3'-10" was defined. It was assumed that it' flood' water - exceeds the 3'-10" level, 'all EC' S equiproent would.be disabled. Flooding C scenarios which are arrested before reaching the 3'-10" level, have been' found to contribute negligibly in the core ' damage fr'equency.' Functional event trees were used in.the Shoreham internal' f1' od PNA to o .model the plant response gi,ven an. internal 'fided initiator. The flood, initiator freque.ncy was calculated based on two types of internal ' flood ~ precursors: online 'mainte, nance and rupture of p,iping, valves or pumps. These - precursor frequencies are described in,Section 2.'2.' 'Given the occurrence of these' flood precursors, the progr,ession.of events.,was modeled using initiator., ' event trees. Details of the initiator event trees are presented in Section

2. 3.'

. Since all.th'e ECCS systems are assumed lost given a 3'-10" flood; the only ' '.. available m'eans for cooling the. core'are the feedwater and the condensate pump inj ection. The availability of these two systems depe'n'ds' ort the state of the . MSIVs and on the. ultimate source of the flood (condensate storage tank or suppression pool). l

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l 8ecause'of these' depehdences,'the end states 'of the ini~tiato'r vent trees e were classified into six categories each of which becomes the entry condition for the functional event trees. Table 2.1.3 summarizes the information in a matrix form. Each row of the matrix depicts one of the 17 types of internal ee g O ( 8 e - = - - - - - - - - ~ - - = .m.-m --e . s

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) 2-3 flood precursors, the columns represent the six entry conditions to the functional event trees. The six entry conditions can be grouped into manual shutdown, turbine trip and MSIV closure. Two possible entry conditions are considered for each of these three initiators: flooding due t.o water from the condensate storage ta'nk'(CST) and flooding due'to wat,er'from other sources. Based on [hese s.,ix entry conditions, six, functional event t;rees were de-vel oped. An example is given in Figure 2.1.1. ~ 2.2 SNPS-PRA Quantification of the Frequency of Flood Initiators Twd types' of, flood initiators were cons.ide, red in the.S!'PS-P,RA. 1. F1'oods initiated by an accide.ntal loss of isolation (valve' opening), while, a component in the El'evation 8 area is dismantled for main. tenance.. 2. Floods initiated by a.rupt.u're in th'e ~ press.urized or the non-4 pressurized.part of the piping. ~ 2.2.1' Maintenance-induced Flood Initiators The frequency of the first. type of initiator was calculated by estimating the frequency of maintenance of va'riousl components based on operating experience d'ata. The,LER data ba'se iri. Ref.2 identifies the observed'. failures-f' rom turbine-driven and motor-driven pump failures. The dat,a used in the. SNP.S-PRA are summarized in Table 2.2.1. There are four failure modes for pumps! i.e., leak' age / rupture, does not ' start, loss of functi6n, and does not continue to run. The hourly LER failure rates characterize,the leakage / rupture failure mode, while demand failure rates consider other f ailure mo' des. ' - 1 The following LER rates are found for the four failure. modes in o. s. o, motor-driv.en and turbine-driven standby pumps. Motor Driven Pumos - Leakage / rupture: 6 events /6,777,627 h'rs. = 8.9x10-7/hr. - Does not start, loss of function, and does not continue to run: (5+4+6) events /(13,644 demands)=1.1x10-3/ demand l g 1

2-4 SNPS-PRA assumed that these pumps are in standby status until there i's a demand. The' number of demand used in SNPS-PRA are 12 on the ' average per year (four scheduled tests plus eight other occurrences). Hence, the maintenance ~ frequency for motor driven standby pumps per year is calculated as (8. 9x1'0-7 failure /hr)*(24 hr/ day)*(365 day /yr) + ~ (1.1x10-3/dem'and)*(12 demands /yr) k 2.0x10-2 failure / year. ~ Turbine Driven Pump ~ Similiarly; the maintenance frequencj for. turbine driven standby pumps p'er.

  • year is calculated *as 0.079 failure / year.

8 .There are two mot'or driven pumps' associated with the Core Spray. System, four motor driven pumps with the LPCI System', and four motor driven pumps as-sociated with the Service Water System in which two are linked as a pair to ['theRHRH. eat Excha,nger System. There is on.ly one turbine driven pump as-sociated with the HPCI System and one with the RCIC System. Tabl e 2. 2. 2 ', ,. summarizes the SNPS.-PRA frequencies. associated with ' major maintenance , operation.s. based upon the above evaluation and a conserv,ati.ve, estimate of hea't exchanger on,line maintenance. ^2.2 2~. Rupture-Induced Flood Initiators The frequencies of the initiators caused by lo'ss of system integrity from ~ breaks or ruptures wer'e derived from WASH-1400 failure rates of. major com. ponents involving exter'nal. leak and external' ruptures, bas'ed on a'ssuinptions, mad.e in NUREG/CR-1363 '(Refe'rence 3). This'information has been summariz'ed in Table' 2.2.3. The calculation of each initiator is done by identifying the. appropriate l- , type and length of piping and number o'f components susceptible to rupture and ~. l. summing the estimated yearly rupture rates. As an example; the total riumber of valves involved in the HPCI' discharge system are 3,(2 MOV's and 1 Check Valve); there is no pump involved (Table 2.2.5) and the' total length of piping is76'. Referring to Table 2.2.3, the rupture failure r. ate for 100' of pipe section is 4.3x10-l'l/hr 'and for external failure of a valve is l. . e ee + .w.m _e-e er, en . ene oo.** N w-ausumen-

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2-5 1.3x10-9/hr. The total length of pipe in the HPCI Discharge System is es-timated to be 76' (Table 2.2.5). (3 valves)*(1.3x10-9/hr) + 76'/100' (4.3x10-ll/hr) = 3.9x10-9/hr or 3.5x10-5/yr. Since the flow rates thiough. suction line breaks are time depexdent (f.es, a function of the varying water baad in'th, source) and a strong fun'etion of ~ .the break shape and size, a ' simp 1;fif4 model based on historical experiince and en'gineering judgement is used in the Shoreh.am PRA to describe the con-ditional probability'of break size. Table 2 2.4. summarizes the. classes of. break. size examined. These probabilities; are combined with the frequencies estimated for. initiators associated with core spray, HPCI, RCIC, LPCI, and Service Water Rup,ture/ Leak Suction Systen.f ailure to obtaih.the init' ating' event, frequencies f for non-pressurized pipi,ng.. Table 2.2,.6 sumarizes the frequencies of initiators due to the loss of sys' tem in,tegr,ity from breaks or ruptures. 2.3 Initiator Event Tree.s* The probability of causing a' flood due to component under maintenance or the probability of'not arresting the' flood is calculated wi.th,the help of' initiator Event Trees. These trees are 'shown in Figures 2.3.1 through 2.3.17. A discussion of the P, D, E. I, and A events.in the' event t.rees foll'ows. a. Event P' r Operator removes power from equipment and valves. ,The removal of power from equi'pme'nt and its 'isolatio'n v'alves is a re- '. quired procedure during a maintena'nce. in bo'th foss,il 'and nuclear power. stations.. The equipment and, i, solation valves are electrically discon-i necte' fr,om their associated power supply by pulling and tagging the' - 'd appropr:iate. bieaker at the MCC.. A s.econd qualified person. verifies . the* correct implementation of the tagging order and placement of' the clearance tags. j A human error probability (HEP).of 0.01 is assigned for this operator action.. This value is determined'using the probability data given in NUREG/CR-1278(4) (p.20 '23). l l j l t .....n...

a 2-6 b. Event D - System not demanded. During the maintenance process there is a possibility that the safety j systems will be demanded because.of a transient challenge. Isolation. valves will, automatically open if the operator has failed to remove power from the isol.ation valves (Event P). c'. " Event E - Operato.r maintains isolation'. .During on-line maint.enance with the equipment disassembled, the isol.a-tion valves need to be maintained in closed position throug.hout the duratipn of the' maint'enance process..However, an operatoi er~ror could . I, inadvertently open isolation val'vhs. SNPS concludes that it is un'likely that the operator will manually open these valves locally in the RB and fail to notice the flood. Opening "of the isolation. valves at the MCC is also concluded by SNPS. to be unlikely. The. remaining. possibility is that the valve is opened from the control ' room (given Even.t P). The panel switch, could be activated by th,ree. events. These events are: the operator mistakenly operates the . switch; a command' fault' to the valve; or the operator in' advertently cperat'e's the switch. 'Th.e, probabilities for these. events.are' 10-3, i d, and.10-2,.respectively.. d. Event I .F.lood annunciation. The excessive water in reactor bui.lding is a'nnunciated by a.l'anns in the control room. The.probabilit'y of the operator to fail to' notice the alarm (the light is in a "b'ack" panel) is assessed at 10-3, l

e. ' Event A' - Operator diagnoses and responds to isolate the flood.

l The operator ~must identi.fy the source of and isolate the. flood before ' 'it readhes the 3-10"* level.' 'Th'is event is co$si.dered by' SNPS under two conditions 'as foll'ows. 1. Operator isolates flood after auto occurrence, e.g., turbine trip or MSIV closure (Event A ). Multiple alanns will occur in the A, control room at the same time as the flood alann. l i e m- _y,,,,g._ .g_ .m..

6 3 2-7 i 2. Operator isolates flood after manual occurrence, e.g., power oper-ation or manual shutdowo (Event Ag). Only the flood related alarms will annunciate'in the control room. The HEP data provided in SUREG/CR-1278(4) (1982 Edition, Chapter 12) 'are applied by SNPS for their ' evaluation Figure 2.3.18 and Table E 2.3.1 show the time var.ving cumulative, HEP,for b.oth the, single and the inultiple occurrenct conditions. 4 S O g g O O g e 9 9 9 0 9 e s 9 e e e 4 e o e 8 g 4 0 e p 9 e 8 e 6 0 9 e O t D 4 t e g + g 4 8 e j-8 e e 8 e e e 6 e f e 9 e e

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6 ~~ 1.i n...... BallIhlem 34WIee Slaltlk ((estillemet feOS.') MMIM tefilelass Islisell? senests. .msns w om- :mau ec(m> m(m a(Cums n one stw wr (As(teAlle ITP( M '5T11(n 8 isAlle as M5ruseu as ht Stenssi test 10. 01: ChesDillim oritaIOR Wilth43M asLT $dtpale(E ' IN M,es Tr) 9(aCleA lP. I olismanGa se, en s taes 14. M M. M5ael$ lu. angesettalte gsris agg5 gggggg' j 4 tai le a seassent 9e11 la. A A seitv Aest - ilflee me u s.esuena .usenu mir a ssuw - w(acann o I g III sts It I 1 I A e I m. .j-1 e.e, ,s. e s m.). 8 sensniaa sannares (n) .:'Aa 4.II-7 5.e e e ne e.es mi i.u. '*4 a 8 3-inia f.rt.: 's-ai e m. i meist inte til !8" 2 tia . 2.u-7 , t-e e l 8 "3 7.It-s i.e m e L ~ e 6 (5) s.. e.., t I A 4 83 3 l'54 4.6f.8 5-e ' e 8 penly (g esset s-g,, l j e.15 a me) ggg g,3g,g steintus As rosaa se y t musuus } f. i..v n n e - i.is.s i.e - z.rt-. s-e - s.u-r l j .i-i e i. iw.. a,... 2,.... ~ i r i ~ Figure 2.3,9 Init'lator Event Tree for Postulated Flooding Se-1 ~ ~ .l. quences Initiated lay a, Service llater Line Bre.ak si s y l 3 s t ~*

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j Figure 2.3;.11 Initiator Event Tree for Postulated Flooding Sequences - initiated by a Maximusi RCIC Suction 1.ine Bl'eak s i e e i 4 y b l

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  • I e

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i i i t Figure 2.3.13 Initiator Eveiit Tree for Postulated Flooding Sequences g Initiated lay a large llPCI Suction Line Break ~ 4 I 1 4 l 1 l a l I I

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e. eel lett 6.af.9 es e L

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i, i ..e ..i. t l. g, Figiare 2.3.17 Inttlator' Event' Tree for Postulated Flooding Sequences I Initiated by a large~LPCI Suction Line Bre&k L G s y

t-. e i .......l,......,s..?....__..+.i.._......a N (.- e \\ 2-26 s 1 z' ~ k Ys 1.0 \\s. t,g200 s W A!2-12*S 117.2 MU 4 I for single event. i.. U

    • g vants.1278 REY.2 Mt?

s for e.e anone of. ulti- \\' ple event $ \\.. 5 A ,,.1 .g .g:. s,s.. g .s 5\\ W c.* C. \\ w k -- '%'I.*!PLE DDT

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x* t a.: .\\ ~~m a. ta.1.. -t 1 i s' ( . i., .a go ga 3a ao so 4 10 Ti. ns (Minutes ) 1 t s I 1 g '. F.igure 2.3.14 Comparison of the NEPs Associated with Operator Actions for Singular. Events and.' Coincident Multiple Events 4 I Q 'i. ,s. . s.> T.., l i .__..m.

.o 2-27 Table 2.1.1 Summary of Potential Water Sources and Types of Initiators Which may Lead to Release of Excessive Water in the Elevation 8 Compartment ~ No. of Source Quantity (Gallons) Lines Systems Involved ' Suppression Pool,. 160,000* 8 CS,LPCI,RCIC HPCI 4 CS,HPCI,RCIC' pondensate Storage Tank (CST) 550,000 Reactor Primary' System ** a) 42,928-b). 152,,9.28 Scree,nwell (,Long Island Sound) Unlimited 4-Service Water Water Fire' Protection System ,Stor, age Tank' 600,000

Many, Fire Main
  • Total water' volume in t'Se suppression'po61'at the high water level mark is

' 608,500 ga11o6s., However, only a portion of.the water can be drained- ',l. through, ECCS pump suctfon piping. l t

    • Figure (a) includes water'from the. bottom of the core to normal ' water level

? in the RPV. Figure-(b) includes (a) plus condenser hotwell water. 8 g g e g e g 3 e D s .? ,l 't I* ! / '" l - p e e 9 g .m e .y -,,,-c.- ,,w-y-,-.,. ~ ~..,,, -y .--------.m--. +s--,-

2-28 Table 2.1.2 Summary of Internal riooding Initiator Types: Source, Pathway, Flowrates, and Time to Critical Flooding Depth Elevation 8 Flooding Time . Flow Rate (Minute,s*) t source Location gixn* 3'-10" Sup'pression Pool HPCI, Pump Suction 9600 ' 17.6 l RCIC Pump Suction 1500 - 10.6 ~ LPCI Pump suction t (Max /Large)** 17000/8500 9.4/19.0 CS Pump Suction. 13000 12.0 1 LP.CI' Pump Suction 10500 I5.0 (1 Pump Runout) CS Pump Discharge 6850 23.0 (1 Pump Runout) Co'densate'St' rage' n o Tank (CST). HPCI Pump Suct' ion

  • '13.0/27.0 (Max /Large)**'

1200/6000 RCIC Pump Suction 2100 76.0 CS Pump Suction (Max /Large)**. 1200/6000 13.0/27.0 HPCI Pump Discharge - 4350 37.0 -(Design) t Service Water RHR' Heat Exchanger 8000 20.0 .(PumpRunoat) WFPS 'Ru,pture of18" Pipe

  • 4000 40.0
  • These ' flood times were calculated based on a failure of, the sump pumps to successfully oper' ate and a 41,600 gallon per foot depth in the reactor building given in the Shoreham FSAR.
    • Lahe, flow rates ass med to be 1/2 maximum' flow. '

\\ e -e .-e...

a 2-29 Table 2.1.3 Sumary of System Event Tree Entry States by Initiator Type SYSTIM EVfMT TRC[ [NTRY C0tmlil0N FRCQUCNCY (Per ha Tt) InitlA!0R M.0 ' N-C T d' TC 50 5C T,gg " ' t.8a l 0* 8 1.8:10'8 7.6:10'8 4.3:10*8 ' ~ T,,,. ' 5.rair 5.2 air - 2.5mic) 5.Osic6 r r T '3'.0:10*8 l.lal0*A gg3 T 5.8 10*I 4.3:10*8 gg, T .3.6:10'8 4.1:10-8 it$ 1.0a10'I 1.3a10'I T gg T 6.4a l'0'I 3 5:10'# ggy. T 1.1a10 5 2.0 10-5 9.0:10*8 ggg T 1.3a10*8 2.fa10*I 5.Salo'I gg T 2.3 10 2.850 1.410'8-fLIO 1.810., ,3.4 10'8 8 ,1,.5a10 .. ngg T 1.0a10*I 2.1 10*I TL12 , 2.6:10*8 7.8a10 8 Tggg3 .T 1.6510*I

2. gal 0'8 gg,

'T 4.4a10'8' 8' ,2.5a10' ftl5 .T 1.141 8.1:10*I

  • 6'.6al0'I R16 T

2.4al0*I 8.8:10*I 2.4 10'I ggg TOTALS 1.6410 5 8.2al(I 2.2 10-5 3.4:10'I 1.7a10 5 5.5a10'8 s ) = 6 e 6

= 2-30 Table 2.2.1 LER Data for BWR Standby Pumps for the Period of January 1972 Through April 1978 Does Not Standby . Leakage. Does Not Loss of Continue Standby Pumps ~ Demands Hours. . Rupture Start . Function To Run Motor.' Driven , 13,644'* 6,777,627. 6 5 4 6 Turbine Driven

  • 1.,820 868,03,3 1

. ' 6' '. 5 . Table.2.2.2 Frequency.of Online Mafor Maintenance , System in the Reactor, Building. Frequency (Per Initi,ator System' Year) SNSP-PRA Event Tree . Core Spray (Motor Driven)- t 0.042' ..TFL3 ALP.C'I.(Motior Driven). O'. 08.4. ..TFL4 " TFL2 HPCI (Turbine Driven),, 0.079 RCIC (Tuibine Driven') 0'.079 . TFL1 ' Servi,eWater'(RHR'or c RBCLCW HK),(Motor Driven) .0.042 TFLS.. l, s ~

2-31 Table 2.2.3 Summary of Failure Rates for Major Components Involving External Leak and External Rupture Total Failure Rupture

  • Parameter Rate Rate /Hr (Mean)

Reference Failure Rate /Hr Pipe' Failure Section 8.5E-10 WASH-1400 4.3E-11 (100') External Failure.of a Valve, ,2.7E-S WASH-1400 1.3E-9 ~ External Failure of 1.5E-10' a Pump' 3.0E-9 WASH-1400

  • Based upon the operating experience to date, given that a failure occurs, the ratio of external leaks to complete failures appears to be in the range of 20 to 1.

This is substantiated by.the specific dat review cited in th.e text for.v'alues (la to 1) and data. published by Bush 4g)' on pipes (4 to.1 up to .30 to 1). Because the. internal ' flood e7aluation is based upon initiators with substantial, flooding rates, i.e., short' operator response times, only

  • the catastrophic or large ex'ternal rupture failures are treated in this-evaluation.

l-Table 2.,2.4. Condit.ional Probability of Pipe Break Size ~ Break Conditica. 1.' Flow Rate Probabilityf, Size Characterization - Maximu'm Guillotine Break 100% 0.05 ~ Large Substa'ntial Rupture 50% 0.10 Smal l.* Localized Rupture in Ductile Material 13% 0.85 i i

  • Remainder' of the conditional probability was all'ocated to small breaks.

e g ee w ~,

2-32 Table 2.2.5 Initiating Event F'requency Estimates Involving Component Leak / Ruptures VALVES PIPING ESTIMATED ~ INITIATOR SOURCE LENGTH (FT)/ FREQUENCY /, MOV MAN CMX PUMPS SECT /DIA (IN) YR , rwc1 01senarge, CST /SUPP. .2 0 1 0 .76/1/14 3.5E-5 T(5 ~ y CS l Discharge SUPP '4 0 2 0 128/2/12~ 6.9E-5 T yg7 LPCI 2.5E-4 01senarge. SUPP 14 4 4-0 240/6/16 ~ ' ' ' ' ' ' ' ~ " T FL8 ~ Service Service Water-Water 4 4 4 'O 7,15/3/10-20 1.4E-4 y FL9 , 1,.,1E-5 g WFPS WFPS 'l , 157/2/6-8 *

  • Eti.0 RCIC**

3 - CST 1 1 1 1 70/1/6 3.5E-5 Suction FL11 HPCI** Suction CST ** 1-1 -1 1 87/1/1df 3.5E-5 Tytyg,TFL13 'CS **.

t..,,

120/2/12' ; 4.9E-5. ' Suction

  • CST *. ' '

.2 '2 ,2 TFL14,TFL15 LPCI** Suction SUPP 4 4 L20/2/20 5.2E-5' TFL16,Tpt17

  • CST is assumed to be the source.
    • Suction failures are also classified by flow rate.

O

_....- ~.. 2-33 Table 2.2.6 Calculated Frequencies for Initiating Events Resulting from System Ruptures (SNPS-PRA) Initiator Frecuency (Per RX Yr) Press 6rized Piping. HPCI Discha.rge Break, TFL6,' 3.5x10-5 CS Discharge Break, TFL7 .6.9x10-5 ,2.5x10-4 LPCI Discharge Break, TFL8 ~ .SW' Discharge Break,'TFL9 1.4x10-4 1.'1x10-5 'WFPS Discharge Break, TFL10 Non-Pressurized Piping RCIC S'uction Failure, TF11 (max) 1.'7 5x10-6. HPCI Suction Failure, TF12 (max) 1.75x10-6. HPCI Suction Failure, TF13 (large) 3.5x10-6. -CS Suction Failure,, TF14 (' ax)

2. 5x10-6.

m '4' 4.7x10-6. [ CS Sueti,bn Failure, TF14 ('laige) LPCI Suction Failure,'TF16 (max) 2.6x10-6. LPCI Suction Failure,' TF17 (large) ' 5.2x1'0-6. '* Modified based upon engineering judgement made on the size of low p'ressure - suction line breaks.- 5 i

li.

I . 1 1

  • elm

= .e. y,,. ~_

z P 2-34 Tabl e 2.3.1' ' THE PROBABILITY T)(AT FLOCD REMAINS UNIS0 LATED FOR X ' MINUT AFT.ER AUTOMATIC PLANT ACTION: E.G., RIRBINE TRIP OR MSIV CLO.SURE X P(fgr, multiple event) P(f'or' single event) 1 1.0 1 0.1 10' 15t + 2nd = 0.'54.' 0.11 .0.01 ~ ' 20 0.01,i - 1.1E-3 3,0 2.0E-4 i 60 0.0011 . r '1500 - 1.1E.4 - 1.1E 4 .~ s + i n J ee 0 w - -

...,..,.,---.,.,,,,n

.-_n-.-,_,-- --,-_-,.n,,..,,_,---..,--c.

. _. ~ - ~ a 3-1 3.0 BNL ACCIDENT REVIEW AND SEQUENCE QUANTIFICATION This section discusses the quantification and review of the internal flooding accident sequences in the SNPS-PRA due to system maintenance and pipe ruptures. The section.is organized as follows. Subsection 3.1 presents a s'ummary. of the appr'oac'h used by BNL. to calculate the initiator frequencies. S.ubsection 3.2 discusses BNL quantitative review'of~the initiator eve,nt trees, and Subsection 3.3 presents the functional event tree analysis and evaluation. 3.1 Flood Precursor Frequency This review revi, sed the, assessment of the frequency of the flood initia-tors in two ways. First the experiential data' for the estimation of the var-. ious failur'e rates were revised to in'clude recent'avents. Second, the models for calculating the frequency of' floods (or probability 'per.ye'ar of rea'ctor. ope,rationF have 'beeri improved by r,emoving unnece,ss'ary conservatisms'. As 1.t was already. discussed 'in Section 2.2, two types of initiators were., con-sidered: a) maintenance-induced initiat'o'rs; and b), ruptureeinduced initiators. The revised frequencies 'for these types 'ck in'itiators are presented'in the following two subsections. 3.1.1 Mairiteriance-Induced Flood Initiator $ A flood can be, initiated during the,ma'in,tenance of a component of the ECCS or'of another s'ystem in th'e Elevatior) 8 area jf the maintenance process c requires dismantl.ing of'the component and one of the isola. tion valves o' pens ' ' ' inadvertentiy whil'e the component is being maintained.

  • The components that contribute to thes'e' initiators are-the pumps and the heat exchangers in 'the Elevation 8 area., These ' ave standby components that can fail ir} a time-dependent fashion,while on standby. Periodic tests are performed to check -their ' operability and if found failed they are put dnder l' repair.

A Markov model that describes the stochastic beh'avior of these components has been developed and quantified. The important characteristics of this model are as follows: f = e D D ++ --e.-.= e e e e. e e.. wwwiw-- v --r- -c ,,p- ---_,.m- -y s ,--,in-. ,y., ...p-._ ,___-,,,-.wy_w--w. ,-y.. m,-., yeeey--n ..y.-.y e

I eu% 3-2

1) The component can be'in six states (see Figure 3.1.1).

ii) In state 1 t'he component (pump, heat exchanger) is available, that is ready to start operating if asked to do so. iii,) The component'while on, standby can fail with exp'onentially dis-tributed times to failure. A failure brings the component into ". state 2'(seeFigure3.1.1). iy) The failure.rema. ins undetectable until a test is perfomed or a real' challenge is' posed to the component. A test that will find the cont., ponent in state 2 will initiat'e a repair action. The same will hap-pe'n folloWing a real demand for the component. y) There are three repair states. States 3 and 3' in which the com- ,ponent is under repair *while' the reactor is online, and State 4 where the ' component is' under repair with the.' reactor shutdown. vi).Following a test' that finds th.e. component failed' and before' the' dis-mantling of the component,, all the.appropria'te motor operated valves must be closed and' their breaker's racked out from the corresponding l. ..MCCs. ; There'is,'however, a ' chance' that.the. operator will not remove the breakers.from the.MCCs. leaving then the. MOVs able to. open fol, If th' probabil'ity,of such an error is P,. lowing a signal to do so. e then a test brings t' e component from State 2, to State 3 *with h l probability 1-P (bre'aker removed) a'pd to State '3' with probability p, The component remains in Statis 3 or 3' until the repair is completed

  • vii) and then it returns to State 1. or until the allowable outage time,is exhausted and then the component transit to State 4lwhere the repair I

i 'f continues with the reactor shutdown. When the repair is completed, : i the reactor is brought ba'ck online and the c6mp'onent returns to State

1. '(Transition 4 to 1).

G t 8 e O --.*~ =~g, -.m s - mr~ m

m. we.-,=e m

- s

i, es 3-3 Quantification The solution of the model requires the quantification of the following . parameters.

1) The catastrophic failure rate A.

This failure mode implie,s such ' failures that. require' major maintienance (disman'tling) of the com-ponerit. The SNPS-PRA used the da'ta presented in Table 2.2.1 from Ref.. 2. BNL has updated this' table using ' additional data included in an

  • updat.ed version of R.ef. ~2('Ref.6).

The new data are summarized,in. Table 3.1.1.. Naximum likelihood estimators for the failure rat'es number of fa,ilures A=(total, operat.ing time) yield A=5.7x10-5/hr for Turbine Driven Pumps and A=3.3x10-6/hr for Motor Driven Pumps . The mean times to repair were assumed 100 h'rs and 50 hrs for the ii) ttirbine. driven and the reactor driven pumps l, r'espectively. 'Thds, y y=10-2/hr for Turbine Driven Pumps' .and. n=2x10-2/hr for Motor Driven Pumps. iii) In the BNL' revision of the SNPS-PRA, the frequency of trahsients ' involving MSIV closure, ha~s been assessed at.4.42/yr. ~Thus,.the ' f,requency of. tr.ansients on an hourly basis is ,A =5.0x10-4/hr D 1.,y)-', Testst.are performed.every 3 months (4~ times a yepr) fgr, both motor - i i driven and. turbine-driven pumps. The allowable outage times are 14 and 7 days for turbine-driven and motor-driven pumps, respecti,vely. 1. , - += .-wy_.--- _-.8 ,.. = .w, + - - = + +_ 4..-p

g. +

. =~- ~ :_ a m. .a 34 v) The probability of not racking out' the breakers of the isolation valves (P) is assessed in the S.NPS-PRA as 10-2 The same value is-used in these requantifications. vi) The mean time for inadvertentiy activating a particular switch in the contiol room hhs been assumed equal to 10,000 hrs., This' implies a rate.of A =104/hr. o Quantification of the Mairkoyian mod.el'with.the nuinerical values of.the paramet'ers mehtioned, ab6ve yields the probabilities per year for the various maintenance induced floods.. The results are tabulated in Table -3.,l'2. Additional assumptions are: the ' Core Spray System consists of two ' motor driven pumps, the LPCI consists of four mdtor driven pumps,and that,RBCLCW heat ~ exchangers are equivalent to motor driven pumps. '3.1.2 kupture-Induced Flood Initiators A flood cah be initiated if a rupture occurs at 'any point in the pressure boundary of the va,rious systems in the Elevation 8 area. Such' a rupture will-involve,one of'the following three types of components:.1) piping; 2) valve; I e 'and 3) pump. The.model a's,sumes. that'.ca,tastrophic ruptures occur, in the,fol,- lowing way'. A component fails in such a way that if it is demanded to ope-- rate then "a catilstrophic rupture (large enough to allow the flow rates neces-sary.for t,he fTood sizes of interest to this' analysis) will' occur. That is, the component transits first in a rupture-vulnerable' state and then, when a de-mand occurs, it ruptures.~. A Na'rkov model that decribes this stochastic behavior has been developed. and quantified. The model is graphically depicted in Figure 3.1.2.. The basic characteY.istics of the' model are as follows: g (i) The system in question (HPCI, RCIC, LPCI, CS, RHR, RBCLCWHX) is in state where it is available to perform its function. 1

  • e

' ' } - ~. -. -^ e. 3-5 (ii) The system transits to State 2, which is a rupture vulnerable state with failure rate A *R (iii) If a-demand occurs while.in State 2 a flood is initiated. A demand occurs whenever a transient, a manual shutdown or a test occurs. We ,distingutsh thre'e flood states: -State. 3, which 'is a rupture trig-gered by a transient :involvin' 'an MSIV clos,ure; State 4, which 1,s'a g rupture trigger 5d by a t'urbine-trip transient; and State 5 which is rupture trig'ered by a manua,1 shutdown or an' equipment, test. ~ g Thi solution,of 'this'model ' yields 'the probabilities that the system will occupy States

  • 3, 4 and 5 denoted by P, P, Pg*, respectively. The'se 3

T probabilities at the end of one-year period provide the frequency of rupture-initiatid flood precursors. The, expression for these probabilities is R )/A '(1-e-At)fx] Pj(t,) = F [(1.e-A R R ~ (1). .A-A ' where i = S, T F is the. number of tests per' ys'ar. Aj jis the rate of irrival of a tran'sient of ty'pe 1. (1'=S',T). l I' ' 1 "is' the r' te of catastrophic 'rutpure' failure in the~ system R a ~ 'and A is the rate of arrival.of any tran'ient (1=A +A +A ) s 3 T M For the m'anual. shutdown the corresponding expression is AAMR XR-F(t.)=F[A-Ag(1-e-AN)/A-(l'-e-At)fx+ A-A (e-A T EAT)] M R R R (2) Quantification For a.given* systdm' having piping of length' L, n'y vilves and h pumpi. I p the failure. rate Ag is equal to AR

  • LA'+D A +np p (3)

A vv .m 9 .e,r...

_,j'J. 3-6 where A, Ap are the catastrophic rupture failure rates for valves and y pump and A' the same failure rate per unit of piping length'. A search of the LER, has indicated that at least one pipe rupture (welding failure) has occurred in the ECCS pip.ing in the 215 accumulated BWR years (see Ref.8). This provides a maxjmum likelihood estimator for the ruptu're failure rate- ~ of (1/215yh5.31x10-7/hr). Assuming, as in the SNPS-PRA, that only one out o'f every twenty ruptures will create a bre'ak. 'laige enou'gh to generate floods, 'of the. sizes. of' concern to this' analysis the catastrophic piof ng rupture ra'te becomes A=2.7x10-8 'This of course is applicable for the total l'ength, of safety related piping (denoted'by L). ~ F,or a pa'rticular system witlT a total of. piping length 1, then the. . catastrophic rupture ' rate for piping' becomes 1"=(f)x2.7x10-8 /hr (4 ) ' where /L denotes.the fraction of the total length.of the piping that belongs. ~to the particular system.' 4-For.the rupture rates of the valv.es and the pumps, the WASH-14,00 v,alues. weie used,(s'ee Table.G.4 4, in SN' S-PRk)'. Using the length of piping, number P of valves and pumps provided in Table G.4-5 'of the SNPS-PRA, and by virtue of - Eqs.1-3. ' The total failure rate AR for the various, systems alor}g with the - probabilities P, PT and Pg were ca,1culated.

  • The results are tabulated 3

in. Table 3.1.3. A total of 13.51 transidnts per year 'were assumed (4.42 MSIV closures, ....4.89' turbine 1 trips,and,4.2 man'ual, shutdowns).

  • p The splitting between maximum and'lar'ge ficods for initiators TFL12-TFL13,
  • TFL14-TFL15, TFL'16-TFL17 was done as in the SNPS-PRA, that is,1 to 2.

The additional factor of 20 used in the SNPS-PRA to account for non-pressurized piping,is not assumed in the BNL quantification. ,s- -e-- -. - ~ w

~ * ...... -...:.. ' _ :=. ..= - -2 e~- -- m.a 3-7 i!- BNL Quantitative Rev'iew of the Initiator Event Tree 3.2 The quantita'tive review of the initiator event trees is discussed in the following subsections. 3.2.1 Review of Flooding Alarm Related Procedures 'The RB water level is detected by two RB water level, monitors installed on the RB floor. The flood alarps are activated by the monitors when.the water l.evel is more t.han 0 5 in. above the floor. The. sump al. arms will be activated . when water level reach'es the sump a.larm setpoints instal' led.at a level right ' ,below the level that activates the' RB flood alahns. S0mp alann se'nsors are instal' led at.various locations in the RB. ^ . The imnediate operator action specified in the Alarm Response Procedure (ARP56,71) is to initiate the Suppr'ession Pool Leakage Return System. The re-quired subsequent actions are: .f. Monitor RB water level to determine approxirate leak rate. Use'su'm'p' alarms to supplement the information obt'ained from.the above . instruments to ascertain the' approximat'e loc'ation of the' leak. I' '2. 'Nonitor parameters (such' as ;line pressure and flow rate) of the safety

  • systems as' a leak would affect 'the. system' parameter.s.

Isolate.the source of leakage' per proc'edure listed below in Stm) 3. 3. If required and plant condition: permi,t, dispatch an 6perator to the RB. ' floor to visually ' locate the so' rce of leakage. Isolate.using the ap-u .propriate system procedure listed below.. System. HRCI, Procedure No.SP23.202.01 'i L'eakage indication:.. Abnormal suction or: discha,rge piping pressure,. i . Excessive HPCI Loop'Livel Pump Flow or low dis- .cha'rge pressure. e 6 em 6 e i l -':^ ~ * - = ~ *.. -. r s e:.c -,__;,..;~.s ,..,em_.,._,.m,w_,,,.

4 - ' = - z.z....:. = .:_.~ ~ c::... =:_m ..... ' [. 3-8 Reactor building sump high water levels in vicin-ity of leak. Reactor building flooding alarm. Leakage isolation: If in standby, isolate the HPCI system by secur-ing the HPCI Loop Level Pump &nd then cl',osing CST Suction Valve (MOV-031). If the system is operating, secure per shutdown procedure and then isolate as ' described above. RCIC, Procedyre No.SP23.119.01 Leakage indicatic'n: Ahnormal suction or disch&rge piping pre'ssure Excessive HPCI Loop Livel Pump. Rqactor building sump high water levels. Reactor. building flooding alarm. Leakage isolation: If in standby, isolate.the RCIC system by secur-ing the'RCIC' Loop Lev 51 Pump 'and then elos'ing CSTSuction. Valve (Rby-031).. If the system is operating, secure per shut'down 'procedur:eandthenisolat'ehsdesc'ribedabove. RHR, Procedure.No.SP23.121.01 Leakage indication: Heat exchanger servi'ce water side temperature in' onsistencies. c Abno.rmal RHR system flow for mode.of operation. Ab' normal RHR system' pressures for. mode of oper-

ation, Reactor water level inconsistencies for mode of operation.

Sump high. level alar,ms. Reactor building flooding alarm. Leakage isolation: Isolate the leakage by shutting down the affected loop in accordance with.the appr.opriate procedure G 9 -e -m-. ,,+e-e-a r, m, w e..+.,.,- e,.,-ean.-,--e .=e, , -, e -e r w. ,,.-,.,,,,-,,,-.-.----e--

'~ 3-9 for the mode in which it was operating and then systematically shutting valves to isolate areas of the system fo.und above to be possible sources of 'l eakage. The above isolation procedure may. require inter- ~ mittent oper'ation of the leakage return system to. 'obse'rve the effects on water buildup When the 1.eakage has been isolated return the un-affected portions ~(as required) to service.- BNL has found that SNPS alarm response, procedures specify. general guidelines for monitdring. system parameters for determining the leakage loca-tiort and for initiating the Jeak', age isolation. However, the proce'dures faji to include specific requirements 'for operators to systematically c. heck the operati.on' parameter.s of relevant syst, ems. 'Since there are'm'any' system para. meter indicators in the control room,, the operators may possibly~ fail to ob-serve the. indication of an abnomal' system parameter. When the abnomal condition is severe enough to actu' ate the alam o'f a particular sy' stem parameter,,the correspondlng -Alam : Response,Pi ocedure will ~ then be fo,llowed by. operators. However, BNL, has reviewed.the relevant.Alam'., t 2-Response Procedures for abnomal system parameters, and found that these procedures do.not contain steps that should be followed under 'RB flood con-ditions.' Th'ese procedures provide guidel'ines for con'ditions other than RB.',. flood, such as water source abnormal or iso'laYion valves abnormal, etc. The '. operator risponses to tile flood could be ' delayed or confused when these Alarm . Rssponse. Procedures are followed. 3.2.2 Requantification -il. .I t , The revised initiato,r frequencies are. applied for evaluating the. sequence frequencies of the initiator event tree. In addition to the critical flood-depth of 3'-10" used by SNPS, BNL also evaluated the sequ6nce frequencies cor-respondilig to flood depth of l'-10" and l'-3". This is because, as indicated in Table 3.2.1, flood heights of I'-10" and l'-3" will disable several vital. 4 egum 9

3 systems such as HPCI and RCIC. The times for the flood to reach 3'-10", l'-10", and l'-3" depth were calculated based on the leakage flow rates de-termined in SNPS PRA. The calculated times are shown in Table 3.2.2. The HEP, values used by SNPS.are identical to the nominal HEP values provided in'the Probabilistic Risk Analysis Procedure Guide (D (see Figure 3.2.1 and Table 3.2.3). BNL fee'Is that the HEP could be higher than' the nominal HEP values' because.'the' flooding alarm related procedures fail to '. pro' vide specific guidelines to. identify and*to isolate the flood source (se'e Section 3.2.1). ' l The HEPs under the multiple alarm and the singl,e,alann conditions are listed in Tables 3.2.4 and 3.2.5. 3.3 BNL Review of Functional Event Tree This section is, divided.into three subsections. Section 3.3.1 provides a. qualitative ' review.of, the Shoreham'In' ernal Flood event tree analysis and Sec ~ t tion 3.3.2 presents the BNL revised time phased ' event, trees..Section 3.3.3 describes the res,ults obtained from th'e quantification of the BNL event trees. .3.3.1 Qualitative Review. Iri general, BNL'is of thef opinion that the methodolo.gy[used in'the ~ Shoreham Internal Flood Analysis'is consistent with that of the state-of-the-art and the approach is reasonable. The analys.is for the inte'r-nal,' flo'6d postulated much severe scenarios'than those 'of the Shoreham FSAR., The' Shoreham Iriternal Flood functio'na1, eye.nt tree' a'nalysisis ' based pi edominantly on the event trees developed for the internal. ' vent initiators, e namely, turbine trip,,MSIV closure and mandal shutd'own. These internal flood functional ev,ept tre,es only model flood scenarios where. the flood water height' at! Elevation 8' exceeds 3'-10". While it ap' pear's that thh Shoreham. functional ., everit trees do provide a representative modeling of the plant response, it is not well substantiated that floods.that are arrested' before reaching 3'-10". .will result in negligible core vulnerable frequency. 5

  • e O

TJ 3-11 Table 3.3.1 enumerates the vital equipment that has been identified in the Shoreham analysis. The components that are located at the lowest elevation are presented first. It can be seen that at the l' level, both the RCIC and HPCI vacuum pumps and condensate pumps are expected to be disabled. However, it isNiudged that their failures.do not lead to the failure,of the respective high pressure systems. Sinfilar argu'ments apply to the loop level pumps of' the low pressure core spray, HPCI and the RCIC systems as w' ll. At approximately e 2',,' instrumentation for both high p,re'ssure injection systems are submerged and hence.'resulting in failure of both.sy' stems. At; 3'-10" instrume.n'tation for 'both LPCS and.RHR is submerged.. lea' ding to the failure of t. hose low pressure systems. In the Shoreham analysis the critical height of 3'-10" is selected. However, since both HPCI and RCIC ha've fail,ed at about the 2' level, these ibut' an .. scenarios with tirmination of the flood prior to 3'-10" may not. contr e insignif.icant ' amount to 'the core vulnerable frequency. In the BN'L' revised \\ event trees, a time-phased approach is used to include the contribution from flooding below the 3'-10" igvel. ' Another area of concern ' stems from the treatment of' propagation of f ailures iri the Shoreham anal.ysiso, As noted' in Table 3.3.1, at 'the l level,, 4-480V ' pumps are expect 9d to e.xpe.rience, electrical shor.ts. The Shoreham an-alysis did noc investigat.e.any. cascading failure which may.resu,lt ffom the ' electrical shorts.. BNL reviewed the electrical drawings and elementary drawings for so.me af the systems. It appears that for. each pump there'is only on'e blectrical breaker which separ'tes it ' rom the rest of the loads in the a f same motor control center (MCC).. Random failure of this breaker to open could ~ result in the propagation of the short circuit fault upstream to the MCC, other MCCs and the load ce'nter. BNL's review of the electrical diagrams indicates that failure of the breaker to open will result in tr,ipping the t breaker at the load center. D.iscus'sions with Shoreham engineers suggested. that ther,e may possibly be an additional breaker per pump that is in. series with the first breaker. 'However, this was not confinned by BNL. In the BNL revised event trees, only one breaker is assumed and its failure ih modeled, explicitly. BNL 'did not review.the spraying effects due to water cascades from higher elevations. -,o

LI. I e, 3-12 3'.3.2 BNL Time' Phase Event Tree Thi determination of the time periods which are critical to the con-sideration of the progression of the flood is based on the vital ' equipment . location list (Table 3.3.1). Three heights were selected for,the BNL anal-ysis: at the l'-3" level, at the l'-10" level, and at the 3'-10" level. If, the flood is terminated prior to reaching the l-3" level, no impact is as-sumed for any equipment and the plant will'be shutdown, t'his is, Phase 1. How-ever,. i.f,the flood wa' er. exceeds the l'-3" levil but is termin,ated before the. t l'-10" level,'th's is Phase II. Phase. III en. tails the fa.ilures 'of bqth HPCI and RCIC system as well as the loss of power to the MG set recirculation pump.' fluld coupler before arresting the flood below the 3'-10" level. Any flood level which e.xceeds the 3'-10" level, it is treated in Phase IV. The event trees 'of these four, phases are. presented in Figures 3.3.1 through'3.3.4. Given that the flood is terminated in Phase I, BNL assumed that the rdactor has a high probability (0.9) that-it 'will' be manually shu't - down. Ten percent, of, the 1!ime... it may result. in a MSIV closure event. These two branches of the Phase J event trees are tra'nsfe,rred to the respective' j. l internal event tree, Figure 3.3.1.. Figure 3.3.2' depicts' the Phase II ' functional dient' tree'.which considers the various acc.i' dent mit.igation systems. Moreover, owing to ths fact that a number of the 480V pumps will be flooded,.the possibility of a breaker failure to isolate.i;he fa' ult is also evaluated.' It is assumed,t, hat the breaker' failure to op'en probability is 1x10-3 and there ar'e.a total of.five pumps in Division 1 and two pumps'in Division II that will be short circuited. probability of 0.5 is also assumed that failure of a load center in a division would lead to failure of other equipment. connected to that division. In the '. i ! event of;a MSIV ciosure, the feedwater system:is considered to be} unavailable. The probab,ility that the reactor will bd manually shutdown is also assumed to be 0.9 for the' maintenance induced flood events. Figure 3.3.3 illustrates the functional event trhe used to describe' the Phase III events. The major difference between this event tree and the Phase e

o m 3-13 II tree is the high pressure systems. In the Phase III events, both the RCIC and the HPCI systems are unavailable due to the failure of respective instrumentation. The' probability that the reactor will be manually shutdown is assumed to be 0.5 for the maintenance induced flood events. The Phase IV event tree '.is presented in Figure 3.3.4. This tree is drastically different from the other ones in that it only' considers the feedwater system, the depressurization function and ths PCS. Al.l the other. systems are disabled dpe to flooding. The likelihood that the reactor will be ~. ~ manually shutdow'n' is the same as in Phase III for maintenance-induced ficods'. 3.3.3. Quantitative Analysis Based on the development 'of the revised ficod initiator frequency', the BNL time-phased event tree and the mgdified human response to arrest flood,

  • . quantitative,results are obtained.

In the BNL an,alysis, ther,e are 1.7 different flood precurs' ors. Similar'to the Shoreham classif.ication', ' he first t five precursors are online mainten,an'ce related; the, remaining. twelve of them ,are rupture related. A detailed discussion on the BNL. flood precursors.is' given in Sectiori 3.1. Owing to the ways that these. flood precursors; ahe calcblated,'.the. ini-tiator event t'rses' have* been. modified tio ~inc'lude ' niy th'res functioris:' the', o flood alann annunication, I; operator action to isolate flood, A;' and reactor status. The entry con.dition to the different ti.me phase. event trees is deter. ~ mined'by the A function (see Sectio'n 3.2 for details). Each of the 17' flood pr'ecursors were'.evaIuated with 'the initiator event tree and the four' time phase event trees. The una'vailability values for the yarious event trees are the same as those,used in the Shoreham, analysis except as noted in the.last section. >t t .( ? i Whe'n the time phase event tre'es were quantified for the 17 flood pre-cursors, the results are the conditional fre.quency of core vulnerable given the particular flood precursor. These frequencies are summarized i'n Table ~ . ~ - -.. e w - + v-e r n

, - I..: ~. _ .: n 2-1.2 .n,_,,_,_-- .u 3-14 3.3.2. The seventeen precurso'rs are listed as rows while the four phases are shown as columns. Within' each precursor, contributions from manual' shutdown, MSIV closure or turbine trip are also shown. For instance, the conditional frequency of care,vulner'able with operator arresting the flood prior to 3'-10" but after }'-10" ' Phase.III, for TFL1 is '2.0(-5) given the reactor is' manually shutdown. However, if instead of a m'anual shutdown, the plant, experiences a MSIV. closures then the conditional frequen'cy is 8.5( 4) As expected, the conditional. frequency consistently increases as the flood progresses to higher elevat. ions. In oth'er words, the conditional ' frequency of Phase LV is always larger than any of the other phases., Another noteworthy - observation is.the unusually large conditional frequency of core vulnerable for the LPCI system' induced flood, i.e., TFL4 and TFL8. The TFL9 and TFL5 values.are also large since they disabled the LPCI systems as well. The core,vulberable frequer1cy given the BNL reyf' sed flood precursors,' initiator event trees and time phase event ' trees is 'shown"in Table 3.3.3. In this table, the 17 precursors are depicted on the left with the 4 phases de-picted as columns. Each precursor also identifies the contributions from t'he. ~ vario'us. plant states. C' ore vulnerable f'requency contributions from Phase,I c 'and.J I :a r'.e very. spal'1,' i.n..,the order, ' (10-9 Contrib'ut{ons, from, Phase'III ', o are not insignificant.but not substantial,'approximately 10-6 Seventy per-cent of the total core vulnerable frequency (70% of 2.0(-5)) is attributable. ~ to LPCI system maintenance or r0ptsre induced flopd. The mai,ntenanc,e, con-tribution to flood is about.37% w'hile the' balance is due to rupture. 'It appears al'so that failure to properly model the fault propagation of the short circuits through the breakers does not have a sign'ificant effect'on core vulnerable frequency. 4; I 3.4 Uncertainty Estimates This section presents a limited uncertainty assessment on 'the BNL quantitative analysis for the core vulnerable frequency due to reactor. building flcoding. s

  • n.

D . - - ~ ~ - .~

. _. l. L. = .. _ ~ .7 3-15 A rigorous propagation of the uncertainties is outside the scope of the present review. The BNL approach for the uncertainty evaluation consisted of the following general steps. 1. The uncertainties. in the human errors as well as the split ratio be-tween'the manual shutdown and the MSIV closure event were quantified by fitting lognormal distributions to evaluate uncert.ainty measures (meari and var,iance). An error factor of 10 was applied to human.er-rors and the spli,t ratio.' ~

2.
  • Human erro'rs of t' e. following operator action's were included for the.

h uncertainty evaluation: Operator maintains isolation valves in closed position.during the online maintenance (Event E, see Section 2.3). Operator diagnoses and re'sponds to, isolate th'. flood (Event. A, see e Section 2.3). l . Operator depressurizes. the Reactor Pressure Vessel (Event X, F.igures 3.3.2-3.3 4)'. j 3. The u'ncertaintiesiin the core:vulne'rab1'e frequency werel evaluated us- ~ ~ 'ing the mafor accident sequence ~s. and thi distributions ' ass'ess'ed 'in Step 1. '.The SAMPLE code was,used for the estimaton of. uncertainties. The mean, c.

median, 5%.and 95% probability intervals fo.r the core vulnerable frequency are shown as follows'.

.Mean 1.91E-5 = 1.90E-6 Median = 5% Conf.idence 2.19E-7 = f t 95% Confidence,= 7.51E.-5, e e .ry ,..,,,m.

-2.;_. b_.. _ & ,x 2-3-16 1 UP A MAINT'. R v AD' 4 SHUTDOWN p 1 2 DOWN * + +. / 'jh g j (t_p) ' MAINT. R ONI.!N I' X ~ BREAKER. / OUT MAINT. .f R ON-X 3' 'LINE BREAKER u . '.., IN A: Comp. Failure Rate. x x .p:. Comp. Repair, Rate .y A: Transient (MSIV clos.) Rate' D ' P: Prob. of not racking out Breaker . FI'.000 INITIAT l: Rate of inadv. operation of' 5 o ED switch A0T: Allowable Outage Time' i i i Figure 3.1.1 State Transition Diagram, for Component-Maintenance Induced Floods.. e e O .~, .. ~.. -.

._._ r. 2 s.... - - 2.

r.. _. -

. _. s - 3-17 I 1 i UP l A. R ,7 RUPTURE VULNER-1 2 ABLE s s s \\ A A N S T M a \\ \\ .g Y S T M FLOOD., ptg0D. FL60D A '3 'MSIV CL ' 4

l..

5 1 l TURB. T MANUAL SHUTDOWN .t ..t Figure 3.1.2 State Transition Diagram for Rupture-Induced Floods.' ~ .ep s ..-.g, -a.=+e,..gz men==**=_.**+ e

-e=*

--e megnam e-u. w- .,y---.y,-,--,ry 4-- r--= =:- - e,-v--

3-18 1, __._7 s - s \\ s s s, \\ . s 10 \\. \\ \\ g \\- s Upp'er Sound 10 c.. .q~~~. .\\ ~ .\\ ticminal. . 1'0" Value + \\. \\ ,..g .u Lowe.r 5ound. lj NI N N . ; s. N s. 10-2 1 l'0 100 1000 Minutes. Figure 3.2.1 P.ro'blem-solving human er'ror probability vs time ' screening values. 1* .a lt .s .t_ 8 e

.i i.! Phase'I ! g t t. l l i i i . i, i l 'P r j i i e I t-i a R. Stat.' l [ .t t ? i t .7

- i l

i t f i I l.-. s i '1 c r l . + i } l 'I + l - l l l .i 1 e j. l-I' l ? 8 3 1 6 g ) g i i 3, MANUAL. i Transfer to E/T i 1 i I i t l 3 -l 1! l I i t

  • l 1

i. i j i s - t .j t.. t l i i -l I l l-l i l 3 i 8 i 8 j l Li I i is l i j t I .I w [- } 8 e i I e i l i l j.-. i i I l l i MSIV i-

Tiansfer to t/T a

e { -j . i I' - t I l i -, 'l l i e i i i l i l l l i '}. i 30 i I i ..i e. i I, I i i i .i I. . I. t. l [ a

t..

( j-i .i j t, 4 l l Phase I of Inte' rn'l Flood; Functional Event. Tree i I s.. s Figure 3.3.1 a g:

i..

i

3. -

t i, i i .I e I i 3 t i I i i-t. 1 i i I: i l l I l 8 i 1 I i l i i I ? 1 I i l i

/ l l ~ I.. R !C M P- ~ Q Divl Div U U.', X V y.y W Z!'s 3 R HI CS CI cond R l ,P - j -l

Stat.

BKR *BKlt 2 6 ,I I I-l I i.. l 1 i .l j' I I l, t j j - l.- 1 I.. I l. l e i. t i l e i, . e t l. l l - 1 t l i I i i t l c i w i l I. mc f, a I S. Q ( s _ ( l i i i l l. -l. i i l 3 l- - I, 1 'l i 1. .i ~ s s l ~ Figure 3.3.2 Phase.II of lInteFnal ~ ]. ~ ~ ,l~ Flood Functionali lr ~ ~. Event Tree i ) i . M. 1 I :^ ~' 9 l' i .j l. n I 4I l l f~-- j

..g 1; l 1 l. r I i.. 4. R C' M P _.'Q Divl Div2 U ; U.X Y 'V V ! W ' Z.W i S ta t. BKR BKR R H.- CS CI cond R. P j .i i + + i l L 4 i I 1 i. i-i i i -l l.. j i i I l I ...i i t' I i

  • include prob.

l l of losing the t a.. 6 1 division. i i i I l e i. i, 1 i l I-i l l l I i. e.- [ l - 1 i + i e. g i i I. I i i 1,- I li i i r: t i ~ i. l -l i i It 1 i it i i l w i 1 i .i l 5' { i I. l

. l.l j,

.i q l i i

)

I p: . \\. i i I l l t i i F t [ i p l g y .i. l

t.,.

? i i e -l g 5 l j l i i I '.i -l-l l l'l l i i. e i. l l l 2 1 i i 6 -i. 8- ., Figure 3.3.3 '~ j-I PhaseIIIofjInternal Flood Functional. Event g I P Tree 3. J 8* .l l l ; e . i i.. i ~ j {..- ~l f g~ ~ l h i i r s-I i s j j-! l l.' i

'i -..) l i i -l l* e ~ i I I l-ll. 1 's t ..I-I .P 1 X Vicondl Z. W C' M PC '- i. l t li: I l - i i '. i l l l 4 i i

J i-1 g

l l. [ ii 1 I i l: y-l

il N

t I ~ \\ i

i l

I e l l ., i i ,1 a 3 i. i e i l s l l e 1 i t i. t I l I l 1 0 i I ( l'igure 3. 3.4 Ph'ase IV of Interria' Flood Functional Event Tree l ) e = i c I 0 o + .. s\\ l r l' 6

(,- 3-23 Table 3.1.1 LER Data for BWR Standby Pumps for the Period of January 1972 Through September 1980 - Does Not Standby Standby Leakage Does Not loss of . Continue Pumps Demands Hours ~ Rupture - Start - Function To Run Motor , Driven 20,321 10,453,806 9 8 8 9 Turbine.- 23 34 '.-.' 25 Driven 2,860 1;439.491 Table 3.1.2 Frequency of' Maintenance - Induced Flood Precursors ' System Initiator Event Trees . Probability per Year. TFL1'P'.D 1.05x10-4 1. RCIC - TFLI P.Eo 2.10x10-5 2.10x10-5 TFL1 P.E' '. L TFL2 P.D 1.05x10-4 2.' HPCI TFL2 P.Eoj l 2.10x10-5,. ,2.1Dx10~.g TFL2 P'.E. t 1.89x16-5 3. Core' Spray TFL3 P.D. (2 motor.drivenpumps) TFL3 P.E 1.87x10-6 o. 3.78x10-5

4.. LPQI '

TFL4 P.D ' '(4 motor' driven) T'FL4. P.E 3.74x10 6 o 5. Service Water TFL5 P.D 1.89x10-5 TFL4 P.Eo 1.88x10-6 (RHR or RB(LW HX)

.. 2 motor driven pumps

} 3 0 e em ,u h* ....e .e ..s .,gep. ..e e. .w. ,p ,,.,,m.

  • 3.

m.m 3-24 Table 3.1.3 Flood Precursor Frequency. 4 Pipe Valves Pump Total AR Ps PT Pg TFL'6 1.2(-9) 6.5(-9). 0 7.7(-9)

1. 6 (-5 )'. 1.7(-5) 1.5(-5)

TFL7 2.0(-9) 1.3(-8) 'O 1.5('-8) 3.1(-5) 3.4(-5) 2.9(-5) - ' TFL8 3.7(-9). 2.9(-8) 0 3.2(-8) 6.5(-5) 7.3(-5) 6.2(-5) TFL9 1.1(-8).2.3(-8) 6.0(-10) 1.3(-8) 2.6(-5) 2.'9(-5) 2.5(-5) TFL10-2.4 (-9') 1.3(-9)~ 0 3.7(-9). 7.5(-6) 8.4 -6) 7.2(-6) ( TFL11 1.1(-9) 9.1(-9) 15(-10) 1.0(-8) 2.1(-5) 2;4(-5,) ~'.2.0(-$) .TFL12 1.4 (-9) 3.9{-9) 1.5(-10).5.5(-9) 3.7(-6) 4.0(-6) 3.6(-6) TFL13 7.3(-6) 8.0(-6)- 7.1(-6) TFL14 '1.9(-9) 5.2(-9) 3.0(-10) 7,4 (-9) 5.0(-6) 5.6(-6) 4.'8(-8) .'TFL15', 1.0(-5). 1.1(-5)- 9.6(-6)

  • TFL16 1.9(-9)',,5.2(-9) 6.0('-10) 7.7(-9)

' 5.2(-6) 5.8(-6)- 5.0(-6) 1.0(-5)

1. 2,(-5 )
1. 0(-5) ~ -

TFL17 e s / .t l 4 8 l .-l

u

_._._.. m ..r,1 c 3-25 Table 3.2.1 MAJOR ELEVATION 8 EQUIPMENT LIST POSTULATED EQUIP. EQUIPMENT DESCRIPTICN PART NO. DISABLED TWE HEIGHT' FW5 ,Ficor Drain Sump Pumps ' ' 1G11*P-035A-0

  • l'-0" l

1G11*P-036A-F Dry F1'co' Orain Tank Pumps 1G11*P-151A,3 l'-0" r Radwaste Equip. Drain Sump & Pump to Porous Sump. 1G11*P-224A,3 '1'-1"

    • HPCI Pump,

1E41*P.015' HPCI Vacuum Pump 1E41*P-075 l'-0"' HPCI C,cn. Pump. 1E51*P-076 l'-0" ' **' RCIC Pu::;p 1E51*P-015 l ',0 " ' RCIC Vacuum Pump 1E51*P-076 RCIC Cen. Pump' 1E51*P-077 l'-0"

    • RRR Pump Motors IE1I*P-014A-0 5'-4"
    • Leakage Return Pump G11 *P-270.

. 3'-9", Core Spray. Leop Level Pumps ', 1E21*P-049A,3 l'-3" Drywell Equip. Drain Tank Pumps 1G11*P-0332A,B l'-2" RCIC Loop Level Pump. 1E51*P-051 l'-4" HPCI Loop Level' Pump '- 1E41*P-b50 2'-3" l TURBINES l HPCI Turbine 1E41*-TV-002 6'-0" t I

    • RCIC Turbine 1E41*-TU-005

- 4'-0"

  • e 9

,.--.eew-a -wy,y-+e.,---y%..,-,,,e

_. ~ .= .... r. :.=.: ... :.~:. := _.. = - 3-26 Table 3.2.1 (Continued) MAJOR ELEVATION 8 EQUIPMENT LIST POSTULATED EQUIP. EQUIPMENT DESCRIPTION PART NO. ' OISABLED HEIGHT' ~ TYP' 30T0a CONTROL CENTERS". Su=p Pumps and Cooling Water Pumps to Recirc. 1R24-1101 1*-6" ~ Pump MG-Set. Fluid Coupler 1R24-1201 l ' -6 ".' ~ TiNXS' Floor Drain Sump Tank 1G11*TX-050A,B 1G11*TX-056A'-C Orwell Floor Orain Receiver 1G11*TX-057 Salt Water Dr'ain Tarik 1G11*TX-190 ~ .,.:.i, s. Orwell Equip. Orain Receiver 1G11*TX-049. HEAT EXCHANGER HPCI' Barometric Con. Vacuum Tank 1E41*E-036 RCIC Barometric Can'. Tank 1E51*E-038 IE114*E-024A,8, i ' RHR Heat Exchanger,, 3 RSCLC2 Heat Exchangers IP42*-011A,3 Orw ell Equip. Orain Cooler 1G11*E-094 I em ..---..-~~-,.._n

"~ ~ ~ '~' ~~~ ~ ~ ~ " ' ~ ' -: -.... = ~ L . ~..,.-..:.: ;:... u -_: 3-27 Table 3.2.1. (Continued) MhJOR ELEVATION 8 E'Q I? MENT LIST POSTULATED E;UIP. EQUIPMENT DESCRIPTION PART NO. DISABLED TY,PE HEIGHT" EL'EC. PANELS

    • RCIC Instr. Rack 1H21'PNL-017 2'-0"
    • RCIC Instr. Rack
  • 1H21*PNL-037 2'-0" 3'-10" *.
    • Core Spray Rack 1H21*PNL-01'
    • Core Spray Rack'

' 1H21*PNL-01'9 3'-i0" c ' ELEC. 'P:NELS " RHR Inst. Rack'.'A 1H21*PNL-O'18 3'-10"- 3',-10" " RHR Inst. Rack S. 1H21*PNL-021 . k l'-10" 1H21 PNL-036

    • HPCI Inst. Rack A
    • HPCI Inst. Rack.B'..

1H21*PNL-14 1'-10" Equipmen't required for operation of the identified system. Heignts are taken frem a physical survey measurement from the bottom of the, component to ficar level. l --- Non-electrica.1 ccmponent i . em ~ - ~ ' .---qw .yyr,m.= --v-vw-w

. A- - ~ ..:...==.:.. l 3-28 Table 3.2.2 Times to Flood Depth of 3'-10", l'-10", and l'-3" in Reactor Building Water Time (min.) to Flood Depth of System ~ Source Leakage Location 3'-10" l'-10" l'-3" ~ HPCI S.P.- pumpsuction(max.) 17 7.9 5.4 . S. P. pump suction (large 34 15.8 10.8 pump suction (max.) 13 6.4 4.4 CST CST pumpsuction(large) 27 12.8 8.7 ~ pump discharge 37

  • 17.5 11.9 RCIC S. P.

,pumpsuction(max.) 110.0 50.8 34.6 S.P. pump ' suction (large) 220.0 101.6 69.3 CST pump suction (max.) 76.0 36.3 24.8 CST pump suction (large) 152.0 72.6 49.5 LPCI S. P.

  • pump suction'(max.)

9.4 4.5 3.1 S.P. pumpsuction(large) 19.0 9.0 6.1 15 7.3 5.0-pump discharge CS-S.P. pumpsuction(max.). .12 , 5.9 4.0 S. P. pump suction (large) 24 11.8 8.1 pumpsuction(max.)) 27 12.8 13 6.4 4.4 CST CST ' pump, suction * (lirge 8.7 23 11.1 76. pump discharge - SW SW RHR heat exchanger. 20 9.5. 6.5 WFPS WFPS rupture of 8, pipe 40 , 19.1 13.0 s, . Note:

1. Large flow rates is.1/2 of ' aximum flow rates."

m ' 2. Flood tim'e,s were calulated based on a 41,600 gallons per foot depth ~. in the reactor building. ) 3.'S.P. =. Suppression Pool i CST. = Condensate Storage Tank 'SW = Servi'ce Water System WFPS = Watep { ire Rrotection System Tanks. .,6 g .l .....4....?_

.~ .....u.. _i. _.--._u. -.. _..__.._.__:... .._._._ m... 3-29 Table 3.2.3 Human Error Probability: Screening Values -Problem-solving 8 e, Time Nominal Value Errcr Facter- <1 min. 1 10 min. 5E-1 5 1E-1 10 20 min. 10 .30 min.- -1E-2 -10 60 min. 1E-3, 1500 min. TE-4 30-Procedural Errors- !!cminal Value-Errce Fac cr .'lE-3 ('rli t:h P.eco.very) 3 IEf2 -('.it thout P.ecovery) 3 =ia-b 6 g h G g g 8 e, e g O 3 g c l e D .m b - *==-- ,w.., .,e w_ -g-or,--y, p-+r 9..q,,y,. .--.-9 --.9__g w y. O

'- ^

- w.
. ~.::. _ _ -:

.h. _. :::

==.;-- -c .._ 2.w. m 3-30 Table 3.2.4 HEP (Event A) Single Alarm Condition Manual Shutdown (Nt' REG /CR-1278) l'-3" l'-10" 3'-10" _ m. TFLI. 10-3 10-3 2.0x10-4 TFL2 1 1 0.1 TFL3 1 1 0.1 - . TF L4 '. 1 .1 1 1 1 ~10.TFL5. ' TFL6 0.1 0.1 10-3 1 0.1 10'-2 TFL7 TFL8 1 1 0.1 TF.L9 1 1 10-2 10-3. TFL10 0.1 0.1 ~ 10-3,. '10-3 2x10'd TFL11 TFL12 1 1 0.1 . TFL13 d.1 O.1 10-3 1 1 0.1 TFL14 ~ TFL15 1 0.1 10-2 '.TFL16' 1 l' 1 TFLI.7 .1 1 0.1 } ' ' I

  • 8 s

a e - - - - -. ~. __-.m

~... m. ..~......--..: 3-31 Table 3.2.5 HEP (Event A), Multiple Alarm Condition -(Nominal Value, PRA Procedures Guide) l'-3" l'-10"' '3'-10" TFL1. 10-2 10-2 10-3 ~ TFL2 1 1 0.5 0.5 TFL3 1 1 1 1 . r TFL4 TFL$ '

1. '

~ 051 ' 'l ' 2 TFl.6 O.5 0.5 TFL7 1-0.5 - ' 0.1. TFL8-1 1 O.5 TFL9 1-1 b.I .0.5" 0.5 10-2. TFL10 10-2 10-3 .TFL11 10-2 1. 0.5 ~ TFL12. l 'iFLib. 0.5 O.5 '10-2 ,TFL14 1 1 0.5

TFL15 1

0.' 5 0.1 ' ' - al' 'TFL16 I 'TFL17 ~1 1 .0.5 t 0 0 + ' t e e e e 8 6 g E j j i 8 4 s e j 9 D 9 g 0 e e e l e e e

  • e**-

.*=r .-w.. A . w. w p.- y -r.- y ,m..f g_-gw--4 .._s y .m-g

s.,,..;...

. ~..... s f 3-32 Table 3.3.1 Vital-Equipment Locations at Elevation 8 l' 'HPCI vac. pump - '~ cond. pump RCIC vac. pump ' cond. pump l'-3" l'-3" CS loop level pump l'-4" RCIC. loop pump *.

  • l'-6" recir. pump M-G. set l'-10" HPCI instrumentation l'-10" 2'

.R'CIC instnamehtation h 2'-3 HPCI loop level pump 3'-10" RHR instrumentation 3 -10 i CS in.strumentation 9 O g D e e p

.?

4 e e e g g e o 6 e g

  • 0 e

G s a a b l =e ) I ...-~..-..? r.. - ~~ - - - ~ ~ - -= ~ ~= ~ ~ ~ ~. - - - - *. ~. - - - ?.

-1___- =r...- 3-33 Table 3.3.2 Conditional Frequency of Core Vulnerable (1 of 2) Phase I Phase II Phase III Phase IV TFL1 Manual 5.8(-7) 2.7 - 2.0 - 7.3 -3) MSIV 3.2(-6)

8. 7 -

8.5 I.2 -1.) TT 7.7(-7) 2.2 - 2.1 3.3(-2) TFL2 Manual ' 5.8 - 2'.2(-6) 2.0(-5 7.3(-3 ' MSIV 3.2 - 6.S(-5)

8. 5 (-4 1.2(-1

'TFL3-Manual 5.8(-7) 1.1(-6) 2.2(-5) 7.3(-3) ' MSIV 3.2(-6) 1.1(-5)

9. 5(-4 )

1.2(-1-) TFL4 Manual' 5.8(-7) 3.9(-4 )

5. 2(-4 ),

7.3(-3) MSIV 3.2(-6) 2.0(-2) 2.6(-2) 1.2(-1) TELS Manual 5.8(-7) . 3.9.(-4 )

5. 2(-4 )

7.3 (-3 ) MSIV 3.2(-6). ' 2.0(,2).

2. 6,(-2) 1.2(-1)',

TFL6-Ma.nual 5.8(-7)

2. 2-(~ 6 )

2.0( 5) 7.3(-3) MSIV..

3. 2'(-6),

6.8(-5) 8.5(,4) 1.2(-1) TT 7.7(.-7) 1.6(-5) 2.1(-4 )

  • 3.3(.-2) 5.8(-7) '!.

1.1(-6) -

2. 2(-5 ) '..

73(-3). TFL7 Manual MSIV 3.2(-6) 1.1(-5) 9.'5 (-4 ) - 1.2 -1) ' TT.:, . 7.7(-7). g .3. 2 (~-6), ., '. 2. 3 (-4 ). 3.3 -2)... TFL8 Manual 5.8 - 3.9(-4 5.2 -4 ) 7.3 - MSIV

3.2 -.

2.0(-2 2.6 -2) 1.2 -

  • TT

. 7.7(,-7 ). . 4.7(-3 6 2..3) '3.3(-2) ' '5. 2(-4 ) .7.3(5 ' TFL9 Manual

  • 5.8 -
3. 9 -4 )

MSIV 3.2 - .2.0 -2)

  • 2.6(-2 1.2(
  • ~

TT

  • 7.7(-7) 4.7(-3) 6.2(-3 3.3(-2)

TFL10 Manual ~ 5.8(-7) 1.1(-6) 2.2(-5)' 7.3(-3) 3.2(-6) 1.0(-5)

9. 5 (-4 )

1.2(-1) MSIV .TT yj 7 7(-7) 3.2(.6) ,3. 3 (-4 ). 3.3(-2) 2 TFL'11,Same as TFL1 TFL12 Same a's TFL6 TFL13 Same as TEL6 eem

~ ' ~ ... - s... _ l 3-34 Table 3.3.2 Conditional Frequency of Core Vulnerable 1 (2 of 2) Phase I Phase II' Phase III Phase IV TFL14 Manual

3. 2'(.-6 )

1.1(-6)- 2.2(-5) 7.3(-3) 5.8 -7) MSIV ( 1.1(-5)

9. 5 (-4 )

1.2(-1)" TT 7.7(-7) 3.2(-6) 2.3(-4 ) - 3.3(-2) TFL1,5 - Same as TFL14 TfL16 Same as TFL8 TFL17' Same as TFL8 8 e s 9 G 8 e e e 8 e g 9 e a n O h e t 0 8 1 e i i, e e g e e 4 e e l-i . s 5 6 S e 9 e .e S o 4.e e--- .+e..-* , -... =., - = - .-- 77 ~===****=*we.m..

r,..,ny

--v ,,e.- --m ~svv r- - -,,,~

... s .a. - t ~ 3-35 Table 3.3.3 Core Vulnerable Frequency (1of2) P-1. P-2 'P-3 P-4 TOTAL TFL1 Man. 7.3(-11) 0 1.7(-11) 1.2( 9). MSIV 4.5(-11) 0 8.2(-11) 1.3(-8) .1.2(-10) T, 9.9(-11) 1.5(-8) 1.4 (-8)' TFL2 ' Man. 0 0 9.1(-10). 2.1(-7) ' MSIV 0 .0 3.9(-8)' 3.4 (-6) - l. 0 F. 4.0(-8) 3.6(-6)- 3.7(-6) ' TFL3. M.an. 0 0 1.2 -10) ' 3.5(-8) MSIV 0.. 0 5,2 -9) 5.8'(-7) 0 V 5.3(-9) 6.1(-7) 6.2(-7). TFL4 Man. 0 0 0 1.5(-7). MSIV 0 0 O 2.5(-6) 'E 'T '2.7(-6) 2.7:(-6) 6 TFL5 Man.

  • 0 0'

4.9(-9) 6.9(-9) MSIV 'O 0 2.5(-7).. 1.1( -7~) I' ,T 2.5(-7) 1.1(-7) - 3.7(-7) ' TF'L61 Man' O 1.'6(-10) MSIV. 0 1.4 (.-8 ) 5.6(-8) U 1.4 (-8) 5.6(-8) .7.0(-8)' TFL7 Man. 0 1.4 (-9) MSIV 0 3.9(-10) 2.6(-8 ). - 8.3(-7) I 3.9(-10) 2.6(-8)

  • }8.3(-7)

'8.6(-7) TFL8 Man. 'O 2.3(-8)' D' 1.5(-8) MSIV 'O 0 1.6(-6). 4.3(-6) TT 0 0 2.3(-7) " 1. 2(-6 ) V U 1.8(-6) 5.5(-6) 7.3(-6) TFL9, Man., 0,.,., O . 6.5(-9) 1 2(.-9) g O ' 9.2(-7) ' 3.4 (-7 ) MSIV' '0 TT' 0 0 1.8(-8) 9.9(-8) 3 3 9.4 (-7 ) 4.4 (-7 ) 1.4 (-6)

  • Lessthan1.0'(-10).

.a.'-* __._..a ^ 3-36 Table 3.3.3 Core Vulnerable Frequency ~ (2 of 2) P-1 P-2 P-3 P-4 TOTAL .TFL10 Man. O MSIV 0 3.9 s9) 1.4 (-8 TT 0 9.4 -10) 3.6(-9 I'. 4.8(-9) 1.8(-8) 2.2(-8) ,0 .TR 1'l Man. - *- MSIV. 0 1.'6( -1.0-) 1.4 -8) TT 3.2.-9) '. 0 ' E 1.6(-10)' ' 1.7(-8)

1. 7 (,-8) '

1.1(-9) TFL12 Man, 0 0 MSIV 0 0 4.5(-9) 4.9(-7) 3 E, 4.5(-9), 4.9(p7) - 5.0(-7) . Man. 0-TFL13 0 ' 6.6(-9) 2.5(-8) 3.1(-8) MSIV 1.8(-9) TFL14 Man. 0 0. MSIV 0 0 7.3(-9) 6.9(-7) 5 7.3(-9) 6.9(-7) 7.0(-7) 4.6(-10) TFL15 Man; .0 0 . 8.4-( -9 )

i.. 2.6(-7).

.2.8(-7) 'MSIV V 8.4 (-9)- 2.6(-7) TR.16 Man. 0 0 0 1.8 -8)' 0 0 9.2 -7 MSIV O .TT' 0 0 0 1.9(-7 E F V-1.1(-6), 1.1(-6) TFL17 Man, 0 0 2.3-9) 3.8(.9 MSIV 0, 0 2.5 -7) 6.6(-7 TT 0 0 3.7 '8)- '2.0(-7 0 0

2. 9(-7) 8.6(-7) 1.2(-6)

./- l, j' I,

  • Lessthan1.0(-10).

t e . eo g O m -,--_,_.--,m e -,.. _,. ~ ' ~ * ~ ~, _ -.

_3..i. 7._ e ~ - ;. g_.. 4-1 4.0

SUMMARY

l-l BNL reviewed the internal flood analysis which is a part of the Shoreham PRA and found that assumptions, methodology, and results are reasonable. BNL re-e' valuated the flood precursor frequency us'ing ricent LER data and a more accurate methodology. This methodology avoids some of the'.conservatisms in the SNPS-PRA approach. A slight increase in the. initiator,fr'equency is calculated'because of th4 revis,ed data. Similarly,' based on the PSA Pro'cedure Guide, the HEP was reviewed and only l ~ 2 minimal changes were mad to the'Shoreham HEP values used in the' analysis. As, e f6r the functional' event trees, a time phase a'pproadh was adopted to better, model the progression of the flood events. l Results are. summarized in Table 4.1. This table'can be divided into three 'pa rt s. Part' A provides a comparison between.'the Shoreham Fesults and those' obtained'in the BNL review. The.BNL v'a' ue is about 5 times that of th.e - l Shoreham frequency, 2.0(-5) vs 3.,9(-6). The conti-libutions from t.he differ'ent ,- ' plant states are 'aiso presented. The major increase in the total core vulnerable frequency in the BNL analysis is attributable to th,e. increase in' flood precursor frequencies. Part B comp' ares only, the contributions fr'om the*. BNL Phase'IV result's'with'the'Shoreham values. 'It' can 'be' inf erred tihat' bf' ' neglecting the initial three phases, the core. vulnerable frequency will be underestimated by 3kl0'-6 or about.18L Part C shows the contributions of - core vulnerable frequency for different' plant states due to maintenance a'nd rupture,ind'uced f1'oods. In.the Shoreharii analysis 41% of the core vulnerable ' I-frequency is calculated to be caused by maintenance related floods while the-BNL analysis shows 37%. An uncerta'inty estimation h,as been carried out assuming,lo'gnormal distributions ~. An error factor o'f 10 was applied to the ~ operator errors and' the spli.t ratio for the manual shutdown and the MSIV clo,sure event following

  • 0 1

i l _; n _ .. ~

-n il . e, ' 4-2 the Reactor Building flooding. The results of the uncertainty assessraent-for - the core vulnerable frequency are as follows. 1.9E-5 tiean = - 1.9E-6 Median = 5% Confiden*e 2.2E-7 c = 95% Confidence,= 7.5E'-5 O t 8 $ e 0 e 8 8 e e 8 e e , e 8 9 3 e e e S O g g e C e e 6 4 g F e e 4 9 5 4 g e O 4 0 $ g 4 8 g e i l + 1 e I 9 9 8 4 9 g g 5 e e 8 4 e, g g 8

  • e e

e e e e t e-g e 9 e s e e 6 g 0 4 6 ,I ' l.1 i .I l t t a e 8 9 e e 0 O t e 5 se 4 D E

V. , o y.

U.

4-3 Table 4.1 Summary of Core Vulnerable Frequency e Shoreham BNL Part A Ma'nual ,8.5(-8) 4.8(-7) MSIV 3.0(-6) 1.8(-5) TT. 7.7(-7)' 2.0(-6) .3.9(-6) 2.0(-5) Total BNL (only Shoreham Phase IV 4 Part B Manual 8.5(-8).

4. 5(-7)'

MSIV 3.0.(-6) -1.5(-5) TT 7.7 (-7 ) '. 1.7(-6) Total 3.9(-6) 1.7(-5)' 4 Shoreham - BNL Part C' Manual Mainte' nance ,'.3.9(-8) 4.1(-7) Rupture - 1.6(-7) 7.0(-8) '.9(-6) MSIV Maintenance 1.5(-6) 6 Rupture 1.4(-6) 1.1(-5) '\\ t TT Mainten$nce 0 0 Rup,ture 6.7('7) 2.0(-6) Total Maintenance, 1.6(-6) 7.3(-6) Rupture 2.3(-6) 1.3(-5) .m .~ e- -w- ..--s ~ -.. .an-n~.

x.- - 1 d.h'.'*. -.w. cz-R-1 REFERENCES 1. PRA Shoreham Nuclear Power Plant, LILCO, June 24, 1983. 2. U.S. Nuclear Regulatory Commission, " Data Summaries of Licensee Event Re-ports'of Pump's at U.S. Commercial Nuclear Power Plants: January 1, 1977 to , April 30,1978,". NUREG/CR-1205, January 1980. . 3. W. H. Hubble,' C. Miller, " Data. Summaries.of. Licensee Event Reports of. Valves at U:S. Commercial Nuclear Power Pl'4nts,".NUREh/CR-1363, Volume 1, Ju'ne 1980.. ~ '4.. Swain, A. D., Guttmann, M. E.', " Handbook of Huma'n Reliabi,11ty Analysis With Emphasis on Nuclear Power Plant Applications," NUREG/CR-1278,1982. 5. Busch, S. H., " Pressure Vessel Reliability," Trans. ASME J., Pressure

  • . Technology,' January 1975.

'.S. Nuclear Regulatory Commission, " Data summaries of Li,c.ensee Event Re-U 6. ports of. Pumps at U.S. Commercial' Nuclear Power Plants: Janua'ry,1,1977 to ' September 30, 1980," NUREG/CR-1205, September l'981. U 7.- "Probab'ilistic Safety Aqalysis Procedures Guide," NUREG/CR-2815, September '1983. 8. " Nuclear Power Experience," Div,ision of Petroleum Inforination Corporation, ~ Denver, Colorado, August 1981. 8* e d t l i g 9 e ee 4

m M' s, s. s MAY 9 gM Docket No. 50-322 MEMORANDUM FOR: Albert Schwencer, Chief Licensing Branch #2 Division of Licensing FROH: Ash'ok Thadani, Chief Reliabiitv and Risk Assessment Branch Division "of Safety Technology

SUBJECT:

SHOREHAM FLOODING

Reference:

(1) Memorandum dated Ma.r.ch 30, 1984, from A. Thadani to A. Schwencer, "Shoreham Flooding." In Reference 1 we transmitted our findings on' the Shoreham flooding These findings were based on the Brookhaven National Laboratory issue. (BNL) evaluation provided to us in a draft report. The purpose of this ,~ Our memorandum is to transmit to you the final BNL report on this issue. conclusions reported to you in Reference 1 still remain valid. Ashok Thadani, Chief Reliability and Risk Assessment Branch Division of Safety Technology

Enclosure:

BNL Final Report on Shoreham Flooding cc: H. Denton Olstribution R. Mattson Central file D. Eisenhut RRAB Rdg T. Speis Echow E. Chellfah cChow'CHRON

  • l R. Caiuso Busiik 9

i M. Campagnone A. Busiik W d / /A hadani f R. Frahm b , y,r ~ " Vi W"/

  • 1

...R..R..A..B : DS.T..*..... ...R.RAB : DST RRA8:DSTP/ sum = >...E.C.h.9.w....e.1...... AB,u,s,1,,i,,k,N,,,.. , A,T,h,ad,a n,,,,,,,, c,,,c, > w 1 E .. 5L.9....a.4.... 54.. 8..L84.. ..si.. .....,La4. usce i,.i2 s o*n > V OFflCIAL RECORD COPY ......~...a.

r 1 - e. ~ #' . ^ ' ENCLOSURE 1 - l . LETTER REPORT ON THE ~ REVIEW 0F THE SEQUENCES FOLLOWING A RELEASE *0F EXCESSIVE WATER IN ELEVATIO.N 8'0F THE., REACTOR BUILDING IN THE. SHOREHAM NUCLEAR POWER STATION. K. Shiu Y. Sun E. Anavim I A. Papazoglou Risk E' valuation Group Department of Nuclear. Energy Brookhaven National. Laboratory. Upton, New York.11973 .l.. s. .~ t April 19d4, ' Prepared for ~ U.S.' Ndelear Regulatory Commission - ' Washington, D.C. 20_555 Contract No.DE-AC02-76CH00016 8 0

  • 6 U'W )c-v- Y (

1 LJ /f ?'~

c c - / l ABSTRACT i The core vulnerable risk resulted from Reactor Building flooding events is addressed as a part of the SNPS PRA.(1) The analysis was reviewed and re. evaluated at.BNL and the results are presented in this report. The BNL ' review includes both qualitative and quantitative analyses of flooding initiato,rs, operator errors, and. accident, sequences which result in a-vulnerable' core state. An esticdte of the uncertainty for, the core vulnerable risk is also. included. e e / o e 9 O g g e e I g o, g s e s e s e e e g ' e e e iii g I ......s ...4 .-.y..-

r . ~.. _.. -.. ~ 'C d TABLE OF CONTENTS Page A BS TR A C T................................................................. i i i L I S T OF F I G UR E S.....................................................,..... v i L I S T OF TA B L E S........................................................... v i i i 1.0 I NT R OD U C T I O N........................................................ 1 -1 2.0 SNPS lETH000 LOG Y AND ANAL YS IS...................................... 2-1 2.1 O ve r vi ew...............................'....................... 2-1

  • 2.2 SNPS-PRA Quantification of the Frequency of Fl' od, Initiators..

2-3 o 2.2.1,, Maintenance-Induced Fl ood Initiators...............'.... 2-3 2.2.2 Ru ptu r,e-Induced.Iri t ti a to rs............................. 224 2.,3 Initiator Event Trees....J.t.........................,......... 2-5. 3.'0 B L ACCIDENT REVIEW AND SEQUENCE' QUANTIFIC ATION........ k.':......... 3-l 3.1 Flood, Pre curso r Frequen cy........................... :......... 3-l',

  • 3.1.1 Maintenance-Induces Fl ood Ini tiato rs...................

3-1 3.1.2 Rupture-Induced Flood. Initiators...........:....... 3-4 3.2 BNL Quantitative Review of the Initiator Event Tree........... 3-7 3.2.1 Review of Flooding Alam Related Procedures............. 3-7 3.2.2 Requantification.................'.................~..... 3-9

3. 3, BNL R e vi ew o f Fu'n ctional Even t Tre e............................, 3-10
3. 3.1. ' Qual i tative Revi ew....... ;............................

3-10 3.3.2 BNL Time Ph a se E vent Tree....... /............'.......... 3-12 3.3.3 Qua nti ta tiv e An al ys i s................. ;........ J....... 3-13 3.4 U n c e rt a i n ty E s t i ma t e s...........'..'........... '...... '..'.....*.... 3 -14 4.0 S U M t% R Y...................................... '....................... 4 - l ' R E F E R E N C E S.................................................*.........'.... R - 1. g 0 6 e g g e a 8 e 4 ,a 8 6 9 5. i i e e s e e 9 y e -~

4 .-~ a LIST OF FIGURES Figure No. Title Page 2.1.1 System event tree for manual. shutdowns with greater than 3'-10" of water in the Reactor Building (Source = CST)......................................... 2-8 2.3.1 TFL1: Initiator event tree for postulated flooding sequences initiated during RCIC maintenance.......... 2-9 ~2.3.2 T.2 . Initiator event tree for postulated floodi.ng FLsequences ' initiated by an error during HPCI major. maintenance......../.................................. 2-10 2.3.3 TFL3: Initiator event tree for postulated flooding sequences initiated by an error during core spray - maj o r mai nt e na nce.................................... 2-11 2.3.4 TFL4 : Initiator event, tree fo'r postulated flooding sequences initiated by an error durin .maintehance..........................g.LPCI major ................. 2-12 2.3.5 TFL5:. Initiator event tree fo'r postulated flooding sequences initiated by an error during se,rvice water major maintenance (i.e., heat exchangers )............ 2-13 .2. 3.' 6 Initiator event tree for postulated flooding sequences initiated, by a HPCI discharge pipe break;........'... 2-14 2.3.7 Initiator event t'ree'for postulated flooding seque.nces-initiated by a 'CS discharge pipe break..... s.......... 2-15

2. 3.&

Initiator event trees for postulated flooding sequences. initiated by a LPCI' discharge pipe break.........i..- 2-16 2.3.9 Initiator event tree.for postulated flooding sequences ini.tiated by a service water li ne break............... 2-17 2.3.10 Initiator, event tree for ' postulated flooding sequences i ni ti ated by a, WFPS break... y..... i...... J ;......... 2-18 . Initiator e' vent. tree for postulated flooding. sequences - 2.3.11 initiated 'by a maximum RCIC suction line ' break....... 2-19 2.3.12 Initiator event tree for postulated flooding sequences in.itiated by a maximum HPCI suction line break....... 2-20 Initiator eve'nt. tree for postulated flooding sequences 2.3.13 initiated. by a large HPCI suction line break......... 2-21 2.3.14 Initiator event tree for postulated floodi'ng sequences initiated by a maximum core spray sucti.on line. break.. 2-22 2.3.15' Initiator event tree for p'ostulated flooding sequences initiated by a large core spray suction line failure. 2-23 ^ 2.3.16 Initiator event tree for postulated floo' ding sequences initiated by a maximum LPCI suction line break....... 2-24 2.3.17 Initiator; event. tree for postulated flooding sequences ! initiated by a large LPCI suction line break......... 2-25 2.3.18 Comparison of the HEPs associat'ed with operator i actions for singular events and coincident multiple events................................................'. 2-26 e 9 vi

e f f e LIST OF FIGURES (Cont.) Figure No. Title Page 3.1.1 State transition diagram for component-mainten-ance induced f1oods.................................. 3-16 3.1.2 State transition diagram for. rupture-induced floods... 3-17 3.2.1 Problem-solving human error probability vs. time screening va1ues..................................... 3-18 3.3.1 Phase I of internal f,lood functional event tree....... 3-19 3.3.2 Phase II of internal flood' functional event tree...... 3-20 3.3.3 Phase III of internal flo'od. functional event tree..... 3-21 3.3.4 Phase IV of internal flood func'ttonal event tree....... 3-22 ', 0 e 0 8 e 8 0 e g 4 e e 4 0 9 e 8 e e' e O g en W 8 e t e 0 W e 8 4 0 g e g e +44 4 S - ' g 9 w g g a 5 9 9 8 a e D 9 9 9 eg g 4 0 8 e e t k e 0 0 ?.' I .i D 5 e e 9 vit -- - - ~..~ -- -- -. ~.-.- ~..,, :. ~.. _

e.

a..

. o. LIST OF TABLES Table No. Title Page Summary of Potential' Water Sources and Types of 2.1.1 Initiators which may Lead to Release of Excessive Water i n the El evation 8 Compa rtment................. 2-27 2.1.2 Summary of Internal Flooding Initiator Types: Source, Pathway, Flowrates, and Time to Critical Flooding 0 e p t h............................,................... 2 - 28 2.1.3 Summary of System Event Tree Entry States by' I ni ti a t or Ty pe..... s.........'.......................... 2-29 2.2.1 LER Data for BWR Standby Pumps for the Period of Janua ry 1972 Through April 19784....................... 2-30 ' 2.2.2 Frequency'of Online Major Maintenance System in the Reacto r Buil di ng..................................... 2-30 2.,2.3 Sunnary of Failure Rates for Major Camponents Involving External Leak and External. Rupture.....'.... 2-31 2.2.4 Conditional ' Probability of Pip'e Break Size............. 2-31

2. 2'. C Initiat,ing Event ' Frequency Estimates' Involving Component Leak /Ru ptures..............................

2-3 2 2.2.6 Calculated Frequencies for Initiating Events Re-sulting f rom System Ruptures (SNPS-PRA).............. 2-33 2.3.1 The, Probability that Flood Remains Unisolated for X Minutes After Automatic Plant Action, e.g., Turbine Tri p o r MS IV Cl os u re................................. 2-34 3.1.1 LER Data for BWR Standby Pumps for the Period,of January 1972 through September 1980.................. 3-23 3.1.2 Frequency of Maintenance Induced Flood Precursors... 3-23, '3.1.3 Fl ood Precurso r Frequency.........'..........,.......... 3-24 Majo r El evati on 8 Equi pme nt Li st..'..................... 3-2 5 /3.2.1-

  • - '*: 43.2.2

Times to. Flood Depth of 3'-10'4 l'-10", and' 1!-3'h in ' Re a ct o r B u i l di ng............. '....................... '. 3-28 3.2.3 Human Error Probability: Screening Values............ 3-29 3.2.4 HEP (Event A). Single Alarm Condition Manual Shutdown '(NUREG/CR-1278)...........................;..........,. 3

  • 3.2.5 HEP (Event A), Multiple Alann Co'nditi'on (Nominal Value, PRA Procedures Guide).........................

3-31 3.3.1 ' Vital Equi pment Locations. at El evation 8.............'.. # 32

3. 3. 2 '

Conditional Frequency of Core Vulnerable...........,... 3-33 3.3.3 Co r'e Vul'n erabl e F req ue ncy............................. 3-3 5 4.1 Summa ry of Co re Vul ne rabl e Frequency.................. 4 -3 i ) i .il, e 0 O, e viii e omeeeamme m. een s +se eene m *gess-w **+em=-eene 7 - ; *me _ _ _ #a e..*N*****W***-* 8'8** N _ _ _ -*i

C.
  • o e

1.0 INTRODUCTION

At the Shoreham Nuclear Power Station (SNPS) the majority of safety-related equipment are located in the Reactor Building (RB). The Shoreham Reactor Building is a cylindrical, building surrounding the liARK II containment structure. Water leakage from equipment in the reactor building will drain to Elevation 8 (the lowest level of the RB) via openings and . stairwells since t'here is no structural separation between s.afety systems.. Floodi.ng of the Elevation 8 compariment may potentially disable al'1 the ECCS because they are locat'ed in the Elevation 8 compariment. The SNPS-PRA(l') has includ,e'd ' flooding as a' Sommon bode event which may disabl.e the $CCS equipment.,The SNPS PRA assumes'that a critfcal flooding depth of 3'-10" from' tne RB floor will disable all the ECCS equi.pment., Operator diagnosis and isolat' ion of the floo' ding before it reaches 3'-10" - depth is considered in SNPS-PRA.' Because of the pote..tially si.gnificant impact, the ~SNP5's evaluation of, the core melt, risk due to RB flood,ing warrants a specialL* review. A -fi el d ' tri p to the Shoreham plant has been made by ENL perso,nnel for o'btaining' detailed ' t informa' ion' on 'the equipmen't and power control layouts in the ~RB', especially t in the' Elevation 8 compartment. BNL has ' determined' that there are 'three flooding depths (:1!,23"; l',-10", and 3'-10"),that are. critical,, des are'thus ' tc.the . availability of. va'rious ECCis equiprut. Th'e initiator, event 't s revised accordingly. t N t BNL alsp identified that the random failure of a equipment. protection .s circuit br,eaker coinsiding wit,ti the RB flood event may cause the prppagation of failures to equipment powered by separated Moto'r Control Centers (MCC). 'This potential common m, ode failure event has also been modeled in.BNL event trees. ,. 'Shoreham Plan,t Procedure Gui, des ;, relevant to the RB flooding have been re,,. viewed by BNL. BNL found that these proce' dure guides fail to require a sys-system parameter indfcators in the control room f'ol' lowing a tematic check ofs s RB Flooding Alarm annunciation. This',nay cause the operator. to ignore an aenomal system' parameter,,especially under a multiple alarm situation (such as a turbine trip). 9 t g g .... ~. - ...---..;=- -.a.................... c_*__oj

e 1-2 BNL's revised event trees, q'uantitative evaluation of core vulnerable risk due to RB flooding events, and an uncertainty estimate for the core vulnerable risk are presented in this report. The report is organized as follows: Section 2 'sumarizes the SNPS-PRA ap-proach.to the flood sequence identif,1 cations and quantification. Section 3 pr'esents the BNL revision both in the methodology and in the quantification., Finally, Section 4.0 summarizes the'tesbits.. t 8 0 0' e e a b 6 e 9 g 9 4 g a e g g S 9 g 6 e e g D 9 8 0 0 g 9 0 9 g g 6 9 8 g g a a e f t I me D g 8 0 0 4 g e 8 e O e e e s 6 6 ,g 9 e e e e e O l t O g e S 0 e 4 6 B 9 9 8 e 6 o 4 9 + e ,g.,, p e an M # 8 * *N O**Wmee ** *W @ ' * * ' ' * * "8'-*** O' m nma. e

p 2 2-1 2.0 SNPS METHODOLOGY AND ANALYSIS 2.1 Overview The SNPS methodology for deteruining the contribution to the risk of the internal floods can be divided into,tihree steps. 1. Identification of water sources and pathways to Elevation 8 com-pa rtment.

2.. Evaluation of operators responses and assessment.of likelihood. of ar-resting' the flood.
3.,'Ev'aluatio'n of system responses a'nd identificatidn of the sequqnces leading to a core. vulnerable state given a flood.

In the Shoreham PRA approach it was determined that flooding at location.s ' other 'than Elevation 8 would.be bounded by the anal.ysis of ficoding at the ~ lowest lev'el of the reactor. building Elevation 8, since.the flood water will. .t drain and cascade down.to that' level through stairwells and openings. All the ' evaluitt' ions of flood are hence focused on equipment at the Elevation 8 level.. The volu'me of, water required 'to flood the reactor building Elevation 8 com'partment, with aT1 equipment'and piping inst,alle, is.estimat'ed to be o e 41,600 gallons in SNPS-PRA f or each foot of depth. The fol. lowing' draii. age systems are available to receive.the initial volume,of flood water:

  • Reactor Building' Floor ' Sumps

. - Reactor Buil' ding Equipment Sumps, .' - Reactor Building Porous Con' crete Sumps. These systems have total. sump capacity of 4,650. gallons, and total sump pump capacity of 640 gallons per minute, howeyer, they are not included in the. i analysis. The potential water. sources which may release excessive water in Ele-vation 8 are summarized iri Table 2.1.1. For each'of these sources, a pathway inve.stigation has been performed in the SNPS-PRA, to define the potential for =-.---.j----,.

e. f 2-2 flood at. Elevation 8. Table 2.1.2 summarizes the water sources as evalua'ted in the Shoreham PRA. For each water source the largest possible flow rate has been determined and the time required for,*the flood to reach the 3'-10" level in Elevation 8,,have been estimated. These times are also given in Table 2.1.2. These time,s provide the ba5'is for estimating the probability of successful prevention of' flood a't the 3'-10" level by ope'rator actions. A survey of.all, vital equipment by Shoreham ident'ified a number.of , components for the various acciderit mitigatior) systems which could potentially , ~, be submerged in the. event' of.an internal, flood. Ba' sed on this informa. tion, ~ ~ the critical. height of 3'-10" was defined. It was assumed that if' flood water exceeds the 3'-10" level, 'all EC' S equiprgent would.be disabled. Flooding C sce5arios which are arrested before reaching the 3'-10'" level, have been found to contribute negligibly-in the core 'da' mage frequency.' Functional event trees were used'in 5he Shoreham internal' flood PRA to ' model the plant response given'an. internal 'flded initiator. The flood, initiator freque.ncy was calculated based on two types of internal ' flood precursor's: odline 'mainte, nance and rup'ture of p,iping, valves eY pumps. Thes'e - precursor frequencies are described in, Section 2.'2.' 'Given the occurrence of these' flood ' precursors, th', progr,essico,of events: was modeled using initiator., e ' event trees. Details of the initiator event trees are presented in Section .2.3. . Since all.th'e ECCS systems are assumed lost'given' a 3'-10" flood; the only

  • available m'eans for cooling the, core'are the feedwater and the condt.Cate pump injection. The availability of these two systems depe'n'ds' ort the state of the MSIVs and on i;he. ultimate source of the flood (condensate storage tank or suppression pool).
l..

8ecause'of these depehde.nces,'the erid states 'of the initiatoE ' event trees ' were classified into six categories each of which becomes the entry condition for the functional event tr'es. Table 2.1.3 summarizes the information in a e matrix form. Each row of the matrix depicts one of the 17 types of internal + O A e.--, --sm.~

w..-

..m .,, + - - -, ~ * - +~~.v------*-- ... = ~

i 2-3 flood precursors, the. columns represent the six entry conditions to the functional event trees. The six entry conditions can be grouped into manual shutdown, turbine trip and MSIV closure. Two possible entry conditions are considered for each of these three initiators: flooding due t.o water from the ~ condensate storage ta'nk'(CST) and flooding due'ta water'from other sources. Based on these s.,ix entry conditions, six, functional event t;rees were de-veloped. An example is given in Figure 2.1.1. 2.2 SNPS-PRA Quantification of the Frequency of Flood Initiators Twd types' of, flood initiators were consside, red in the.SNPS-P,RA. 1. F1' cods initiated by an accide.ntal loss of isolation (valve' opening) while & component in the Elevation 8 area is dismantled for main. tenance. 2. Floods initiated by a.ruptu're in th'e ' press.urized or the non-pressurized.part of the piping. 2.2.1 Maintenance-induced Flood Initiators The frequency of the first. type of in,itiator was calculated by estimating the frequency of maintenance [of va'rious components based on nperatinn experience d'ata. The.LER dat' ba'se iri'Ref.2 identifies the ob. served'.failuresi' a f' rom turbine-driven and motor-driven. pump failures. The dat,a used in the-SNP.S-PRA are summarized in Table 2.2.1. There are four failure modes for pumps l 1.e., leak' age / rupture, does not ' start, loss of function, and does not continue to run. The hourly LER failure rates characterize,the i . leakage / rupture failure mode, while demand failure rates consider other f ailure modes. ' - The follo. wing LER r.ates are found for the four failure. modes in s. s., motor-driv.en.and turbine-driven standby pumps. Motor Drhen Pumps - Leakage / rupture: 6 events /6,777,627 h'rs. = 8.9x10-7/hr. - Does not start, loss of function, and does not continue to run: (5+4+6) events /(13,644 demands)=1.1x10-3/ demand I g e

m. a e

-s< _,s

1 i. ( 2-4 SNPS-PRA assumed that these pumps are in standby. status until there i's a demand. The number of demand used in SNPS-PRA are 12 on the ' average per year (four scheduled tests plus eight other occurrences). Hence, the maintenance ~ frequency for motor driven standbj pumps per year is calculated as (8. 9x1'0-7 failure /hr)*(24 hr/ day)*(365 day /yr) + (1.1x15-3/ demand)*(12 d'emand,s/yr) h 2.0x10-2 failure /y' ear. Turbine Driven Pump Similiarly; the maintenance frequengj for turbine driven standby pumps p'er. .. ~ year. is calc,ulated as 0. 79 failure / year. ,There are two not'or driven pumps

  • associated with the Core Spray. System, four motor driven pumps with the LPCI System', and four motor driven pumps as-sociated with the Service Water System 16 which two are linked as a pair to

, the RHR H' eat Exchanger System. bere is on.1y one turbine dr.iven pump as-sociated with the HPCI System and one with the RCIC System. Table 2.2.2* . summarizes the SNPS.-PRA frequencies. associated with ' major maintenance operations. based upon the above evaluatio'n and a conseryati.ve, estimate of heat exchanger on,line maintenance. ~2.2.2;; Rupture-Indu'ced Flood Initiators. ,s The frequencies of the initiators caused by lo's cf system integrity.from s breaks or ruptures were, der.ived from WASH-1400 failure rates of. major com-ponents invol'ving exter'nal.leali and externaI ruptures, bas'ed on a'ssumptions made in NUREG/CR-1363 (Refe'rence 3). This'information has been. summarized in Table' 2.2.3. The calculation of each initiator is done by identifying the. app opriate . type and length of piping and number o'f components susceptible to rupture and summing the estimated ye'arly~ rupture rates. As an example; the total riumber of valves involved in the HPCI

  • discharge system are 3,(2 MOV's and 1 Check Valve); there is no pump involved (Table 2.2.5) and the' total length of piping is 76'. - Referring to Table 2.2.3, the rupture failure r. ate for 100' of pipe section is 4.3x10-II/hr, 'and for external failure of a valve is

=. ~

9 1 2-5 1.3x10-9/hr. The total length of pipe in the HPCI Discharge System is es-timated to be 76' (Table 2.2.5). (3 valves)*(1.3x10-9/hr) + 76'/100' (4.3x10-1l,/hr) = 3.9x10-9/hr or 3.5x10-5/yr. ~ Since the flow rates thr*ough. suction line breaks are time d endent (i.ea, a function of the varying water head in'the source) and a strong fun'ction of' .the break shape and size, a 'simplifid4 model based on historical experiince and engineering judgement is used in the Shoreh.am.PRA to describe the con- ' ditional probability' of break size. Table 2 2.4. summarizes the.. classes of. break size examined. These probabilities; are combined with the frequenci,es estimated for - initiators associated with core spray, HPCI,, RCIC, LPCI, and Service Water Rup.ture/ Leak Suction, System.f ailure to. obtaih.the initiating event, frequencies for no.n-pressurized pipi,ng.'. Table 2.2 6 summarizes tihe fre,quencies of initiators due to the loss of system in,tegr,ity from breaks or ruptures. 243 Initiator Event Tree.s' The probability of causing a flood due to component under maintenance or .the probability of *not arrestin.,g th,e.' f.lood 'is' calcu.lited wi.th the he. lp of' , initiator Event Trees. These trees are 'shown in Figures 2.3.1 through 2.3.17. A discussion 'of the P, D, E. I, and A events. in the' event t.rdes follows. a. Event P'- Operator removes power from equipment and valves. ,The removal of. power from equ'ipme'nt and its'is,olation v'alves is a re- '. qu' ired proc'edure during a maintena'nce. in bo'th foss,il 'and nuclear power. stations. The equipment and, i, solation valves are electrically discon-nected fr,cm their associated power supply by pulling and tagging the' - appropr:iate bieaker' at th'e MCC., A s.econd qualified person perifies . the* correct implementation of the tagging order and placement of' the clearance tags. ' A human error probability (HEP).of 0.01 is assigned for this operator action. This value is determined'using the probability data given in NUREG/CR-1278(4) (p.20 -23). i ] ..--..-.--.~.~,,.,,,.,:--rr------- - - -. - -. - -- = c --s.. ...--..,,..~,,

2-6 b. Event D - System not demanded. During the maintenance process there is a possibility that the safety . systems will be demanded because. of a transient challenge. Isolation valves will, automatically open if the operator has failed to remove power from the isol.ation valves (Event P). j c '.

  • Event E - Operator maintains isolation'.

During on-line maintenance with the equipment disassembled, the isola-tion valves need to be maintained in closed position throug,hout the duratipn of the' maint'enance process..However, an operatoi er'ror could ., inadvertently open isolation val'vhs. SNPS concludes that it is un'likely that the operator will manually open these valves locally in the RB and fail to notice the flood. Opening 'of the isolation. valves at the MCC is also concluded by SNPS. to be unlikely. The, remaining. ' ossibility is that the v'alve,'is opened from the control p ' room (given Event P). The panel switch could be acti'vated by three. events. These events are: the operator mistakenly operates the . switch; a co'mmand' fault' to the valve; or' the operator in' advertently operat'es the switch. 'The, probabilities for these. events.are' 10-3, 10-4, and 10-2,.respectively.. d. Event I .F.loolf annunciation. The excessive water in reactor building is a'nnunciated by a.1' arms in the control room. The.probabilit'y of the operator to fail to notice the alarm (the light is in a "b'ack" pr.nel) is assessed at 10-3,

e.
  • Event A - Operator diagnoses and responds to isolate the flood.

The operator'must identi.fy the source of and isolate the. flood before ' 'it readhes the 3-10"* levsl.' 'Th'is svent is co$sidered by' SNPS under two condit, ions'as foll'ows. 1. Operator isolates flood after auto occurrence, e.g., turbine trip or MSIV closure (Event A ). Multiple alanns will occur in the A, control room at the same time as the flood alarm. 1-ee G S 9 .ee =o-- =.. amom-e e mee -

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i m,

y,,. 1l Inte 3.n.a 3.C u j 4 3.St 5/ l! s; Tr- ~rO.6H. A, Iutell( IRIP (I) I IIA s, I-C H Ill .s I-C N ~ i 8.87 OK . (CS) I A, M5f t sl05uRf 10.088 o.nei 3.s !l 158 I.I(-8 5.C H BlitAINS Al P0ldt se 1'

  • n Gticieer t

^ j 5 eelmata n C

  • 4.0( 7. 5-C = l.3f-P j

i.p w I. iw,r.. i,...i.a,.4.... ir...

4 l

Figure. 2.3.6 Initiator. Event. Tree for Postulated Flooding Serguences j ilisi . Initiated by a 1,lPCI Discharge Pipe Dreak 1e j 1 l ~ i O. O (f' ) 1

= _ ~_ e 2-15 r g .,.sa s a. e s a .e a s s 3 .N . o w U By e o e c aa'e4 e eo e o g5 U. g& i & E =m U ,M e =. w =. e .... w w. c a M. ..E.*w: .w. = GS , g t :m m

  • W w ww -e C

w - ,C_ g. w y! =6 E ge w c < = o + w5 5-e c: 6 o gn '1 E 8'w e ag g 1. .R..z. e e w X w e z --o eo v= C. .s.a c. gg WE ' c. m O C O w. e 6 3 m

  • =

e 6g y C.C a 5 u ga g 2 E m U* 'e N .e k e 'S S 8 .a I. b

=

e e a >- M U E-- 3 E -' "J a e C

  • f5

.E ..o ~- a== t i w $ E h.'". h. g. ..g D. h w.m a w .g 3 -e

== 8 w LT 3 CC w N g. = us

o. g i

= am W g..E g y E -g ,= .a y, ywI.,. a g aw a 'g a g a w-g a. .c.c E _.a i. g g e = sa o5w==E W .= r -a h#3 y e XX,33 -. p { . at e 2 m. E m' g gg-I., N N I,

  • k w

3 33 a. C 5 g 3s=~ z 1 ~ z ut *E E e. =m. E e e

== Imeem ) w x l .i g! 3.W= = s s. a 4 y

d. w--S M

_~ g ~. a e-y

+. 6 4 , = 2 - I, ~. I + e o l l ~ .IulilATOR lythi IME l I BulllAIOR' DI Atlet 5IATUS ((ONDilleleAL Pace.) M Atled assingluG Iulically 5t( At ettlet: :M AE ettle*< tel At CCita*. Ilfue $(fpa sert (At(UL Alls Ifrf Of IIIIIM ~ IME A5 blittWaSt As pf 5Pfue55 Int 10, an (Ohbillage OrfmAIOR Dist&hAlde Inti)qu(V a DlWHANG( 10 Eat t(E 10, ce M. mises15 Iss, AssNlse(I Allis 150 Al($ EI AEI" i (Fer Sait) I, l l#I 10 4 MAfdaAs i:4 55 IN. A A M5lW AND ' It000 5I AM, y . "I A.E Sansliense gunslut full (t oient WCocaelite 88 8 4.j 4-Isla M i 5 i A 8 1 i* M*

  • 4 f

0,8 - l., IMA 3.ltl5 M-0 L' A mensae Swisentas (M) i o.3 .i g., g, 0.003 IMI

  • 3.3t e)

M-0 L l g

  • 50f a

l Inla 1.ar.7 E.M-e/Sais ~.

  • l fu Ota 11NltlNE 1tIP (1)

I

  • ggg
  • 2.nf. 5 10 L

lll r.ec.; 80 t i OK f a 0.3e (5) I A M5iv (t05emi IO.26 3 15A. 4.71: 6 5-0 L ' II "'""I 151 Mmint Al toute s0 4.M.e 50 L = ci icimi i j i Steelnais M.0

  • 8.lt.6. I'.0's 2.M.S. 5 0
  • 9 of.6

^ ~

  • latimJud la time pervleusly evaluated event,treet.

1 I e 1 l .l FI'gure 2.3.8 Initiator Even't Trees for Postulated Flooding Se-t quences Initiated by a I.PCI Discliarge Pipe Dreak 1 .l. 1 I,

i l

n 3 n (

o S 2-17 s e o e e o e e o e e o e g . f' ,gy e o e e o e o.e c.x t*.. $hx he 4 a.' 4 h~' m m. 5 ~ c =~ = en c.c= .a w

W.

~. ~.... a. . =.. eo oc 4 4'. 3 3 sw

  • =3=

a o-e w f.t. 4.: 4 4 '4 J u. z w .w mo ~8 y ca ME

  • 9 1 22

- 3 3.. e3 w$ 3-x x x w -- s- - - e - 3 3o 3 *3 = .e u ' m-o> ga

c. 6

-E c bm W A. o 4** .5 y,= .o N a 8m b en -= ..o g 1*2 '>- 'o -g e = k-e. au = E ecm- = e e o = 5 m: a ~.

-

o, W* g t = W -o-L Iggg 8 E. E. .o., w g. - ~- = e o e E g w u g e,' 4 = 4 a y ..a.c g .. - a a g. -o g o - e :1 w c= e = - zo

== w-x = >* C" w W a

  • .w a.w

} s =- = a r ~. m s g 1-s w$4. 5 m. I 8 *e = = 8 ' . -= y w w*3.Y e- ~

  • N w

e n. = I '5 4 $ 2g255 'g I c o u E a. E2$I g = = "E 3I 5 L m e Ue EE z { = W N EE 45 E

  • e e

Ir;223 I e' e e E g e 5 3 n p 3 -as u .e =. - y.i 13

=. w ~ 1

,f. ? ~ w. l g 3* p . 2 a e m g -m, .=8*W..u.* s... --s.. e.2 -p m. e yw' g

-~ . - ~ ~. o S. '2-18 ~ ( z -g . s eo e e e.. e e. s c. e e ,BX 9*

  • * * * *
  • e *

= t E a a = -c e o. a. w e. > e e o e e .C 2 EM. TT T T T T T -m l wc

  • sw"
x. 3
3.
  • x. *.
  • y =.
  • y,u was e-c.

e e. e ygg w-gam ~ m c., .3 WE 2.* -6 a ~3 e a, m v . a < a 3 3 x

o - = o - cA = L g .%g

  • g.8..

oc a = e e. a 5 -3 9

g. m a

o-f

6. a.

g 8 *.N s. c W W w.c -- ~ a =< =q 5m 3 cw 5 = E 3 3 _~ o e

> c e

3 . e u.: v c ,c b@ ~= X 8=-~'= a o-3- 8 a e a a a .-g..a -. e e .e .e 8 8 3 .m .a 4. a. e- @ 4.s b 4 E 3 gg tf. 5. a_I 3 g as

      • 4 4.
  • Mno el

.W' s.n w p

==x a .= 8, g . o,* e ~t g was

  • e 3

3 Y g = e ? 3 (a. ,,, e,,. g g M 83Im t y-e 4 I N =sg ~. h. e EC Eq" s e e 3 g . t m_ 5 e' 3 c 3! m83 -l E! e. e-g a 1 g WlEe3S=3 3 -~s = - z g c. . so .a - i

      • 3"e3 e,

E=g s a a o v. q x n d. n. .t R = . ~3 ~ + M* U" y b" .WD e. 5

=

y -e. - meg e... egee e.e e a neo.4 e es.. e...

e. e me.

e g -- ..=""'**9 &9-** --.-*.,-,,,.m**e--,..w. , - - - -, w -v-e ID.<***e e.q..* ~ ...e 5 .-- +, - - + -. e--,

e i._ t

l.

) 3 l ~ 3 )> 8 .198881 A10E (Viul lett. lhlllAIGA bl Aflim SIAlut (tGeelliossAt rang.) gg Atlast tul DissG lullt.mily ~NI A8

    1. l At Siti AE

. gig,g 5(fp4 Wl-(Attut Alta Ivel Of . ST5itM ( 00lh Alner (DOpd astf Islipsissf r $(f3ut h([-.

CJC gag g gen, antitet A1 St um115 MDes put

,,,nemillf,W g gqg pe gg Isl5818s50 30 4181 suit 30, 10 is ,,g g g,,, gg,g agg g pg 4(Ices (p,, p. y, y t ,8 Mu ppset H11$1AH5 ID tR hlSIA Ib 9($O4 I% IN, g g gg,gg SIA)US } A HA8MsAt let. A list. A MilW

  1. 88,

$sastluna. alms.selP (s mste( NI UW.Ni tt O ,IIII I M I 5 l A A gggy a. I IMA 1.I(-18 ,M-C

  • R ygg.f 10.3 MAsmaat 5sensinuit (n)

Inne. , 5-C a ,63 ini 3.7(.9. eMC g a e'.ch i 3. IMIA 7 l!.le 1.C g . R Of l.ht.6/este i lueelms tale (1)' n. coil e IIA 9.lf.le. I.C R gg 4 lit 2.ti.9 . I.C . R b I to ~ e.ss I - (5) .'o.tmil 154 - 1.sc.le 5.C n w it.(Insief e.la

  • ** )

151 c.et.le 5-C . n l. ein4 Ins Al enura a uttalslets i i 4 siewAmit Il-C ' l.et.s. T.C.*3'ac.s. 5.C. l.sg.,

  • Isocle. led in tim prevleusly evalueled event trees.

l i Figure 2.3.11 Initiator Event Tree for Postulated Flooding Sequences ' Initiated by a Haximist RCIC Suction Line Br'.eak b e t q I J l l I J

. -. -. - ~. - n.. ~...n,... 4 e 2-20 e x .G . s a z a e z =. m z. M .. c.t g = .m U O Y Q. Q W b.b bp W84.9 g =* g e

  • e 3

E. 5 z.= x

  • C e

a - U o* MM -55'. ? ~. 2 AU. ,e - gg,. ..::,.= . s.. x. . r. x. =. x. a4 er M I, Q.g e-g nc e s u. ,3 . ~ Y* 2 -s e .E g 6 co = E s. 1Ex- - .x. .= w- ,u .g -u

== um g,2

  • =,;

4 o-

c. u C

5, e. e> a 6= w y o P ~ % =S 3*

e. 1 z

. e= . s.. < = a-6x e a d 6 5 3 3 = a E C #U o a --e- >x -e..<=g S. e w = .u.a.::- E gr e e ss u.c.c.. i c g ee 12 2 .g. -j ,3-ta -c j wa s s =V=.c. = -.c e e-g e 2 'sa*5 e g cs=.- e - cy. 358a3 4 M. e .W e c .-==g=E a = 4 =:: m. w a 3 y 9, 3: = g. 5 3 g d-gE Re. a w wa = .=r 5 I M 5il= ' z 3 g Yz a o 3 e sy a 31mee x u. = E-E m = ~ I e t, -= z.12-2 g =. w y =1 1 s x-g l I ) 8

        • -m=e.4p ae
  • w____

OM w n.. ~ r

, I - ) - t j 8 l lattlA18R EWhf Itf( laillAfee al Atlest SIAlu5 (CotStil010AL P900.) St ACIOS BullethG INI(GRilf esPCI UE AE OCCle se( AK CCCups neg Ag attles ft00s 5(spatuCE CAICIA tite IVP( ef' SVil(M Distf.hAIDA fnEqut1CV 5f tp41sCE f 5ta[Iless 15 M SP0ss'.C A5 M5PON'4 tot 10. 011 . (chellines OPIBAIOR pgACig - sk(AK 10 A stAsNIAl. Set 15 IN'. A A M539 AND 10000 ~er SeVr) L Iht.

10. 0AI(A05 10. 0A et.

Af 54115 lit. AssimmeCI Alto lietAlf5 SIAlW5 l Anf.(

  • Slaslighet -

(tmellet 18lP Ctesues dt(OtJell(B g tII M I 5 I A R f Ill3 q - I.C I. ...(.. I 4 HAsasAt 51millsass (n) IMAA 8.7( 9 5-C N s.'esii IMI 4.1[.9 M -C ll

'"I i

IMit 2.6(-9 5-C H I .n l 1.5(-6/s.Vr OK* s N ' 0.11 . Itatist IRIP fil IIA a, T-C H H e.a01 Ill s I-C H 8 88 4 ~ Oc gg ,0.lt ,MSIV C(0$uht n ISA 6.5( 8 5C N 15 0.003 ItsseAlms' AT Peuta so ISI l.st.9 5-C H s! g gg,, g SiteenaV: M C

  • 2.6(.S. 5-C = 7.St.S
  • lacluded la tN peevleusly evaluated event treet.

i . 1 1 g Figure 2.3.13 Initiator Event Tree for Postulated Flooding Sequences Initiat'ed lly a large llPCI Suction Line II.reak I O I 7

r -- --~ ............... ~., _,,. - ~.. ~~_. _.. - e e. 4 2-22 e e E e as s e e a s e s s e g o. M tu e .s U'.3d e.-.e. e.. .e. 53 as, =6 2' E-c*:::: o = zu mc C e m-e '* g

  • g e a.

e cJ M. j **.e',. j $ffe g j," w w a e g oo a.2 4 -4 4 a.; w"* .C*- e ea u; U w3 E = M-o a wj Tm a tu = r s 1g 5 as

s,r e

m 6 3 C. um "W 5I* O a c:., 6 w w-a ,a C O2 d d e >=

. s. v 3,,, iC, - a W g ve Wa f

e e-
*g 8.

-j rm 3 Y e e yn - 2 s 3-g Z =. , =a-M,o 3.e::: e .. m na c. c

5. g)* =. s -

-a E. 3 > h a .g. I + 't = ua a * -y 9us 8*i e e e 5T

  • h oy 3e g

ww., Q gg, + g, y-ee w g I 1 aa -w M,a,, g - n= 3 = -s fw,J.s=

e. m 9

-r a. Iw w =r y f.- 3 &e ~ a 3 e 4 c. =Ne s c 82Rg El ag ,,, y m Efis v E,* l 9 A i = a

kast casI3 1

L 6 e b 5 5.='a m = 3 I = e. e

==W a g I

Isss 3

r 5 55.' ,,, %.i.t ,4 - i. .e '1

g lI M _g,.

g = i 3 - Y ,y1 5-w I ~ f D .e, W. e,e

  • .e ew

. me *N e.e= ae menuee m* * * * * - Mme.eMe -

  • .M W Dee, er F**M-**

MS -*.-W M *, @,

.g. j .i + l 141tlAlqn telut IN( 4 lufflates M ALI4mt $1Alus lietell10ssAs Ptos ) bl At tom BullDING 1811(Git lIf

  • P 3'

($ Sutfitus nMAE g(CisteM AE C((tst'.SM AE ACCtHt* fifWe Ol't hAIGR M AClflR t Isat bkl A6 AS M$rtue.I ".5 D(5 Pot 51 last IG. On (Onellloll lint Al(5

  • SIAlv5 5(fp4NCE

'CAlttEAl(D Irri af 5V51(M i SAWLE le,0 ell (ADS so. OS 8t.- 45ae 15 is, Assase(I Alle fisste taillAIOR th(qulNCV 5(QlEIICE le A 8tAfuaAt %L4 35 IN. A .A h51v s M titi.nlille (Per se te] Subtlpas tuleslut 1818 (1054est T ill5 t n 5 I i e !LtL _ m. int ,3.et e 8e4 t S.el _.. 8mseAt Suuionisein), g e.3 g, ,, g,, 3,, g c.cos ing .I.or-e 88-e L , 8. 3 i INIA 4.4t.9 5-8 L 4.st-s/s. v. N .i GE. ,e.Il R yg, g m timatat tale (1) W S col Til e 1 -0.* t 0.05 M nggy. .g s.Il Elesumt 151 I 15A 5.4(-s 5-0 L I** 15l

  • 3.5E.le
  • 50 L

8tnnlN5 Al Find N s 0 ugg gggggg j, + 8 ( Slettants 84 0

  • 4.4t 8. 5-0 ; ).5(-S i
  • laclieded la teis reselessly evelvated event trees.

g j i Figure 2.3.15. Initiator Event Tree for Postulated Flooding Sequences . Initiated by A Large Core Spr'ay Suction tihe Failure .l iil l j l t

n -e 9 t. 2-24 z t = G a o v c = B M. c- ,,,,5- = .s a z. c-e. C O3 Au m = 6 . Ci c .. S. ="

== W ". . =. Cb g g g'

8'

- c; =. *..,* g*. , W r ". was 5g - ' a - w-- CC = "3 6 -5 W- -c c .a ' 5.3 a - m w 'W C --5.d u-3 3 1 .x. .. s

w W

ea - c -e

2 -.

MA Eg m o- ' 5-E 2 cv e e c. g-3 g-E.. te .E - sa 8,=

    • a 8

't o-3 $y~- M 9 . s" g u 2' E_ e ~ -u ~* g = a.. u :- 2 c 8 e =* = >w. =E:w{ a 6.= x -. w a .,I E a{. u -J u - a = ce c5 ua ~5x ea + M. a. y -E g aw .v2-fg

3. s 3_

I_yy.*w ..fc n c e.o ~. a pue eu .e. e wi , *U.qg 3 .e g = - 3 w a - a.- E = = I 5 m. gg* =.e z ~ w ~ ~ es5+ g g n g

latin!

.'. Y -E ~ g. g.3* a T e_ s E I'g a 3 es s E N J.1-z g 2 4 L 6 C E =, l u-e

  • =eo*3 4

1 .c x =- 2+ $#~~ 6 I g -M 5 1 2 f i i g i e E a. i

  • u a

l $y 5 8 4 i j Y". -. M_ 1 3 g.i ~

  • 4.D 4

l e i

= m Q j i. t,; .(; i c. = = + j laillAlum (V(ut Ip(( INIllAIOR pt ACION SIAIUS (Copeilliestat "relle.) MAtton positpinG Intf&allV ? 3 ipa't 5tstileuepeM AE ettedes te( AE OCCs:Atdel AE SCCMAi fl0018 OPSAAIGA M ACIGA l Ileat set AE gg gsrese5( AS MSPon51 MiaC 10. OR Colullilius a 1544 Al(5 'SIAIUS 5(fluttett ' CArtisAlte IIP ( Of ' SV5IE M i 1A8CE lit Oltt! Ass ag ne pf 50t l Ituts 15 Ill Atuassfl Alti It000 f(51Gf1410A l'ettp(leCV SitPILIICE l 40 A pensanal IN A Ilmalell 4 M519 s at torJsllfl' (Per Rs Tr) 8 5emtismes IhlP (B estait i l gg, 8 8 T,, p. M 'I 5 .I A , R g ') t on* 4 _ _.J. L 0.8 InA

2. M - 7 M-0

'*'"*I 8'"88 l'!I ~ 83 f inAn g.g,m 50 I t l-0.sini ~ test 6.iit.S M0 L 0.3 L l! . f.9C.9 50 IMill 8 nsa '1 functnt Isle (f) . ' O.54 i s.n. spa.Vr 8 i I

0. ens ggg

,,C,, L I Ill. 4.9(.9 I-0 L e,3s -'I 0.54 .it I l'Il' 5-O L l n 0 M (5) I g,,. ' 154 . I.u(.7 0.it a.nni. ,3, ,,,g,, istda Ictel t. .f """'"I # I U "" " * " e SillHAlif 8 M O

  • 2.4(.7. 5 0 = 0.et.P.5 0 *2.M.t*

t( ~ f. 8 l

  • lecluJud la the prevleunty evaleased event tract.

.j ,i : l) ', ~ ~ , ~ Figilre 2.3.17 Initiator Tcent' Tree for Postulated Flooding Seglsences Initiateti lay a Large~ LPCI Suction Line Brehk .] g 1

w e t 2-26 i 1 1.0. l ,*.'s g LEGDo g U., WRE3-1273 Art. 2 MEP for single event s. \\ NUREG-1273 RD.2 NEP \\ for t.e sec:na of.=ulti-sie events

  • a gg 1 D

ce g. c gg <= \\. \\, e = c. UIPLE DENT .* f"t., c= MEP e". 2 Q* 'o tg u c d -fact! nm = nu m o, , =, .s, s .'s.,,m C. lo** - e 1 e .to.g to 20 so 4 50 W Time (Minutes) '. F.igure 2.3.18 Comparison of the HEPs Associated with. ' Operator Actions f6r Singular. Events and.* i Coincident. Multiple Events 4 e O +.-.e, ._.ap.-+

    • 1 m

-. =. . ~........ -_m

i p s i '2-27 Table 2.1.1 Summary of Potential Water Sources and Types of Initiators Which may Lead to Release of Excessive Water in the Elevation 8 Compartment ~' No. of Source Quantity (Gallons) Lines Systems Involved Suppression, Pool 160,~000* 8-CS,LPCI,RCIC,HPCI 4 CS,HPCI,RCIC' Condensate Storage Tank (CST) 550,000 Reactor. Prima ry ' System ** a) 42,928-b). 152,,9,28 Scree,nwell (Long Island Sound) Unlimited 4 Service Wate'r Wat'er Fire' Protection System ,'Stor. age Tank' 600,000

Many,

. Fire Main s.

  • Total wa'ter' volume in tTie suppression *po61 at the high water level mark is

'608,500 gallons., However, only a portion of.the water can be drained' I

  • ' I.

thr.ough.ECCS pump sucti'on piping.

    • Figure (a)includhswa.ter from the. bottom. of the core to normal" water level in the RPV. Figure-(b) includes -(a) plus condenser hotwell water.

0 e g f i' :. .) - .l\\ i' ) r F 4 ee . n w.2 m.-ea-.. ....,..,__.,..,.e .,_.,,-..%.W-....mr.,.,,,..,,...m.,,- .w..,%

.._ -m._ _ _. u. - i. 2-28 Table 2.1.2 Summary of Internal Flooding Initiator Types: Source, Pathway, Flowrates, and Time to Critical Flooding Depth Elevation 8 Flooding Time . Flow Rate (Mi nute,s*) source Location opm* 3'-10" Suppression Pool HPCI. Pump Suction 9600

  • 17.6 RCIC Pump Suction 1500 '

10.6 ~ LPCI Pump S'uction 1 (Max /Large)** 17000/8500 9.4/19.0 CS Pump Suction 13000 12.0 LPCI' Pump Suction 10500 I5.0 (1PumpRunout) CS Pump Discharge 6850 23.0 (1PumpRunout) ~ Condensate' Storage Tank (CST), HPCI Pump Suct' ion

  • '13.0/27.0 (Max /Large)**

1200/6000 ~ RCIC Pump Suction 2100 76.0 CS Pump Suction (Max /Large)**. 1200/6000 13.0/27.0 HP.CI Pump Dischargh - 4350 37.0, ~ - (Desi.gn) f.. t Service Water RHR' Heat Exchanger 8000 20.0 .(PumpRunout) WFPS 'Ru,pture of8" Pipe '4000 40.0 i

  • These / ood times were calculated based on a failure of. the sump pumps to fl successfully oper' ate and a 41,600 gallon per foot depth in the reactor building given in the Shoreham FSAR.

il l 1

    • Lahje. flow rates assunied' to be 1/2 maximum flow. '.

..t e

  • e

-w

_ _ 7.-.. e i 2 29 Table 2.1.3 Sumary of System Event Tree Entry States by Initiator Type SY11tM [V[NT 1 ACE [NTRY ConolTION FRIOuCNCY (Per Rs Tr) N-C T d' T-C. 5-0 5-C INiilATOR M-0 T ' " 1.8a10 3.8:10 7.6:10 4.3 10*8 ~ ygg TTL2' 5.7a10*I 5.Ja10*I' 2.5al(I '5.0x10 T 3'.0a10 1.1x10',5 gg3 T 5.8aI0*I . 4.3 10 gg, i .3.6a10 6.1x10-8 gg$ 1.0a10-'

1. sale' T,t, 1,gg 6.4al'0'I 3 5:10*7 T,1, i.iai0 5 2.0;ies.

... 0mics yg, . 1.3a10 2.fa10*I 5.Sa10*I T T 2.3a10*' 2.8aiO*8 1.410. FL10 1.8a10 g, .3.4 10 1,.5a10 .IAll T 1.0:10'I . 2.la10*I TL12 ggg3 ,2.6s10*! 7.8a10, ^ T .T 1.6 10 ggg4 2.ga10 'T 4.4a10 ,2.5a10 ' TL15 4.la10*I

  • 6'.6a10*I

.T 1.141 fLl6 i 2.4alo'I 4.8:10*I

  • 2Ja10'I gggy 10!AL5

'1.6a10 5 8.2a10'I 2.2a10 3.410 l.7.alg 5 5.5a10 -e s g s -- - i l s ' si w. ,..i' ' 'i...i.ii.i.... 1.h_'___1_u-.-nm _n----

A--

2-30 Table 2.2.1 LER Data for BWR Standby Pumps for the Period of January 1972 Through April 1978 Does Not Standby Standby . Leakage. Does Not loss of Continue Pumps - Demands Hours. Rupture Start . Function To Run Motor'.. Driven 13,644'- 6,777,627. 6 5 4 6 Turbine .1,,820 868,03.3 Driven 1 . 6-5 i . Table.2.2.2 Frequency.of Online Mafor Maintenance System in' the Reactor Building. Frequency (Per Initi,ator Sy' stem, Year) SNSP-PRA Event.T.ee, . Core Spray (Motor Driven) t 0.042 .TFL3

,LP.CI.(Motor Driven),

, O'.08.4. ,,TFL4 HPCI (Turbine Driven) ' 0.079 ' TFL2 RCIC (Turbine Driven') 0'.079 . TFL1

  • Servi'ce Naterl(RHR or RB;CLCW HX),(Motor Drive.n)

.0.042 TFLS... .i. j' t e g oe

  • s g

m - m .. - - +

_ u =.:. a.: _ _ : -

.,_ _.. = _. = _..

x_ s.2 i m 2-31 Table 2.2.3 Summary of Failure Rates for Major Components Involving External Leak and External Rupture. Total Failure Rupture

  • Parameter Rate Rate /Hr (Mean)

Reference Failure Rate /Hr ' Pipe' Failure Section 8.5E-10 WASH-1400 4.3E'-11 (100') External Failure.of WASH-1400 1.3E-9 a Valve 2.7E-S., ~ External Failure of 1.5E-10' a Pump

  • 3.0E-9 WASH-1400
  • Based upon the operating experience to date, given that a failure occurs, the ratio of external leaks to complete failures appears to be in the range of 20 to 1.

This is substantiated by*.the specific data review cited in the text for.v'alues (18 to 1) and data. published by Bushl.DI on pipes (4 to.1 'up to .30 to 1). Because the internal ' flood e' valuation is based upon initiators with substantial. flooding rates, i.e., short' operator response times, only, the catastrophic or large external rupture failures are treated in this-evaluation. l. ji i. Table 2.2.4. Condit.ional Probabfifty of Pipe Break Size Break i Conditional.' Size Characterization-Flow Rate Probability Maximu'm Guillotine Break 100% 0.05 Large Substa'ntial Rupture 50% 0.10 Small

  • Localized Rupture in Ductile-Material 13%

0.35 e i t

  • Remainder' of the conditional probability was alTocated to small breaks.

i O e e e e =*v.e.,. e ,e, +w-,,-. ,4 _,.,m

-2 = i 2-32 i Table 2.2.5 Initiating Event Frequency Estimates Involving Component Leak / Ruptures VALVES PIPING ~ ESTIMATED " INITIATOR SOURCE LENGTH (FT)/ FREQUENCY /, MOV MAN CHK PUMPS SECT /DIA (IN) YR

HFCL 01senarse,

CST /SUPP. .2 0 1 0 .76/1/14 3.5E-5 T '5 FL Discharge SUPP '4 '0 - ' -

  • CS 2

0 128'/2/12 6.9E-5' T pt7 LPCI 2.5E-4 T ' ' ' ~ ' ' ~ ' ' ~ ~ ' ' " - ' '240/6/16 Discharge. SUPP 14 4 4 0 FL8 Service Service Water. Water 4 4 4 '0 - 7,15/8/10-20

  • 1.4E-4

.T yt.g d WFPS WFPS ' 1' 157/2/6-8 ' ,1;.1E-5 ' i ~~ 3 fil,0 T RCIC** Suction CST 1 1 1 1 70/1/6 3.5E ; T dt l y r l l HPCl** Suction CST ** 1 1 1 1 87/1/16I 3.5E-5 Tyttj,TFL13

CS **,
c..,,

l-120/2/12' j 4.9E-5 Suction ^ CST

  • 2

'2 2 TFL14 TFL15 ~ LPCI+* Suction SUPP-4 4 L20/2/20 5.2E-5 TFL16,Tpt17

  • CST is assumed to be the source.
    • Suction failures are also classified by flow rate.

0

- - _ _ _ z.__ ~=- ~ i. 2-33 Table 2.2.6 Calculated Frequencies for Initiating Events Resulting from System Ruptures (SNPS-PRA) Initiator Frequency (Per RX Yr) Press 6rized Piping H.PCI Discha.rge Break, TFL6,' 3.5x10-5 .6.9x10-5 CS Discharge Break, TFL7 ,2.5x10-4 LPCI Discharge Break, TFL8 .SW' Discharge Break,'TFL9 l.4x10-4 'WFPS Discharge Break, TFL10 1.1x10-5

Non-Pressurized Piping T

1.'7 5x10-6. RCIC Suction Failure, ' F11 '(max} HPCI Suction. Failure, TF12 (max) 1.75x10-6. HPCI Suction Failure, TF13 (large) 3.5x10-6+ CS Suctfon Failure,, TF14 (' a,x) 2.5x10-0* m 4.5Tx10-6* - CS Sueti,on Failure, TFiii (large)' LPCI Suction Failure,'TF16 (max) 2.6x10-6* ~ ,' 5.2x1'0-6. LPCI Suction Failure,' TF17 (large) '* Modified based upon engineering judgement made on the size of low p'ressure - suction line breaks.- I.

li, c

..t r ee

  1. r<sne -m *

,,~em-., .,e,,,y-&+-y.%- .e* a e

rr e,

-m--

_ _ j__ x---~ .. 7 1 2-34 .. ~. Tabl e 2.3.1' ' THE PROBABILITY THAT FLOOD REMAINS UNISOLATED FOR X MINUTES AFTER AUTOMATIC PLANT ACTION: E.G., TURBINE TRIP OR MSIV CLO,5URE X P(for. multiple event) P(fbr' single event) l I 1.0 1 l t + 2nd = 0.'S4l 0.1 i 10' 0.11 0.01 20 0.01i 1.1E-3 ~ 30 2.0E-4 i 6,0 0.0011 .s

'1500 i.

1.1E-4 1.1E-4 i g e g 8 I g g c j-. \\ l. s e 6 6 9 6 8

c...

m- --m:w.s.,_.2. i 3-1 3.0 BNL ACCIDENT REVIEW AND SEQUENCE QUANTIFICATION This section discusses the quantification and review of the internal flooding accident sequences in the SNPS-PRA due to system maintenance and pipe ruptures. The section.is organized as follows. Subsection 3.1 presents a summary. of the appr'oac'h used by' BNL. to calculate the initiator frequencies. S.ubsection 3.2 discusses BNL quantitative review'of the initiator eve.nt trees, ~ and Subsection 3.3 presents the functional event tree analysis and evaluation. 3.1-Flood Pretursbr: Frequency This review revised the assessment of the fr'equency of the flood initia-tors in two ways. First the experiential data' for the estimation of the var- - ious'failur'e rates were revised to include recent'avents. Second, the models for calculating the frequency of' floods (or probability 'per.ye'ar of ~ reactor. ope, ration)L have 'beeri improved by r,emoving unnece,ss'ary conservatisms. As 1,t was-already. discussed 'in Section 2.2, two types of initiators were. con e sidered: a) maintenance-induced initiat'o'rs; and b), ruptureeinduced initiators. The revised frequencies 'for the.se types ok in'itiators are presented'in the following two subsections. ~, 3.1.1 Mairiteriance-Induced Flood Initiators. A flood can.be, initiated during the, maYn,tenance of a component of the ECCS ,or of another s'ystent in th'e Elevatior) 8 area if the Eaintenance process I requires dismantl.ing of'the component and one of the isola. tion valves o' pens ' inady'ertentiy whil'e the component is being maintained. I

  • The components that contribute to thes'e' initiators are. the pumps and the h' at exchangers in 'the Elevation 8 area..

These ' r'e standby components that a c can fa'il ir} a time-dependent fashion' while on standby. Periodic tests are performed to check their ' operability and if found failed they are'put under l' repai r. A Markov model that describes the stochastic behavior of these components ,has been developed and quantified. The important characteristics of this f model are as follows: I eein

3-2 1).The component can be'in six states (see Figure 3.1.1).

11) In state 1 t'he component (pump, heat exchanger) is available, that is ready to start operating if asked to do so.

~ iii) The component while on, standby can fail with exponentially dis-tributed times to failure. A failure brings the component' into '. state 2'(see Figure'3.1.1). iy) The failure.rema. ins undetectable until a test is perfonned or a real' challenge is~ posed to the comhonent. A test that will find the conf., ponent in state 2 will initiat'c a repair action. The same will hap-pe'n following a real demand for the component. y) There are three repair states. States 3 and 3' in which the com- , ponent is under repair khile' the reactor is online, and State 4 where the' component is' under repair with thefreactor shutdown. vi).Following a test' that finds tha. component failed' and before' the' dis-mantling of the component,, all the.appropria'te motor operated valves must be closed and' their breakers racked out from the corresponding ~ l. ,NCCs. ; There'is,'however, a ' chance that.the, operator will not remove the. breakers.from the.MCCs. leaving then the. MOVs able to. open fol, lowing a signal to do so. If the probabil'ity,of such an error is P,, then a test brings t'e component from State 2, to State 3'with h probability 1-P (tre'aker removed) a'pd to State '3' with probability p, vii) The component remains in Statis 3 or 3' until the repair is completed and then it returns to State 1. or until the allowable outage time is exhausted and then the component transit to State 4lwhere the repair - r continues witit the reactor shutdown. When the repair is completed, 1 the reactor is brought back online and the c6mponent returns to State

1. (Transition 4 to'1).

l-j \\ I e ,e c -,-n ..n- ,,---,,~e-a- -.~

7.... -.7 3-3 Quantification The solution of the model requires the quantification of the following parameters.

1) The catastrophic failure rate A.

This failure mode implie,s such ~ ' failures that. require major maintenance (dismantling) of the com-p.onerit. The SNPS-PRA used the da'ta presented in Table 2.2.1 from Ref. 2. BNL has updated this table using additional data included in an

  • updat.ed version of Ref.'2'(Ref.6). The new data are summarized,in.

Table 3.1.1.. Naximum likelihood estimators 'for the failure rat'es number of fa,ilures 1=(total. operating time) yield ~ A=5.7x10-5/hr for Turbine Dri.ven Pumps and A=3.3x10-6/,hr for Motor Driven Pumps _ The mean times to repair were assumed 100 h'rs and 50 hrs for the 11) ttirbine. driven and the reactor driven pumps', r'espectively. Thu,s,. y=10-2/hr fo'r' Tu'r'bi ne" Driven ' Pumps ' ~ .and. u=2x10-2/hr for Motor Driven' Pumps. ) iii)' In the BNL' revision of the SNPS-PRA, the frequency of tra'nsients involving MSIV closure, ha's been issessed at.4.42/yr. 'Thus.th'e ' f,requency of. tr.ansients on an -hourly basis is ,A =5.0x10-4/hr D J,. 1,v). Testst.ar.e performed,qvery 3 months (4 times a yepr) f9r, both motor -,. driven and turbine-driven pumps. The allowable outage times are 14 and 7 days for turbine-driven and motor-driven pumps, respecti,vely. ae 6 s

= 34 v) The probability of not racking out'the breakers of the isolation valves (P) is assessed in the S.NPS-PRA as 10-2 The same value is used in these requantifications. vi) The mean time for inadvertently activating a particular switch in the contiol room h&s been assumed equal to 10,000 hrs., This'. implies a rate.of A =104/hr. o Quantification of the McFkovian mod.el'with.the nutnerical values of. the paramet'ers mehtioned ab6ve yields the probabilities per year for the various maintenance induced floods.. The resul_ts are tabulated in Table 3.1'2. ~ Additional assumptions are: the' Core Spray System consists of two motor driven pumps, the LPCI consists of four mo' tor driven pumps,and that,RBCLCW heat exchangers are equiv'alent t'o motor driven pumps. 3.1.2 Rupture-Induced Flood Initiators ~ A flood can be initiated if a rupture occurs at any point in the pressure boundary of the va,rious systems in the Elevation 8 area. Such'a rupture will' involve.one of'the following three types of components: 1) piping; 2) valve; I and 3), pump. The.model as,sumes. that '.ca,tastrophic ruptures occur in the,fol,- lowing way'. A component fails in such a way that if it is demanded to ope-rate then 'a catastrophic rupture (large enough to allow the flow rates neces-sary.for t,he flood sizes of interest to this' analysis) will' occur. That is, the component transits first in a' rupture-vulnerable's' tate and then, when a de-mand occurs, it ruptures.. A Na'rkov model that decribes this stochastic behavior has been developed. and quantified. The model is graphically depicted in Figure 3.1.2.. The basic characteYistics of the model are as follows:.t (1) The system in question (HPCI, RCIC, LPCI, CS, RHR, RBCLCWHX,) is in state where it is available tc perform its function. 9 ee 4 e

.. u. 0 Sam 3-5 -(ii) The system transits-to State 2, which is a rupture vulnerable state with failure rate A

  • R (iii) If a demand occurs while in State 2 a flood is initiated. A demand occurs whenever a transient, a manual shutdown or a test occurs. We

, distinguish thre'e flood states:. State 3, which 'is a rupture trig-gered by a transient involving 'an MSIV closure; State 4, which 1,s'a rupture triggerid by a t'urbtne-trip transient; and State 5 which is ~ iupture trig'ered by a manua,1 shutdown or an' equipment, test. g The solution,of 'this'model ~ yields 'the probabilities that the system will occupy States *3, 4 and 5 denoted by P, P, Pg respectively. These S T probabilities at the end of one-year period provide the frequency of rupture-initiatid flood precursors. The, expression for these probabilities is -[(1-e-At')/Ag_'(i.,-At)fx] R .Pj(t,) ~= F 'A-A ' (~1). R where i = S, T F is the number of tests per' ysar. Ag jis the rate of arrival of a tran'sient of ty'pe 1, (i=S,T), A "is' the r' ate 'of cat'astrophic 'rutpure' failure 'in the sy' stem i R 'and A is the rate of arrival.of any tran'ient (A=A +A +A ) s 3 T M For the manual.shutdoM the corresponding expression is AAMR AR-P(t.)=F[A-A(1-e-AN)/Ag-(1-e-At)fx+ A-A (e-A T e T)] M R A R R (2) Quantification For a given' system' having piping of length' L, n'y,va'lves and h pumpi. p the failure rate AR.is equal to AR

  • U '+"v v+np p (3)'

l A 1 4 =

.= 0 3-6 where A, A are the catastrophic rupture failure rates for valves and y p pump and A' the same failure rate per unit of piping length'. A search of the LER, has indicated that at least one pipe rupture (welding failure) has occurred in the ECCS piping in the 215 accumulated BWR years (see Ref.8). .. This provides a maxjmum likelihood estimator for the. ruptu're failure rate- ~ of (1/215yN5.31x10-7/hr). Assuming, as in the SNPS-PRA, that only one out of every twenty ruptures will create a bre'ak 'lahe enou'gh to generate floods, 'of the. sizes. of' concern to this' analysis, the ca,tastrophic pi,pi,ng rupture ra'te becomes A=2.7x10-8, Thi,s of course is applicable for the total fength, of safety related piping (denoted by L). F,or a pa'rticular sfstem with a total of piping length 1, then the. . catastrophic rupture ' rate for piping' becomes A"=(f)x2.7x10-8 /hr (4 ) ' where 5/L denotes the fraction of the total length.of the piping that belongs. 'to the particular system; 'For.the rupture rates of the valves and the pumps, the WASH-14.00 v.alues weie used,(s'ee Table.G.4-4, in SNPS-PRh)'. Using the length of piping, number ' of vitives and pumps provided in Table G.4-5 'of the SNPS-PRA, and by virtue of

  • for the various. systems 'along with the l

Eqs.1-3. ' The total failure rate AR probabilities P, PT and Pg were ca.lculatei The result; are tabulated 3 in Table 3.1.3. l l A total of 13.51 transiints per year were assumed (4.42 MSIV closures, .4.89 turbine ~. trips,and,4.2 manual, shutdowns).

  • p The splitting between maximum and'larga ficods for initiators TFL12-TFL13,
  • TFL14-TFL15, TFL'16-TFL17 was done as in the SNPS-PRA, that is, I to 2.

The additional factor of 20 used in the SNPS-PRA to account for non-pressurized piping,is not assumed in the BNL quantification. I l (

~ -~.--.... . w.. - 3-7 3.2 BNL Ouantitative Review of the Initiator Event Tree The quantita'tive review of the initiator event trees is discussed in the following subsections. 3.2.1 Review of Flooding Alarm Related Procedures 'The RB watei level is detected by two RB water level,1:enitors installed on the RB floor. The flood alar,ms are activated by the monitors when.the water level is more than 0 5 in. above the floor. The. sump al.ams will be activated when water level ' reach'es the sump alarm setpoints instal' led.at a level right ,below the level that activates the' RB flood alanas. SQmp alarm sen' sors ar[ instal' led at.various locations in the RB. The immediate operator action specified in the Alarm Response Procedure (ARP5671) is to initiate the.Suppr'ession Pool Leakage Return System. The re-quired subsequent hctions are:

i. Moriitor RB ster level to determine approximate leak rate. Use' sum'p alams to supplement the infonnation obt'ained from.the above instruments to ascertain the' approximate location of the' leak.

I' '2. honitor parameters (such' as iline pressure and flow rate) of the safety ~ ' systems as a ' leak woul'd affect 'the system' parameters. Isolate,the source of leakage per procedure listed below in Step 3. 3. If' required and plant condition: permi,t, dispatch an operator to the RB. floor to visually' locate the so'urce of leakage. Isolate.using the ap- .propriate system procedure listed below.. System. HRCI, Procedure No.SP23.202.01 Abnormal suction or. discharge piping pressure,. L'eakage indication: i-.- Excessive HPCI Loop

  • Level Pump Flow or low dis-cha'rge pressure.

~ e w-.--- ---ee n y p m-g y-y + y y. ow en. ~-,- y m

1 __...._... _. m._-- e ~ 3-8 . Reactor building sump high water levels in vicin-ity of leak. I . Reactor building' flooding alarm. Leakage isolation: . If 1,n standby, isolate the HPCI system by secur-ing the HPCI Loop Level Pump &nd then c1',osing CST Suction Valve (MOV-031). . If the system is operating, secure per shutdown procedure and then isolate as 'dese,ribed above. RCIC, Procedyre No.SP23.119.01 - Leakage indicatio'n: . Ahnormal suction or disch&rge piping pre'ssure. . Excessive HPCI Loop Level Pump. . Rqactor building sump high water levels. . Reactor building flooding alarm. . If in standby, isolate.the RCIC system by securl Leakage isolation: ing the'RCIC' Loop Lev 51 Pump 'and then closing CST Suction. Valve (MOV-031).. . If the system Is operating, secure per shutdown 'procedur'e 'and then isolat'e $s -desc'ribed.above. , s RHR, Procedure.No.SP23.121.01 Leakage indication: . Heat exchanger servi'ce water side temperature in'onsistencies. c ,.' Abnormal RHR system flow for mode.of operation. .. Abnomal RHR system' pressures for. mode of oper-

ation,

. Reactor water leve,1 inconsistencies for mode of operation. . Sump high. level alar,ms. g . Reactor building flooding alam. Leakage isolation: . Isolate the leakage by shutting down the affected loop in accordance with.the appr.opriate procedure .e 8

i 3-9 for the mode in which it was operating and then systematically shutting valves to isolate areas of the system found above to be possible sources - of 'l eakage. The above isolation procedure'may. require inter-mittent oper'ation of the leakage return system to. Jobse'rve the, effects on water buildup. When.the leakage has been isolated return the un-affectedportions'(asrequired)toservice. BNL has found that SNPS alarm response, procedures specify general gui.delines for monitdring. system parameters for determining the leakage loca-tion and for initiating the Jeak. age isolation. However, the proce'dures faji . to include specific requirements'for operators to systematically check the operati.on' paracaters of relevant systems. Since there are m'any' system para. j meter indicators in the control room,, the operators may possibly' fail to ob-serve the. indication of an abnomal' system parameter. When the abnormal condition is severe enough to actuate the alarm of a particul'ar sy' stem parameter,,the corresponding -Alarm : Response Piocedure will then be followed by.operhtors.. However, BNL. has reviewed.the relevant'. Alarm, ~ Response Procedur.es for abnomal system parameters, and found that these procedures do,not contain steps that should be followed under 'RB flood con-ditions.' Th'ese procedures provide guidel'ines for con'ditions other than RS.',. flood, such as* water source hbnormal or iso'laYion valves abnormal, et.c. The operator responses to the flood could be ' delayed or confused when these Alarm Response, Procedures are followed. 3.2.2 Requanti fication ~ l .i l* T'ha revised initiator frequencies a're. applied for e'valdating thk sequence frequencies of the initiator event tree. In addition to the critical flood- - depth of 3'-10" used by SNPS, BNL also evaluated the sequ6nce frequencies cor,- respondiitg to flood depth of l'-10" and l'-3". This is because, es indicated in Table 3.2.1, flood heights of l'-10" and l'-3" will disable several vital. e' e,. w. -,7 w,, ..m ..,.~..-.,_..o ._, m __7.,,, m. -_o,_ ,y..,., .y._.,.,.,#_.-,,-.,.~,,~.,..w.

y .,7 3 4 systems such as HPCI and RCIC. The times for the flood to reach 3'-10", l'-10", and l'-3" depth were calculated based on the leakage flow rates de-termined in SNPS PRA. The calculated times are shown in Table 3.2.2. The HEP, values used.by SNPS are identical to the nominal HEP values .' ' provided in'the Probabilistic Risk Analysis Procedure Guide (7) (see Figure 3.2.1 and Table 3.2.3). BNL feels that the HEP could be higher than' the . nominal HEP values because.'the* flooding alarm related procedures f, ail to '. pro' vide specific guidelines to. identify and'to' isolate the flood source,(se'e Section 3.2.1). - s Tfie HEPs under the multiple alarm and the singl,e,alam conditions are listed in Tables 3.2.4 and 3.2.5. 3.3' BNL Review of Functional Event Tree 'This section is. divided.into three su'bsections. Section 3.3.1 provides a. ~ qualitative' review.of, the Shoreham'In' ernal Flood event tree analysis and Sec ' t tion 3.3.2 presents the BNL revised time' phased' event; trees.',Section 3.3.3 describes the res,ults obtained from th'e quantification of the BNL event trees.

3.3.1 Qualitative Review..

Ib general, BNL'is of thel opinion that the methodolo.gylused in'the ~ Shoreham Internal Flood Analysis'i.s consistent with that of the st. ate-of-the.-art and the approach is reasonable. The analys.is for th' inte'r-e nal,' flo'6d postulated much severe scenarios'than those 'of the Shoreham FSAR. The' Shoreham Iriternal Flood functio'nal eye.nt tree' a'nalysisis ' based ' predominantly on the event trees developed for the internal. event initiators, namely, turbine trip, MSIV closure and mandal shutd'wn. These internal flood o functional ev,ent tre.es only model flood scen.arios whe,re. t,he floo,d water height' i -at Elevation 8' exceeds 3'-10". While 1,t ap' pear's that thh Shoreham. functional i .,eventtreesdoprovidearepresenta1iivemodelingoftheplant' response, itis f ~ not well substantiated that floods.that are arrested' before reaching 3'-10". .will result in negligible core vulnerable frequency. f Oe f + .~,-.---,,_--mmm-, ,,,,,--_,_,-....._v..,.

3-11 Table 3.3.1 enumerates the vital equipment that has been identified in the Shoreham analysis. The components that are located 'at the lowest elevation are presented first. It can be seen that at the l' level, both the RCIC and HPCI vacuum pumps and condensate pumps are expected to be disabled. However, it is' judged that thejr failures.do not lead to the failure,of the respective high pressure systems. Sinfilar argu'ments apply to the loop level pumps of' the low pressure core spray, HPCI and the RCIC system's as well. At app'roximately 2', instrumentation for both high p,re'ssure injection systems are submerged and hence.resulting'in failure of both sy' stems. ' At; 3'-10" ins'trume$tation for }', 'both LPCS and,RHR is submerged.. lea' ding to the failure of-(hose low pressu.re systems. In the Shoreha.m analysis the critical height of 3'-10" is selected. Hewever, since both HPCI and RCIC have failed at about the 2' level, these scenar'ios,wi.th termination of the flood prior to '3'-10" may not contribut' e an insignificant ' amount to 'the c' re vulnerable frequency. In the BN'L' revised o eve'nt trees, a time-phased approach is used to include the contribution from ' flooding below the 3'-10" level. ' Another area of concern ' stems from the treatment of propagation of failures in the Shoreham anal,ysis.., As noted in Table 3.3.1, at 'the l level,, 4-480V ' pumps are expectgd to e,xpe,rience, electrical shorts. The, Shoreham en, alysis did not -investigat,e,any, cascading failure which may. result ffam the ' electrical shorts.. BNL reviewed the el' ctrical drawings and elementary e drawings for so.me of the systems. It appears that for. each pump there'is only one electrical breaker which separates it 'from the rest of the loads in the same moi!ar control center (MCC).. Random failure of this breaker to open could result in the propagation of the short circuit fault upstream to.the MCC, ' other MCCs and the load ce'nter. BNL's review of the electrical diagrams indicates that failure of the breaker to open will result in tr,ipping the breaker at the Ioad center. D.iscussions with Shoreham engineers suggested,.. that there may possibly be an additional breaker per pump that is in, series . with the first breaker. However, this was not confimed by BNL. In the BNL,. revised event trees, only one breaker is assumed and its failure ij modeled, ' explicitly. BNL 'did not review.the spraying effects du'e to water cascades from higher elevations. 4

~_. i 3-12 3.3.2 BNL Time' Phase Event Tree The determination of the time periods which are critical to the con-sideration of the progression of the flood is based on the vital ' equipment .locatiort list (Table 3.3.1). Three heights were selected for,the BNL ar}al-ysis: at the l'-3" level, at the l'-10" level, and at the 3'-10" level. If the flood is terminated prior to reaching the l-3" level, no impact is as- ~ 'sumed for any equipment and the plant will'be shutdown, t'his is, Phase 1. How-i ever,.1.f.the flood wa'ter., exceeds the l'-3" level but is termin,ated before the. l'-10" level,'this is Phase II. Phase. III entails'the fa.ilures 'of bqth HPCI and RCIC system as well as the loss of power to the MG set recirculation pump.- '. fluid coupler before arresting the flood below the 3'-10" level. Any flood level which e.xceeds the 3'-10" level, it is treated in Phase IV. The e. vent trees'of these four, phases are presented in Figures 3.3.1 th' rough'3.3.4. Given that the flood is terminated in Phase I, BNL assumed that the reactor has a hi.gh probability (0.9) that it will be manually shut - down. Ten percent. of. the tiime.. it may result in a MSIV closure event. These two branches of the Phase J event trees are transferred to the respective' 4

interna' event tree, Figure 3.3.1..

l Figure 3.3.2' depicts the Phase If furictional event tree.which considers ~ the various ace.fdent mit.igation systems. Moreover, owing to thi fact that a number of the 480V pumps will be flooded,.the possibility of a breaker failure to isolate.i;he fa' ult is also evaluated. It.is assumed.t. hat the breake V , failure to open probability is 1x10-3 and there ar'e.a total of.five pumps in Division 1 and two pumps'in Division II that will be short circuited. probability of 0.5 is also assumed that failure of a load center in a division-would lead to failuie of other equipment. connected to that division. In the l event of a MSIV ct'osure, the feedwater system is considered to be! unavailable. I '. i The probab.ility that the reactor will be manually shutdown is also assumed to be 0.9 for the'maintena'nce induced flood events. Figure 3.3.3 illustrates the functional event trhe used to describe' the Phase III events. The major difference between this event tree and the Phase 4 f Dep ...n., ,,-.--..n,-,, .,-.-.,.,,.n.,

. = 3-13 II tree is the high pressure systems. In the Phase III events, both the RCIC and the HPCI systems are unavailable due to the failure of respective instrumentation. The probability that the reactor will be manually shutdown is assumed to be 0.5 for the maintenance inducea flood events. The Phase IV event tree'.is presented in Figure 3.3.4. This tree is drastically different from the other ones in that it only~ considers the feedwater system, the depressurization function and ths PCS. Al,1 the other, systems are disab. led due to flooding. The likelihood that the reactor will. be manually shutdown' is the same'as in Phase III for maintena'nce-induced ficods'. ~ 3.3.3 huantitative Analysis Based on the development of the revised flood initiator frequency', the BNL ime-phased event tree and the modified human response,to arrest flood, quarititative.results are obtained.. In the BNL an.alysis, there are 1.7 t different flood precurs' ors. Similar to the Shoreham classif.ication', ' he first! five precursors are online maintenan'ce related; the, remaining. twelve of them .are rupture related. A detailed discussion ch the BNL. flood precursors.is' given in Sectiori 3.1. Owing to the ways that these flood precursors; ahe calc'ulated,'.the. ini-

tiator event t'rees' have'been. modified tio 'inc'lude only th' ee functioris
' the'

~ r flood alann annunication, I; operator action to isolate flood, A;' and reactor l status. The entry con.dition to the different ti.me phase. event trees is deter, mined'by the A function (see Sectio'n 3.2 for details). Each of the 17' flood pr'ecursors were'.evafuated with 'the initiator event tree and the four time phase event trees. The una'vailability values. for the various event trees are the same as those,used in the Shoreham analysis exce;:t as noted in the last section. ? .) .t When the time phase event trees were quantified for the 17 flood pre-cursors, the results are the conditional frequency of core vulnerable given the particular flood precursor. These frequencies are summarized th Table i l 1 l a m .m.. e-,-. -~ -,--,- ,-nw,

...---..-a-3-14 3.3.2. The seventeen precurso'rs are listed as rows while the four phases are shown as columns. Within each precursor, contributions from manual shutdown, MSIV closure or turbine trip are also shown. For instance, the conditional 4 1 frequency of core vulnerable with operator arresting the flood prior to 3'-10" but after } '-10" ' Phase.III, for TFL1 is '2.0(-5) gi.ven the reactor is' manually shutdown. However, if instead of a m'anual s'hutdown, the plant, - experiences a MSIV. closures then the conditional frequency is 8.5(-4) As expected,.the conditional. frequency consistently increases as the flo'od progressestohihherelevat. ions. In, oth'e'r words, the conditional ' frequency of Phase I.V is always larger than any of the other phases., Another noteworthy - observation is.the unusually large conditional frequency of core vuine'rable for the LPCI system induced flood, i.e., TFL4 and TFL8. The TFL9 and TFL5 values,are also large since they disabled the LPCI systems as well.

  • The ' core. vulnerable frequericy given the BNL revised flood precursors,'

initiator event trees anc' time phase event trees is shown"in Table 3.3.3. In a this table, the 17 precursors are depicted on the left with the 4 phases de-picted as columns. Each precursor also identifies the contr'ibutions from t'he, vario'us. plant states. C' ore vulnerable, frequency contributions from Phase,I and. !!.ar'e very. smal1,' i.n..the qrder of 10-9 Contrib'utions, from, Phase'III' are not insignificant,but not substantial,'approximately 10-6 Sev.enty per-cent of the total core vulnerable frequency (70% of 2.0(-5)) is attributable ~ The mai,tenance, con-to LPCI system maintenance or ruptore induced flopd. n ~ tribution to flood is about.37% while the balance is due to rupture. 'It appears al'so that failure to properly model the fault propagation of i i the short circuits through the breakers does not have a sign'ificant effect'on core vulnerable frequency. . i l, I 3.4 Uncertainty'Es'timates This section presents a limited uncertainty assessment on 'the BNL quantitativ,e analysis for the core vulnerable frequency due to reactor. bu'ilding flooding. 4 0 0 } g -9 t _._,r,,s-. .,-,-,n-,-,-e,e-,,-, .-n-.--<,,-..-- - - m-- en n-,, --,--,-,n.-,w-

= _. _. - - a 3-15 A rigorous propagation of the uncertainties is outside the scope of the present review. The BNL approach for the uncertainty evaluation consisted of the following general steps. 1. The uncertainties. in the human errors as well as the split ratio be-tween'the manual shutdown and the MSI'V closuie event were quantified by fitting lognonnal distributions to evaluate uncert.ainty measures (meari and var.iance). An error factoi of 10 was appli,ed to human er-rors and the spli,t ratio.' 2. Human erro'rs of t' e, following operator actions were included for the. h uncertainty evaluation: Operator maintains isolation valves in closed position.during the online maintenance (Event E, see Section 2.3). Operatoi diagnoses and re'sponds to, isolate the. flood (Event. A,- see Section 2.3). Operator depressurizes. the Reactor Pressure Vessel (Event X, ~ F.igures 3.3.2-3.3.4). 3. Th'e u'ncertainties!In the core:vulnerab1'e frequency werel evaluated us- 'ing the mafor accident sequences. and 'thh distributions ' ass'ess'ed 'ih Step 1. .The SAMPLE code was used for the estimaton of. uncertainties. The mean, c'

median, 5%.and 95% probability intervals fo.r the core vulnerable frequency are shown as follows.

.1.91E-5 Mean = ~ 1.90E-6 Median = 5%Confjdence,= 2.19E-7 g 95% Confidence = 7.51E.-5 e e e 4D g 6 4 ,o ..y

3-16 1 UP x n MAINT*. R X x-4 l

  • D SHUTDOWN u

t oown. y u +. y J. .p, N ' /* X BREAKER.- / OUT ' MAINT. ..j R ON-g ' 3' 'LINE BREAKER u Jr .., IN

x. x 1: Comp. Failure Rate

.v: Comp. Repair. Rate ,y A: Transient (MSIV clos.) Rate' D ' P: Prob. of not racking out Breaker ' FLOOD INITIAT-A: Rate of inadv. operation of' 5 o t ED

switch, A0T: Allowable Outage Time i

i i Figure 3.1.1 State Transition Diagram, for Component-Maintenance Induced Floods.. 6 e e e f w ee s-w wg.eW,-, y,--q- ,,ww-c----c-,eg -y, .-e

gy-.g

_4-w--g+ --+-e e e t 7*-yatr--+ mew sw--

1 ___. 3-17 .s I 1 i UP j A. R %f RUPTURE 1 2 .VULNER- ~ i ABLE s s s \\ A A N S T M g \\ .g v + S FL600., FLf00. ~ FL600 I '3 'MS.IV CL ' 4 5 l l l TURB. T MANUAL SHUTDOWN i g Figure 3.1.2 State Transition Diagram for Rupture-Induced Floods. .... t

3-18 1

7 s

- s s \\. s s s. s, \\ . s 3 ~ 10~ \\. s \\ g \\- s q ~ Up;ter' Sound 10.. c ~g

  • um.

.\\ ttominal. l'0" Value .\\. / g ,.g l.cwe. 5cund. llN si r N,. g. ,g N 10<3 1 10 100 1000 tiinutes. ~ Fi gure,3.2.1 P.ro'blem-solving human error probability vs time ' screening values. .,6 .l l .s .t a. O o

F = .l g i i .j t Phase'I' t l I . ~. e i i i l t . i 1 i 1 t R. Stat.' i i = l. i I .i i i g [I . i g i i l-j. l 1 i 8 l l t i i e. i, i. e .-i 8 WuluAl. i 1.. i i ] 4 1 '. Transfer to E/T j .i .6 .l .[ l l l l-t } } i. l I i i r i j I i j . i 8 j i 's .} I i r i t i i l-i i } e t i i t e i 8 !w ' 4 i i I in I -i j. i l } i i i ~ 8 e i I I l I j' I g j i. g i-i t l l i MSIV t- ' :Tiansfer to E/T L l i l -l I- .i I l } j 1. i i g i 1 t,, g 1 i 1-l 8 I i. l .C i' 1 i ..I I 1.I i g. 3 6 .i l- .i 3 . i. i. l t .t [ i

1..

t, t ( f I l l. t l 1 i i Figure. 3.i.1 Phase i of litterna'l FloodiFunct'onal Event. Tree j. s i ).i 2 g i I s; I g g' t ,l [ i I 8 i s ' - I g. t { 1 .e .e l i l l' I. f i i i a P g g' l 8 I 1 1 l 1 1 f

j l l W Z 1 - l i-

  • V

' V.j 6 r. i j' i U.' i R !C M P. Q Divl Div2 U CI V 11, X cond R , Stat. BKR BKR ,P .j CS 2 ~ R i 1 t I l l

6..

I t 8 g 4.. l g - ; .l l i i e '.- l.. i t l i I U r i. l i i i i l i l 1 t l l i i l ,l-l . I, 3 g 3 t i j i g-t 1 l~ ~ I I 3 l } I- .i ly I c m M. l i \\ l 1 l ] . g.. . -i.. j - i 1 s .i l.. .l . i 1 I 4 !t i .i 1. .i I -t l j Fiigure 3.3.2 Phase II of Internal j l-i Flodd Functionali l i J Event Tree ) j i --), l --.+ 1 --l i 4 l .n i l l l i i

L' .c IM AGE EV ALU ATION g TEST T ARGET (MT-31 KN w c V ~, s-c ), 3 sg ) 4 l.0 '1 1.l l l '.'.$ l.: i c, Ik lIllII

  • 4

>i g, d'\\ \\' g'q -) a ,s x N \\\\' 6's-(\\ '/

  • P t

'Gl

i.. !

.i. - t-n 4. R C' M P .'Q Divl Div2 U ' U.X Y 'V V ! W ' 2.'W i i Stat.- BKR BKR R

11. -

CS CI cond R. P - i 1 3 ? i I g i .I [ i I t l. i l 1 i. -1 l. I I .? r. e l i l l i

  • include prob.

l l i l l of losing the 6 i l l division. i i I i 1 i i l i I,. i l ? I j .t i s l i I 8 l t i I i. I i i 1 3 i l ..- I t - 1 -l i i I. i i i l i w i l l e I ,i l.l. i. .i i. \\ n i i t g i 8 . \\. i + I i i i i j .i ~ l n 4 1 i, l. ( I l i l 1 I l l l l i .i i i. l - l 1 l -l l i ~ ii l j j I -i. 8-Figure 3.3.3 Phase III ofj Internal l . l Flood Functional. Event I i - 't r-Tree i i ) l 'j@' h li l l. I. .i . i.. .l i. ..I a ! I e i i s .i l l

i..t a

s l l I

1 l i 1 ^l I i 1 8 I i b g l i C' M P i ~ X V3 Z-con'd W i i l l PC '- i l l l i 1 i l i u i l l l 8 i e ] l l l-i o l l i \\ 1 .s l Y j ... i m m m i I l l j j

l..

l 1 l l l i 4 l -- 3 l. .l ~ ', l -1 I t g e s i i I 4 l -t i I i-i l i l i l l l Pha.se IV of' Interria' \\ i Figure 3. 3.4 Flood Funct.ional Event Tree \\ 1 i e 8 8 t 8 I 'e ,; 6k

3-23 Table 3.1.1 LER Data for BWR Standby Pumps for the Period of January 1972 Through September 1980 Does Not Standby Standby Leakage Does Not -Loss of Continue Pumps Demands Hours ~ Rupture - Start Function To Run Motor , Driven 20,321 10,453,806 9 8 8 9 Turbine. ' 23 34 '. l 25 Driven 2,860 1;439,491 Table 3.1.2 Frequency of' Maintenance - Induced Flood Precursors - ~ System Initiator Event Trees Probability per Year. TFL1'P'.D . 1.05x10-4 ' 1. RCIC - TFLI P.Eo 2.10x10-5 2.10x10-5 TFL1 P.E' L TFL2 P.D 1.05x10-4 2.' HPCI ' TFL2 P.Eo.j 2.10x10-5,. .,2.10x10-5 TFL2 P.E. L 1.89x16-5 3. Core' Spray TFL3 P.D. (2motordrivenpumps) TFL3 P.E 1.87x10-6 ~ o 3.78x10-5

4.. LPQI '

TFL4 P.D '(4 motor' driven) T'FL4. P.E 3.74x10-6, o 5. Service Wate'r TFL5 P.0 1.89x10-5 (RHR or RB(LW HX) TFL4 P.E 1.88x10-6 o . 2 motor' driven pumps .i f. O -e n y

.....=.. -:uu~ 3-24 Table 3.1.3 Flood Precursor Frequency. Pipe Valves Pump Total AR Ps PT Pg TFL'6 1.2(-9) 6.5(-9), 0 ,7.7(-9) 1.6(-5)' 1.7(-5) 1.5(-5) TFL7 2.0(-9) 1*3(-8) 'O 1.5('-8) 3.1(-5) 3.4(-5) 2.9(-5) ' TFL8 3.7(-9). 2.9(-8) 0 3.2(-8) 6.5(-5) ' 7.3(-5) 6.2(-5) 2.'(-5) 2.5(-5) TEL9 1.1(-8) 2.3(-8) 6.0(-10) 1.3(-8) 2.6(-5) 9 TFL10-2.4 (-9')

1. 3 (-9)~

0 3.7(s9). 7.5(-6) 8.4 -6) 7.2(-6) ( TFL11 1.1(-9) 9.1(-9) 1.5(-10) 1.0(-8) 2.1(-5) 2:4(-5,)' '. 2.0(-d) .TFL12 1.4 (-9) 3.9{-9) 1.5(-10).5.5(-9) 3.7(-6). 4.0(-6) 3.6(-6) 7.3(-6) 8.0(-6)- 7.1(-6) TFL13 TFL14

1. 9-(-9) 5.2(-9) 3.0(-10) 7.4 (-9) 5.0(-6) 5.6(-6) 4.'8(-8) 1.0(-5),

,1.1(-5)- 9.6(-6) .TFL15 TFL16 1.9(-9)l,5.2(-9) 6.0(-10) 7.7(-9) ' 5.2(-6) 5.8(-6)- 5.0(-6) 1.0(-5)

1. 2,(-5 )

1.0(-5) TFL17 9 6 8 g 8 0 e e e O e e e 4 0 9 t O S e 6 D e-O e 9 -- ~. - -

3-25 1 4 Table 3.2.1 MAZR ELEVATION 8 EQUIPMENT LIST ~ POSTULATED EQUIP. EQUIPMENT DE5CRIPTION P ART NO. DISABLED TWE HEIGHT" euMPs ~' , Floor Drain Sump Pumps 1G11*P-035A-0

  • l'-0" 1G11*P-036A-F 0.ry F1'co' Drain Tank. Pumps 1G11*P-151A,3 l'-0" r

Radwaste Equip. Drain Sump & Fump to Porous Sump 1G11*P-224A,3 'l'-1" l HPCI Pump, 1E41*P.015' HPCI' Vacuum Pump 1E41*P-075 l'-0" ' 1E51*P-076 l'-0" HPCI Cen. Pump. ' "' RCIC Pump 1E51*P-015 .RCICVacuu$iPump' 1E5'1*P-076 l '.0 " RCIC Cen."Pum'p' 1E51*P-077 l -0 " RHR Pump' Motors 1E1I*P-014A-0 5'-4" Leakage Return Pump G11 *P-270. 3'-9", Core Spray.Leop Level Pumps

  • 1E21*P-049A,3 l'-3" Drywell Equip. 0 rain Tank Pumps 1G11*P-0332A,9 "l'-2"

~ RCIC Loop Level Pump. 1E51*P-051 l'-4" i HPCI Lcop Level' Pump ' 1E41*P-b50 2'-3" o lTURSINES HPCI Turbine 1E41*-TU-002 6'-0"

    • RCIC Turbine 1E41*-TU-005

'4'-0" ,r- .-,-e ,.c.,,,.-,,--------,-,--g-,~ - - -, = - -.

r

4 t 3-26 l ~ Table 3.2.1 (Continued) MAJOR ELEVATION 8 EQUIPMENT LIST POSTULATED EQUIP. EQUIPMENT DESCRIPTION PART NO. OISABLED HEIGHT' ~ TYPE MOTOR CONTROL CENTERS" ~ Sumo Pumos and Cooling Water Pumpt to Recirc. 1R24-1101 l'-6" Pump NG-Set, Fluid Coupler 1RZ4-1201 l'-6",' TiNXS' Floor Drain Sump Tank 1G11*TX-050A,8 1G11*TX-056f-C Orwell Floor Drain Receiver 1G11*TX-057 Salt Water Or'ain Tarik 1G11'TX-190 Orwell Equip. Drain Receiver l'G11*TX-049 ~ HEAT EXCHANGER HPCI' Barometric Con. Vacuum Tank 1E41*E-036 l RCIC Baremetric Con'. Tank 1E51*E-038 IE114*E-034A.S, i RH.R Heat Exchanger, 3 1742*-011A,3 RSCLCs Heat Exchangers Orp e'll Equip. Orain Cooler 1G11*E-094 me e = ,r.a,.,em. ,+w- +,--+----,-,-v-mw-

__..._.._..__.__u 9 3-27 t!' Table 3.2.1 (Continued) MhJORELEVATION8E' QUI?MENTLIST f POSTULATED EQUIP. EQUIPMENT DESCRIPTION . PART NO.

DISABLED, HEIGHT' TY,PE EL'EC.

PANELS " RCIC Instr. Rack 1H21*PNL-017 2'-0" '** RCIC Instr. Rack

  • 1H21*PNL-037 2'-0"
    • Core Spray Rack 1H21*PNL-01

3'-10" '. ~ 1, 21*PNL-01'9 3'-10"

    • Core Spray Rack' H

E' ~c P:NELS 1H21*PNL-O'18 3!-10" " RHR Inst. Rack.;A 3',-10"

    • RHR Inst. Rack S.

1H21*PNL-021

    • HPCI Inst. Rack A 1H'21*P8L-036 I'-10"
    • HPCI Inst. Rack,B' 1H21*PNL-14 1'-10"

~ Equipment required for operation of the identified system. Heignts are taken frem a physical survey measurement frem the bottem of the. component to ficar level. ---, Non-electrical component i t G s t = i I

3-28 Table 3.2.2 Times to Flood Depth of 3'-10", l'-10", and l'-3" in Reactor Building Water Time (min.) to Flood Depth of System, Source Leakage Location 3'-10" l'-10" l'-3" ~ ^ HPCI S.P.- pump suction (max.) 17 7.9 5.4 . S. P. pump s4ction large 34 15.8 10.8 CST pump suction max.) 13 6.4 4.4 CST pump suction (large) 27 12.8 8.7 ~ 11.9 37 17.5 pump discharge RCIC S. P. ,pumpsuction(max.). 110.0 50.8 34.6 220.0 101.6 69.3 S.P. pump suction (large) CST pump' suction (max.)

  • 76.0 36.3 24.8 CST pump suction (large) 152.0 72.6 49.5 -

LPCI S. P. pump suction'(max.) 9.4 4.5 3.1 pump suction (large) .'19.0 9.0 6.1 S.P. pump discharge 15 7.3 5.0-l CS. S.P; pump suction (max.). .12 , 5.9, 4.0 S. P. pump suction (large)- 24 11.8 8.1 13 6.4 4.4 pump suction (max.)) CST pump, suction-(lirge 27 12.8 CST '

8. 7.

23 11.1 7.6. pump discharge - SW SW RHR heat exchanger, 20 9.5. 6.5 40 19.1 13.0 ,ruptureof8,] pipe!, WFPS WFPS . Note:

1. Large flow rates' is 1/2 of ' maximum flow rates."
  • 2. Flood tim',s weire calulated based on a 41,600 gallons per foot depth e

in the reactor building. 3."S.P. = Suppression Pool CST. = Condensate Storage Tank 'SW = Servi'ce Water System WFPS=Waten{ireRrotectionSystemTanks g e b

1 3-29 Table 3.2.3 Human Error Probability: Screening Values -Problem-solving ' Time Nominal Value Errer Facter* <1 min. 1 10 min. -5E-1 5 lE-1 10

  • 20 min.

10 .30 min.' -1E-2 60 min. IE-3 10 30- ~ 1500 min. TE-4 Precedural-Ereces- !!c=inal Value-Error Fac::r 3 IE-3 (*Jith P.eco.very) IE-2 (Ulthout' P.ec:very) ~ 3 G e s g e ,S O O S j j g l D e e e = t a.es-e s-- -e -. =.. -..,., ,-e y,_ ...w.g--,-9. m ,g ,,--q 7-wp. e_,,.-,7w. ,-.g-,-y y - - - +, .y9y-- y

- us _ t-3-30 Table 3.2.4 HEP (Event A) Single Alann Condition Manual Shutdown (Nt' REG /CR-1278) l'-3" l'-10" 3'-10" T'FLI. 10-3 10-3 2.0x10-4 1 0.1 TFL2 1 TFL3 1 1 0.1 1 1 1 WL4 1 1 '10-2 .TFL5 T$L6 0.1 0.1 10-3 ~ TFL7 1 0.1 10-2 TFL8 1 1 0.1 1 1 10-2 TF.L9 10-3. 0.1 0.1 T.FL10 10-3 '10-3 2x104 ~ TFL11 TFL12 1~ 1 0.1 .d.1 O.1 3 TFL13 1 1 0.1 TFL14 NL15 1 0.1 10-2 . TFL16 1 l' '. 1 TFLI.7 .1 ,1 0.1 5 g ~! l t } ? - ' I t s s O 6 6 0

-.' o'.. 3-31 Table 3.2.5 HEP (Event A), Multiple Alarm Condition -(Nominal Value, PRA Procedures Guide) I'-3" l'-10"' 3'-10" TFL1. 10-2 1,0-2 10-3 TFL2'. 1 1 0.5 0.5 TFL3 1 1 I 1 . r TFL4 061

  • TFL$'

1. 'l 10-2 . TFL6 0.5 0.5 TFL7 1-

0. 5 -

- 0.1. TFL8 1 1 0.5 TFL9 1-1 d.r .0. 5' ' O.5 1012.

  • TFL10

~ TFL11 10-2 10-2 . 10-3 .^ 1 1. 0.5 TFL12. 'TFL13 O.5 0.5 10-2 TFL14 1 1 0.5 r 1 0.5 .0.1 .TFL15 21' ." TFL16 I '1 .'TFL17 1 1

0. 5,,

8 8 8 lg g s

o.. 3-32 Table 3.3.1 Vital Equipment 1.ocations at Elevation 8 l' 'HPCI vac. pump -

  • ~

cond. pump RCIC vac. pump ' cond. pump l'-3" l'-3" CS loop level pump l ' -4 " RCIC. loop pump *., I'-6" recir, pump M-G. set I'-10" HPCI instrumentation l'-10" . 2' R'CIC instrumehtation 2'-3" HPCI loop level pump 3'-10" RHR instrumentation t 3 -10 CS instrumentation 0 5 t e e e g 4 , g 6 9 s 0= + 9 0 g i e S S 8 e 9 , g g 0 4 8 e e e 9 -g S 4 e D 4 e 9 8 b I'l )- 9

  • 9 e

D e 6 e 6 e e e e e 9 8 a e e e 8 m...,.

,o.. 3-33 Table 3.3.2 Conditional Frequency of Core Vulnerable (1 of 2) Phase I Phase II Phase III Phase IV TFL1 Manual 3.2 - )ll 5.8 - 2.7 - 2.0 - 7.3 -3) MSIV 8.7 - 8.5 T.2 -1.) TT 7.7 ,1 2.2 - 2.1 3.3 -2) i TFL2 Manual

  • 5.8(-7) 2.2(-6) 2.0('-5) 7.3(-3)'

MSIV 3.2(-6) ,6 4(-5) ., 8. 5 (-4 ) 1.2(-1) -TFL3-Manual 5.8(-7) 1.1(-6) 2.2(-5) 7.3(-3)

  • MSIV 3.2(-6) 1.1(-5)
9. 5(-4 )

1.2(-1-) ~ TFL4 Manual

  • 5.8(-7) 3.9(-4 )
5. 2(-4 )

7.3(-3)- MSIV 3.2(-6) 2.0(-2) 2.6(-2) 1.2(-1) TFL5 Manual 5.8(-7) . 3.9.(-4 ) 5.2(-4 ) 7.3 (-3 ) MSIV 3.2(-6) 2.0(,2).,

2. 6.(-2) 1.2(-1)'..,

TFL6 Manual - 5.8(-7) 2.2 - 2.0( 5) 7.3(-3) MSIV..

3. 2'(-6),

6.8 - 8.5( ) 1.2(-1) TT 7.7(.-7) 1.6 - 2.1( ) 3.3(-2) 5.8(-7) *! 1.1 -6)

  • TFL7 Manual 2.2 -5).

7 3 -3) MSIV 3.2(-6) 1.1-5)-

9. 5 -4 ) ~ -
1. 2 -1)

, ".. TT. .. 7.7(-7). .3.2-6) .,. 2. 3 -4 ). 3.3 -2).. TFL8 Manual 5.8(-7) 3.9 5.2 7.3 -3)* _3.2(-6). 2.0 - 2.6 - 1.2 -1) MSIV ,, ' TT .7.7(-7). 4.7 -

6. 2..

'3.3 -2) ' '5. 2 TFL9 Manual 5.8(-7) ' 3.9 1.25))' .7.3 3 MSIY " 3.2(-6) 2.0 - '2.6 - 1 TT '7.7(-7) 4'.7 - 6.2 - 3.3 -2) TFL'10 Manual 5.8(-7 1.1(-6 2.2 - 7.3(-3) 3.2(-6 1.0(-5 9.5 1.2(-1) MSIV l . TT, j 77(-7 3.2(6 , 2. 3 3.3(-2) TFL'11,Same as TFL1 TFL12 Same as TFL6 TFL13 Same as TFL6 .e e

. _ ~.... 1 3-34 Table 3.3.2 Conditional Frequency of Core Vulnerable (2 of 2) Phase ! Phase II' Phase III Phase IV TFL14 Manual 5.8 -7 1.1 2.2 -5 7.3 '3 MSIV 3.2.-6 1.1 -5

9. 5 -4 1.2 -1 "

TT 7.7(-7) 3.2(-6) 2.3(-4 3.3(-2) TFL15 Same as TFL14 ./ . TfL16 Same as TFL8 TFL17' Same as TFL8 i ..t f l

i

9.
  • o 3-35 Table 3.3.3 Core Vulnerable Frequency (1 of 2)

P-1. P-2 'P-3 P-4 TOTAL 4. 51 -1 1 )l 7.3(-11 0 1.7(-11) 1.2(.9). TFL1 Man. MSIV 0 8.2(-11)

1. 31' -8 )

.1. 2i;-10 ? I, 9.9(-11)

1. 5I,-8) 1.4 (-8)'

TFL2 ' Man. 0 0 9.1(-10). 2.1( I) ' MSIV 0 0 3.9(-81' 3.4 (-6). ,T-T. 4.0(-8? 3.6(-6)- 3.7(-6) ' TFL3 M.an. 0 5,2(9)10) ' 3.5(-8) 0 1.2 MSIV. 0.. 0

5. 8'(-7 )

i.- D I 5.,31, -9 ) 6.1(-7) 6.2(-7) TFL4 Man. 0 0 0 1.5(-7) 0 O

2. 5 i' -6 ),

~ MSIV 0 F, F 'F 2.71,-6. ) 2.7(-6) ~ TFL5 Man. ' 0 0* 4.9(-9) 6.9(-9) MSIY 'O 0 2.5(-7). 1.1(-7) E' ,I 2.5(-7) 1.1(-7) - 3.7(-7)

t TFL61 Ma n'.

0 1.'6 (-10)

MSIV, 0

1.41'.-8 ) 5.6 -8 E -)) .7.0(-8)' 7 3 1.4 i,-8) T'FL7 Man. 0 1.4 (-9) MSIY 0 3.9(-10) 2.6(-8).. ' 8.3(-7) i 3.9(-10) 2.6(-8) ','8.3 (-7 ) 8.6(-7) TTL8 Man. 'O 'l D' 1~. 5 (-8) 2.3 - '

  • MSIV
  • O 0

2.3(-7)6). 4.3 - 1.6 TT 0 0 L " 1. 2 - I' i

1. 81, -6 )

5.5(-6)* 7.3(-6) TFL9, Man., ,0 0, ' 9.2(1,-7 . 6.5 -9 12(.-9 MSIV' 'O,' O 3.4 (-7 TT' 0 0 1.8L-8 9.9(-8 3 V 9.41;-7) 4.4 (-7 ) 1*4(-6) ~

  • Lessthan1.0'(-10).

. ~ 't. 3 t 3-36 Table 3.3.3 Core Vulnerable Frequency (2of2) P-1 P-2 P-3 P-4 TOTAL .TFL10 Man. 0 MSIV O 3.9(s9) 1.4 (-8) 3.6(-9) 0 9.4(-10) TT I I' 4.8(-9) 1.8(-8) 2.2(-8) ,0 .TFL1'1' Man. - *-

MSIV,

,0 1.'6(-1,0') 1.4(-8) ~. 0 3.2(.-91,. TT 7 I 1.6(-10) l.7(-8) 1.7(,-8) 1.1(-9;l TFL12 Man. 0 0 MSIV 0 0 4.5(-9) 4.9(-7 l . I .I, 4.5(-9), 4.9(-7;l-5.0(-7) TFL13' Man. 0 0 ' 6.6(-9) 2.5(-8) 3.1(-8) MSIV 1.8 (,-9') TFL14 Man. 0 0. 7) MSIY 0 0

7. 3( -9 'i
6. 94 I

'I 7.3(-9;i

6. 9 L, -7 )

7.0(-7) 4.6(-10).., TFL15 Man; d. . 8.41'-9 ? ,1 . 2.6J-7), l 'MSIV 0

V 8.4 i,-9f
2. 6 L,-7 )

2.8(-7) TFL16 Man. 0 0 0 1.8(-8)' 0 9.2 f'-7ll MSIV O 0 TT 0 0 O 1.91 -71 T T-1.11,-6 ;' 1.1(-6) N TFL17 Man. 0 0 2.3 - ) 3.8(.9 - MSIV 0, 0 2.5 ) 6.6(-7 '2.0J-7 TT 0 0 3.7 Y I I

2. 9(-7J
8. 6 L,-7 )

1.2(-6) .n i, p 3,

  • Lessthan1.0(-10).

a e j j

  • =

6 ~,-..,,ym.. ...__.,,-,_..____,,_,,__,_,,.,,,_,__,_,,-_,,_,7- ,_..,,,,m

.2-ql e ' 4-1 4.0

SUMMARY

BNL reviewed the internal flood analysis which is a part of the Shoreham PRA and found that assumptions, methodology, and results are reasonable. BNL re-e' valuated the flood precursor frequency us' ng ricent LER' data and a more i accurate methodology. This r$ethodology avoi,ds some of the'.conservatisms in the SNPS-PRA approach. A slight increase in the. initiator,fr~equency is calculated'because of thi revised data. Similarly,' based on the PSA,Probedure Guide, the HEP was reviewed and only minimal changes were madii to the Shoreham HEP values used in the' analysis. As, for the functional' event trees, a time phase a'pproach was adopted to better, model the progression of the flood events. Results are, summarized in Table 4.1. This table 'can be divided into three ' arts. Part A provides a comparison between'the Shoreham results and those' p obtained'in the BNL review. The BNL v'aiue is about 5 times that of th.e.' Shoreham. frequency, 2.0(-5) vs 3.9(-6). The contFibutions from t.he differ'ent ' plant states are'aiso presented.' The major increase in the total core vulnerable frequency.in the BNL analysis is attributable to th,e. increase in flood precursor frequencies. Part B comp' ares only, the coritributions fr'om the. BNL Phase '!V results'with 'the' Shoreham values. 'It'can be* inferred that'by' neglecting the initial three phases, the core. vulnerable frequency will be underestimated by 3kl0'-6 or about.1BL Part C shows the ' contributions of - core vulnerable frequency for different plant states due to maintenance and rupture. ind'uced f1'oods. In the Shoreham analysis 41% of the core vulnerable - frequency is calculated to be caused b maintenance related floods while the BNL analysis shows 37%. An uncertainty estimation has been carried out assuming.lo'gnomal ~- distributions. An error factor of 10 was applied to the ' pe.rator errors and o the spli.t ratio for the manual shutdown and the MSIV closure event following p E 9 e 6 g

n. _.... c

9.
  • o s

4-2 the Reactor Building flooding. - The results of the uncertainty assessraent for the core vulnerable frequency are as follows. 1.9E-5 tiean = 1.9E-6 Median = 2.2E-7 5% Confideni:e = 95% Confidence = 7.55-5 t 6 I., .l t s-c 0 0 /.i .I' t t 0 g .-l 6

i 3 T.

  • 4-3 Table 4.1 Summary of Core Vulnerable Frequency Shoreham BNL Part A 4.8(-7)

Ma'nual ,8.5(-8) MSIV 3.0(-6) 1.8(-5) TT. 7.7(-7)' 2.0(-6) .3.9(-6) 2.0(-5) Total ~ BNL (only Shoreham Phase IV p B 8.5(-8).

4. 5 (- 7)'

Manuel MSIV

3. 0,(-6) 1.5(-5)

TT 7.7 (-7 ) ', 1.7(-6), , Total 3.9(-6) 1.7(-5)~ Shoreham - BNL Part C' Manual Mainte' nance 3.9(-8) '4 1( 7) Rupture - 1.6(-7)' 7.0(-8) '.9(-6) MSIV Maintenance 1.5(-6) 6 Rupture 1.4(-6) 1.1(-5) l t TT Maintenance , 0 0 P.up,tur e 6.7(J7) 2.0(-6) Total Maintenance. 1.6(-6) 7.3(-6) Rupture 2.3(-6) 1.3(-5)

  • 6 s

e o

y h. *, i \\ R-1 REFERENCES 1. PRA Shoreham Nuclear Power Plant, LILCO, June 24, 1983. 2. U.S. Nuclear Regulatory Commission, " Data Summaries of Licensee Event Re-ports'of Pump's at U.S. Commercial Nuclear Power Plants: January 1,1977 to , April 30,1978,".NUREG/CR-1205, January 1980. 3. W. H. Hubble, C. Miller, " Data. Summaries of. Licensee Event Reports of. ' Valves at U;S. Commercial Nuclear Power Plants," NURED/CR-1363,. Volume 1, Ju'ne 1980.. ' '4.. Swain, A. D., Guttmann, M. E., " Handbook of Huma'n' Reliability Analysis-With Emphasis on Nuclear Pow'er Plant Applications," NUREG/CR-1278, 1982. 5. Busch, S. H., " Pressure Vessel Reliability," Trans. ASME J., Pressure '. Technology, January 1975.. '.S. Nuclear Regulatory Commission, " Data sunmaries of Li,censee Event Re-U 6. ports of. Pumps at' U.S. Commercial'. Nuclear Power Plants: Janua'ry,1,1,977 to ' ' September 30, 1980," NUREG/CR-1205, September l'981. U 7.- "Probab'ili,stic Safety Analysis Procedures Guide," NUREG/CR-2815, September '1983. 8. " Nuclear Power Experience," Div,ision of Petroleum Infarination Corporation, Denver, Co.lorado, August 1981. 8 i >i ? s ? e 6 0 96 .}}

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