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' 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
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Risk Evaluation Group Department of Nuclear . Energy ,
Brookhaven National Laboratory . ,
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Upton, New York 11973 ., . .
. April 1984, -
Prepared for - ";
U.S.' Nuclear Regulatory Commission;- -
Washington, D.C. 20555 .
. . Contract No.DE-AC02-76CH00016 t
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. 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. . .
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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
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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
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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 . . .
2.3.4 '
TFL4 Initiator event tree fo'r postulated flooding *
- . .- 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
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2.3.12 Initiator event tree for postulated flooding sequences in.itiated by a maximum HPCI suction line break....... 2-20 2.3.13 -
Initiator eve'nt. tree for postulated flooding sequences ,
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 .
o E 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't ional event tr.ee........ 3-22 .
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~ . LIST OF TABLES Table No. Title s Page' 2.1.1 -
Summary of Potential ' Water Sources' and~ Types of
,) 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 9 2.2.3
- Sumary of Failure Rates for Major Components Involving Elternal Leak and External Rupture.....'.... 2-31" ,
Conditional ' Probability of Pipe Breald Size. . .. ... ..... . 2-31
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. - 2.2.4 1-2.2'.5 Initiat.ing Event' Frequency Est'imates' Involving -
4
. Cmpo.nent 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 osure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-34 -
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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,
'3.1.3 Fl ood Precurso r Frequen cy,. . . . . . . . .'. . . . . . . . .'. . . . . . . . . . . 3-24
.J, 3.2.1:
1 Maj o r El evati on 8 Equi pment Li st. l. . . . . . . . . . . . . . . . .'. . . . 3-25 -
.' 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 4
3.2.4 YEP (Event A). Single Alarm Condition Manual Shutdown (NUREG/CR-1278)....................................... .
3-30 .
' HEP (Event A), Multiple Alarm Co'ndition (Nomtral .
, 3.2.5
- Val ue , PR A Procedures Gui de ) . . . . . . . . . . . . . . . .~. . . e . . . . . 3-31 3.3.1 , ,
Vital Equi pment Locations. atil evation 8. .. . . . . .'. .. l. 3-32 .
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 -
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l 4.1 Summa ry of Core Vul nerabl e Frequency. . . . . . . . . . . . . . . . . . 4-3
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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'i ng warrants a special' review. A ~fiel d ' trip .
I I:0 the Shoreham plant has been made by BNL perso' n nel for o'btaining detailed ' -
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informa't ion' on 'the equipmen't and power conti ol la,youts in the 'RB, especially
- in the' Elevation 8 ccmpartment. BNL has ' determined that there are 'three flooding depths (:1'.D"; l'-10", and 3'-10". ) that are c. ritical, to .the 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).
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. 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
The report is organized as follows: Section 2 summarizes the SNPS-PRA ap- l 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. . '
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( 2.0 SNPS METHODOLOGY AND ANALYSIS -
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2.1 Overview -
The SNPS methodology for determining the contribution to the risk of the l
internal floods can be divided into ,three steps. .
.- 1. Identification of water sou'rces and pathways to Elevation 8 com-
. pa rtment. .
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, 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 .
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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
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compartment, with aT1 equipmentna' pd ' iping insi:alled, is .estimat'ed to be .
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. ;. -
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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
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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..
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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' CS equiproent would.be disabled. Flooding 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 o PNA to
.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).
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l 8ecause'of these' depehdences,'the end states 'of the ini~tiato'r e vent trees 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 .
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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.
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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. .
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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 '
1 - -
f ailure mo' des. ' - - -
The following LER rates are found for the four failure. modes in .
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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
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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
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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 ,
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, - _ . . - . - . - - - - , _ . _ - _-.--.,.-m,,.,~ . . . - - - .....-,,,,.-._,--...-..--,.,-.~,-.__,-,-,-.----4..---,~_ ,.
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' fating' event, frequencies .
. 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' 'd fr,om their associated power supply by pulling and tagging the' -
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 AA, ). Multiple alanns will occur in the control room at the same time as the flood alann. ,
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- 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) ,
E
'are applied by SNPS for their ' evaluation Figure 2.3.18 and Table 2.3.1 show the time var.ving cumulative, HEP ,for
- b.oth the, single and the -
inultiple occurrenct conditions. , .
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j Figure 2.3;.11 Initiator Event Tree for Postulated Flooding Sequences
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g Figure 2.3.13 Initiator Eveiit Tree for Postulated Flooding Sequences Initiated lay a large llPCI Suction Line Break
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. ' figure 2.3.16 Initiator' Event Tree for Postulated Flooding Sequences ft .
~
lattiated by a N xim m LPCI Suction Line areak
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g, Figiare 2.3.17 Inttlator' Event' Tree for Postulated Flooding Sequences I Initiated by a large~LPCI Suction Line Bre&k L . ..
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, g ' . F.igure 2.3.14 Comparison of the NEPs Associated with .-
Operator Actions for Singular. Events and.' >
Coincident Multiple Events -
. 4 I
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.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 , . ,
. pondensate Storage Tank (CST) '
550,000 -
4 CS,HPCI,RCIC' 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- '
through, ECCS pump suctfon piping. -
l t
',l . !
- 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 ' '
e . -
3
. D s . *
. .? ,'
,' , ! / .
'"', l - .
,l 't I*
p e
e 9 - g
.m e
.y - -,,,-c.- - ..,,,- ..,,,,.,-, ,,,, ,,w- . , . ~ ~ . . , , , -y ,_
+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 t ' .
. Flow Rate (Minute,s*) -
3'-10"
, source Location gixn* -
Sup'pression *
, Pool HPCI, Pump Suction 9600 ' 17.6 ,
RCIC Pump Suction 1500 ~- ,
10.6 l t -
LPCI Pump suction '
. (Max /Large)** 17000/8500 9.4/19.0 1 -
CS Pump Suction . -
13000 12.0 .
LP.CI' Pump Suction 10500 -
I5.0 .
(1 Pump Runout)
- CS Pump Discharge 6850 ,
23.0 (1 Pump Runout) .
Co'densate'St' n orage' -
Tank (CST). HPCI Pump Suct' ion . ,
(Max /Large)**' 1200/6000 ,- * '13.0/27.0 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 ,
- 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 '
' ~
5.rair r 5.2 airr-2.5mic)
. T,, ,. ' ,
.- . . . . . 5.Osic6
. T gg3 '3'.0:10*8 , . ,
l.lal0*A .
T gg, 5.8 10*I . . . 4.3:10*8 T
it$
.3.6:10'8 4.1:10-8 ,
T .
- 1.0a10'I 1.3a10'I
. gg . .
T ggy . 6.4a l'0'I ,,
3 5:10'# ,
T ggg ,
1.1a10 5 2.0 10-5 9.0:10*8 T gg . . 1.3a10*8 ,
2.fa10*I 5.Salo'I T
fLIO 2.3 10 2.850 . 1.410'8-
.. ngg . . ,
, ., ,, 1.8108 ., .
,3.4 10'8 *
,1,.5a10
T TL12 1.0a10*I . . 2.1 10*I T ggg3 , 2.6:10*8 ,,
7.8a10 8
,,. .T gg, 1.6510*I . . 2. gal 0'8 .
'T ftl5 4.4a10'8' ' 8'
,2.5a10' . . ,
.T R16 1.141 . 8.1:10*I
- 6'.6al0'I T ggg 2.4al0*I ,
8.8:10*I 2.4 10'I 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
l
=
l l
. l 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.' .
, 13,644'*
. Driven -
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 ..
HPCI (Turbine Driven) , , 0.079 -
" TFL2 RCIC (Tuibine Driven') 0'.079 . TFL1 ' . .
Servi,eWater'(RHR'or c
. RBCLCW HK) ,(Motor Driven) .0.042 ,,
TFLS . . .
. . . s .
. . . . l,
_. __ .. _ _ _ _ _ _ _ _ _ _ ~ ._.. _ . _ _ ._ _ . _ . . - . . _ _ , . . -
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 (100') - -
8.5E-10 WASH-1400 4.3E-11 .
External Failure.of
- a Valve , , ,2.7E-S WASH-1400 1.3E-9
~ *
. External Failure of '-
a Pump' 3.0E-9 WASH-1400 -
1.5E-10'
- 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.'
Size . Characterization - -
Flow Rate . Probabilityf,
~
, 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
- 3.5E-5 01senarge , CST ~
/SUPP. .2 0 1 0 .76/1/14 T(5 y CS l ,
Discharge SUPP
'4 0 2 0 128/2/12~ 6.9E-5 -
Tyg7 LPCI .
01senarge. SUPP 14 '
4 '
4- 0
~ ' ' ' ' ' ' ' ~ "
240/6/16 '
2.5E-4 T
FL8 -
~ '
Service Service .
Water- ,
Water 4 4 4 'O 7,15/3/10-20 1.4E-4 -
y - - .
FL9 .
WFPS . WFPS 'l . . , 157/2/6-8 * , 1,.,1E-5 g
.. > *Eti.0 .
RCIC** .
Suction CST 1 1 1 1 70/1/6 3.5E-5 3FL11 - . . . .
HPCI** .
Suction CST ** '
1- 1 -1 1 87/1/1df 3.5E-5 Tytyg,TFL13 . .
'CS **. ; . t. . , , .
' Suction
- CST *. ' ' .2 '2 ,2 120/2/12' ; 4.9E-5 .
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)
- l Initiator Frecuency (Per RX Yr) ,
Press 6rized Piping . ,
. HPCI Discha.rge Break, TFL6 ,' 3.5x10-5 CS Discharge Break, TFL7 .6.9x10-5 LPCI Discharge Break, TFL8 ,
,2.5x10-4 ,
~
.SW' Discharge Break,'TFL9 .
1.4x10-4 . .
'WFPS Discharge Break, TFL10 1.'1x10-5
HPCI Suction Failure, TF12 (max) -
1.75x10-6.
HPCI Suction Failure, TF13 (large) 3.5x10-6.
. -CS Suction Failure,, TF14 ('max) ,
- 2. 5x10-6. .
'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 . 1.0 15t + 2nd = 0.'54.'
0.1 .
10' .
~'
20 -
0.11 ,
.0.01
- - 0.01,i - 1.1E-3 -
. 3,0 .
60 0.0011 2.0E-4 i
- .r 1.1E.4 - 1.1E 4 .
'1500 -
. .~
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). I 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
lowing a signal to do so. If th' eprobabil'ity,of such an error is P, .
then a test brings t' eh component from State 2, to State 3 *with '
l probability 1-P (bre'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 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 ii) . The mean times to repair were assumed 100 h'rs and 50 hrs for the
. 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 D=5.0x10-4/hr -
i 1.,y)-', Testst.are performed .every 3 months (4~ times a yepr) fgr, both motor - .
. .. . 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 .
Ao =104/hr.
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-I involve ,one of'the following three types of components: .1) piping; 2) valve; ,
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. -
. l 1 .
- e l
... .. : - '- - - -^ - - -
' ' } - ~. - . .
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' g'an MSIV clos,ure; State 4, which 1,s'a
, rupture trigger 5d by a t'urbine-trip transient; and State 5 which is '
- ~
', rupture trig'ered g by a manua,1 shutdown or an' equipment, test. '
Thi solution ,of 'this'model ' yields 'the probabilities that the system will
. occupy States
- 3, 4 and 5 denoted by 3P , P T , Pg*, respectively. The'se probabilities at the end of one-year period provide the frequency of rupture- ,
- initiatid flood precursors. The, expression for these probabilities is Pj(t,) = F [(1.e-A R )/AR '(1-e-At)fx]
.A-AR'
' ~
(1).
,, 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
' 1 R"is' the r' te a of catastrophic 'rutpure' failure in the~ system
' ~
I' '-
'and A is the rate of arrival.of any tran'ient s (1=A +A 3 T+AM) '
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)]
R .
M R R (2)
Quantification . .
For a.given* systdm' having piping of length' L, n'y vilves and hppumpi . I the failure. rate Ag is equal to AR
- LA'+Dvv A +npAp -
(3)
.m 9
.e,r... .. . - _ - _ _ . _ _ _ _
_,j'J. . . - . - . . -- . . . - .- - . -
3-6 where A y , Ap are the catastrophic rupture failure rates for valves and 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' PS-PRk)'.
Using the length of piping, number 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 ,3 PT and Pg were ca,1culated.
- The results are tabulated .
. 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 i
3-7 3.2 BNL Quantitative Rev'iew 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 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: ,
Monitor RB water level to determine approxirate leak rate. Use'su'm'p'
.f. ,
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' urce of leakage. Isolate .using the ap-
.propriate system procedure listed below.. *
- System.
HRCI, Procedure No.SP23.202.01 -
- ' i -
'i L'eakage indication: .. Abnormal suction or: discha,rge piping pressure,.
. 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'consistencies.
. 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 rw. ,,.-,.,,,,-,,,-.-.----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 -
- t 2- then be fo,llowed by. operators. However, BNL, has reviewed.the relevant.Alam' .,
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't ernal Flood event tree analysis and Sec ~
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.e ' vent initiators, ,-
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 asew' ll. At approximately -
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
.. scenarios with tirmination of the flood prior to 3'-10" may not. contr ibut'e an 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 a it 'f rom the rest of the loads in the .
' ~
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't er. exceeds the l'-3" levil but is termin,ated before the . ,
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',t' he first 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 o ' niy th'res functioris:' the' ,
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). ,
.. , 1 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 '
? :' < - i .
t .(
- l 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 I l
~ . ~ - - . . _ _ _ _ _
__ 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 vulnerablec f'requency contributions from Phase,I '
'and.J I :a r'.e very. spal'1,' i.n..,the order,o ' (10-9 Contrib'ut{ons, from, Phase'III ' ,
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 ht' e. following operator action's were included for the. .
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'e . flood (Event. A, see ;
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 .
Median = 1.90E-6 . .
. . 5% Conf.idence ,
= 2.19E-7 .
f .
,. , t .
95% Confidence,= 7.51E.-5 , . .
- e e
,..,,,m.
-2.;_. b_.. _ _ & ; . . . . . ,x . . _ _ _ _. 2- ...
3-16 1 UP .
A . ,
MAINT'. .'
R v .
AD' .- .
4 '
p .
SHUTDOWN
+ ' '
1 2 DOWN * ' +. .
, , $ /
jh g .
j (t_p) ' MAINT.
R ONI.!N I'
~
X ,
BREAKER. .
' ' / -
OUT , ,
. MAINT. .f RX ON- , .
3' 'LINE .
, BREAKER u
. . . ' . . , IN
- x x A: Comp. Failure Rate. -
.p: . Comp. Repair, Rate ,
.y ,- -
A:D Transient (MSIV clos.) Rate'
" , ' P: Prob. of not racking out Breaker
'. . FI'.000 INITIAT l: Rate of inadv. operation of' -
o switch
. 5 ED . ,
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 . . .
1 VULNER- .-
2 s ABLE -
s .' .
s .
\
- N A A S .
T .
M .
a *
\
. \ .
, .g .. .
Y S M T . . .
FLOOD., ptg0D. FL60D .
A
'3 'MSIV CL 1
4 l
& l. . 5 . & .
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.
- - .,y---.y,-,--,ry - - - - - , - - - , - - - - , 4-- r--= =:- - e,-v--
. 3-18 _
1, s
-s%
__._7 .
\
s -
s % *
. s, \ .
. s . . .
10 .-
.' \. \
\ g .
\- s 10 c. .
.q~~~. Upp'er Sound
~
.\
.\ .
ticminal . .
. 1'0" .- .
. Value
+
\.
\ , .
,..g ,
.u . .
N -
Lowe.r 5ound. lj NI .. .
. ; s.
N .
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.!
g .
t . -
, Phase'I ! i -! -
t.
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. . 3, Transfer to E/T . '
i 1
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Phase I of Inte' rn'la Flood; Functional Event. Tree s .. s Figure 3.3.1
- i . I -
. i. .
i i i
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. t e
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i, ; !.
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- . i I ;
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!C M P- Q Divl Div U R U .' , X V - 3 I .. R CS CI cond R ,P -
~
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- . -l
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6 -
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Figure 3.3.2 Phase.II of lInteFnal
]. '
~
~
~.
l
~
,l~ Flood Functionali Event Tree lr :
) -
.- <. i ,
i .
~' - 1 I :^
M. l' i .j 9 -
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- 4. R C' M P _.'Q Divl Div2 UR ; UH.- .X Y CS CI cond R. P
., j i . . -
BKR BKR S ta t. * * .
.- +
+
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- include prob. - .
...i i :
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a . . .
6
- - i ; . 1 division . . .-
. i
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'~
.. . ., Figure 3.3.3 PhaseIIIofjInternal I . . Flood Functional. Event g I j- - -
3.
- ! Tree -
P J 8*
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e i.. ..i .l ..
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