ML20211C769
| ML20211C769 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 09/12/1984 |
| From: | Speis T Office of Nuclear Reactor Regulation |
| To: | Duval R ENERGY, DEPT. OF |
| Shared Package | |
| ML20209C800 | List:
|
| References | |
| CON-FIN-A-0801, CON-FIN-A-801, FOIA-87-6 NUDOCS 8702200210 | |
| Download: ML20211C769 (34) | |
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. TSpels DST CHRON Mr.RichardA.Duval,ActingMa'naher JHalvorsen San Francisco Operations Office VZeoli, ORM U. S. Department of Energy DDandois, ORM 1333 Broadway, Wells Fargo Building Oakland, California 94612
Dear Mr. DuVal:
Subject:
LLNL Technical Assistance to the Division of Safety Technology NRR, NRC, " Review of Seabrook Nuclear Power Plant Probabilistic Risk Analysis" (FIN A-0801) The enclosed NRC Form 173, Standard Order for DOE Work, is hereby submitted in accordance with Section III.B.2 of the DOE /NRC Memorandum of Understanding dated February 24, 1978. Funding authorization in the amount of $75,000 to begin work on this subject project was transmitted on April 3, 1984 and $156,000 was transmitted on August 28, 1984. The purpose of this letter is to transmit FY84 funds in the amount of $29,000 for continuation of the project. If you have any questions concerning acceptance of this order, please contact Ms. Halvorsen on FTS 492-7932. Sincerely,
/S/
l Themis P. Speis, Director Division of Safety Technology , Office of Nuclear Reactor Regulation g , gg,4 o,
Enclosure:
NRC Form 173 D42lo cc: R. Barber, HQ/ DOE f J. M. Johnson, LLNL A. Garcia, LLNL, L-95 {g22 0 870211 SHOLLYB7-6 PDR l /#ff RRAB: DST RRAB: DST RRABgDS AD S NRR,7 ,D:h' , orne p
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fe' ORok R NUugt R u s Nycts AR ascut AToae couuissioN NRC eonu 173 41 181 20-84-681 OAit
. STANDARD ORDER FOR DOE WORK ggg ACCOUNTING CITATION ISSUED SY: INRC Officol _
APPRO*RIATION SYMBOL ISSUED TO: lOOE Of ficel Office of Nuclear Reactor 31X0200.204 San Francisco Operations Ofc. Regulation ,, ,,,,,, 20-19-40-41-5 PERFORMING ORGANIZATION AND LOCATION "" Lawrence Livermore National Laboratory A bE-4 Livermore, CA . WORK PERtOO . THIS ORDE R FIXE 0 S ESTIMA7Eo O FIN TlTLE FROu: TO: Review of Seabrook Nuclear Power Plant 05/15/84 01/15/84 Probabilistic Risk Assessment OBLIGATION AVAILABILITY PROVIDED BY: S 29,000 A Twr$ ORoER 8 TOT"AL OF ORDERS PLACEO
" ADON PRIOR SYM80L" TO HRST AND THE THIS FOUR OATEOlG4TS WeTHOFTHE 7,g3g,g$g THE PERFORMING ORGANIZ ATION S
Y"s e A NS!a*$"[iNDO /c (TOT AL A & 81 5 2,868,256 C. TOTAL ORDERS TO DATE S 260,000
- 0. AMOUNT INCLUDED IN "C" APPt.lCASLE TO THE ** FIN NUMBER" CITED IN THIS ORDE R.
FINANCf AL FLExis#La TV: B FUNDS AUTHORIZE 0. WILL NOT BE REPROGRAMMED SETWEEN FfNS. LINE D CONS N tEvEL uP TO s50K. LINE C CONSTITUTES A LIMITATION O FuNoS MAv eE REPROGRAsuEoNOT TO ExCEEo .sosOF F ON 08 LIGA 78ONS AUTHORIZED. STANDARD TERMS AND CONDITIONS PROV10ED DOE ARE CONSIDERED PART OF THIS ORDER UNLESS OTHERWISE NOTED. ATTACHMENTS. SECURITY-THE FOLLOWING ATTACHMENTS ARE HERE8Y MADE A PART OF THl5 ORDER: .
@ WORK ON THIS ORDER 15 NOT CLASSIFIED D 5TATEMENT OF WOR K , O WORK ON THis OR'OER INVOLVES CLASSIFIED INFORMAf TON. NRC FORM 187 is ATTACHED D ADDITION AL TERMS AND CONDITIONS C OTHER 4
REMARMS This order provides incremental funds to continue work associated with this project 20-84-620. in accordance with LLNL proposal dated 6/5/84 as modified by Work Order D. Dandois, After acceptance, e send to the Office of Resource Management, ATTN: and provide a c y to the Office of Nuclear Reactor Regulation, ATTN: K. McGrath ACCEPTING ORGANIZATION
<7 I . l. i M'"Jf NG AtfTHORITY r
" N N Sheis( N tor %-l1-h ftTLE bivisionofSafetyTechnology D A TE NRCFORM 17311 781 P
r STATUS OF SEABROOK PRA SEISMIC, WIND, AND EXTERNAL FLOOD REVIEW by John W. Reed Martin W. McCann, Jr.
)
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Presented to U.S. Nuclear Regulatory Commission - Bethesda, Maryland. F6I A - 2 7 - M (, Bjn August 28, 1984 Jack R. Benjamin & Associates,Inc. g Consulting Engineers Mounto.n Boy P,czo. Suite 501 I 444 Costro Street. Mounto n v.ew Co forn.c 94041
i (- a 6 SEISMICHAZARDANALYSIS l STATUSOFREVIEW 2 e READ PRA REPORT SECTION e READ llM. 89 HAZARD EVALUATION FOR SEABROOK e PREPARING HISTORICAL SEISMICITY ANALYSIS WITH UNCERTAINTY t e SEISMICITY CONSULTANT HAS REVIEWED ZONATION AND SEISMICITY PARAETER ESTIMATES FINDINGS TO DATE e SOURCE ZONATION TENTATIVELY APPEARS REASONABLE e ESTIMATES OF MAXIttJM MAGNITUDE ARE A MAJOR CONCERN OF THE SEISMICITY CONSULTANT e USE OF PEAK GROUND ACCELERATION TO SCALE A SITE-SPECIFIC RESPONSE SPECTRA IS APPROPRIATE e THE EARTHQUAKE DURATION FACTOR FAILS TO TAKE INTO ACCOUNT FREQUENCY DEPENDENCE e SIGNIFICANT DIFFERENCE BETEEN THE d ($@) HAZARD STUDY AND THE UTILITY STUDY Jack R. Bonjomin & Associates,Inc. O Consulting Engineers D
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'e EXIB5'AL FLOOD ANALYSIS STATUS On Review e READ PRA REPORT SECTIONS e RECEIVED RECENT FIA STUDY DRAFT REPORT e READ FSAR SECTIONS FINDINGSTODATE e APPROXIMATE, POINT ESTIMATE PROBABILISTIC ANALYSIS WAS PERFORMED e FREQUENCY ESTIMATES ARE BASED ON STATISTICAL EXTRAPOLATIONS OR ASSLNED VALUES e Piti AND ftN PAR #ETERS TENTATIVELY APPEAR REASONABLE e SENSITIVITY OF FLOOD ELEVATIONS TO HYDRAULIC PARAMETERS ERE NOT INVESTIGATED l
l EXPECTATIONS FROM PLANT IRIP e LOCATE EXPOSED SAFETY-RELATED STRUCTURES AND EQUIPMENT e SURVEY FLOOD PROTECTION STRUCTURES e IllSCUSSIONS WITH SYSTEMS PRA ENGIEERS Jock R. Senjamin & Assoclotes, Inc. E Consulting Engineers D
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SEISMIC FREILITY ANALYSIS STATUS OF REVIEW e READ PRA REPORT SECTIONS e REQUESTED FRAGILITY CALCULATIONS (HAVE NOT RECEIVED) e SYSTEMS EQUATIONS REQUIRED FOR SENSITIVITY ANALYSIS FINDINGS TO DATE e NO SLIDING ANALYSES PERFORED e SOME COMPONENT CAPACITIES APPEAR LOW e RELAY CHATTER NOT CONSIDERED FAILURE e PRA MEAN FREQUENCY OF CORE ELT IS 2.89-5 EXPECTATIONS FROM PLANT IRIP e SEE CRITICAL COPPOENTS e FAMILIARIZATION WITH PLANT e DISCUSSIONS HITH SYSTEMS / STRUCTURAL PRA ENGINEERS 1 Jock R. Bonjomin & Associates,Inc. I Consulting Engineers B
TABLE 9.2-3. LEA 3 ROOK KEY E't:01URES/:( *:1: 'S FOR SEISP.;C MA.YS!! Median Acceleration Capacity Symbol Structure / Equipment a 0R O U
@ Reserve Auxiliary Transformers 0.30 0.25 0.62
@ Unit Auxiliary Transformers 0.30 0.25. 0.62
@ Switchyard 0.40 0.25 0.54
@ Switchgear 0.41/1.52* 0.32 0.31/0.48
@ Motor-Driven Emergency Feed Pumps 0.66 0.40 0.56
@ Steam-Driven Emergency Feed Pump 0.66 0.40 0.56
@ Spray Additive Tank 0.75 0.40 0.32
@ 120V AC Instrument Buses 0.75 0.42 0.36 -
@ 480V Motor Control Centers 0.78/> 2
- 0.36 0.61 h 480V Transformers, Buses 0.79 /> 2
- 0.37 0.72 h RWST 0.86 0.40 0.33 PCC Heat Exchangers 0.99 0.37 0.49
@ Diesel Fuel Oil Day Tanks 1.03 0.39 0.48 h RHR Pumps 1.07 0.34 0.65
@ Safety Injection Pumps 1.07 0.34 0.65 h Charging Pumps 1.07 0.34 0.65 h Control Room Evaporator Units (diesel generator building) 1.18 0.16 0.50 h Reactor Internals 1.50 0.38 0.44 h Diesel Generators 1.51 0.36 0.35 -
h Steam Generators 1.71 0.36 0.39 h Service Water Cooling Tower Fans 1.71 0.41 0.39 h Reactor Coolant Pumps 1.74 0.35 0.32 h Reactor Building Crane 1.75 0.25 0.55 h MSIVs 1.86 0.41 0.41
- Fragility shown is for recoverable chatter and trip, respectively.
i Structural failure is significantly greater. 0917P080383 9.2-18
i WIND ANALYSIS STATUS OF REVIEW e READ PRA REPORT SECTION e REQUESTION NSSFC DATA FOR SITE FINDINGS TO DATE e TORNADO WIND HAZARD TENTATIVELY IS REASONABLE e BUILDINGS DESIGNED FOR 360 res e PRA MAN FREQUENCY OF CORE ELT IS 1.13-8 EXPECTATIONS FROM PLANT IRIP e SEE SAFETY-RELATED STRUCTURES l e SEE EXPOSED SAFETY-RELATED EQUIPENT e DISCUSSION WITH SYSTEMS PRA ENGIEERS e INSPECT MISSILE POPULATION gP/LA
' %uIL mc pd It'Y f"l0 x
mah Qwb&# / (# y s4 l g s Jock R. Bonjomin & Associates,Inc. Consulting Engineers E S J
PROGRESS REPORT ON REVIEW 0F SEABROOK STATION PROBABILISTIC SAFETY ASSESSMENT P. R. DAVIS AUGUST 28, 1984 fo2.a- F7-ood Bha ,
o 4 AREAS 0F REVIEW
- 1. FAILURE DATA A. COMPONENT DATA B. SYSTEM DATA C. COMMON CA!JSE QUANTIFICATION
- 2. ACCIDENT SEQUENCES A. ASSUMPTIONS, ANALYSIS, PHENOMENA ASSOCIATED WITH SEVERE ACCIDENT PROGRESSION
- 3. DEPENDENCIES A. INITIATING EVENT B. INTERSYSTEM DEPENDENCY C. INTERCOMPONENT DEPENDENCY (COMMON CAUSE, COVERED UNDER 1.C. ABOVE)
- 4. REVIEW 0F MAJOR RESULTS AND ASSUMPTIONS
PROGRESS TO DATE (8/28/84)
- 1. FAILURE DATA A. COMPONENT DATA --
COMPLETED COMPARISONS MADE WITH ALTERNATE DATA SOURCES (SSPSA DATA BASES PROPRIETARY) EVALUATIONS MADE OF DATA WITH NO COMPARISONS AVAILABLE FOR OUTLIERS OR UNKNOWNS, EVALUATION OF SIGNIFICANCE OF COMPONENT FAILURE DETERMINED FROM DOMINANT ACCIDENT SEQUENCES JUDGMENT PROVIDED ON SIGNIFICANCE OF DISAGREEMENTS B. SYSTEM DATA APPROACH
. COMPARE RATES
. REVIEW QUANTIFICATION
. IDENTIFY OUTLIERS
. ASSESS SIGNIFICANCE
. REQUANTIFY COMPARISONS COMPLETED WITH ALTERNATE DATA SOURCES FOUR SYSTEMS REVIEWED
. AC POWER
, DC POWER
. SERVICE WATER
. PRIMARY COMPONENT COOLING
. INSTRUMENT AIR I
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s COMPONENT DATA REVIEW FINDINGS
- 1. SSPSA DATA BASE PROPRIETARY
- 2. OF THE 59 COMPONENTS, 37 DATA COMPARISONS WERE FOUND
- 3. ALL FAILURE RATES APPEAR REASONABLE OR NOT SIGNIFICANT BASED ON PRELIMINARY REVIEW
- 4. SSPSA IMPORTANT COMPONENT (VALVES, PUMPS, DIESEL GENERATORS) FAILURE RATES SOMEWHAT CONSERVATIVE
- 5. INCONSISTENCIES AND UNUSUAL VALUES FOUND FOR MEAN/ MEDIAN RATIOS
i I j COMPARIS0N OF SSPSA MEAN TO MEDIAN RATIO WITH OTHER DATA SSPSA RANGE FROM OTHER DATA RANGE COMPONENT MEAN/ MEDIAN Low HIGH NUMBER FACTOR 1 SINGLE CONTROL ROD 5.5 IE-4 4.6E-5 2 2.2 PIPE RUPTURE s 3" 9. IE-9 8.6E-9 5 8.6 i PIPE RUPTURE a3" 9. 4.56E-11 8.5E-10 5 20. VALVE Disc 5.4 -- -- 0 --
! RUPTURE BISTABLE SPURIOUS 13.2 -- --
0 -- OPERATION 1 Bus FAILURE 1.5 IE-8 3E-5 4 3000. RELAY FAILURE 2.2 4E-6 IE-3 5 250. E i l l i l
h SYSTEM FAILURE RATE REVIEW
. AC POWER --
REVISION IN COMMON CAUSE CONTRIBUTION WILL RAISE CMP BY LESS THAN FACTOR OF 2
. DC POWER --
NO RISK CHANGE
. SERVICE WATER SYSTEM --
NO RISK- CHANGE
. PRIMARY COMPONENT COOLING --
INDETERMINATE ISOLATION OF NON-ESSENTIAL LOADS [3 FACTOR APPLIED TO AIR-OPERATED
- VALVES
, INSTRUMENT AIR --
NO RISK CHANGE i l l J I
TABLE 6.3-2. BETA FACTOR DISTRIBUT'10NS , a Component Failure 5th . 95th i Mean Median Percentile Description Mode Percentile High Pressure Fail to Start 5.88-2 1.09-2 5.03-2 1.36-1 Injection Pump Fail During 6.40-2 8.78-3 5.29-2 1.59-1 Operation s (SI and CC) ,, Containment Fail to Start 1.25-1 8.70-2 1.69-l' 2.81-1 Fail During 2.23-2 6.45-4 , 8.29 7.50-2 Spray Pump , Operation Service k'ater Fail to Start 1.11-1 4.58-2 1.05-1 1.95-1 Pump Fail During 7.62-2 3.44-2 7.27-2' 1.30-1 Operation , Fail to Start 3.65-2 , 7.34-4" 1.00-2 9.45-2 Component Cooling Pump Fail During 2.32-2 6.51-4 8.43-2 7.65-2 Operation RHR Pump Fail to Start 6.67-2 1.05-2 5.72-2 1.53-1 Fail During 2.76-1 1-33-1 2.74-1 3.79-l' Operation Fail During 1.18-1 4.48-2 1.11-1 2.16-1 Emergency Feed- s water Pump Operation (turbine-driven and motor-driven) Fail to Open/ 4.23-2 2.22-2 4.08-2 6.72-2 Motor-Opera ted Valve Close on Demand Diesel Generator Fail to Start 1.46-2 4.57-3 -1.33-2 2.93-2 Fail to Run 3.25-2 1.41-2 3.08-2 5.69-2 1.11-1 4.58-2 1.05-1 1.95-1 Reactor Trip Fail to Operate Breaker on Demand 1.25-1 1.00-3 5.00-2 5.00-1 Generic -- Component
- Excludes driver. .
NOTE: Exponential notation is indicated in abbreviated form; i .e. , 5.88-2 = 5.88 x 10-2,
o
.\ ~
s, , DEPENDENCIES REVIEW (COMMON CAUSE FAILURES) s. wx
, - g) _- 1 ' , [7 FACTORS NOT REASONABLE FOR FOLLOWING COMPONENTS
[ '4 ' FAILURE SSPSA N COMPONENT MODE RATE OTHER Ls PCC PUMP START 0.0365 0.307 I1)
} OPERATE 0.0232 0.349(1)
DIEGEL- START 0.0146 0.08(2) 0.16(3) ,0,08(4) GENERATORS OPERATE 0.0325
. IN PCC PUMP CASE, NO CHANGE IN SYSTEM FAILURE s . IN DIESEL GENERATOR CASE, EMERGENCY AC POWER LOSS PROBABILITY INCREASED OTHER PP.0BLEMS
. $0 COMMON CAUSE CONTRIBUTION FOR " PASSIVE" COMPONENTS
. COMMON CAUSE OMITTED FROM SOME COMPONENTS (VENTILATION FANS, SOME VALVES, BATTERIES)
- 1. NUREG/CR-2098, " COMMON CAUSE RATES FOR PUMPS"
. 2. NUREG/CR-2099, " COMMON CAUSE RATES FOR DIESEL GENERATORS"
- 3. NUREG/CR-1362, " DATA SUMMARIES OF LERS OF DIESEL GENERATORS AT NUCLEAR POWER PLANTS"
- 4. NUREG/CR-2989, " RELIABILITY OF EMERGENCY AC POWER SYSTEMS AT NUCLEAR PLANTS" T
TABLE E.3-:i.
SUMMARY
OF ACCIDENT SEQUENCES WITH SIGNIFICANT C M AND CORE MELT FREQUENCY CONTRIBUTIONS
$heet I of 2 Sequence Ranking Initiating Additional System Failures /
. Event Humn Actions Resulting Dependent Failures r qu n y latent Early (per reactor year) Health Health Risk Risk l (O7 Loss of Offstte Onsite AC Power, No Recovery of AC Power Component cooling, high pressure makeup 3.3-5
- Power Defore Core Damage (ECCS), reactor coolant pump seal LOCA, (/4/] 1 1 1:entainment filtration and heat removal.
Loss of Offsite Service Water, No Recovery of Offsite Onsite AC power, component coolf ng, high 9.2 6 2 2
- Power Power and low pressure makeup (ECCS), reactor *
(t/) coolant pump seal LOCA, containment flitratton and heat removal. Small LOCA Residual Heat Removal None. 8.9-6 (J,q} 3 *
- u Control Room None Component cooling, high and low pressure 8.7 6 [3,F) 4 3
- Fire makeup (ECCS), reactor coolant pump seal LOCA, containment filtration and heat
,ro removal.
Y Loss of Main Solfd State Protection System Reactor trip, emergency feedwater, high 8.3-6 ('38) 5 4
- y Feedwater and low pressure makeup (ECCS), contain-ment flitration and heat removal.
Steam Line Operator Faflure to Establish Long Tecm 5.6-6* *
- Break Inside m (2/h 6 Jontainmen Heat, Removal 2 [ .
Reactor trip Component Cooling High and low pressure makeup (ECCS), 4.6-6 7 5
- reactor coolant pump seal LOCA, contain-ment flitration and heat removal, Loss of Offsite Traf n A Onstte Power Train 8 Service Train R onsite power, component cooling. 4:4-6 (/T 8 6
- Power Water, No Recovery of AC Power Defore hf gh and low pressure makeup (ECCS).
Core Damage reactor coolant pump seal LOCA, contain-ment flitration and heat removal. l Loss of Offsite Train B Onsite Power Train A Service Train A onsite power, component cooling, 4'.4-6 (/.T 9 7
- Power Water, No Recovery of AC Power Defore high and low pressure makeup (ECCS),
Core Damage reactor coolant pump seal LOCA, contaln-ment flitration and heat removal. PCC Arce Fire None Component cooling, high and low pressure makeup (ECCS), reactor coolant pump seal ' 4.1-6 [/,P) 10 8
- LOCA, containment flitration, and heat removal.
.I
- Negligible contribution to risk.
NOTE: Esponential notation is indicated in abbreviated form; f.e., 3.3-5
- 3.3 x 10-5, 00 W 122?n3
TABLE 2.3-5 (continued) - Sheet 2 of 2 Sequence Ranking In t ting Additto iy t Failures / Resulting Dependent Failures equ n y Latent Early (per reactor year) Core g 14el t Health Health Risk Risk [ . Partial Loss of Component Cooling High and low pressure makeup (ECCS), reactor 11 9
- Main Feedwater coolant pump seal LOCA, containment flitra- 3.8-6 (l.7) tion, and heat removal.
Cable Spreading Component conling, high and low pressure 10
- Room Fire None makeup (ECCS), reactor coolant pump seal 3.5-6 [/.5) 12 LOCA, containment ffitration, and heat removal.
Loss of One DC Emergency Feedwater, No Recovery of Diced and feed cooling, Train A 3.2-6 h. 13 *
- 1 Dus Emergency or Startup Feedwater containment filtration and heat removal.
Reactor Trip Operator Failure to Establish Long Term None. 3.0-6 (/,3) 14 *
- Heat Removal.
Turoine Trip Component Cooling Iffgh and low pressure makeup (ECCS), 2.8-6 h.2) 15 11 , y reactor coolant pump seal LOCA, contain-
. ment filtration, and heat removal, w
[La Loss of Service Water None , Component cooling, high and low pressure makeup, reactor coolant pump seal LOCA, 2.3-6 (/) 16 12
- containment filtration, and heat removal.
f
, Partial loss of Feedwater Operator Failure to Establish Long Term lleat Removal None. 2.3-6 ([) 17 - * * /
Turbine Building Fire Onstte AC Power, No Recovery of AC Power Defore Core Damage of* e power, component cooling, high and low pressure makeup (ECCS), reactor 2.3-6 (f) 18 13
- coolant pump seal LOCA, containment filtra-tion, and heat removal.
Sna11 LOCA Train B Safety Features Actuation. Traf n A Residual lleat Removal Train A high and low pressure makeup and residual heat removal; train 8 containment 2.2-6 (l) 19 *
- l filtration and heat removal.
Small LOCA Train A Safety Features Actuation. Train D Residual IIcat Removal Traf n B high and low pressure makeup and residual heat removal; train B containment
,2.2-6 (l) 20
- filtration and heat removal.
Turbine Trfp Reactor Trip, Failure to Mar.ually Scram Functional inability to provide adequate 1.9-6 [.F/ 25 *
- Reactor and to Effect Emergency Doratton high pressure makeup.
Interfacing None Low pressure makeup, residual heat 1.8-6,(.I) 27 14 1 Sys*. ems LOCA removal, containment Isolation and flitration.
*Negligtble Contribution to risk.
NOTE: Exponential notation is indicated in abbreviated form; f.e., 3.8-6 = 3.8 x 10-6, 0901P12??A3
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*4 s6t24CC *2t01995t6616945t 5C4t554 f 10 e 114 5o 0 9 e 3 1
* * * " * * - * * * * * - * ' * * * * * * * " ' * * ' * * - * * * * ' * **
- c 1 t f4 P5t 1't12E cett!i5 *T *i * * *Ct521 *411
- t81
- e p
0 3 q b , 7 u e 7 l 0S y4 va 4 < 4 d d 1 d a5a ss Ca s s3Ctss v3C C u2} Cs 1 tsittf22ilis sOO3s e s cs s i t v.* !vt=i *: = : ?
?*
- 3se99A tt p 6 aosp a6h L +
ewe t 2l e0$ l(jl Illl!llil lll
aur I.E , d ), Kwowlug Do . FEom A*4 n Docupa.4r
%*eecs Ptacar Damage shte pa.m N C S , G.),
Exnedue. A;; vs. Ew. E,c, L,z c 6.ai f.= 2. ,c to pereewdue. (nearera. rF k;* , EXAMI*S*- hameusar Asx. S.syear S+n4 e, G) IF E E;z, L;ta s hXAmeME b liM b .ML. l 4.a.L,t-Fea. Donowaar EmnL/Teme. ges y se S h h. Cx)
t Number Matrix Operation Results 1 (F = (IM Frequency of occurrence of plant damage states. 2 (8 = (FC = (IMC Frequency of occurrence of release categories. 3 4" = 4'S = (I MCS Frequency of exceedance of damage (master assembly levels (repeated for each damage index). equation) 4 NC Conditional frequency of each release
' category given each initiating event.
5, CS Conditional frequency of exceedance of each damage level given each plant damage state (repeated for each damage index). t 6 MCS Conditiona1 frequency of exceedance of each damage level given each initiating event (repeated for each damage index). 7 Frequency of occurrence of plant states as (fM a result of initiating events. 8 Frequewy of occurrence of release (fMC categories as a result of initiating events. 9 Fmquency of exceedance of each damage (fMCS level as a msult of each init(ating event l (rupeated for each damage inden). 10 Fmquency of occurrence of release
${C categories from each plant state.
l l l 11 Frequency of exceedence of each damage
${CS level as a eesult of each. plant state .
(repeated for et.:h damage indeal. 12 Fewquency of exceedance of each damage
$$$ level as a result of each release category (mpeated for each damage index).
.. - ' ' ~
e i 9 l ,, ,_ 1
l . l NSIGHTs 1
. No cor sers
. 09,7 someciCAL VALUES
. No rmeoarAN CE.
INFORMATiott l 1
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