ML20049A281

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Responds to 800528 Request for Investigation of Possible Strategies Re Reopening of NEPA-related Issues on Selected Plants W/Cps But Which Are Candidates for Risk Assessment re-evaluation.Three Alternative Strategies Proposed
ML20049A281
Person / Time
Issue date: 06/19/1980
From: Vollmer R
Office of Nuclear Reactor Regulation
To: Harold Denton
Office of Nuclear Reactor Regulation
Shared Package
ML19250E879 List:
References
FOIA-81-416 NUDOCS 8009290209
Download: ML20049A281 (2)
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? DISTRIBUTION i Central Files LSofer RVollmer Reading SGoldberg, ELD RVollmer CWoodhead ECase J(g ) 91980 DEisenhut Dross DMuller {

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MEMORANDUM FOR: Harold R. Denton, Director Office of Nuclear. Reactor Regulation EM ~ y

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.g. FROM: Richard H. Vollmer, Director Mi Division of Engineering @L h.

SUBJECT:

REOPENING OF NEPA ISSUES ON RISK-STUDY PLANTS v ~ ~;. On May 28,1980 you requested that we investigate possible strategies regarding- - p$ reopening of NEPA-related issues on selected plants that have cps but which are ,y/j candidates for re-evaluation under your program of risk assessment for the-plants ~ "m. on sites that have a high population density. The review areas in which there have been significant' changes since the cps were issued and thus are potentials j, for reopening are: 3 1. The alternative site process--which has become better defined and, in the case of two cps, has utilized more stringent population density and distribution criteria; .4 /

p;s 2.

The "need for poser" justification--which might now be less defensible ' -sg than when the plant was licensed due to a general decrease in load 3,8 growth; and "~ 3. The relative costs of generating electricity with coal and nuclear--- which are now closer, thus making the nuclear decision less obvious. The possible alternative strategies that we could pursue, in the event siting / environmental issues are raised, include: 1. Allow reopening of environmental issues. This alternative projects the image that NRC is a strong regulator responsive to changing times and pressures. 'On the other hand, this option may not be legally defensible since NEPA does not' include a i continuing review function and a valid decision made at a valid review point, such as the CP would not be subject to reveview. (The justi-g i fication for reopening issues at the OL stage is substantial new - ~> f - l information that significantly affects the environment.) This alter-j, native would result in instability in the licensing process, could () result in a high manpower expenditure, and wouid be difficult to Np l i}/ manage since there would be no obvious point at which to cease rereview.1 j l ~ l V v hPPICE b tuiNawE W OE Adia== = A e v=muoy ' ~~ > oavE ( _NRC PORM 318 (9-76) NRCM 0240 W u. s. sovERNMENT PRWTINe OPPtCEs says - 639-e24, (

~ { ~ . ~ ~. s .~ old R. Denton JUN ' S 1980 2. Do not allow reopening of environmental issues. The pros and cons of this alternative are essentially the opposite of those noted under alternative 1, above, and need not be repeated.' Unfortunately, if the rereview of risk at densely populated sites is opened under the guise of NEPA, holding back on any. reasonable other issues would not be consistent with reopening the NEPA consideration of risk in the first place.. Allow considerat4'on of only a limite'd number of issues for which a. ~ 3. showing of definite changes can be made. ~ With proper management this siternative probably could be limited to one or two salient issues, the acceptability of which would be decided on a generic basis; but a problem would be that these same one or two generic issues could then be raised on all cases, whether or not they're high population sites. Neverthe-In my opinion, there is no obvious best alternative that stands out. less, alternative 3, which is not particularly satisfying to anyone, would at least show a regulatory interest in following up on the possible environmental concerns raised following the risk studies. For this reason, I recomend your consideration of this as the appropriate strategy to follow..If you agree, we can develop criteria for implementation of this option with assistance from OELD. There are some members of the staff who are concerned over the appearance that population considerations are being used as the sole basis for triggering detailed risk assessment of selected plants. There is an implication that population is the dominant factor in determining the societal risk of plant operation. Some of the technical staff content that the total risk represented by a plant is a weak function of the population, which for existing sites can affect risk by no more than an order of magnitude and that other factors, such as emergency preparedness, l mechanical design, and maintenance and operation could affect risk by a greater amount. I raise this point merely to suggest you exercise caution in the j manner in which you express your basis for deciding on which plants to pursue i risk analysis. One additional problem, with regard to Bailly, we have to contend with is an extension to the construction permit. I would recommend that such consideration be based solely on issues which deal with reactor safety. I would, therefore, consider it appropriate to require additional measures for safeguarding the' surrounding population such as those that may arise out of the Zion and Indian Point studies unless, of course, it can be shown, because of the large close-in industrial population, that adequate protection is not possible, or for other reasons that it is not needed.. ortstmal sis =ed br: --) i Mehmed H. YeUmar -c Richard H. Vollmer, Olrector / Division of--Engineer 1 1g o,,,c. R,

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6/ L80 o n s,. o

G .T t Limerick Generating Station { Preliminary Risk Assessment PRESENTATION OF RESULTS = 12/9/80 Pottstown Holiday Inn s AGENDA s I. Introduction Vincent Boyer Philadelphia Electric Company Saul Levine NUS Corporation

11. Limerick Plant Richard Mulford Description Philadelphia Electric Company Ill. Features of the Limerick Roger McCandless Boiling Water Reactor General Electric Company IV. Descript!on of Gene Hughes the Analysis Science Applications,Inc.

l V. Discussions Roger McCandless of Results General Electric Company VI. Summary and Vincent Boyer Conclusians Philadelphia Electric Company 7 4 on os a 2 % 9 (

LIMERICK GENERATING STATION PRELIMINARY RISK ASSESSMENT e May 8,1980 - Letter Requested ) Assessment of Plant / Site Combination s o May 21,1980 - Meeting Discussed Approach

  • July 11,1980 - Meeting Clarified the Study Approach e December 9,1980 - Presentation of Results l

Lirnerick Generating Station Preliminary Risk Assessment PRA PARTICIPANTS s e Philadelphia Electric Co. e General Electric Co. e Science Applications Inc. o NUS Corporation ~. e Bechtel Power Corporation e Chicago Bridge & Iron Co. o Fauske & Associates,Inc. o Meteorological Evaluation Services, Inc. o Center for Planning Research

i WASH 1400 OVERVIEW e First Comprehensive Study of Probabilities and Consequences of Nuclear Plant Accident Events o Directed by Norman Rasmussen 1 MITll and Saul Levine I AECl e Started in 1972 and Completed in 1975 o 70 Man Years and Four Million Dollars c Compared Nuclear Plant Risk / Consequences to the Risk / Consequences of Natural and Man-Made Events 4

Limerick Generating Station s Preliminary Risk Assessment CASES STUDIED e WASH 1400 at Limerick Site e Limerick Plant at Limerick Site L.

.i Limerick Preliminary Risk Assessment Wash 1400 Risk Comparison 1 L MAN-CAUSED RISK 100 \\ TOTAL NATURAL RISK 1000 1 10,000 FREQUENCY 1 (everits/ year) 100,000 1 1 MILLION l 1 10 MILLION N 1400 BWR l 1 N 100 MILLION \\ 1 1 BILLION 1 10 100 1000 10,000 FATALITIES i 03235.s _,. -,,.. -.,. -...,.,...,.. _, -. ~ - - - - - -, _. _ - - - -,... -. - -.. - - -. -- ----_ _ _

Limerick Generating Station c Preliminary Risk Assessment WASH 1400 PLANT AT LIMERICK SITE e WASH 1400 Probabilities e WASH 1400 Release Fractions e WASH 1400 Methods o Limerick Site Population o Limerick Site Meteorology

Limerick Generating Station s Preliminary Risk Assessment LIMERICK PLANT AT LIMERICK SITE e Limerick Plant Features e Limerick Site Features o Updated Data dn Methods A 4

Limerick Generating Station Preliminary Risk Assessmerit RESULTS e Limerick Site s Higher Population e Limerick Plant Better Than WASH 1400 Plant ~.

  • Limerick Plant-Site Together Lower Risk Than WASH 1400 BWR e Limerick Generating Station Presents Minimal Risk to Public 4

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1 l Limerick Preliminary Risk Assessment s l FEATURES OF THE LIMERICK I .f BOILING WATER REACTOR h o s / Roger McCandless .1 General Electric Co. 4 s, g r, _L.,_L..,.-,------ --,E-~~-*---~~*= " ' ~ ' " " " ' ' ' ' ' ~ ' " ~ ' ' ~ ' ' ' ' ~ ~ ' ~ ' ~ ~ '

..x _,, \\ LIMERICK PLANT FEATURES M A R K 11 CONTAINMENT = a... d...'U l'iii d.: X W a " ' CONCRETE REACTOR BUILDING I -~b L.. TURBINE GENERATOR =. ..q A i k, q" ?. f T C Ni.,') . g _ __... g c a :......,.. n -N -- g !;. !. I l l. l ,i u b'p.;' J j MAIN ?'s ' ',' CONDENSOR y j

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k,{glD1[@72&13 '"i x.'I .Q)i; ,.3, 5l l 'f I.. t FEEDWATER SYSTEM i ~~m ' ' ' ' ~

  • SUPPRESSION POOL i

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l s S N t O a e IT H C ya N c U e F D e e N r v o o O C m IT o e t C R r E e t T a O W R y P p l p E u R S O C s ~ l l i . : i: i J l ii1 ilI!; 4 !li

g COOLING SYSTEMS N Turbine Water ' ' '-"'~' -;,,J. Condensor I ),'-! 3 'd i A ^ . y, q s t ,L t 'N. , e ..~.,t.' /i.,, t Control Rod Reactor Core Feedwater injection isolation Cooling System System System

y--.------------- r i EMERGENCY COOLING SYSTEMS i \\ I'. ; E i ., f.;r y.:. v.-' , s' 4.. s

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P i.YI k 4h di' ' A lA A A A A A E I e a a a a a 5.;4j!;.}Yh.,,.,,,7, '[j.%,4l;%}j$'yeji. [ YdhD '5 ';'.I ) Condensate..., I ,t ,... Am 3,.,.. ........, g. u. 4.. 5'. StorageTank H i High Pressure Low Pressure Low Pressure Coolant injection Coolant injection Core Spray System System System B547.38 i 1 8

SYSTEMS TO SUPPLY WATER TO CORE HIGH PRESSURE 4 Feedwater 4 High Pressure injection 4 Reactor Core isolation Cooling 4 Control Rod Drive Cooling 13 Pumps LOW PRESSURE Pa 4 Coolant Injection ,d Flood 4 Core Spray 4 Condensate w 5 5

l l t DECAY HEAT REMOVAL x Bam_ - st... ser.ty/n.ii.evaiv. ? 9- -... s t..m Residual Heat Removal System Core i h t Heat Exchangern Pumps suy.ssion B54 7.40 4 _.--.-,---_--_-,,---w-,---.. .----.---em--=-.---.-==-----.m.-w.---w= 3,ew--= e-+-+--- ---e----r-ws-+ e--- ee

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I s FISSION PRODUCT CONTROL FEATURES

  • Suppression Pool e Containment Sprays e Bypass Leakage Control o Secondary Containment e Containment Leakage Filtration

EMERGENCY OPERATION l e Simple System i

  • One Vessel i

l e One Loop

  • Reactor Water Level Measured i

j e Simple instructions to Operator I B 54 7.4 --_.-_--m_,---.-- - - *----- -, - - -. - -.---.m,-u

l CONCLUSIONS l l I e Multiple Water Supplies ) l ~ % Heat Removal Capability I

  • Fission Product Control i

e Simple Emergency Operation f 14

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Limerick , Preliminary Risk Assessment ~ METHODOLOGY i i Gene Hughes Science Applications, Inc. 8547 27 i 1 i

i LIMERICK ANALYSIS e Systems -- Limerick (BWR/4) e Procedures -- Limerick Projection e Operating Experience Data -- Philadelphia s l Electric Company Where Applicable i e Containment -- Limerick Mark Il e Sequences -- Limerick Specific i e Containment Analysis -- Limerick Specific e Consequences -- Limerick Site-Specific Risk -- Limerick Specific l ......4.- ...--...-..._._,-m%me..

Fault Success Trees r-Event Trees

  • Criteria

~ i a s Event Specific Plant Results Frsquency of y Release Release Fractions A Offsite Analysis F Summation of Sequences 9 Risk Curve 854731

1 7 F I I l Fault Success l l Trees & Event Treese Criteria l L N J s Event Specific Plant Results Frequency of Release y [ Release Fractions b. Offsite Analysis y Summation of Sequences y Risk Curve E !47.34 g I l

ACCIDENT INITIATORS l

  • Plant initially Operating at Power i
  • Plant Safety Challenge Occurs j

- Normal Transient - Small Loss-of-Coolant Accident i - Large Loss-of-Coolant Accident - Transient Without SCRAM I Allows Specific Capability l to be Analyzed i

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es.1w} l u@- x -. i l Turbine Trip Large ' s' MSIV Closure S m all Loss of Feedwater Very Small ~ IORV Loss of Offsite l Power indepth Analysis for Transients Assured

\\ ..~. PLANT-SYSTEM RESPONSE e Demands on Core Cooling s i e Demands on Containment Cooling

  • Demands on Pressure Control (Reactor and Containment)

~. e Demands on Core Criticality Determined for Each Initiator and Event Sequence l l ) i

SUCCESS CRITERIA e Define Minimum Systems to Achieve Shutdown e Determined for Each initiator 'eeBased Upon Realistic Analyses e Use Both Safety Systems and Normal Operating Systems Status: Developed and Reviewed for Each initiator

Limerick Preliminary Risk Assessment SUCCESS CRITERIA Transients [with SCRAM;l Require: l

  • Reactor Core isolation Cooling i

s O' e High Pressure Coolisiit (injection l ' e 1 fo 3 Feedwater Pumps O' o 1 or 2 Low Pressure Core Spray Loops O'

  • 1 of 4 Low Pressure Coolant Injection Pumps or o 1 Condensate Pump and e Main Condenser 1 of 2 Residual Heat Removal Loops 1

EXAMPLE OF AN EVENT TREE l Functions \\ A B C D E A .l AE Event AD Sequences AC 1 ACD 1 AB

,.., y. [. System Lsvel Fault Trees i 8

  • CRD INJECTION
  • HIGH PRESSURE COOLANT INJECTION l

e FEEDWATER ~

  • REACTOR CORE ISOLATION COOLING i

e CONDENSATE e LOW PRESSURE CORE SPRAY / e AUTOMATIC DEPRESSURIZATION

  • EMERGEiNCY SERVICE WATER f

SYSTEM l o POWER: AC (NORMAL), , esBESIDUAL HEAT REMOVAL AC (EMERGENCY), DC i - LOW PRESSURE

  • DIESELS COOLANT INJECTION

~ - CONTAINMENT SPRAY e STANDBY LlOUID CONTROL - RHR SERVICE WATER 1 I FAULT TREES ARE LOGIC MODELS FOR FUNCTIONS )' n. I 032358 o'

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success Fault Trees r Event Trees

  • Criteria

\\ r Event Specific i l Frequency of Plant Results l Release p Release [ , s' Fractions Offsite Analysis if Summation of Sequences l If Risk Curve

a 1 L l i TO DETERMINE FREQUENCY OF EACH EVENT SEQUENCE s

  • Select input Probabilities for Fault Trees 1

)

  • Run "Wam Code" Series to Combine Probabilities to Determine the System Level Failure Probability
  • Quantify Event Tree Sequences by Combining

-- Initiator Frequency -- System Level Fault Tree Values Frequency of Each Event Determined d i l

CRITERIA FOR SELECTION OF DATA s i e PECO Specific l

  • Most Recent Data I

i e Generic BWR Data i e WASH-e400 i 1 l Result: Best Available Data Used l i

l OVERVIEW OF ACCIDENT SEQUENCE PROBABILITIES e Attempt to Maintain the Same Criteria for Successful System Operation as We Perceive was Used in Wash-1400 s - No Heroic Actions - Once a System Falls the System Stays Failed - Little Credit Given for Operator Action Within 30 Minutes .e Changes Only in the Following - Differentiation in the Types of Accident initiators - System Success Criteria - Containment Overpressure Relief - New Data (Components, Maintenance, Diesels Offsite Power) 1 Result: Limerick-Specific Event Trees Quantified with Best Available Data .uu I

Fault Success Trees > Event Trees

  • Criteria p

s l Event Specific l I Plant Results I Frequen;y of I I Release Rei ase g g ,. Fractions g L-~_ Offsite Analysis U Summation of Sequences U Risk Curve 0547.32 . - - _. _ -.., _ -. - _ -. ~ -,,.. _ _ _.....,... _ _., -.. - - -.. - -... _.

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1 i Boll PVMelt inter 4 Core Uncovery Pressure Vessel Core-Concrete and Meltdown Melt-Through Interaction s i Mass and Energy Flows Contempt-Lt Heat Transfer and Atmosphere Exhange l i Containment Conditions (Reactor Vessel, Drywall, Wetwell, Misc. Compartments) l m :. ) i \\ j d' +

4 ~s., CORE MELT ANALYSIS . m.g 3 .L"g s i o t., n h ,i f 7 %lRrc8l ,i 5- ' r /' 3 x'., / 8 ( e ',i e o1 = =, I

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j....' A !141' l 'QY( L. ) V W l F Es. -..nr. 4, g .l e.. u cowreen '1 lrAH 1. i. M ARK il CONTAINMENT

y e ' GENERIC ACCIDENT SEQUENCES s e Melting Before Containment Failure e Melting After Containment Failure e Transient Without SCRAM e Transient Without SCRAM (Heat Removal Failure Case)

j i EACH GENERIC EVENT ANALYZED TO DETERMINE e Time of Release

  • Fraction of Core Content Released
  • Path and Energy of Release

) l I l 1 l 4 1 i

l i N . BASED UPON D TAILED EVENT SEQUENCE PHYSICS s e Compartment Flows e Plant Conditions l e Phases of Event l are Used to Establish Release Fractions l l

RELEASE FROM FUEL 4 WASH-1400 Methodology \\

1. Eight Species Classes Examples:

Halogens (1, Br) Tellurium Group (Te, Se, Sb)

2. Four Components Fuel Rod Rupture Fuel Meltdown t

Steam Explosion Concrete Penetration

3. Two Physical Forms Gaseous Particulate I

i DECONTAMINATION FACTORS [DFD impact on Lime.:ck s i

c.-tamination Source Release Fractions Primary System Retention No Credit Taken -- Inability to Quantify Effect l

Suppression Pool Scrubbing Credit Taken for Saturated as Well as Unsaturated Pool s Primary Containment and Same DF as Used in Wash-1400 Retention but Retention Times in Some Cases are Longer Standby Gas Treatment Same DF as Wash-1400 ) E!47 2 l

r DECONTAMINATION FACTORS \\ Meltciown VesssiFa... <. Conditions Release V a po.-iz e. t :- __ -) Containment Failure at j s End of Release 100 10 [ i .y Containment Failure i Initiates Release 10 10 = _ EM t.

CONTAINMENT FAILURE MODES e No Failure at All e Drywell Region s e Drywell and Wetwell Region e Wetwell Air Space ~# e Wetwell [ Drain Pool;I e Leak [Small;I

  • Leak [Large I

l e Steam Explosion [In-Vessel? e Steam Explosion [In-Containment;'

  • Hydrogen De'lagration Failure

i' t I l r l' ii i : e ) ) 6 6 c t n n ( ( e u K E C e i O 7 y'7 g 6 & Yppa v i Eq e S y b me t t r ds a ne u n a e l t e si aGr mya T t S S F 1 ek ga r a e L L l tos oo PL ll e er wlu yir a DF l le e r wu l t iea WF ek ga ra e L L -on l o H pi xs E de er a su Hui l aa CF e n r t r u ian es t e s v nmOre o C p e l r e-r u s es s v s e e V Orp et l ro e CM i lIi l 1 i t.l l!: i! 1

l i "CRAC" CALCULATION o Specify Fractions of isotope Groups l e Time of Release s e Energy of Release

  • Evacuation Model
  • Weather e Population Calculate Consequences

..g* 4 1

t "CRAC" CODE e Runs 91 Trials Summed s e Sum isotope Categories

  • Sum-Up Meteorology Selected for Each Case

's Sum-Up Population Sectors e Weights with Frequency / Stores -- " Bookkeeping Code" Result N Risk Curve i

ADDITIONAL SAFETY MARGIN (Not Evaluated) '* Nuclear Regulatory Commission Actions in Response to Three Mile Island s e Institute for Nuclear Power Operations Recommendations .

  • Nuclear Safety Analysis Center Actions Being Taken (Now in Progress)

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~ ^ Limerick Preliminary Risk Assessment DISCUSSION OF RESULTS Roger McCandless General Electric Co. ff

Limerick Preliminary Risk Assessment DISCUSSION OF RESULTS Site Differences e Plant Design Differences

  • Data Differences e Methodology Differences

LIMERICK SITE COMPARISON (Relative to WASH 1400) e Realistic Site

  • Higher Population e Different Weather e Same Evacuation Model 1

t l k LIMERICK DESIGN COMPARISON (Relative to WASH 1400 BWR 1

  • MK ll Reinforced Concrete Steel-lined Containment e Larger Standby Gas Treatment System o Containment Overpressure Relief

'e More and improved Safety / Relief Valves ' e improved Piping e improved Shutdown System e Spray Pond for Emergency Cooling Water e improved Emergency Pump Capability e Four Dedicated Emergency Diesel Generators e More Reliable Offsite Power

x Lim; rick Prsliminary Risk Asssssmsnt site comparison 1 L MAN-CAUSED R SK 100 s TOTAL NATURAL RISK 1000 3 1 10,000 FREbUENCY 1 (events / year) 100,000 1 1 MILLION WASH 1400 BWR AT LIMERICK SITE 1 10 MILLION N 1 N 100 MILLION WASH 1400 B \\ 1 1 BILLION 1 10 100 1000 10,000 FATALITIES j ouasi

l: o l-LIMERICK DATA COMPARISON l 0 Relative to WASH 1400)

  • Larger Data Base

~

  • Initiating Frequencies I

e Equipment Reliability e Maintenance Times !I

^ LIMERICK METHODOLOGY COMPARISON JRelative to WASH 1400) e improved Computer Models s o More Comprehensive Treatment of Transients 'e Updated Decontamination Factors e Updated Treatment of Hydrogen / Steam Physics \\

  • Use of Emergency Operator Guidelines 1

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)1 ~ ' ~ Lim 2 rick Prcliminary Risk Asssssmsnt Methods & Data Comparison 1 L MAN-CAUSED RISK 100 TOTAL NATURAL RISK g 1000 3 1 10,000 FREOD'ENCY 1 (events / year) 100,000 1 1 MILLION WASH 1400 BWR WITH UPDATED METHODS -.. AND DATA AT LIMERICK SITE 1 ~% l 1 10 MILLION ^ 1 \\ 100 MILLION WASH 1400 BW 1 1 BILLION 1 10 100 1000 10,000 FATALITIES 03235 4

1 LIMERICK DESIGN COMPARISON !l Relative to WASH 1400 BWR) s e MK ll Reinforced Concrete Steel-lined Containment e Larger Standby Gas Treatment System e Cont' inment Overpressure Relief a e More and improved Safety / Relief Valves i e Improved Piping e Improved Shutdown System o Spray Pond for Emergency Cooling Water e improved Emergency Pump Capability e Four Dedicated Emergency Diesel Generators e More Reliable Offsite Power

Y l Limerick Preliminary Risk Assessment I Design Features Comparison 1 L MAN-CAUSED RISK 100 s TOTAL NATURAL RISK 1000 3 1 10,000 FREQUENCY 1 (ev.ents/ year) 100,000 1 1 MILLION WASH 1400 BWR WITH UPDATED METHODS AND DATA AT LIMERICK SITE * 'DOES NOT INCLUDE. 10 MILLION LIMERICK DESIGN FEATURES 100 MILLION LIMERICK AS DESIGNED 1 1 BILLION 1 10 100 1000 10,000 FATALITIES 032356

^ ." l ~ Limerick Preliminary Risk Assessment Risk Comparison Summary 1 L MAN CAUSED R SK 100 h g TOTAL NATURAL RISK 3 1000 i 1 10,000 FREQUENQY 1 (events / year) 100,000 1 WASH 1400 BWR WITH UPDATED METHODS AND DATA AT LIMERICK SITE

  • 1 MILLION I

'DOES NOT INCLUDE-10 MILLION LIMERICK DESIGN FEATURES 100 MILLION WASH 1400 BWR d LIMERICK AS DESIGNED ) 1 BILLION 1 10 100 1000 10,000 FATALITIES 03235 6 [

1 i l l Limerick l Preliminary Risk Assessment s

SUMMARY

AND CONCLUSIONS l Vincent Boyer ~ Philadelphia Electric Co. i l l l

i ) Limerick Preliminary Risk Assessment Limerick / Wash 1400 Risk Comparison 1 \\ L MAN-CAUSED RISK 100 TOTAL NATURAL RISK 1000 3 1 10,000 FREQUENCY 1 (events / year) 100,000 4 1 1 MILLION 1 -WASH 1400 BWR 10 MILLION N 1 N 100 MILLION LIMERICK N 1 1 BILLION 1 10 100 1000 10,000 FATALITIES -}}

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