Text

Forwards Mods to 238 Nuclear Island PRA Rept to Be Added to Rev,Per Discussion W/D Yue.Rept Provided for Accident Consequence Analysis Review to Allow NRC Review in Technical Update Meeting.Viewgraphs from 820622-23 Meetings Encl
	ML20129E458
Person / Time
Site:	05000447
Issue date:	07/16/1982
From:	Holtzclaw K; GENERAL ELECTRIC CO.
To:	Frahm R; NRC
Shared Package
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References
	FOIA-84-175, FOIA-84-A-66 NUDOCS 8506060499
	Download: ML20129E458 (53)
	v • d • e

r-EN@t'.$lh%E!F.PTEle I

s s

NUCLEAR WER 4

SYSTEM S.DIVISIO N GENERAL ELECTRIC COMPANY,175 CURTNER AVE.. SAN JOSE, CALIFORNIA 95125 MC 682, (408) 925-2506 July 16, 1982 L

h Mr. R. Frahm U.S. Nuclear Regulatory Commission.

Washington, DC 20,555

Dear Ron:

Attached are the modifications to the 238 Nuc1 car' Island Probabilistic Risk Assessment report that Steve Stark and I discussed with you and Dave,

Yue last Monday.

We are planning to make these changes in the GESSAR

' report as part of a future revision.

In order to allow the-NRC and its consultants an opportunity to review the changes prior to next week's Technical Update Meeting in Bethesda, we afe providing copies to you and the key reviewers involved with accident consequence analyses.

As we mentioned on Monday, the changes noted in the attachment were made to include the results of the GE Pool Scrubbing Program, the effects of containment sprays, and the use of the central estimate dose model.

~

It is our intent to cover these changes in detail at next week's Technical Update Meeting.

If you have any questions before that time, please contact me at (408) 925-2506.

Very truly yours, fS N}W X. W. Holtzclaw Principal Licensing Engineer Nuclear Safety and Licensing Operation 9

KWH:sem/B071611 h

~

J. Meyer (NRC) / 8. #2A 4[

R. Bari (BNL)

\\.

cc:

@ Yue (NRC Q )

Q J. Mitchell (NRC q'

8506060499 841203 T

PDR FOIA CURRANB4-A-66 PDR Of

.h Q,*.P A " f-

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GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION

- Claas III

)

s' l %f wi h

Table 1 V

ESTIMATED CORE DAMAGE AND RISK COMPARISON Assessed

-. Risk Frequency Early Latent 3

of Event Fatalities Fatalities Per Reactor per Reactor per Reactor Event Year Year Year I.

CORE DAMAGE RSS BWR/4 Mark'I 8

-5

-2

@ composite site 44x10 N1x10

%5x10 RSS6 BWR/4 Mark I

. i tI O~*

-5

-6 4

-3

@ site #6C

%4x10 1.2x10 4.8x10 h to' &

BWR/6 Mark III

-6 d-

-5

@ site #6c 5x10

,0 1.7x10 II.-

U.S.

NATURAL BACKGROUND continuous 0

814 RADIATION "With WASH-1400 Methods (calculated from the reported curves) bThe total accident-caused fatalities over the lifetime of the

. exposed population or the calculated excess cancers in the.same population from one year of background radiation cComputed with the GE'CRAC code e

e e

S e e p shdAe prevuudy mpedeg in 9-OJ-Q uv n

., o o vuus.nn umu

' GENERAL ELECTRIC COMPANY

+

. PROPRIETARY.INFORMATION t..'

Clasa III 10-3 4

10 M

A C

6 10-5 8

04 Y.,

E WASH 1400 BWR ATSITE 4 F

5 10-4

.m N

o w

8WR/6 AT SITE 6 E

i-o N

.10-7 w

\\

10-a l

l l

l 10-8 0

10 101

2 3

4 I

10 10 10 10 L.ATENT F ATAllTIES PER YEAR (X)

Figure 1.

Comparison of Risk for the WASH-1400 BWR and BWR/6 xvi

,, ~,_,

.. -, _ -.,.,. - ~,....., -...,.

c

_ _m-GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION Clasc I.II i

10~3 Y

U

~

Gy V

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j,'

510-5

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AT COMPOSITE SITE

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g WASH 1400 swn AT SITE e 10-8

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gh 10 101 2

0 4

10 ig LATENT FATAUTIES IX) PER YE AR Figure 2.2-2.

Site Comparison of Latent Fatality Curves for WASH-1400 BWR Releases 15.D.3-33/15.D.3-34

<. a o - a vu ur.,an

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GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION t,.*

CLASS III 10-3 4

10 s

K A

4 E 10 l

c

+

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~

IE 5

h 10-8 U

=

5 Y

i S

E BWR/6 AT SITE 6' "8

k go-7 104 3,_;

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100 1

2 s

4

~

10 10

,a

,o 3,

LATENT FATALITIES PER YE AR IX)

Figure 6.1-1.

Comparison of Risk for the WASH-1400 BWR and BWR/6 15.D.3-130

GESSAR II 238 NUCLEAR ISLAND GENERAL ELECTRIC COMPANY

~

PROPRIETARY INFORMATION Class III 7.1.2 The Risk Curves (Continued) suppression pool scrubbing and in-vessel' fission product '

retention, there were no early fatalities among the CRAC runs and'no BNR/6 CCFF early fatality risk curve can be constructed a

The risk curve' for latent fatalities is presented in Figure 7.

and is discussed in Section 7.2 and 7.3.*

7.1.3 Risk

~

~ i))nr

, Risk is calculated as the product of the assessed frequency of OS,bf release categories per reactor year and the estimated average.

f}d

' number of consequences per release category, summed over all release categories.

The risk,for BWR/6 is' calculated to be 1.7 x10~ lifetime latent f atalities per reactor year and is further discussed in Sections 7.2 and 7.3.

J

~

The number of latent fatalities per year was calculated by dividing the total lifetime latent fatalities per accident by 30 to allow comparison with WASH-1400 results (Section 6.1.2) 15.D.3-136

I ovuuuna As unu GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION t..,

CLASS III to-3 M

io

=

X A

5 2

5 5o-5 8

G 5

-l wAss i4oo nWR AT SITE a r

5 to-5

_B b

5 g

g

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y, swr /s Ar sers :

k ev so-*

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0 iot to2 to3 to to d

5o t ATENT FATAUTIES PER YE AR IX)

Figure 7.1-2.

Comparison of Risk for the WASH-1400 BWR and BWR/6 15.D.3-139

- ~ - - ~ '

~'

g urssAR II 238 NUCLEAR ISLAND GENERAL ELECTRIC COMPANY

PROPRIETARY INFORMATION Class III 7.2 COMPARISON WITH WASH-1400 (Continued) in a saturated suppression pool.'

In the RSS, BWR risk was evaluated for a composite of sites, which was meant to represent a compo' site of all BWR sites in the United States.

In the BWR/(

analysis, the risk was evaluated at a specific site (RSS Site 40 The difference-in risk due to the site difference is small as can be seen by comparing the curve for the WASH 1400 BWR at the composite sihe with the curve for the WASH-1400 BWR at Site'#6 (section 2.6).

n'

j Another difference in the evaluation of risk was the use,of-e

~

an updated version of the CRAC code in the BWR/6 analysis.

The

^ '

y difference in risk due to the use of a different CRAC code was.

smal-1 and is shown in Appendix F.4.

Figure 7.1-2 compares the RSS and BWR/6 CCFF risk curves for atent fatalities.

The risk of latent fatalities for the BWR/6 is ess than the risk for the WASH-1400 BWR at Site 6 by a factc of abou 2 (Table 7.2-1).

This reduction is primarily due to the add onal prevention and mitigation features of the BWR/6 -

Mark III design.

Another' measure of risk is the assessed average number of conse-quences (early and latent fatalities) per reactor year.

The RSS

~

provided no evaluation of average number of consequences specif-ically fbr the BWA.

Only risks for ths combined average for the BWR and PWR plants were provided (Reference 7.2-1).

To provide a basis for comparison with the WASH-1400 BWR, the average numbe of consequences was estimated from the RSS BWR CCFF curves.

The BWR/6 risk is lower by several orders of magnitude as,shown in Table 7.2-1.

15.D.3-141

GENERAL ELECTRIC COMPANY t,-

PROPRIETARY INFORMATION Class III Table 7.2-1 ESTIMATED CORE DAMAGE AND RISK COMPARISON Risk Assessed Frequency Early Latent of Event Fatalities

, Fatalities Per Reactor Per Reactor Per Reactor Event Year Year Year I.

CORE DAMAGE RSS BWR/4 Mark I

-5

-5"

-2" e composite site.

s4x10 six10 s5x10

,y, l.l 10 g[

~3 RSS BWR/4 Mark I.

-5

-6 6 site 46C N4x10 1.2x10 4.8x10

~ 4~

200 BWR/6MargIII

-6

-5 9 site 46 5x10 0',

1.7x10 II. U.S. NATURAL BACKGROUND

-continuous 0

814-RADIATION "With WASH-1400 Methods (calculated from the reported curves).

bThe total accident-caused fatalities over the lifetime of the exposed population or the calculated excess cancers in the same population from one year of background radiation. '

CComputed with the GE CRAC Code.

6 g

I e

s b

f 15.D.3-143 t

s

':JO NUCLEAR 1SLAND GENERAL ELECTRIC COMPANY PROPRIETARY INFORMAT20N CLASS III 1

10

/

,,0 FLOOo o

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TC RNADO ACTUARIAL

=

ASSESSMENT HURRICANE 1,~I ~,

EARTPQUAKE

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gg-4 METEORITES gwR/8 LATENT

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. NOTE NO EARLY FATALITIE5' RECORDEO FOR EMRMI U

10

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ga NUMBER OF FATALITIES (N)

Figure 7.3-1.

Risk Comparison Between Natural Hazards o

and BWR/6 15.D.3-146

GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION e,

CLASS III l

1o AVIATION 0

MARINE 10 ACTUARIAL ASS'ESSM ENT MOTOR VEM

=---.

R AILRO ADS ESTIMATED to'l MIN!NG DAM PAILURE o-2 7 k CHLORINE TRANSPORT s.%.

v

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BWR/S LATENT RECORDED FOR SWR /S

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Figure 7.3-2.

Risk Comparison Between Man Made Hazards and BWR/6 15.D.3-147

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e GE 238 NUCLEAR ISLAND (GESSAR II)

SEVERE ACCIDENT SUBMITTAL GE - NRC TECHNICAL UPDATE MEETING #3

-s BETHESDA, MD JULY 22 & 23, 1982 t

4 S-l?

KFH-1 '

7/22/92 n

INTRODUCTION TECHNICAL UPDATE MEETING OBJECTIVES...

o DISCUSS DETAILS OF GE SUBMITTALS ADDRESSING SEVERE ACCIDENT ISSUES o

BWR/6 STANDARD PLANT PROBABILISTIC RISK ASSESSMENT o

PLANT FEATURES INCLUDING PLANT IMPROVEMENTS o

SEVERE ACCIDENT PREVENTION AND MITIGATION CAPABILITIES o

FISSION PRODUCT SCRUBBING TEST PROGRAM e

o FACILITATE STAFF REVIEW 0F TECHNICAL SUBMITTALS FOR GESSAR II SAFETY EVALUATION REPORT 4

9 KHH-2 7/22/82

,]

t REVIEW 0F HIGHLIGHTS GE - NRC TECHNICAL UPDATE MEETING #2

SUMMARY

o BWR/6 STANDARD PLANT PRA PROVIDES:

o SUCCESS' CRITERIA FOR INITIATING EVENTS o

CLASSIFICATION OF CORE DAMAGE SEQUENCES BY EVENT TYPE AND PHENOMENOLOGY o

CONTAINMENT EVENT TREES THAT DEFINE RELEASE PATH AND FREQUENCY o

RESULTS OF ANALYSES o

VERY LOW PROBABILITY OF HYDROGEN GENERATION o-BWR FEATURES PRECLUDE STEAM EXPLOSIONS o

LOCAL AND GLOBAL COMBUSTION AND LOCAL DETONATION MORE PROBABLE THAN GLOBAL DETONATION o

MOST PROBABLE LOCATION FOR LOSS OF CONTAINMENT INTEGRITY AT HIGH ELEVATION IN DOME AREA o

CONTAINMENT FUNCTION MAINTAINED FOR SEVERE ACCIDENTS o

SUPPRESSION POOL RETENTION CAPABILITY QUANTIFIED THROUGH GE SCRUBBING TEST PROGRAM f

o USE OF REALISTIC DECONTAMINATION FACTORS RESULTS IN LOW 0FFSITE DOSES KMH-3 L

7/22/82 L:

AGENDA TECHNICAL UPDATE MEETING #3 o-INTRODUCTION K. W. HOLTZCLAW o

FISSION PRODUCT TRANSPORT MODELING D. A. HANKINS o

CONSEQUENCE ANALYSIS D. A. HANKINS o.

PRA RESULTS D. A. HANKINS o

SUMMARY

AND CONCLUSIONS K. W. HOLTZCLAW o

QUESTIONS AND ANSWERS D. A. HANKINS o

AGENDA TECHNICAL UPD TE MEETING #4 K. W. HOLTZCLAW l

D KFH-4 7/22/82

. s a

BWR/6 STANDARD PLANT PROBABILISTIC RISK ASSESSMENT FISSION PRODUCT TRANSPORT MODELING AND CONSEQUENCE ANALYSIS D. A. HANKINS JULY 22, 1982 e

e G 9

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J GESSAR II 22A7007 238 NUCLEAR ISLAND R v. 2 GENERAL ELECTRIC COMPANY

.s PROPR Y INFORMATION Cla a III TASKI FREQUENCY OF SECTION 3 CORE DAMAGE i f FREQUENCY OF TASK #1 MA0lOACTIVE TASKIV SECTION 4 RELEASE SECTION $

3 l

CONSEQUENCES g

OF RADIOACTIVE 1

RELEASE TASK til M AGNITUDE OF SECTION $

RADICACTtvE RELEASE i

Figure 1.5-1.

Major Tasks of the PRA 15.D.3-19 e

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 2 GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION Class III N

J DECONTAMINATION FACTORS i l QUANTIFICATION CORE AN ALYSIS OF OF RE LE ASE RADIOACTIVE FIS5tONPRODUC*

e IN#LANT w

F R ACTIONS RE LE ASE INVENTORY TRAMRT SY RELEASE FRACTIONS CATE GORIES a l RE LEASE

~

SUMMARY

o CORE FREQUENCY OF CORE DAMAGE o

5 X 10-5/ REACTOR YEAR o

DEMONSTRATES EFFECTIVENESS OF BWR/6 SAFETY FEATURES o

RISK CURVES o

NO CALCULATED EARLY FATALITIES

-o LATENT FATALITY CURVE BELOW WASH-1400 4

1 9

4 Kk'H-6 7/22/82

GESSAR II 22A7007

+

238 NUCLEAR ISLAND GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION CLASS III Table 7.1-1 BREAKDOWN OF THE ASSESSED' FREQUENCY OF CORE DAMAGE PER REACTOR YEAR Frequency of Core Damage Event oer Reactor Year Class Description BWR/6 I

Transients e Loss of off-site Power 4.1xlO

~

~7 e All others 5x10

~I I

Drywell LOCA 2x10 C

II Loss of Heat Removal

[2x10-8,f Following all events

-10 III ATWS with Standby' Liquid 9x10 Control but without core cooling

-8 IV ATWS without Standby Liquid 5x10 s

Control but with core cooling

-11 V

Ex-Containment LOCA 2x10

,s'_[/

VI '

Containment LOCA

/

5x10 fXlO

~ ' - -='

-6 TOTAL 4.7x10 k

l 15.D.3-137 i

l

e 4

,~

~

i ASSESSE D FRE OUENCY OF CORE DAMAGE /RE ACTOR YE AR 9

5 5

5 k

I E

1 1

1

'A L

I I

I d b en ISOLATION 1

V' 4

4 C

i N

w n

a 1

h 60SS OF OFF-51TE POWER 4

Y Y

40 W

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%M OZ l

0 M M O

INADVE RTENT OPEN S/RV WWW e.

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McM P-l M O IA C

SMALL BRE AK MMMMM U

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t BWR/6 PROBABILISTIC RISK ASSESSMENT

SUMMARY

& CONCLUSIONS e

ek e

I O 4

9 9

XYH-5 7/22/82 E.1

9

SUMMARY

(CONTINUED) o-COMPARIS0N WITH WASH-1400 o

METHODOLOGY BASICALLY THE SAME o

INPUT AND ANALYSIS DIFFERENCES BWR/6 MK III VS BWR/4 MK I

~

RSS SITE 6 VS COMPOSITE OF SITES BWR/6 FAULT TREES - ANALYZE MORE COMPONENTS AND FAILURE MODES LARGER NUMBER OF ACCIDENT CLASSES AND RELEASE CATEGORIES BWR/6 PRA INCLUDES UPDATE ASSESSMENT OF INITIATING EVENT FREQUENCY ADDITIONAL OPERATING EXPERIENCE REVISED COMPONENT FAILURE PROBABILITIES MORE REALISTIC SUCCESS CRITERIA MORE REALISTIC TREATMENT OF FISSION PRODUCT TRANSPORT MODELING o.

COMPARIS0N OF RESULTS

~

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EARLY FATALITIES f

LATENT FATALITIES l

I KWH-9 7/22/82 k -

or.bbAM 11 238 NUCLEAR ISLAND y

GENERAL ELECTRIC COMPANY I

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PROPRIETARY INFORMATION Class III go-2 l

i j

io-*

N A*

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Comparison of Risk for the WASH-1400 BWR and BWR/6 xvi

n i.

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SUMMARY

-(CONTINUED) 1 s

o COMPARISON OF OTHER RISKS c

o AVERAGE NATURAL BACKGROUND EXPOSURE

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o.

NATURAL AND MAN-MADE HAZARDS

- 0 NEW CURVES is P0OL SCRUBBING DFs O

o CONTAINMENT SPRAY DF o

CENTRAL ESTIMATE DOSE MODEL o

PER EVENT BASIS

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8

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KWH-11 7/22/82

eczeens 66 ~

238 NUCLEAR ISLAND GENERAL ELECTRIC COMPANY

9 PROPRIETARY INFORMATION j,#

CLASS III I

10 10 FLOOD TORNADO ACTUARIAL

=

ASSESSMENT HURRICANE 10"I

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METEORITES

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l' 108 10' 10 to 10 toe 10 2

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NUMBER OF FATALITIES (N)

Figure 7.3-1.

Risk Comparison Between Natural Hazards and BWR/6 15.D.3-146 y

<,__.--____..t

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238 NUCLEAR ISLAND GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION class III i

I 10 AVIATION 0

MARINE 10 ACTUARIAL PIRE MOTOR VEH

. ASSESSMENT RAILROADS

. ESTIMATED 10-1 INING DAM FAILURE

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Risk Comparison Between Man Made Hazards and BWR/6 15.D.3-147

~

SUMMARY

(CONTINUED) l[

o COMPARISON TO PROPOSED NRC SAFETY GOAL NRC PROPOSED NUMERICAL BWR/6 PRA PARAMETER' GUIDELINE RESULT CORE MELT PROBABILITY 1 X 10 "/RY 5 X 10-5/RY I,NDIVIDUAL EARLY 5 X 10-7 s0 FATALITY RISK INDIVIDUAL LATENT 2 X 10 5 2.1 X 10 12 FATALITY RISK SOCIETAL EARLY 1 X 10 "

$ 0 FATALITY RISK (THEORETICAL DEATHS)

SOC'IETAL LATENT 3.2 1.7 X 10 5 FATALITY RISK (THEORETICAL DEATHS)

G KHH-34 7/22/82

/h CONCLUSIONS o

LOW FREQUENCY OF CORE DAMAGE - ATTRIBUTED TO BWR/6 PREVENTION FEATURES o

CONSEQUENCE OF GE SAFETY APPROACH MULTIPLE PUMPS AND HEAT REMOVAL CAPACITY o

SYSTEM' CAPABILITIES EXTEND BEYOND REGULATORY REQUIREMENTS o

NO EARLY FATALITIES - ATTRIBUTED TO BWR/6 MITIGATIVE FEATURES o

MAINTAIN CONTAINMENT FUNCTION o

LIMIT FISSION PRODUCT RELEASE o

LOW RISK - ATTRIBUTED TO BWR/6 PREVENTION AND MITIGATION CAPABILITIES o

BWR/6 RISKS SUBSTANTIALLY LOWER THAN MAJOR NATURAL AND MAN-MADE HAZARDS o

RISK DUE TO EXPOSURE TO AVERAGE U.S. NATURAL BACKGROUND SUBSTANTIALLY LARGER THAN RISK FROM POSTULATED BWR ACCIDENTS o

BWR/6 STANDARD PLANT RISKS WELL WITHIN PROPOSED SAFETY GOAL KWH-15 7/22/82

..e A

W BWR/6 STANDARD PLANT-

. PROBABILISTIC RISK ASSESSMENT f

+

QUESTIONS AND ANSWERS D. A. HANKINS J

JULY 23, 1982 9

5 s

6 0 1

1 y

1

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f Iff,I QUESTIONS o-RISK CONTRIBUTION BY RELEASE CATEGORY.

O SENSITIVITY ANALYSES CRAC INPUT (HEIGHT OF. RELEASE)

CRAC INPUT (TIME OF RELEASE) o DF CALCULATION FOR CRACKS-o CORRAL /CRAC INTERFACE (DURATION OF RELEASE)

~

o PRIMARY SYSTEM PLATE 0VT o

EFFECT OF EVALUATION ASSUMPTIONS o

SITE PARAMETERS DAH-1 7/23/02

s.

GE_ PROPRIETARY INFORMATION BWR/6 PRA RISK RESULTS BY RELEASE CATEGORY RELEASE LATENT FATALITIES MAN-REM PER CATEGORY.

EVENT PER REACTOR YEAR REACTOR YEAR

\\'

g 'y 3 l

b Ta V U' W

lilGilEST PROBABILITY SEQUENCES INVOLVE CONTINU0US POOL SCRUBBING DAH-2 7l23/G1

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y DF CALCULATION FOR LEAKAGE PATHS L

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I L

i i-

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NO CREDIT TAKEN FOR

't\\ y LARGER CRACKS j

-\\ [Y) 0\\

t-e REALISTIC CREDIT f.

FOR SMALL CRACKS j

t DAH-5 N

7/23/82

\\

c

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CRAC/ CORRAL INTERFACE o

CORRAL CALCULATES PERCENT FP RELEASE VERSUS TIME

- 0 SET DURATION OF. RELEASE f

v

\\,

t t

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CONSERVATIVE RELEASE TIME FOR H'IGH BOILING POINT FISSION PRODUCTS M

S

~

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DAH-6 7/23/82 A

~

IN-VESSEL PLATE 0VT

( 00 o-DF = 10 FOR LOW FLOW RATE EVENTS BASED ON SURRY ANALYSIS (NRC MEM0 JUNE 1981) o BWR. PRIMARY SYSTEM SURFACE AREAS LARGER o

BWR PRIMARY SYSTEM TEMPERATURES COMPARABLE OR LOWER 9

S COMPARISON OF BWR AND SURRY INDICATES BWR DF IS HIGHER O

DAH-7 7/23/82 E

'J *.

,.1, Table 2.1 Potential effects of some natural processes in the PCS during accidents Processes Etiects' Aerosol agglomeration I.ow aerosol concentrations would usually not be rapidly decreased by and depositiosi these processes; however, high aerosol concentrations could be rapidly decreaseil in the PCS if steam flow rates were low Conelensation onto (and Accidents with low temperatures in the PCS could involve substantial evaporation from) deposition of gases onto acrosols if enough aerosols were present; particles accidents with high temperatures throughout the PCS would not involve u.

such deposiLlon.

Condensation onto (and Accidents with low temperatures in the PCS could involve substantial evaporation from) deposition onto surfaces if larne enoueh surface areas were available; surfaces accidents with high temperatures throughout the PCS would not involve such deposition.

Chemical reactions and The chemical form of a species could change and'thus alter the radioactive decay likeliliuod of retention in the PCS.

Scrubbing Passage through water could substantially reduce both the gases and acrosols escaping Irom the PCS.

"For details, see Table C. I of Appendix C.

M i

DAH-8 A

7/23/82 M

~

REACTOR VESSEL INTERNALS ETEAal OATER LIFT 80e8 LIJO

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me ase.esaa mv

/

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J" xg q

'. g, w; j f =.t"a g

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sesam ourta

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't.

j y', =

A44-I bt k'[,PSEDWATERInsLET pq'

='~'="

j.,

. wm f

usuascoo,unt gij I

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.i coat enav ramaan Q =a

....;r 9

i.

i t

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o.g v -

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l nr puis inecincourion f

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ji i N

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-h seurno6aosoeum #

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BMAS PLUM teoINT04 DAH - 7 7/23/82

EFFECT OF EVACUATION ASSUMPTIONS o

EVENT CLASS I SMALL BREAK WITH HYDROGEN DETONATION-HIGHEST CONSEQUENCE RELEASE CATEGORY (I-SB-E1)-

o DOSES T.0 BONE MARROW f

f k

J 1

h

)/ /J EVACUATION HAS LITTLE EFFECT t

),

.MH-19 ',

7/23/12 m.

_9-

.=,

,e

,i e__

.*s.

COMPARISON OF SITE METEOROLOGY 9

~

ACUTE 80NE MARROW DOSE r-Distance Site (mi) 1 2

4 5

6 7

\\

1 h

l-LATENT WHOLE BODY DOSE I

-m 9

Pss SITE 6 HAS AVERAGE METEOROLOGY l

DAH - 11 7/23/82,,

I e

l j;.

CONCLUSIONS

.o HIGHEST PROBABILITY SEQUENCES INVOLVE CONTINUOUS POOL SCRUBBING o

CONSERVATIVE DF FOR DRYWELL CRACKS o

USED CONSERVATIVE DURATION OF RELEASE o

IN-VESSEL RETENTION SHOULD BE SIGNIFICANT 9

o EVACUATION IS UNIMPORTANT

~

o SITE 6 HAS AVERAGE METEOROLOGY G

0 DAH-12 7/23/82

.

ML20129E458

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