ML20203N166
| ML20203N166 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs, 05000000 |
| Issue date: | 03/10/1986 |
| From: | Knight G AFFILIATION NOT ASSIGNED |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20198G688 | List:
|
| References | |
| NUDOCS 8609230178 | |
| Download: ML20203N166 (85) | |
Text
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' GARY D. KNIGHT 118 TOLLCATE WAY
- FALLS CHURCH, VIRCINIA 22046 I
March,10, 1986 t
t Nuclear Regulatory Commission 1711 H Street, N.W.
Washington, D.C. 20555
~
Dear Sirs:
1 I am writing this letter for inclusion into the record of the request of the Baltimore Gas & Electric Company' to have the radius of the emergency planning zone for the Calvert Cliffs Nuclear Power Plant reduced from ten miles to two.
- My wife and I, together with my parents, Capt and Mrs. 0.T.
- Knight (USN, Ret.) own a weekend home at Box 220K White Sands Dr.,
Lusby, Md. 20657. Our home is four houses in from St. Leaonard Creek.
We have no real problem with the request of Baltimore G&E in this matter, and we understand that the Commission has deferred further consideration of this request pending the completion of *-
various technical studies.
However, we would like to' interject into the proceedings a practical consideration, as opposed to a legal one. Should the NRC grant Baltimore G&E's request literally (i.e., an EPZ of exactly two miles in circumference from the plant) our home, and perhaps a dozen or two others lying immediately outside the two mile arc.would not have to be notified of any nuclear incident at the pl, ant.
" This is disturbing in that the only escape route, by land, ,
from the White Sands area is toward the plant (from outside the EPZ to a point far inside the.EPZ to intersect with Maryland Rt. 4).
While BG&E may argue that if the scientific data are sufficient to convince the NRC to reduce the EPZ from 10 miles to 2 miles, then White Sands residents just outside the new 2 mile EPZ should not concern themselves.
However, given the variable of wind and weather and the imprecise nature of trying to predict the adjacent impacts of an unknown event, we residents of White Sands would prefer not to rely upon arbitrary, uniform EPZ circles.
Accordingly, we would support the BG&E request only if the Commission requires BG&E to notify those " landlocked" residents of White Sands situated outside the two-mile EPZ as if we were inside the uniform circle. In this manner, we could make our own decision about whether to not worry #
and stay put -- or to evacuate until all doubt has passed.
- 8609230178 860910 PDR COMMS NRCC _ _
CORRESPONDENCE PDR
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F Nuclear Regulatory Commission March 10, 1986 ,
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W4 re We would respectfully request tha't this , letter be made a gi We would also appreciate receiving a copy part of the hearing.
pc of the Commission's final disposition regarding this matter, pr al Very truly yours, cu, n e.
Gary D. Knight cc:
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Agenda-Calvert Cliffs EPZ Reduction Meetin6 BG&E/NRC April 2,1936 COFFEE.
~
- 1. Introduction (Montgomery) 9:30 o Purpose of meeting _
o Current status
- 2. Philosophy of November 13 BC&E Submittal -
9:45 (Montgomery and Mirsky)
- o. Basis for exemption o Meeting the underlying purpose of the rule (NUREG-0396) '
for design basis accidents for severe accidents o New material circumstances regarding severe accidents (updated source terms)
~'
- 3. Updated Source Terms - Background 10:30
_ (Gardner and Warman) .
o History -
o Major reports o Present status LUNCH 12:00
- 4. Severe Accident Source Terms for Calvert Cliffs 12:30 (Metcalf and Warman) o Sequence selection o Comparison of plant features .
o -Sequence frequencies o- Source terms (NUREG-0956 and V-Sequence)-
o Summary of inputs to analysis
- 5. Consequence Analysis: Establishing an 2:30 Appropriate EPZ Distance (Mirsky) o Description of CRAC Code -
o Assumptions and Results
- 6. Impact on Emergency Plan (Forgette) 3:30 o Changes to plan o' Changes to response -
o Contacts with county and state Attendance Dave Jaffe (NRC)
Druce Montgomery (BC&E)
Steve Mirsky (BC&E)
Tom Forgette (BG&E)
Dick Gardner (S&W)
Ed Warman (S&W)
Jim Metcalf (S&W)
John Neumann (BG&E) 1
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PURPOSE OF MEETING o BASIC PHILOSOPHY OF SUBMITTAL -
o HISTORY AND STATUS OF SOURCE TERM ISSUES ,
o APPLICATION OF SOURCE TERMS TO CCNPP o SELECTION OF APPROPRIATE EPZ.
o IMPACT ON EMERGENCY PLAN 9
O M
e
s -
CURRENT STATUS o DENTON LETTER DATED FEBRUARY 14 HOLD REVIEW FOR NUREG-0956, IDCOR/AIF o BGaE REPLY DATED MARCH 27
, WILL WAIT FOR NUREG-0956 DECOUPLE FROM IDCOR/AIF, NUREG-IISO t-0FFERS MEETING WITH D'ENTON o NRC SCHEDULE (REPLY TO REPRESENTATIVE BEVILL)
BEGIN EARLY '87, FINISH LATE '87 HOLD REVIEW FOR NUREG-il50 AND COMMENTS o IMPLICATIONS OF NRC SCHEDULE e'
e
_ . , . , , . , , . . ,y. -- . , _ - ,y - , , . , ,--. , . , -
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CHANGING THE CCNPP EPZ LICENSING APPROACH SPECIFIC EXEMPTION UNDER 10 CFR 50,12
~ 10 MILE EPZIS NOT NECESSARY TO ACHIEVE THE UNDERLYING PURPOSE
_ 0F THE RULE NEW MATERIAL CIRCUMSTANCES NOT CONSIDERED WH. N THE RULE WAS ADOPTED H0W.THE NEW CIRCUMSTANCES (SOURCE TERMS) APPLY TO THE RULE AND TO CCNPP -
BASED ON NEW SOURCE TERMS, UNDERLYING PURPOSE OF RULE IS MET WITH 2-3 MILE EPZ 4
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_ , , , ._.,...m__ , ,, . ,___,__ .- - -, ,_ -, , -_ _.m . _ . _ . , . _ ., , , , _ - .
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CHAN@NG THE CCNPP EPZ
~
ANALYTICAL APPROACH
. SEEK THE ADVICE / CONFIRMATION OF EXPERTS ON PROPOSED -
APPROACH
, RELY AS MUCH AS POSSIBLE ON WIDELY ACCEPTED RESULTS USE NRC-FUNDED OR NRC-APPROVED STUDIES
, CONSIDER PLANT-SPECIFIC FEATURES T0: .
o SHOW THAT THERE IS NO REASON WHY THE NEW SOURCE TERMS SHOULDNT APPLY TO CCNPP 5 DEVELOP PLANT-SPECIFIC SOURCE TERMS WHERE GENERIC SOURCE TERMS D0NT EXIST OR WHERE APS, NRC, ETC, SAID SOURCE TERMS ARE PLANT-SPECIFIC DONTINV0KE OR INVITE NEW RESEARCH e
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50% , ,
0 10 20 30 40 DISTANCE IMILES!
l I l t t 0 5 10 15 20 APPROX. TIME OF CLOUD ARRIVAL (HOURS)
Figure 18. Centerline Dose Versus Distance for Licensing Calculation of DBA/LOCA at 2 Hours Assuming 5 Percentile Meteorology end Straight Lino Plume Trajectory.
50% Curve is Median of 67 Actual Site Calculations 10% Curve is Highest 10% of Calculations
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,3 I-36 .
E. Emercency Plannino Consideration Derived ft~1.
TheReacterSaketvStudy OfASH la001 The Reactor Safety Study (RSS) attempts to provide a detailed quanti,tative assessment of the probability and consequences of.
" Class 9" accidents. The study conc'uded that the public risk from nuclear reactor accidents was dominated by accidents in which there was substantial damage to the reactor core and 1 that the p'robabilities of such accidents were very small.*
Since emergency planners are encouraged to develop response' plan; '
which will be flexible enough to respond to most accident ,
situations, some understanding of " Class 9" accidents and the relationships between them and emergency pl'anning is needed.
The Reactor Safety Study developed the mathematical techniques and data base to provide an understanding of these relationships.
To obtain an appreciation for the distances to which or areas within which emergency planning might be required, a perspective on the relative probabilities of certain critical doses as a function of distance frem the power plant for these accidents
- Probability of a " core-melt" accident was estimated to be approxi-mately 1 in 20,000 ($ x 10-5) car reactor year. There is a large uncertaintv on tnis numcer.
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, , I-37 is needed. 'A set of such curves has been prepared for cli of the RS5 accident release categories (figure I-11). These
, curye's include both Pressurized and Goiling Water Reactor ( A4R
& SWR) accidents. Ooses are given for the critical values for which emergency planners should be concerned. One and five rem whole body doses correspon1 to the lower range of the PAGs; 50 rem whole body corresponds to the dosage at which early illnesses start to occur; and 200 rem whole body is the dose at which significant early injuries start to occur. As can be seen from figure I-11, core melt accidents can be severe, but the orobab'ility of large doses drops off substanti-ally at about 10 miles from the reactor. Similar conclusions can be reached by evaluating the otrer critical organs of lung and thyroid shown in figures I-12 and I-13, respectively.
For the lung, the doses of 5, 25, 3C0 and 3000 rem were plotted
.as a function of distance and probability of occurence. For the thyroid, the reference doses of 5, 25, 300 rem, which correspond to the Icwer and upper PAG 1evels, and the guide-line exposure used for siting purposes are presented.
Given a core melt accident, there is about a 707. chance of exceeding the PAG doses at 2 miles, a 40% chance at 5 miles.
and a 305 chance at 10 miles frem a power plant. That is, t
the probability of exceeding PAG doses at 10 miles is 1.5 x 10-5 e
72 .
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1 to 100 tc00 OtSTANCE IMILESI Figure 111. Conditional Probability of Exceeding Whota Dody 00sc Versus Distance. Probsbilities are Conditional on a Core Melt Accident (5 x 10,5),
ishole body dose calculated includes: external dosn to the whole body due to the passing cloud, exposura to rcdionuclidns on ground and the desrto the whole bcdy from inhaled radionuclides.
Dose calculations assumed no protective actrons tab en, and straight tina plume traisctory.
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- 7. .
e I-41' per reactor year * (one >; hance in 50,00) ner reactor-year) frcn the Reactor Safety Study analysis.
Based'in part upon the above informatian the Task Forco judged
~
that a 10 mile plume EPI would be aonropriate to deal with core m.elt accidents.
Potential ingestion doses to the thyroid (through the cow / milk
, pathway) from ccre melt accidents are afven in ' figure I-14 The distance for which emergency plann.r.g is needed is not easily ,
determined from the information given in the figure. It is 3
evident that deses can potentially be cuite high out to ,
considerable distances.
The current PAG for milk ingestion is 30 rem thyroid to an individual and 10 rem thyroid to a suitable sample of the population (usually calculated on the basis of an infant's thyroid). Given a core melt accident, there is a near 100% chance of exceeding the 10 rem thyroid PAG frem milk ingestion at I mile. about an 80% chance at 10 miles and a 40%
chance at 25 miles frcm a power plant. A planning basis for milk ingestion on the order of 25 miles would therefore approximately correspond to the 10 mile plume exposure distance
- There is a large uncertt inty on this number.
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CALVERT CLIFFS and NUREG.-0396 ' Calculation 2 hr. Whol'e Body Dose for Licensing . Calculation
. of DBA/ LOCA or 2 Hours Assuming 5 Percentile Meieorology and Straight line Plume Trajectory
. ( Reproduced from Fig.1-8, NUREG. -0396)
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! THE SOURCE TERM
- THE QUANTITY, TYPE, AND TIMING OF POTENTIAL RELEASES OF
! _ RADIOACTIVE MATERIAL i '
! TO THE ENVIRONMENT l
DESCRIBES THE SOURCES OF RADIOACTIVITY FOR USE IN CALCULATIONS OF ACCIDENT CONSEQUENCES i
OC & T 99 & #% , 1
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- Source Terms in the Regulatory Process i 1960 1970 1980 1990 ,
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i Sandse i
Siting Study NUREG/CR-2239 a
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} Source Term
! Reassessment j NUREG-0712 k
- Severe Accident Research Program j
j y SNL. Accident
{
i 4 Consequence Study WASH-740 ~
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Reactor Safety SourceTerm .
' Study Reassessment Study WASH 1400 NUREG-09Ei6 TID-14844 in-Plant Release Assumptions
\ \.
Source Term i Containment WASH-1400 Performance Credit Release Tables Analytical Procedure i
Siting Regulations Emergency Planning
- 10CFR100 Environmental impact Manned Regulatgry 1 Regulatory Guides Statements 3*PN"*"'*'I'"
Design Basis Accident PRA Analyses Assumptions
- l l
Figure 2.1 History of source-term assessment and' relationship to regulatory process.
! i
, e' J COMPARISON OF SOURCE TERMS i 1 4
PERCENT OF CORE INVENTORY RELEASED TO ENVIRONMENT l BATTELLE SWEC i ~
NUREG-0956 TID-14844 REVISED PROPOSED WASH-1400 SANDIA (DRAFT) DESIGN BASIS INDIAN INTERIM
! FISSION PRODUCTS PWR-2 SST1 TMLB-y SEO. SOURCE TERM POINT SOURCE TERM NOBLE GASES (Xe & Kr) 90 100 100 100 100 100
l TELLURIUM-ANTIMONY 30 64 50 1.0 0.1 1.0 4
BARIUM-STRONTIUM 6 7 1 1.0 0.1 1.0 i
i RUTHENIUM 2 5 0.08 1.0 0.1 l 1.0 LANTHANUM 0.4 0.9 0.2 1.0 I
0.1 ,
0.4 v
g m= =ee ee
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4 SOURCE TERM REASSESSMENT
- EARLY EVENTS
!I TMI 3/79 i
l STRATTON, MALINAUSKAS, CAMPBELL LETTER 8/80 EPRI SYMPOSIUM 11/80 BABBITT LETTER 12/80 NUREGs 0771 AND 0772 6/81 l PUBLIC UTILITIES FORTNIGHTLY LETTER 5/82
! ANS SPECIAL COMMITTEE FORMED 6/82 ICONTT ll- E.A. WARMAN PAPER 11/82
. ASTPO FORMED IN NRC I
12/82 '
FIRST BMI- 2104 PEER REVIEW 1/83 9
, t' '
l SWEC SOURCE TERM ACTIVITY ,
! PROJECT i
l
- WORK INITIATED
- 1981
! PROJECT ESTABLISHED: EARLY 1983 AVERAGE EFFORT,1983-4:, 15 EQUIVALENT FULL TIME BREADTH OF EFFORT: SEQUENCE ANALYSIS,
- THERMAL HYDRAULICS, AEROSOLS, STRUCTURAL ANALYSIS, CHEMISTRY, OFF-SITE CONSEQUENCES i
! STEERING COMMITTEE . ,
-l l I
CONTINUING CONSULTANT REVIEWS l . .
i 4
l SWEC SOURCE TERM ACTIVITY PARTICIPATION IN THE NATIONAL EFFORT
- NRC PEER REVIEW PROCESS (BMI-2104)
! SUBSTANTIVE PORTIONS OF ANS SPECIAL COMMITTEE i REPORT l PUBLICATION OF ORIGINAL PAPERS ,
BRIEFINGS OF APS STUDY GROUP PARTICIPANT IN NRC CONTAINNIENT LOADS -
j -
WORKING GROUP l BRIEFING OF OECD / CSNI SPECIAL SOURCE TERM i
! GROUP IDCOR PROGRAM EXPERT AND SPECIAL SOURCE TERM
, REVIEWS
! AIFiEMERGENCY PREPAREDNESS SUBCOMMITTEE .
INDUSTRY AD HOC COMMITTEE ON SOURCE TERM POLICY
i SWEC SOURCE TERM ACTIVITY l CONTRACT WORK I
! EPRI - OECD / CSNI GROUP OF EXPERTS ON SOURCE TERMS - AEROSOL CODE EXERCISE i EG&G - COST IMPLICATIONS OF SOURCE TERM i
IDAHO REDUCTION
- EPRI - METHODOLOGY AND APPLICATIONS FOR i
SEVERE ACCIDENT ANALYSIS PROGRAM: CALVERT CLIFFS, OCONEE, SUSQUEHANNA ANALYSES IDCOR 85 - SOURCE TERMS AND EMERGENCY PLANNING TASK ZION AND SEQUOYAH ANALYSES RATIONALE FOR REDUCTIONS IN EMERGENCY PLANNING 8
DOCUMENT THE NEW TECHNICAL BASIS' FOR REDU'CED SOURCE TERMS BALTIMORE - TECHNICAL SUPPORT FOR EPZ REDUCTION G&E
I i
KEY SOURCE TERM REPORTS
- o BATTELLE COLUMBUS STUDY - BMI-2104 - JULY 1984 I o ANS SPECIAL COMMITTEE ON SOURCE TERMS - NOVEMBER 1984
- o IDCOR TECHNICAL
SUMMARY
REPORT - NOVEMBER 1984 o APS STUDY GROUP - MARCH 1985 i.
- o NRC REASSESSMENT - NUREG-0956 (DRAFT) - JULY 1985
- . o OECD/CSNI SPECIAL TASK FORCE REPORT - 'IN PUBLICATION l o IDCOR TECHNICAL REPORT -~ REASSESSMENT OF IN PUBLICATION 0F EMERGENCY PLANNING REQUIREMENTS WITH f
i -PRESENT SOURCE TERMS ,
I o NRC - SEVERE ACCIDENT RISK REBASELINING/
IN PREPARATION
. RISK REDUCTION (SARRP) - NUREG-1150 i !
i 1 .
t e
9.
g MAJOR FINDING SOURCE TERMS HAVE BEEN OVER- ,
~
ESTIMATED BY LARGE FACTORS BOTH IN GOVERNMENT AND INDUSTRY
' PUBLICATIONS. WITH A SMALL NUMBER OF EXCEPTIONS, SOURCE TERMS CAN BE REDUCED'BY MORE THAN AN ORDER OF MAGNITUDE TO SEVERAL ORDERS DF ,
. MAGNITUDN. NOBLE GASES ARE EXCEPTIONS. -
1 t
- t. . .
COMPARISON OF RESULTS THE WASH 1400 PWR 2 AND BWR 2 RELEASE CATEGORIES ARE COMPARED WITH 60 ANALYSES MADE BY 9 ORGANIZATIONS
. FROM 4 COUNTRIES CONSIDERING 21 VARIATIONS ,
OF 4 FAMILIES OF SEQUENCES
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IN 11 PLANTS -
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COMPARISON OF CALCULATED IODINE RELEASES. .
TO ENVIRONMENT FOR PWR ACCIDENTS INVOLVING RELEASES INTO CONTAINMENT to -
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PRE. EXISTING OR EARLY LATE o.3 CONTAINMENT BREACH CONTAINMENI SRfACH z -
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5 g mM WASH BMI SWEC IDCOR CIA KfK BMI SWECIDCOR NYPA KlK 1400 (8) (8) (2) (2) (1) (1) (4) (3) (2) (2)
PWR.2
- ORGANIZATION PERFORMING ANALYSIS-(NUM8ER OF ANALYSES)
G,
d COMPARISON OF CALCULATED IODINE RELEASES TO ENVIRONMENT F.OR BWR. MARK ]K ACCIDENTS INVOLVING RELEASES INTO CONTAINMENT Lo -
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(1)
BWA 2 ORGANIZATION PERFORMING ANALYSIS (NUMBER OF ANALYSES)
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LEGEND
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r l
7.-' TWO SPECIFIC AREAS REQUIRING ~
}C2 FURTHER STUDY i . o MARK-I AND II BOILING WATER REACTORS i
e LARGE DISCREPANCY IN LIMITED
{ DATA LARGELY ATTRIBUTABLE TO l MODELING DEFICIENCIES l '
e RETENTION IN REACTOR VESSEL l INTERNALS - IMPORTANT .
O INTERFACING SYSTEMS LOCAs (E.G.
Y SEQUENCE) l e VERY PLANT SPECIFIC 1 '
i e RETENTION IN RCS & INTERFAblNG l SYSTEM IMPORTANT*
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, ,- i APS CONCLUSIONS
! 1 i
- BOTH SCIENTIFIC AND CALCULATIONAL PROGRESS MADE SINCE RSS i
e SOME NEW CALCULATIONS INDICATE LOWER SOURCE TERMS BECAUSE:
CONTAINMENTS ARE STRONGER THAN ASSUMED l lN RSS .
PREVIOUSLY NEGLECTED PHENOMENA HAVE BEEN INCLUDED ADDITIONAL RETENTION SITES HAVE BEEN INCLUDED .
i i
i i
APS CONCLUSIONS 2
e FOR MOST SEQUENCES AND MOST RADIONUCLIDES, l PHENOMENA CONSIDERED SINCE RSS REDUCE THE I SOURCE TERM .
! e BUT NON-VOLATILE RELEASES FROM THE CORE-l CONCRETE INTERACTION MAY BE LARGER
- STUDIES HAVE NOT TREATED ALL TYPES OF REACTORS NOR ALL TYPES OF CONTAINMENTS IN EQUAL DETAIL.
"lT IS IMPOSSIBLE TO MAKE THE SWEEPING GENERALIZATION THAT THE CALCULATED SOURCE '
TERMS FOR ANY ACCIDENT SEQUENCES INV'OLVING ANY REACTOR PLANT WOULD ALWAYS BE A SMALL FRACTION. . . ." .
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APS COMMENT ON SEQUENCE SELECTION ;
. . . NRC- AND ITS CONTRACTORS HAVE SELECTED THE . '
SEQUENCES REASONABLY WELL.'HOWEVER, SEVERAL OF 4
THESE SEQUENCES NO LONGER APPEAR TO BE' RISK DOMINANT, AND OTHER SEQUENCES HAVE BECOME RELATIVELY MORE IMPORTANT. IN ORDER TO MAKE SURE
) THAT THE RISK DOMINANT SEQUENCES HAVE BEEN ,
ADEQUATELY IDENTIFIED AND INVESTIGATED, WE .l l,
STRONGLY URGE ANOTHER ITERATION. . . . SEQUENCES THAT MIGHT BE CONSIDERED INCLUDE CONTAINMENT-ISOLATION FAILURE AND CONTAINMENT BYPASS i SEQUENCES. . . ." .
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f J
I Program Relationships Research Regulat!on
( m,, 7 g 3 l 885-29 4 y
( l Funeenainey esu se.7 I SAND 84-0410 y I
I F St .aosv.ud.uon 3 ,
l k oRNum-es.2 ) Source Term l
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} Reassessmasd % ! = Source Torne-Based
{f Connainment Working Group Loads 3 l i
l
(- NUREG-1079 ) g i l l
l(Connainmene working Group Performance) y n
( NUREG-1037 ) I I
Savne Accian r 3 Risk N) I .
- j APS Review l Severe Accident
! ( ) Risk Reduction j ;
Regulaim1f_ _
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! STATUS OF SOURCE TERM TECHNOLOGY I
i i (1) MAJOR PROGESS SINCE THE REACTOR SAFETY STUDY
! (2) AREAS WHERE SOURCE TERM INFORMATION IS SUFFICIENT l
l (3) AREAS OF DISAGREEMENT BETWEEN CURRENT STUDIES i
j (4) RECOMMENDATIONS FOR DEALING WITH UNCERTAINTIES
! (5) APPLICABILITY OF SOURCE TERM TECHNOLOGY FOR VARIOUS PLANTS 4
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(1) MAJOR PROGRESS SINCE THE REACTOR SAFETY STUDY (WASH-1400) f 1 0 CONSIDERABLE PROGRESS REALIZED AT COST OF CONSIDERABLE COMPLEXITY
)
O SOURCE TERM IECHNOLOGY IS VASTLY IMPROVED OVER RSS FOR ALL PLANT TYPES
~
O PROGRESS HAS BEEN UNEVEN, WITH MUCH MORE KNOWN'ABOUT LARGE DRY PWRS.
4 j 0 SOURCE TERMS ARE NOW LOWER BECAUSE OF:
MORE REALISTIC TREATMENT OF CONTAINMENTS l
l INCLUSION OF PREVIOUSLY NEGLECTED PHENOMENA INCLUSION OF ADDITIONAL. RETENTION SITES ,
i i
i I
- I l
I
,e -
i j CONTAINMENTS FAIL. IF AT All. AT LATER TIMES l
A) IT IS NOW GENERALLY BELIEVED THAT LARGE SCALE FAILURES OF REACTOR
! CONTAINMENTS WILL NOT OCCUR UNTIL THEIR YIELD STRESSES ARE EXCEEDED l -- AT INTERNAL PRESSURES ABOUT 2 1/2 TIMES GREATER THAN NOMINAL DESIGN PRESSURES. (BY CONTRAST THERE ARE MANY ACCIDENT SEQUENCES IN THE RSS
! IN WHICH LARGE SCALE EARLY FAILURE WAS ASSUMED.)
B) ACCIDENT-INDUCED PRESSURES ARE NOT EXPECTED TO EXCEED YIELD STRESSES l
UNTIL MANY HOURS AFTER THE REACTOR PRESSURE VESSEL MELT-THROUGH, AND
! STEAM EXPLOSIONS LARGE ENOUGH TO CHALLENGE THE CONTAINMENT ARE NOW CONSIDERED VERY UNLIKELY. ,
l l
i i C) DELAYED CONTAINMENT FAILURE ALLOWS TIME FOR NATURAL, PASSIVE MITIGATING l
PROCESSES TO ACT.
l D) DELAYED CONTAINMENT FAILURE ALLOWS TIME FOR OPERATORS TO ACT TO I
MITIGATE THE COURSE OF THE ACCIDENT.
l
I '
- i' I
I SIGNIFICANCE OF TIMING OF CONTAINMENT FAILURE I O NO CONTAINMENT FAILURE - ESSENTIALLY No SOURCE TERM.
4 0 LATE CONTAINMENT FAILURE (1.E., HOURS AFTER CORE DAMAGE) - VERY LOW SOURCE TERMS, (E.G., SMALL FRACTION OF ONE PERCENT).
1 0 EARLY CONTAINMENT FAILURE - NOW CONSIDERED EXTREMELY REMOTE POSSIBILITY WITH MUCH LOWER SOURCE TERMS THAN THE RSS WITH i COMPARABLE ASSUMPTIONS.
1 0 FAILURE TO ISOLATE CONTAINMENT (I.E., PRE-EXISTING OPENING) .
SOURCE TERMS MUCH LOWER THAN THE RSS (COMPARABLE TO EARLY CONTAINMENT FAILURE SOURCE TERMS) 1 -
PATHWAY TO ENVIRONMENT VIA CONTIGUOUS SYSTEMS AND STRUCTURES ,
COMBINED PROBABILITY OF ACCIDENT SEQUENCE OCCURRING COINCIDENT ,
WITH FAILURE TO ISOLATE CONTAINMENT IS VERY LOW, E.G., LESS THAN I 1 X 10-7 PER REACTOR YEAR. .
1 t
I
' i N
A
. M S E H U A H P I W T N E S I S T E S N H L S G C A H I T O N G I S E I D , I T W O R T H R U
N S H R A A O S E - D .
G M P . T A H E O S) C -
T T N N T R NS U S I N O I R F AC RR D
O A W E I M T D E O N S T( R , D N N I I T P S E I E C - M E T A T F N U TE N S A T E O I D C T O A I N R O U S I G C O S S R D Y S O C T M A P O S S E S C
- R R I L S N U O E N P T F B A I D e A F V O N O O N A I N A F N L R R E L H S O L O . . A E P M A E S I O G N S V T
. O C B I S O N N A E O A N '
N I F S C O I H R M . W OI E M T I I T T U E) .
H E O F F R T A T R S F S P H N O O A E R C K O S C D T Z H E U L N I F S . E N C I H R O I N F .
O T ES S A A R Y T T S S O N OS A E O A O S O I N T A DR E S R P C R T T O N N L C A E , S E A I E I E E E I E V D S U A E S S M M N H R L H E T O ( H O T T O I T U T R T C U P C A D D F A C U G S N S E E E O N O R N F U D I I O I F R R I I D I O D O T S D F T P S Y H O R N E N E T T E H TI S R P O R O E D E A N T I P C O I H E E H H E A L W S N H T T H V T T T R A Y N O D P A T O X MN L O I N O S F R T D E D R O A I S A I N O D 2 A N E E I N S S S E N l
f H A E S H T A SI I T U D T A I T G A . T N F N F N N H R B L E F F E F O E S NO A E F T O F M I C M T O S L Y U O E O O N D T N I C L F R N T I M A E T A L G O G A F A E D I D A E N F I E N T O E R I N N O T R I O T U I N T T C G A T N U L A D L O N S C
. O E T P S R E C O D A CI D M A U I OS D I N E E S E I R O S PC O E S I V E RC M R E C Y RR MH U O R U E P L O T L W R E E E S P M E A C E E C O P V H R S X E H N N H H F N L M E T A A E T I A I T T I F I S 0_
n I O O 0 0 0 O 0
- I,l.! i! i \
,e -
INCLUSION OF ADDITIONAL RETENTION SITES O RETENTION WITHIN THE RCS 1 0 SUPPRESSION POOL SCRUBBING MODELS VERIFIED BY EXPERIMENTS O IMPROVED MODELS OF RETENTION IN ICE BEDS O MULT! COMPARTMENT ANALYSES OF CONTAINMENTS AND PWR AUXILIARY BUILDINGS AND BWR REACTOR BUILDINGS.
O THE IMPORTANCE OF THE DISPOSITION OF WATER IN CONTIGUOUS STRUCTURES FOR THE CONTAINMENT BY-PASS SEQUENCES, INCLUDING THE POTENTIAL FOR FLOODING OF THE PIPE BREAK LOCATION.
O THE ROLE OF FIRE PROTECTION SPRINKLER SYSTEMS IN FISSION PRODUCT RETENTION.
I
FIGURE D.3 RELATIVE CONTRIBUTION TO WHOLE BODY EXPOSURE RISK FROM W ASH.140C PWR RELEASE CATEGORIES 10 _ _ _
~_. _............._._ ... _
PWR 2 . . _. .
PWR.3
..... .. . l 0I , P W R . ,. .-
Z .__
O C 01 .
3 - . ..
co ____
PWR 4 cm .
- **- -- --- ~ ~
Z -- - - --
O . ._. ...
pyR. 5 P W.R.. 6... - - . . . ...
y .
a
- C 00. _ . . _ _ ,
PWR.8 cs ,_, . . , . . ,
P WR .7 '
l 4
v 0001 . _ . _ _ . __ . .
~
DWR9
. nd af, w,.,.W ,Wg I RELE ASE C ATEGORY I f s
f i
?
i .e
._.--.__,,,.-,-_.__...._w. _ __-- _ - ,m __.,.,,,_,.,,_.,.___,--.,.,___,-,__m_ _
m .J~ - +A--_.- w-t 6
FIGURE D-4 RELATIVE CONTRIBUTION TO WHOLE BODY EXPOSURE RISK FROM WASH-1400 BWR RELEASE CATEGORIE .
N .
1.0 SWR-3 BWR-2 D
0.1
,,W,,
Z 9
- 5 E .
z O 0.01 l
s nWR-4 ac 0.0 01 BWR-5 1
0.0001 RELEASE CATEGORY 85-44605 l esw-- w wg+wan-- ,eg-.rwp.3-- -w+-wewwww--,w.-,.------ggww
- i:
i
- 12) AREAS WHERE SOURCE TERM INFORMATION IS SUFFICIENT l
l 0 THERE IS SUFFICIENT INFORMATION TO DEMONSTRATE THAT MOST CONTAINMENTS ARE STRONGER THAN PREVIOUSLY THOUGHT.
- 0 THE INVENTORY OF RADIOACTIVITY IS ASSOCIATED WITH RELATIVELY SMALL l UNCERTAINTIES, AND INFORMATION IS CONSIDERED SUFFICIENT.
I O THE RELEASE RATES OF FISSION PRODUCTS FROM THE FUEL ARE, FOR THE
- MOST PART, GENERIC TO DIFFERENT LWR REACTORS. EXAMPLES OF DIFFERENCES
- ARE THE AMOUNTS OF ZIRCONIUM, THE COMPOSITION OF CONTROL RODS (E.G.,
AG-IN-CD, B4C, AND HF ARE ALL USED IN LWRS), AND THE PRESENCE OF BORIC l, ACID IN PWRS.
i 0 INFORMATION IS SUFFICIENT TO CONCLUDE THAT THE DOMINANT FORMS OF THE VOLATILE FISSION PRODUCTS IMMEDIATELY FOLLOWING RELEASE FROM THE FUEL
]
ARE CSI, CSOH AND TE IN ALL TYPES OF LWRS.
l O MOST IMPORTANT IN-VESSEL THERMAL HYDRAULIC PHENOMENA ARE RECOGNIZED AND i I HAVE BEEN INCLUDED IN THE IDCOR METHODOLOGY AND ARE BEING INCORPORATED IN THE NRC METHODOLOGY.
l 4
1
, e' (2) AREAS WHERE SOURCE TERM INFORMATION IS SUFFICIENT (Cour'n)
O NATURAL CIRCULATION WITHIN THE RCS IS RECOGNIZED TO BE IMPORTANT IN THE ANALYSIS OF TRANSIENTS IN WHICH THE RCS REMAINS PRESSURIZED PRIOR TO REACTOR VESSEL MELT-THROUGH. ' .
O CONSENSUS EXISTS ON NEARLY COMPLETE EARLY RELEASE OF I & CS FROM FUEL, TE MAY BE RELEASED IN-VESSEL OR EX-VESSEL (1.E., AFTER PRESSURE VESSEL MELTH-THROUGH). ,
0 EXISTING EX-VESSEL CODES ARE ADEQUATE TO PREDICT LONG-TERM OVERALL THERMAL-HYDRAULICS AS THEY IMPACT ON CONTAINMENT LOADING. PHENOMENA DESCRIPTIONS RELEVANT TO SHORT-TERM THERMAL-HYDRAULIC BEHAVIOR ARE LESS WELL DEVELOPED, ESPECIALLY WITH RESPECT TO THE EFFECT OF STEAM SPIKES, DIRECT HEATING, AND HYDROGEN BURNING.
O AEROSOL BEHAVIOR IN THE CONTAINKENT ATMOSPHERE IS ONE OF THE BETTER DEVELOPED AREAS IN SOURCE TERM METHODOLOGY, AND IS WELL SUPPORTED BY EXPERIMENTAL DATA.
! O INFORMATION ON THE LONG-TERM BEHAVIOR OF IODINE IN CONTAINMENT IS WELL DEVELOPED. ADDITIONAL INFORMATION ON RADIOLYSIS EFFECTS, IMPURITIES l
l AND ORGANIC IODIDE PRODUCTION IS BEING DEVELOPED.
O e l
l e
t'.
i i l
l~
(3) AREAS OF DISAGRFFMENT BETWFFN CURRENT STUDIES <
l 0 THE MATHEMATICAL MODELS REPRESENTING THE PHYSICAL PHENOMENA GOVERNING THE PROCESSESJ I
k '
i 0 THE CORRELATION AND SPECIFICATION OF THE PHYSICAL PROPERTIES OF THE VARIOUS MATERIALS OF INTERESTJ i
J 0 THE INADVERTENT OMISSION OF POSSIBLY IMPORTANT PHENOMENAJ ..
! O THE BOUNDARY AND INITIAL CONDITIONS RELATED TO THE SPECIFICATIONS OF THE ACCIDENT SEQUENCES AND THE PLANT GEOMETRYJ O THE NUMERICAL APPROXIMATIONS INVOLVED IN THE APPLICATION AND SOLUTIONS I 0F THE VARIOUS MATHEMATICAL REPRESENTATIONS OF THE PHENOMENA.
1
1 Fission Product Thermal Hydraulic l' Transport , ,
Behavior ,
g p______________,
!. ORIGEN MARCH l
! Fission Product d N Inventory in Fuel ! g , ,v; ,, ,,
l "" Reactor Coolant System, == 1 E Molten Core, and gl Containment a
ml l CORSOR l .
l i i E i :
I l Release Retained from Fuel l
g in Fuel l E
l l
I I lll i
I .
l la l, .'
TRAPMELT l MERGE l l
l Reactor Coolant System suo m. --- em -m .,, Detail d Temperature.
Pressure, and Flow in l*
Transport and g g Reactor Coolant System g I Retention u l
! l L____ _ _ _ _ _ _ _ . _ _ _ _ _l i l I i l ON l VANESA Release from m- Detailed Core- um)
Concrete Temperature l a Core. Concrete Melt i 4 l and interactions fl l l
=i , , 4 '
NAUA, SPARC, ICEDF l
l Containment Transport j and Retention g g
I I L_.___ _. _____._____.I 1 r ~
Release of fission products to the environment: Source Term j
Figure ES.1 BMI-2104 suite of codes as used in the source term reassessment.
i xviii 5
___ __________ ______-__ ____ _ _____ _ _ _________- - - - _ _ _ _ - .- _ . ~ _ .
e NRC CODE SulTE'(AS USED IN NUREG-!!5b)' ,
- . s !
, i , i' 4
NAUA
, , r,_ '
~
) Trar.gort & Retention la Contiguous Structures 1 ,
i p 7 b i, __
MARCH-3.0 TRAPMERGE
- Overall Detailed Temperature Reactor Coolant' System Pressure and Flow in NAUA, SPARC, ICEDF
~
Molten Core, and Reactor Coolant System -
,- Containment g Containment Transport ).,
' ""d "*E*"EI""
Reactor Coolant System 1 CORCON Treatment of Transport and Retention Detailed Temperature including Revaporization j and Interactions Prior to Vessel Meltthrough i .
A ORIGEN * '
VANESA , .
Fission IProduct Invent.in Fuel J > Release from Core-Concrete Melt O
e
, s' CODE SUITE USED BY SEW FOR CALVERT CLIFFS V SEQUENCE ANALYSIS I
MAAP - 2.0B ) THREED-ST NAUA GC)
Aerosol Transport Thermal Hydraulic Multinode Model of Thermal Hydraulics in Auxiliary in Auxiliary Building 5 Mass and Energy Release from RCS Building (incl. natural (incl. Fission Product convection and Release to retention / release) Environment) -
. i Core / Concrete ,
J
! Aerosols From .
BMI-2104 Analysis WIth CORCON/VANESA 1 e
i
I l
,t' DIFFERENCES AFFECTING IN-VESSEL REIFASES l
l 0 CORE MELT PROGRESSING (INCLUDING SLUMPING MODEL) l 0 FRACTION OF ZIRCONIUM REACTED O NATURAL CIRCULATION WITHIN THE RCS 0 COUPLED THERMAL HYDRAULICS / AEROSOL TRANSPORT AND REVAPORIZATION v
e t
l
l DIFFERENCES AFFECTING EX-VESSEL REIFASES e (AFTER VESSEL MELT-THROUGH) l 0 RATE AND MAGNITUDE OF RELEASE OF NON-VOLATILE FISSION PRODUCTS DURING CORE / CONCRETE INTERACTION.*
j 0 RATE AND MAGNITUDE OF RELEASE OF CONCRETE AEROSOLS.
i l 0 EFFECTIVENESS OF SCRUBBING BY OVERLAYING WATER.
O Ex-VESSEL RELEASE OF TELLURIUM (TIED TO FRACTION OF ZIRCONIUM REACTED IN-VESSEL).*
I PRESENTLY DOMINANT FACTORS FOR BWR MARK IS AND llS.
i I 4 .
i .
t ' .
l (11) RECOMMENDATIONS FOR DEALING WITH UNCERTAINTIES i
O LATE CONTAINMENT FAILURE - UNCERTAINTIES NOT SIGNIFICANT O EARLY CONTAINMENT FAILURE RANGE OF DEBATE IS NARROW FOR PWRS AND MARK 111 BWRS.
UNCERTAINTIES AS REFLECTED IN THE DIFFERENCES IN CURRENT STUDIES HAVE
) NOT TRANSLATED INTO LARGE DIFFERENCES IN SOURCE. TERMS.
) - RANGE OF DEBATE IS LARGE FOR MARK I AND 11 BWRS, INDICATING UNCERTAINTIES AFFECT CURRENT SOURCE TERMS VERY LOW PROBABILITY OF OCCURRENCE 1
O CONTAINMENT ISOLATION FAILURE UNCERTAINTIES ESSENTIALLY THE SAME AS FOR EARLY CONTAINMENT FAILURE PROBABILITY OF OCCURRENCE MORE PLANT SPECIFIC 0 CONTAINMENT BYPASS FAILURE LESS AFFECTED BY UNCERTAINTIES IN EX-VESSEL / RELEASES RETENTION IN CONTIGUOUS STRUCTURES DOM,INANT CONSIDERATION, THEREFORE, !
UNCERTAINTIES IN THAT AREA MOST IMPORTANT.
. an.. - O
-C
- W O o U at at *t3 9 at - t$
h e W '4 3 W W N W N *ts 4 W ts W o 9 9 W O *t$ W at W 3 z N 4
- W o O t$ W E at W 3 N & W Q. W N
- N t E W
< o N -t 4 N M D 4 W o t N
- e 3 9 t$ t$ W at *ts t cW t$
0 N 3 W E O W t$ O E w W U *ts o 5 N N s .t % 4 M 9' t O O c 2 C.M NWWW o t$ W N.
9 W o 43 9 h 4 %
- W *N N N N o W N t$ 3
- 9 . 3 N u) N w W '4 . e ts N o 43 ats W 9 *ts "ts W 3 W o N O 4 N g t$ R 9 9 E W t a*
N at 9 3 e g tg
, m 9 E
& W O h t$ N 'o W o 9 o o W U W Z at 9 W W ts -Q *ts at W w at t oN at
- '4 at m -D E Q. N 9 W ad U o aug, N &N N a* 2 t$ E at "t3 Q.
- w t$ -Q N 9 W h o t$ W W 9 2 W H W W W m *ts M O -Q W D N N C 9 W N t$ t$ "t3 W W *t$ t$ *t$ 5 3 N at > t$ W U N W & 3 E W W z o W N N W Z hN o 9 o *4 9 2 t$ c W E W h 2
- C N E ad N .* W t$ W -Q *d = t$ at t$ t$ Q. -O E = 4 t$ e a* o W Z z Q. W 4 N *
- O W o W W -2 C 41 E W d . W ad at e W W N 4 ts W o at 2o 5 o W t$ at & W N N ,
M 't3 O N W
- o Q. 9 W W O t$ 3 W Z
- O 3 o 9 H 9 w at E E 2 t:n 2 C t$ t$ 'e o' N mW W W w t$ 3 at t$ W N
- o W W E E O N O WE H NW 9 t$ N W W W *t3 9 W *eo 'e at act 6 O E O E E *
- W . 9
- O h W .
C:2 9 o 2 N o -o t$ W 4 WN o WW 2
- o N W N u 9 3-w c 4 -t W W h W D Q. -t W W N W W
- 4 E W *ts *ts E ts 3 *d Q. W 9 N j at t$ W N 3 o = = * - W -
- N. ._ W W Q. W N a* N N o t$ at *t$ -Q W t$ 4 O N U W t$ at e hw
- at b t WWN W Q. W at W 2 t$ t$ W g W W o N W3 M N
ts t$ Q. at E O A W ts *4 W OWW 9 -c = ts 'e W W WWW Q. W N at at W o O W W E W E L -C C C t$ c
.cd o 3 W t$ E 3 E 9 W W W # 4 W 3
- W t$ o O E o N 9 4 E ts E f E *ts W 3 *ts 1.- o * 'e W W N W E W W W W .* O
- s* t$ o W W 3 3 c:n N N e O N at H E O O at t$ t$ N 9 *ts W D. 2
- Q. W
-t o se 'a W -t W ts W 3 9 E N E o 'e W ad W 'e o o -Q W E D # *w W N
- C4 N N o N O
= a
, . - . . , , = _ ,.
, ,~
te APPLICABILITY OF SOURCE TERMS FOR VARIOUS PLANT TYPES O A SINGLE SET OF INCORRECT SOURCE TERMS (WASH-1400) IS IN USE TODAY TO REPRESENT ALL PWRS AND BWRS.
O IHESE (RSS) SOURCE TERMS SHOULD BE ABANDONED NOW AND BE REPLACED BY BEST ESTIMATE SOURCE TERMS FOR EACH PLANT-TYPE BASED ON THE BEST SOURCE TERM TECHNOLOGY AVAILABLE TODAY.
0 THAT TECHNOLOGY HAS PRODUCED SOURCE TERMS FOR LARGE DRY PWRS, MARK III BWRS, '
AND TO A LESSER EXTENT ICE CONDENSER PWRS, WHICH:
- ARE SUBSTANTIALLY LOWER THAN THE RSS CATEGORIES, ARE IN REASONABLE AGREEMENT BETWEEN VARIOUS GROUPS OF INVESTIGATORS ,:
WITH THE EXCEPTION OF CONTAINMENT BYPASS SEQUENCES, WHERE, I
THE RETENTION OF FISSION PRODUCTS IN CONTIGUOUS STRUCTURES IS GREATER l IN INDUSTRY SPONSORED ANALYSES THAN IN PUBLISHED NRC SPONSORED ANALYSES, NOT OWING TO FUNDAMENTAL DIFFERENCES IN TECHNOLOGY BUT DUE TO THE
' INCLUSION OF MULT! COMPARTMENTS AND OTHER PHYSICAL EFFECTS, WHICH HAVE NOT BEEN INCLUDED IN THE NRC SPONSORED ANALYSES, TO DATE.
t I;
4 4s
,e o I
! APPLICABILITY OF SOURCE TERMS FOR VARIOUS PLANT TYPES (CONT'D) .
)
O THE TECHNOLOGY IS SUFFICIENTLY ADVANCED TO REPLACE THE RSS SOURCE TERMS WITH NEW MUCH LOWER SOURCE TERMS FOR THESE TYPES OF PLANTS NOW. ,
l j 0 THE " RANGE OF DEBATE" HAS BEEN NARROWED CONSIDERABLY WITH REGARD TO THESE TYPES OF PLANTS, AND BEST ESTIMATE SOURCE TERMS CAN BE DEVELOPED BASED ON j
l THE DATA NOW AVAILABLE.
i l 0 FOR CONTAINMENT BYPASS SEQUENCES THE PRESENT STUDIES ARE MUCH ADVANCED FROM j THE RSS V-SEQUENCE TREATMENT, BUT ILLUSTRATE THAT PLANT-SPECIFIC DETAILS PLAY IMPORTANT ROLES IN THE SOURCE TERMS FOR THESE SEQUENCES.
O WITH RESPECT TO MARK I AND 11 BWRS, THE " RANGE OF DEBATE" HAS NOT BEEN
' NARROWED SUFFICIENTLY, TO DATE, TO RESULT IN AGREEMENT ON BEST ESTIMATE ,
SOURCE TERMS FOR THESE PLANT TYPES. HOWEVER, THIS IS ONLY A TEMPORARY SITUATION. ANALYSES BEING PERFORMED OR PLANNED, SUCH AS NRC'S NUREG-1150 EFFORT, SHOULD SUBSTANTIALLY IMPROVE THE KNOWLEDGE AND UNDERSTANDfNG OF THE
- SOURCE TERMS FOR THE ACCIDENT SEQUENCES FOR WHICH CURRENT SO'URCE TERM ESTIMATES DIFFER t SUBSTANTIALLY.
. I 1
~
36A16 HEVISED) -
i SOEDULE POR lillREG-1258_al5 IWut Suppelt V
{ ACTIVITY SBARY PEACll 8971151 SE88 Mall BMIS 98tF ZION kASAkhE ,
i
- 1. gg C C C C C C ,
i 2. SOURCE TEM C C C C C C
! 3 11511116 .
1
- 3. gRg gE 4 6 8 3 4 3
- 4. gETERMCODE C C C C C 4/25/86
! 5. RE1 EASE CHARACTER- C C C C C 5/0/96 l ISTICS.
l 6. ' CONSEQUENCE CALC. C C. C 3/4186 4/25186 5/29/95
- 7. R g g CIDENT C C C C C 5/36/86 '
, 8. gigTREES C 3/21/86 3/28/86 4/25/86 4/30/86 4/29/96 :
- 9. BASELINE RISK C 3/20/86 3/24/86 4/8/86 '
4/30/86 5 /2 5/31.
C6LCULATIoll 10.' RISK / RISK. RED. C 4/1/86 4/4/06 5/15/86** 5/16/86** 7/1/88*
o TA8LES -
4/22/86** 5/15/86** 6/13/86** 6/20/86 5/30/86 8/4/85
- 11. DETAILED RPT. RISK / RISK BED. DRAFT 'l
. RISK / RISK RED 6/18/86 7/14/86 8/8/06 8/18/86 7/30/86 9/29/80
- 12. DETAILED RPT.. FINAL .
..ci erare ermas trNtlTIVITY ANALYSIES AIID UllCERTAINTY RANGES
, e' BASIS FOR SEQUENCE SELECTION i-i i e Consideration of a spectrum of potential accidents.
i e Observation that severe core damage accidents dominate risk.
o Contribution of sequences to overall core melt frequency in Calvert Cliffs IREP study.
., e Availability of Containment Spray System to reduce source terms. ,
o Probability of Occurrence (AB sequences neglected).
4 O
S t ,
! t ,
] LEGEND OF SYM80LS USED IN DESCRIBING ACCIDENT SEQUENCES Initiating Events System Failures
]
- S2 = Small-small LOCA D" = High Pressure Safety injection l
(1.9" in dia.)
i Ty = Loss of Offsite Power F= Containment Spray System (Recirculation)
T High Pressure Safety Recirculation System 2 = Loss of PCS H=
T = Transients requiring primary K= Reactor Protection System 3
relief Tg = All other transients L= Auxiliary Feedwater System T = ss 5 VDC bus 11 M= Power Conversion System DC Q= Relief valves fall to reciose ,
U= Chemical Volume and Control 5,ystem C= Containment Air Recirculation and Cooling System ,
C'= Containment Spray System (Injection) f
- s :.
i e' ' ,
i CALVERT CLIFFS DOMINANT ACCIDENT SEQUENCES (From IREP Study - NUREG/CR - 3511) l REP 1 REP t FREQUENCEY FREQUENCY BEFORE AFTER % TOTAL RECOVERY RECOVERY CM
,) SEQUENCE DESCRIPTION (/YR) (/YR) FREQUENCY i
ATWS(PSF)
2.8E-5 2.8E-5 20 l,
S2-52 S FH 5 7E-5 -
1.lE-5 9 2
T2-82 TL 1.8E-4 7.lE-6 6 2
Tg-173 Tg KU 6.7E-6 6.7E-6 5 Tg-147 Tg ML 3 4E-4 6.3E-6 5 i
4 i Ty-81 .65 T yq-D"CC' l.3E-5 5 3E-6 2.4E-5 4.9E-6 4 i Ty -82 TLy Blackout ---- 2.4E-4 4.4E-6 3 Tg-152 TgKQ 4.3E-6 4.3E-6 3 ,
! T -139 T KU 3.7E-6 3.7E-6 3 l 3 3 T "I' 2 3E-6 2 3E-6 2 l 3 3 l T -ll3 T ML 8.5E-5 1.7E-6 1 3 3 S2.-59 52D" 2.8E-6 1.6E-6 i Ty -85 T jLCC' 5 9E-5 1.0E-6 , 1 Sequences 7.8E-6 6 l Below Cutoff ---- ----
l Total ----
3 8E-3 1.3E-4 100 ,
. (
, e' i
GROUPING OF ACCIDENT SEQUENCES FOR EVALUATION i
IMPORTANT CHARACTERISTICS FRE ENCY GROUP (%) DESCRIPTION RELATIVE GX) SOURCE TERMS ATWS 33 Anticipated transients without Containment sprays are available, scram 6. sequences with failure
' resulting in low source terms j'
4 of reactor protection system-TML 32 Transients with loss of feed- Containment sprays are available, water resulting in low source terms.
S2H/S2D" 12 Small LOCAs with loss of ECCS Containment sprays are available, resulting in low source terms S FH 9 Small LOCA with recirculation No containment recirculation-2 failure spray .
l i:
TMLB'* 3 Station blackout No containment Spray V ** Interfacing System LOCA Containment bypass
- Differs from traditional definition of TMLB' in the delayed failure.(af.ter about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) of the AFW system.
- Dismissed in IREP on probabilistic grounds (1 x 10-11 yr -1).
s . g
F
, r. ,
SUMMARY
OF ENGINEERING EVALUATION OF SELECTED SEQUENCES ATWS:
e Containment Spray System (CSS) & Containment Air Recirculation & Cooling l
r System (CARCS) remain available.
e BG&E implementation portions of ATWS role applicable to CE plants -
l o CCNPP Control Room Procedures call for:
i 1) Manual SCRAM
- 2) De-energization of motor-generator sets for SCRAM circuitry ,
i 3) Manual initiation of boron injection j .
- e BG6E has refurbished SCRAM breakers and improved testing & maintenance e Therefore, ATWS frequency measureable reduced from IREP study.
i l
9 O
L 9
?
i ;;
. g i i
SUMMARY
OF ENGINEERING EVALUATION OF SELECTED SEQUENCES-Cont'd.
i i TML:
s e Five sequences involving transient induced loss of PCS and subsequent
~
. loss of feedwater.
l' e Feed and bleed cool'ing not guaranteed with primary relief valves cycling.
following loss of secondary heat removal. .
j e ECCS makeup postulated to fail.
4 l
e Containment spray system (CSS) assumed to substantially redsfce source terms.
ri h
e i
F. ,
I, 4,
i e' t
SUMMARY
OF ENGINEERING EVALUATION OF SELECTED SEQUENCES - Cont'd.
J S2H/S2D": ,
e Small break LOCA (S2 ) assumed to occur.
e Reactor SCRAM and AFW initiation.
e High pressure ECCS injection successful for S H but not for S2 D".
i e Loss of reactor water inventory is assumed to be immediate for S2 D".
(Switchover at 4 to 12 hrs. depending on size of break) i e Core uncovery and fuel damage result from lack of makeup.
I
. e Containment Spray System (CSS) substantially reduces source terms for both
! t D". -
l f2H and S 2 ,
t l
t i
i t
i.
f-
- 5
1
, e'
SUMMARY
OF ENGINEERING EVALUATION OF SELECTED SEQUENCES-Cont'd.
1 S2FH:
i j e Small break LOCA (S2 ) assumed to occur.
l e Reactor SCRAM, AFW, and high pressure ECCS initiation.
j; i
)
e Decay heat removal via steam generators.
o Reactor vessel water inventory maintained. ,
l e At RWST depletion and switchover to recirculation:
- Containment spray recirculation assumed to fail (F)
ECCS recirculation assumed to fail (H)
~
(Switchover at 4 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> depending on size of break) i
. e Core uncovery and fuel damage result from lack of makeup.
e Containment Spray System (CSS) assumed inoperable due to failure of recirculation portion of system.
l}
l 1 l
e No credit taken for fission product removal by.CARCS, l
l .
. i
' t
t a
e'
SUMMARY
OF ENGINEERING EVALUATION OF SELECTED SEQUENCES-Cont'd. .
1
- Station Blackout:
) e Loss of offsite power postulated to occur.
c e Reactor SCRAM and subsequent loss of onsite emergency power. [
1 i
! e Power conversion system not available without AC power.
j e AFW continues makeup to steam generators - decay heat removal via relief valves.
e Subsequent to 4 hrs station de power postulated to be lost due to battery depletion.
e AFW fails on loss of de control power. ,
! e Core uncovery and fuel damage occur assuming ac power not restored. t
]
e CSS and CARCS assumed unavailable due to loss of ac power. -
i e qequence differs from traditional TMLB' in the delayed failure of AFW by at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, taking credit for availability of_de power.following loss of all ac power.
it
' b
e' ,
Station Blackout Cont'd.:
e Assumption of de power from batteries for > 4 hr based on loads under blackout :
conditions substantially lower than DBA-LOCA conditions, o Design basis capacity for de power supply is 2 hrs. ,
o Fully qualified reserve battery with additional 2 hrs capacity available.
e IREP study assumed 4 hrs.
e Actual test discharges - several hours. '
s e Separate NRC approved source of offsite power (SMECO) used for_ hotel services .
to administrative areas.
e SMECO power could be used to recharge station batteries.
I e Taking credit for SMECO power would reduce probability of station,blackou,t l below IREP study. .
I j ,
e .
I
SUMMARY
OF ENGINEERING EVALUATION OF' SELECTED SEQUENCES-Cont'd. !
i .i' 4
V Sequence Survey:
Pipe No. of Valve Elev. Within System /Flowpath Size (in). Failures Req. Break Room Conunents ,
}
l HPSI 6 4 Low HPSI pump discharge pipin; is code service class CC l
(2485 psig rating) or DC
' (1600 psig rating) 3 LPSI 6 4 High Includes one check valve which is normally open, f but which will close if l reverse flow exceeds 300 gym. ,
CVCS/ Letdown 2 2 High Letdown flow must pass !
through 2 air-operated valves.
CVCS/ Charging 2 3 Low Charging pump discharge 1 lines'are code service [
class CC(2485 psig rating l
The 3 valves are all chec.
valves.
SCS/Leddown 14 2 High Schedule 10 piping down-stream of 2 flex-wedge MO Disk failure character- !
istic of check valves !, .
assuned. ZionPRAfrequencyl
. . (1 x 10~ ) used to calculate consequences for CCNPP.
1
' s,
M I8.
Table 2-5 .
Compactsose of soney. 2sose. amo Catyt#f Cl.IFFS PLAfff. FEATURES ressure auct2 11 sit ceivert Cliffe ,
- t. Powoc (tout) ,
2440 3250 2700
- 2. N555 configuretten/owller . 3-leep M 4-leep M 2-toop CE *
! 3. Contetsument free welues (fge) g,g a ge* 2.8 x 10' 2.0 m SO'
- 4. Solnforced or poet-teneloned contefament Aetnforced Post-tenstoned Post-tensioned
- 5. Contatnment operating pressure (pete) to-ft Ate Ate
- 6. Containment deelge preneure (pete) SO 42 45 !
- 7. Contatnment fatture pressure (pele) 135 150 140
- 8. In-core Instrumentatten penetretten lecetten Gotten Settee Top
- s. Cawlty condition (given sprey failure) Dry Flooded Dry j tO. Conerete type Baseltla Ltesetone 5liIcoous * ,
l tt. Spr.y wit - t - - e ve. o
- 82. ser.y recire. sadopendent of ECCS recere. veo see v.e f
- 93. ESt cone.tne.at unit coeiere/ flit.c. me vea vos l
- 14. No. of high pressure welwee in ECCS discharge 2* 4 4*
f 15. ECCS too pressure line break lacetton submerged Vee feet inweettgeted See Note 5 -
- 94. SHR (or shutdown cooling) systee Inglde contaleusent vee $$e 90 s l 17. 'No. of high pressure welwee (n 909/$CS letdeun path SS/A 2* 2m l
- 88. SHe/SC5 letdown aeroeit locetten submerged st/A Bio too f!
- 89. Contiguous structure free volume (f t') 10.000
- 1.4 m 90* 2.5 m 10' Engineered Ausi1tery Aux 81Iery .'
5efoguarde mutidtag sullding Sullding - ,
(t) Two perellel cold lege per leep.
(2) CaCoe content comparetste to beoeftle. >
(3) IncIud4ng twe check weIwee and 8 SOW fn cold leg fIou path Iactied open.
(4) Including one normally open volghted check velve closing on 300 mesa reverse flou.
(5) High pressure pump suction submerged. Seu pressure pump discherp met submerged.
(6) Interlocked to preclusse open6ng wtth h00h RCS pressure.
(7) seel-2904 esaused apprestaately 200.000 ft' (apparently includig epsench sprey pump house and main eteem wolve house).
I 4
5
- .l.
i' II. , i i
DI AGRAMATIC REPRESENTATION OF SOURCE TERMS IN CCNPP ANALYSIS 4
j . .
I i
In Containment impaired Cont.(B) *P =0.005/yr (Represented by TML8'-S from NUftEG-0956)
No Sprays (S FH, Elackout) Early Overpressure (6e) P =0.005/yr(Represented by TML8'-6e from NUREG-0956)
- 2 e {
in Containment l With Sprays Hydrogen Burn (y) ** Pc=0.001/yr(Represented by S2 D-y from NUREG-0956)
(ATWS, TML, S H/Sy D")
2 Late Overpressure (
' r------- 3-) _
Low Source Term Sequences in Containment l -- Neglected in CCNPP Analysis '
With or Without e Basemat Heltthrough (c)
Severe Core Sprays *** '
Damage Accidents I l No Failure a_.-___ -_ - p I
Unsubmerged Release ,
Plant-Specific High Range V Sequence
' Containment Bypass (V)
Submerged Release Plant-Specific Low Range V Sequence
- From NSAC Oconee Study
- Coqtainment Failure at Time of Early Burn, Concurrent Loss of Spray *
- With Sprays Operating, Containment Failure Unlikely I
e f
i 4
1i
. L
i' 1 .
TABLE 2-6 ,
l 1 . PROBABILITY OP OCCURRENCE OF SEVERE ACCIDENT SEQUENCES AT CCNPP (Per Reactor Year) l Mode of CCNPP ,
Release from Probability Cont. Event Probability i Sequence Containment of Event . Tree Probability of Release !'
~
TMLB'-& e Early 1.5 x 10-Ma) 5 x 10~3ICI 7.5 x 10'I j Overpressure TMLB'-# Isolation 1.5 x 10-N*I 5 x 10*NS 7.5 x 10-8 Failure !
S 2 D-y Early 1.0 x 10-4(b) Ix10~NCI 1.0 x 10 ~7 l Overpressure V Bypass Via 1 x 10-7 N/A 1 x 10-7 i Interfacing !'
Syst. LOCA (a) Based on combining the 4.4 x 10 4 probability of a TMB' hequence with the 1.1 x 10~3 probability of an 5 2FH sequence from tl IREP Study for CCNPP.
i.
! (b) Based on combining the probabilities of all core melt sequences with containment sprays operating (ATWS, transients with loss .;
- FW and 52 D"/S2H) for the IREP Study for,CCNPP.
.(c) Based on NUREG-09%.
(d) Based pn EPRl/ NSAC analysis for Oconee (NSAC-60 Oconee PRA, . lune 1984). ,
. . j:
F L
. . . _ . - . - . . . _. . . - . . _ 1 10 DINE RELEASE FRACTIONS'FOR IN-CONTAINHENT SEQUENCES l
1.0 -
O.9 . ,
0.8 ,
0.7 .
3 E -
5 u
> 0.6 , ,
E .
E W
3 w 0.5 E
o z
C 0.4 - ,
13 1
w 03 .
i' 0.2 O.l
~~
X u
^
O
% e WASH- NUREG- IDCOR NUREG- NUREG-1400 0956 TMLB'-S 0956 0956
! PWR-2 TMLS'-S Zion TMLE'-de 5 2 0-y Surry Surry Surry r
IODINE RELEASE FRACTIONS FOR V SEQUENCES
- 1. 0 -
09 ,
0.8 .
~
- 0.7 , ;
O E
g 0.6 . .
E E
5 E 0.5 -
U 8
o 0.4 .
. E-u Y
u.
0.3 -
0.2 .
0.1 -
m
~
X
> 0 _
WASH- NUREG- NUREG- CCNPP CCNPP 1400 0956 0956 High Low PWR- 2 V Seq. V-Seq. Range Range UnSub. Submerged V-Seq. V-Seq.
Surry Surry
SELECTION OF METHODOLOGY FOR CCNPP-V SEQUENCE ANALYSIS o NRC Methodology Does Not Include Revaporization After Vessel'Heltthrough In-vessel o Use MAAP to calculate:
Release o MAAP Reasonably - Event Timing Approximates NRC - RCS Thermal Hydraulics
.In-vessel Methodology & - Mass & Energy Release Includes Revaporization i
o NRC Methodology predicts o- Use NUREG-0956 data for i higher release from ex-vessel release adjusted Ex-vessel for auxiliary building retention Release core / concrete interaction observed in CCNPP analysis.
than MAAP. .
. I o Neither NRC Methodology o Use Stone & Webster THREED-ST nor MAAP includes free code for Thermal Hydraulics in convective flow in Aux. Bldg. Aux. Bldg. - Incl. Free Convectic; Retention & effects of H2 burns.
In Aux. Bldg.
o NRC-NAUA and MAAP i aerosol models differ o Use NAUA for Aerosol Behavior in Aux. Bldg. with modification to treat spray removal.
I
s ==-
r (s i L t,a 1::s - -
l i,du c =
!, U, t-s y 1 3. 2 ~
llv. .
i ,
..k .
.f. 3 $ lg 'I g l
! .r 3 j 3YM t= '" U fiIg((Fl/E$ ls 8
-s s : . 1 a:*:an ~' :l . %'
3 .! 4 S j O
ti e re2 **
e [ [ g I, !
.T *
's i i. 1 [ {>
.' l \ r h A
,c A.g
- y i
v%; .
p#15
., y
=
\
-. - II g; l=
-4 I, %
if l , -
y s
- i. t--
r
- [
'i l
.I lr b '_C .dp p"IS
- a '
. i.
% h a .
3l, . .l li -.
' O -F -- -
- l 11
',l3 '= . j g..! E j cI s l
r.
m.
- g. ,9 , . , . g .
c a 4
jitj
- i. .
g <= si q& Ja
... a la .
Y j
h i,
n 1
.=_ y _.
gris 3/
i.. _.
4
, r 4,
gl . ., ,
~
f-N 'll l kA Y i,i
,:. 1 e
d Q ,3 55l w- 7 11 ,
pa.__f i
}Y&. Nj.ml[
i g ~ a m.. , y Ac fh
- ' I 1 fi ;' !D f L ;'
- ll - == Iv 8 ~
se i. J-1! -
- d
~
ilt a y i ./
s
'P #_ _
3 .g r -- =
jn u,s b
' .. l h
n t
I,:{ h+J:n[_ i u
gchr.asl__ >
m , s .c .. 3u _ -
_ ,, L' ~ ~ ~
-mm u- .v y, I1' l
h lN : Y-t.
g p ry N .C -S y 's v... ~ ~p 1.,,
I ,f~
l, k w,g :
]" a s co x - a. e i
r ,
w g -8~ 5 f.e il .
9 7 s m aammsi. % :l
(
.'i
! y ...
.E . I er g, y ,
-o . .m ,
g t' l
CALVERT CLIFFS V SEQUENCE I -
AUXILIARY BUILDING THERMAL HYDRAULIC H0 DEL t
I I .
t .
- L
\ l' s
v . - - _ _ - :
J , ,
<str
/// st
- e i s BREAK LEVEL ..
(
I I i (;
, .,. i, , J w ., _y J. ..
s s .. ,
p ,,
.-. o
~p 7 n, h'.
. CALVERT CLIFFS V SEQUENCE ~'
. - REVISEI) (11 N0DE) H0 DEL- .
a-x, u'
c e
~~ ,
' x: '..y A-i' d , 7, 1 1
- + , x::a
.=
- - \
, , p c7
- - - ,.: e J,7 h ,
u ,, .
4
, .}
= - - -
- . J '
.g .
,.. i .,
- . __ --- a .
, , , , , , ,- 9 , t r i a, 4
1 i
a
- u .
BREAK LEVEL >
SUBDIVIDED INTO 5 N0 DES .
3 9
I l
$\
i I t
, e' i
. i IMPORTANT FEATURES IN V SEQUENCE ANALYSIS 6 3 e Large Multicompartmented Auxiliary Building (2.5 x 10 fg ), ;
e Multi-level /Multicompartment flowpath to environment.
e Large Break LOCA (14 in.) - openings in building result in Free Convection pattern.
e Extensive fire suppression system including:
- Sprinklers on all levels in potential flowpaths. !
- Individually activated > 212 F.
- + 300 sprinklers on elevation 5' (break elevation)
- 63 sprinklers actuated during blowdown. ;
140 sprinklers actuated during accident sequence.
- 600,000 gal. minimum water supply.
q De-inerting results in H2 burns .
s Direct aerosol removal (washout) :
e Diffusiophoresis onto water droplets. ,.
i E _(
l CALVERT CLIFFS ,
V-SEQUENCE ANALYSIS
~ '
SEQUENCE m LARGE (14") m LOSS 0F AUX BLDG SELECTION PIPE BREAK INTEGRITY &
FREE CONVECTION I
- \.
SCONSIDERATION > LOSS OF STEAM INERTING ;
_- 0F FIRE PROTECTION & HYDROGEN BURNS -
SPRINKLER SYSTEM '
-> LARGE DROPLET SIZE ;
i
& SELECTIVE REMOVAL .
4 EARLY EXHAUSTION OF y SPRINKLER WATER SUPPLY
+ CONSIDERATION OF DIFFUSIOPHORESIS ON DROPLETS
-9 FINER N0DALIZATION >BETTER DEFINITION OF 0F AUX BLDG- AEROSOL REMOVAL
> LONGER OPERATION OF SPRINKLERS o
4
...-- x BG&E METHODOLOGY FOR GENERATING . ,
NOV. 18, 1985 EPZ SUBMITTAL FIGURE l-2 CURVE l 9
- 1. RUN CRAC2 CCNPP MODEL FOR EACH OF FOUR SOURCE TERMS INDIVIDUALLY .
~
- 2. GENERATE CONDITIONAL PROBABILITY CURVES AT 200 REM WHOLE BODY
- 3. MULTIPLY CONDITIONAL PROBABILITIES BY SEQUENCE PROBABILITY
. 4. SUM SEQUENCE PROBABILITIES FOR FOUR SOURCE TERMS TO GENERATE SINGLE CURVE OF ANNUAL PROBABILITY OF EXCEEDING 200 REM WHOLE BODY VS DISTANCE O
M e
e-
. . - - . -. - - . .. . ...~.,
CALVERT CLIFFS CRAC2 MODEL o SOURCE TERMS NUREG-0956(TMLB-(,TMLB-f,S2D-7)
SWEC CALCULATED V o 1983 CCNPP MEASURED METEOROLOGICAL DATA o NO EVACUATION ,
o NORMAL ACTIVITY o NORMAL SHELTERING o DID NOT USE POPULATION DISTRIBUTION, HEALTH EFFECTS, OR PROPERTY DAMAGE CALCULATION CAPABILITY o CALCULATE CONDITIONAL PROBABILITY CURVES (ASSUMES ACCIDENT OCCURS) FOR WHOLE BODY DOSE 5
l l
,-w --
' i r i :! ' i -, , I :L i6 } ':ifk E
S G T A C M E A F D F
E Y T
H R T E L P A O E R H P
- N' E
N I
m L
T N U O Y O I CE NS T \
L A EN E L / GO U RP D P ES e O O ME ER M P 2
C .
A R
N/
C
. +
Y R
T E
M I
S O
- D A'
\ /
N O
A I N T T O A C E I D I L T
. R P A N
R E
H V E H
P V E
D V DM I
T S D NA A O U UT E H O ON W T L RO A C GC N/
I s
E V
I NT OC I A TO PI E I DS RAA CRE S L EFE DOR
. . _ . . . . ._ ._. _ .a HEALTH EFFECTS ,
o EARLY FATALITY B0NE MARROW LUNG GI o LATENT FATALITY (CANCER) -
~ ~
o EARLY INJURY WHOLE BODY LUNGS LOWER'GI WALL i PROPERTY DAMAGE o EVACUATION COST i
i o MILK AND CR0P DISPOSAL
~
o DECONTAMINATION o LAND USE PR0HIBITION (INTERDICTION)
. ~ . - . - .. .. - . . --. .. . _
POPULATION o DISTRIBUTION OF POPULATION IN 34 RADIAL ZONES (0.-5 TO 500 MILES) AND 16 EQUAL ANGULAR SLICES OF EACH ZONE AROUND PLANT 4
EMERGENCY RESPONSE
~
_ o WHEN EVACUATE o SPEED OF EVACUATION o EVACUATION DISTANCE o SHELTERING ACTIVITY LEVEL CLOUD' SHIELDING GROUND SHIELDING BASEMENT / HOUSES
\
. . - . . . - --.l_-.--.__-..-.--.-..-.- . . - . . - . . - . - . . _ - . - - .
y . _ _ . . . . . . . . . . .. . . -.. . ._ . - _ . __ . _ . . . . . . _ .
.c DOSIMETRY o INPUT FROM ATMOSPHERIC DISPERSION CLOUD DEPLETION GROUND CONTAMINATION o CALCULATES DOSES WHOLE BODY INTERNAL ACUTE LATENT 4
i e
L GROUND CONTAMINATION o GROUNDSHINE o CLOUDSHINE o INHALATION o INGESTION' I
oo O
e
. ....._. .- . _ ._. __ . ... .._ _ . ~
i 1
d CLOUD DEPLETION o IS0 TOPE DECAY
, o RAIN DEPLETION (WET DEPOSITION) i o DRY DEPOSITION l
I
.-- ,...a--.i --. -.-,,f.., , ..,,,,-,,,,,,.,,_y ,---w.-----. . 7~,- w.,y. ,---,w.y- e,-,wwy. , .,,m. p.g w -,-w --. - - * - - , .----w
t
. _ . _ _ . . _ _ ._ _.. __ ._ . 1 -
. ...g
~
ATM0 SPHERIC DISPERSION
~
o RADI0IS0 TOPE DILUTION IN ATMOSPHERE .
o USES WEATHER DATA .
o ACCOUNTS FOR BUILDING WAKE EFFECTS AND PLUME RISE
- o PROVIDES SOURCE FOR CLOUD DEPLETION AND GROUND CONTAMINATION 4
i l
l M
e 1
___,,_.,_..._i_.
~
.__l~'"~__'._______.,._.._._.___._ _ . . _ _ . . _ _ _ . . _ _ . _ . , _ _ _ _ . ._
- . - : . . . . . . ~ . . . . . . . - .. . --
WEATHER DATA o METEOROLOGICAL DATA FROM SITE DATA TAKEN EVERY HOUR OF A YEAR I
RAIN WIND DIRECTION -
) .
WIND SPEED
. STABILITY CLASS (A, B, C, D, E, F OR 6)
BASED ON TEMPERATURE DIFFERENCE AND WIND SPEED, A MEASURE OF WEATHER DISPERSION EFFECT DN RELEASED RADI0IS0 TOPES 9
l o
- ^^ - ^ - - - -
DESCRIPTION OF RADI0 ACTIVE RELEASE o CORE INVENTORY BUILT-IN TO CRAC2 PWR, 33,000 MWD /MTU, 3412 MWT .
RUN ORIGEN CODE 54 IS0 TOPE CAPABILITY o FRACTION OF IS0 TOPES RELEASED FROM " CONTAINMENT" '
o TIME AFTER ACCIDENT OF RELEASE FROM " CONTAINMENT" o TIME PERIOD OF RELEASE FROM " CONTAINMENT" T
( -
~
8 4
i i ii I; l , 6 * ( i,\i ,\ i L
. E S G T A C M E A F D F
E Y T
H R T E L P A O E R H P
' Ns e
E N
I L
T N U O Y D I CE NS T \
L A EN E L /' GO U RP D P ES O O ME t
M P ER g
2 C
A R N/ .
C Y
R T
E M
I S
O D
~
s'
\ /
N O
A I N T T O A C E I D I N L T RO P A N
R E
H V EI HS PR V E
D V DM I
- i N#
1 I E
V I
NT OC I A .
TD PI E I DS RAA CKE S L EFE DOR
. . . _ . . _ . . . . . ~~ . _. ~ .._.~ _. ~_~ ~<
CRAC2 j (LALCULATIONOFREACTORACCIDENT10NSEQUENCES)
(DEVELOPED BY SANDIA FOR NRC) o MODELS:
_ RELEASE OF RADI0IS0 TOPES FROM " CONTAINMENT" DISPERSION DOWNWIND OF PLANT DEPOSITION ON GROUND
. EFFECTS OF AIRBORNE AND DEPOSITED RADI0 ISOTOPES ON MAN AND ENVIRONMENT o CONSEQUENCES INCLUDE:
EARLY AND CONTINUING BIOLOGICAL EFFECTS LATE (CANCER) BIOLOGICAL EFFECTS GENETIC EFFECTS ECONOMIC IMPACTS a