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. PROPOSED RESEARCH PROGRAM
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t FISSION PRODUCT TRANSPORT PROGRAM --
LIMERICK PLANT ANALYSES TASK to U.S. NUCLEAR REGULATORY COMISSION January 25, 1984 i
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e BATTELLE Columbus Laboratories 505 King Avenue Columbus, Ohio 43201 l'
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PROPOSED RESEARCH PROGRAM on k
FISSION PRODUCT TRANSPORT PROGRAM --
1 LIMERICK PLANT ANALYSES TASK Ti
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4 U.S. NUCLEAR REGULATORY COP 94ISSION I
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BATTELLE l
Columbus Laboratories January 25, 1984 INTRODUCTION Battelle proposes to undertake an adoitional task under the existing program Fission Product Transport (FIN No. B6747). The purpose of the task is to apply the methods of source term analysis developed for the Source Term Reassessment Study (STRS) to the Limerick Plant.
These analyses are particularly timely for two reasons:
(1) Source terms have been investigated for severe accident sequences in BWRs of the Mark I and Mark III containment design types in the STRS but not for a Mark II design.
(2) BNL has been performing analyses in support of l
NRC-NRR as part of the Draft Environmental Statement for the Limerick Generating Station using WASH-1400 based methods. The planned studies will allow BNL to benchmark or update these analyses.
BACKGROUND In the Source Term Reassessment Study a number of accident g
L sequences were selected for analysis within five plant designs. For each accident sequence, consistent analyses were performed for the 4
1 2
processes controlling the release and transport of radionuclides within the plant and their release to the environment.
The first major step in the process was the selection of types aj of nuclear power plant designs to be considered and a specific plant to i
represent each type. The types to be considered were:
large, dry PWRs; j
Mark I BWRs; Mark III BWRs; and ice condenser type PWR designs. The specific plants chosen to represent each type were the Surry and Zion, j
Peach Bottom, Grand Gulf, and Sequoyah plants. These selections were I
made on a combined basis of typicality of design and availability of design details needed for analyses.
Accident sequences were chosen for each plant design based on risk significance and on a desire to have a range of physical conditions represented by the analyses. The plants selected and the accident sequences considered are listed below:
PWR Large Dry PWR Large Dry Containment Containment BWR Mark I (Surry)
(Zion)
(Peach Bottom)
2 V
TW TMLB' PWR Ice Condenser BWR Mark III Containment (Grand Gulf)
(Sequoyah)
TC S HF 2
TPI TMLB' TQUV TML Following the selection of plants and sequences the required plant design data were collected and thermal hydraulic analyses performed for the accident sequences. Overall thermal hydraulic conditions on a time-dependent basis were estimated with the MARCH 2 code, N and detailed thermal hydraulic conditions for the primary system estimated with the MERGE code which was developed specifically for use in this program.
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.The time-dependent core temperatures were used as input to another code, CORSOR, developed for the STRS which predicts time and temperature dependent mass releases of radionuclides from the fuel within
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the pressure vessel. Releases during core-concrete interactions of
- j-radionuclides remaining with'the melt were provided by Sandia National i
j Laboratories-using their computer code VANESA.
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Using the MARCH /ERGE predicted thermal hydraulic conditions 9.
and the CORSOR predicted radionuclide release rates as input, a newly I.!
developed version of the TRAP-ELT code was used to predict vapor and lf particulate transport in the primary coolant circuit.
Transport and deposition of radionuclides in the containment and ex-containment structures were calculated using a modified version of the NAUA-4(2) code. Decontamination in suppression pools and ice 4.
beds was predicted using the SPARC and ICEDF codes, respectively.
The basic stepwise procedure described above-is illustrated in 4
Figure 1 which shows the relationships among the computational models.
The calculations were of a "best estimate" type using input derived, to the extent possible, from experimental measurements. Types of data i
employed in the analyses include vapor deposition velocities, aerosol f
deposition rates, aerosol agglomeration rates, fission product release rates from fuel, particle sizes formed from~ vaporizing / condensing fuel materials, engineering correlations for heat and mass transfer, and
,i physical properties of various fuel, fission product and structural materials.
I APPROACH 4
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The methods of analysis to be used in this task are the same asthoseusedpreviouslyforthePeachBottom(MarkI)andGrandGulf
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(Mark III) plants. The specific sequences to be analyzed were selected J.
cooperatively by the NRC, BNL, and BCL staff. The sequences were chosen S.
to be representative of a number of accident classes identified in the Limerick Generating Station PRA and analyzed previously by BNL. The
~7 - sequences selected for analysis are: TQUV, TC-y, TC-y', and S QVV.
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Plant. data are to be provided by BNL.
In addition to performing the
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SELECTION OF TYPE 5 0F PLANTS ir SELECTION OF SPECIFIC PLANTS i
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,h' 9r II' SELECT!0m 0F ACCI ENT SEQUENCES J
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,r SPECIFICATION OF PLA?.i -
INVENTORY GE0 METRY AND ACCICENT SEQUENCE PHENOMENA ORIGEN 0VERALL THERMAL HYOPAlt!CS MARCH 9r f
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l PRIMARY SYSTEM RELEASE FROM FUEL THER.9AL HYDRAULICS COR$0g MERGE
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l CORE. CONCRETE I m MCTION PRIMARY SYSTEM TRANSPORT
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%f CORE CONCRETE RELEA$E CONTA! MENT TRANSPORT yggg$g NRUA.4 Mod 1fied l
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RELEASE TO ENVIRONMENT l
FIGURE 1.
INFORMATION FOR RELEASE. TRANSPORT, AND DEPOSITION CALCULATION t
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5 identified analyses, BCL will assist BNL in learning how to use the STRS computer codes.
j STATEMENT OF WORK 3
The first' step in this task is the collection of, data. Some of the input data for the MARCH analyses was prepared previously by BNL_
for their earlier analyses. However, more detailed data are required for the codes used in the Source Term Reassessment Study than for the WASH-1400 based methods.
The analyses for each sequence will be performed in the following order: thermal hydraulic (MARCH 2, MERGE), fission product source (CORSOR, VANESA), reactor coolant system transport (TRAP-MELT),
and containment transport (NAUA, SPARC). BNL staff will become involved with code operation to the extent practical in order to assist in the transfer of experience in their use.
Following completion of the analyses _, the results will be f
documented in a final letter report.
f TIME AND COST To acconinodate the schedule requirements of the NRR staff, it is important that these proposed calculations be completed as soon as 7
possible.
It is estimated that they can be completed by February 20 and this completion date sets the overall schedule for thermal hydraulic, fission product release, and transport and deposition calculations.
To complete the proposed calculations and transmit the results informally by letter will require an appropriation of $137,956, including a fixed fee of $9,110. Appropriate cost forms are enclosed.
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REFERENCES (1)
Wooton, R. O., et al, " MARCH 2 Code Description and Users' Manual",
Draft (December,1982).
'.'j (2)
Bunz, H., Koyro, M., and Schock, W., "A Code for Calculating ti Aerosol Behavior in LWR Core Melt Accidents, Code Description and j
Users' Manual".
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BACKGROUND TO BNL FISSION PRODUCT SOURCE TERM CALCULATIONS FOR LIMERICK
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Fission product release fractions for Class 1 j
l LGS - PRA NUREG-3028 CURRENT CALCULATION
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CyY I-T/DW I-T/WW I-T/W FAILURE MODE CY i
1 OXIDATION RELEASE.. Yes Yes
_No No No Xe - Kr
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DF for 12 100 100 100 100 100 l
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DF for Aerosols 100 100 l
100 100-100 l
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.2.42 1st Vap. Release 1
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CONSEQUENTLY, THE LONGER-TERM DAMAGE INDICES (LA-TENT FATALITIES, PERSON-REM, ETC.) ARE LOWER IN THE LGS-DES THAN WOULD HAVE BEEN CALCULATED USING THE LGS-PRA SOURCE TERMS BROOKHAVEN NATIONAL LABORATORY l} g)l ASSOCIATED UNIVERSITIES, INC.(Illl 8h'*-
n
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CONTAINMENT FAILURE OCCURS RAPlDLY (THREE FAILURE LOCATIONS CONSIDERED) e COOLANT INJECTION FAILS e
CORE MELTS INTO A FAILED CONTAINMENT i
BROOKHAVEN NATIONAL LABORATORY l} g)l j
A5500ATED UNIVERSITIES, INC.(Illl
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Table 4.7*
Fission product release fractions for Class IV (failure location W below wetwell waterline)
LGS - PRA NUREG-3028 CURRENT CALCULATION
~ Cp" --'
' ~ ~~ C y " '
FAILURE M3DE-4 IV-T/W OXIDATION RELEASE Yes Yes
~
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I I
i l
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Table 4.6*
Fission product release fractions fo'r Class IV (failure location WW) 0 l
LGS - PRA NUREG-3028 CURRENT CALCULATION C y '_,,, _ C y' IV-T/WW FAILURE MODE-4_
4 OXIDATION RELEASE Yes Yes Yes
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, - - ~.
Table 4.5*
Fission product release fractions for Class IV (failure location DW) l LGS - PRA NUREG-3028 l CURRENT CALCULATION Cy Cy FAILURE MODE 4,..
4
-IV-T/DW OXIDATION RELEASE Yes Yes Yes
, - - - - - - - - - Xe *-Kr-
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Table 4.11* A comparison of fission product release fractions for Class IV sequences initiated by LOCAs and Transients NUREG-3028
. CURRENT CALCULATION FAILURE MODE cay LOCA IV-A/DW IV-T/DW Xe
-Kr----
- 9989
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