ML20203N171
| ML20203N171 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs, 05000000 |
| Issue date: | 06/19/1986 |
| From: | Cook T, Metcalf J, Mirsky S BALTIMORE GAS & ELECTRIC CO., STONE & WEBSTER ENGINEERING CORP. |
| To: | |
| Shared Package | |
| ML20198G688 | List:
|
| References | |
| TP-86-61, NUDOCS 8609230181 | |
| Download: ML20203N171 (29) | |
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WEPZ REDUCTION AT THE i' :;CALVERT CLIFFS NUCLEAR i.
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18931-573480111-B3 EPZ REDUCTION AT THE CALVERT CLIFFS NUCLEAR POWER PLANT J. Metcalf Stone & Webster Engineering Corporation and S. Mirsky T. Cook Baltimore Gas & Electric Company INTRODUCTION AND
SUMMARY
A plant-specific analysis of offsite radiological consequences was performed for severe core damage accidents postulated at the Calvert 1
Cliffs Nuclear Power Plant (CCNPP) to support a request for exemption from the 10CFR50.472 requirement for a plume exposure pathway emergency planning zone (EPZ) radius of 10 miles (16.1 km).
The results of the analysis usin's new severe accident source term knowledge showed that an EPZ radius of 2 miles (3.2 km) affords the same relative dose savings as a 10-mile (16.1-km) EPZ radius using the earlier source term calculations 3
of WASH-1400.
It also showed that the risk of significant injury to an individual is less for the reduced EPZ for CCNPP than had been calculated previously for the 10-mile (16.1-km) EPZ.
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18931-573480111-B3 Source terms and release fr,equencies for the CCNPP offsite consequence analysis were taken from available NRC-sponsored analyses where possible.
For the case of the interfacing system loss of coolant accident (LOCA),
no NRC-sponsored analysis could be applied to CCNPP, and a plant-specific analysis was performed.
IMPORTANCE OF INTERFACING SYSTEM LOCAs TO EPZ REDUCTION
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An interfacing system LOCA may occur with the failure of one or more elements of a light water reactor (LWR) primary coolant pressure bounda-ry, resulting in the discharge of reactor coolant into an < interfacing fluid system.
A rupture outside the containment may then occur in thid interfacing system, establishing a direct path from the reactor to the contiguous structure which surrounds the break location.
If the rupture cannot be isolated, an unrecoverable coolant loss may occur.
An impor-tant example of such an event, and one which has been reanalyzed several times -6, is the Surry V sequence identified in WASH-1400.
4 7 and the The American Nuclear Society Special Committee on Source Terms American Physical Society Study Group on Radionuclide Release from Severe Accidents at Nuclear Power Plantss stressed the importance of. such i
j sequences despite their low probability.
Indeed, for a large dry con-i tainment pressurized water reactor (PWR),
interfacing system LOCAs bypassing containment and severe accident sequences involving failure to isolate the containment have been identified as the most important sequences relative to source terms, i.e., release of radioactive material to the environment.
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18931-573480111-B3 I' -
A comparison of CCNPP and Surry (discussed later) indicated that the two plants would exhibit ~ similar source terms for sequences involving an initial release into the containment (including failure to isolate the containment), but not for interfacing system LOCAs.
A CCNPP-specific analysis for interfacing system LOCAs was therefore undertaken.
CCNPP-SPECIFIC INTERFACING SYSTEM LOCA ANALYSIS At Surry, the most likely interfacing system LOCA involved a low-pressure emergency core cooling system (ECCS) failure in the safeguards building, with subsequent submergence of the break by drainage from the refueling water storage tank (RWST). The probability of such an event occurring at-CCNPP is virtually nil because of the piping and contiguous structure configurations, but rupture of the shutdown cooling system (SCS) letdown line butside containment, although extremely unlikely, is possible.
The CCNPP SCS letdown line rupture may or may not be submerged by break effluent.
Moreover, the rupture would occur in the auxiliary building, I
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'which is considerably larger than the safeguards building at Surry; As the SCS letdown line at CCNPP is Schedule IOS 14-inch (30.5-cm) diameter piping, it appears unlikely that it would remain intact after exposure to full reactor coolsot system (RCS) pressure. For the unsubmerged case it is assumed that the rupture occurs at the first welded attachment after the failed containment isolation valve.
For the submerged case, the rupture is assumed to occur in one of the flood protected ECCS pump rooms l
located in the lowest level of the auxiliary building.
A failed pump
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seal would result in the same break location as the submerged pipe rupture but is considered less likely.
I Because of the large postulated rupture size, building pressurization is rapid, and multiple vent paths to the environment,at different elevations in the auxiliary building may result. Multiple-vent paths permit - free
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convective exchange between the building and the environment and dramati-cally affect the analysis.
The vent paths assumed in this analysis include the truck acce'ss door located at grade and several ventilation openings located at two different elevations on the roof.
To carry out the CCNPP-specific SCS letdown line break analysis, Stone &
Webster (S&W) assembled a customized code package using IDCOR, NRC, and S&W proprietary codes.
Each code used 'was selected for its suitability for the particular analytical requirements.
The thermal-hydraulic analysis of the auxiliary building was carried out using the S&W proprietary THREED' computer code and an auxiliary building model with 11 nodes (Figure 1).
THREED is basically a. modification of RELAP 4 Mod 5 with all subroutines pertaining to the core and the primary i
f system removed.
CONTEMPT-LT heat sink models have been added, as well as fan and cooler models to permit treatment of ventilation systems.
The code has been used for equipment qualification calculations on several 10 dockets and has been benchmarked against HDR Containment Experiment 42 The particular version of THREED employed in this study includes hydrogen as an additional noncondensible to air.
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18931-573480111-B3 Mass and energy release rates to the auxiliary building were taken from a MAAP 2.0B12 analysis of the sequence.
This analysis also provided the event timing. Mass and energy release rates and event timing are largely independent of the flooding status of the ruptured letdown line, and therefore the MAAP 2.0B analysis applies to both the submerged and the unsubmerged pipe rupture cases.
In the case of a pump seal failure, the sequence would progress at a much slower pace.
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Fission product transport and release in the auxiliary building were calculated using the NAUA412 computer code modified by S&W to include diffusiophoretic deposition as an additional leak term in the airborne depletion expression.
The fission product release from the RCS is that from the NAAP 2.0B analysis except for the three " nonvolatile" fission product groups as defined in BMI-2104.
NRC-sponsored calculations (BMI-2104) typically calculate a greater release for these groups (Ba-Sr, Ru-Mo, and La).than does MAAP (which does not treat La,at all) and it was therefore decided to base the " nonvolatile" releases from the RCS on the V sequence analysis presented in BMI-2104.
Additional conservatism in the fission product release calculation is found in the assumption that all of the tellurium is released before vessel failure.
A unique feature of the CCNPP analysis is the inclusion of fire protec-l tion sprinkler effects.
CCNPP employs an extensive, seismically de-i i
signed, wet pipe sprinkler system in the auxiliary building, with a
, dedicated water supply and two fire protection pumps located in an autonomous structure.
Spray operation affects both thermal-hydraulics l
and fission product transport in the auxiliary building. Aerosol removal l
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by impaction and by condensation on both passive heat sinks and water droplets wais included.in the analysis by specific modifications to NAUA4 to account for impaction.
Impaction exhibits a very sharp " cut-off" behavior in terms of particle size. For a given size droplet, particles below a certain size.will-move along the streamlines around the falling droplet and will be unaffected; above that size, removal becomes very efficient because of the very large number of droplets.
Removal is l
therefore - dependent on the particle size distribution of the RCS release and the rate at which small particles agglomerate to sizes above the threshold for removal by impaction.
Since MAAP 2.0B does not provide a particle size distribution for. the RCS release, this also was taken from the BMI-2104 V sequence analysis for Surry.
2 For the unsubmerged break case, fire protection sprinklers are calculated to be actuated in the vicinity of the rupture (Figure 2, Node 2) shortly after the rupture occurs. The combination of sprinkler operation and the free convective exchange of the building and ambient atmospheres permits combustible mixtures of air and hydrogen to form in the break node (Node 2) and two adjacent nodes.
Several hydrogen burns were calculated to occur.
The volumetric flows out of the building during and immedi-ately after these burns are included in the NAUA4 analysis, i
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The prevailing free convective flows through the auxiliary building are depicted in Figure 3.
Note that several of the thermal-hydraulic model nodes are not relevant to the fission product transport and release (NAUA4) analysis, and the fission product transport model is therefore
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18931-573480111-B3 smaller.
The nodal assignaients for the fission product transport model are as follows:
Fission Product Transport Thermal-Hydraulic Model Node Number Model Node Number 1
2 2
3 3
7 4
9 5
11 Additional fire protection sprinklers are actuated in the fission product transport model Nodes 2 and 3 (thermal-hydraulic model Nodes 3 and 7) at i
the time of the first hydrogen burn (t = 13500 seconds).
Table 1 shows the ' attenuation of CsI in the auxiliary building for the unsubmerged case. Note that approximately 78 percent of the CsI released from the RCS is retained in the first node where it first encounters the fire protection sprinklers.
The effective decontamination factor (DF) for the remainder of the building is approximately 4.5.
For the sub-merged case, the DF is assumed to be sufficiently large to result in a release of essentially only the noble gases.
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18931-573480111-B3 SOURCE TERMS FOR SEQUENCES OTHER THAN INTERFACING' SYSTEM LOCAS i
For large dry containment PWRs, source terms sufficiently large to produce mean doses in excess of 200 rem (indicating significant risk of early injury) beyond one mile from the point of release are likely to occur only if the containment is bypassed, is initially impaired, or fails early (i.e., at or near the time of vessel failure). Containment bypass sequences for CCNPP were discussed in the previous section. Source terms' for CCNPP sequences involving an impaired containment or early
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containment failure are discussed below.
For CCNPP, severe accident source terms for impaired containment sequenc-es and sequences involving early containment failure were based on the recent NRC-sponsored risk analysis of the Surry plant that was presented in NUREG-095618, Appendix D. The source terms in this recent analysis were taken for the most part from BMI-2104, with some additional source terms taken from supplemental analyses.
The important source terms are as follows:
1.
TMLB'-p - representing sequences where the sprays are not operating and the containment is impaired 2.
TMLB'-6, - representing sequences where the sprays are not operating and the containment fails early 3.
S D-y - representing sequences where the sprays operate ini-tially but fail at the time of early containment failure sums a wesem 8
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To apply these source terms to CCNPP it was first necessary to demon-strate that CCNPP-is similar to Surry in regard to both the calculation of source terms and the assessment of conditional probabilities for various containment failure modes.
It was then necessary to determine
- the frequency of the CCNPP severe accident sequences that could lead.to
.the potentially large source terms described above.
SURRY AND CCNPP SIMILARITIES 4
The important features establishing similarity between Surry and CCNPP may be divided into three groups:
General Features, Features Important t
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to Severe Accident Progression, and Features Important to Containment Performance, as follows:
4 1.
General Features 2440 Mw(t) for Surry vs. 2700 Mw(t) for CCNPP Power l
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10s fe - (50970 m ) for Containment Free Volume - 1.8 X 3
s (56630 m ) for CCNPP Surry vs. 2.0 X 10s ft 2.
Features Important to Severe Accident Progression l
In-core Instrumentation Penetration Location - Lower head for Surry vs. upper head for CCNPP l
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,e Cavity Condition-(Given Spray Failure) - Dry-for Surry and
3.
Features Important to Containment Performance Containment Construction - Reinforced for Surry vs. post-tensicned for CCNPP Containment Operating Pressure 11 psia _ (0.069-0.076 MPa) for Surry vs. atmospheric for CCNPP Containment Design / Failure Pressure - 60/135 psia (0.41/0.93 MPa) for Surry vs. 65/140 psia (0.45/0.96 MPa) for CCNPP Post-Accident Status of Containment Unit Coolers
- Unavailable for Surry vs. available for CCNPP i
In comparing these features, the following may be noted:
1.
The free volume of the containment per unit power of the core 3
8 for both plants is approximately 740 ft /Mw(t) (21 m /Hw(t)).
2.
CCNPP has no penetrations in the lower head, making it differ-ent from Surry. This might be expected to lead to a later
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b 18931-573480111-B3 calculated head failure for CCNPP.
However, in the BMI-2104 1
analyses, the failure of the lower head was calculated as if the penetrations did not exist, and therefore these analyses appear to be more applicable to CCNPP than to Surry.
3.
Cavity conditions and concrete type are similar (both basaltic and siliceous concrete are low in CACO, leading to a low 3
noncondensible gas production during concrete degradation).
Therefore, core / concrete interaction after vessel failure should be approximately the same.
1 4
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The containment construction is different. Recent experimental
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investigations sponsored by EPRI suggest that " leak-before-break" may be somewhat more likely for post-tensioned designs than for reinforced designs, but this would have little effect on.early containment failure since pressurization rates associ
ated with early containment failure are quite high.
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The subatmospheric operating pressure for Surry makes operation with an impaired containment less likely for Surry than for CCNPP.
6.
The containment capability for CCNPP is slightly greater than l
l that for Surry, but given Surry's lower containment operating pressure, the capabilities are comparable.
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.No credit has baen taken for the operation of the CCNPP safety-L related containment unit coolers for sequences in which the 9
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core has been degraded and the sprays have failed (e.g., S FH).
4 For these sequences, the unit coolers would be exposed to conditions for which they have not been qualified, and taking credit for 'their operation might be subject to question.
However, ac T!!I-2, the containment unit coolers were exposed to such conditions and! continued to remove heat from the contain-
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ment atmosphere.
Therefore, it is likely that the CCNPP i
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containment unit coolers would reduce the conditional probabil-s
~ failure for sequences such as S FH when ity of containment N'~
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comparest to comparable sequences for Surry.
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Taking the above 'into consideration, the following conclusions may be drawn:
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Source terms calculated for a given "in-containment" sequence it. Surry may be applied te(similar sequences for CCNPP.
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The conditional probability of an impaired containment at Surry i
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(0.002 from NUREG-0950, Appendix D) should be replaced by'O.005 s,,
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for CCNIP, based on the EPIi/NSAC study for Oconee".
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For cectsin sequences involving loss'of containment sprays, the I*
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conditional probability of containment failure for Surry may be i
i somewhit high when applied to CCNPP.
The conditional probabil-I ii ity of " leak-before-break' ' for Surry may be somewhat low when s
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applied to CCNPP.
However, both of these conclusions apply to late containment failure modes, and are not of concern for the i
source terms under consideration. - Therefore, the conditional probabilities of early overpressure failure for sequences with sprays failed and of early overpressure failure with subsequent spray failure for sequences with sprays initially operating will be taken as is from NUREG-0956, Appendix D (0.005 and 0.001, respectively) and applied to CCNPP.
FREQUENCIES ASSOCIATED WITH IMPORTANT SOURCE TERMS The Interim Reliability Evaluation Program (IREP) study for CCNPP1A estimated the total core melt frequency (CNF) to be 1.3 x
10-4/ reactor year. Approximately 90 percent of the severe accident (core i
melt), sequences for CCNPP fall into the following categories:
1.
Anticipated Transients without SCRAM - 33% (ATWS (PSF), T KU, 4
T KQ, T KU, and T KQ)
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2.
Transients with Complete Loss of Feedwater - 32% (TDC, T L.
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Small LOCAs with Loss of ECCS - 12% (S H and S D")
2 3
4.
Small LOCA with Loss of Recirculation Sprays and Core Cooling - 9% (S FH) 2
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18931-573480111-B3 L
5.
Station Blackout - 3%
For categories 1 through 3, the containment sprays will be initially operating, and the concern is for an early containment failure leading to loss of sprays. The conditional probability of this occurring is 0.001 i
(refer to previous section), and the associated Surry source term is that for S D-y.
Therefore, the frequency of this source ters as applied to CCNPP is:
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CCNPP CE x (Percentage of CW for Categories 1 through 3/100) x 0.001 = 1.0 x 10-7/R-yr For categories 4 and 5, the containment sprays will not be operating and the concern is for either an impaired containment or an early containment
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failure. The conditional probabilities for each of these containment failure modes is 0.005 (refer to previous section), and the associated Surry source' terms are those for TMLB'-p and TEB'-8,,
respectively.
Therefore, the frequency for each of these source terms as applied to CCNPP is:
CCNPP CE x (Percentage of CE for Categories 4 and 5/100) x j
0.005 = 7.5 x 10-s/R-yr The remaining frequency requiring definition is that for the SCS inter-facing system LOCA bypassing containment (V(SCS)). This sequence was not addressed in the CCNPP IREP study, but a similar configuration was evaluated in the Zion PRA a (sequence 16-V) with an assessed frequency of l
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18931-573480111-B3
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1.0 x 10-7 per reactor-year. This is the frequency assigned to the CCNPP i
V(SCS).
No attempt was made to define a conditional probability of pipe rupture or seal failure, or of a submerged or unsubmerged release. Two limiting offsite consequence calculations were done; one with the fre-quency of 1.0 x 10-7 per reactor-year assigned to the unsubmerged pipe rupture case, and one with the frequency of 1.0 x 10-7 per reactor-year assigned to the submerged seal failure case.
This is discussed further in the last section.
1 Table 2 provides a summary of the input to the offsite-consequence analysis.
OFFSITE CONSEQUENCE ANALYSIS The CCNPP consequence analysis was performed for both design basis accidents (DBA) and the severe accidents discussed earlier in this paper.
The consequences for DBAs were evaluated for 67 nuclear power plant sites in NUREG-039617 The doses were related to the 1-and 5-rem protective action guidelines (PAGs) by plotting two-hour whole body dose as a l
function of distance from plant sites. These DBA/LOCA doses were calcu-lated with a five percentile meteorology and straight line plume trajec-tory.
Results are presented in, Figure 1-8 of NUREG-0396 for 50 percent of the 67 sites (median) and the highest 10 percent of the 67 sites. The specific CCNPP DBA/LOCA two-hour whole body dose taken from the FSAR s i, i
added to these two curves from NUREG-0396 and presented in Figure 4.
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38931-573480111-B3 The CCNPP severe accident offsite consequence analysis was performed with the Calculation of Reactor Accident Consequences (CRAC2) " computer code.
CRAC2 was developed by Sandia National Laboratory for the NRC.
CRAC2 models the release of radioisotopes from a containment, their dispersion
.s downwind of the plant, and the subsequent deposition of these radioiso-topes onto the ground.
CRAC2 calculates the effects of airborne and
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deposited ' radioisotopes on people and on the environment.
These conse-quences can include:
early, continuing, and late (carcinogenic) biologi-cal effects, genetic effects, and economic impact.
CRAC2 explicitly models the mechanisms contributing to ground contamina-tion doses (grounshine, cloudshine, inhalation, and ingestion).
Clou'd depletion is another important process which is accounted for by CRAC2 in evaluating isotope decay, rain depletion, and dry deposition.
CRAC2 simulates atmosphere dispersion by using site-specific meteorological data, including radioisotope dilution, and providing for building wake effects as well as plume rise.
Site-specific meteorological data consist of information on rain, wind direction, wind speed, and stability class taken every hour for at least one year. Stability class, categorized by the letters A through G, is based on temperature differences at different elevations and wind speeds.
Stability class is a measure of the disper-sive effect of weather on released radioisotopes.
The radioactive release in CRAC2 consists of a fraction of the core inventory of up to 54 isotopes available in the code that are released along with the time period of the release and the time after a postulated accident that release occurs.
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18931-573480111-B3 U
The CCNPP severe accident consequence analysis utilized the radioisotope release (source terms) for the four sequences previously discussed and presented in Table 2.
The CRAC2 model used Calvert Cliffs meteorological j
i data from 1983.
No evacuation, normal activity, and normal sheltering were assumed.
The economic impact, health effects, and population distribution options were not utilized. CRAC2 calculated the conditional probability curve (i.e., assuming that the postulated accident occurred)
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for receiving a 200-rem whole body dose for each of the four severe accident sequences.
Each individual sequence curve was then multiplied by its probability and the resultant values were summed for the four sequences to generate a single curve. This curve, presented in Figure 5,'
is compared to the analogous Figure 1-11 curve from NUREG-0396. Figure 5 relates the probability of exceeding a 200-rem whole body dose per reactor year for different distances from the release point (i.e.,
CCNPP). Two CCNPP cases are plotted:
one using the higher (unsubmerged)
V sequence source term and one using the lower (submerged) V sequence source term.
The, closeness of the two curves illustrates the large contribution from the in-containment source terms.
EPZ REDUCTION RATIONALE A major factor in the establishment of the existing 10-mile (16.1-km) EPZ radius presented in NUREG-0396 is that the DBA whole body dose should not exceed the PAGs at the limit of the EPZ.
Figure 4 shows that the DBA dose for CCNPP does not exceed even the lower (1-Rem) PAG at a distance of,2 miles from the point of release. This finding supports a reduction in the EPZ radius from 10 to 2 miles (16.1 to 3.2 km) for CCNPP.
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A second major factor in establishing the 10-mile EPZ is the distance at which the 200-rem curve begins to decrease abruptly in Figure 1-11 of NUREG-0396.
Both CCNPP curves decrease abruptly at a distance of about 2 miles (3.2 km) vs. 10 miles (16.1 km) for the NUREG-0396 curve.
This indicates a dose savings for a 2-mile (3.2-km) EPZ at CCNPP, based on recent source terms, comparable to the 10-mile (16.1-km) EPZ dose savings reported in NUREG-0396. For either case, the pro 6 ability of experiencing a 200-rem dose at any distance is substantially lower for CCNPP with current source terms than was the case in ~ NUREG-0396 with WASH-1400 source terms.
This finding also supports a reduction in the EPZ radius from 10 to 2 miles (16.1 to 3.2 km) for CCNPP.
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18931-573480111-B3 REFERENCES 1.
Letter from Baltimore Gas & Electric Company to U.S. Nuclear Regula-tory Conusission, November 18, 1985.
2.
Nuclear Regulatory Commission, " Code of Federal Regulations," 10CFR 50.47, Revised January 1, 1985.
~
~
3.
U.S.
Nuclear Regulatory Commission,
" Reactor Safety Study: An Assessment of Accident Risks in U.S.
Commercial Nuclear Power i
Plants," WASH-1400 (NUREG-75/014), October 1975.
4.
Drozd, A., Elia, F. A., Jr., and Metcalf, J.E., "The V Sequence: An Engineering Viewpoint," presented at the ANS Topical Meeting on Fission Product Behavior and Source Term Research, Snowbird, Utah, July 15-19, 1984.
5.
Electric Power Research Institute, "Surry Source Term and Conse-quence Analysis," Final Report, EPRI NP-4096, June 1985.
6.
Gieske, J.A.,
et al.,
"Radionuclide Release Under Specific LWR Accident Conditions," Vol. V, Draft Report, BMI-2104, July 1984.
7.
" Report of the Special Committee on Source Terms," American Nuclear Society, September 1984.
stoms a wee m a 19
.3
[
18931-573480111-B3 8.
" Report to American Physical Society of the Study Group on Radionuclide Release from Severe Accidents at Nuclear Power Plants,"
Draft Report, February 1985.
9.
- Boyle, J.C.,
and Wong, W.,
"THREED - A Subcompartment Transient Response Code," Stone and Webster Engineering Corporation, NU-092, May 1984.
10.
- Bamdad, F.,
et al., "A Methodology for Analysis of Environmental Effects of a Pipe Break Outside Containment on Systems in Harsh /
Nonharsh Environments," presented at the Third International Meeting
~
on Reactor Thermal Ilydraulics, Newport, Rhode Island, October 15-18, 1985.
11.
" Modular Accident Analysis Program - MAAP-2.0B", Industry Degraded Core Rulemaking (IDCOR) Program, (In Publication.by Atomic Industri-al Forum.)
12.
Bunz, H., Kayro, M., and Schoch, W., "NAUA-Mod 4: A Code for Calcu-lating Aerosol Behavior in I,WR Core Melt Accidents," KFK-3554, August 1983.
13.
U.S. Nuclear Regulatory Commission, " Reassessment of the Technical Bases for Estimating Source Terms," Draft Report, NUREG-0956, July 1985.
~
smus a weemn 20 l
6 18931-573480111-B3 14.
Nuclear Safety Analysis Center, "Oconee PRA, A Probabilistic Risk Assessment of Oconee Unit 3," NSAC/60,.Palo Alto, California, June 1984 15.
Interim Reliability Evaluation Program: Analysis of the Calvert Cliffs Unit 1 Nuclear Power Plant, NUREG/CR-3511, SAND 83-2086, March 1984.
16.
Zion Probabilistic Safety Study, Revision 0, Chicago, kilinois, Commonwealth Edison Company, September 1981.
17.
U.S. Nuclear Regulatory Commission, " Planning Basis for the Develop-ment of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants," NUREG-0396, December 1978.
18.
Calvert Cliffs Nuclear Power Plant Updated Final Safety Analysis Report, Revision 4, July 15, 1985.
19.
Calculation of Reactor Accident Consequences, Version 2, CRAC2 Computer Code, NUREG/CR-2326, SANDSI-1994, February 1983.
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A 21 1-18931-573480111-B3
.s TABLE 1 CsI ATTENUATION IN THE CCNPP AUXILIARY BUILDING SCS LETDOWN LINE RUPTURE - UNSUBMERGED CASE Total inventory (MAAP):
24.1 kg Released from:
Mass:
Leaving / Entering:
me
- 1st Node 4.5 kg (23%)
- 2nd Node 2.75 kg (61%)
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- 3rd Nods 1.8 kg (65%)
- 4th Node 1.15 kg (64%)
- 5th Node
.95 kg (83%)
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TABLE 2 SEVERE ACCIDENT SOURCE TERMS AND PROBABILITY OF OCCURRENCE FOR CCNPP Range of V Sequences i
TMLB'6e(*}
THLB'-S(*)
S,D-y(*)
Low (b)
High(C)
Accident Sequence Probability, Per 7.5 x 10-s 7.5 x 10-s 1 x 10-7 1 x 10-7(d) 1 x 10-7(d)
R actor Year i
Time of Rel, hr.
2.5 2
2.5
>10(*)
4 II) ff)
Duration of Rel, hr.
10 lo 3
1 3
l Fission Product Group Fraction of Core Inventory Released to Environment 1
l In-Kr 0.85 1.00 0.50 1.00 1.00 I-Br 0.07 0.022 0.005
<0.001 0.039 l
i Co-Rb 0.058 0.013 0.0001
<0.001 0.038 Te-Sb 0.055 0.11 0.01
<0.001 0.042 Ba-Sr 0.01 0.058 0.03
<0.0001 0.0009 i
Ru 0.0013 0.0053 0.001
<0.0001
<0.0001 i
La 0.00017 0.0002 0.0009
<0.0001
<0.0001 j
4 (a) ' Release fractions from NUREG-0956.
(b) Low range of potential V sequences includes small leakage paths (e.g., 0.1 ftz) at a pump seal submerged in a flood-protected pump compartment.
j (c) HighrangeofpotentialVsequencesinbasedona14-in.-diameterSCSpipeopening-unsubmerged.
(d) 1 x 10-is the total probability of the range of V sequences' for SCS.
The probability of experiencing the high-range source term would be substantially lower than 1 x 10-7 I
(e) Time of release is a function of postulated size of opening in SCS.
(f) Analyzed in CRAC2 with a 2-hr. release duration.
1 19050-522023611-B3 1 of 1 A
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g FIGURE 1 THERMAL-HYDRAULIC MODEL OF THE AUXILIARY BUILDING O 11 NODES O 17 JUNCTIONS 131' A H = 28' 103' A H = 22' 17 8
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CALVERT CLIFFS AND NUREG.-0396 CALCULATION 2 HR. WHOLE BODY DOSE FOR LICENSING CALCULATION OF DBA/LOCA AT 2 HOURS-ASSUMING 5 PERCENTILE METEOROLOGY AND STRAIGHT LINE TRAJECTORY (REPRODUCED FROM FIG.1-8, NUREG.-0396) 4 g
k i
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3
- g SITE CALCULATIONS W
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=
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CALVERT CLIFFS l
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8 10 12 14 DISTANCE (MILES) k
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FIGURE 5
-s 10 CALVERT CLIFFS AND NUREG.-0396 d
- PROBASILITY OF EXCEEDING 200 REM -
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WHOLE BODY DOSE PER REACTOR YEAR m===
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