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Nonproprietary Chapter 6 to RESAR-SP/90 Westinghouse Advanced Pwr,Pda Module 16, Offsite Consequence Analysis
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Issue date:	09/30/1985
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WESTINGHOUSE ADVANCED PRESSURIZED WATER REACTOR CONSEQUENCE ANALYSIS SECTION 6.0 O

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O SECTION 6.0 CONSEQUENCE ANALYSIS 6.1 OVERVIEW &

SUMMARY

A quantitative assessment of severe accident consequences has been performed employing radioactive source terms derived from postulated accident scenarios IO for a reactor plant of the Westinghouse APWR design. Modeling considerations of the analysis resulted in the selection of two representative reactor plant sites (Byron and Salem) from a survey of existing sites. These sites provided representative settings in terms of geographic location, population distribution, and meteorological conditions that would represent a broad spectrum of potential site locations.

The performance of . the APWR consequence analysis is being limited to the airborne pathway of fission product transport. Review of studies undertaken during previous consequence analyses has generally found the comparative consequences of the rainout to fish flesh and of the liquid pathways to be sufficiently small, with respect to consequences from airborne pathways as to be negligible without affecting the validity and usefulness of the risk studies. This conclusion is expected to be true for the W APWR plant at most _

sites, and results from the low potential for melt-through of the containment, basemat and subsequent radionuclide release to liquid pathways as well as to the small population dose estimates one would expect to calculate due to consumption of contaminated fish.

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i O 6.2 AIRBORNE PATHWAYS CONSEQUENCE ANALYSIS 6.2.1 GENERAL O The airborne pathways consequence analysis is carried out for estimating the population health effects resulting from the release cf fission products from the containment into the environment subsequent to a postulated core melt accident. The methodology utilized in this study is consistent with major-safety studies on nuclear power plants such as the Reactor Safety Study (Reference 1) and the Zion and Indian Point Probabilistic Safety Studies (References 2 and 3).

6.2.2 AIRBORNE PATHWAYS CONSEQUENCE MODEL The CRAC2 computer code (Reference 4) was used to estimate the airborne pathway consequences. This code is an updated version of the CRAC code that was employed in the Reactor Safety Study. The CRAC code is developed, O supported and continually improved by Sandia National Laboratories. All the modifications recomended by Sandia up through December 20, 1982 (Change 42.82/12/20) as well as changes for environmental calculations have been implemented in this version. Conversion of the code for use on the CRAY-1 computer was carried out. The code has been properly verified using the sample problems provided by Sandia. The physical phenomena modeled in the*

code have been described elsewhere, (References 4 and 5). Consequently, no attempt is made to describe the details of the code in this report; however a very brief overview is provided below.

O Transport of the fission products released from the containment due to the prevailing wind is modeled. Vertical rise of the plume depends on the release energy associated with the particular release category. Time dependent motion of the plume is simulated making use of the meteorological data - wind speed, O turbulence, etc. Radioactive decay of the nuclides and their daughter buildup are also accounted for.

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One of the improvements of the CRAC2 code over the original CRAC model is the O

provision for " bin sampling" of the meteorological data. The meteorological data file contains pertinent information for each of the 8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months in a year. Performing plume dispersion calculations for each set of this data d would be prohibitively expensive. What is normally done therefore is to pick approximately 100 samples and to perform a calculation for each of these start times. The sampling can be random, at regular intervals or by bin sampling.

In the latter method, all of the hourly data are first classified into groups or bins (there are 29 bins defined in CRAC2) on the basis of similarity of weather sequences. Then samples are picked from each bin. This insures against exclusion of important weather sequences or inappropriate weighting of the weather sequences. Bin sampling is utilized in the current study using four samples from each bin.

As the plume travels through the atmosphere, deposition of the particulate radioactive material takes place. When rain or snow occurs, the deposition l

rate is enhanced, depending on the rate of precipitation. The deposition af fects both airborne and ground concentrations of the radioactive material.

Both dry deposition and wet deposition are modeled by the code. Noble gases are not removed by deposition.

The radiation doses received by individuals are from the passing radioactive cloud (plume) and the material deposited on the ground. The cloud doses could, be due to direct radiation and due to inhalation of the radioactive material suspended in the air. These processes will last only during the passage of the cloud over the affected population. Doses f rom the deposited radioactive material are via three paths: direct radiation from the radionuclides, inhalation of resuspended material and ingestion of contaminated food and water. The CRAC2 code simulates all of these dose paths.

In order to assess the ef fect on the entire population, the individual doses are combined with the population distribution. The area within a 350 mile radius around the reactor is modeled on a circular grid that has ' finer radial divisions closer to the plant for better resolution. The entire grid consists e._

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of 33 rings (or circular intervals) and 16 sectors (circumferentially around O

W APWR-PSS 6.2-2 September,1985 87420:10/090985

r O the rings) fonning a total of $28 discrete areas. The population in each area is modeled based on the 1980 census data as described in Section 6.3.2.3.

Several protective action measures to reduce it.e radiation doses are modeled.

O These include evacuation of the nearby population to prevent or limit the cloud dose and early ground dose, and sheltering of the non-evaceees to limit the doses they receive. The emergency response model is typically comprised of various strategies and is described in Section 6.3.3.2. Long term O relocation of people, and interdiction and decontamination of land in the contaminated area are other steps that could limit the radiation doses and are modeled in CRAC2.

The population health effects are determined from individual radiation doses and dose response characteristics. The health ef fects that are focused upon in this study are acute fatalities, bone marrow dose, w' hole body dose, thyroid dose, latent cancer fatalities (excluding thyroid cancers), and total population whole body dose. Early fatalities are dominated by bone marrow dose. The latent cancers occur over a period of several decades. CRAC2 also calculates specific economic consequences of radioactive release which will not be addressed in this study.

6.

2.3 REFERENCES

1. U.S. Nuclear Regulatory Connission, ' Reactor Safety Study: An Assessment of Accident Risks in U.S. Nuclear Power Plants" WASH-1400 (HUREG/75/014),

October 1975.

Volumes 1-12. -

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4. Ritchie, L. T., J. D. Johnson and R. M. Blond, " Calculations of Reactor Accident Consequences Version 2 -

CRAC2 - Computer Code Users Guide,"

NUREG/CR-2326, SAND 81-1994, Sandia National Laboratories.

5. Aldrich, D. C., J. L. Sprung, et. al., " Technical Guidance for Siting Criteria Development," NUREG/CR-2239, SAND 81-1549, Sandia National Laboratories.

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T O 6.3 GENERAL INPUT AND HETH000 LOGY 6.3.1 SITE SELECTION The characteristics of the area surrounding a particular site can play an integral part in the calculation of estimated consequences for a postulated severe accident sequence. Since the W APWR design is not presently associated with any particular site, it was both' necessary and desirable to select a site which was considered to be representative of possible future site locations.

With regard to the essential items of site description such as geography, meteorology, and demography, it was decided that no one site could 'be an ideal ,

representative. As a result, two sites have been chosen for use in the analysis in an attempt to represent as many diverse site conditions as possible.

The accessability of pertinent data for use in a consequence analysis was a primary concern, and as a result the two sites selected were chosen f rom a O survey of existing reactor sites. The site locations for which the consequence analysis has been perforned are Byren, IL and Salem, NJ. These sites are considered to be representative of future reactor locations by virtue of the following attributes: 1) Geography - the Salem,' NJ site has a east coast location, while the Byron, IL site is located inland on a river shore in the midwest, 2) Heteorology - the Salem, NJ site has a coastal' climate with warm summers and mild winters, while the Byron, IL site has a continental climate with warm sumers and cold winters; 3) Demography - the Salem, NJ site has a relatively high surrounding population distribution with l.

2.6 million persons residing in a 0 to 40 mile radius, while the Byron, IL site has a medium population distribution with approximately 632,000 residents in the same O to 40 mile radius. The combination of these various characteristics as embodied in the two sites selected for analysis is considered to provide two locations which are likely to give an indication of the consequences of severe accidents at future reactor plant construction sites. Once a site is actually selected for construction, it will be necessary to confirm that the chosen site's characteristics are approximately encompassed by the two sites examined in this analysis.

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6.3.2 SITE CHARACTERISTICS O

6.3.2.1 GEOGRAPHY The Salem site is located on the southern part of Artificial Island on the east bank of the Delaware River in Lower Alloways Creek Township, Salem County, New Jersey. The point of intersection of the centerlines of the two containment buildings and the auxiliary building is located at Latitude 39*27'46' North and Longitude 75*32'08" West. While called Artificial Island, the site is actually connected to the mainland of New Jersey by a strip of tideland formed by hydraulic fill f rom dredging operations on the Delaware River by the U.S. Army Corps of Engineers. The site is 15 miles south of the Delaware Memorial Bridge, 18 miles south of Wilmington, Delaware, 30 miles southwest of Philadelphia, Pennsylvania, and 7-1/2 miles southwest of Salem, New Jersey. The location of the site with respect to major cities in the northeast is shown in Figure 6.3-1. The 700 acre site in which the Salem Nuclear Generating Station is located is owned by Public Service Electric &

Gas. The minimum distance between the reactors and the exclusion area boundary (property line) is 1270 meters. Additional infonnation pertaining to the site and surrounding area is available in the Salem FSAR (see Reference 1).

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The Byron site is located in North Central Illinois in Ogle County near the Rock River and approximately 50 miles southwest of Chicago. The center of the reactor containment buildings is located at Latitude 42' 04' 30' North and Longitude 89* 16" 55' West. The main site area, owned by Commonwealth Edison, occupies 1398 acres while the transmission and pipeline corridor to the Rock River occupies an additional 384 acres. The city of Rockford is the nearest significant population center and is located approximately 17 miles northwest of the site. The location of the site with respect to nujor cities in the midwest is shown in Figure 6.3-2. The minimum exclusion area boundary is 445 meters as measured f rom the outer containment wall. Additional inf ormation pertaining to the site and surrounding area is available in the Byron FSAR (see Reference 2).

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l The selection of these two geographical sites for the CRAC2 analysis provides a set location which is then characterized by the nature of the offsite 3

surroundings. As a prerequisite for this, a circular grid is established as described in Section 6.2.2. For the present analysis, the grid was defined by ,

33 rings (350 miles) and 16 sectors and input via the CRAC2 subgroup SPATIAL to provide discrete areas for input of offsite data. Overlay of this circular ,

grid onto a site-centered map was then used to determine the state and f

habitable land fraction for each of the discrete areas. This data was entered into the CRAC2 code through subgroup TOPOGRAPHY. The specification of the state for each area then permits computation of economic effects through I direct access to regional cost data assoc'iated with each state as available in the CRAC2 subgroup ECONOMIC. The data supplied in subgroup ECONOMIC is identical to the standard data set in the CRAC2 User's Guide (Reference 3). A sample of the actual data as input for the SPATIAL, TOPDGRAPHY, and ECONOMIC subgroups appears as part of Table 6.3-1 and Table 6.3-2.

l 6.3.2.2 METEOROLOGY The Salem FSAR (Reference 1) describes the Salem -site as being lccated in a region intersecting two climatic zones. They are humid continental and humid f subtropical. Both zones have characteristics of warm suneners and mild winters. Sununer maximum average temperatures are near 80 degrees Fahrenheit, and the coldest month is .lanuary with average daily temperatures o f.

approximately 32 degrees Fahrenheit. Most rainfall has been shown to occur in the sunener months.

i The Byron FSAR (Reference 2) describes weather at the Byron site as typically continental, with cold winters, warla suneners , and frequent short-period i fluctuations in temperature, humidity, cloudiness, and wind direction. The annual average temperature in the Byron site area as represented by Rockford j weather station . data is 48.1 degrees Fahrenheit, with extreme temperatures having ranged from a maximum of 103 degrees . Fahrenheit to a minimum of minus j 22 degre'es Fahrenheit. According to the Byron FSAR, annual precipitation has averaged about 37 inches for the 40-year period of 1937 through 1976 with minimum and maximum amounts being 23.25 inches and 56.48 inches respectively.

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CRAC2 requires meteorological data for use in calculational models which O

simulate weather sequences by determining the rate at which the radioactive plume will travel, disperse, and be depleted. This meteorological data must be in the form of 8760 hourly weather observations (one year) and consist of measurements of such parameters as wind speed, wind direction, stability class, and precipitation. For CRAC2 purposes, such data is normally attached as an input file which is then available for calculations made in the dispersion / deposition submodel.

The meteorological data in the present CRAC2 analysis for the Salem and Byron sites is obtained f rom the nearest of the 29 National Weather Service (NWS) stations as utilized in the reactor siting study of Reference 4. The Washington. 0.C. NWS station was used for Salem meteorology, while the Moline, It NWS station was the source of the Byron data. As detailed in Appendix A, Section A.3 of Reference 4 the weather observations used were those of a Typical Meteorological Year (TMY) and were tailored for compatability with CRAC2.

The SITE input subgroup of CRAC2 classifies the above meteorological data O

according to a bin sampling method on the basis of similarity of weather sequences. Samples (4 in the current analysis) were then chosen from each bin to insure against exclusion of some weather sequences or inappropriate weighting of them. .

An additional important aspect of a site's meteorological description is the wind rose. For the purposes of CRAC2 analysis, the wind rose represents the probability of the wind blowing towards each of the 16 sectors of the CRAC2 spatial grid. The centerline of each of these sectors are the 16 cardinal l

directions on a compass (N, NNE, NE, etc.). The data used in this analysis for both the Salem and Byron sites, was reported in Appendix A, Section A.4, t

Table A.4-1 in the Siting Study of Reference 4. The wind rose values are input as part of the CRAC2 subgroup POPULATION and appear as part of the CRAC2

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sample input in Table 6.3-1 and Table 6.3-2.

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6.3.2.3 DEMOGRAPHY The of f site population distributions for the Salem and Byron sites as used in l the analysis were prepared from 1980 census data. As noted earlier in Section O 6.3.1 the two sites selected represent medium and high offsite population distributions within the range of what may be considered suitable for future sites of nuclear plant construction. The area within 40 miles of the Salem site contains approximately 2.6 million people while the same 40 mile radius O around Byron con *.ains approximately 632,000 people. At the Salem site, the

'ninimum distance between the reactors and the exclusion area boundary (property line) is 1270 meters. The nearest population center of 25,000 is Wilmington, Delaware, 18 miles north of the Salem site. At the Byron site, the minimum distance from the reactor containment to the exclusion area boundary is 445 meters. The nearest population center of 25,000 is Rockford, Illinois, approximately 15 miles northeast of the Byron site.

The CRAC2 compatible method for specifying the offsite population distributions consists of assigning population figures to each of the 528 discrete areas of the spatial grid as described in Section 6.2.2. The 1980 U.S. Census Data was processed into this required form.through application of a modified version of the SECPOP computer code as developed by the il .S.

Department of Connerce and the Environmental Protection Agency. The SECPOP code when supplied with a site -location (latitude and longitude) and with a, spatial grid centered at the site location assigns population figures to the grid areas in accordance with the strategy described in Reference 4, Appendix A Section A-2. For both the Byron and Salem sites, portions of the discrete areas of the spatial grid protruded intc Canada and necessitated the addition O of population figures to those assigned by SECPOP. The final population figures, as well as the windrose for each of 16 sectors, appear in the POPULATION subgroup in Table 6.3-1 and Table 6.3-2 as part of the CRAC2 sample input.

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6.3.3 ATMOSPHERIC RELEASE AND EMERGENCY RESPONSE 6.3.3.1 ACCIDENT SOURCE TERMS The basic component of all W APWR accident source terms is the inventory of Oi, 1

radionuclides in the reactor at the time the fission chain reaction ceases. '

The CRAC2 input subgroup, ISOTOPE, permits the input of initial activity inventories as well as other parameters for up to 54 radionuclides which contribute to accident source terms. End-of-cycle equilibrium inventories were generated by calculations performed with the ORIGEN computer code for the 3800 MWt W APWR. For this analysis the radionuclides were catagorized into six leakage groups according to their characteristics and in agreement with the fission product groupings produced by the MAAP code output (see Section 5.6.2). All of the isotopes of Cesium were assigned to the CSOH leakage group and not to the CSI leakage group. This was done in consideration of the substantial excess of Cesium present beyond the amount comitted to Cesium Iodide formation. In addition, the leakage fractions of the CSI and CSOH groups as calculated by MAAP were comparable for all accident sequences.

Therefore, the leakage f raction applied to the CSOH leakage group (and thus all Cesium) conservatively approximates the total amount of Cesium released.

The parent isotope and half-life are also entered as data in the Isotope

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subgroup for each radionuclide to facilitate decay calculations. The depletion rate of each radionuclide in an eventual radioactive cloud is modeled in CRAC2 through use of two additional inputs for dry deposition velocity and for the coefficient which determines rain depletion. The values of all of the above parameters as input to CRAC2 through the ISOTOPE subgroup can be found in the CRAC2 sample input of Table 6.3-1 and Table 6.3-2.

Two other CRAC2 input values which affect the atmospheric release are used to determine the initial behavior of the released plume. These two parameters are entered through the CRAC2 input subgroup DISPERSION and are the W APWR ,

reactor building (containment) length and height. The other options available to CRAC2 in the DISPERSION input group were not used. '

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l The CRAC2 input values supplied in the subgroup LEAKAGE provide the essential l characteristics defining a particular source term for atmospheric release.

j The W APWR base cases consisted of CRAC2 analyses of various accident l

- sequences which result in a release of radionuclides to the atmosphere. For l purposes of the base analyses only, accident sequences were assigned a

! probability of occurance of 1.0. The accident sequences used to create release fractions for the CRAC2 base case analysis were TE. AE, SE, AEFC, j

' SEFC, and several special sequences not analyzed with MAAP which are a j V-sequence and two basemat f ailure cases - (BM1) with sprays, and (BM2) no sprays, i

I j The accident analysis models in the MAAP code produce results which provide j: four of the release (source term) characteristics for each of the accident sequences. These four ' source term characteristics are input to the CRAC2

. LEAKAGE subgroup, and are: 1) time between reactor shutdown and release to j

the environment; 2) duration of the release (maximum of 10 hrs. allowed in j CRAC2); 3) fraction of each isotope leakage group to be released; and, 4)' sensible heat rate of the release, k

j lhe MAAP computer code calculated results for each of six radionuclide leakage groups in the current analysis. The amount of each group released to the

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environment as a function of time is a primary product of the MAAP i calculations. Beginning at the time of reactor. shutdown (time zero) the MAAP, models simulate the progression of events in a core melt sequence which eventually lead to environmental release. The time of release af ter shutdown l

as determined by MAAP is not identical for all six of the leakage groups, but j CRAC2 permits only one time of release initiation to be input. Therefore i engineering judgement was necessary to determine when release of a significant l amount of radionuclides would begin. Here, a significant amount is defined as a release fraction equal to lx10-6 of shutdown inventory. Once such a l

l magnitude of release was attained for any one leakage group, all leakage groups were assumed to be, gin their release to the environment.

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Duration of the release to the environment was also obtained from the MAAP O

calculational results, specifically from the data describing the fractional amounts of leakage as a function of time. The end of release to the environment for a specific leakage group was taken as the time when the release fraction of that group became approximately constant. This time, and hence the duration of the release, again differed for each leakage group.

Since CRAC2 allows only one value of release duration for the entire release (all leakage groups), a selection of the shortest time over which significant release occurred was made. In cases where the release duration for all leakage groups exceeded 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , the CRAC2 naximum permissable value of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months was used. Using the shorter release durations in CRAC2 produces l

conservative estimates of the distances to which thresholds of radiation dose l and ground contamination occur.

The release fractions for each of the six leakage groups become constant 1 (environmental release ceased) after. differing amounts of time as noted above. The values of the individual group release fractions at this point for the particular accident sequence being studied comprise the inputs for the CRAC2 analyses. The release fractions for the noble gas (Xe-Kr) leakage group were assumed to be 1.0 for all of the CRAC2 analyses performed. This assumption is especially conservative for the base AE, AEFC, SE, SEFC, and TE analyses where the release fractions calculated by MAAP were of the order of 10- or smaller. In cases where the release fraction did not become constant,.

the MAAP results were then obtained from runs for a time period of two days or longer. It was assumed that within this expanded time frame some action would be taken to terminate the release or at least, as indicated by the past majority of accidents, prevent the release from exceeding the value calculated for two days. An additional note here is that a uniform rate of release is the means by which the CRAC2 models allow radionuclides to escape to the environment.

In the process of the present }{ APWR work, modifications to the MAAP code have been made to enable calculation of the amount of sensible entrijy and decay heat accompanying the release of radioactive gases to the environment. These two sources of thermal energy contribute to the plume rise as calculated by the CRAC2 atmospheric dispersion models. While CRAC2 specifies the input of W APWR-PSS 6.3-8 September, 1985 87420: 10/090985

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!O f sensible heat rate only, the addition of the decay heat rate permits more

] accurate estimation of the thermal content of the plume. The heat rate utilized in all of the present CRAC2 analyses was calculated by applying the j MAAP values of total sensible energy and decay heat uniformly over the I% duration of the release.

Containment failure height for the W APWR was assumed to be below ground j level. With this assumed release height (LEAKAGE subgroup) the CRAC2 computer l

code assumes the plume to be entrained in the building wake and resets the i release height to gr;ind level.

A final input parameter required as part of the LEAKAGE subgroup is the l

j warning time. This variable represents the time elapsed from issue of l official warning to the beginning of atmospheric release, and is used in J

evacuation modeling. While the time of initiation of atmospheric release was a

determined from the MAAP code results as described above, the time at which official warning is given can vary. Different accident sequences and various i

l site specific emergency planning policies impact the time at which a general emergency is declared. To ensure both consistancy and conservatism in the present analysis, the declaration of a general emergency (official warning) l

{ was assumed to take place at the time of core uncovery (top of core). This ,

l time is readily available as part of the MAAP results which indicate the reactor vessel water level as a function of time. The warning time as defined in CRAC2 is then obtained by simple comparison with the time of atmospheric

l release. Tables 6.3-3 through 6.3-17 list all of the various CRAC2 inputs to j the LEAKAGE subgroup for each of the accident sequences modeled by the MAAP

^.VO) code and considered in the analysis.

i l The V-sequence source term was not specifically analyzed for the W-APWR

design. The source term characteristics used in this study were derived from 1

i an unpublished MAAP code analysis of the V-sequence source term for a l Westinghouse 4-loop 3411 Megawatt Thermal reactor design with the RHR system routed to an Engineered Safety Features Building. Plant designs'with the RHR I

system in a separate ESF building exhibit a significantly higher . source term

! for the V-sequence than those with the RHR in the general auxiliary building area due to the smaller volume of the ESF building which results in W APWR-PSS 6.3-9 September, 1985 8742Q:lD/090985 1

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signficantly smaller fission product residence times compared to large O

auxiliary building designs. While the release fractions were taken from the unpublished study, the other release characteristics (timing, etc.) were modified to reflect:

1. A full pipe rupture for the }{ APWR design as opposed to an RHR pump seal failure such as the IDCOR Zion reference analysis.

2. Different water volumes available for ECCS injection in the W APWR design.

The release characteristics used in this study are sununarized in Table 6.3-18. These are judged to conservatively umbrella the W APWR design response to the V-sequence.

The basemat failure source terms were also not specifically analyzed for the W APWR design. The two cases considered were derived from previous PRA studies for Zion, Indian Point, Sizewell B and Millstone Unit 3. The source term characteristics used in those studies are also used in this study and can be traced directly to those given in the Reactor Safety Study (WASH-1400). The basemat f ailure source terms are considered to be very conservative estimates since they are derived directly from WASH-1400. These source terms should represent the containment atmosphere fission product inventory at the time of basement f ailure with some reduction due to the filtering ef fect of the soil.

However, the WASH-1400 containment inventories are one or more orders of magnitude higher than those predicted by the MAAP code for similar sequences at long times af ter core melt as would be representative of basemat failure times. Comparison of these ' WASH-1400 source terms with the MAAP late containment failure source term in many cases shows the basemat failure source term to be greater when, in fact, the basemat failure source term would always be expected to be less due to the soil filtration mechanism. A sununary of the release characteristics are given in Table 6.3-19 for the case of containment sprays operational and Table 6.3-2 for the case of no containment sprays. l i

A complete description of the various source terms for the base case accident sequences of this analysis involves designation of all the appropriate l parameters in the ISOTOPE, DISPERSION, and LEAKAGE subgroups of CRAC2. This l report section has detailed both the sources of these CRAC2 inputs and the

'dAPWR-PSS 6.3-10 September, 1985 8742Q: 10/090985 1

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! methodology applied in their derivation. The CRAC2 sample input provided in lables 6.3-1 and 6.3-2 provide a listing of the subgroup inputs as utilized !

j for the analysis.

l l 6.3.3.2 EMERGENCY RESPONSE STRATEGIES i

j Specification of the various input parameters in the CRAC2 subgroup EVACUATE l provides the means of examining the effectiveness of various emergency f response strategies. In the present case of the W APWR reactor plant with l representative site locations of Byron, IL and Salem, NJ the emergency i response strategies employed represent a modification of those specified in ;

, the siting study of NUREG/CR-2239 (Reference 4). The most significant modification is a product of recent revisions to the CRAC2 code which enable ,

. the user to define a sheltering region beyond the maximum evacuation distance. Additionally, a duration of exposure to external groundshine may be [

specified for the sheltering region as well'as for the region beyond it. This [

greater flexibility in the method of analysis permits an expanded range of emergency response strategies.

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lable 6.3-21 sumarizes the emergency response strategies investigated in the

present CRAC2 analysis for W APWR. The numbers entered in the table represent the actual CRAC2 inputs used to specify the individual emergency response '

! strategies. The distance from the reactor site at which evacuees complete ,

j t j their evacuation does -not appear in the table, but a value of 15 miles was i

j .used in CRAC2. As in NUREG-CR-2239 (Reference 4), strategies 1, 2, and 3 were respectively assigned 30 percent, 40 percent, and 30 percent weightings. When 8

considered to be combined into a "Sumary Evacuation" strategy they represent l

a best estimate for consequence predictions. Identical delay times and response speeds were used in the siting study of NUREG/CR-2239. (Reference 4).

They are based on ' a statistical analysis of evacuation ' data gathered by the

{ EPA. Emergency response strategy 4 with an extended delay time of 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and ;

a response speed of 1 mph could represent a response impeded by severe weather p or a seismic event. Emergency response strategy 5 is investigated to provide

! an estimate of the consequences resulting from no public action in, response to-I i an accidental release. This strategy may -also be examined to investigate the health effects. resulting from a small fraction of the population not

participating in an evacuation.

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The EVACUATE subgroup of the CRAC2 input bases its calculation of latent effects and evacuation costs on the emergency action data defined in the last emergency response strategy entered which was strategy 2 of Table 6.3-21.

Therefore, in CRAC2 runs for the present analysis, the data representing emergency response strategy 2 (3 hr. delay,10 mph) was entered last because this strategy was shown in NUREG/CR-2239 to yield results nearly identical to those obtained for the "best estimate" summary evacuation.

The attenuation of dose to a person from either airborne or ground-deposited radioactive material depends upon the distance f rom the radiation source and upon the geometry and composition of the material through which the radiation m'st u pass. The most comprehensive single study of the effect of structures on shielding from airborne and ground-deposited radioactive material was performed as part of the Reactor Safety Study, Appendix VI (Reference 5). The Reactor Safety Study surveyed numerous sources and compiled shielding factors for structures of varying composition (wood, brick) as well as for conrnuting and outdoors. The variance in' the percentage of brick or wood housing units by geographical region was also considered, in addition to the frequency of public time spent at principal locations or activities. The result of this work was two tables which weighted all of these elements and arrived at average shielding factors from a passing cloud and from ground contamination for each of five geographic regions. Because of their makeup these average shielding factors are, for the -present analysis, considered to represent shielding factors for persons carrying on normal activity. The location of the Byron site places it in geographic region III, while the Salem site f alls into geographic region II.

The CRAC2 EVACUATE subgroup requires the input of four separate shielding factors for both cloud and ground shielding. These four shielding factors are for stationary evacuees, moving evacuees, persons sheltering, and persons taking no emergency action. For all but the moving evacuee category the Reactor Safety Study average shielding f actors representing normal activity were employed for the two W APWR representative sites. Thus, the Byron site analysis used 0.74 for cloud shielding and 0.31 for ground shie,lding. The Salem site analysis should have used 0.80 for cloud shielding and 0.35 f or ground shielding, but inappropriate values were inadvertently iriput f or the analysis. The actual numbers used were: 1) cloud shielding, stationary W APWR-PSS 6.3-12 September, 1985 i 8742Q: 10/090985

evacuees and persons taking no emergency action - 0.75, persons sheltering -

0.7; 2) ground shielding, stationary evacuees and persons taking no emergency l action - 0.33, persons sheltering - 0.05 (characteristic of large structures l

as opposed to f amily housing units). The effect of using these values is discussed in Section 6.4.3; The shielding factors utilized for coving evacuees at both sites was a comuting shielding factor of 1.0 for the passing cloud and 0.5 for ground contamination.

The breathing rate used for all persons was 2.66 x 10-4 m3 3-1 , which is an average figure for people during normal activity. These values, along with the other EVACUATE subgroup parameters, may be seen in Tables 6.3-1 and 6.3-2 as part of the CRAC2 sample input for the Byron and Salem sites.

6.

3.4 REFERENCES

1. Salem Generating Station, Updated - Final Safety Analysis Report, Public Service Gas and Electric Company.

2. Byron /Braidwood and Byron Station, Final Safety Analysis Report, Comonwealth Edison Company.

3. Ritchie, L. T., J. D. Johnson and R. M. Blond, " Calculations of Reactor Accident Consequences Version 2 - CRAC2 - Computer Code Users Guide,",

NUREG/CR-2326, SAND 81-1994, Sandia National Laboratories.

l

4. Aldrich. D. C., J. L. Sprung, et. al., " Technical Guidance for Siting Criteria Development," NUREG/CR-2239, SAND 81-1549, Sandia National Laboratories.

5. United States Nuclear Regulatory . Commission, " Reactor Safety Study, Appendix VI - Calculation of Reactor Accident Consequences," WASH-1400 (NUREG/75/014), October 1975.

s ..

O W APWR-PSS 6.3-13 September, 1985 B742Q:10/090985-

_ . . _ , - - A. - ._ __

-O TABLE 6.3-1 CRAC2 SAMPLE INPUT i

BYRON SITE l

(a C e

9 I

W APWR-PSS 6.3-14 SEPTEMBER, 1985 r

h

. . . . . - - - - . . ~ . . . . . - . . . - - - . . . - . . - . -.~~.--.---.----......~.--~~-__...~~-.._-u -

...---..----xua.

TABLE 6.3-1 (CONTINUED) ,

CRAC2 SAMPLE INPUT l 6 BYRON SITE o

(4,C

@ i f

i i

I i

9 !

t 9 -

l r

g .W APWR-PSS 6.3-15 SEPTEMBER, 1985 :

I I

1 __ _. . _ _ _ _ _ _ . _ . _ . . _ _ _ _ . - . , . _ _ . . _ _ _ . . . . _ . _ _ _ _ _ _ _ _

l l

l TABLE 6.3-1 (CONTINUED)

CRAC2 SAMPLE INPUT BYRON SITE (a.c l

l

. O O

O, e.

j W APWR-PSS 6.3-16 SEPTEM3ER, 1985 i

e 4 4-. 4 -$4JM---p-M4hAMaJ.*-%-m Ma a 4M.A. MJ m ma_.&_L-56. 44.d--==--M h2 h-wm-EA.-.<-Ana6.-3.En.ww- e a haJe=,ma p L uh-6Ama- -m- -.-.-mm w# m hs.a,-- _--u,,

I

- TABLE 6.3-1 (CONTINUED)

O CRAC2 SAMPLE INPUT BYRON SITE l

(4.C '

6 l i

i l

-f Y

e .

I

- I k

9 :

i e

E g

e -

1 .

t g ~y APWR-PSS 6.3-17 SEPTEMBER, 1985

O TABLE 6.3-1 (CONTINUED)

CRAC2 SAMPLE INPUT BYRON SITE W

I l I l I l >

_l l 5

E PWR-PSS 6.3-18 SEPTEMBER, 1985 l

l j

l l

TABLE 6.3-1(CONTINUED) !

CRAC2 SAMPLE INPUT BYRON SITE (

i

('8 , C ,

i a

I O !

l l

f t

f 9 l

~

6 h

6.3-19 SEPTEMBER, 1985

{.WAPWR-PSS l i !

j .

1 .. !

O TABLE 6.3-1 (CONTINUED)

CRAC2 SAMPLE INPUT BYRON SITE i I l C

I I I I i

I l l

t

{ y APWR-PSS 6.3-20 SEPTEMBER, 1985 l

. - . .. - - -. - . - . .~.- - .- .

. . -.-. . -... ._ _ -... -. . - -- --. -..- - - . - -i t

i i

f TABLE 6.3-1 (CONTINUED)

CRAC2 SAMPLE INPUT l

) BYRON SITE t

j

! (a.c j l ;

I I

I r

[

i

@ l i

t i

i

!9 .

l O. !

I W APWR-PSS 6.3-21 SEPTEMBER, 1985 1

_._._w,,.w --J----

W TABLE 6.3-1 (CONTINUED)

CRAC2 SAMPLE INPUT BYRON SITE (4,C S

6 6

a

.[ y AP'n'R-PSS 6.3-22 SEPTEMBER, 1985

j - _ . . . . _ . _ . _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ - ._.____.___ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

9 -l TABLE 6.3-1 (CONTINUED)

CRAC2 SAMPLE INPUT BYRON SITE i

i e

(a.C ;

f i i h

.}

i t

9 !

I

. i b

1

5 l

e t' 9

l f

g I

i

. -l i

9, f .

I g W APWR-PSS 6.3-23 SEPTEMBER, 1985 ;

1 .

I

)

l

- - - a -

O TABLE 6.3-1 (CONTINUED)

CRAC? SAMPLE INPUT BYRON SITE

- 9 (a.C D

0 l

9

{ W APWR-PSS 6.3-24 SEPTEMBER, 1985

- - - . - - - - - - ----- - w.,- , ,- - -- -- , - - - w- - -

I l

e l i

TABLE 6.3-1 (CONTINUED)

CRAC2 SAMPLE INPUT BYRON SITE I

1 (a.c ,

9 i i

j

\

l 9 i i

l i

L

f l !

I i !

l@

r t

t O. I r

. . l i

W APWR-PSS 6.3-25 SEPTEMBER, 1985 ;

l ;

\

n, - - . - . - ~ . , - - - . , - . . - - - - - - . - - - - . . .. -- - - ..-. - . -...- -.-..- - - -

O TABLE 6.3-1 (CONTINUED) ,

CRAC2 SAMPLE INPUT I BYRON SITE

. l ta.c l

t W APWR-PSS 6.3-26 SEPTEMBER, 1985

{

l

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

s I

i l

l l

O l TABLE 6.3-2 f CRAC2 SAMPLE INPUT SALEM SITE l

(4,C ,

. O !l i

O '

.t i

O :

L i

l , ,

i e

' W APWR-PSS 6.3-27 SEPTEMBER, 1985 l l ,

I

O TABLE 6.3-2 (CONTINUED)

CRAC2 SAMPLE INPUT SALEM SITE J I (a .c I I I l I

W APWR-PSS 6.3-28 SEPTEMBER, 1985

O i

TABLE 6.3-2(CONTINUED)

CRAC2 SAMPLE INPUT SALEM SITE .

(a.c !

@ l l

6 i

.1 i

r l

j i

t ..

W APWR-PSS 6.3-29 SEPTEMBER, 1985 l 1

)

5

,-em--w.,+_'_ -

,.w.w,, ,w------wgr

1 TABLE 6.3-2 (CONTINUED)

CRAC2 SAMPLE INPUT SALEM SITE (a.c e

i l

W APWR-PSS 6.3-30 SEPTEMBER, 1985 l

1

?

l l

l 4

TABLE 6.3-2(CONTINUED)

CRAC2 SAMPLE INPUT SALEM SITE ,

i (4,C I f

1 l

. i N

6

I i !

i A 4

9 .

i

+

!O l !

6.3-31 SEPTEMBER, 1985, lWAPWR-PSS ,

i 4

1 . . .

TABLE 6.3-2 (CONTINUED)

CRAC2 SAMPLE INPUT SALEM SITE d I l (a.c l I db w

l i I I l 7 l l t

W APWR-PSS 6.3-32 SEPTEMSER, 1985 1

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

1

?

^

TABLE 6.3-2(CONTINUED)

CRAC2 SAMPLE INPUT l SALEM SITE r

(a.c i

!O !

i

. t 8

I

. 1 i

1 O i 1

c l

t

. I f

W APWR-PSS 6.3-33 SEPTEMBER,~1985 r __ . __

O TABLE 6.3-2(CONTINUED)

CRAC2 SAMPLE INPUT .

SALEM SITE H I (a.c l I l

i

\

l l l

l l l l l

. l I 6.3-34 SEPTEMBER, 1985 W APWR-PSS m

. . , - . _ . = = - . . _ - - . _ . .

i l

G l TABLE 6.3-2(CONTINUED)

CRAC2 SAMPLE INPUT SALEM SITE o

3 l

(4.C 4

. l l

I

1 i

G G

i E-l yAPWR-PSS 6.3-35 SEPTEMBER, 1985 j 1 -

2 1

1 1 _ . - - , _ _ . _ _ . . . _ -

TABLE 6.3-2 (CONT 1 HUED) t CRAC2 SAMPLE INPUT SALEM SITE (a.C

, \

I i y APWR.PSS 6.3 36 SEPTEMBER. 1985 I

\ O .

TABLE 6.3-2 (CONTINUED)

CRAC2 SAMPLE -INPUT SALEM SITE g

l e

6 7

1 e

S 9

l I

f .

I' 8, I W APWR-PSS 6.3-37 SEPTEMBER, 1985 l l 1

p .

t. ----. - - . . _ . .

. TABLE 6.3 2 (CONTINUED)

CRAC2 SAMPLE IHPUT 1 SALEM SITE (a.c 1

5

\

i

\

t g

5 I -w APWR-PSS 6.3-38 SEPTEMBER 1985 a

9 TABLE 6.3-2(CONTINUED)

CRAC2 SAMPLE INPUT SALEM SI*.E

@ l I

(a.c '!

t i

i P

i-P f

t i

l I

!~

e P

j .

i W APWR-PSS 6.3-39 SEPTEMBER, 1985 t

l i

-rrwr----o.,-veww=w-swee--=n-**--e-++w-----e+----

l TABLE 6.3-3 CRAC2 LEAKAGE SUBGROUP INPUT TE BASE O

(a.c) f 9'

O l

l 9

l O

-W APWR-PSS 6.3-40 J PTEMBER, 1985

I l

!8 TABLE 6.3-4 CRAC2 LEAKAGE SUBGROUP INPUT TE CONTAINMENT ISOLATION FAILURE O 1

. (a.c)

O O

t t

O i

O .

I W APWR-PSS 6.3-41 SEPTEMBER, 1985 4 ,

i 1 -

1

l TABLE 6.3-5 CRAC2 LEAXAGE SUBGROUP INPUT TE EARLY FAILURE O

(a c)

O O

O

. O W APWR-PSS 6.3-42 SEPTEMBER, 1985

. . . _ . . . - - . . . . - - . - . - _ - - . ~ . . - . . - - . . - - . - . - . . . ~ - . . - . . . . - - -. . _ _ - . - = - - . . = _ - _ - - . - . _ _ -

, ino u o.,-o CRAC2 LEAKAGE SUBGROUP INPUT TE LATE TEMPERATURE FAILURE (a c)-

O ,

e l

1 O .

i i i l t

(

O !

4 y,APWR-PSS 6.3-43 SEPTEMBER, 1985

-r +- w = e w e e, w-e --

CRAC2 LEAKAGE SUBGROUP INPUT e

AE BASE O

(a.c)

O 6

0 t

e S

O O

x - a-ess e.2-o sE, m um. 1,Bs e

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

O TABLE 6.3-8 CRAC2 LEAKAGE SUBGROUP INPUT AE CONTAINMENT ISOLATION FAILURE 0

(a.c) !

O ,

e O

- l i

i O

i O .

O W APWR-PSS 6.3-45 SEPTEMBER, 1985 i

k l -

TABLE 6.3-g CRAC2 LEAKAGE SUBGROUP INPUT AE EARLY FAILURE O'

(a.c) l 9

i I

t G

l l

I t

I

! O l

5 l

W APWR-PSS 6.3-46 SEPTEMBER, 1985 i

l 1

,. ~

1 l

< I AULL b.1- 10 l

CRAC2 LEAKAGE SUBGROUP INPUT AE LATE FAILURE (a,c)

Q :

l l

i 1

O '

l O

i i

W APWR-PSS 6,3-47 SEPTEMBER, 1985 i

,_---m._,

4 _ . _ . _ _

TABLE 6.3 11 O

CRAC2 LEAKAGE SUBGROUP INPUT AEF EARLY FAILURE O

(a.c)

O l

9 i

l 9

O l

l i

. O 9

x ~~ss e.a.o sE,mm. ms O l

a.a. .a-.s -.a-w-n ._ a. , ._. a. a s. .x ---.a. w.a--

.mu4 a. --+.w--a2-.s--. .-a,. .ea . s .....n = -.u..- -----n a--

1 i

4

, O TABLE 6.3-12 CRAC2 LEAKAGE SUBGROUP INPUT AEFC BASE

. O (a.c) ;

f

}

i O :

l l !

i i

I O

e e8 O y APWR-PSS 6.3-49 SEPTEMBER, 1985

r TABLE 6.3-13 CRAC2 LEAKAGE SUBGROUP INPUT AEFC CONTAINMENT ISOLATION FAILURE O

(a c)

O O

' O

: O 6.3-50 SEPTEMBER, 1985 YAM-PSS l

TABLE 6.3-14 CRAC2 LEAKAGE SUSGROUP INPUT AEFC EARLY FAILURE O ,

(a.c) 5 i

l

[

O i

t e

G i

O '

'O .

O y_APWR-PSS 6.3-51 SEPTEMBER, 1985 ;

[

.~c... $

CRAC2 LEAKAGE SUBGROUP INPUT SE BASE O\

(a.c)

O I

l O

S l

l O

O m

W APWR-PSS 6.3-52 SEPTEMBER, 1985

~

1

i l

O TABLE 6.3-16 I CRAC2 LEAKAGE SUBGROUP INPUT !

i SEFC BASE (a.c)

O .

[ t t

l O .

L f

I l

I O l I

g 5 9 0 O

,WAPWR-PSS 6.3-53 SEPTEMBER, 1985 i

I

(_ _ - - _ _ . . _ - . _ - _ . . _ _ -.

,e m ... 1, o CRAC2 LEAKAGE SUBGROUP INPUT SEFC EARLY FAILURE O

(a.c)

O e

O O

I O

O W APWR-PSS 6.3-54 SEPTEMBER, 1985

_a - -

-e a- -:e. -- ---,emw-wasu-aw- =m e - - * --aa--m _m n-=.- - - - - ~ ~ . > -. - -- e->-.- a---a--4 a.=-s e,=-a E -.a- se-- - 4 2*a-i i

TABLE 6.3-18 CRAC2 LEAKAGE SUBGROUP INPUT V-SEQUENCE ,

I (a.c)

O i

{

O f

e e

i O i 4

I i i l

e i

O yA,..ss e.3.,e se,1<,,<,. 1,e,

)

l TABLE 6.3-19 9 >

CRAC2 LEAKAGE SUBGROUP INPUT AEFC BASEMAT FAILURE O

(a.c) !

O O

l l

O e

W APWR-PSS 6.3-56 SEPTEMBER, 1985

,. _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ . . _ _ . _ _ . . _ _ . _ _ _ . . _ . . _ _ _ , _ _ _ _ _ . . . _ . . - _ _ _ _ _ _ = . - _

l i -

1 e,

IABLt b.J-4U CRAC2 LEAKAGE SUBGROUP INPUT 1

AE BASEMAT FAILURE l

(a,C)

r --

@ l e

6

9 t

1 l

l 8 .

e

\l l@ '

6.3-57 SEPTEMBER, 1985 W APWR-PSS .

t b

-- . -,w -e- era v re weo w - = .ww w-

TABLE 6.3-21 O

EMERGENCY RESPONSE STRATEGIES Strategy Response Type and Delay Time Response External Grounds Number Application Region Before Response Speed Exposure Duration l' Evacuation (0-10 mi.) 1 hr. 10 mph -

Sheltering (10-17.5 mi.) - -

0.5 day No Response (17.5 mi.+) - -

0.5 day a

2 Evacuation (0-10 mi.) 3 hr. 10 mph -

Sheltering (10-17.5 mi .) - -

0.5 day No Respoase (17.5 mi.+) - -

0.5 day 8

3 Evacuation (0-10 mi.) 5 hr. 10 mph -

Sheltering (10-17.5 mi.) - -

0.5 da No Response (17.5 mi.+) - -

0.5 da 4 Evacuation (0-10 mi.) 5 hr. 1 mph -

Sheltering (10-17.5 mi.) - -

0.5 day No Response (17.5 mi.+) - -

0.5 day 5 No Response (entire area) - -

0.5 day a) Strategies 1, 2, and 3 comprise the respective 30% /40% /30% weight of the Sununary Evacuation strategy.

O i l

l O

W APWR-PSS 6.3-58 September, 1985 8742Q:10/090985

O FIGURE 6.3-1 SALEM SITE WITH RESPECT TO MAJOR CITIES amm lO -

JYR4C/M Recx rect * *'

O NEW SLBANY, g,,,, 803 TON NORK ggggw e

Hastforo e a tt (

s t M7ad .

(gy, I p ,

PfNM3Y N/R *

}\ ,

%RK .

' N*I Nd W yo

."#" rn ,

o ;

W

~j ) ariparic ocran f Bsirm Sont. Iconi. Isont swies y

NfST V/En/M/A MO1 y

k Mt canotasroy

.j ecanoao

"'" "'^" '

O c, -

t 2 g W APWR.PSS 6.3-59' SEPTEMBER, 1985 l l

O FIGURE 6.3-2 BYRON SITE WITH RESPECT TO MAJOR CITIES O

( .

I f Wisconsin n' f

e. A fp& s, .-

N

,a == ~-

osio ILLINOIS

- =

MISSOURI KENTUCKY U

O W APWR-PSS 6.3-60 SEPTEMBER, 1985 e

. O 6.4 CONSEQUENCE ANALYSIS 6.4.1 DAMAGE INDICES l The damage indices selected for study in this analysis are all health l consequences, economic consequences will not be addressed. The specific health consequences to be presented and discussed are the mean results for early fatalities, population with bone marrow dose greater than 200 REM, population with whole body dose greater then 25 REM, population with thyroid dose greater than 300 REM, latent cancer f atalities (due to early and chronic exposure, excluding thyroid cancers), and population total whole body MANREM.

It is important to note that all of the consequences to be presented are conditional and must be weighted by the frequency of release to have significance from the standpoint of risk. ,

6.4.2 BASE ANALYSIS RESULTS I The mean result for each of the six health consequences selected for examination in Section 6.4.1 will be presented for all accident sequences (18) whose leakage to the environment were modeled with CRAC2. The CRAC2 input for these accident sequences was as discussed in Section 6.3. Six health consequence tables are presented for both the Byron and Salem sites, with four of these including the mean result for the sununary evacuation strategy,,

impeded response case ( 5 h,r . delay, 1' mph), and the no emergency response case. The health consequence tables for latent cancer fatalities and -

population total whole body MANREM provide results which are limited to just the sunrnary evacuation strategy and the no emergency response case.

Tables 6.4-1 and 6.4-7 present the conditional mean number of early fatalities calculated by CRAC2 for the Byron and Salem sites respectively. Results appear for all eighteen accident sequence source terms for three emergency response scenarios. As is evident from the values calculated t$y CRAC2, the estimated mean number of early fatalities for the majority of source terms is *

[ ] The maxima for both site locations are [ (a,c)

] for the summary evacuation, impeded response, and no response scenarios respectively [ ] (a,c) i W APWR-PSS 6.4-1 September, 1985 8742Q: 10/090985

Tables 6.4-2 and 6.4-8 present the conditional mean number of the population O

receiving a bone marrow dose greater than 200 REM as calculated by CRAC2 for the Byron and Salem sites respectively. Results appear for all eighteen accident sequence source terms for three emergency response scenarios. As is i

evident from the values calculated by CRAC2, the estimated population with bone marrow dose greater than 200 REM for the majority of source terms is (a,c) [ ] or [ ] The maxima for both site locations are (a.c) [ ] for the summary evacuation, impeded response, and no (a c) response scenarios respectively [ ] l Tables 6.4-3 and 6.4-9 present the conditional mean number of the population receiving a whole body dose greater than 25 REM as calculated by CRAC2 for the Byron and Salem sites respectively. Results appear for all eighteen accident sequence source terms for three emergency response scenarios. The maxima f or (a,c) both site locations are [ ] for the summary evacuation, (a.c) impeded response, and no response scenarios respectively [

]

Tables 6.4-4 and 6.4-10 present the conditional mean number of the population O

receiving a thyroid dose greater than 300 REM as calculated by CRAC2 for the Byron and Salem sites respectively. Results appear for all eighteen accident sequence source terms for three emergency response scenarios. The noxima f or (a,c) both site locations are [ ] for the sumary evacuation,.

(a,c) impeded response, and no response scenarios respectively [

]

Tables 6.4-5 and 6.4-11 present the conditional mean number of latent cancer f atalities (excluding thyroid cancers) as calculated by CRAC2 for the Byron and Salem sites respectively. Results appear for all eighteen accident sequence source terms with outcomes based on early exposure only and early

,lus chronic exposure for two emergency response scenarios. The maxima for (a,c) both site locations for the 3-hour delay,10 mph emergency response are [ ]

(a,c) due to early exposure only, and [ ] due to early and chronic' exposure [

] The maxima for both site locations for the no emergency (a c) response case are [ ] due to early exposure only, and [ ] due to early (a,c) .and chronic exposure [ ]

W APWR-PSS 6.4-2 September, 1985 87420: 10/090985

O Tables 6.4-6 and 6.4-12 present the conditional mean populaton total whole body MANREM as calculated by CRAC2 for the Byron and Salem sites respectively. Results appear for all eighteen accident sequence source terms for two emergency response scenarios. The maxima for both site locations are O [ ] for the summary evacuation, and [

) for the no emergency response case.

(a,c) 6.4.3 UNCERTAINTY AND SENSITIVITY ANALYSIS The consequence analysis is performed using CRAC2, a complex computer model which treats a large number of diverse factors. These include site meteorology, radionuclide deposition parameters, population distribution (within 350 miles), plume dispersion, site evacuation characteristics',

shielding factors, radionuclide dose pathways, and dose conversion factors to calculate population health ef fects. Point estimate values (as opposed to a distribution of values) are used for most of the CRAC2 inputs. Although the intent was to use best estimate values for inputs, there is a tendency to select conservative estimates when values are uncertain. The input values used in the consequence calculations would thus tend to be conservative, lhe models describing atmospheric processes are based on substantial observational data. Variability in weather is treated by statistical sampling of a large body of weather data and performing consequence calculations for l each sample. Thus uncertainties in the weather data are relatively small.

l The processes of atmospheric dispersion and of radionuclide deposition on the other hand are treated by models based on experimental data. For atmospheric dispersion processes, the model represents a fit to data which has a O significant degree of variability, and hence uncertainties. The large sampling base employed in CRAC2 tends to reduce overall uncertainty in dispersion calculations to a relatively low level. Deposition processes on the other hand depend on particle size of the material being treated.

Particle sizes for radionuclides once they reach the atmosphere are not O 6.4-3 W APWR-PSS September, 1985 8742Q: 10/090985 4

clearly known; hence there is signif h.c uncertainty in calculations of the rate of deposition. Of particular concern are weather conditions which produce high early fatality estimates. These scenarios are a result of rain washing out radioactive material and depositing it in a local area; principal dose effects result from ground shine from the deposited radioactivity.

Estimates of concentration associated with ground deposition are probably skewed to the high side since the variability in rainf all in a storm is not accounted for in the model (deposition over a larger area than calculated is likely).

Also critical is modeling of evacuation. There is appreciable uncertainty associated with such modeling. One problem with modeling of evacuation processes is the lack of data regarding the way an evacuation would proceed under the variety of possible dose exposure regimes. For rain scenarios this f is particularly critical. For example, no credit is taken for the potential for avoidance of rain areas in the evacuation path based on either dose surveys and/or directions given to the evacuating population. Dose effects would be smaller for some scenarios if people are sheltered in place rather than evacuated. Such distinctions are not accounted for in the evacuation and dose models.

f The biological transport models and dose models employed were developed from a substantial data base and represent the state-of-the-art. They have relatively low levels of uncertainty associated with them.

Treatment of uncertainties associated with the CRAC2 calculations would require the identification of the uncertainties for all of the inputs as a minimum, and propagation of these uncertainties through the CRAC2 calculations. The present analysis was performed with the objective of providing a realistic assessment of the possible health consequences which could result from aW APWR core melt accident. This was, as previously mentioned, done as a best estimate ef fort utilizing conservative estimates of input values where uncertainty existed. Consequently, a specific Jjncertainty analysis was not deemed essential and was not performed. The analysis results as calculated with CRAC2 are considered to present a suitable repr'esentative estimate of W APWR accident consequences.

W APWR-PSS 6.4-4 September, 1985 8742Q:10/090985

O One assumption to which the calculation of accident consequences was seen to be sensitive was the quite conservative assumption that the entire noble gas l

inventory was released (release fraction = 1.0) during the base accident I sequences in which the containment did not fail (AE, AEFC, SE SEFC, TE). For _

all of these accident sequences the MAAP code calculated release fractions of the order of 10-3 or smaller released over a time period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The maximum duration of release allowed in CRAC2 is just 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . These conservatisms, in conjunction with early release start times and short warning O times for the no containment f ailure cases serve to magnify the level of the CRAC2 calculated results. This is the direct reason for the calculation of a

[ ] of early fatalities for the SE and SEFC no containment (a,c) failure cases. Additionally, the frequency distribution entered in the site matrices for the no containment failure category (represented by SEFC) is likewise af fected and thus quite conservative as confirmed by supplementary CRAC2 analysis.

An additional input for which the sensitivity of CRAC2 results was examined were the cloud and ground shielding factors as employed in the Salem site analysis. As noted in Section 6.3.3.2 inappropriate shielding f actors were inadvertently input for the CRAC2 Salem site calculations. Further analyses were done to investigate the effect of these inappropriate values and they revealed only a minor influence on results.

N O

O O W APWR-PSS 6.4-5 September, 1985 8742Q:10/090985

e TABLE 6.4-1 BYRON SITE MEAN NUMBER OF EARLY FATALITIES (all results are conditional)

SUMMARY

IMPEDED NO ACCIDENT EVACUATION RESPONSE EMERGENCY SEQUENCE STRATEGY (5 hr. delay. 1 mph) RESPONSE TE TE CONT. 150. FAILURE (a,t)

TE EARLY FAILURE TE LATE TEMP. FAILURE AE AE CONT. 150. FAILURE AE EARLY FAILURE AE LATE FAILURE AEF EARLY FAILURE AEFC AEFC CONT. 150. FAILURE AEFC EARLY FAILURE SE SEFC SEFC EARLY FAILURE V SEQUENCE BASEMAT FAILURE 1 (AEFC)

BASEMAT FAILURE 2 (AE) e M APWR-PSS 6.4-6 September, 1985 07d?n.1n/non00C

O -

TABLE 6.4-2 BYRON SITE POPULATION WITH BONE MARROW DOSE GREATER THAN 200 REM (mean number - all results are conditional)

SUMMARY

IMPEDED NO ACCIDENT EVACUATION RESPONSE EMERGENCY SE0VENCE J RATEGY (5 hr. delay.1 mph) RESPONSE TE TE CONT. 150. FAILURE (a,c) '

TE EARLY FAILURE TE LATE TEMP. FAILURE O AE AE CONT. 150. FAILURE AE EARLY FAILURE AE LATE FAILURE AEF EARLY FAILURE AEFC AEFC CONT. 150. FAILURE AEFC EARLY FAILURE SE SEFC SEFC EARLY FAILURE V SEQUENCE BASEMAT FAILURE 1 (AEFC)

O BASEMAT FAILURE 2 (AE)

O W APWR-PSS 6.4-7 September, 1985

TABLE 6.4-3 e

BYRON SITE POPULATION WITH WHOLE BODY DOSE GREATER THAN 25 REM O

(mean number - all results are conditional)

SUMMARY

IMPEDED NO ACCIDENT EVACUATION RESPONSE EMEPGENCY SE00ENCE STRATEGY (5 hr. delay.1 mph) RESPONSE TE TE CONT. 150. FAILURE (a,c)

TE EARLY FAILURE TE LATE TEMP. FAILURE AE AE CONT. 150. FAILURE AE EARLY FAILURE AE LATE FAILURE AEF EARLY FAILURE AEFC AEFC CONT. 150. FAILURE AEFC EARLY FAILURE SE SEFC SEFC EARLY FAILURE V SEQUENCE BASEMAT FAILURE 1 (AEFC)

BASEMAT FAILURE 2 (AE) e W APWR-PSS 6.4-8 September, 1985 87420:10/090985

O -

l l

TABLE 6.4-4 !

BYRON SITE l

POPULATION WITH THYROID DOSE GREATER THAN 300 REM (mean number - all results are conditional)

O

SUMMARY

IMPEDED NO .

ACCIDENT EVACUATION RESPONSE EMERGENCY SE0VENCE STRATEGY (5 hr. delay, 1 mph) RESPONSE TE TE CONT. 150. FAILURE (a,c)

TE EARLY FAILURE TE LATE TEMP. FAILURE AE AE CONT. 150. FAILURE AE EARLY FAILURE AE LATE FAILURE AEF EARLY FAILURE AEFC AEFCCONT.150.FkILURE AEFC EARLY FAILURE SE SEFC SEFC EARLY FAILURE V SEQUENCE BASEMAT FAILURE 1 (AEFC)

BASEMAT FAILURE 2 (AE)

O W APWR-PSS 6.4-9 September, 1985 07 8 96 1 n /00000 f

l l

TABLE 6.4-5 O!

BYRON SITE LATENT CANCER FATALITIES (exluding thyroid cancers - mean number - all results are conditional)

EVACUATION WITH 3 HR DELAY. AT 10 MPH NO EMERGENCY RESPONSE ACCIDENT DUE TO EARLY DUE TO EARLY AND DUE TO EARLY DUE TO EARLY AND SE0VENCE EXPOSURE ONLY CHRONIC EXPOSURE EXPOSURE ONLY CHRONIC EXPOSURE TE TE CONT. 150. FAILURE (a,c)

TE EARLY FAILURE TE LATE TEMP. FAILURE AE AE CONT.' 150. FAILURE AE EARLY FAILURE AE LATE FAILURE AEF EARLY FAILURE AEFC AEFC CONT. 150. FAILURE AEFC EARLY FAILURE

-SE SEFC SEFC EARLY FAILURE V SEQUENCE BASEMAT FAILURE 1 (AEFC) 8ASEMAT FAILURE 2 (AE)

O e

W APWR-PSS 6.4-10 September,1985 87420:10/090985

O TABLE 6.4-6 BYRON SITE

,O L

l POPULATION TOTAL WHOLE BODY MANREM

( (mean number - all results are conditional)

SUMMARY

NO ACCIDENT EVACUATION EMERGENCY SE0VENCE STRATEGY RESPONSE TE TE CONT. 150. FAILURE ( a ..c )

TE EARLY FAILURE o TE LATE TEMP. FAILURE p AE I O AE CONT. 150. FAILURE AE EARLY FAILURE .

AE LATE FAILURE AEF EARLY FAILURE AEFC AEFC CONT. 150. FAILURE AEFC EARLY FAILURE SE SEFC SEFC EARLY FAILURE V SEQUENCE BASEMAT FAILURE 1 (AEFC)

BASEMAT FAILURE 2 (AE)

O W APWR-PSS 6.4-11 September, 1985 B742Q:1D/090985 l

O TABLE 6.4-7 SALEM SITE MEAN NUMBER OF EARLY FATALITIES O

(all results are conditional)

SUMMARY

IMPEDED NO ACCIDENT EVACUATION RESPONSE EMERGENCY SEQUENCE STRATEGY (5 hr. delay. 1 mph) RESPONSE TE TE CONT. 150. FAILURE (a,c)

TE EARLY FAILURE TE LATE TEMP. FAILURE AE AE CONT. 150. FAILURE AE EARLY FAILURE AE LATE FAILURE AEF EARLY FAILURE AEFC AEFC CONT. 150. FAILURE .

AEFC EARLY FAILURE SE SEFC SEFC EARLY FAILURE V SEQUENCE BASEMAT FAILURE 1 (AEFC)

BASEMAT FAILURE 2 (AE)

O W APWR-PSS 6.4-12 September,1985 87420:10/090985

O TABLE 6.4-8 SALEM SITE O POPULATION WITH BONE MARROW DOSE GREATER THAN 200 REM (mean number - all results are conditional)

SUMMARY

IMPEDED NO ACCIDENT EVACUATION RESPONSE EMERGENCY SEQUENCE STRATEGY (5 hr. delay. 1 mph) RESPONSE TE TE CONT. 150. FAILURE (a,c)

TE EARLY FAILURE TE LATE TEMP. FAILURE AE O AE CONT. 150. FAILURE AE EARLY FAILURE AE LATE FAILURE AEF EARLY FAILURE AEFC AEFC CONT. 150. FAILURE .

AEFC EARLY FAILURE SE SEFC SEFC EARLY FAILURE Y SEQUENCE BASEMAT FAILURE 1 (AEFC)

BASEMAT FAILURE 2 (AE)

O .

O W APWR-PSS 6.4-13 September, 1985 87420:10/090985

TABLE 6.4 9 SALEM SITE POPULATION WITH WHOLE BODY DOSE GREATER THAN 25 REM (mean number - all results are conditional)

SUMMARY

IMPEDED NO ACCIDENT EVACUATION RESPONSE EMERGENCY SEQUENCE STRATEGY (5 hr. delay.1 mph) RESPONSE TE TE CONT. 150. FAILURE

(, )

TE EARLY FAILURE TE LATE TEMP. FAILURE AE AE CONT. 150. FAILURE AE EARLY FAILURE AE LATE FAILURE AEF EARLY FAILURE

. AEFC AEFC CONT. 150. FAILURE ,

AEFC EARLY FAILURE SE SEFC SEFC EARLY FAILURE V SEQUENCE BASEMAT FAILURE 1 (AEFC)

BASEMAT FAILURE 2 (AE) -

O O

W APWR-PSS 6.4-14 September, 1985 87420:10/090985

f.

N

'O I

TABLE 6.4-10 SALEM SITE O

POPULATION WITH THYROID DOSE GREATER THAN 300 REM (mean number - all results are conditional)

SUMMARY

IMPEDED NO ACCIDENT EVACUATION RESPONSE EMERGENCY SE00ENCE STRATEGY (5 hr. delay. 1 mph) RESPONSE TE TE CONT. 150. FAILURE (a,c)

TE EARLY FAILURE I TE LATE TEMP. FAILURE AE AE CONT. 150. FAILURE AE EARLY FAILURE AE LATE FAILURE AEF EARLY FAILURE AEFC -

AEFC CONT.. ISO. FAILURE -

AEFC EARLY FAILURE SE SEFC O SEFC EARLY FAILURE V SEQUENCE BASEMAT FAILURE 1 (AEFC)

BASEMAT FAILURE 2 (AE)

O W APWR-PSS 6.4-15 September,1985 e,n n.,ninencer

TABLE 6.4-11 O

SALEM SITE LATENT CANCER FATALITIES (exluding thyroid cancers - mean number - all results are conditional)

EVACUATION WITH 3 HR DELAY. AT 10 MPH NO EMERGENCY RESPONSE ACCIDENT DUE TO EARLY DUE TO EARLY AND DUE TO EARLY DUE TO EARLY AND SE0VENCE EXPOSURE ONLY CHRONIC EXPOSURE EXPOSURE 9NLY CHRONIC EXPOSURE TE TE CONT. 150. FAILURE (a,c)

TE EARLY FAILURE TE LATE TEMP. FAILURE AE AE CONT. 150. FAILURE AE EARLY FAILURE AE LATE FAILURE AEF EARLY FAILURE AEFC

. AEFC CONT. 150. FAILURE AEFC EARLY FAILURE SE SEFC SEFC EARLY FAILURE V SEQUENCE BASEMAT FAILURE 1 (AEFC)

BASEMAT FAILURE 2 (AE)

O M APWR-PSS 6.4-16 September,1985 87420:10/090985

f 1

TABLE 6.4-12 SALEM SITE ,

O POPULATION TOTAL WHOLE BODY MANREM (mean number - all results are conditional) l l

SUMMARY

NO l ACCIDENT EVACUATION EMERGENCY l SE0VENCE STRATEGY RESPONSE TE TE CONT. ISO. FAILURE (a,c)

TE EARLY FAILURE TE LATE TEMP. FAILURE AE AE CONT. ISO. FAILURE O' AE EARLY FAILURE 4

AE LATE FAILURE AEF EARLY FAILURE AEFC AEFC CONT. ISO. FAILURE AEFC EARLY FAILURE SE SEFC SEfC EARLY FAILURE V SEQUENCE

BASEMAT FAILURE 1

! (AEFC)

BASEMAT FAILURE 2 (AE)

~

I. .

}

W APWR-PSS 6.4-17 September, 1985

, 8742Q: 10/090985 i

k -._ . , . _

j l

i

!O j 6.5 APPLICATION OF RESULTS OF RELEASE CATEGORIES i This section describes the application of the results of the fission product ,

l

. and consequence analyses to the definition of the release category fission fd product source terms. The release category definition provides the link d

between the containment and site matrices in that the end points of the containment event tree represent the dependent variable in the containment matrix and the independent variable in the site matrices.

1 6.5.1 RELEASE CATEGORY APPLICATION 1

The definition of the release category source terms includes the definition of

! the fraction of the total core inventory of fission products released to the .

atmosphere, the time of the release initiation, the duration of the release

j. and the warning time available for offsite emergency actions. The definition also includes the energy of the release as 'it can impact the near-site i consequences.

lO

]

The containment event tree end points define basic release category bins which j may be considered and quantified through the fission product and consequence f analyses. These release categories were defined in Section 5.6.2 and are l reiterated in Table 6.5-1. Fission product and consequence analyses were

performed for a number of event sequences as presented in Sections 6.3.3 and 6.4.2. Based upon the results of these analyses, the quantification of the

fission product. release c'ategory parameters I's perf ormed for each release category as described in the following sections.

d The basis for the classification of the event sequences into release categories is detailed in Section 5.6.2. Since the containment event tree defines numerous sequences which yield the same release category, a representative sequence must be selected to represent all sequences in a given release category for the consequence analysis and risk. qualification process.

The basis for selection of the representative sequences used in thi_s study was to analyze the dominant representative sequences and chose the sequence with O

W APWR-PSS 6.5-1 . September, 1985 9115Q:10/091085

O the most severe consequences for a given release category. The results of the definition of the release category representative sequences are summarized in Table 6.5-2.

6.5.1.1 BYP RELEASE CATEGORY As described in Section 6.3.3, no specific analyses were performed for this release category. The release category definition was taken from other previous analyses that are from a representative plant that should conservatively' approximate the W APWR.

6.5.1.2 UCS RELEASE CATEGORY The UCS release category represents an event sequence with an unisolated containment and with the containment spray system operational for fission product scrubbing. Only one event sequence ( AEFC, unisolated) was analyzed for this release category and the results were used directly in the composition of the site matrices.

6.5.1.3 UCN RELEASE CATEGORY

- The UCN release category represents an event sequence with an unisolated containment and without the containment spray system operational for fission .

product scrubbing. Two event sequences were analyzed for this release category, TE and AE. The results of the consequence analyses for these event sequences, as presented in Section 6.4.2, indicate that the consequences for the AE sequence were generally more severe than the TE case due to the earlier release time, shorter release duration and shorter warning time. In those few instances in which the TE consequences were more severe, the dif ference was judged to be insignificant relative to the magnitude of the consequences.

Therefore the AE (unisolated) event sequence analysis was chosen to represent this release category.

O W APWR-PSS 6.5-2 September, 1985 9115Q:10/091085

O 6.5.1.4 ECS RELEASE CATEGORY The ECS release category represents an event sequence with an early containment failure and with the containment spray system operational for fission product scrubbing. Two event sequences were analyzed for this release category, SEFC and AEFC. The results of the consequence analyses for these event sequences, as presented in Section 6.4.2, indicate that the consequences for the SEFC sequence were generally more severe than the AEFC case due to the higher iodine, cesium and tellurium release fractions. In those instances in which the AEFC consequences were more severe, the difference was judged to be insignificant relative to the magnitude of the consequences. Therefore the SEFC (early) event sequence analysis was chosen to represent this release category.

6.5.1.5 ECN RELEASE CATEGORY The ECN release category represents an event sequence with an early O containment failure and without the containment spray system operations for fission product scrubbing. Three event sequences were analyzed for this release category, AE, TE and AEF. The results of the consequence analyses for these event sequences, as presented in Section 6.4.2, indicate that the

- consequences for the AE sequences were always more severe than the TE and AEF event sequences due to the shorter release initiation time, longer release .

duration and shorter warning time. Therefore the AE (early) event sequence analysis was chosen to represent this release category.

6.5.1.6 ICS RELEASE CATEGORY The ICS release category represents an event sequence with a containment f ailure at an intermediate time frame with containment spray operational for fission product scrubbing. No event sequences were specifically analyzed for this release category. Due to the lack of quantitative analysis results, the early failure release category definition (ECS) was conservatively chosen to represent the parameters describing this release category. This is, judged to O

W APWR-PSS 6.5-3 September, 1985 9115Q:lD/091085

be conservative in that the intermediate failure would result in a longer time O

to release initiation thereby allowing more time for fission product re. tov'al mechanisms to act on the inventory in the containment prior to release.

Therefore, the SEFC (early) source term was used to describe this release category.

6.5.1.7 ICN RELEASE CATEGORY The ICN release categor y represents an event sequence with a containment failure at an intermediate time frame without containment spray operations for fission product scrubbing. No event sequences were specifically analyzed for this release. category. Due to the lack of quantitative analysis results, the early failure release category definition (ECN) was conservatively chosen to -

represent the parameters describing this release category. This is judged to be conservative in that the intermediate failure would result in a longer time to release initiation thereby allowing more time for fission product removal mechanisms to act on the inventory in the containment prior to release.

Therefore, the AE (early) source term was used to describe this release category.

6.5.1.8 LCF RELEASE CATEGORY The LCF release category represents an event sequence with a late containment failure. This category considers event sequences both with and without the containment spray system operational for fission product scrubbing. Due to the late time of the containment failure, the natural fission product removal processes in the containment are nearly as effective in reducing the fission product inventory as the containment sprays. Two event sequences were analyzed for this release category: AE and TE. The results of the consequence analyses for these event sequences, as presented in Section 6.4.2, indicate that the consequences for the AE sequences were always more severe than the TE event sequence due to the shorter release initiation time, longer release duration and shorter warning time. Therefore the AE (late) event sequence analysis was chosen to represent this release category. -

W APWR-PSS 6.5-4 September, 1985 O

9115Q:10/091085

- _ _ _ _ _ - _ _ ____1___-_--__-

O 6.5.1.9 BMF RELEASE CATEGORY The BMF release category represents an event sequence with no containment failure above the ground level such that fission products could be released O

directly to the atmosphere, but where the core debris penetrates the containment concrete basemat. This results in the pressure relief f rom the containment going to an underground area where the fission products are liable to be filtered by the soil prior to release to the atmosphere. Two basemat failure sequences were analyzed in the consequence analysis section (Section ,

6.4.2) representing a case with containment spray operational (AEFC) and a ,

case without sprays (AE) as defined by WASH-1400. The results of the consequence analysis show that the case without containment spray is always more severe, in terms of consequences, than the case with sprays available.

Therefore, the AE basemat sailure case was chosen to represent this release category.

6.5.1.10 NCF RELEASE CATEGORY O The NCF release category represents an event sequence in which the containment 1, integrity is maintained throughout the course of the event sequence and no containment failure is postulated. Five event sequences were analyzed, as

. presented in Section 6.4.2, which include the AE, TE, SE, AEFC and SEFC plant damage states. The results of the consequence analysis show that the SEFC case results in more severe postulated consequences than any of the other cases. Thus the SEFC (no f ailure) case was chosen to represent this release category.

6.5.1.11 RELEASE FRACTIONS AND FINAL CATEGORY DEFINITIONS Table 6.5-3 gives the release information used for each of the accident sequences analyzed. The data is provided for each accident sequence and also includes the data for the bypass and basemat sequences. A final table is provided, Table 6.5-4 which gives the release information corresponding to each final release category as found in the containment matrix. -

O W APWR-PSS 6.5-5 September, 1985 91150: 1D/091085

TABLE 6.5-1 O

RELEASE CATEGORY DEFINITIONS SYMBOL DESCRIPTION BYP CONTAINMENT BYPASS OR SG TUBE RUPTURE, EARLY OR LATE MELT UCS UNIS0 LATED CONTAINMENT WITH SPRAYS UCN UNIS0 LATED CONTAINMENT - NO SPRAYS ECS EARLY CONTAINMENT FAILURE WITH SPRAYS ECN EARLY CONTAINMENT FAILURE - NO SPRAYS ICS INTERMEDIATE CONTAINMENT FAILURE WITH SPRAYS ICN INTERMEDIATE CONTAINMENT FAILURE - NO SPRAYS O

LCF LATE CONTAINMENT FAILURE BMF BASE MAT FAILURE NCF NO CONTAINMENT FAILURE O

O O

W APWR-PSS 6.5-6 September, 1985 91150: 10/091085

O TABLE 6.5-2

ACCIDENT SEQUENCES SELECTED TO DEFINE EACH RELEASE CATEGORY RELEASE ACCIDENT SEQUENCE SEQUENCE SELECTED TO ,

CATEGORY ANALYZED AND CONSIDERED DEFINE RELEASE CATEGORY O BYP NONE- USED EXISTING ANALYSES - -.

UCS AEFC, (unisolated) AEFC (unisolated) ,

UCN TE. AE, AEFC (unisolated) AE (unisolated)

ECS SEFC, AEFC (early failure) SEFC (early failure)

~

ECN TE, AEF, AE (early failure) AE (early failure)

ICS NONE- USED EARLY FAILURE SEFC (early failure)

ICN NONE - USED EARLY FAILURE AE (early failure)

LCF TE, AE (later failure) AE (late failure) ,

BMF NONE - USED EXISTING ANALYSES (WASH-1400) AE (of WASH-1400)

NCF AEFC, SEFC, TE, SE, AE (no failure) SEFC (no failure)

O .

O L W APWR-PSS 6.5-7 September, 1985 9115Q: 10/091085

WESTINGHOUSE PROPRIETARY CLASS 2 TABLE 6.5-3 ACCIDENT SEQUENCE RELEASE INFORMATION

, R'e'1 ease Release Accident Start Warning Release MAAP Fission Product Sequence Time ilme Duration Release tractions Category thrs.) (hrs.) (hrs.) XE-KR CSI TE SR RU CS0!

4 W APWR-PSS 6.5-8 September, 1985 8742Q:1D/090985 0 9 O O O 9 O

1 f l I1l1 5

8 9

O 1

1 r C e b

m e

t p

e S

O U R

O t

c u s R d n S o o r i P t c

n a o r i f s

N s e O i s 2 I f a T e E S A P l l S M A e A R A R L ) 0 M C d E e N Y u I R n A i E T t S E n A I 9 I o E S -

R C L C 5 P ( E O R 6 R 3 O P E

S U

C f

5 6

E

- E C

N E

U Q

i L E G B S N A R I T T K T N -

S E E E D X

.W I C

C A n e o ) *

, s i a t .

e a s l r r e u h R D (

e g s n )

a i .

e n e s O l r m r e a i h RW l (

e s )

a t .

e r e s l

e a m r t l h

'R S i t

' 5 8

O 9

0 9

S 0 S /

t e y P 0 n c r - 1 e n o R :

d e g WQ i u e P 2 c q t A 4 c e a A S C W_ UT O

' lllllll

ll 1 5

8 9

1 r

S e C bn e

t P

e S

U R

t c

u s R d n S o o r i F t c

n a o r i F s

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D Y

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O T H A G C N R I E K T S -

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. W L E

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s i a t .

e a s l r r e u h R D (

e g s n )

a i .

e n e s l r m r e a i h R W T (

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s a t .

e r e s l a m rh e t 1

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O 6.6 SITE MATRIX COMPILATION A set of six site natrices was assembled for both the Byron and Salem sites.

The six na trices per site represent the six health consequences (damage O indices) as selected for examination in Section 6.4.1. The matrices are composed of separate columns for each of the accident sequence source terms as assigned to the release categories in Section 6.5. The values appearing in the release category columns are the conditional probabilities of exceeding O the corresponding levels of consequence (damage magnitudes) as indicated in the first column. Each column is the actual CRAC2 output frequency distribution as obtained for the sununary evacuation scheme.

The six health consequences for which site matrices have been generated are:

1) early fatalities. 2) population with bone marrow dose greater than 200 REM,

3) population with whole body dose greater then 25 REM, 4) population with thyroid dose greater than 300 REM, 5) latent cancer fatalities (excluding thyroid cancers), and, 6) population total whole body MANREM. The site matrices for the Byron site appear at the end of this section as Tables 6.6-1 through 6.6-6. The site matrices for the Salem site follow, and appear as Tables 6.6-7 through 6.6-12.

. These site matrices as calculated for the two representative sites in the CRAC2 analysis are utilized in Section 7.0, Assembly of Plant Risk. ,

O O .

O W APWR-PSS 6.6-1 September, 1985 91150: 10/090985

BYRON SITE MATRIX O

EARLY FATALITIES (CONDITIONAL PROBABILITY DISTRIBUTION)

O a

(a.c)

O l

o O

9 O

W APWR-PSS

6.6-2

4 POPULATION WITH BONE MARROW DOSE GREATER THAN 200 REM i (CONDITIONAL PROBABILITY DISTRIBUTION)

O (a.c)

O J

o O

O O .

O v ~a-ess e.e.3 um"*a "88

BYRON SITE MATRIX POPULATION WITH kHOLE BODY DOSE GREATER THAN 25 REM (CONDITIONAL PROBABILITY DISTRIBUTION)

O a (a.c)

O O'

9 O

O Q

O BYRON SITE !! ATRIX POPULATION WITH THYROID DOSE GREATER THAN 300 REM (CONDITIONAL PROBABILITY DISTRIBUTION) l (a.c)

O O

I O W APWR-PSS

1 6.6-5 f I

BYRON SITE MATRIX O

LATENT CANCERS (EXCLUDING THYROID CANCERS)

(CONDITIONAL ,ROBABILITY DISTRIBUTION)

O (a.c)

Ol O

O O

y m R. s, se,1,sete. 1,Be O

I I

O BYRON SITE MATRIX l POPULATION TOTAL WHOLE BODY MANREM (CONDITIONAL PROBABILITY DISTRIBUTION)

(a,c)

O 4

O f

O O .

O W APWR-PSS SEPTEMBER, 1985

BYRON SITE MATRIX e

POPULATION TOTAL WHOLE BODY MANREM (CONDITIONAL PROBABILITY DISTRIBUTION) im 9 (a.c) e i e G

e O

W APWR-PSS '

6.6-8

O SALEM SITE MATRIX EARLY FATALITIES (CONDITIONAL PROBABILITY DISTRIBUTION) ~

O (a.c)

O i

l i

i O i 1

I i

O O .

S O y m a. SS ,.,., Ste m ma. m s

TABLE 6.5-8 SALEM SITE MATRIX POPULATION WITH BONE MARROW DOSE GREATER THAN 200 REM (CONDITIONAL PROBABILITY DISTRIBUTION)

(a.c) 9 Ol 9

O O

Q

s O . - - .

SALEM SITE MATRIX POPULATION WITH WHOLE BODY DOSE GREATER THAN 25 REM (CONDITIONAL PROBABILITY DISTRIBUTION)

(a.c)

O O

o

[

O O .

O y A. R.PSS

,.,.11 S m _ . 1,e, . .

I SALEM SITE MATRIX POPULATION WITH WHOLE BODY DOSE GREATER THAN 25 REM (CONDITIONAL PROBABILITY DISTRIBUTION) l l

(8,C)

O O

G e

0 6

4 g

O SALEM SITE MATRIX POPULATION WITH THYROID DOSE GREATER THAN 300 REM (CONDITIONAL PROBABILITY DISTRIBUTION)

O (a.c)

'O O

4 O

O .

e O y m a-PSS

.. 1, StPuMua. ms 1

(

SALEM SITE MATRIX O

POPULATION WITH THYROID DOSE GREATER THAN 300 REM (CONDITIONAL PROBABILITY DISTRIBUTION)

O (a.c)

O O

9 O

O S 0

IMOLL O.V-11 SALEM SITE MATRIX LATENT CANCERS (EXCLUDING THYROID CANCERS)

(CONDITIONAL PROBABILITY DISTRIBUTION)

(a.c) lO O

G O

O g

SALEM SITE MATRIX O

POPULATION TOTAL WHOLE BODY MANREM (CONDITIONAL PROBABILITY DISTRIBUTION)

O (a.c)

O O

O g

SALEM SITE MATRIX POPULATION TOTAL WHOLE BODY MANREM (CONDITIONAL PROBABILITY DISTRIBUTION)

(a c)

O O

9

1 I i

6.7

SUMMARY

AND CONCLUSIONS The performance of the CRAC2 analysis in conjunction with the core and containment analysis provide the means of conducting a more comprehensive evaluation of the enhanced ability of the W APWR design to mitigate the I l

j consequences of a severe accident. The environmental release of severe !

! accident ' source terms generated by the core 'and containment analysis were

! modeled with CRAC2 utilizing the offsite surroundings of the Byron, Il and s

{ Salem, NJ reactor sites. The primary -considerations in selection of these 5 locations were their overall suitability with regard to possible future reactor siting .and their representation of a spectrum of offsite characteristics.

i

( a . c )~

h iO The overall results indicate that the conditional probability of any acute fatalities for either site is [ ] for many sequences. The maximum mean '(a,c) number of acute fatalities for Byron is [ ). The maximum mean number of (a,c) i acute fatalities for Salem is [ ]. For latent cancers at BYRON the maximum "(a,c) mean number calculated is [ ). At SALEM's site the maximum mean number is (a.c)

[ ). These means correspond to reasonable evacuation. (a,c) i j The Plant Risk Assembly of section 7.0 incorporates the probability of occurrence of the release categories as defined by the various accident sequences. Accordingly, the Plant Risk Assembly enables estimation of the total public health risk posed by W APWR power generation.

O, f,

O W APWR-PSS 6.7-1, September, 1985

91150: 10/091085 i

,

ML20135J098

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SUMMARY

2.3 REFERENCES

3.4 REFERENCES

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ML20135J098

Text

SUMMARY

2.3 REFERENCES

3.4 REFERENCES

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

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SUMMARY

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