ML19309C281

Forwards Evaluation of Emergency Planning Zone Distances Per NUREG-0396 Guidelines.Requests Emergency Planning Zone & Ingestion Pathway Be Established at 5 & 30 Miles, Respectively

ML19309C281
Person / Time
Site:	Fort Saint Vrain
Issue date:	04/01/1980
From:	Warembourg D PUBLIC SERVICE CO. OF COLORADO
To:	Grimes B Office of Nuclear Reactor Regulation
References
P-80066, NUDOCS 8004080399
Download: ML19309C281 (27)

Text

{{#Wiki_filter:1 pubue servlee company eOdondo 16805 ROAD 19% PLATTEVILLE, COLOR ADO 80651 April 1, 1980 Fort St. Vrain Unit No. 1 P-80066 Mr. Brian K. Grimes Direct 6r, Emergency Preparedness Task Group Office of Nuclear Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

Fort St. Vrain Unit No. 1 Emergency Planning

REFERENCES:

1) P-79205 2) P-800ll

Dear Mr. Grimes:

In Reference (1), P-79205, we expressed our concern with the exten-sion of the EPZ to a ten (10) mile radius and the ingestion pathway to a fif ty (50) mile radius. We did not receive any acknowledge-ment of this letter, however, on January 8, 1980, we met with Messrs. Roe, Kuzmycz, and Williams in Bethesda to discuss our concerns. As a result of this meeting we agreed to evaluate the EPZ for Fort St. Vrain with reference to the guidelines of NUREG-0396 to provide fur-ther justification for smaller EPZ and ingestion pathway radii for Fort St. Vrain. We have completed our evaluation, and we are transmitting herewith five (5) copies of this evaluation for your review. The attached evaluation is self explanatory and fully supports our or'31nal con-tention that the EPZ for Fort St. Vrain can be considerably smaller than that for a 1000 MWe LWR which was utilized as basis for developing the guidelines of NUREG 0396. As a brief summary our evaluation indicates that the comparable EPZ for Fort St. Vrain would be about five (5) miles and a comparable in-gestion pathway would be about thirty (30) miles. The evaluation is based on conservative estimates of source term releases, worst case meteorological conditions and extremely conservative PCRV (Prestressed Concrete Reactor Vessel) leakage rates. l r;C \\ e1

I Mr. Brian K. Grim:s April 1, 1980 On the basis of our evaluation it is requested that the Fort St. Vrain EPZ and ingestion pa.:hway be established as five (5) and thirty (30) miles, respectively. It should be noted further that we have been through our bases with State officials, and we have complete agreement with the State con-cerning the inherent safety of Fort St. Vrain versus a LWR. We have also discussed our position with the local FEMA representatives, and again have concurrence that there is no justification for penalizing Fort St. Vrain with water reactor criteria. It is requested that you give this matter your immediate attention. As you are aware we are in the process of finalizing our Emergency Response Plan with the State as well as the emergency planning for station. Our planning is presently based on the smaller EPZ and ingestion pathway exposure radii and any changes in this criteria could have a significant impact on final acceptance of our plans, both for the station and the State. Very truly yours, ^ L SNMn Don Warembourg Manager, Nuclear Production Fort St. Vrain Nuclear Generating Station DW/alk I

EVALUATION OF EMERGENCY PLA! DING ZONE DISTANCES APPLICABLE TO T11E FORT ST. VRAIN NUCLEAR GENERATING STATION 1. Introduction In an NRC letter to all reactor licensees dated November 29, 1979, a request was made for information regarding estimates for,, evacuation of,various,,. areaa around nuclear power reactors. NUREG-0396, EPA 520/1-78-016 specified, based upon studies of a 1000 MW(e) LWR, that the emergency planning zone (EPZ) for plume exposure pathway should be about 10 mile radius about the plant and about 50 mile radius for the ingestion exposure pathway. Both design basis accidents with various active engineered safety features, and the accident release categories of the Reactor Safety Study (Wash 1400) were considered in determination of the specified distances. A factor in establishing the about 50 mile radius ingestion exposure pathway is that distance has been evaluated and discussed in detail in Chapter 2 of Facilities Safety Analysis Reports and in Environmental Reports. The purpose of the evaluation presented herein is to determine the appropriate EPZ distances for Fort St. Vrain which provide the general public equivalent pro-tection that would be afforded by the 10 mile and 50 mile EPZs for a similar sited 1000 MW(e) LWR. At the onset of the investigation it was believed the principle differences would be a smaller source term due to size, 330 FN(e) at 39" efficiency I versus 1000 MW(e) LVR, and time to perfocm emergency actions due to the relatively slow heacup rate provided by the large graphite core of an HTGR. It was confirmed by the detailed study that the FSV smaller source term coupled with unique slou and gradual core heacup characteristic of the plant, did in fact justify sharply re-duced EPZ distances for Fort St. Vrain and significantly longer times to perform offsite emergency actions. m

• O

.-g am o =imeP gm se i. esmee = wmassum ..-e-ey = - w we -

i I 2. Conduct of Study i This section describes the. methodology, parameter assignments and assump-tions used in the conduct of this EPZ study. Section 2.1 describes the accident j scenario used as the basis for FSV EPZ determination as well as the description of what was assumed to be a comparable accident scenario for the LWR. Section 2.2 describes the conservatisms applied to the accident scenario. Section 3 describes the conservativa mateorlogical conditions assumed to prevail during the duration of the EPZ determining event. Dose computation methodology and parameter assignments and assumptions appear in Sections 4 and 5 respectively. The appropriate EPZ distances for FSV are presented in Section 6. Conclusions of this study are presented in Section 7. 2.1 Accident Scenario Used to Establish the EPZ for FSV l The FSV Design 3 asis Accident #1 (DBA #1) which was the only accident iden-l tified and analyzed in the FSAR that results in extensive core damage, was selected as the representative scenario used to establish the plant's EPZ. This event is l postulated to be initiated by a "non-mechanistic" permanent loss of forced cir-culation while operating at full power. The reactor is scrammed by the plant protection system and all attempts to restore forced circulation using the multiple heat sinks, circulators, and motive power for the circulators fail. When it becomes apparent to the plant operators that the loss of forced circulation is permanent, e.g. when restoration of cooling would cause steam generator damage, the primary coolant system would be depressuri=ed to storage in a controlled 4 l manner through the helium purification system. The reserve shutdown system would be operated after this initial period to assure an adequate shutdown margin. I l I

Because of the large heat sink provided by the graphite core, considerable time is availtble to initiate primary coolant depressurization and to restore forced circulation. The FSV FSAR specifies the time available to initiate de-pressurization to be 5 hours, which was later amended by PSC letter P-77250 dated December 22, 1977 to be 2 hours. The reduction in time was due to the re-evaluation of the capability of the helium purification system to process primary coolant during the planned depressurization and release of the clean primary coolant to the reactor building ventilation stack. Thus, the de-pressurization of the PCRV is now initiated after 2 hours and completed 7 hours i later (or 9 hours from the onset of the accident). The fuel is slow to heat up due to the large heat sink provided by the core graphite. A peak average active core te=perature of 5400 F is reached about 80 hours after the onset of the accident. At this temperature, the core structural integrity and geometry are not compromised since the vaporization i temperature of graphite is 6900 F. Peak activity released to the primary coolant, considering decay, is reached about 24 hours into the accident. PCRV containment integrity is maintained through heat removal by the liner cooling system in the " redistribution mode" which maximizes cooling in the top i head of the PCRV. Leakage of primary coolant from the PCRV is assumed to occur at a conser-vatively high leakage rate of 0.2% of the primary coolant inventory per day. The reactor building ventilation system maintains continuous venting and processing of the reactor building environment at 1.5 building volumes /hr during I the entire period of the accident. I

2.2 Con'irvetiem9 Applied to DBA #1 A:Tumptiona 2. 2._1 Primary Coolant Leakage Rate During DBA #1 The FSV FSAR DBA #1 (appendix D, page D.1-56) assumed an arbitrarily con-servative and non-mechanistic estimate of PCRV leakage after the intentional depressurization by assuming that the liner has failed completely (or does not exist) and only concrete permeability controls the leakage. An internal 5 psi Pressure differential was assumed which purportedly gave a PCRV leak rate of -5 8.33 x 10 fraction per hr (0.2 / day). Reference was made to Question II.7 of Amendment No. 9 of the FSV FSAR for the calculation of the permeation rate for the FSV PCRV concrete under these conditions. Examination of Question D.2 revealed simply the conclusion that a 5 psi positive differential pressure led to 0.2:/ day and 2 psi positive differential pressure led to 0.08%/ day. Question II.7 also did not provide details of the calculation of the 0.2 / day rate. However, considerable detail and a derivation was provided for the analysis of leakage rate tests at high pressure. The fol-lowing equation was provided(egn. 14 on page II.7-8): 1\\ -5 W (1b/ day) = 1.13 x 10 g -6 + 2.2 x 10 p 2,p 2 (eqn. 14) Where AP = PCRV inside pressure in psig APo = PCRV inside pressure in psig for which the net compressive stress in concrete = 0 A = Face area of concrete, ft X = Concrete thickness, ft P - Permeation or high side pressure, psia g P = Ambient or low side pressure, psia 2 Nu=crical values were inserted for P = 845 psig with the assumption that APo g was approximately equal to P in the following equation (eqn. 15 on next page): g 9 )

~ -5 9000 857.5 -7 9000 2 2 w = 1.13 x to x 1" + 9. x 0 (857.5 - 12.5 ) 10 12.5 10 = 0.43 + 602 = 600 lb/ day (egn. 15) The first item to note is that the coefficient for the second (laminar i flow) term is in error which is most likely a single error in transcribing -7 from equation 14 to 15 since equation 13 has the 9.1 x 10 coefficient. Equation 15 should read: W = 1.13 x 10 x 0 -0 2 2 1" + 2.2 x 10 (857.5 - 12.5 ) 0 2. 0 = 0.043 + 1445 = 1450 lb/ day (egn.15 revised) j l The second item is that the AP/APo term has been dropped in going from i eqn. 14 to egn. 15, which is significant if it is assumed that these equations are appropriate for evaluating the leak rate at P =_5 psig. Including this factor, g we find: Pressure F 1 lb/ day %/dav j (psig) Eqn 14 15 15 Revised 14 15 15 Revised Given App D; j Amend 9 Question D.2 3 5 .0019 .13 .30 .001 .07 .17 .20 Amend 9 3 l Question D.2 i i 2 .0003 .046 .107 .0001 .025 .059 .08 l I 4 h ~.. m oo ~.a

Since equation 14 is the appropriate equation, the 0.2 / day leak rate is 1 conservative by a factor of 200. For purposes of this evaluation, the historic 0.2 / day is assumed to exist as an upper limit of all potential contaminated primary coolant leakage including permeability through the PCRV concrete. This is judged to be conservative since the primary coolant with any significant activity is contained within the PCRV i or helium purification components contained in wells within the PCRV. I l.' For this study, the PCRV leak rate of 0.2:/ day is conservatively assumed to apply for the total accident duration period beginning at t = 0 hours. In actuality, this arbitrarily conservative estimate of PCRV leakage would not apply during the period of controlled PCRV depressurization or in the subsequent post depressurized period prior to an assumed liner failure. However, for con-servatism this PCRV leak rate will be applied uniformly for all time. 2.2.2 Radionuclide Source Terms for DBA-1: As previously stated, the fuel within the graphite core is slow to heatup during DBA#1. After about 7.5 hours, some portions of the slowly heating core will have reached the FSAR fuel particle coating failure temperature of 1725 C (3137 F). Once reaching the FSAR fuel particle coating failure temperature, the fission products are assumed, for purposes of this evaluation, to be released in i a time dependent fashion normalized per the TID-14844 core release assumptions. For release to the primary coolant within the PCRV, the TID-14844 conditions assume a release of 100% of noble gases, 50% of the iodines and 1% others. I Consistent with TID-14844 release assumptions, 50% of the iodines plateout f within the primary coolant system resulting in a depletion of the iodine to 25: l of core inventory in the resultant PCRV leakage to the reactor building atmosphere. l I l

Tharo TID-14844 rolracco and pictscut argumpticna cra vary conc 3rvrtiv2 for an HTCR but never the less have been utilized in this evaluation. This conservatism in release and plateout assumptions is particularly evident when the halogen release fractions estimated in the FSV FSAR (Appendix D) are compared to TID-14844 values. The FSAR estimates a core halogen available leakage fraction of 5.5% which is a factor of 4.5 below the TID-14844 effective core release fraction of 25%. The FSAR fuel particle coating failure temperature of 1725 C is also conservative compared to test data cimulacing a core heatup accident. Figure 2-1 3 illustrates the margin from the test data to the 1725 C failure temperature used in this evaluation. Also shown is a NRC model (NUREG-0111) for TRISO UC2 particle which is similar to the FSV FSAR model. 3. Meteorology 3.1 Site Soecific Meteorology l The Fort St. Vrain site meteorology was described in detail in the FSV FSAR Section 2.8. The most likely condition on an annual basis (Table XVII [ of FSAR) are winds from the N sector of the wind rose which occur 17% of the time. The most likely stability category is D (Table XIIV of FSAR), and the I average wind speed at 200 ft. elevation averaged over all directions is 6.51 mph (2.9 m/s) as given in Table II of the FSAR. Figure 17 in Section 2.8.3 of the i l FSAR gives the probability that the dilution factor will exceed a given value. t l Using the dashed curve shown in Fig. 17 (FSAR Sec. 2.8.3) which includes the effect of using corrected anemometer wind velocity indicates that a dilution factor -3 l of about 3.5x10 s/m3 at 590m will be exceeded only 5% of the time (the 5% cut-off). This corrected cut-off cerresponds to stability category F with a wind speed of about 2.4 m/s. Thus the dilution factor for stability category F { with 1 m/s vind speed will be exceeded considerably less than 5% of the time and represents a " worst case" type condition, i I s

3.2 NRC Staff Meteorology from SER The Fort St. Vrain Safety Evaluation Report (SER) (Ref. 4) prepared by the then AEC staff includes the following meteorology and breathing rate assumptions for accident condition calculations at the LPZ (16,000m) 3 Time Period Dilution F2ctor (s/m ) Breathing Rate ( ) -6 ~ Most Unfavorable 8 hrs. 7.8x10 3.47x10 ' Next Most Unfavorable -6 -4 16 hours 3.9x10 1.75x10 ~0 1 to 4 days 2.4x10 2.32x10 ~ 4 to 30 days 7.8x10 2.32x10 ~7 ~4 30 to 180 days 4.6x10 2.32x10 Plotting these dilution factor values against the time period over which they represent an average value (Fig. 3-1), it appears that there is approximately at time dependence, as would be expected, when allowing for wind meander and change of direction. This can be also be seen from the equations for averaging the dilution factor for wind meander over a 22.5 sector ($ = for ground level release) and for.an annual average, (Eqn. 6, Sec. 2.8.3 of FSAR) both of which depend inversely on o, ( the vertical diffusion parameter). Since at long times the diffusion process is dominated by random walk effects, c2 is expected to be proportional to t, and so t would be proportional to t 3.3 Extention of NRC Staff Meteoroloev to Other Distances To derive the time dependence of the dilution factor at other distances, the time dependence of the NRC staff meteorology was estimated from a curve drawn through the 16,000m points from the SER as shown in Fig. 3-1. This gave a ~' t dependence. The reference point used for the other distances was the 0.5 8 hr. value, which was calculated from 4(8hr) = !! o o G with G = 1 m/s and ya o o, for all distances taken for stability category F. This is equivalent to 4MyMM "MSW a e

• 9 6FW-M H

An O4-V9e

1 l-l taking the short term dilution factor for ground level release, ~ Ha O r as the 2 hour value, and then extrapolating to 8 hours by the factor \\g = 0.5. Table 3-1 gives the values at other times obtained from $(t) = $(8) where t is in hours. The breathing rates used were taken from the SER and are given above in 3.2. 3.4 Conservative Meteorology Assumptions Listed below are the conservatisms used in obtaining the dilution factor in Table 3-1. a) Worst case type meteorological conditions (Category F, n = 1 m/s) used are likely to occur much less than 5% of the time, b) Downwash conditions are assumed to persist over the entire time, c) The effective stack height is taken to be zero (i.e., ground level release). d) Bouyancy effects of He relative to air are neglected. e) Dilution by the building wake is neglected. 4. Dose Computation Methodology Plume exposure doses were calculated using the computer program TDAC (Ref. 1). TDAC is a computer program which utilizes analytical methods to calculate the time-dependent radiological impact due to routine or accidental release of radionuclides from nuclear power plants or other potential sources of radioactivity release. The basic model used by TDAC is comprised of eight volu=es which are in-terrelated by a network of flow paths. The program calculates the amount of activity in each volume at specified times and determines radiological doses are specified distances based on activity entering volume 8 and specified meteorology. The calculated doses are external whole body gamma, external whole body beta, and the internal inhalation commitment doses to the skeletal system, thyroid, lung, 4 l gastro-intestinal tract and total body. The required input data consists of I

three types: (1) flow model data such as flow rates and filter efficiencies, (2) nuclear parameters such as decay constants, branching ratios, and disintegration energies, and (3) radiological dose input such as windspeed, atmospheric dilution factors, distances, and breathing rates. The calculation of activity in each volume is based on the assumption of instantaneous homogeneous mixing. The calculation of radiological doses is based on the semi-infinite cloud approximation. All calculations are based on the analytical solution of the coupled linear differential equations governing the activity in the different volumes. The TDAC code, with appropriate parameter assignments was also used to evaluate the plume exposure doses for the Reference LWR of this study. i 1 f For this study, it was observed that the dose categories of whole body gamma external plume exposure and thyroid inhalation and ingestion dose _ control and for this reason other organ dose categorica are omitted from this presentation. I 5. Parameter Assienments and Assumotions Table 5-1 provides a listing and su==ary of the parameter values and analysis assumptions of this study. The characteristics of the FSV siting event, the DBA #1, and the reference LWR event, DBA/LOCA are described. These events i are felt to bound the envelope of emergency preparedness planning. 1 I i l We a

6. Applicable EPZ Distances for Fort St. Vrain 6.1 Evaluation of Evacuation EPZ for FSV The EPZ (Emergency Planning Zone) determination for Fort St. Vrain is based on the estimated accident doses in the environment and their comparison to the re-commended Protective-Action-Guides-(PAGs) contained in Ref. 3, Table 5.2. These PACS for the general population consist of dose ranges in the event of an emergency condition, along with recommended actions to be taken by the responsible local authorities. Figures 6-1 and 6-2 show the comparison between FSV and a reference 1000 MW(e) LWR of the cumulative dose vs. distance. Also shown on Figure 6-1 are the 1 and 5 Rem whole body gamma PAG dose levels. Below 1 Rem, no protective action is required according to the PAG, while the state may issue an advisory, and environmental radiation levels are monitored. For doses of 1 to less than 5 Rem, the general population should seek shelter and await further instructions, evacuation of particularly susceptible elements of the general population should be considered, access should be controlled, and environmental radiation levels i monitored. For 5 Rem or higher whole body gamma doses, mandatory evacuation of i the population in predetermined areas is recommended. Figure 6-2 is similar to Figure 6-1 except that it is for inhalation thyroid dose and the PAG levels are 5 and 25 Rem. Clearly, thyroid dose is the controlling factor since the distance to which protective action is required is larger. From Fig. 6-2, the O to 30 day inhalation thyroid dose for FSV of 25 Rem occurs at about 4100m (2.5 mi.). I j The above data is also summarized in Table 6-1. i l For a 10 mi. LWR evacuation radius (16,000m), the dose for the reference LWR is about 11 Rem inhalation thyroid. The distance for FSV for the same dose as the reference LWR at 10 mi. is about 7500m (4.7 mi.). Therefore an evacuation radius of about 5 mi. for FSV would afford a comparable degree of public protection at _. ~.

the 10 ti. LWR dictraco. There is also a considerable difference in the time of the activity released from FSV compared to the reference LWR as shown in Fig. 6-3 for the inhalation i thyroid. This figure shows that for the LWR high thyroid doses (100 rem) have already been received at 1 mi. in the first hour, while 12 hours should be avail-able even at 590m for FSV, before the lower PAG (5 rem thyroid) is reached al-lowing an orderly and organized evacuation to be done if necessary. I i 6.2 Ingestion Pathway EPZ Evaluation In addition to direct plume exposure, the population surrounding the nuclear I power plant can be subjected to pathway doses from contamination of the sur-rounding crops, milk, and food stuffs subsequent to the atmospheric release of radioactivity in the accident. The protective action guides for exposure from food stuffs or water (Ref. 3) define the guidance to be exercised by the utility in its assessment and control of the post accident ingestion pathways. The most severely restrictive ingestion pathway for FSV and the reference 1 f LWR accident examined was found to be the iodine-grass-cow-milk-infant thyroid I pathway. The predicted thyroid inhalation dose vs. distance for both the reference LWR and FSV is displayed in Fig. 6-2. The iodine-grass-cow-milk-infant pathway i may be related to I-131 plume exposure dose by using an approxi= ate factor of 300 (Ref. 2) and recognizing the fact that the thyroid dose reported in Fig. 6-2 I is nearly all due to I-131 exposure.* I I l The plume thyroid inhalation dose, which is proportional to ingestion pathway dose, for the reference LWR is 0.85 Rem.**The distance at which this dose level is reached for FSV is about 30 miles and is consistent with FSAR Chpater 2 discussions of demographic studies at distances up to 30 miles from the site. I

• I-131, at 50 miles, contributes 89% of the reference Lin plume inhalation dose commitment and correspondingly 97% for FSV.
  • Ref. 2, Table I-2 extended to 50 mile radius.

t 7. Conclusions lt is concluded from this EPZ study that a significantly smaller evacuation i radius can be justified for the Fort St. Vrain plant than for a reference 1000 MW(e) l' A further conclusion from this study is that the unique characteristics of the LWR. FSV plant provide a slow gradual heatup of the core allowing considerable time (on the order of tens of hours) for orderly notification, monitoring and eva-l cuation to proceed. Also, a smaller radius for the controlling ingestion pathway, the grass-cow-milk-infant thyroid, can be derived which offers the public the same protection as a similarly sited LWR. In both cases, thyroid dose is the limiting constraint. These conclusions are based on siting event analyses using generally conservative core release fractions and meteorology. Table 7-1 sum-marizes the applicable emergency planning zones for the plume exposure pathway and the ingestion exposure pathway for Fort St. Vrain providing equivalent pro-tection to the general public. g l l References l 1. "TDAC - An Analytical Computer Program to Calculate the Time Dependent j Radiological Effects of Radionuclide Release", D. W. Buckley, i General Atomic Co. Report GA-D13476, 1976. l 2. NUREG-0396, " Planning Basis for the Development of State and Local Covernment Radiological Emergency Response Plans in Support of Light Water Nuclear Plants". 3. " Manual of Protective Action Guides and Protective Actions for Nuclear j Incidents", Environmental Protection Agency Report, EPA-520/1 l 001, 1975. 4. FSV Safety Evaluation Report, Division of Reactor Licensing, USAEC, Jan. 20, 1972. I t my== wuo

• TABLE 3-1 ATHOSPHERIC DILUTION FACTORS
• FOR FSV DBA 11 AND REFERENCE LWR DBA/LOCA BASED ON NRC STAFF VALUES USED IN THE FSV SER DISTANCE (H)

TiltE (HR) 590 1600 4800 8000 16,000 24,140 32,200 48,300 64,400 80,500 8 7.2-4 1.43-4 2.76-5 1.65-5 7.89-6 5 09-6 3 77-6 2.5-6 1 9-6 1.5-6 16 5 1-4 1.02-4 1.96-5 1.17-5 5.61-6 3.62-6 2.68-6 1.8-6 1 3-6 I.0-6 24 4.1-4 8.35-5 1.61-5 9.60-6 4.59-6 2.96-6 2.19-6 1.4-6 1.1-6 8.6-7 f 34 3 5-4 7 03-5 1.35-5 8.09-6 3.87-6 2.50-6 1.85-6 1.2-6 9 1-7 7 2-7 40 3.2-4 6.49-5 1.25-5 7.46-6 3 57-6 2 30-6 1 70-6 1.1-6 8.4-7 6.7-7 52 2.8-4 5.7-5 1.1-5 6.56-6 3 14-6 2.02-6 1.50-6 9.8-7 7.4-7 5.8-7 58 2.65-4 5.4-5 1.04-5 6.21-6 2 97-6 1 92-6 1.42-6 9.3-7 7 0-7 5.5-7 l 100 2.05-4 4.13-5 7 95-6 4.75-6 2.27-6 1.47-6 1.08-6 7 1-7 5.4-7 4.2-7 400 1.03-4 2.09-5 4.01-6 2.4-6 1.15-6 7 4-7 5.47-7 3 6-7 2 7-7 2.1-7 720 7 7-5

1. 5 ',- 5 3.0-6 1.8-6 8.59-7 5 54-7 4.1-7 2.7-7 2.0-7 1.6-7 f

1500 5.3-5 1.09-5 2.09-6 1.25-6 5 98-7 3.86-7 2.85-7 1 9-7 1.4-7 1.1-7 4320 3 2-5 6.45-6 1.24-6 7 42-7 3 55-7 2.29-7 1.69-7 1.1-7 8.4-8 6.6-8 ? ~b 3

• Values in units of S/m are listed in exponential form such that, e.g., 7 2-4 means 7 2 x 10 i

i

Table 5-1 r EPZ Study - Assumptions and Parameter Selections i Parameter Description j FSV i l Source Term f Power level, MW(t) design 879 MW(t) i Accident scenario assumed DBA #1 (FSV FSAR Appendix D, page D.1-56) t Release of fission products fram Time dependent release

• over a 475 l

the FSAR fuel particle hour period Release fraction from fuel to TID-14844 fractions employed primary coolant system Noble gases 100% Iodines 50% Others 1% Plateout reduction factor. .._ _. TID-J4844 reduction _of_50% applied to (Iodine only) Iodine transported from the primary coolant system to the reactor con-tainment building. j PCRV leak rate 0.2% volume / day l (total duration of accident) Reactor building vencilacion system characteristics Purge rate 1.5 volumes / hour (en for total duration of the accident) Filter efficiencies Noble gases 0% Halogens 90% Particulates 95% Additional plateout and gravita-Ignored tional settling of condensible nuclides in the confinement building l j

• Time-dependent source term identified by designation FSVTID*FUELREL. and discussed in engineering file C-70-001, F. S. Dombek.

i 1 I i t

Table 5-1(Cont.) Parameter Description Reference LWR

• Source term Power level, MW(t) 2958.

Accident scenario assumed DBA/LOCA Release of fission products from Near instantaneous release of the core matrix to the reactor TID-14844 core fractions building i Core inventory, available for TID-14844 assumptions (including i release from containment iodine reduction due to plateout) Noble gas 100% Iodine 25% l Available for leakage from the reactor containment immediately Isotopes after LOCA (C1) I-131 1.9+07 I-132 2.7+07 1-133 4.2+07 I-134 4.8+07 ? I-135 3.8+07 I Kr-83M 1.3+07 Kr-85M 3.2+07 i Kr-85

7. 3+05 l

Kr-87 6.3+07 l Kr-88 8.8+07 i Kr-89 1.1+08 Xe-131M 6.5+04 Xe-133M 4.0+06 Xe-133 1.7+08 Xe-135M 4.5+07 Xe-135 2.8+07 Xe-137 1.5+08 Xe-138

1. 5+08 I

I

• While not specifically identified as the Sun Desert Nuclear Power Plant, the reference LWR of this study has many characteristics str.ilar to this current PWR design.

1

i.. Table 5-1(Cont.) Parameter Description Iodine composition Elemental 91% Particulate 5% Organic 4% Containment building leak rate 0-24 hours 0.2%/ day >24 hour 0.1%/ day Containment spray characteristics Initiation time 150 see Sprayed region 70% Unsprayed region 30% Mixing rate sprayed to unsprayed 2 hr-1 region Elemental Iodine removal rate 10 hr-1 Particulate Iodine removal rate 0.45 hr-1 Organic Iodine removal rate 0 Iodine DF (elemental) 100 Resultant single volume LWR 3.0 hr-1 spray removal constant used this study Post LOCA containment purge e characteristics (to reduce combustiele gas concentration) Purge interval 336 to 720 hours Purge rate 50 SCFM Purge filter efficiency Elemental & Methyl Iodine 70% Radionuclide leakage from ESF Ignored area i I

,, Table 5-1(Cont.) Parameter Description Specific Site and Dose Parameters (Used for both FSV & Reference LWR dose estimates) Atmospheric dilution factors See Table 3-land Figure 3-1 employed Breathing rates Reference LWR 3 0-8 hour 3.47 x 10-4 m /sec 3 8-24 hour 1.75 x 10-4 m /sec 3 24-720 hour 2.32 x 10-4 m /sec l FSV* l 0-40 hour 2.32 x 10-4 3 m /sec 40-52 hour 3.47 x 10-4 3 m /see 3 t - -- 52-58-hour 1.751c-10-4-m /sec f 3 58-720 hour 2.32 x 10-" m /sec Average wind speed 1 m/sec Dose code utilized TDAC - GA-D13476 1 i Decay & buildup estimate Considered Fallout deposition enroute to Considered for Iodine only (see dose receptor Reg. Guide 1.111 Figure 6) Conversion factor for infant-300 (Ref. NUREG-0396 Append'x I, milk thyroid Iodine ingestion page I-19) pathway (ratio of thyroid dose i commitment factor for milk pathway to the inhalation plume exposure pathway) l l

• Note: FSV breathing rates and atmospheric dilution factors have been permuted to maximize the off-site dose, i.e.,

the times of highest { breathing rate and poorest atmospheric dilution correspond to the peak i t in time dependent release of fission products from the fuel to the primary coolant. t l i

Table 6-1 Summary of Derived EPZ Evacuation Distances for FSV and a Reference 1000 MW(e) LWR Whole Body Gamma - Direct Plume Exposure 9 No Protective Action (PAG <1 rem whole body) EPZ Distance FSV LWR 0-8 Hour < EAB( 2.2 Mi. 0-24 Hour < EAB 2.5 Mi. 0-30 Day 0.6 Mi. 3.0 Mi. Evacuation Required (PAG _> 5 rem whole body) EPZ Distance l FSV LWR 0-8 Hour < EAB 0.9 Mi. 0-24 Hour < EAB 1.0 Mi. 0-30 Day < EAB 1.2 Mi. Thyroid Inhalation - Direct Pluma Exposure No Protective Action (PAG <5 rem Thyroid) EPZ Distance l FSV . LWR. _ _ _ l 0-8 Hour < EAB 7.5 Mi. 0-24 Hour 1.2 M1. 9.0 Mi. l l 0-30 Day 8.1 Mi. 16.0 Mi. t a Evacuation Required (PAG ?. 25 rem Thyroid) j EPZ Distance FSV LWR 0.-8 Hour <EAB 3.1 Mi. 0-24 Hour 0.5 Mi. 3.5 Mi. 0-30 Day 2.5 Mi. 5.3 Mi. (1) FSV EAB (Exclusion Area Boundary Distance) 590 meters. l I t ___

Table 7-1 FSV EPZ Distance Compared to Generic LWR EPZ Distance of NUREG-0396( } Accident Phase Generic LWR Emergency Comparable FSV Emer-Planning Zone Distance gency Plangpng Zone NUREG-0396 Distance i I3) Plume Exposure Pathway About 10 miles radius About 5 miles Whole body (external) Thyroid (inhalation) Other organs (inhalation) Ingestion Pathway About 50 mile radius About 30 miles ( ) Thyroid, whole body, bone marrow (ingestion) (1) NUREG-0396, EPA 520/1-78-016, " Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Power Plants", Joint Report USEPA and USNRC. Collins, H. E. et. al., Dec. 1978. (2) FSV EPZ distances were obtained by (1) assuming that the Reference LWR of this study meets the intent of NUREG-0396 for design basis accidents, (2) estimating the Reference LWR dose at 10 miles (plume) and 50 miles (ingestion) for the DBA/LOCA of this study and then, (3) determining the corresponding equi-dose distance for the FSV-DBA #1 event. (3) Thyroid inhalation controls with about a 5 mile EPZ. EPZ distance for whole body is approximately 2 miles and other organ doses are negligible. (4) The grass-cow-milk-infant thyroid exposure pathway doninates for the FSV environmentally released radionuclide source term.

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