ML20140B251
| ML20140B251 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway  |
| Issue date: | 09/09/1981 |
| From: | Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| SLNRC-81-94, NUDOCS 8109140123 | |
| Download: ML20140B251 (22) | |
Text
.
e 9
SNUPPS Standardized Nuclear Unit Power Plant System 5 Choke Cherry Road Nicholas A. Petrick Rockville. Maryland 20850 Executive Director (301)869 8010 September 9, 1981 SLNRC 81-94 FILE: 0541 SUBJ: AEB Review Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Docket Nos.: STN 50-482, STN 50-483, and STN 50-486
Dear Mr. Denton:
In discussions with Dr. Gordon Edison, NRC Project Manager for the SNUPPS applications, it was learned that the Accident Evaluation Branch required additional information in order to complete their review of the SNUPPS FSAR. The enclosure to this letter provides the requested information and will be included in the next revision to the SNUPPS FSAR.
Ver truly yours,
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Nicholas A. Petrick
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D. T. McPhee, KCPL "55 *"
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ilEvlstD CHAPT 2 i SNUPPS j
P 2.3 METEOROLOGY 2.3.4 SHORT-TERM (ACCIDENT) DIFFUSION ESTIMATES 2.3.4.1 Objective l
l The objective of this section is to provide an envelope of the short-term atmospheric dispersion factors (x/Qs) for the two t
sites for the postulated accident analyses presented in Chap-(
ter 15.0.
l 2.3.4.2 Calculations 2.3.4.2.1 Site Boundary and LPZ l
The short-term atmospheric dispersion factors (x/Qs) are based on onsite meteorological data for the two sites.
The diffusion l
equations and assumptions used in the calculations were those I
outlined in draft NRC Regulatory Guide 1.145, " Atmospheric l
l Dispersion Models for Potential Accident Assessment at Nuclear 1
Power Plants."
Table 2.3-1 lists the limiting X/Qs for the l
Wolf Creek and callaway sites.
The detailed procedures used in the calculations are given in Section 2.3.4.2 of each Site I
i Addendum.
2.3.4.2.2 Control Room Intake l
The basic model employed for the distribution of relative concentrations (x/Qs) within a building wake at the SNUPPS l
control room intakes following an accident is given by Reference 1 to be:
(1) c
(
X/Q =
gy 2
reference cross-sectional building area, m Where A
=
reference wind speed, m/sec V
=
nondimensional concentration coefficient K
=
c K
is a function of nondimensional space coordinates x/L, y/L, aRd z/L, building configuration, wind direction, and source configuration.
The K field for a given building configura-tion, source configurEtion, and wind direction is considered I
to be invariant.
Accordingly, K values determined by wind tunnel tests with a model struct6re are expected to be the same as those that would be obtained with a geometrically l
similar building in the full-scale atmosphere in the same 'ind w
direction, with a similar leak.
The SNUPPS contiguous building arrangement is shown in Figure 2.3-1.
The K data used in the c
Rev. 7 9/81 2.3-1 l
l
Hl483 SNUPPS analysis for low level release are presented in Figure 2.3-2 and were derived from two sets of tests.
One used rectangular prisms (Ref. 2 ),
the other used a model of the EBR-II complex (Ref. 1).
Both tests were described and portions of the data presented in Reference 3.
The K data for the unit vent release from the top of the contEinment were extracted from Figure 10 of Reference 1 and are presented in Table 2.3-2.
The value of A used in conjunction with K in Figure 2.3.2 and Table 2.3-2 is the SNUPPS equivalent of t8e EBR-II area, A=
1.12 D2 = 2280 m2 with the diameter of the reactor D = 45.1 m.
The value of V used in conjunction with Figure 2.3-2 is the mean velocity of the approach flow at an elevation correspon-ding to the anemometer elevation of the EBR-II model tests.
Reference 3 reports this elevation to be 62 feet or 0.77D above the top of the dome.
The SNUPPS equivalent height becomes 63.4 + 0.77 x 45.1 = 98.lm above ground.
The V values were obtained by extrapolating wind speeds at anemometer elevations equivalent to 98.1 meters by the power law.
ui (98.1/zi)"
(2)
V
=
Where mean speed at elevation zi, m/sec ui
=
anemometer elevation at a given site, m zi
=
atmospheric stability exponent n
=
Values of n were arbitrarily assumed for the various stability classes as follows:
Pasquill Stability Class A
B C
D E
F G
n 0.20 0.25 0.29 0.33 0.40 0.50 0.60 A cumulative frequency distribution was constructed for the X/Q values calculated by equations 1 and 2 above, using 3 years combined onsite meteorological data.
The corresponding highest i
5 percent, 10 percent, 20 percent, and 40 percent X/Q values I
for each site are given in Table 2.3-3.
2.
3.5 REFERENCES
1.
- Halitsky, J.,
- Golden, J.,
- Halpern, P.,
(1963):
" Wind Tunnel Tests of Gas Diffusion From a Leak in the Shell of a Nuclear Power Reactor and from a Nearby Stack," N. Y.
University Department of Met. & Ocean, GSL Rep. 63-2 under USWB Contract Cwb-10321 2.
- Halitsky, J. (1963):
" Gas Diffusion Near Buildings,"
ASHRAE Trans. 69: pp. 464-484 l
l Rev. 7 2.3-2 9/81 l
.. _ _ =._._ -._. - _
db SNUPPS i
1 3.
Slade, D.
H., ed. (1968):
" Meteorology and Atomic Energy,"
U. S. AEC Division of Technical Information TID-24190 1
l t
1 i
I
\\
)
I i
e i.
l I
l Rev. 7 l
2.3-3 t
9/81 f
SNUPPS TABLE 2.3-1 3
LIMITING ATMOSPHERIC DISPERSION FACTOR, X/Q (sec/m )
Site Boundary Callaway Wolf Creek 0-2 hr.
2.OE-4 1.5E-4 Low Population Zone l
0-8 hr.
2.6E-5 1.9E-5 8-24 hr.
1.7E-5 1.3E-5 24-96 hr.
7.2E-6 5.3E-6 96-720 hr.
2.0E-6 1.5E-6 i
l Rev. 7 9/81
a
.,u mm, SNUPPS TABLE 2.3-2 VARIATION OF INTAKE K WITH WIND DIRECTION c
UNIT VENT RELEASE Wind Direction Wolf
- Creek Callaway N
O 1.5 NNE O
0.5 NE O
O i
ENE 0.5 0
E 1.5 0
ESE 2.5 0
SE 1.5 0
l.
SSE 0.5 0
S 0
0 l
SSW 0
0 SW 0
0 WSW 0
0 W
0 0
WNW 0
0.5 NW 0
1.5 NNW 0
2.5 t
l I
Rev. 7 9/81
I Ak6 SNUPPS TABLE 2.3-3 RELATIVE CONCENTRATION (X/Q) AT CONTROL BUILDING AIR INTAKE From Low Level Release SITE Percentage Wolf Creek Callaway 5
5.33 7.18 10 3.62 5.28 20 0.66 1.66 40 0
0 For Unit Vent Release Percentage Wolf Creek Callaway 5
1.14 1.33 10 0.68 0.90 20 0.17 0.41 40 0
0 3
- Units for X/Qs are 10 4 m/sec l
l Rev. 7 9/81
P'" PS
/4 cA iM 9 i
MATHEMATICAL MODELS USED IN THE ANALYSIS - Mathematical models used in the analysis are described in the following sections:
a.
The mathematical models used to analyze the activity creleased during the course of the accident are described
-in Section 15A.2.
b.
The atmospheric dispersion factors used in-the analysis were calculated, based on the onsite meteorological measurements program described in Section 2.3 of each Site Addendum, and are provided in Table 15A-2.
c.
The thyroid inhalation total-body immersion doses to a receptor exposed at the exclusion area boundary and the outer boundary of the low population zone were analyzed, using the models described in Sections l
15A.2.4 and 15A.2.5, respectively.
d.
Buildup of activity in the control room and the integrated doses to the control room personnel are l
analyzed, based on models described in Section 15A.3.
IDENTIFICATION OF LEAYAGE PATHWAYS AND RESULTANT LEAKAGE ACTIVITY - For evaluating the radiological consequences of a postulated LOCA, the resultant activity released to the con-tainment atmosphere is assumed to leak directly to the en-vironment.
l No credit is taken for ground deposition or radioactive decay during transit to the exclusion area boundary or LPZ outer l
boundary.
15.6.5.4.1.2 Radioactive Releases Due to Leakage from ECCS and Containment Spray Recirculation Lines Subsequent to the injection phase of ESF system operation, the water in the containment recirculation sumps is recirculated by the residual heat removal, centrifugal charging and safety
(
injection pumps, and the containment spray pumps.
Due to the operation of the ECCS and the containment spray system, most of the radiciodine released from the core would be contained in the containment sump.
It is conservatively assumed that a leakage rate of 2 gpm from the ECCS and containment spray l
recirculation lines exists for the duration of the LOCA.
This leakage would occur inside the containment as well as inside the auxiliary building.
For this analysis, all the leakage is l
assumed to, occur inside the auxiliary building.
Only trace quantities of radioiodine are expected to be airborne within the auxiliary building due to the temperature and pH level of the recirculated water.
However, 10 percent of the radioiodine in the leaked water is assumed to become airborne and exhausted from the unit vent to the environment through safety grade filters.
15.6-27
1 S'."JPP S LC:JOW Radiological Consequences of ECCS/CS Recirculation Line Leakage -
i The assumptions used to calculate the amount of radioiodine released to the enviroranent are given in Table 15.6-6.
The dose models are presented in Section 15.A.
The offsite doses at the site boundary and LPZ and the doses to control room personnel from this pathway are given in Table 15.6-8.
15.6-27a
T':??3 LD;'! [Cf I
15.6.5.4.2 Identification of Uncertainties and Conservatisms in the Analysis The uncertainties and conservatisms in the assumpticas used to evaluate the radiological consequences of a LOCA result princi-pally from assumptions made involving the amount of the gaseous fission products available for release to the environment and the meteorology present at the site during the course of the accident.
The most significant of these assumptions are:
a.
The ECCS is designed to prevent fuel cladding damtge that would allow the release of the fission products contained in the fuel to the reactor coolant.
Severe degradation of the ECCS (i.e., to the unlikely extent of simultaneous failure of redundant components) would be necessary in order for the release of fission products to occur of the magnitude assumed in the analysis.
b.
The release of fission products to the containment is assumed to occur instantaneously.
c.
I.t is assumed that 50 percent of the iodines released to the containment atmosphere is adsorbed onto the internal surfaces of the containment or adheres to internal components; however, it is estimated that the removal of airborne iodines by various physical 15.,6_-26
E!rJFPS TABLE 15.6-6 (Sheet 2) 3.
Sprayed volume, percent 85 4.
Unsprayed volume, percent 15 5.
Sprayed-unsprayed mixing rate, CFM 85,000 6.
Containment volume, ft8 2.5E+6 d.
Activity released to containment Isotope Curies I-131 2.24E+7 I-132 3.40E+7 I-133 5.00E+7 I-134 5.85E+7 I-135 4.55E+7 Xe-131m 9.25E+5 Xe-133m 4.93E+6 Xe-133 2.00E+8 Xe-135?n 5.55E+7 Xe-135 1.91E+8 Xe'-137 1.82E+8 Xe-138 1.70E+8 Kr-83m 1.48E+7 Kr-85m 4.62E+7 Kr-85 1.46E+6 Kr-87 8.32E+7 Kr-bd 1.14E+8 Kr-89 1.42E+8 e.
ECCS/ recirculation leakage 1.
Leak rate (0.47 hour5.439815e-4 days <br />0.0131 hours <br />7.771164e-5 weeks <br />1.78835e-5 months <br />s-30 day), gpm 2.0 2.
@ 0.47 hour5.439815e-4 days <br />0.0131 hours <br />7.771164e-5 weeks <br />1.78835e-5 months <br />, curies I-131 4.46E+7 I-132 5.9^S+7 I-133 9.672+7 I-134 8.08E+7 I-135 8.67E+7 3.
Sump volume, gal.
460,000 4.
Fraction iodine airborne 0.1 5.
ESF filter efficiency, i 90.0 l
IV.
Control room parameters Tables 15A-1 and 15A-2
Sh'JF PS fr i
e TABLE 15.6-8 RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT-ACCIDENT Doses (rem) callaway Wolf Creek I.
Exclusion Area Boundary (0-2 hr) a.
Containment leakage (0-2 hr)
Thyroid 87 65 Whole body 2.9 2.2 b.
ECCS recirc. leakage (0.47 hr-2 hr)
Thyroid 27 20 Whole body 0.081 0.061 II.
Low Population Zone Outer Boundary (0-30 day) a.
Containment leakage (0-30 day)
Thyroid 55 41 Whole body 1.0 0.75 b.
ECCS recirc. leakage (0.47 hr-30 day)
Thyroid 59 44 Whole body 0.059 0.044 III.
Control Room (0-30 day) a.
Containment leakage (0-30 day)
Thyroid 16 11 Whole body 0.43 0.31 Beta-skin 7.8 5.5
- b..ECCS recirc. leakage (0.47 hr-30 day)
Thyroid 3.1 2.1 Whole body 8.4E-4 6.4E-5 Beta-skin 7.5E-4 5.7E-4
rm7 FPS APPENDIX 15A ACCIDENT ANALYSIS RADIOLOGICAL CONSEQUENCES EVALUATION MODELS AND PARAMETERS 15A.1
, GENERAL ACCIDENT PARAMETERS This section contains the parameters used in analyzing the radiological consequences of postulated accidents.
Table 15A-1 contains the general parameters used in all the accident analyses.
For parameters specific only to particular acci-dents, refer to that accident parameter section.
The site specific, ground-level release, snort-term dispersion factors (for accidents, ground-level releases are assumed) are based on Regulatory Guide 1.145 (Ref.1) methodology and represent the worst of the 5 percent overall site meteorology and the 0.5 percent worst-sector meteorology and these are given in Table 15A-2 (see Section 2.3.4 and each site addendum for additional details on meteorology).
The core and gap inven-tories are given in Table 15A-3.
The thyroid (via inhalation pathway), beta skin, and total-body (via submersion pathway) dose factors based on References 2 and 3 are given in Table 15A-4.
15A.2 OFFSITE RADIOLOGICAL CONSEQUENCES CALCULATIONAL MODELS This section presents the models and equations used for calcu-lating the integrated activity released to the environment, the accident flow paths, and the equations for dose calcula-tions.
Two major release models are considered:
(1) a single l
holdup system with no internal cleanup and (2) a holdup system wherein a two-region spray model is used for internal cleanup.
15A.2.1 ACCIDENT RELEASE PATHWAYS The release pathways for the major accidents are given in Figure 15A-1.
The accidents and their pathways are as follows:
LOCA:
Immediately following a postulated loss-of-coolant accident (LOCA), the release of radioactivity from the con-tainment is to the environment with the containment spray and ESF systems in full operation.
The release in this case is calculated using equation (12) which takes into account a two-region spray model within the containment.
The release of l
radioactivity to the environment due to assumed ESF system leakages in the auxiliary building will be via ESF filters and is calcula'ted using equation (5).
l 15A-1
408/ Mg
,c,-.7 3 15A.3 CONTROL ROOM RADIOLOGICAL CONSEQUENCES CALCULATIONAL MODELS Only radi,ation doses to a control room operator due to postu-lated LOCA are presented in this chapter since a study of the radiological consequences in.the control room due to various postulated accidents indicate that the LOCA is the limiting Cast.
15A.3.1 INTEGRATED ACTIVITY IN CONTROL ROOM Radioactivity is brought into the control room via the control room filtration system which draws in air from the control building.
Radioactivity is brought into the control building through safety grade filters via the control room pressurization fan which draws in outside air.
Additional radioactivity is brought into the control building via an assumed inleakage rate.
The activity concentrations at Qe control bui.' ding for each time interval are found by multiplying -the activity release to the environment by the appropriate X/Q for that time interval.
The flow path model is shown below.
F5 E
N
,1NTAFE l FILTER V
FILTER 8
I R
F CONTROL 4
CONTROL O
BUILDING ROOM N
M INLEAKAGE F
F+E 2
3 4
4 E
N T
- 2XHAUST(F + F I 3
2 once activity is brought into the control building, mixing within the control building is afforded by the control room pressurization fan.
Only one-half of the control building volume is considered as the mixing volume.
The control room filtration system fan takes air from the control building and the control room (recirculation) and discharges to the control room through the control room filtration safety grade filters.
The CRVIS starts both trains of the control room pressurization system and the control room filtration system.
For the deter-mination of doses to control room personnel, the worst single failure has been ascertained to be the failure of the filtra-tion fan in one of the two filtration system treins.
Operator action is required to isolate the system train with the failed filtration fan.
Prior to isolation, a potential pathway exists allowing air from the control building to enter the Rev. 7 15A-8 9/81
Luft NS6 P
control room, bypassing the control room filtration filters.
After isolation, one control room pressurization fan and one control room filtration fan operate for the duration of the accident.
Owning to this single failure of the control room filtration fan, the assumed failure of one of the two containment spray (CS) trains, and one.of the two containment cooling sy' stem (CtCS) trains, inherent in the LOCA anal.ysis parameters, given in Table 15.6-6, should not be' applied in this analysis.
With both trains of CS and C,CS operating, more volumetric coverage of the containment spray and more mixing between the new sprayed and unsprayed regions would be expected, thereby giving much greater iodine removal within the containment atmosphere, prior to leakage to the outside environment.
However, the doses to control room personnel have been based on the LOCA analysis parameters given in Table 15.6-1.
The activity in the control building and control room is calculated by solving the following coupled set of first order differential equations.
dACB(t)
CB +)
I
+ F ] X/Q [R(+)] + A A
I)-
3 A
[(1-7) F
=
2 4g CR dt dA I}
CR A
I}
lACB(t) - A
[(1-7) A.3f + A3u
=
4 CR dt where ACB(t) activity in control building at time t,
=
curies ACR(t) activity in control room at time t, curies
=
filter efficiency, fraction
=
3 filtered intake rate, meter /sec F
=
3 s
unfiltered intake (inleakage), meter /sec F2
=
3 atmospheric dispersion factor, sec/ meter X/Q
=
activity release rate in Ci/sec as given in R(t)
=
l Equation 3 of Section 15A.2.2 or Equations 11 and lla of Section 15A.2.3 total removal rate from the control building, A
=
3 sec where
=
Ad+A32 + A3f + A3u 1
isotopic decay constant, sec A
=
d l
MA Days s m ys t
T outleakage to atmosphere from the control A
=
31 building (=(F
+ F )/V with V
- "9 2
CB CB control building mixing volume in meter ),
3 1
sec A'
filtered flow from control building /sec),
into
=
control room (=F I F
in meter 3
CB, 3
sec unfiltered flow from control building into A
=
3u 3
control room (=F /Y F
in meter /sec),
3 CB, 3
sec total removal rate from the control room, A
=
4 sec
=
Ad+A
+A where r
41 recirculation removal rate (=7F /V with A
=
3 CR r
F being recirculation flow rate through 3
3 filter with efficiency 7 in meter /sec and 8
V being control room volume in meter ),
CR 2
sec leakage to control building from the A
=
4g 2
control room (=F3 + F )/VCR), sec Upon solving this coupled set of differential equations, the integrated activity in the control room (IACR) s e erm ned by the expression St IA I)*
ACR(t) dt CR O
l This IACR(t) is used to calculate the doses to the operator in 1
the control room. 'This activity is multiplied by an occupancy factor which accounts for the time fraction the operator is in the control room.
l 1
Rev. 7 15A-10 9/81
/8
[
SINFPS l
i 15A.3.2 CONTROL ROOM THYROID DOSE CALCULATIONAL MODEL l
Control room thyroid doses via inhalation pathway are cal-
]
culated using the following equation:
IIACRij) 3 x0 D
Thi I
TH-CR l
CR i
j 4
e l
t l
l l
t 15A-10a l
l w
--=---g-w--,-w--
=
- =_-,,,
-,__.p.
g)M SNUPPS where DTh-CR =
control room thyroid dose in rem and
'BR breathing rate assumed to be always 3.47 x 10 4
=
~
3 meter /second V
=
CR v lume of the control room in cubic meters l
DCF
=
Thi thyroid dose conversion factor for adult via inhalation in rem /Ci for isotope i IA
.=
CRi3 ntegrated activity in control room in Ci-sec for isotope i during time interval j O.
=
J control room occupancy fraction during time interval j 15A.3.3 CONTROL ROOM BETA-SKIN DOSE CALCULATIONAL MODEL The beta-skin doses to a control room operator are calculated using the following equation:
D S-CR pi IACRij) 3 x0 CR 1
)
where D and DCF E-CR pg are the beta-skin doses in the control room in rem and the bet a-skin dose conversion factor for isotope i in rem-meter /Ci-sec, respectively.
The other 3
symbols are explained in Section 15A.3.4.
15A.3.4 CONTROL ROOM TOTAL-BODY DOSE CALCULATION Due to the finite structure of the control room, the total-body gamma doses to a control room operator will be substan-tially less than wh~at they would be due to immersion in an infinite cloud of gamma emitters.
The finite cloud gamma l.
doses are calculated using Murphy's method (Ref. 4) which l
models the control room as a hemisphere.
The following equa-tion is used:
D'B-CR
- T V
GF) f Yi IACRij ) x 03 CR 1
)
l 15A-11
LS/2 A2fdf SI.TPPS where dose reduction due to control room geometry GF
=
factor 1173/(V )
GF
=
2 volume of the control room in cubic feet l
V
=
3 D
al-body dose in the control room in rem, TB-CR and other quantities have been defined in subsections 15A.2.5 and 15A.3.4.
15A.3.4.1 Model for Radiological Consequences Due to Radio-active Cloud External to the Control Room This dose is calculated based on the semi-infinite cloud model which is modified using the protection factors described in Section 7.5.4 of Reference 5 to account for the control room walls.
15A.4 REFERENCES 1.
USNRC Regulatory Guide 1.145, " Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," August 1979.
2.
USNRC Regulatory Guide 1.109, Rev.
1,
" Calculation of Annual Doses to Man from Routine Releases of Reactor l
Effluents for the Purpost of Evaluating Compliance with i
10 CFR Part 50 Appendix I," October 1977.
3a.
- Kocher, D.C.,
" Nuclear Decay Data for Radionuclides Occurring in Routine Releases from Nuclear Fuel Cycle Facilities," ORNL/NUREG/TM-102, August 1977.
l 3b.
- Berger, M.J.,
" Beta-Ray Dose in Tissue-Equivalent Material l
Immersed in a Radioactive Cloud," Health Physics, Vol. 26, l
pp. 1-12, January 1974.
4.
- Murphy, K.G.
and Campe, K.M., " Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19," Paper presented at the 13th AEC Air Cleaning Conference.
5.
" Meteorology and Atomic Energy 1968," D. H. Slade (ed.),
USAEd Report, TID 24190, 3968.
l l
SMUPPS TABLE 15A-1 PARAMETERS USED IN ACCIDENT ANALYSIS I.
General 1.
Core power level, Mwt 3565 2.
Full-power operation, days 1000 3.
Number of fuel assemtlies in the core 193 4.
Maximu.m radial peaking factor 1.65 5.
Percentage of failed fuel 1.0 6.
Steam generator tube leak, 1)/hr 500 II.
Sources 1.
Core inventories, ci Table 15A-3 2.
Gap inventories, Ci Table 15A-3 3.
Primary coolant specific activities, pCi/gm Table 11.1-5 4.
Primary coolant activity, technical specifi-cation limit for iodines - I-131 dose equiva-lent, pCi/gm 1.0 5.
Secondary coolant activity technical specifi-cation limit for iodines - I-131 dose equiva-lent, pCi/gm 0.1 III.
Activity Release Parameters 1.
Free volume of containment, ft3 2.5 x 108 2.
Containment leak rate i.
0-24 hours, % per day 0.2 l
ii.
after 24 hrs, % per day 0.1 t
IV.
Control Room Dose Analysis (for LOCA) 1.
Control building i.
Mixing volume, cf 150,000 ii.
Filtered intake, c fc.
Prior to operator action (0-30 minutes) 1000 After operator action (30 minutes -
720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />) 500 iii.
Unfiltered inleakage, cfm 300 iv.
Filter efficiency (all forms of iodine), %
90 2.
Control room i.
Volume, cf 100,000 ii.
Filtered flow from control building, cfm 400 iii.
Unfiltered flow from control building, cfm Prior to operator action (0-30 minutes) 400 After operator action (30 minutes -
720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />) 0
Wd k S*R?PS T.'.3LE 15A-1 (Sheet 2 )
.iv.
Filtered recirculation, cfm 1600 v.
Filter efficiency (all forms of iodine), %-
40 V.
Misce'laneous 3
Table 15A-2 1.
Atmospheric dispetsion factors, x/Q sec/m 2.
Dose conversion factors 3
i.
total body and beta skin, rem-meter /
Ci-sec Table 15A-4 ii.
thyroid, rem /ci Table 15A-4 3
3.
Breathing rates, meter /sec i.
control room at all times 3.47 x 104 ii, offsite 4
0-8 hrs 3 47 x 10 -
8-24 hrs 1.75 x 10 4 24-720 hrs 2.32 x 10 4 4.
Control room occupancy fractions 0-24 hrs 1.0 24-96 hrs 0.6 96-720 hrs 0.4 W
-,-.-4.--
,~.
=..----..-.-___.*r
-r e
t r
- ~ + -
v-
-e
SNUPPS
& g ;y;3,$
TABLE 15A-2 LIMITING SHORT-TERM ATMOSPHERIC DISPERSION FACTORS (x/Qs) FOR ACCIDENT ANALYSIS FOR SNUPPS SITES 8
~
(sec/ meter )
Location Type /
Time Interval Wolf (hrs)
Callaway Creek Site boundary 0-2 2.0E-4 1.5E-4 Low-population zone t
0-8 2.6E-5 1.9E-5 8-24 1.7E-5 1.3E-5 24-96 7.2E-6 5.3E-6 96-720 2.0E-6 1.5E-6 Control room (via containment leakage) 0-8 7.18E-4 5.33E-4 8-24 5.28E-4 3.62E-4 24-96 1.66E-4 6.60E-5 96-720 0
0 Control room (via unit vent exhaust) 0-8 1.3E-4 1.1E-4 8-24 9.0E-5 6.8E-5 24-96 4.1E-5 1.7E-5 96-720 0
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