ML20150E047
| ML20150E047 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Catawba, 05000000 |
| Issue date: | 12/19/1986 |
| From: | Birch M, Hampton J DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20150D105 | List: |
| References | |
| PROC-861219, NUDOCS 8803280071 | |
| Download: ML20150E047 (42) | |
Text
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DECEMBER 29, 1986
Subject:
Offsite Dose Calculetion Manual Revision is The General Office Radwaste Engineering Staff is transmitting to you this date, Revision 13 of the Offaite Dose Calculation Manual. As this revision only affects Catawba N u,c l e a r Station, the approval of other station managers is not required. Please update your copy No. , and discard the affected pages.
REMOVE THESd PAGES INSERT THESE PAGES Remove all pages from the Insert the attached package in CATAWBA section (Appendix C) the CATAWBA section (Appendix C) of the manual and save of the manual and replace the the following figures: blank pages with the "saved "
figures.
FIGURE C1.0-1 FIGURE C1.0-2 FIGURE C5.0-1 (1 of 2)
FIGURE C5.0-1 (2 of 2)
NOTE: As this letter, with it'e attachments, contains "LOEP" information, please insert this letter in front of the September 9, 1986 letter.
Aproval Dater /2 /I'!((S Approval Date / E - / k* bb
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Effective Date: 1/1/87 Effective Detot 1/1/87 A hl Mary L. Birch J. W. Ham ton, Manager System Radwaste Engineer Catawba Nuclear Station 4
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- if you have any questions concerning Revision 13, please call Jim Stewart at (704) 373-5444 49g4 //[ Ihd 7)[
James M. Stewart, Jr.
1 Associate Health Physicist Radwaste Engineering Enclosures 8803280071 870106 PDR ADOCK 05000269 P PDR J
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JUSTlFICATIONS FOR REVISION 13 Section C2.1.2 Updated section for clarity purposes: no change in (Page C-5) Intent.
Section C2.2.2 Updated section to provide an exact location.
(Page C-7)
Section C3.0 Updated section with more exact monitor information.
(Page C-8)
Section C3.1.2 Changed wording for clarity purposes: no change in Page C-9) Intent.
Section C3.1.3 Changed wording for clarity purposes no change in Page C-9) intent.
Section C3.1.4 Changed wording for clarity purposes no change in Page C-9) intent.
Section C3.1.5 Changed wording for clarity purposes: 'no change in Page C-10) Intent.
Section C4.1 Changed the word "shull" to "may" to allav the computer (Page C-12) codes "LADIAP" and/or "GASPAR" to be us'd.
Sectiori C4.2.1 Updated section based on actual plant operating data.
(Page C-12) Deleted reference to Cobalt since fuel failure has occurred.
Section C4.3.1 Updated section based on actual plant operating data.
(Pages C-12 & Deleted reference to Cobalt since fuel failure has C-13) occurred. i Section C4.3.2.1 Updated section based on actual plant operating data.
(Pagu C-14)
Section C4.3.3.2 Updated section based on actual plant operating. data.
(Peges C-14 &
C-15) 1 1
i Section C4.4 Updated section to calculate an actual number per NRC (Pages C-15 thru request.
C-21)
Section C5.0 Corrected typo and changed land use census dates.
(Page C-22)
Tables C5.0-2 Changed sampling location based on latest land use and C5.0-3 census.(see attached letter)
Table C5.0-3 Typo errors changed location of "particulates".
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, DUKE POWER COMPANY C ATAWB A NUCLEAR STATION Po. sox ase CLOVER. S C. actio trLt'uoNr P i isossasitasa December 8, 1986 TO: J. M. Stewart
SUBJECT:
Catawba Nuclear Station Revision to Offsite Dose Calculation Manual File No.: CN-778.02 Please incorporato the following changes into the current revision to the Offsite Dose Calculation Manual:
t
- Add broadictf vegetation sample location #200.
- - Omit bcoadleaf vegetation sample collection #203.
Technical Specf.fications require collection of broadleaf vegetation samples "nearest each of two different offsite locations with the highest predicted annual average ground level D/Q, if milk sampling is not performed".
The two locations with the highest predicted annual average D/Q are NNE and NE of the Station. Broadleaf vegetation is currently collected at the site boundary NE (#201)a ' nd' addition of a collection location at the site boundary NNE (#200) would provide the specified locations. Broadleaf vegetation collection at the site boundary in the SE section (#203) will be deleted with the addition of collection from location #200. Milk sampling ;
is not performed in either the NE or NNE sectors.
b) h. dow! L r.T r t, ...
W. P. Deal, Station Health Physicist l Catawba Nuclear Station JSI/bjp xc: G. T. Mode-G. L. Courtney C. V. Wray R. E. Sorber l
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APPENDIX C CATAUBA NUCLEAR STATION SITE SPECIFIC INFORMATION I l
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APPENDIX C - TABLE OF CONTENTS Page C1.0 CATAWBA NUCLEAR STATION RADWASTE SYSTEMS . . . . . . . . C-1 C2.0 RELEASE RATE CAL _CULATION . . . . . . . . . . . . . . . . C-4 C3.0 RADIATION MONITOR SETPOINTS . . . . . . . . . . . . . . C-8 C4.0 90SE CALCULATIONS . . . . . . . . . . . . . . . . . . . C-11 C5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING . . . . . . . . . C-22 l
_Rev. 13 1/1/87
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C1.0 CATAWBA NUCLEAR STATION RADWASTE SYSTEMS 1
C1.1 LIQUID RADWASTE PROCESSING -
The liquid radwaste system at Catawba Nuclear Station ,(CNS) is used to collect and treat fluxd chemical and radiochemical by products of unit operation. The system produces effluents which can be reused in the plant or discharged in / l small, dilute quantities tc the environment. The means of treatment vary with waste type and desired product in the various systems:
A) Filtration - All waste sources' are filtered 'during processing. In some cases, such as the Floor Drain Tank (FDT) Subsystem of the Liquid Waste '
(WL) System, filtration raay be the only treatment required.
B) Adsorption - Adsorption of halides and organic chemicals by activated charcoal (Carbon Filter) is used primarily in treating waste in the Laundry and Hot Shower Tank (LHST) Subsystem of the WL System. FDT waste may also be treated by this method.
C) Ion Exchange - Ion exchange is used to remove radioactive cations from solution, as in the case of either LHST or FDT waste in the WL System after removal of organics by carbon filtration (adsorption). Icn exchange is also used in removing both cations (cobalt, manganese) and anions (chloride, fluoride) from evaporator distillates in order to purify the distillates for reuse as makeup water. Distillate from the Waste Evaporator in the WL System and the Boron Recycle Evaporator in the Boron Recycle System (iiL) can be treated by th' method, as well as FDT, LHST waste, and letdown. -.
D) Gas Stripping - Removal of gaseous radioactive fission products is accomplished in both the WL Evaporator and the NB Evaporator.
E) Distillation - Production of pure water from the waste by boiling it away from the contaminated solution which originally contained it is accomplished by both evaporators. Proper control of the process will yield water which can be reused for makeup. Polishing of this product can be achieved by ion exchange as pointed out above.
F) Concentration - In both the WL and NB Evaporators, dissolved chemicals.are concentrated in the lower shell as water is boiled away. In the case of the WL Evaporator, the volume of water containing waste chemicals and radioactive cations is reduced so that the waste may be more easily and cheaply solidified and shipped for burial. In the NB Evsporator, the dilute boron is concentrated to 4% so that it may be reused for makeup to the reactor coolant system.
Figure C1.0-1 is a schematic representation of the liquid radwaste system at Catawba.
C-1 Rev. 12 9/19/86
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J Figure C1.0-1 i
<w - + - - - , ,-= - , - . - - - . --e-,,- -- -
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l' Table C1.0-1 1
ABBREVIATIONS S
Systems:
CM'- Condensate System KC - Component Cooling NB - Boron Recycle RL - Low Pressure Service Watsr RN - Nuclear Service Water Systems 1:
WC - Conventional Waste Water Treatment WL --Liquid Waste Recycle WP - Turbine Building Sump WS - Nuclear' Solid Waste Disposal Tanks:
BA - Boric Acid Tank FDT - Floor Drain Tank LHST - Laundry and Hot Shower Tank MST - Mixing and Settling Tank NCDT - Reactor Coolant Drain Tank RHT - Recycle Holdup Tank '
RMT - Recycle Monitor Tank RMVST - Reactor Makeup Water Storage Tank SGDT - Steam Generator Drain Tank
- VUCDT - Ventilation Unit Condensate Drain Tank WDT - Waste Drain Tank WEFT - Waste Evaporator Feed Tank i WMT - Waste Monitor Tank i
l l Table C1.0-1 l l
Rev. 4 l 7/18/84
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C1.2 GASE0US RADWASTE SYSTEMS The gaseous waste disposal system for Catawba is designed with the capability of processing the fission-product gases from contaminated reactor coolant fluids resulting from operation. The system shown schematically in Fig. Cl.0-2 is designed to allow for the retention and subsequent decay of the gaseous fission products generated from the reactor coolant system via the chemical and volume control system and/or the boron recycle system in order to limit the need for intentional discharge of high level radioactive gases from the waste gas holdup tanks. Sources of low-level radioactive gaseous discharge to the environment include periodic purging operations of the containment, the euxiliary building ventilation system, the secondary system air ejector and decayed WG Tanks. With respect to purging operations, the potential contamination is expected to arise from uncollectable reactor coolant leakage.
With respect to the air ejector, the potential source of contamination will be from leakage of the reactor coolant to the secondary system through defects in steam generator tubes. The gaseous waste disposal system includes two waste gas compressors, two catalytic hydrogen recombiners, six gas decay storage tanks for use during normal power generation, and two gas decay storage tanks for use during shutdown and startup operations.
C1.2.1 Gas Collection System The gas collection system combines the waste hydrogen and fission gases from the volume control tanks and that from the boron recycle gas stripper evaporator produced during normal operation with the gas collected during the shutdown degasification (high percentage of nitrogen) and will cycle it through the catalytic recombiners to convert all the hydrogen to water. After the water vapor is removed, the resulting gas stream will be transferred from the recom-biner into the gas decay tanks, where the accumulated activity may be contained in six approximately equal parts. From the decay tanks the gas will flow back to the compressor suction to complete the loop circuit.
C1.2.2 Containment and Auxiliary Building Ventilation Nonrecyclable reactor coolant leakage occurring either inside the containment cr inside the auxiliary building will generate gaseous activity. Gases result-ing from leakage inside the containment will be contained until the containment air is released through the VQ or VP system. The containment atmosphere will l be discharge.d through a charcoal adsorber and a particulate filter prior to i release to the atmosphere. I l
Gases resulting from leakage inside the auxiliary building are released, with-l out further decay, to the atmosphere via the auxiliary building ventilation l system. The ventilation exhaust from potentially contaminated areas in the l auxiliary building is normally unfiltered. However, on a radiation monitor alarm, the exhaust is passed through charcoal adsorbers to reduce releases to the atmotphere.
C1.2.3 Secondary Systems l l
Normally, condensate flow and steam generator blowdown will go parallel l
C-2 Rev. 12 9/19/86 l
through 4 of the 5 condensate polishing demineralizers to remove activity and harmful ions from the water. Noncondensable gases will be taken from the, secondary system by the condenser steam air ejector and are passed through a radiation monitor to the unit vent.
Figure C1.0-2 is a schematic representation of the gaseous radwa:,te system at Catawba, l
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C2.0 RELEASE RATE CALCULATION Generic release rate calculations are presented in Section 1.0; these calcula-tions will be used to calculate release rates for Catawba Nuclear Station.
C2.1 LIQUID RELEASE RATE CALCULATIONS There are two potential release points at Catawba. They are as follows:
- 1. Liquid Waste Effluent Discharge Line (WL)
- 2. Conventional Waste Water Treatment System Effluent Line (WC)
C2.1.1 Liquid Waste Effluent Discharge Line (WL)
There are three low-pressure service water pumps with a minimum flow rate of 16,500 gpm each and four nuclear service water pumps with a minimum flow rate of 9,000 gpm each which provide the required dilution water needed for a release. The LPSW system flow rate monitor has a variable setpoint which term-inates the release by closing the isclation valve (1 WL124) should the dilution flow fall below the setpoint. The following is a typical equation which can be used to calculate a discharge flow, in gpm.
n f5FRL [o I i } [ 0.9 ]
i=1 HPC.I where:
f = the undiluted effluent flow, in gpm.
F RL
= actual low pressure service water flotrate, in gpm.
o = the recirculation factor at equilibrim (dimensionless), 1.027. '
120 cfs = 1.027 g_
y,QR = 1 + 4400 cfs g-H where:
Q = averag dilution flow (120 cfs)
R QH= Ver ge fl W past Wylie Dam (4400 cfs)
C. = the concentration of radionuclide, i, in undiluted effluent as determined by laboratory analyses, in pCi/ml.
MPC. * = the concentration of radionuclide, i, from 10CFR20, Appendix B, Table II, Column 2. If radionuclide, i, is a dissolved noble gas, the MPC. = 2.0E-04 pCi/ml.
1 0.9 = factor used to reduce the WL flowrate (f) to allow the WC system to simultaneously make 10% of the stations releases.
C-4 Rev. 12 9/19/86
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C2.1.2 Conventional Waste Water Treatment System Effluent Line (WC)
The conventional waste water treatment system effluent is potentially radio- l active; that is, it is possible the effluent will contain measurable activity '
above background. It is assumed that no activity is present in the effluent until indicated by radiation monitoring measurements and by periodic analyses of the composite sample collected on that line. The water sources listed below that are normally discharged via the conventional waste water treatment system and/or the Turbine Building Sump will be diverted if they will cause the WC dischange to exceed administrative limits designed to ensure that station release limits will not be exceeded,
- a. Containment Ventilation Unit Condensate Effluent Line The containment ventilation unit condensate effluent line could potentially discharge into the Turbine Building sump, but if activity is detected above its monitor's setpoint, the discharge will be terminated and an alarm actuated. The containment ventilation unit condensate tank will then be pumped to the RHT or WMT, recirculated, sampled, processed thru the WL system if necessary, and then discharged through the liquid waste effluent line and monitored.
- b. Auxiliary Feedwater Sump Pumps and Floor Drain Sump Pump Line Normally the discharge line coming from these sumps will discharge into the Turbine Building sump, but if activity is detected above its monitor's setpoint, the discharge flow will automatically be routed to the floor drain tank for processing and later be discharged through the liquid waste effluent line. Subsequent radioactive releases may be allowed to discharge into the TBS if administratively controlled to assure that release limits are not exceeded,
- c. Steam Generator Blowdown Line Normally the discharge from the Steam Generator Blowdown will be pumped to the Turbine Building Sump, but if activity is detected above its monitor's setpoint, each blowdown flow control valve, the atmospheric vent, and the valve to the Turbine Building Sump will close, thus terminating the discharge. Blowdown can only be continued by venting the i steam to "D" heater and pumping the liquid to the condensate system. !
- d. Turbine Building Sump Discharge Line ,
l Normally the discharge from the Turbine Building sump will go into the conventional waste water treatment system, but if activity is detected above its monitor's setpoint, the sump pumps A, B, and C will stop and an alarm actuated. The Turbine Building sump discharge line can then either be routed to the floor drain tank for processing, routed directly to the liquid waste effluent discharge lire, or allowed to continue being dis-charged via the circuit with proper administrative controls implemented to assure that release limits are not exceeded.
C-5 Rev. 13 )
1/1/87
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i C2.2 GASEOUS RELEASE RATE CALCULATIONS The unit vent is the release point for waste gas decay tanks, containment air-releases, the condenser air ejector, and auxiliary building ventilation. The condenser air ejector effluent is normally considered nonradioactive; that is, it is unlikely the effluent will-contain measurable activity above background. l It is assumed that no activity is present in the effluent until indicated by radiation monitoring measurements and/or by analyses of periodic samples collected on that line. Radiation monitoring alarm / trip setpoints in con- ,
junction with administrative controls assure that release' limits are not 1 exceeded; see section C.3.0 on radiation monitoring setpoints.
The following calculations, when solved for flowrate, are the release rates for noble gases and for radioiodines, particulates and other radionuclides with half-lives greater than 8 days; the most conservative of telease rates calculated in C2.2.1 and C2.2.2 shall control the release rate for a single release point.
C2.2.1 Noble Gases l
{Kg [(X/Q)D] g
< 500 mrem /yr, and 1
{(L1 + 1.1 Mf) [(X/Q)Qf ] < 3000 mrem /yr 1
where the terms are defined below, i 1
C2.2.2 Radiciodines, Particulates, and Other Radionuclides With T 1/2 ) 8 Days l
{Pg[WD] g
< 1500 mrem /yr l 1
where:
K g
= The total body dose factor due to gamma emissions.for each identified noble gas radionuclide, in mrem /yr per pCi/m3from Table 1.2-1.
L.
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= The skin dose factor due to beta emissions for each identified noble gas radionuclide, in mrem /yr per pCi/m 3from Table 1.2-1.
M = The air dose factor due to gamma emissions for each identified noble I
gas radionuclide, in mead /yr per pCi/m3 from Table 1.2-1 (unit conver- ;
sion constant of 1.1 mrem / mrad converts air dose to skin dose). l P.
= The dose parameter for radionuclides other than noble gases for the ;
inhalation pathway, in mrem /yr per pCi/m3 and for the food and ground plane pathways in m2 .(mrem /yr) per pCi/sec from Table 1.2-2. The dose factors are based on the critical individual organ and most restrictive age group (child or infant).
= The release rate of radionuclides, i, in gaseous effluent from all D*. release points at the site, in pCi/sec.
C-6 Rev. 12 9/19/86
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- l. '. *3 (X/Q) = 3.10E-05 sec/m3 . -The highest calculat'ed annual. average relative concentration '(dispersion parameter) for any area at or beyond the unrestricted area boundary. The location is the NNE sector
@.0.5 miles.
W = The highest calculated annual average dispersion parameter for estimating the dose to an individual at a location in the unrestricted area where the total inhalation, food and ground plane pathway dose is determined to be a maximum based on operational source term data, land use surveys, and NUREG-0133 guidance.
W = 3.1E-05 sec/m3 , for the inhalation pathway. The location is the NNE sector @ 0.5 miles.
W = 1.1E-07 meter 2, for the food and ground plane pathways. The location is the NE/NNE sector @ 0.5 miles.
D.=kC.f+k2 1 t 1 = 4.72E+2C.f 1 where:
C. = the concentration of radionuclide, i, in undiluted gaseous effluent, 1
in pCi/ml.
f = the undiluted effluent flow, in efm d i = conversion factor, 2.83E4 ml/ft 3 k2 = conversion factor, 6El sec/ min C-7 Rev. 13-1/1/87
C3.0 RADIATION MONITOR SETPOINTS Using the generic calculations presented in Section 2.0, final effluent radiation monitoring setpoints are calculated for monitoring as required by the Technical Specifications.
All radiation monitors for Catawba are off-line except EMF-50 (Waste Gas System) which is in-line. These monitors alarm on low flow; the minimum flow alarm level for both the liquid monitors and the gas monitors is based on the manufacturer's recommendations except EMF-50. These monitors measure the activity in the liquid or gas volume exposed to the detector. Liquid monitors are independent of flow rate if a minimum flow rate is assured. Gaseous monitors are dependent on pressure or vaccum. Particulate monitors are dependent on flow rates.
Radiation monitoring setpoints calculated in the following sections are expressed in activity concentrations; in reality the monitor readout is in counts per minute. Station radiation monitor setpoint procedures which correlate concentration and counts per minute shall be based on the following re.la t.ionship :
c = E 22 x 10 Ue V where:
c = the gross activity, in pCi/ml r = the count rate, in cpm 2.22 x 106 = the disintegration per minute per pCi e = the counting efficiency, cpm /dpm V = the volume of fluid exposed to the detector, in ml.
For those occurrences when simultaneous releases of radioactive material must be made, monitor setpoints will be adjusted downward in accordance with Station Procedures to insure that instantaneous concentrations will not be exceeded. ;
C3.1 LIQUID RADIATION MONITORS C3.1.1 Waste Liquid Effluent Line As described in Section C2.1.1 on release rate calculations for the waste l liquid effluent, the release is controlled by limiting the flow rate of l effluent from the station. Although the release rate is flow rate controlled, the radiation monitor setpoint shall be set to terminate the release if the effluent activity should exceed that determined by laboratory analyses and used to calculate the release rate. A typical radiation monitor setpoint may be calculated as follows:
e$"[ $ 2.48E-05 pCi/ml where:
c = the gross activity in undiluted effluent, in pCi/ml C-8 Rev. 13 1/1/87
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a f = the flow from the tank may vary from 0-100 gpm but, for this ' calculation, o is assumed to be 100 gim.
MPC = 1.0E-07 pCi/ml, the MPC for an unidentified mixture a = 1.027 (See Section C2.1.1) N F = the dilution flow may vary as described in section C2.1.1, but is conservatively estimated at 25,500 gpm, the minimum flow available.
C3.1.2 Containment Ventilation Unit Condensate Effluent Line - EMF 44 As described in Section C2.1.2 on release rate calculations for the containment ventilation unit condensate effluent, it is likely that the effluent will contain measurable activity above background. It is assumed that effluent activity is less than the monitor's setpoint until indicated by a radiation a la rm. Since the tank contents are discharged automatically, the radiation monitor setpoint will be set at 1.0E-06 pCi/ml (the monitor's minimum practical setpoint) plus background to assure that release limits are not exceeded.
C3.1.3 Auxiliary Feedwater Sump Pumps and Floor Drain Sump Pump - EMF 52 As described in Section C2.1.2 on release rate calculations for the auxiliary feedwater sump pumps and floor drain sump pump effluents, it is possible that the effluent will contain measurable activity above background. It is assumed that the effluent activity is less than the monitors setpoint until indicated by a radiation alarm. Since the sumps are discharged automatically, the radiation monitor setpoint will initially be set at 1.0E-06 pCi/ml (the monitor's minimum practical setpoint) plus background to assure that no activity is unknowingly discharged into the Turbine Building sump. The set-point may be changed af ter initial detection to allow positive control of effluent releases using the guidance given in Section C3.1.5.
C3.1.4 Steam Generator Blowdown Line - EMF 34 As described in Section C2.1.2 on Release Rate Calculations for the Steam Generator Blowdown, it is possible that the effluent will contain measurable activity above its monitors setpoint. It is assumed that no activity is present in the effluent until indicated by radiation monitoring. Since the Steam Generator 31owdown line is discharged automatically, the radiation ;
monitor setpoint will be initially set at 1.0E-06 pCi/ml (the monitor's minimum practical setpoint) plus background to assure no activity is unknowingly l discharged into the Turbine Building sump. The setpoint may be changed after I detection to allow positive control of effluent releases using the guidance given in Section C3.1.5.
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C3.1.5 Turbine Building Sump Discharge Line - EMF 31 As described in Section C2.1.2 on release rate calculations for the turbine building sumps, it is possible that the effluent will contain measurable activity above its monitor setpoint. Since the sump contents are discharged automatically, the radiation monitor setpoint will be initially set at 1.0E-06 pCi/ml (the monitor's minimum practical setpoint) plus background to assure:
that no activity is unknowingly discharged into the WC system. Should radioactive effluent releases need to be made from the TBS via the WC system a typical monitor setpoint may be calculated as follows:
M xF c$ g $ 1.42E-06 pCi/ml where:
c = the gross activity in undiluted effluent, in pCi/ml f = the undiluted effluent flow in gpm; for this example the flow is from the Turbine Building Su.aps and is asstmed to be 250,000 gallons / day or 2175 rpm.
MPC = 1.0E-07 pCi/ml, the MPC for an unidentified mixture o = 1.027 (See Section C2.1.1)
F = the dilution flow, in gpm, availabic to dilute the undiluted effluent discharge flow (f); for this example it is assumed that 2550 gpm (10% of the stations RL minimum flow) will be used to dilute the discharge of the WC system. This flowrate will allow the WC system to discharge 10% of the stations MPC and dose limits.
C-10 Rev. 13 1/1/87 i
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! C3.2 GAS MONITORS The following equation shall be used to calculate noble-gas radiation monitor setpoints based on Xe-133.(Historical data shows that Xe-133 is the predomtnant l isotope):
P K(X/Q)Q. 1
< 500 (see-Section C2.2.1) J'-
Q = 4.72E+02 Cg f (see Section C2.2.2)
C. 1
< 116/f' l
where:
C = the gross activity in undiluted effluent, in pCi/ml 1
f = the flow from the tank or building sources, in_cfm
'K = from Table 1.2-1 for Xe-133, 2.94E+2 mrem /yr per pCi/m 3 X/Q = 3.1E-05, as defined in Section C.2.2.2 As stated in Section C2.2, the unit vent is the release point for the contain-ment purge ventilation system, the containment air release and addition system, the condenser air ejector, and auxiliary building ventilation.
For releases from the containment purge ventilation system, a typical radiation monitor setpoint may be calculated as follows:
C.1 < 116/f = 6.5E-04 where:
f = 151,000 cfm (auxiliary building ventilation) + 28,000 cfm (containment purge) = 179,000 cfm For release from the containment air release and addition system, the waste gas decay tanks, the condenser air ejectors, and the auxiliary building ventilation, a typical radiation monitor setpoint may be calculated as follows:
C.1 < 116/f = 7.7E-04 where:
f = 151,000 cfm (auxiliary building ventilation) a C-11 Rev. 12 9/19/86
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C4.0 DOSE CALCULATIONS C4.1 FREQUENCY OF CALCULATIONS Dose contributions to the maximum exposed individual shall be calculated every 31 days, quarterly, semiannually, and annually (as required by Technical Spec-ifications) using the methodology in the generic information sections. This methodology shall also be used for any special reports. Dose. projections may ,l be performed using simplified estimates. Fuel cycle dose calculations shall be performed annually or as required by special reports. Dose contributions may be calculated using the methodology in the appropriate generic information sections.
C4.2 DOSE MODELS FOR MAXIMUM EXPOSED INDIVIDUAL C4.2.1 Liquid Effluents For dose contributions from liquid radioactive releases, dose calculations based on operational source term data and NUREG-0133 guidance indicate that the maximum exposed individual would be an adult who consumed fish caught in the discharge canal and who drank water from the nearest "downstream" potable water intake. The dose from Cs-134 and Cs-137 has been calculated to be 90% of that individual's total body dose.
C4.2.2 Gaseous Effluents C4.2.2.1 Noble Gases For dose contributions from exposure to beta and gamma radiation from noble gases, it is assumed that the maximum exposed individual is an adult at a controlling location in the unrestricted area.where the total noble gas dose is determined to be a maximum.
C4.2.2.2 Radioiodines, Particulates, and Other Radionuclides T 1/2 > 8 days For dose contributions from radioiodines, particulates and other radionuclides; it is assumed that the maximum exposed individual is a child or infant at a controlling location in the unrestricted area where the total inhalation, food and ground plane pathway dose is determined to be a maximum based on operation-al source term data, land use surveys, and NUREG-0133 guidance.
C4.3 SIMPLIFIED DOSE ESTIMATE C4.3.1 Liquid Effluents For dose estimates, a simplified calculation based on the assumptions presented in Section C4.2.1 and operational source term data is presented below. Updated operational source term data shall be used to revise these calculations as necessary.
, m D
WB = 6.38E+05 I (F g)(T )g (CCs-134 + 0.59 CCs-137) f=1 C-12 Rev. 13 1/1/87
- f. . .$ .
where:
6.38E+05 = 1.14E+05 (U,y/Dy + U,g BFg ) DF,g7 (1.10) where:
1.14E+05 = 106 pCi/pCi x 103ml/kg 8760 hr/yr ;
U,y = 730 kg/yr, adult water consumption Dy = 37.7, dilution factor from the near field area to the nearest potable water intake.
U,g = 21 kg/yr, adult fish consumption BF1 = 2.00E+03, bioaccumulation factor for Cesium (Table 3.1-1)
DFau. = 1.21E-04, adult, total body, ingestion dose factor for Cs-134 (Tabic
3.1-2) >
1.10 = factor derived from the assumption that 90% of dose is frem Cs-134 '
and Cs-137 or 100% + 90% = 1.10 And where:
FA=F+f where:
f = liquid radwaste flow, in gpm o = recirculation factor at equilibrium, 1.027 (see Section C2.1.1)
F = dilution flow, in gpm And where:
Tg = The length of time, in hours,.over which CCs-134, CCs-137, and Fg are averaged.
CCs-134 = the average concentration of Cs-134 in undiluted effluent, in i pCi/ml, during the time period considered. ;
C = e average ncen r ti n s-137 in undiluted e Uluent, in pCi/ml, Cs-137 {
during the time period considered. l l
0.59 = The ratio of the adult total body ingestion doce factors for Cs-134 and Cs-137 or 7.14E-05 + 1.21E-04 = 0.59 C-13 Rev. 13 1/1/87 l
- ~ _ , , , , . . _.y_ , , . , .
_ . - _ . r _ , - , _ , . , , . _ - . - , -
- - . . . - - - . - . .-. . . - . - . ~ . . .. . _ -
.. . *f.
C4.3.2 Gaseous Effluents Meteorological data is provided in Tables C4.0-1 and C4.0-2.
C4.3.2.1 Noble Gases For dose estimates, simplified dose calculations based on the r.ssumptions in C4.2.2.1 and operational source term data are presented below. Updated operational source term data shall be used to revise these calculations as necessary. These calculations further assume that the annual average dispersion parameter is used and that Xenon-133 contributes 40% of the gamma air dose and 80% of the beta air dose.
D
=3.47E-10[D}Xe-133(2.50)
D p=.1.03E-09[D}Xe-133(1.25) where:
3.47E-10 = (3.17E-8)(353) (X/Q), derived from equation presented in Section 3.1.2.1.
1.03E-09 = (3.17E-08) (1050) (X/Q), derived from equation presented in Section 3.1.2.1.
= 3.1E-05 sec/m 3 , as defined in Section C2.2.2 X/Q
[D}Xe-133 = the total Xenon-133 activity released in pCi 2.50 = factor derived from the assumption that 40% of the gamma air dose is contributed by Xe-133.
1.25 = factor derived from the assumption that 80% of the beta air dose is contributed by Xe-133.
C4.3.2.2 Radioiodines, Particulates, and Other Radionuclides with T 1/2 ) 8 days i
For dose estimates, simplified dose calculations based on the-assumptions in C4.2.2.2 and operational source term data are presented below. Updated operational source term data shall be used to revise these calculations as necessary. These calculations further assume that the annual average l dispersion / deposition parameters are used and that 80% of the dose results from i Iodine-131 ingested by the maximally exposed individual via the vegetable !
garden pathway at the controlling location. The simplified dose estimate for i exposure to the thyroid of a child is:
J D=7.51E+02w(D)I-131(1.25)
I where:
w = 1.1E-07 = D3 for food and ground plane pathway, in m2 from Table C4'.0-2 for the controlling location (NNE sector at 0.6 miles).
C-14 Rev. 13 1/1/87
1 (D)I-131 = the total Iodine-131 activity released in pCi.
7.51E+02 =(3.17E-08)(Rf[D/Q])withtheappropriatesubstitutionsfor child-vegetablegardenpathwayfactor,Rf[D3]forIodine-131. See Section 3.1.2.2.
1.25 = factor derived from the assumption that 80% of the total inhalation, food and ground plane pathway dose to the maximally-exposed individual is contributed by I-131 via the vegetable garden pathway.
C4.4 FUEL CYCLE CALCULATIONS As discussed in Section 3.3.5, more than one nuclear power station' site may contribute to the doses to be considered in accordance with 40CFR190. ..The fuel cycle dose assessments for Catawba Nuclear Station must include gaseous dose contributions from McGuire Nuclear Station, which is located approximately thirty miles NNE of Catawba. For this dose assessment, the total body and maximum organ dose contributions to the maximum exposed individual from the combined Catawba and McGuire liquid and gaseous releases are estimated using the following calculations:
Dg (T) = Dg (1,) + Dg (1 ) + Dg (g,) + Dg(gc )
DM0(T) = DM0(I m ) + DM0(I c ) + M0(8 m )
- M0(8c }
where:
Dg(T) = Total estimated fuel cycle whole body dose commitment resulting from the combined liquid and gaseous effluents of Catawba and McGuire during the calender year of interest, in mrem.
DM0(T) = Total estimated fuel cycle maximum organ dose commitment result-ing from the combined liquid and gaseous effluents of Catawba and McGuire during the calender year of interest, in mrem.
C4.4.1 L,1 quid Effluents Liquid pathway dose estimates are based on values and assumptions presented in Sections B4.3.1. and C4.3.1. Operational source terms shall be used to update these simplified calculations as necessary.
C4.4.1.1 McGuire's Liquid Contributions Based on operational history, the Catawba fuel cycle maximum exposed individual whole body doss resulting from McGuire liquid effluent releases (Dg (1 )) is estimated using the simplified dose calculation given below:
Dg (1,) = (8.78E+05) ( Fg)(Tg ) (CCs-134 + 0.59 CCs-137 where:
C-15 Rev. 13 1/1/87
8.78E+5 = 1.14E+05 ( U, + U,f x BFg ) ( DF,g ,) ( 1.49 )
where:
1.14E+05 = ( 1.0E+06 pCi/uCi x 1.0E+03 ml/kg ) / ( 8760 hr/yr )
U,y = 730 kg/yr, Adult water consumption U,f = 21 kg/yr, Adult fish cons:unption BF; = 2.00E+03, Bioaccumulation factor for Cesium (Table 3.1-1)
DF ait
= 1.21E-04, Adult total body ingestion dose factor for Cs-134 (Table 3.1-2) 1.49 = Factor derivei from the assumption that 67% of the dose is derived from Cs-134 and Cs-137 or 100% / 67% = 1.49 where:
Fg=f/F where:
f = McGuire's liquid radwaste flow, in gpm F = 1.97E+06 gpm, the average flow past Lake Wylie Dam i
where: '
T and g=F8760 hours, the time period of time over which CCs-134 , CCs-137 g are averaged.
CCs-134 = The average concentration of Cs-134 in McGuire's undiluted effluent, in uCi/ml, during the calender year of interest.
CCs-137 = The average concentration of Cs-137 in McGuire's undiluted effluent, in uCi/ml, during the calender year of interest.
0.59 = The ratio of the adult total body ingestion dose factors for Cs-134 and Cs-137 or 7.14E-05 / 1.21E-04 = 0.59 Based on operational history, the Catawba fuel cycle maximum exposed individual ;
maximum organ dose (Adult-GI-LLI) resulting from McGuire's liquid effluent '
releases (Dgn(1 )) is conservatively estimated using the simplified dose calculation for the infant thyroid given below:
DM0(I m ) = ( 6.95E+05) ( Fg)(Tg ) (CI-131) where:
6.95E+05 = { 1.14E+05 ( U, ) (DF,1 ) (1.33) where:
C-16 Rev. 13 1/1/87
1.14E+05 = ( 1.0E+06 pCi/uCi x 1.0E+03 ml/kg ) / ( 8760 hr/yr )
U,y = 330 kg/yr, infant water consumption DFat.t = 1.39E-02, infant thyoid ingestion dose factor for I-131 (Table 3.1-2 )
1.33 = Factor derived from the. assumption that 75% of infant thyroid dose is from I-131 or 100% / 75% = 1.33 where:
Fg=f/F where:
f = McGuire's liquid radwaste flow, in gpm F = 1.97E+06 gpm, the average flow past Lake Wylie Dam where:
Tg = 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br />, the time period of time over which CI-131 and Fg are averaged.
C = The average concentration of I-132 in McGuire undiluted I-131 effluent, in uCi/ml, during the calender year of interest.
C4.4.1.2 Catawba's Liquid Contribution Based on operational history, the Catawba fuel cycle maximum exposed individual whole body dose resulting from Catawba's liquid effluent releases (DWB (Ic )) is estimated using the simplified dose calcuation given belew:
DWB(I c ) = ( 6.38E+05 ) ( Fg)(Tg ) (CCs-134 + 0.59 CCs-137 )
where: i 6.38E+05 = 1.14E+05 ( U, /Dy + U,f x BF1 ) ( DF,g ) ( 1.10 )
where: l l
1.14E+05 = ( 1.0E+06 pCi/uci x 1.0E+03 ml/kg ) / ( 8760 hr/yr )
U aw
= 730 kg/yr, Adult water consumption Dy = 37.7, dilution factor from the near field area to the nearest potable water intake (Rock Hill Water Intake)
U af
21 kg/yr, Adult fish consumption BF
i 2.00E+03, Bioaccumulation factor for Cesium (Table 3.1-1)
DF ". = 1.21E-04, Adult total body ingestion dose factor for Cs-134 (Table 3.1-2)
C-17 Rev. 13 1/1/87-
, , -.,1 em,v-,- --n ,-- -,c- -------,n- n - + ~~+v~~c ~~ ' ' ' ' * - -*
l 1.10 = Factor derived from the assumptiom that 90% of the dose is derived from Cs-134 and Cs-137 or 100% / 90% = 1.10 and where:
Fg=(f)(o)/(F+f) where:
f= Catawba's liquid radwaste flow, in gpm o= Recirculation factc,r at equilibrium, 1.027 (See section C2.2.1)
F= Catawba's dilution flow, in gpm and where:
T and g=F8760 arehours, the time period of time over which CCs-134, CCs-137 averaged.
g -
CCs-134 = The average concentration of Cs-134 in Catawba's undiluted effluent, in uCi/ml, during the calender year of interest.
CCs-137 = The average concentration of Cs-137 in Catawb'a's undiluted effluent, in uCi/ml, during the calender year of interest.
0.59 = The ratio of the adult total body ingestion dose factors for Cs-134 and Cs-137 or 7.14E-05 / 1.21E-04 = 0.59 Based on operational history, the Catawba fuel cycle maximum exposed individual maximum organ dose (Adult, GI-LLI) resulting from Catawba's liquid effluent releases (DM0(I c )) is estimated using the simplified dose calculation given below:
DM0(1 c ) = ( 1.89E+06 ) (Fg ) (T g) (CNb-95)
I where: l l
1.89E+06 = ( 1.14+05 ) ( U /D 9+ U,f x BFf ) (DF,g) (1.25) l l
where:
1.14E+05 = ( 1.0E+06 pCi/uCi x 1.0E+03 ml/kg ) / 8760 hr/yr U,y = 730 kg/yr, Adult water consumption D" = 37.7, Dilution factor from the near field area to potable water intake.
U,f = 21 kg/yr, Adult fish consumption BF1= 3.00E+04, Bioaccumulation factor for Niobium (Table 3.1-1) j C-18 Rev. 13 l 1/1/87 I
~*
- .~ . ..
DF1. = 2.10E-05, Adult GI-LLI ingestion dose factor for Nb-95 (Table 3.1-2) 1.25 = Factor derived from the assumption that 80% of adult GI-LLI dose is from Nb-95 or 100% / 80% = 1.25 where:
Fg = (f) (a) / ( F + f - )
where:
f= Catawba's liquid radwaste flow, in gpm o= Recirculation factor at equilibrium, 1.027 F= Catawba's dilution flow, in gpm where:
Tg = 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br />, the time period of time over which C Nb-95 and Fg are averaged.
CNb-95 = The average concentration of Nb-95 in Catawba's undiluted effluent, in uCi/ml, during the calender year of interest.
C4.4.2 Gaseous Effluents Airborne effluent pathway dose estimates are based on the values.and assump-tions presented in Sections B4.3.2. and C4.3.2. Operational source term data shall be used to update these calculations as necessary.
C4.4.2.1 McGuire's Gaseous Contribution Based on operational history, the Catawba fuel cycle maximum exposed individual whole body dose resulting from McGuire's gaseous effluent releases (DWB(E m )) is estimated using the simplified dose calculation given below:
DWB(8 m ) = ( 9.32E-06 ) ( w ) ( Xe-133 ) F) ( 1.67 )
where:
w= 1.50E-07 = (X/Q) for the plume immersion factor pathway factor, in sec/m3 which corresponds to a location 5 miles SSW of the McGuire site (See table B4.0-1)
Q Xe-133 = The total Xe-133 activity released from McGuire during the calender year of interest, in uCi.
9.32E-06 = ( 3.17E-08 ) ( K.1X/Q] ), with appropriate substitutions for whole body expcsure in a semi-infinite cloud of Xe-133. See Section 1.2.1.
C-19 Rev. 13 1/1/87
+---r- s.y e-,3+ v- ,weyy ,- w w-- - -
-m y v y --
,, '(.
- . Sp = 0.7 = External radiation shielding factor for individuals.
, 1.67 = The factor derived from the conservative assumption (based on l' historical data) that 60% of the whole body dose to the maximally
- exposed individual is contributed by Xe-133.
Based on operational history, the Catawba i i cycle maximum exposed individual ,
maximum organ dose (Adult-GI-LLI) resulting from McGuire's gaseous effluent releases (Dgg(g )) is conservatively estimated using the simplified dose calculation for,the infant thyroid given below:
s DM0(8 m- ) = ( 1.84E+04) ( w ) ( QI -131) ( 1.05 )
where:
w = 3.80E-10, D/Q for the food and ground plane pathway, in m' , for a location 5 miles SSW of the McGuire site (Table B4.0-2) s QI -131
= The total I-131 activity released from McGuire during the calender year of interest, in uCi.
C 1.84E+04 = ( 3.17E-08 ) ( Rg [D/Q] ) with appropriate substitutions for C
goat's milk in the grass goat-milk pathway, Rg [D/Q) for 1-131. See Section 3.1.2.2.
1.05 = The factor derived from the conservative assumption (based on historical data) that 95% of the total inhalation, food and ground plane pathway dose to tne. maximally exposed individual is contri-buted by I-131 via the goat-milk pathway.
C4.4.2.2 Catawba'; Gaseous Contribution Based on operational history, the Catawba fuel cycle maximum exposed individual whole body dose resulting from Catawba's gaseous effluent releases (DWB(8c )) IO estimated using the simplified dose calculation given below:
s DWB(8 c ) = ( 9.32E-06 ) ( w ) ( QXe-133) ( F) ( 2.50 )
where:
w = 3.10E-05 (X/Q) for the plume immersion factor pathway factor which otresponds to a location 0.5 miles NNE of the Catawbs site (see Table C4.0-1) s Q Xe-133 = The total Xe-133 activity released from Catawba during the calender year of interest, in uCt.
I C-20 hav. 13 1/1/87
__m _ _ _ . . __ . . . . _ _ _ __ - _ . _ _
y r'
. c.
g -
9.32E-06 = (3.17E-08) (K g
[X/Q]), with appropriate substitucions for whole body exposure in a semi-infinite cloud of Xe-133. See Section 1.2.1.
Sp = 0.7 = External radiation shielding factor for individuals.;
2.50 = The factor derived from the conservative assumption (based on historical data) that 40% of the whole body dose. to the maximally.
exposed individual is contributed by Xe-133.-
Based on design basis operation, the Catawba fuel cycle maximum exposed individual maximum organ dose (Adult-GI-LLI) resulting from Catawba's gaseous effluent releases (DM0(8 c )) is conservatively estimated using the simplified dose calculation for the child thyroid given below:
DM0(8 c ) = ( 7.51E+02 ) ( w ) ( I-131 ) ( 1.25 )
where:
w= 1.1E-07 = D/Q for the food and ground plane pathway in -2 m , for a location 0.6 miles NNE of the Catawba site (see Table C4 0-2).
QI -131
= The total I-131 activity released from Catawba during the calender year of interest, in uCi.
V 7.51E+02 = ( 3.17E-08 ) ( R f (D/Q] ) with appropriate substitutions for V
the child-vegetable garden pathway factor, R.[D/Q] 1 for I-131.
i See Section 3.1.2.2.
1.25 = The factor derived from the assumption that 80% of the total inhalation, food ano ;;round plane pathway dose to the maximally, exposed individual is contributed by I-131 via the vegetable garden pathway.
C-21 Rev. 13 1/1/87
. j TABLE C4.0-1 (1 of 2)
CATAWBA NUCLEAR STATION DISPERSION P'sRAMETER (X/Q) FOR LONG TERM RELEASES > 500 HR/YR OR > 125 HR/QTR (sec/m3 )
Distance to the control location, (miles)
Sector 0.5 1._0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 N 2.6E-5 6.5E-6 2.7E-6 1.5E-6 9.7E-7 6.9E-7 5.2E-7 4.1E-7 3.3E-7 2.8E-7 ENE 3.1E-5 8.1E-6 3.3E-6 1.8E-6 1.2E-6 8.2E-7 6.2E-7 4.9E-7 4.0E-7 3.3E-7 NE 3.0E-5 7.8E-6 3.2E-6 1.8E-6 1.1E-6 8.0E-7 6.0E-7 4.7E-7 3.9E-7 3.2E-7 ENE 1.5E-5 3.9E-6 1.6E-6 8.9E-7 5.7E-7 4.1E-7 3.1E-7 2.4E-7 2.0E-7 1.6E-7 E 1.4E-5 3.7E-6 1.5E-6 8.4E-7 5.4E-7 3.8E-7 2.9E-7 2.3E-7 1.9E-7 1.6E-7 ESE 9.0E-6 2.3E-6 9.5E-7 5.3E-7 3.4E-7 2.4E-7 1.8E-7 1.4E-7 1.2E-7 9.7E-8 SE 9.2E-6 2.4E-6 9.8E-7 5.4E-7 3.5E-7 2.4E-7 1.8E-7 1.4E-7 1.2E-7 9.8E-8 SSE 1.1E-5 2.9E-6 1.2E-6 6.4E-7 4.1E-7 2.9E-7 2.2E-7 1.7E-7 1.4E-7 1.1E-7 S 2.5E-5 6.4E-6 2.6E-6 1.5E-6 9.3E-7 6.6E-7 5.0E-7 3.9E-7 3.2E-7 2.7E-7 SSW 1.7E-5 4.4E-6 1.8E-6 1.0E-6 6.4E-7 4.5E-7 J.4E-7 2.7E-7 2.2E-7 1.8E-7 SW 1.3E-5 3.4E-6 1.4E-6 7.4E-7 4.7E-7 3.3E-7 2.4E-7 1.9E-7 1.5E-7 1.3E-7 WSW 7.0E-6 1.8E-6 7.2E-7 3.9E-7 2.5E-7 1.7E-7 1.3E-7 1.0E-7 8.2E-8 6.8E-8 W 8.9E-6 2.3E-6 9.3E-7 5.0E-7 3.2E-7 2.2E-7 1.7E-7 1.3E-7 1.1E-7 8.7E-8 WNW 6.6E-6 1.7E-6 6.8E-7 3.7E-7 2.4E-7 1.7E-7 1.3E-7 9.8E-8 8.0E-8 6.6E-8 NW 1.0E-5 2.6E-6 1.1E-6 5.9E-7 3.8E-7 2.7E-7 2.0E-7 1.6E-7 1.3E-7 1.1E-7 NNW 1.3E-5 3.3E-6 1.4E-6 7.5E-7 4.8E-7 3.4E-7 2.6E-7 2.0E-7 1.6E-7 1.4E-7 Rev. 4
~7/18/84
TABLE C4.0-1 (2 of 2)
A C_ATAWBA NUCLEAR STATION The values presented in this table were generate'd by using the computer program X0QD0Q (NUREQ/CR-2919) which implements NRC Regulatory Guide 1.111 (1977) and the following assumptions:
- 1. Data Collection Period, 12/17/75 to 12/16/77.
- 2. Ground Level Releases.
- 3. Height of The Vent's Building = 47 meters.
- 4. Open Terrain Recirculation Correction Factors.
1 l
l Rev. 13 1/1/87 l
. [
TABLE C4.0-2 (1 of 2)
CATAWBA NUCLEAR STATION DISPERSION PARAMETER (IUQ) FOR LONG TERM RELEASES > 500 IIR/YR OR > 125 !!R/QTR (meter 2)
Distance to the control location, (miles)
Sector 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 N 6.4E-8 1.6E-8 5. 6T?-9 2.8E-9 1.6E-9 1.1E-9 7.5E-10 5.6E-10 4.3E-10 3.4E-10 NNE 1.1E-7 2.7E-8 9.6E-9 4.7E-9 2.SE-9 1.8E-9 1.3E-9 9.5E-10 7.4E-10 5.8E-10 NE 1.1E-7 2.6E-8 9 . T', -9 4.62-9 2.7E-9 1.8E-9 1.3E-9 9.3E-10 7.2E-10 5.7E-10 ENE 4.1E-S 1.0E-8 3.6E-9 1.EE-9 1.1E-9 6.9E-10 4.9E-10 3.6E-10 2.8E-10 2.2E-10 f E 3.6E-8 8.8E-9 3.2E-9 1.6E-9 9.3E-10 6.1E-10 4.3E-10 3.2E-10 2.4E-10 1.9E-10 f ESE 2.5E-8 6.0E-9 2.2E-9 1.1E-9 6.3E-10 4.2E-10 2.9E-10 2.2E-10 1.7E-10 1.3E-10 SE 3.0E-8 7.3E-9 2.6E-9 1.3E-9 7.7E-10 5.0E-10 3.5E-10 2.6E-10 2.0E-10 1.6E-10 SSE 3.8E-8 9.3E-9 3.3E-9 1.7E-9 9.7E-10 6.4E-10 4.5E-10 3.3E-10 2.6E-10 2.0E-10 S 7.2E-8 1.8E-8 6.3E-9 3.1E-9 1.8E-9 1.2E-9 8.5E-10 6.3E-10 4.8E-10 3.8E-10 SSW 6.6E-8 1.6E-8 5.8E-9 2.9E-9 1.7E-9 1.1E-9 7.3E 5.8E-10 4.4E-1(, 3.5E-10 SW 5.7E-8 1.4E-8 5.0E-9 2.5E-9 1.5E-9 9.6E-10 6.7E-10 5.0L-19 3.9E-10 3.1E-10 WSW 2.4E-8 5.7E-9 2.1E-9 1.0E-9 6.0E-10 4.0E-10 2.8E-10 2.1E-10 1.6E-10 1.3E-10 V 2.8E-8 6.7E-9 2.4E-9 1.2E-9 7.0E-10 4.6E-10 3.2E-10 2.4E-10 1.9E-10 1.5E-10 tant 1.9E-8 4.6E-9 1.7E-9 8.2E-10 4.8E-10 3.2E-10 2.2E-10 1.6E-10 1.3E-10 1.0E-10 NW 2.9E -8 7.0E-9 2.5E-9 1.3E-9 7.3r.-10 4.8E-10 3.4E-10 2.5E-10 1.9E-10 1.5E-10 NEW A.IE-8 9.9E-9 5.6E-9 1.8E-9 1.0E-9 6.8E-10 4.8E-10 3.6E-10 2.7E-10 2.2E-10 Rev. 4 7/18/84
TABLE C4.0-2 (2 of 2)
, CATAWBA NUCLEAR STATION Thevaluespresentedint\istableweregeneratedbyusingthecomputerprogram X0QD0Q (NUREQ/CR-2919) which implements NRC Regulatory Guide 1.111 (1977) and the following assumptions: -
- 1. Data Collection Period, 12/17/75 to 12/16/77.
- 2. Ground Level Releases.
- 3. Height of The Vent's Building = 47 meters.
- 4. Open Terrain Recirculation Correction Factors.
Rev. 11 8/31/86
..-.--.~.-~.v_- . - . .
f . ' ^ a .l .
L TABLE C4.0-3 * ,,
(1 of 3)
CATAWBA NUCLEAR STATION ADULT A g DOSE PARAMETERS (mrem /hr per pCi/ml)
NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG- GI-LII H 3 0.0 4.58E-01 4.58E-01 4.58E-01 4.58E-01 -4.58E-01 4.58E-01 NA 24 4.11E+02 4.11E+02 4.11E+02 4.11E+02 4.11E+02 4.11E+02 4.11E+02 CR 51 0.0 0.0 1.28E+00 7.65E-01 2.82E-01 1.70E+00 3.22E+02 MN 54 0.0 4.39E+03 8.37E+02 0.0 1.31E+03 0.0 1.34E+04 MN 56 0.0 1.10E+02 1.96E+01- 0.0 1.40E+02 0.0 3.52E+03 FE 55 6.64E+02 4.59E+02 1.07E+02 0.0 0.0 2.56E+02 2.63E+02-FE 59 1.05E+03 2.46E+03 9.45E+02 0.0 0.0 6.89E+02 8.21E+03 s C0 58 0.0 9.08E+01 2.04E+02 0.0 0.0 0.0 1.84E+03 CO 60 0.0 2.61E+02 5.75E+02 0.0- 0.0 0.0 4.90E+03 4
NI 63 3.14E+04 2.18E+03 1.05E+03 0.0 0.0 0.0 4.54E+02 N1 65 1.28E+02 1.66E+01 7.56E+00 0.0 0.0 0.0 4.20E+02 CU 64 0.0 1.02E+01 4.77E+00 0.0 2.56E+01 0.0 8.66E+02 ZN 65 2.32E404 7.38E+04 3.33E+04 0.0 4.93E+04- 0.0 4.65E+04 ZN 69 4.93E+01 9.44E+01 6.56E+00 0.0 6.13E+01 0.0 .1.42E+01
- BR 83 0.0 0.0 4.05E+01 0.0 0.0 0.0 5.83E+01 BR 84 0.0 0.0 5.25E+01 0.0 0.0 0.0 4.12E-04 BR 85 0.0 0.0 2.16E+00 0.0 0.0 0.0 0.0 RB 86 0.0' 1.01E+05 4.71E+04 0.0 0.0 0.0 1.99E+04 RB 88 0.0 2.90E+02 1.54E+02 0.0 0.0 0.0 4.00E-09 RB 89 0.0 1.92E+02 1.35E+02 0.0 0.0 .0.0 1.12E-11
, SR 89 2.28E+04 0.0 6.54E+02 0.0 0.0 0.0 3.66E+03 SR 90 2.84E+05 0.0 7.62E+04 0.0 0.0 0.0 1.62E+04 SR 91 4.20E+02 0.0 1.70E+01 0.0 0.0 0.0 2.00E+03 SR 92 1,59E+02 0.0 6.88E+00 0.0 0.0 0.0 3.15E+03 Y 90 5.97E-01 0.0 1.60E-02 0.0 0.0 0.0 6.33E+03 Y 91M 5.64E-03 0.0 2.18E-04 0.0 0.0 0.0 1.66E-02 Y 91 8.75E+00 0.0 2.34E-01 0.0 0.0 0.0 4.82E+03 j Y 92 5.24E-02 0.0 1.53E-03 0.0 0.0 0.0 9.18E+02
- Methodology for table provided by: M. E. Wrangler, RAB:NRR:NRC on 3/17/83 4
a TABLE C4.0-3 (1 of 3)
I Rev. 4
- 7/18/84 3
. . _ _ _ - ._ __ ~, _ . - _ . _ _ . __
- o' e?.
i
, TABLE C4.0-3 ;
(2 of 3)
-CATAWBA NUCLEAR STATION ADULT A g DOSE PARAMETERS (mrem /hr per-pCi/ml) a .c NUCLIDE B0NE LIVER T.B0DY. THYROID KIDNEY LUNG GI-LII Y 93 1.66E-01 0.0 4.59E-03 0.0 0.0 0.0 5.27E+03 ZR 95 3.07E-01 9.85E-02 6.67E-02 0.0 1.55E-01 0.0 3.12E+02 ZR 97 1.70E-02 3.43E-03 1.57E-03 0.0 5.18E-03 0.0~ 1.06E+03 NB 95 4.47E+02 2.49E+02 1.34E+02 0.0 2.46E+02 0.0 1.51E+06 M0 99 0.0 1.13E+02 2.14E+01 0.0. 2.55E+02 0.0 2.61E+02 TC 99M 9.41E-03 2.66E-02 3.39E-01 0 .' O 4.04E-01 1.30E-02 1.57E+01 TC 101 9.68E-03 1.40E-02 1.37E-01 0.0 2.51E 01 7.13E-03 4.19E-14 RU 103 4.84E+00 0.0 2.08E+00 0.0 1.85E+01 0.0 5.65E+02 RU 105 4.03E-01 0.0 1.59E-01 0.0 5.20E+00 -
0.0 2.46E+02-RU 106 7.19E+01 0.0 9.10E+00 0.0 1.39E+02 0.0 4.65E+03 AG 110M 1.23E+00 1.14E+00 6.78E-01 0.0 2.24E+00 0.0 4.66E+02 /
TE 125M 2.57E+03 9.32E+02 3.45E+02 7.74E+02 1.05E+04 0.0 1.03E+04 TE 127M 6.50E+03 2.32E+03 7.92E+02 1.66E+03 2.64E+04 0.0 2.18E+04 TE 127 1.06E+02 3.79E+01 2.28E+01 7.82E+01 4.30E+02 0.0 8.33E+03 TE 129M 1.10E+04 4.12E+03 1.75E+03 3.79E+03 4.61E+04 0.0 -5.56E+04 L TE 129 3.01E+01 1.13E+01 7.34E+00 2.31E+01 1.27E+02 0.0 2.27E+01 TE 131M 1.66E+03 8.12E+02 6.77E+02 1.29E+03 8.23E+03 ,0.0 8.06E+04 TE 131 1.89E+01 7.90E+00 5.97E+00 1.55E+01 8.28E+01
~ '0. 0 2.68E+00 TE 132 2.42E+03 1.56E+03 1 47E+03 1.73E+03 1.51E+04 0.0 7.40E+04 I 130 2.88E+01 8.50E+01 3.35Et01 7.20E+03 1.33E+02 0.0 7.32E+01-I 131 1.59E+02 2.27E+02 1.30E+02 7.43E+04 3.89E+02 0.0 5.98E+01 I 132 7.74E+00 2.07E+01 7.24E+00 7.24E+02 3.30E+01 0.0 3.89E+00 I 133 5.41E+01 9.41E+01 2.87E+01 1.38E+04 1.64E+02 0.0 8.46E+01 l I 134 4.04E+00 1.10E+01 3.93E+00 1.90E+02 1.75E+01 0.0 9.57E-03 I 135 1.69E+01 4.42E+01 1.63E+01 2.92E+03 7.09E+01 0.0 4.99E+01 CS 134 2.98E+05 7.09E+05 5.80E+05 0.0 2.29E+05 7.62E+04 1.24E+04 CS 136 3.12E+04 1.23E+05 8.86E+04 0.0 6.85E+04 9.39E+03 1.40E+04 CS 137 3.82E+05 5.22E+05 3.42E+05 0.0 1.77Et05 5.89E+04 1.01E+04 CS 138 2.64E+02 5.22E+02 2.59E+02 0.0 3.84E+02 3.79E+01 2.23E-03 BA 139 1.14E+00 8.14E-04 3.35E-02 0.0 7.61E-04 4.62E-04 2.03E+00 i
! l l
t I I l
l Rev. 4 7/18/84 i
- , .l .
e TABLE C4.0-3 (3 of 3)
CATAWBA' NUCLEAR STATION ;
ADULT A,17 DOSE PARAMETERS (mrem /hr per pCi/ml)
NUCLIDE B0NE LIVER T. BODY THYROID KIDNEY- . LUNG GI-LII BA 140 -2.39E+02 3.00E-01 1.57E+01 0.0 1.02E-01 1.72E-01 4.93E+02 BA 141 5.55E-01 4.19E-04 1.87E-02 0.0 .3.90E-04 2.38E-04 -2.62E-10 BA 142 2.51E-01 2.58E-04 1.58E-02 0.0 2.18E-04' 1.46E-04 3.54E-19 4
LA 140 1.55E-01 7.82E-02 2.07E-02 0.0 0.0 0.0 5.74E+03 LA 142 7.94E-03 3.61E-03 9.00E-04 0.0 0.0- 0.0 2.64E+01 CE 141 4.31E-02 2.91E-02 3.30E-03 0.0 1.35E-02 0.0 - 1.11E+02 CE 143 7.59E-03 5.61E+00 6.21E-04 0.0 2.47E-03 0.0 2.10E+02 CE 144 2.25E+00 9.39E-01 1.21E-01 0.0 5.57E-01 0.0 7.59E+02 PR 143 5.71E-01 2.29E-01 2.83E-02 0.0 '1.32E-01 0.0 2.50E+03 PR 144 1.87E-03 7.76E-04 9.49E-05 0.0 4.38E-04 0.0 2.69E-10 ND 147 3.90E-01 4.51E-01 2.70E-02 0.0 2.64E-01 'O.0 2.17E+03 W 187 2.96E+02 2.48E+02 8.65E+01 0.0 0.0 0.0 -8.11E+04 NP 239 3.11E-02 3.06E-03 1.69E-03 0.0 9.54E-03 0.0 6.28E+02 I
I 1 l TABLE C4.0-3 ,
(3 of 3)
Rev. 4 7/18/84 I
C5.0 Radiological Environmental Monitoring The Radiological Environmental Monitoring Program shall be conducted in accordance with Technical Specification, Section 3/4.12.
The monitoring program locations and analyses are given in Tables C5.0-1 through C5.0-3 and Figure C5.0-1.
The laboratory performing the radiological environmental analyses shall parti-cipate in an interlaboratory comparison program which has been approved by the NRC. This program is the Environmental Protection Agency's (EPA's)
Environmental Radioactivity Laboratory Intercomparison Studies (crosscheck)
Program, our participation code is CP. l The dates of the land-use census that was used to identify the controlling-receptor locations was 06/09/86 - 06/12/86. These dates will not be changed unless a subsequent census changes a controlling receptor's location. l C-22 Rev. 13 1/1/87
. .- ~l . .
'% 4 , y TABLE C5.0-1 1 (1 of 1)
- CATAWBA RADIOLOGICAL MONITORING PROGRAM SAMPLING LOCATIONS (TLD LOCATIONS)
SAMPLING LOCATION DESCRIPTION SAMPLING LOCATION DESCRIPTION ,
200 SITE BOUNDARY (0.7M NNE) 232 4-5 MILE RADIUS (4~.1M NE) 201 SITE BOUNDARY (0.5M NE) 233 4-5 MILE RADIUS (4.0M ENE)-
202 SITE BOUNDARY (0.6M ENE) 234 4-5 MILE RADIUS (4.5M E) 203 SITE BOUNDARY (0.5M SE) 235 4-5 MILE RADIUS (4.0M ESE) 204 SITE BOUNDARY (0.5M SSW) 236 4-5 MILE RADIUS (4.2M SE) 205 SITE BOUNDARY (0.6M SW) 237- 4-5 MILE RADIUS .(4.8M SSE) 206 SITE BOUNDARY (0.7H WNW) 238 4-5 MILE RADIUS; (4.2M S) 207 SITE BOUNDARY (0.8M NNW) 239 4-5 MILE RADIUS (4.6M SSW), '
, 212 SPECIAL INTEREST (2.7H ESE) 240 4-5 MILE RADIUS- (4.1M SW) 217 CONTROL (10.0M SSE) 241 4-5 MILE RADIUS ,(44.7M WSW) 222 SITE BOUNDARY (0.7M N) 242 4-5 MILE RADIUS (4.6M W) 223 SITE BOUNDARY (0.5M E) 243 4-5 MILE RADIUS (4.6M WNW) 224 SITE EOUNDARY (0.7H ESE) 244 4-5 MILE RADIUS (4.1M NW) 225 SITE BOUNDARY (0.5M SSE) 245 4-5 MILE RADIUS (4.2M WNW). .
226 SITE BOUNDARY (0.5M S) 246 SPECIAL INTEREST (8.1M ENE) 227 SITE BOUNDARY (0.5M WSW) 247 CONTROL .(7.5M ESE)'
228 SITE BOUNDARY (0.6M W) 248 SPECIAL INTEREST (7.0M SSE) 229 SITE BOUNDARY (0.9M NW) 249 SPECIAL INTEREST (8.1M S) 230 4-5 MILE RADIUS (4.4M N) 250 .SPECIAL INTEREST (10.3M WSW)-
231 4-5 MILE RADIUS (4.2M NNE) 251 CONTROL (9.8M WNW)_ ,
Rev. 6 1/1/85
_. __ _ . _ _ _ . .. . _. _ . ~ _ - .
TABLE C5.0-2 c + .. 4 (1 of 1) v 3 7 CATAWBA RADIOLOGICAL MONITORING PROGRAM SAMPLING LOCATIONS E % i (OTHER SAMPLING LOCATIONS) L 5 % '
CODE: m E m as v O,o m 3 >
E L Uo '
W - Weekly
- SM - Semimonthly o E% $ as
- O -S BW - Biweekly Q - Quarterly Em5 as E O E $ 0 M - Mtathly SA - Seminannually E EM E E T E E I LEf t 5L an- O sg 3 o
_C dr=
5*r-8 L
E L
Eo gampl.39g, Location Descript3og______________,,__________________________________n.
<r e m ca m x u_ cn <a u. ;
200 Site Boundary (0.7m NNE) W M Q .l~
- 201 Site Boundary (0.5m NE) W M 203 Site Boundary (0.5m SE) Deleted 1/1/87 M I 205 Site Boundary (0.6m SW) W 208 Discharge Canal (0.5m S) BW SA SA
- 212 Tega Cay (2.7m ESE) W 213 Fort Mill Water Supply (7.5m ESE)- BW 214 Rock Hill Water Supply (7.3m SSE) BW 215 Camp Steere - Hwy 49 (4.Im NNE) Control BW SA 216 Hwy 49 Bridge (4.0m NNE) Control SA 217 Rock Hill Substation (10.0m SSE) Control W M 218 Belmont Water Supply (13.5m N) Control BW 219 Dairy (6.0m SW) SM 220 Dairy (8.0m WSW) SM 221 Dairy (13.0m NW) Control SM
! 226 Site Boundary (0.5m S) M 252 Residence (0.8m W) Q 253 G.ardens (5-mile radius) M(*)
4 (a) during harvest season i
Rev. 13
. . . '{
TABLE C5.0-3 *d '
(1 of 1)
CATAWBA RADIOLOGICAL MONITORING PROGRAM ANALYSES ANALYSES SAMPLE MEDIUM ANALYSIS SCHEDULE GAMMA ISOTOPIC TRITIUM LOW LEVEL I-131 GROSS BETA TLD
- 1. Radioiodine and Weekly X Particulates X X
- 2. Direct Radiation Quarterly X
- 3. Surface Water Biweekly X Monthly Co:aposite X Quarterly Composite X
- 4. Drinking Water Biweekly X Monthly Composite X X Quarterly Composite X
- 5. Shoreline Sediment Semiannually X
- 6. Milk Semimonthly X X
- 7. Fish Semiannually X
- 8. Broadleaf Vegetation Monthly X
- 9. Groundwater Quarterly X X X
- 10. Food Products Monthly (* X
~~
(a) during harvest season Rev. 13 1/1/87
, , - q, Figure C5.0-1 (1 of 2) 1 i
l l
=c' sla D
5 Figure C5.0-1 (2 of 2) l t
l 1
l