ML20079E569
| ML20079E569 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 01/10/1984 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20079E567 | List: |
| References | |
| PROC-840110, NUDOCS 8401170360 | |
| Download: ML20079E569 (32) | |
Text
{{#Wiki_filter:. _b v . s . APPENDIX C CATAWBA NUCLEAR STATION SITE SPECIFIC INFORMATION aa pg8fo8aaJ386);g A
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APPENDIX C . TABLE OF CONTENTS Page
.C1.0 CATAWBA NUCLEAR STATION RADWASTE SYSTEMS . ..... . . C-1~
C2.0; RELEASE RATE CALCULATION . ... . . .. . . .. .. .. C-4 C3.0 RADIATION MONITOR SETPOINTS . . .. . . . ... . . . , c_a C4.0 DOSE CALCULATIONS- . . . . .. . ... . . . .... . . C-12 C5 . 0 .:' RADIOLOGICAL ENVIRONMENTAL MONITORING . . . .. ... . C-17 e f f 1 i I f
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.C1.0 CATAWBA NUCLEAR STATION RADWASTE SYSTEMS C1.1 LIQUID:RADWASTE PROCESSING!
The liquid radwaste system at Catawba Nuclear Station (CNS) is used to collect and. treat fluid chemical and radiochemical by products of unit operation. The t'
' system produces efflucents which can be reused in'the plant or discharged in small, dilute quantities to~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
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cases, such as the Floor Drain Tank (FDT) Subsystem of the Liquid Waste (WL) System, filtration may 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). Ion , 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 (NB) can be treated by this 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 Evaporator, the dilute boron is concentrated to 4% so that it may be reused for makeup to the reactor coolant system.
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Figure C1.0-1 is a achematic representation of the liquid radwaste system at Catawba! l C-1 j-
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d, . Table C1.0-1 (1 of 2) ABBREVIATIONS j Systems:
.KC-- Component Cooling NB - Boron Recycle NC - Reactor Coolant ND - Residual Heat Removal NI - Safety Injection NR - Boron Thermal Regeneration i NV - Chemical Volume and Control RL - Low Pressure Service Water RN - Nuclear Service Water System WC - Conventional Waste Water Treatment WG - Waste Gas WL - Liquid Waste Recycle WS - Nuclear Solid Waste Disposal Terms:
BOL - Beginning of Core Life. BTRS - Boron Thermal Regeneration System CDT - Chemical Drain Tank L .ECST - Evaporator Concentrates Storage Tank EOL - En'd of Core Life , FDT - Floor Drain Tank FWST - Fueling Water Storage Tank (formerly Refueling Water Storage Tank) LHST - Laundry and Hot Shower Tank MST - Mixing and Settling Tank NCDT - Reactor Coolant Drain Tank RBT - Resin Batching Tank RHT - Recycle Holdup Tank RMT - Recycle Monitor Tank RMWST - Reactor Makeup Water Storage Tank i TABLE C1.0-1
- (1 of 2) i i
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o-TABLE C1.0-1 (2 of 2) 1 ABBREVIATIONS f Terms: (continued) SRST - Spent' Resin Storage Tank
. VUCDT Ventilation Unit Condensate Drain Tank
-WDT - Waste Drain Tank WEFT - Waste Evaporator Feed Tank WMT - Waste Monitor Tank-4 1
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y E, - C1.~ 2 GASEOUS 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. C1.0-2 is designed to allow for the retention, through the plant lifetime, of all the gaseous fission products to be discharged from the reactor coolant system to the chemical and volume control system or the boron recycle system, to limit the need for intentional discharge of radioactive gases from the waste gas holdup tanks. Thus, the only unavoidable sources of low-level radioactive gaseous discharge to the environment will be from periodic purging operations of the containment, from the auxiliary building ventilation system, and through the secondary system air ejec.or. With respect to the former, 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 or 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 be discharged through a charcoal adsorber and a particulate filter prior to release to.the atmosphere. Gases resulting from leakage inside the auxiliary building are released, with-out further decay, to the atmosphere via the auxiliary building ventilation system. The ventilation exhaust from potentially contaminated areas in the auxiliary building is normally unfiltered. However, on a radiation monitor alarm, the exhaust is passed through charcoal adsorbers to reduce releases to the atmosphere. C1.2.3 Secondary Systems Normally, condensate flow and steam generator blowdown will go parallel through 4 of the 5 condensate polishing demineralizers to remove activity and harmful ions from the water. Noncondensable gases will be taken from the C-2
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secondary system by the condenser steam air ejector and are passed through a radiation monitor to the unit vent. c Figure C1.0-2 is-a schematic. representation of the gaseous radwaste system at Catawba. 1 e i' i
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ROOF VENTS 600,200 cf _ _ .,. _ _ e _ _ _ _ _ _ _ _a _ ., C- OUTSIDE AIR %,J
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1 3 ABOVE PRIMARY l -% 1 GROUND UNIT COOLANT g CONDENSER l l OCh HAD! MOM TO R _ STE AM - GENERATOR V] ~~j g CONDENSER _
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77RSINh 8UTLITING~~ - dTEAM JET AIR h TO CONDENSE R WITH EJE CTO RS RAD T N y BAFFLES B LOWDOW N - RECYCLE J V TANK DoWN F SECONDARY SYSTEM (2 UNITS) pE COOuNT SYSW - - (2) H-(DROCEN ,
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CONTROLTANK /W\ / l_.,L \ SHUTDOWN j TANK UNIT 1 STORAGE TANKS l M PLANT T Y v (j VOLUME
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1f 1r .A 8 BORON RECYCLE STORAGE GAS TANKS SHIM 8LEED TAN KS 600 FT.3 A S R PP R Y @ Y / Y ; WASTE GAS SYSTEM C. CONTAINMENT NORMALLY CLOSED UPrFR VOLUME 18.000 cfm
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LOWER VOLUME ? P PURGE OU1 *.ET PURGE INLET A C A P gr 9,400 cfm INTERNATION AL RECIRULATING TRAINS (2) CONTAINMENT VENTILATION SYSTEMS (2) 8,000 cfm EACH "p^y A PE U 4 POTE NTI AL 57.000 cfm - T^ ' " ^ ' " ^ "
- OUTSIDE AREA PER UNIT AIR 34.000 cfm -
-> OTHER AREAS PER UNIT
-UXILf AR Y BUILDING SUPPLY '
FANS AUXILI ARY OUILDING SYSTEM (SHARED) LEGEND: P P R E F I LTE R A HIGH EFFICIENCY PARTICULATE FILTER C CHARCOAL ADSORBER FUEL HANDLING AREA IS NORMALLY UNFILTERED. UPON A RADIATION ALARM BY EMF - 42, THE EXHAUSTWILL BE DIVERTED TO THE FILTERED MODE.
** POTENTIALLY CONTAMINATED AREAS OF THE AUXILIARY BUILDING ARE NORMALLY UNFILTERED. UPON A RADIATION ALARM BY EMF-41.THE EXHAUSTWll L BE DIVERTED TO THE FILTERED MODE.
FIGURE C1.0-2 CATAWBA NUCLEAR STATION GASEOUS RADWASTE SYSTEM DATE COMPILED 8/3/83
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 Thare are two potential release points at Catawba. They are as follows:
- 1. Liquid Waste Effluent Discharge Line
- 2. Conventional Waste Water Treatment System Effluent Line C2.1.1 Liquid Waste Effluent Discharge Line 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 flow rate monitor has a variable setpoint which terminates the release by closing the isolation valve, I WL124 should the dilution flow fall below the setpoint. .The following equation shall be used to calculate a discharge flow, in gpm.
F n ~ f<F RL i# I I i=1 MPC.1
.' ~
where: f = the undiluted effluent flow, in gpm. F RL = actual 1 W Pressure service water flowrate, in gpm, from the sum of the flowrate monitors located in the Control Room. o = the recirculation factor at equilibrim (dimensionlesn),1.027. cfs .027 a=1+OR = 1 + 4400 cfs H where: Qg = average dilution flow (120 Cfs) QH = average fl W Past Wylie Dam (4400 cfs) C. 1
= the concentration of radionuclide, i, in undiluted effluent as determined by laboratory analyses, in pCi/ml.
MPC. l'
= the concentration of radionuclide, i, from 10CFR20, Appendix B, Table II, Column 2. If radionuclide, i, is a dissolved noble gas, the MPC.1
= 2.0E-04 pCi/ml.
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e .. C2.1.2 , Conventional Waste Water Treatment System Effluent Line The conv.entional waste water treatment system effluent is normally considered nonradioactive; that is, it is unlikely 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. These three sources of water that are normally discharged into the conventional waste water system will have their flow diverted if they become radioactive.
- a. Containment Ventilation Unit Condensate Effluent Line Normally the containment ventilation unit condensate effluent line would discharge into the. Turbine Building sump, but if radiation is detected above background, the discharge will be-terminated and an alarm actuated.
The containment ventilation unit condensate tank will then be recirculated, sampled and then discharged through the liquid waste effluent line and monitored or processed thru the WL system.
- b. Clean Area Floor Drain Pump Sumps Normally the discharge line coming from these sumps will discharge into <
the Turbine Building sump, but if radiation is detected above background, the discharge flow will automatically be routed ~to the floor drain tank for processing and later be discharged through the liquid waste effluent line.
- c. Turbine Building Susp Discharge Line Normally the discharge from the Turbine Building sump will go into the conventional waste water treatment system, but if radiation is detected above background, the sump pumps A, B, and C will stop and an alarm actuated. The Turbine Building sump discharge line can either be routed to the floor drain tank.for processing or routed directly to the liquid waste effluent' discharge line.
C2.2 GASE0US RELEASE RATE CALCULATIONS
'The unit veat 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.
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 /trir setpoints in con-junction with administrative controls assure that release limits are not 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 release rates calculated in C2.2.1 and C2.2.2 shall control the release rate for a single release point.
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C2.2.1 Noble Cases 3Kg [(X/Q)D] g < 500 mrem /yr, and 1 3(Lg + 1.1 Mg ) [(X/Q)Q1 ) < 3000 mrem /yr 1 where the terms are defined below. C2.2.2 Radiciodines, Particulates, and Other Radionuclides With T 1/2 > 8 Days s 3Pg [W Qg ] < 1500 mrem /yr 1 where: K. = The total body dose factor due to gamma emissions for each identified noble gas radionuclide, in mrem /yr per pCi/m3 from Table 1.2-1. L.1
= The skin dose factor due to beta emissions for each identified noble gas radionuclide, in mrem /yr per pCi/m3 from Table 1.2-1.
M. = The air dose factor due to gamma emissions for each identified noble 1 gas radionuclide, in mrad /yr per pCi/m3 from Table 1.2-1 (unit conver-sion constant of 1.1 mrem / mrad converts air dose to skin dose). P. = The dose parameter for radionuclides other than noble gases for the 1 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).
-)g = The release rate of radionuclides, i, in gaseous effluent from all release points at the site, in pCi/sec.
3 (X/Q) = 2.60E-07 sec/m . The highest calculated annual average relative concen-tration for any area at or beyond the unrestricted area boundary. W = The highest calculated annual average dispersion parameter for estimating the dose to an individual at the controlling location: 3 W = 1.00E-07 sec/m , for the inhalation pathway. The location is the unrestricted area in the S sector. W = 1.70E-09 meter 2, for the food and ground plane pathways. The location is the unrestricted area boundary in the S sector (nearest residence, and vegetable garden). C-6
y d.
' .b.=..kiC.f 1 1.+ k2 = 4.72E+2C.f' 1 where:
' C: g = the~ concentration of radionuclide, i, in undiluted gaseous eftluent, in pCi/ml .
,. - f. = the undiluted effluent flow, in efm !
4 kt = conversion factor, 2.83E4 ml/ft3 i;
- k' 2 = conversion factor, 6El sec/ min-b i
l" . I + y-h f
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, serm-.,eerr- ,ww , , . + . ,--,ew e, .w,w r-= =-+e---m-*-vere,,--rem-s e e'--- w v w*v-.. -t-,w w - eg v weer**,v - --*< +<ge w ev 4 5 4 -,e -e e-w e - ~ ~* +e m mret 1= w+e et'+er#-ev--**"
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t C3.0 RADIATION MONITOR SETPOINTS ., Using .the generic calculations presented in Section 2.0, radiation monitoring setpoints are calculated for monitoring as required by the Technical Specifications. , All radiation monitors for Catawba are off-line. These monitors alarm on low flow; the minimum -flow ' alarm level for the-liquid monitors is 2 gallons per minute and for the gas monitors is 6 standard cubic feet per minute. These monitors measure the activity in the liquid or gas volume exposed to the detector and are independent of flow rate if a minimum flow rate is assured. c
. Radiation monitoring setpoints calculated in the following sections are expressed in activity concentrations; in reality the monitor readout is in counts per
. minute. The relationship between concentration and counts per minute is estab-lished by a station procedure using the following relationship:
c = 2.22 x 106eV where: c = the gross activity, in pCi/ml r =-the count rate, in cpm 12,22 x 108 = the disintegration per minute per pCi e = the counting efficiency, cpm /dpm ' V = the volume of fluid. exposed to the detector, in ml. 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 liquid effluent, the release.is controlled by limiting the flow rate of 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 that used to calculate the release rate. This setpoint is:
c$"oE $ 2.48E-05 pCi/ml where: i c = the gross. activity in undiluted effluent, in pCi/ml f = the._ flow from the tank may vary from 0-100 gpm but, for this calculation, is assumed to be 100 gpm. MPC = 1.0E-07 pCi/ml, the NPC for an unidentified mixture-a = 1.027 (See Section C2.1.1) ?. C-8 t, ' *
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w * - F = the dilution flow may vary as described in section C2.1.1, but is con'servatively estimated at 25,500 gpm, the minimum flow available. C3.1.2 Containment Ventilation Unit Condensate Effluent Line As described in Section C2.1.2, on release rate calculations for the containment ventilation unit condensate effluent, it is possible but unlikely that the - effluent will contain measurable activity above backgreund. It is assumed that no activity is present in the effluent until indicated by radiation monitoring.
.Since the tank contents are discharged automatically, a maximum tank concentration, which also is the radiation monitor setpoint, is calculated to assure that release limits are not exceeded. The monitor setpoint and maximum tank concen-tration-is:
MPC x F c$ g $ 4.97E-05 pCi/ml where: c = the gross activity in undiluted effluent, in pCi/ml
=
f the flow from the tank may vary from 0-50 gpm but, for this calculation, is assumed to be 50 gpm MPC = 1.0E-07 pCi/ml, the MPC for an unidentified mixture o = 1.027 (see Section C2.1.1) F = the dilution flow may vary as described in C2.1.1, but is con-servatively estimated at 25,500 gpm, the minimum flow available. C3.1.3 Feedwater Pump Sumps As described in Section C2.1.2, release rate calculations for the feedwater pump sumps and floor drain tank effluents, it is possible but unlik e ly tha t the effluents will contain measu-able' activity above background. It is assumed that_no activity is present in the effluent until indicated by radiation monitoring. Since.the sumps are discharged automatically, a maximum sump concentration, which is also the radiation monitor setpoint, is calculated as follows to assure that release limits are not exceeded: c5 gf $ 2.48E-05 pCi/ml
- where:
c = the gross activity-in undiluted effluent, in pCi/ml
- f. = the flow from the sumps may vary from 0-100 gpm but, is assumed to be 100 gpm.
MPC = -1.0E-07 pCi/ml, the HPC for an unidentified mixture o = 1.027 (see Section C2.1.1) C-9 < e i ~+- - y . ----m -er-4 , - - , --,e , --~ - . , . - v -
-,e m-- - , .s , - - - , - - -,,r-~, ee- e - ,, -- -w. -
F
= the dilution flow may vary as described in Section C2.1.1, but is con-servatively estimated at 25,500 gpm, the minimum flow available.
C3.1.4 Turbine Building Sump Discharge Line As described in Esction C2.1.2, release rate calculations for the turbine building sumps, it is possible but unlikely that the effluents will contain measurable activity above background. It is assumed that no activity is present in the effluent until indicated by radiation monitoring- Since the sumps are discharged automatically, a maximum sump concentration, which is also the radiation monitor setpoint, is calculated as follows to assure tha t release limits are not exceeded: c $ " gf $ 1.84E-06, pCi/ml where: c' = the gross activity in undiluted effluent, in pCi/mi f = the flow rate of the sumps may vary from 0-1350 gpm, but for this calculation, is assumed to be 1350 gpm. MPC = 1.0E-07 pCi/ml, the MPC for an unidentified mixture o = 1.027 (see Section C2.1.1) F the dilution flow may vary as described in Section C2.1.1, but is con-servatively estimated at 25,500 gpm, the minimum flow available. A C-10 r--- --y'-- -
C3.2 GAS MONITORS The following equation shall be used to calculate noble gas radiation monitor setpoints based on Xe-133: K(X/Q)Qg < 500 (see Section C2.2.1) kg= 4.72E+02 Cg f (see Section C2.2.2) Cg < 1.39E+04/f where: C f = the gross activity in undiluted effluent, in pCi/ml f = the flow from the tank or building and varies for various release sourcCS, in Cfm K = from Table 1.2-1 for Xe-133, 2.94E+2 mrem /yr per pCi/m3 3 (X/Q) = 2.60E-07 sec/m , the highest calculated annual average relative concen-tration for any area at or beyond the unrestricted area boundary for long term releases. 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 ejecto'r, and auxiliary building ventilation. Since all of these releases are through the unit vent, the radiation monitor on the unit vent may be used to assure that station release limits are not exceeded. For releases from the containment purge ventilation system, the radiation ponitor setpoint may be: C1 < 1.39E+04/f = 7.28E-02 where:
~
f = 151,000 cfm (auxiliary building ventilation) + 40,000 cfm (containment purge) = 191,000 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, the radiation monitor setpoint may be: C.1 < 1.39E+04/f = 9.21E-02 where: f = 151,000 cfm (auxiliary building ventilation) C-11
C4.0 DOSE CALCULATIONS C4.1 FRZ9UENCY OF CALCULATIONS Dose contributiens to the. maximum exposed individual shall be calculateo 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 shall 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, one of the two following cases will apply: 1. i' If the radionuclides Co-58 and/or Co-60 have been detected and Cs-134 and/or Cs-137 have not been detected (i.e., plants without I any fuel failure) dose calculations indicate that the maximum exposed individual would be a child who consumed fish caught in the discharge canal and who drank water from the nearest " downstream" potable " water intake. The dose from these two radionuclides has been cal-culated to be 27% of that individual's total body dose. 3
- 2. If the radionuclides Cs-134 and/or Cs-137'have been detected, dose calculations 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 these two radionuclides has been calculated to be 90% of that individual's total body dose.
C4.2.2 Gaseous Effluents TC4.2.2.1 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 on.the site boundary in each meteorological sectors. C4.2.2.2 Radioiodines, Particulates, and Other Radionuclides T 1/2 > 8 days s For dose contributions from radioiodines, particulates and other radionuclides; it is- assumed that the maximum exposed individual is an infant who breathes the air.and consumes milk from the nearest goat or cow in each meteorological sector. C-12
C4.3 SINPLIFIED DOSE ESTIMATE C4.3.1 Liquid Effluents For dose-estimates, two simplified calculations using the assumptions presented in Section C4.2.1 and source terms presented in the FSAR are presented. Once operational source term data is available, this information shall be used to revise these calculations, if necessary. Case 1 - No Cs-134 or Cs-137 present in effluent. m D WB = .36E+03 I (F g)(T )g (CCo-60 + 0.35 CCo-58) A=1 where: 2.36E+03 = 1.14E+05 (Uaw/Dw +U af BF.) 1 DFat.t (3.70) where: 6 1.14E+05 = 10 pCi/pCi x 103 ml/kg + 8760 hr/yr U, = 510 kg/yr, child water consumption D" = 37.7, dilution factor from the near field area to the nearest potable water intake. U,g = 6.9 kg/yr, child fish consumption BF1 = 5.0E+01, bioaccumulation factor for Cobalt (Table 3.1-1) DF,1 = 1.56E-05, child, total body, ingestion dose factor for Co-60 (Table 3.1-4) 3.70 = factor derived from assumption that 27% of dose is from Co-58 and Co-60 or 100% + 27% = 3.70 And where: ,- F A= F+f where: f = liquid radwaste flow, in gpm o = recirculation factor at equilibrium, 1.027 F = dilution flow, in gpm And where: Tg = The length of time, in hours, over which C Co-58' Co-60, nd Fg are averaged.
.C Co-58 = the average concentration of Co-58 in undiluted effluent, in pCi/ml, during the time period considered.
C-13 w .- e . - - - - - --n.,,.g--, --m-- r--, ,r , , , . -- . , - , - ,
. C Co-60'= during the time period considered.the average concentration of Co 0.35_ = The ratio of the child total body ingestion dose factors for Co-58 and Co-60 or 5.51E-06 + 1.56E Table 3.1-4.
Case 2 - Cs-134 and/or Cs-137 present in effluent, m g WB = 6.38E+05 Ig(F )(T ) (CCs-134 + 0.59 CCs-137) A=1 where: 6.38E+05 = 1.14E+05 (Uaw/D +U w af BF.) 1 DFair (1.10) where: 8 1.14E+05 =-10 pci/pci x 103 ml/kg + 8760 hr/yr U,, = 730 kg/yr, adult water consumption D, = 37.7, dilution factor from the near fiald area to the nearest potable
' . water intake.
y U,g = 21.kg/yr, adult fish consumption i BFg = 2.00E+03,. bioaccumulation factor for Cesium (Table 3.1-1) DF,17 = 1.21E-04, adult, total body, ingestion dose factor for Cs-134 (Table 3.1-2) 1.10 = factor derived from the assumption that 90% of dose is from Cs-134 and Cs-137 or 100% + 90% = 1.10 i And where: I" F
- - A = F +-f l-where
l f = liquid radwaste-flow, in gpm l l a = recirculation factor at equilibrium, 1.017 i l-l F.= dilution flow, in'gpm
- And where
Tg = The length of time, in hours, over which CCs-134, CCs-137, nd Fg 1 L are averaged. t CCs-134 '= the average concentration' of Cs-134 in undiluted effluent, in l pCi/ml, during the time period considered. l C-14 L
C Cs-137 = during the time period considered.the average concentration of Cs-137 i 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 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 estimates using the assumptions in C4.2.2.1 and source terms in the FSAR are presented below. .Once operational source term data is available, this information shall be used to revise these calculations, if necessary. These calculations further assume that the annual average dispersion parameter is used and that Xenon-133 contributes 45% of the dose. D
=2.91E-12[D]Xe-133(2.22)
D p=8.65E-12[D]Xe-133(2.22) where: 2.93E-12 = (3.17E-8)(353) (X/Q), derived from equation presented in Section 3.1.2.1. 8.65E-12 = (3.17E-08) (1050) (X/Q), derived from equation presented in Section 3.1.2.1'.
= 2.60E-07 sec/m 3 , as defined in Section C2.2.2 X/Q
[D]Xe-133=thetotalXenon-133activityreleasedinpCi 2.22 = factor derived from the assumption that 45% of the dose is contributed by Xe-133. C4.3.2.2 Radioiodines, Particulates, and Other Radionuclides with T 1/2 ) 8 days For dose estimates, simplified dose estimates using the assumptions in C4.2.2.2 and source terms in the FSAR are presented below. Once operational source term data is available, this information shall be used to revise these calculations, if necessary. These calculations further assume that the annual average dispersion / deposition parameter is used and that 95% of the dose is from Iodine-131 concentrated in goat's milk. The simplified dose estimate to the thyroid of an infant is: D = 2.00Et04 w (Q)g,131 (1.05) where: w = 2.40E-10 = D/Q for food and ground plane pathway, in m2 from Table C4.0-2 for location of nearest real goat (NW sector at 2.5 miles). C-15
(D)I-131=thetotalIodine-131activityreleasedinpCi. 2.00E+04 = (3.17E-08)(Rf[D/Q))withtheappropriatesubstitutonsfor goat'smilkinthegrass-cow-milkpathwayfactor,R[g[D/Q]for Iodine-131. See Section 3.1.2.2. 1.05 = factor derived from the assumption that 95% of the dose is 4 contributed by I-131. C4.3' FUEL CYCLE CALCULATIONS These calculations shall be performed using models presented in Section 3.3. ' Y t 4 v 4 s A [, C-16 4 a- ,rt -w . - - , + , , -- r r e < a -------,-2,-c p-- -- - m--e-..c>,- c - - - - . <--w-v- re,.- ,--r ww-w., vr-- -,,w- v,-.3., ==v----w ,4-v- e e~,w w ,m.-c
4 t ). TABLE'C4.0-1 L(1 of 1) t
. CATAWBA MUCLEAR STATION -
, DISPERSION PARAMETERs (X/Q) FOR LONG TERM RELEASES > 500 HR/YR OR > 125 HR/QTR-Distance to the control location, (miles) Sector 0.5 1.0 1.5 2.5 2.0 3.0 3.5 4.0 4.5- 5.0 N 9.8E-8 9.5E-8 9.2E-8 7.7E-8 6.5E-8 5.8E-8 4.8E-8 4.4E-8 3.9E-S 3.5E-8 NNE ?. 5E-7 2.6E-7 2.5E-7 2.1E 1.7E 1.5E-7 1.3E-7 1.2E-7 9.8E-8 9.1E-8
- NE 2.3E-7 1.9E-7 1.6E-7 1.3E-7 1.1E-7 9.3E-8 7.9E-8 7.0E-8. 6.2E-8 5.7E-8
, ENE 1.2E-7 1.1E-7 9.5E-8 7.8E-8 6.4E-8 5.8E-8 4.8E-8 4.4E-8 3.9E-8 3.5E-8 E 8.3E-8 6.6E-8 6.6E-8 5.6E-8 4.6E-8 4.2E-8 3.5E-8 3.3E-8 2.8E-8 2.6E-8 ! ESE 6.8E-8 6.0E-8 5.1E-8 4.3E-8 3.5E-8 3.3E-8 2.7E-8 2.4E-8 2.2E-8 2.1E-8 SE 6.2E-8 5.3E-8 4.7E"3 3.8E-8 3.2E-8 2.9E-8 2.5E-8 2.4E-8 2.0E-8. 1 1.9E-8 SSE 8.2E-8 8.3E-8 7.5E-8 5.1E-8 5.1E-8. 4.7E-8 3.8E-8 3.7E-8 3.0E-8 2.8E-8 S 1.0E-7 1.1E-7 1.1E-7 8.6E-8 7.2E-8 6.2E-8 5.3E-8 4.9E-8 4.2E-8 3.7E-8 SSW 1.5E-7 1.4E-7 1.3E-7 1.1E-7 8.7E-8 7.6E-8 6.3E-8 5.8E-8 5.0E-8 3.9E-8 e S*J 1.7E-7 1.7E-7 1.6E-7 1.3E-7 1.0E-7 8.5E-8 7.1E-8 6.3E 5.5E-8 4.9E-8 WSW 9.0E-8 9.2E-8 9.1E-8 7.6E-8 6.0E-8 5.1E-8 4.4E-8 3.9E-8 3.4E-8 3.1E-8 W 8.2E-8 8.4E-8 7.9E-8 6.7E-8 5.2E-8 4.4E-8 3.7E-8 3.3E-8 2.9E-8 2.5E-8
- WNW 7.0E-8 6.6E-8 6.1E-8 5.0E-8 4.1E-8 3.4E-8 3.0E-8 2.6E-8 2.4E-8 2.1E-8 NW 8.0E-8 6.8E-8 6.0E-8 5.0E-8 4.1E-8 3.4E-8 3.1E-8 2.7E-8 2.5E-8 2.2E-8 NNW 1.1E-7 9.3E-8 8.3E-8 7.0E-8 5.9E-8 4.8E-8 4.5E-8 3.8E-8 3.6E-8 i 3.2E-8 4
4
- - - - - ,, - , n - -- ,-
1 TABLE C4.0-2 (1 of 1)
. CATAWBA NUCLEAR STATION' DIPERSION PARAMETER *(D/Q) FOR LONG TERM RELEASES > 500 HR/YR OR > 125 HR/QTR 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 1.4E-9 6.38-10 4.8E-10 2.7E-10 1.9E-10 1.6E-10 1.4E-10 1.2E-10 9.5E-11 7.9E-11 NNE 5.1E-9 2.1E-9 'I.5E-9 8.9E-10 7.2E-l'0 ~ 5.4E-10 4.8E-10 3.8E-10 3.5E-10 3.0E-10 ! NE 6.5E-9 2.6E-9 1.8E-9 1.1E-9 7.5E-10 6.0E-10 4.8E-10 4.1E-10 3.3E-10 3.0E-10 ENE 2.3E-9 8.5E-10 5.3E-10 3.5E-10 2.6E-10 2.1E-10 1.9E-10 1.5E-10 1.3E-10 1.1E-10 4 E 2.2E-9 8.1E-10 6.1E-10 3.0E-10 2.1E-10 1.7E-10 1.5E-10 1.1E-10 9.3E-11 8.0E 4 ESE 1.7E-9 5.4E-10 3.3E-10 2.0E-10 1.4E-10 1.1E-10 9.7E-11 7.3E-11 6.1E-11 5.2E-11 SE 1.6E-9 5.9E-10 3.6E-10 2.4E-10 1.6E-10 1.2E-10 1.0E-10 8.0E-11 6.9E-11 5.6E-11 SSE 1.4E-9 5.9E-10 3.9E-10 2.3E-10 1.7E-10 1.4E-10 1.2E-10 9.2E-11 8.CE-11 6.6E-11 S 1.7E-9 8.3E-10 6.6E-10 3.4E-10 2.7E-10 2.0E-10 1.9E-10 1.5E-10 1.4E-10 1.1E-10 SSW 2.5E-9 9.9E-10 7.7E-10 4.2E-10 3.5E-10 2.5E-10 2.4E-10 1.8E-10 1.7E-10 1.3E-10 SW 6.2E-9 2.4E-9 1.5E-9 8.8E-10 6.6E-10 5.0E-10 4.3E-10 3.5E-10 3.0E-10 2.4E-10 WSW 1.4E-9 5.6E-10 5.0E-10 2.5E-10 2.0E-10 1.6E-10 1.5E-10 1.2E-10 1.0E-10 8.9E-11 W 1.6E-9 6.8E-10 4.8E-10 2.8E-10 1.9E-10 1.8E-10 1.4E-10 1.2E-10 1.0E-10 9.1E-11 WNW 1.2E-9 4.9E-10 3.2E-10 2.0E-10~ 1.6E-10 1.2E-10 9.4E-11 7.8E-11 6.9E-11 5.8E-11 NW 1.6E-9 7.3E 4.8E-10 2.9E-10 2.4E-10 1.8E-10 1.5E-10 1.3E-10 1.1E-10 9.2E-11 NNW 2.0E-9 9.0E-10 6.1E-10 3.7E-10 2.5E-10 2.2E-10 1.7E-10 1.6E-10 1.1E-10 1 1.0E-10 1 4 l
TABLE C4.0-3
- i (1 of 3)
CATAWBA NUCLE /.R STATION ADULT A g DOSE PARAMETEtti NUCLIDE- BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LII H 3 0.0 NA 24 4.58E-01 4.58E-01 4.58E-01 4.58E-01 4.58E-01 4.58E-01 CR 51 4.11E+020.0 4.11E+02 4.11E+02 4.11E+02 4.11E+02 4.11E+02 4.11E+02 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 FE 59 0.0 2.56E+02 2.63E+02 1.05E+03 2.46E+03 9.45E+02 0.0 0.0 6.89E+02 8.21E+03 C0 58 0.0 9.08E+01 2.04E+02 0.0 0.0 C0 60 0.0 1.84E+03 0.0 2.61E+02- 5.75E+02 0.0 0.0 NI 63 0.0 4.90E+03 3.14E+04 2.18E+03 1.05E+03 0.0 0.0 0.0 4.54E+02 NI 65 1.28Et02 1.66E+01 7.56E+00 0.0 0.0 0.0 4.20E+C2 CU 64 0.0 1.02E+01 4.77E+00 0.0 2.56E+01 ZN 65 0.0 8.66E+02 2.32E+04 7.38E+04 3.33E+04 0.0 4.93E+04 0.0 4.65E+04 ZN 69 4.93E+01 9.44Et01 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 BR 84 0.0 5.83E+01 0.0 0.0 5.25E+01 0.0 BR 85 0.0 0.0 4.12E-04 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 6.5LE+02 0.0 0.0 0.0 SR 90 3.66E+03 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 Y 90 3.15E+03
-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 Y 92 4.82E+03 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 TABLE C4.0-3 (1 of 3)
,--..-m..,,,,-n . , .
I . . TABLE C4.0-3 (2 of 3) CATAWBA NUCLEAR STATION ADULT A g DOSE PARAMETERS NUCLIDE BONE LIVER T. BODY THYROID KIDNEY. LUNG GI-LII Y 93 1.66E-01 0.0 4.59E-03 0.0
~
0.0 ZR 95 0.0 5.27E+03 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 NB 95 0.0 5.18E-03 0.0 1.06E+03 4.47E+02 2.49E+02 1.34E+02 0.0 2.46E+02 0.0 1.51E+06 MO 99 0.0 1.13E+02 2.14E+01 0.0 2.55E+02 TC 99M 0.0 2.61E+02 9.41E-03 2.66E-02 3.39E-01 0.0 4.04E-01 1.30E-02 1.57E+01 TC 101 9.68E-03 1.40E-02 1.37E-01 0.0 RU 103 4.64E+00 2.51E-01 7.13E-03 4.19E-14 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 RU 106 0.0 2.46E+02 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 ?.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 4.12E403 1.75E+03 3.79E+03 4.61E+04 0.0 5.56E+04 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 3.35E+01 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 1 132 7.74Et00 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 1 134 8.46E+01 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 2.98.E+05 7.09E+05 5.80E+05 0.0 CS 136 2.29E+05 7.62E+04 1.24E+04 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 CS 138 0.0 1.77E+05 5.89E+04 1.01E+04 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 TABLE C4.0-3 (2 of 3)
TABLE C4.0-3 (3 of 3) CATAWBA NUCLEAR STATION ADULT Ag DOSE PARAMETERS NUCLIDE - BONE 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 BA 142 3.90E-04 2.38E-04 2.62E-10 2.51E-01 2.58E-04 1.58E-02 0.0 2.18E-04 1.46E-04 3.54E-19 LA 140 1.55E-01 7.82E-02 2.07E-02 0.0 0.0 LA 142 0.0 5.74E+03 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- 2.25Et00- 9.39E-01 1.21E-01 0.0 5.57E-01 0.0 1.59Et02 PR 143 5.71E-01 2.29E-01 2.83E 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-C1 2.70E-02 0.0 2.64E-01 0.0 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 0.0 6.28E+02 h TABLE C4.0-3 (3 of 3) l t
-e. --,--y p-- , -m - , +r-,++v 4-----+ r.-,- -e-- ,m , -g- -
, - - - r y -
e s ? ,4 8
- C5.0 Radiological Environmental Monitoring
' The Radiologica1' Environmental Monitoring Program shall be conducted in
, accordance with Technical Specification, Section 3/4.12.
The monitocing program locations ~and analyses are given in Tables C5.0-1 through C5.0-3 and Figure C5.0-1. Site specific characteristics make groundwater sampling, special low-level I-131 analyses on drinking water, and ' food product sampling unnecessary. Ground-
~ water recharge is = from precipitation and the groundwater gradient is toward the effluent discharge area; therefore, contamination of groundwater from liquid effluents is highly improbable. Special low-level I-131 analyses in drinking water will not be performed routinely since the expected I-131 dose from this pathway is less than 1 mrem / year. Food products will not be sampled since lakewater irrigation is not practiced in the vicinity.
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 Intercomparsion Studies (crosscheck)
Program, our participation code is CP. L T s a h i J e = 4 C-17 T T v' - P P y a3w-ymye- n- wr w 9y1--mee< r P- pgrwFgs m4%.gw te ev?- avm wge p m w W 'E E g cW5- - ' gre ?1-t*-e4 3++--+-y e
TABLE C5.0-1 (I of I) CATAWBA RADIOLOGICAL MONITORING PROGRAM SAMPLING LOCATIONS , (TLD LOCATIONS) r SAMPLING LOCATION DESCRIPTION SAMPLING LOCATION DESCRIPTION ~ 200 SITE BOUNDARY (0.7M NNE) 232 4-5 MILE RADIUS 201 (4.IM NE) SITE BOUNDARY (0.5M NE) 233 4-5 MILE RADIUS 202 (4.0M ENE) SITE BOUNDARY (0.6M ENE) 234 4.5 MILE RADIUS 203 (4.5M E) SITE BOUNDARY (0.5M SE) 235 4.5 MILE RADIUS 204 (4.UM ESE) SITE BOUNDARY (0.5M SSW) 236 4-5 MILE RADIUS 205 (4.2M SE) SITE BOUNDARY (0.6M SW) 237 4-5 MILE RADIUS 206 (4.8M SSE) SITE BOUNDARY (0.7M WNW) 238 4-5 MILE RADIUS 207 (4.2M S) SITE BOUNDARY (0.8M NNW) 239 4-5 MILE RADIUS (4.6M SSW) 212 SPECIAL INTEREST (2.7M ESE) 240 4-5 MILE RADIUS 217 (4.lM SW) CONTROL (IO.0M SSE) 241 4-5 MILE RADIUS (4.7M 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 BOUNDARY (0.7M ESE) 244 4-5 MILE PADIUS (4.1M NW) 225 SITE BOUNDAR7 (0.5M SSE) 245 4-5 MILE RADIUS (4.2M NNW) 226 SITE BOUNDARY (0.5M S) 246 SPECIAL INTEREST (8.lM ENE) 227 SITE BOUNDARY (0.5M WSW) 247 CONTROL (7.5M ESE) 228 SITE BOUNDARY (0.6M W) 248 SPECIAL INTEREST (8.2M SSE) 229 SITE BOUNDARY (0.9M NW) 249 SPECIAL INTEREST (8.lM 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)
O
- TABLE C5.0-2 (1 of 1)
- CATAWBA RADIOLOGICAL MONITORING PROGRAM SAMPLING LOCATIONS (OTHER SAMPLING LOCATIONS)
CODE: 6 W - Weekly Q - Quarterly SM - Semimonthly SA - Semiannual M - Monthly __________________________@fMPLING_LgC6TigN,g@CR[PIlgN,,___________________________,,,_______,,,,,,___________,,_______ 200 Site Boundary (0.7m NNE) _ W 201 Site Boundary (0.5m NE) W H 205 Site Boundary (0.6m,SW) W 208 Discharge Canal (0.5m S) M SA SA 209 Dairy (7.0m SSW) SM 210 Ebenezer Access (2.4m SE) SA 211 Wylie Dam (4.0m ESE) ft 212 Tega Cay (2.7m ESE) W 213 Fort Mill Water Supp?y (7.5m ESE) M 214 Rock Hill Water Supply (7.3m SSE M 215 Camp Steere-Ilwy 49 (4.lm NNE) Control SA 216 Ifwy 49 Bridge (4.0m NNE) Control M SA 217 Rock liill Substation (10.0m SSE) Control W M ~218 Belm<it Water Supply (13.5m N) Control M 219 Dairy (6.0m SW) SM 220 Dairy (8.0m WSW) SM 221 Dairy (13.0m NW) Control
- Sfl
TABLE C5.0-3 (1 of 1) - CATAWBA RADIOLOGICAL MONITORING PROGRAM ANALYSES ANALYSES SAMPLE MEDIUM ANALYSIS SCHEDULE GAMMA ISOTOPIC TRITIUM LOW LEVEL I-131 GROSS BETA TLD b i
- 1. Air Radioiodine and Particulates Weekly X
- 2. Direct Radiation Quarterly X a
i
- 3. Surface Water Monthly X Quarterly Composite X
- 4. Drinking Water Monthly X X Quarterly Composite X r
- 5. Shoreline Sediment Semiannually X
- 6. Milk Semimonthly X X i
l
- 7. Fish Semiannually X i
- 8. Broadleaf Vegetation Monthly X 1
4-4
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