ML20012A375

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Rev 25 to, Offsite Dose Calculation Manual,Cawtaba Nuclear Station.
ML20012A375
Person / Time
Site: Catawba, McGuire, 05000000
Issue date: 01/01/1990
From: Birch M, Owen T
DUKE POWER CO.
To:
Shared Package
ML15217A104 List:
References
PROC-900101, NUDOCS 9003090346
Download: ML20012A375 (47)
v • d • e


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December 26, 1989

Subject:

Offaite Dose Calculation Manuel-Revision 25-The General Office Radiation Protection Staff is transmitting.to you this  ;

date, Revision 25 - of the Offsite Dose Calculation Manusi. As this revision only affects Catawba Nuclear-Station, the approval of other station managers-is not required. Please-update your copy No. d33 , and discard the affected pages.

REMOVE THESE PAGES INSERT THESE PAGES Table of Contents Rev. 19 Table of Contents Rev. 25 Figure C1.0-1 Rev. 20 Figure C1.0-1 Rev. 25 Figure C5.0-2 Rev. 4 Figure C1.0-2 Rev. 25 (page 1 Of 2) (page 1 Of 2)

L l' _ Figure C5.0-2 Rev. 20 Figure C1.0-2 Rev. 25 (page 2 Of 2) (page 2 Of 2)

C-7 Rev. 24 C-7 Rev. 25 ,

C-12a Rev. 24 C-12a Rev. 25 C-14 Rev. 24 C-14 Rev. 25 C-15 Rev. 24 C-15 Rev. 25 l C-16 Rev. 24 C-16 Rev. 25 P-C-17 Rev. 24 C-17 Rev. 25

/ C-18 Rev. 24 C-18 Rev. 25

\~ C-19 Rev. 24 C-19 Rev. 25 C-20. Rev. 24 C-20 Rev, 25 ,

C 21 Rev, 24 C-21 Rev. 25 C-22 Rev. 24 C-22 Rev. 25 j Tabte C4.0-2 Rev. 4 Table C4.0-2 Rev. 25 i

C-23 Rev. 24 C-23 Rev. 25 Table- 05.0-1 Rev. 24' Table C5.0-1 Rev. 25 L Table C5.0-2 Rev. 24 . Table C5.0-2 Rev. 25 l Figura C5.0-1 orginal Figure C5.0-1 Rev.-25 l' (page 2 Of 2) (page 2 Of 2)

-NOTE: As this letter, with it's attachments, contains "LOEP" information, 13 please insert this in' front of the December 29, 1988 letter.

Approval Date: _12/21/89 Approval Date: I2/22/09 Effective Date: 1/1/90 Effective Date: 1/1/90 g M Owen, Manager

-Mary L.fB rch T B Radiatica Protection Manager Catawba Nuclear Station If-you have any questions concerning Revision 25, please call Jim Stewart

/T at 704) 373-5444

'A cism /H James-M. Stewart, Jr Scientist 9003090346 900228 Radiation Protection PDR ADOCK 05000269 PNU R u

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P C JUSTIFICATlONS FOR REVISION'25 U T a b l e .o'f= C o n t e n t s ' . Changed incorrect' page number from "11":

to."12". 'f Fig'ure C1.0 Replaced figure with' CAD drawing and added additional information-to reflects actual' station operation.

. Figure C5.0-2. Replaced f.igure with CAD drawing.

(page 1 ofL2); .No changes made.

?? , 1 Figure'C5.0-2 Replaced figure with CAD drawing and

[(page 21of;2) added additional information to r e f l e ct-actual station-operation. 4 Page C-7; Updated sections using dose calculattons

-based on 1989-Effluent Release Data (first nine months) and the 1989 Land.

Use Census Data.

Page C-12a' Updated sections using dose' calculations

+ based on 1 9 8 9 ' F,f f t u e n t Release. Data' (first.nine months) and the 1989-Land

,ft i Use Census Data.

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[ 'P a g e s C-.14 ' t h r u C- 2 2 - Updated-sections using dose calculations-

' based on 1989 Effluent-Release Data (first nine months) and the 1989-Land' -

Use Census Data.

' Tab'le C4-.0-2 Changed " Dispersion" to " Deposition"-

l(1 :o f 2)

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, Page:C-23 Updated-the dates.the l a't e s t Land Use. I Census was preformed.

Table C5.0-1 Changed.per attached August 31, 1989 j letter from W P Deal j Table C5.0-2 Changed per attached August 3 '1 , 1989 3 letter from W P Deal

. -l Figure 05,0-1 Changed per attached August 31., 1989 l (2 of 2)- letter from W P Deal j 1

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' INTERSTATION LETTER :

CATAWBA NUCLEAR: STATION g ]

3 01 Birch: Attentions  ;

M.

R. SL.;

. - Jones s

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SUBJECT:

Catawba Nuclear Station .

Radiological Environmental Monitoring Program (REMP)

Nearby Sampling Location Audit .r .

File No.: -CN-778.00-During athe recent 1989 ; CNS Land Use Census, it became apparent that- certain REMP ~

^ sampling locations nearby CNS 'were not' located exactly as -documented -in the Offsite t

. Dose Calculation, Manual ~(ODCM). Subsequent to the census, a thorough station- audit of all sampling locations - within a one mile- radius of CNS was conducted' as a first step .

'44 ' p) .- sini resolving these 4 discrepancies.. This letter documents the results of the audit, and 3 M. details the -revisions to . the 1ODCM and the REMP Sampling- Program which- are now:

' required.'-

e 'In:'orderf to determine. thecexact position of'each REMP sample clocation' nearby- CNS, a 38-L- 6 Einch by 38 : inch aerial. photograph of. the station and . its environs t e was utilized. ' The i

" photograph 'was taken' on ' February 2,~ 1989, and its legend indicates -an " approximate" scale of one' inch equals-400. feet.. This scale was confirmed to be accurate.

. The CNS Final' Safety Analysis Report ,(FSAR Section 2.1.1.1) defines the' station center:

to be the midpoint between the center of each reactor building. It further indicates z that the . Unit 2 Reactor Building lies due North of the Unit 1 Reactor. Building. Using '

this information as a basis, a sixteen- sector grid was drawn' on the photograph. This

serialE photograph, with the meteorological sectors clearly located, was the same one .'

tutilized during the 1989 Land Use Census.

With 1 the assistance ^ of Applied Science Center personnel,= each REMP sampling ' location 2: within'.a one . mile: radius of CNS was visited, and its exact position was identified and

' plotted ' ~on ~ ' the , aerial photograph. The distance and direction of each location from- ,

. the station ~ center was then determined. The results of this survey are given in Attachment 1.

The 5 survey . results of Attachment I reveal discrepancies with the sample location

- ^ >

information given in the ODCH. The differences are indicated on Attachments 2, 3, and

. . . . 4 ' (ODCM Table C5.0-1, Table C5.0-2, and Figure C5.0-1 (2 of 2), respectively). In 7~% - Jaddition to - several minor differences between distances, both groundwater sampling '

ij : locations- ("200" and 252), and three TLD locations (202, 203, and 225), are located in

' .different . sectors than indicated in the ODCM.

Groundwater Location "200" and TLD Locations ' 202, 203, and 225' are located in the sector neighboring the sector indicated in the ODCM.

5

. , . + , _ - . - - ~. I

f CNS REMP Nacrby Scpling Location Audit j Page 2 of 3 l h l q.

V l The ODCM currently indicates that one TLD is located in the general area of the site l boundary in each of the sixteen sectors, as required by CNS Technical Specification l (TS) 3/4.12.1- (Section 1 of Table 3.12-1). According to the best information and )

methods employed at the time these sixteen TLD locations were established, the TLDs - l were thought to be located as described in the ODCM, one in each sector as required.

With the benefit of a thorough survey of all nearby -locations, utilizing a large aerial photograph having a precisely located sector grid, it is now indicated that the

- E and ESE sectors each have two site boundary area TLD locations and the ENE and SSE sectors have - none. This condition will be rectified during the next scheduled TLD l collection occurring September 14, 1989, by deleting the redundant and more distant TLD Locations 202 and 224, and by adding two new TLD locations (255 und 256) in the ENE and SSE sectors, respectively. Applied Science Center personnel have been i notified to incorporate this change into their sampling program. Although this improvement is now necessary to meet TS requirements, the original condition is not considered as noncompliance with the TS, because the best information available at the time was properly used to establish the locations, and no contradictory evidence has been previously identified.

Groundwater. sampling Location "200" (Littlejohn residence) is located in the N sector, while the ODCM indicates that it is in the NNE sector.- Through March 25, 1988, Location "200" groundwater was collected at the Bolick residence, located in the NNE sector 0.10 mile from the Location 200 TLD, air, and broadleaf vegetation sampling sites. When this groundwater sampling location changed from the Bolick residence to

(-p) the Littlejohn residence 0.17 mile away on June 23, 1988, the new site retained the Location "200" designation because it was reported to also be in the NNE sector, approximately 200 feet f rom the Bolick residence. Now that the new survey indicates l

that the Littlejohn residence is actually in the N sector, groundwater sampling For Sample Analysis Report (SAR)

Location "200" will be renumbered to Location 254.

purposes, renumbering will become effective starting with the first quarter sample collected March 22, 1989, so that the location number will be consistent throughout the year. Applied Science Center personnel have been notified to incorporate this change into their sampling program.

L We request that General Office Radiation Protection personnel revise the ODCM as l indicated in Attachments 2, 3, 4, and 5, and submit in the next Radioactive Effluent l_ Release Report all ODCM revision documentation required by CNS TS 6.9.1.7 and 6.14.2.

l' In addition to the sample location corrections and changes, Attachments 2, 3, 4, and 5 '

I contain a few additional corrections and updates discovered during this audit. For example, CNS and Applied Science Center personnel have expressed a strong desire to

, correct the sector grid of Attachment 4 by rotating the grid spokes 11.25 degrees and

l. recentering the grid to coincide with the station center. Due to the extent and j nature of these ODCM revisions, we would be happy to proofread the revised ODCM pages l prior to their distribution.

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n. CNS-REMP N33rby.S::pling Location Auditi l l-

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<- lIf :- you l have <- any: questions . or l comments regardingl this letter, please contact Brian-

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of Chundriik- at .831-5580.' Thank you. i J , 1

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.W. P. Deal. ( -1

Radiation Protection Manager, ,

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4 Ecc h'W. M. Carter p G. L. Courtney :.

C. F. Lan'

.M. D.-Lane-

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i, APPENDIX C i

CATAWBA NUCLEAR STATION t

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L; SITE SPECIFIC INFORMATION i

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r APPENDIX C - TABLE OF CONTENTS

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Pate C1.0' CATAWBA _ NUCLEAR STATION RADWASTE SYSTEMS . . . . . . . C-1 t

C2.0' RELEASE RATE CALCULATION . . . . . . . . . . . . . . . . C-4 C3.0 RADIATION MONI'IT)R SETPOINTS . . . . . . . . . . . . . . C-8 C4.0 DOSE CA14ULATIONS . . . . . . . . . .. . . . . . . . C-12 l C5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING . . . . . . . . . C-23 i:.

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ABBPIVIATIONS

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I Systems:

CM - Condensate System KC Component Cooling  !

NB - Boron Recycle RL - Low Pressure Service Water ,.

RN - Nuclear Service Water System  !

WC~- Conventional Waste Water Treatment WL - Liquid Waste Recycle WP - Turbine Building Sump WS_- Nuclear Solid Waste Disposal .

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's BA - Boric' Acid Tank FDT - Floor Drain Tank l

LHST - Laundry and' Hot Shower Tank MST - Mixing and Settling Tank

'NCDT - Reactor Coolant Drain Tank I RHT - Recycle Holdup Tank L

l. RMT - Recycle Monitor Tank l

l:- e RHWST - Reactor Makeup Water Storage Tank

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SGDT - Steam Generator Drain Tank i VUCDT - Ventilation Unit Condensate Drain Tank >

WDT'- Waste Drain Tank y WEFT - Waste Evaporator Feed Tank L

WMT - Waste Monitor Tank

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/ Table C1.0-1 Rev. 4 <

7/18/84

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75 (J l C1.0 CATAWDA NUCLEAR STATION RADWASTE SYSTEMS  !

l C1.1 LIQUID RADWASTE PROCESSING l I

The liquid radwaste system at Catawba Nuclear Station (CNS) is used to collect j and treat fluid chemical and radiochemical by-products of unit operation. The j system produces effluents which can be reused in the plant or discharged in l j small, dilute quantities to the environment. The means of treatment vary with .

waste type and desired product in the various systems:

I A) Filtration - All waste sources are filtered during processing. In some l cases, such as the Floor Drain Tank (FDT) Subsystem of the Liquid Waste (WL) System, filtration may be the only treatment required. l l

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 f~N Evaporator.In the WL System and the Boron Recycle Evaporator in the Boron i d 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 ND Evaporator, the '

dilute boron is concentrated to 4% so that it may be reused for makeup to

Figure C1.0-1 is a schematic representation of the liquid radwaste system at ,

Catawba. ,

V C-1 Rev. 12 ,

9/19/86

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) C1.2 GASEOUS RADWASTE SYSTEMS v

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 and subsequent decay of the gaseous  :

fission products generated from the reactor coolant system via the chemical and I 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 auxiliary building ventilation system, the secondary system air ejector and 6ecayed WG Tanks. With respect to purging operations, the potential c.ontamination is expected to arise from uncollectible 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 uso 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 N3 evaporator produced during normal operation with the gas collected during the shutdown degasification (high percentago of nitrogen) and will cycle it through V- 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 recombiner 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 l or inside the auxiliary building will generate gaseous activity. Gases result-l ing from leakage inside the containment will be contained until the containment l- air is released through the VQ or VP system. The containment atmosphere will

be discharged through a charcoal absorber and a particulate filter prior to L release to the atmosphere.

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 auxiliary building is normally unfiltered. However, on a radiation monitor alarm, the exhaust is passed through charcoal absorbers to reduce releases to the atmosphere.

C1.2.3 Secondary Systems O

t i Normally, condensate flow and steam generator blowdown will go parallel G

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.through 4 of the 5 condensate polishing domineralizers to remove activity and ,

k,/.V. harmful' ions from the water. Noncondensable gases will be taken from the F -

secondary system by the. condenser. steam air ejector and are passed through a  !

radiation monitor to the unit vent. )

i Figure C1.0-2 is a schematic representation of the gaseous radwaste system at ,

-Catawba.  ;

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UNITS 1 & 2 Roor WNTS l.600.200 cfm L

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V t I s,. 6.i qj qj ,xU,e --.........................a e I CONTROL TANK &

, REACTOR ~ REA TOR A A A A A A CAS CAS TANKS BORON RECYCLE STORAGE 600 FT3 SHlW8LEED TANKS 100 poig EVAPORATOR y y y y y y CAS STRMfR 9 WASTE GAS SYSTEW l EW*60 l

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P A C CONTAINWENT AIR RELEASE a

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CONTAINWENT VENTILATION SYSEWS (2)

INTERNATL. RECIRCU. TRAINS (2) 4.000 efm E ACH l M -42l .

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AUXIUARY 80tLDING SUPPLY OTHER AREAS pga UNIT d FANS AUXILIARY BUILDING $YSTEM ($HAREp) LICEND:

P = PRESTER Ae HICM. EFFICIENCY PARTICULATE FILTER C e CHARCOAL ABSORBER

  • FUEL MANDUNO AREA l$ NORMALLY UN4TERED. UPON A RADIATION ALARM BY CWF-42. THE EXNAUST wiLL BE DIVERTED TO THE FILTER WODE, e* POTENTIALLY CONTAWINATED AREAS OF THE AUXILlARY BulLDING ARE NORMALLY UNTILERED, UPON A R ADIATION ALARM BY EWF-41. THE EXHAUST WLL BE DlWRTED TO THE FILTERED WODE.

FKlIURE C1.0 2 REVISION 4

/ ,g cATAwsA NUCLEAR STATION CASE 0VS R ADWASTE SYSTEM 7/18/84 i  ; PAGE i or 2 N89121C

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AUXILIARY MONITOR TANK BUILDING r---------------- EMF 58 7

l l l 1 l PARTICULATE & - -

NOBLE CAS l HEPA AND l IODINE SAMPLER MONITOR l l

CAR 8ON FILTER l- l I PROCESS AREAS "

L- 7 rJ HEPA AND l l CARBON FILTER l l I I I FLOW U i

EXHAUST f

BUILDING VENTILATION SYSTEM e I TOTALIZER I 11,000 cfm LA8 HOODS I L____________J GASEOUS WASTE SYSTEM REVISION 25 CATAWBA NUCLEAR STATION 1/1/90 FIGURE C1.0-2 PAGE 2 OF 2 N89121E

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[ F C2.0 RELEASE RATE CALCULATION I

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  • Generic release rate calculations are presented in Section 1.0; these calcula-tions will be used to calculate release rates for Catawba Nuclear Station. l C2.1 LIQUID RELEASE RATE CALCULATIONS There are two potential release points at Catawba. They are as follows:

i

1. Liquid Waste Effluent Discharge Line (WL) '
2. Conventional Waste Water Treatment System Effluent Line (WC)  ;

C2.1.1 Liquid Waste Effluent Discharme Line (WL)  ;

a There are.three low-pressure service water pumps with a minimum flow rate of 16,500 spo each and four nuclear service water pumps with a minimum flow rate ,.

.of 9,000 spo 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 an isolation valve should the dilution _ flow fall below the setpoint. Releases can either be made via EMF-49 which uses isola- '

tion valve IWL124, or EMF-57 which uses isolation valve IWLX28. The following is a typical equation which can be used to calculate a discharge flow, in gpm.  ;

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f5FRL'l I*l I i=1 MPC g

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f = the undiluted effluent flow, in spm.

F " actual I W pressure service water flowrate, in gpm.

RL o = the recirculation factor at equilibrium (dimensionless), 1.027. ,

,,34 Qg ,34 120 cfs = 1.027 4400 cfs >

QH whero:

Q = average diluti n fl w , 20 cis)

R Qg = average flow past Wylie Dam (4400 cfs)

Cg ** the concentration of radionuclide, i, in undiluted effluent as determined by laboratory analyses, in pCi/ml. .

MPC g

= the concentration of radionuclide, i, from 10CFR20, Appendix B, Table II, Column 2. If radionuclide, i, is a dissolved noble gas, the MPCg = 2.0E-04 pCi/ml.

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- () _ X = factor used to reduce the WL flowrate (f) to allow the WC system to release simultaneously. .For example, 0.9 would allow 10% of the stations releases to be WC.

C-4 Rev. 20 7/1/88

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C2.1.2 Conventional Vaste Water Treatment System Effluent Line (WC)

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The conventional waste. water treatment system effluent is potentially radio- ,

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 periodic analysis of the composite sample collected on that ,

line. The water sources listed below that are normally discharged via the l conventional waste water treatment system and/or the Turbine Building Sump will  !

be diverted if they will cause the WC discharge 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 coulc 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 wasta effluent line and monitored.
b. Auxiliary Feedwater Sump pumps and Floor Drain Sump pump Line i

Normally the discharge line coming from these sumps will discharge into the Turbine Building sump, but if activity is detected above its 7s monitor's setpoint, the discharge flow will automatically be routed to the i 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 administrative 1y controlled to assure that relecse 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 abovo its monitor's setpoint, each blowdown flow control valve, the atmospheric vent, and the valve to the Turbine Building Sump will cloze, thus , r terminating the discharge. Blowdown can only be continued by venting the steam to "D" heater and pumping the liquid to the condensate system.
d. Turbine Building Sump Discharge Line 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 line, or allowed to continue being dis-charged via the circuit v4th proper administrative controls implemented to

' assure that release limits are not exceeded.

I v

C-5 Rev. 19 1/2/38

F 7 C2.2 GASEOUS RELEASE RATE CALCULATIONS

l

.The unit vent is the release point for waste gas decay tanks, containment air releases, the condenser air ejector, and auxiliary building ventilation. Tite

> ' condenser air ejector effluent is normally considered nonradioactive; that is, ,

it is unlikely the effluent will contain nessurable activity above background. '

It is assumed that no activity is present in the effluent until indicated by l 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 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 radiododines, particulates and other radionuclides with half-lives greater than 8 days; the most conservatit'e of release 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 r

IK g [(X/Q)Qg ] < 500 mrem /yr, and i

I (Lg + 1.1 Hg ) [(X/Q)h] g< 3000 mrem /yr i

f[]-

where the terms are defined below.

%j C2.2.2 Radiofodines. Particulates, and Other Radionuclides With T 1/2 ) 8 Days IPg [W Qg ] 4 1500 mrem /yr 1

where K

g

= The total body dose factor due to gamma emissions for each identified noble gas radionuclide, in mrem /yr per pC1/m fr?m Tablo 1,2-1.

8 l

r

= The skin dose factor due to beta simissions for each identified noble L

I gas radionuclide, in mrem /yr per pC1/m' from Table 1.2-i.

l

Mg = The air dose factor due to gamma emissions for each identified noble l- gas radionuclide, in mrad /yr per UC1/m' from Table 1.2-1 (unit conver-sion constant of 1.1 mrem / mrad converts air dose to skin dose).

P g

= The dose parameter for radionuclides other than noble gases for the inhalation pathway, in mrem /yr per pCi/m* and for the food and ground plane pathways in m8 -(mrem /yr) per pC1/sec from Table 1.2-2. The dose L factors are based on the critical individual organ and most restrictive j age group (child or infant).

l~ /~N s k) l

.Q = The release rate of radionuclides.-1, in gaseous effluent from all release points at the site, in pCi/sec.

L C-6 Rev. 12 0

9/19/86

l f 1 7)

(X/Q) = 3.10E-05 sec/m . The highest calculated annual average relative 8

concentration (dispersion parameter) for any area at or beyond i the unrestricted area boundary. The location is the NNE sector

@ 0.5 miles. l i

W = The highest calculated annual average dispersion parameter for estimating the dose to an individual at a location in the unrestricted m 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 = 3.0E-05 sec/m , for'the inhalation pathway. The location is 8

the NE sector @ 0.5 miles.  ;

W = 1.1E-07 meter *, for the food and ground plane pathways. The location is the NE sector @ 0.5 miles.

Qg = k:Cg f + ks = 4.72E+2Cg f where:

Cg = the concentration of radionuclide, i, in undiluted gaseous effluent, in WC1/ml.

p '

i f = the undiluted effluent flow, in efm (J

k i

= conversion factor, 2.83E4 ml/ft 8 i

, ks = conversion factor, 6El sec/ min -

1 l<

l >

it i} '

l l

l 1

y~

[gf

,s

'A C-7 Rev. 25 1/1/90

^

,N C3.0 RADIATION MONITOR SETPOINTS i  !

.Using the generic calculations-presented in.Section 2.0, final effluent radiation monitoring setpoints are calculated for monitoring as requitu 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 monitore m e dependent on pressure or vacuum. Particulate monitors are dependent n 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

. relationship:

r

" " 2.22 x 10'e V wherst

[]

k_/

c = the gross activity, in pC1/ml r = the count rate, in cpm 2.22 x 10' = the disintegration per minute per 901 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 setpointo will N 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 - EMF 49 and EMF 57 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 aaalyses and used to calculate the release rate. A typical radiation monitor setpoint may be calculated as follows:

c5 5 2.48E-05 pCi/ml n where:

~

c = the gross activity in undiluted effluent, in pCi/ml C-8 Rev. 20 7/1/88

i i

(N f = the flow f rom the tank may vary from 0-100 gpm but, for this calculation,

-( )

is assumed to be 100 gpm.

MPC = 1.0E-07 pCi/m1, 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 conservatively estimated at 25,500 gpm, the minimum flow available.

i Normally, discharges from the WL system will be limited to either EMF-49 or I EMF-57. Simultaneous releases may occur, however, if proper station procedures are followed to insure that instantaneous concentration limits will not be .

exceeded. >

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 alarm. 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, j, 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 pC1/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 after 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 Blowdown 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 discharged into the Turbine Building sump. The setpoint may be changed after detection to allow positive control of effluent releases using the guidance given in Section C3.1.5.

O V

C-9 Rev. 20 7/1/88

{

C3.1.5 Turbine Building Sump Discharte Line - EMF 31 f ')

Li 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 yCi/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:

c5 ,) $ 1.42E-06 pCf/ml where:

c = the gross activity in undiluted effluent, in pCi/mi f =-the undiluted effluent flow in som; for this example the flow is from the Turbine ^ Building Sumps and is assumed to be 250,000 gallons / day or e175 gpm.

MPC = 1.0E-07 pCi/ml, the MPC for an unidentified mixture o = 1.027 (See rection C2.1.1) f' F = the dilution flow, in gpm, available to dilute the undiluted effluent Q') 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

h 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 predominant l isotope):

-~

K(X/Q)Qg < 500 (see Section C2.2.1)

Q = 4.72E+02 Cg f (see Section C2.2.2)

Cg -< 116/f where:

Og = the grou activity in undiluted effluent, in WC1/ml 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 pC1/m 8 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.

A)'

( For releases from the containment purse ventilation system, a typical radiation monitor setpoint may be calculated as follows:

Cg < 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:

Cg < 116/f = 7.7E-04 where:

f = 151,000 cfm (auxiliary building ventilation) in

.%/ )

s .'

Rev. 12 C-11 9/19/86-

i i

i' ,

1 l

4 I

i ~

C4.0 DCSE CALCULATIONS

(} ~

C4.1 FREQUENCY OF CALCULATIONS L Dose contributions to the maximum individual shall be calculated at least i every 31 days, quarterly, semiannually, and annually (or as required by Tech- {

nical Specifications) using the methodology in the generic information sections.

This methodology shall also be used for any special reports. Dose calculations that are required for individual pre-release calculations, and/or abnormal releases shall not be calculated by using the simplified done calculations. ,

Station dose projections for these types and others that are known to vary from ,

the station historical averages shall be calculated by using the methodology in 6 the generic information section. STATION Dose projections may be performed using simplified dose estimates.

Fuel cycle dose calculations shall be performed annually or as required by special reports. Dose contributions shall be calculated using the methodology in the appropriate generic information sections, i'

C4.2 DOSE N0DELS FOR MAX 1 MUM EXPOSED INDIVIDUAL '

C4.2.1 Liauid Effluents For dose contributions from liquid radioactive releases, dose calculations -

based on operational source term data and NUREG-0133 guidance indicate that the

~_s 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

\m / intake. The dose from Cs-134 and Cs-137 has been calculated to be 89% 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 gar doso is determined to be a maximum.

C4.2.2.2 Radioiodines, Particulates, and Other Radionuclides T.1/2 > 8 days For dose contributions from radiofodines, 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, f

V C-12 Rev. 25 1/1/90

W , 4

. C4.3 . SIMPLIFIED DOSE EST1MATI

(' l C4.3.1 Liauid 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. 9 +05 I (F )(T g )g(CCs-134 + 0.59 CCs-137) t=1 wherst 6.49E+05 = 1.14E+05 (0,,/D, + U,g BFg) DF ait II*I )

where:

1.14E+05 = 10'pci/yCi x 10'al/kg + 8760 hr/yr U,,.= 730 t/yr, adult water consumption D" = 37.7, dilution factor from the near field area to the nearest potable water intake.

7-_z U,g = 21 kg/yr, adult fish consumption i

k- BFg -= 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 l 3.1-2) 1.12 = factorilerived from the assumption that 89% of dose is from Cs-134 and Cs-137 or 100% 6 89% = 1.12 And wheret y&

.A F+f where f'= 11guld radwaste flow, in gpm o =_ recirculation factor at equilibrium, 1.027 (see Section C2.1.1)

F = dilution flow, in gpm

.t O i

\.)

C-12a Rev. 25 1/1/90 1

.. - . _ . . - _ _ _ . _ . . . _ _ _ _ . _ _ _ _ ______.__m

And where:

V; i

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 pC1/m1, during the time period considered.

C = the average concentration of Cs-137 in undiluted effluent, in

    • I pCi/m1, during the time period considered.

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

.\j i

l l'

'O

'( -

t C-13 Rev. 24 1/1/89

t I

L h

I l ( 'j ,

\

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 assumptions 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 calculatiens further assume that the annual average dispersion parameter is used and that Xenon-133 contributes 82% of the gamma air dose and 94% of the beta air dose.

Dy = 3.47E-10 [Q]Xe-133 (1.22)

Dg = 1.03E-09 [Q]Xe-133 (1.06) 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.

js -

= 3.1E-05 sec/m 8 , as defined in Section C2.2.2 i ) X/Q V

[Q]Xe-133 = the_ total Xenon-133 activity released in 901 1.22 = factor derived from the assumption that 82% of the gamma air dose is contributed by Xe-133.

1.06 = factor derived from the assumption that 94% of the beta air dose is contributed by Xe-133.

C4.3.2.2 Radiciodines, Particulates, and Other Radionuclides with T 1/2 > 8 days For done 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 dispersion / deposition parameters are used and that 66% of the dose results  !

frou H-3 ingested by the maximally exposed individual via the vegetable garden

-pathway at the controlling location. The simplified dose estimate for exposure to the thyroid of a child 13: ,

D = 1.28E-04 w (Q)3g.3 (1.M) where:

w = 3.0E-05 = X/Q for vegetable garden pathway, in sec/m from 8

Table C4.0-1 (n'-} for the controlling location (NE sector at 0.5 miles).

C-14 Rev. 25 1/1/90

7 e ,

b

/^

'v e s

(Q)H-3 = the total Tritium activity released in 901. ,

1.28E-04 = (3.17E-08)(R [X/Q)) with the appropriate substitutions for child vegetable pathway factor, R [X[d]forH-3.

See Section 3.1.2.2.

1.99 = factor derived from the assumption that 50% o'f the total l '

inhalation, food and ground plane pathway dose to the maximally exposed individual is contributed by H-3 via the vegetable garden  ;

pathway. j C4.4 FUEL CYCLE CALCULATIONS As dimussed in Section 3.3.5, more than one nuclear power station. site may con; rib:.te to the doses to be considered in accordance with 40CFR190. The fuel i 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

  • t N, Dy3(T) = DWBII)+DWB(I m ) c+ DWB(8 m) + DWBI8c)

(V ~ I8c)

DM0(T) = DM0(I m ) + DM0(Ic ) + DM0(8m ) + DMO where:

DWB(T) = Total estimated fuel cycle whole body dose commitment resulting from the combined liquid and gaseous effluents of Catawba and McGuire during the calendar 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 ,

i and McGuire during the calendar year of interest, in mrem. ,

L C4.4.1 Liould 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 Liould Contributions Based on operational history, the Catawba fuel cycle maximum exposed individual ,

whole body dose ros.ulting from McGuire liquid effluent releases (D WB I m)) is estimated using the simplified dose calculation given below: .

Dy3(1,) = (6.78E+05) ( Fg)(Tg ) (CCs-134 + 0.59 CCs-137 )

U C-15 Rev. 25 1/1/90

,.P-where

-( )

6.78E+05 = 1.14E+05 ( U,,+ U,7 x BFg ) ( DF,gt ) ( 1.15 ) l where:

1.14E+05 = ( 1.0E+06 pCi/uCi x 1.0E+03 ml/kg ) / ( 8760 hr/yr )

U,, = 730 t/yr, Adult water consumption U,g = 21 kg/yr, Adult fish consumption BFg = 2.00E+03, Bioaccumulation factor for Cesium (Table 3.1-1)

DF = 1.21E-04, Adult total body ingestion dose factor for ait Cs-134 (Table 3.1-2) 1.15 = Factor derived from the assumption that 87% of the dose is derived

-from Cs 134 and Cs-137 or 100% / 87% = 1.15 where Fg =f/F where:

px

( f = McGuire's liquid radwaste flow, in gpm F = 1.97E+06 spe, 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 CCs-134 , CCs-137 and Fg are averaged.

CCs-134 = The average concentration of Cs-134 in McGuire's undiluted effluent, in uC1/m1, during the calendar year of interest.

C Cs-137 = The average concentration of Cs-137 in McGuire's undiluted effluent, in uC1/m1, during the calendar 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 maxitsum organ dono (teen liver) resulting from McGuire's liquid effluent l releases (DMO CIm)) is estimated using the simplified dose calculation given below:

.D MOSIm) = ( 8.16E+05) ( Pg)(Tg ) (CCs-134 + .76 CCs-137)

(3 whore:

I \s]

U,,+ U,7 x BFg ) (DFait) (

8.16E+05 = ( 1.14E+05 ( * )

C-16 Rev. 25 1/1/90 e

.f -

(

i f ') where:

1.14E+05 = ( 1.0E+06 pCi/uci x 1.0E+03 ml/kg ) / ( 8760 hr/yr )

U,, = 730 t/yr, teen water consumption ,

U,f = 16 kg/yr, teen fish consumption ,

BFg = 2.00E+03, Bioaccumulation factor for Cs (Table 3.1-1) .

f:

DF"It = 1.97E-04, teen liver ingestion dose factor for Cs-134 (Table 3.1-3) 1.11 = Factor derived from the assumption that 90% of the teen liver dose is from Cs-134 and Cs-137 or 100% / 90% = 1.11 where:

p Fg =f/F where:

f = McGuire's liquid radwaste flow, in gpm F = 1.97E+06 gpm, the average flow past Lake Wylie Dam (3

i ) where:

%J T = 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 and .

t Cs-134, Cs-137 F g are averaged.

CCs-134 = The average concentration of Cs-134 in McGuire undiluted effluent, in uC1/m1, during the calendar year of interest.

CCs-137 = The average concentration of Cs-137 in McGuire undiluted effluent, in uC1/ml, during the calendar year of interest, i l

1 i 0.76 = The ratio of the teen liver ingestion dose factors for Cs-134 and Cs-137 or 1.49E-04/1.97E-04 = 0.76.

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 c(I )) is estimated using the simplified dose calculation given below:

L DWB(I c ) = ( 6.49E+05 ) ( Fg)(Tg ) (CCs-134 + 0.59 CCs-137 )

where:

I 6.49E+05 = 1.14E+05 ( U, / D, + U,f x BFg ) ( DFait ) ( .2) k C-17 Rev. 25 1/1/90

( s where:  ;

( )

'~'

1.14E+05 = ( 1,0E+06 pC1/uC1 x 1.0E+03 ml/kg ) / ( 8760 hr/yr )

0,, = 730 t/yr, Adult water consumption 37.7, dilution factor from the near field area to the nearest

' D" = potable water intake (Rock Hill Water Intake)  !

U,g = 21 kg/yr, Adult fish ccasumption ,

BFg= 2.00E+03, Bioaccumulation f actor for Cesium (Tame 3.1-1)  ;

DF = 1.21E-04, Adult total body ingestion dose f actor for "It '

Cs-134 (Table 3.1-2) 1.12 = Factor derived from the assumption that 89% of the dose is derived from Cs-134 and Cs-137 or 100% / 89% = 1.12 i i

and where:

, Fg =(f)(o)/(F+f) where x f= Catawba's liquid radwaste flow, in gpm

/ \

'% o= Recirculation factor at equilibrium, 1.027 (See section C2.2.1) l F= Catawba's dilution flow, in gpm f I and where:  ;

Tg= 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br />, the tinae period of time over which CCs-134, CCs-137 ""d '

F are averaged.

t L

l C Cs-134 = The average concentration of Cs-134 in Cetawba's undiluted ,

effluent, in uC1/m1, during the calenda: year of interest.  ;

g ,

l C Cs-137 = The average concentration of Cs-137 in Catawba's undiluted effluent,' in uCi/m1, during the calendar 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 (DH0(I c )) is estimated using the simplified dose calculation given i

below:

p DMOI c) = ( 2.26E+06 ) (Fg) (T )g (CNb-95) g C-18 Rev. 25 1/1/90

1 i'

where:

['w/

' ')

g 2.26E+06 = ( 1.14+05 ) ( U,,/ D, + U,g x BFg ) (DF,gg) (1.50) where:

1.14E+05 = ( 1.0E+06 pC1/uci x 1.0E+03 ml/kg ) / 8760 hr/yr U, = 730 t/yr, Adult water consumption i

37.7. Dilution factor from the near field area to potable water i D" =

intake.  ;

U,g = 21 kg/yr, Adult fish consumption BFg= 3.00E+04, Bioaccumulation factor for Niobium (Table 3.1-1)

DF = 2.10E-05, Adult GI-LLI ingestion dose factor for Nb-95 ait .

(Table 3.1 2) 1.50 = Factor derived from the assumption that 75% of adult GI-LLI dose in-from Nb-95 or 100% / 75% = 1.50 where:

fm

\

-Fg = (f) (o) / ( F + f )

V! 'where:

f= Catawba's liquid radwaste flow, in gpm o= Recirculation factor at equilibrium, 1.027

'F = Catawba's dilution flow, in gpm where:

T g = 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 and Fg are Nb-95 averaged.

C = e average c ncentration of E -95 in Catawba's undiluted Nb-95 effluent, in uCi/ml, during the calendar 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.

l 4

C-19 Rev. 25 1/1/90 l

p t ,

/' 'T 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(8 m )) is ,

estimated using the simplified. dose calculation given.below:

s Dy3(g,) = ( 9.32E 06 ) ( w ) ( QXe-133 ) ( F) ( 2.22 )

where:

w= 1.50E-07 = (X/Q) for the plume immersion factor pathway factor, in cec /m which corresponds to a location 5 miles SSW of the 8  :

McGuire site (See table B4.0-1) s Q Xe-133 = The total Xe-133 activity released from McGuire during the  !

calendar year of interest, in uC1.  ;

i 9.32E-06 = ( 3.17E-08 ) ( K,[X/Q] ), with appropriate substitutions for ,

whole body exposare in a semi-infinite cloud of Xe-133. See Section 1.2.1.

Sp = 0.7 = External radiation shielding factor for. individuals.

2.22 = The factor derived from the conservative assumption (based on

m. historical data) that 45% of the whole body dose to the maximally exposed individual is contributed by Xe-133.

i Based on operational history, the Catawba fuel cycle maximum exposed individual maximum organ dose (Adult-GI-LLI) resulting from McGuire's gaseous effluent releases (D,10(E

, m )) is estimated using the simplified dose calculation given below:

s Dgo(g,) = ( 7.23E-05) ( w ) ( Q ,3) (2.63) 3g wheres

'w = 1.5E-07 = E for the food and ground plane pathway, in sec m/', for a location 5 miles SSW of the McGuire site (Table B4,0-1) s Q ,3 3g

= The total 11-3 activity released from McGuire during the calendar year of interest, in uCi. ,

G 7.23E-05 = ( 3.17E-08 ) ( R g[X/Q) ) with appropriate substitutions for V

the adult-vegetable garden pathway, Rg[X/Q) for II-3. See Section 3.1. 2. 2. '

C-20 Rev. 25 1/1/90 V _.- - - _ _ _ _ _ _ _ _ _ -

i G

2.63 = 'ihe f actor derived from the conservative assumption (based on historical dsta) that 38% of the total inhalation, food and ground plane pathw;., dose to the maximally exposed individual is contri-buted by H-3 via the vegetable pathway.

C4.4.2.2 Catawba's Gaseous Contribution 4

Based on operational history, the Cat.awba fuel cycle maximum exposed individual whole body dose resulting from Catawba's gaseous efiluent releases (DWB(8c )) 18 catimated using the simplified doso calculation given below:

DWB(8 c )"(

~

(" (

Xe-133) ( F) ( '

where:

g w = 3.0E-05, (X/Q) for the plume immersion factor pathway facter which corresponds to a location 0.5 miles NE of the Catawba site (see Table C4.0-1) s O = The total Xe-133 activity released f rom Catawba during the Xe-133 calendar year of interest, in uCi.

9.32E-06 = (3.17E-Ot) (K g

[X/Q)), with appropriate substitutions for whole body expnsure in a semi-infinite cloud of Xe-133. See Section 1.2.1, O Sp = 0.7 = External radiation shielding factor for individuals.

2.56 = The factor derived from the conservative assumption (based on historical data) that 39% of the whole body dose ta the maximally exposed individual is contributed by Xe-133.

Based on. dwsign basis operation, the Catawba fuel cycle maximum exposed individual caximum organ dose (Adult-GI-LLI) resulting from Catawba's gaseeus effluent releases (DM0(E c )) is estimated using the simplified dose calculation given below:

s

-05 ) ( w ) ( (

DM0(8 c )=( QH -3 where:

w 3.0E-05 = X/Q for the food and ground plane pathway in m' , for a location 0.5 miles NE of the Catawba site (see Table C4.0-1).

~

Q g,3 = The total H-3 activity released from Catawba during the calendar year of interest, in uCi.

V e 1.23E-05 = ( 3.17E-08 ) ( R [X/Q) ) with appropriate substitutions for V ___

the adult-vegetable garden pathway factor, R [X/Q] for H-3.

See Section 3.1.2.2.

C-21 Rev. 25 1/1/90 1 ,

+. s

  • ' L *
4. - . .

J

[.

--.'i .g -

n s . .

9-J 1, 4 : ,

.h' yc_ *

.1 i

y , ,

.; / [ 1 j-

',O L 1.56'= N factor derived from the assumption that 64% of the total  !

(,- . inhalation,' food and ground plane pathway dose to-the' maximally-j oxposed. individual is contributed by H-3 via the! vegetable. garden

, ' pathway, j

)

I n ,,

ho

  • i s, . =?

5 ,: .

s- .t f D 4

t

,'- .,t t

(= ,

g sf s

u. . : c_ k, _.

t i-4 1

I kJ#

+

t1 9

( 1 f'4 s

' lf o ;a >

"![ ,

7, t

, ,3 f ' " o .. . <

C-22 Rev. 25 i<

W, on %rs, y[ ' ,

5 ,>-

r ci;J w, >

1/1/90-7

a .~ .

_ 'x _.

a

,~ =*

_( s .

y

.A '(

  • n.

TABLE C4.0-1 (1 of'2)

CATAWBA NUCTE.AR STATION ,

DISPERSION PARAMETER (X/01 FOR IDNG TERM'REIE.ASES > 500 HR/YR OR > 125 HR/(TTR (sec/m')

Distance to the control location. (miles)

Swetor 0.5 1.0 M 2.0 2.5 3.0 3.5 4.0 4.5 5.0-N- 2.6E-5 6.5E-6 2.7E 1.5E-6' 9.7E-7 6.9E-7 5.2E-7 4.1E-7 3.3E-7 2.8E-7 NNE 3.1E-5 8.1E-6 3.3E-6 1.8E-6.. 1.2E 8.2E 6.2E-7 4.9E-7 4.0E-7 3.3E 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 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 1.1E-7 5 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 3.4E-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 6.8E-8 W 8.9E-6 2.3E-6 9.3E-7 5.0E 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.SE-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

___ = -_. _ ___ _ _ _ __ __ .

.y < .g 7; ,

_I:

t- t .

?

r I

e

~

f"'p ,

- TABLE C4.0-1 )

h \q'f. (2 of 2) m, CATAWBA NUCLEAR STATION 4

Thi values presented in this. table were generated by using the computer program XOQDOQ

~

L: -(NUREQ/CR-2919) which implements NRC Regulatory Guide 1.111;(1977). and the following  :

Jcscumptions:

l '. c ' Data Collection Period,- 12/17/75 to 12/16/77.  ;

. 2.- Ground Level Releases. l

! 3 ', Height.of_The Vent's Building ~= 47 meters. ,

4. - Open' Terrain Recirculation Correctf or. Factors.

5 Y

5

-_ {

I

.: w

. J -8

'. i r s f

t \

e

!g t

4 h

i e N .t

, I \

. . .[~ k_ / .1 a:

ra.

Rev. 13'

. , : ;~, 7 1/1/87' ,

v ;;s ~. ' '

ide. ,, ......

.25 e- - n

~

.y .,

q p

- ..( v) . 4 J

,%s .

~ ~

t.

TABLE'C4.0-2

"(1 of 2) ,

-CATAWBA NUCLEAR STATION DEPOSITION PARAMETER (D/0) FOR LONG 'IERM RETIAKES > 500 HR/YR OR > 125 HR/QTR (meter ~*).

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 5.6E-9t 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 2.8E-9 1.8E-9 1.3E-9 9.5E-10 7.4E-10 5.8E-10 NE 1.1E-7 2.6E-8 9.3E-9 4.6E-9 2.7E-9 1.8E-9 1.3E-9 9.3E-10 7.2E-10 5.7E-10 ENE 4.1E-8 1.0E-8 3.6E-9 1.8E-9 1.1E-9 6.9E-10 4.9E-10 3.6E-10 2.8E-10 2.2E-10 E 3.6E-8 8.8E-9 3.2E-9 '1.6E-9 9.3E-10 6.1E 4.3E-10 3.2E-10 2.4E-10 1.9E-10 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 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 7.8E 5.8E-10 4.4E-10 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 5.0E-10 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 W 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 WNW 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 7.3E-10 4.8E-10 3.4E-10 2.5E-10 1.9E-10 1.5E-10' NNW 4.1E-8 9.9E-9 3.6E-9 1.8E-9 -1.0E-9 6.8E-10 4.8E-10 3.6E-10 2.7E-10 2.2E-10 Rev. 25 1/1/90.-

o

. , - ~ . . .,.,.,, . .-- - ,,r.. . - , , _ , . . . . - _ . . . .

y. ,:: --

R J:

T-i .k .

t ,

"f . . - '.

j ,7^N  : TABIE C4.0-2

-( ~ ~

(2 of 2)

CATAWBA NUCIEAR STATION P Thy values" presented - in s this table s were generated by . using .the . computer . program -

cX0QD0Q'(NUREQ/CR-2919) which implements NRC Regulatory Guide 1.111 (1977) and the following; assumptions:'

! Data Collection Period,- 12/17/75 to 12/16/77~.

~

!1.

' ~

. 2- ' Ground Level Releases.

3 . -:. iHeight of The-Vent's Building = 47 meters.

~

C

. 4 .' . Open Terrain: Recirculation Correction Factors.

k A

-f

)

4 6

\ * \ m f

s k

i ' . - -

h

.', f 9

.> i

- t Q

(

ir'.

- .c '.

Rev. 11 '

q,.: '

8/31/86-

. a

, . , - - . -6 - - , ,

(J i 1

1 l

ft TABLE C4.0-3

  • i

\

~ ~'

l- l (1 of 3)

CATAWBA NUCLEAR STATION ,

ADULT A DOSE PARAMETERS ait 1

(mrem /hr per pCi/ml)

NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LII i

H 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 L 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 L FE 55. 6.04E+02- 4.59E+02 1.07E+02 0.0 0.0 2.56E+02 2.63E+02  ;

l FE 59 1.05E+03 2.46E+03 9.45E+02 0.0 0.0 6.89E+02 8.21E+03 L C0.58 0.0 9.08E+01 2.04E+02 0.0 0.0 0.0 1.84E+03 I COl60 0.0' 2.61E+02 5.75E+02 0.0 0:0 0.0 4.90E+03

. NI 63 3.14E+04 2.18E+03 1.05E+03 0.0 0.0 0.0 4.54E+02 L NI 65 1.28E+02 1.66E+01 7.56E+00 0.0 0.0 0.0 4.20E+02 +

l CU 64 0.0 1.02E+01 4.77E+00 0.0 2.56E+01 0.0 8.66E+02  ;

ZN 65 2.32E+04 ~7.38E+04 3.33E+04 ' O.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.  ;

p j_s BR 83 0.0 0.0 4.05E+01 0.0 0.0 0.0 5.83E+01 i t i .BR 84 0.0- 0.0 5.25E+01 0.0 0.0 0.0 4.12E-04 l )N / BR 85 0.0, 0.0 2.16E+00 0.0 0.0 0.0 0.0 l 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 l 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 sg_gp 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 i Y 91M 5.64E-03 0.0 2.18E-04 0.0 0.0 0.0 1.66E-02 l- JY 91 8.75E+00 0.0 2.34E-01 0.0 0.0 0.0 4.82E+03

1. Y 192 5.24E-02 0.0 1.53E-03 0.0 0.0 0.0 9.18E+02 a
  • Methodology for table provided by: M. E. Wrangler, RAB:NRR:NRC on 3/17/83 l

N Tt.Bry c4.n.3

. .!s_.) .

f a Rev. 4 7/18/84

=

., - . . - -- - ~

a i

, - f~l TABLE C4.0-3 1K ) -

I:

(2 of 3) o +

CATAWBA NUCLEAR STATION' U

ADULT A,g DOSE, PARAMETERS (arem/hr per pCi/ml) h

.NUCLIDE BONE LIVER T. BODY 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 MO 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 0.0 4.04E-01 1.30E-02 1.57E+01 TC 101 9.68E 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 125M2.57E+03 9 . 3 2E+02 3.45E402 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 f-~s 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 i'- ') . 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 JTE 131' l'. 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.E9E+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 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 1.77E+05 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 BA 139 1.14E+00 8.14E-04 3.35E-02 0.0 7.61E-04 4.62E-04 2.03E+00 s

ss _-)

A Rev. 4

, , 7/18/84

^

i;;;

I, AJ f' 'i TABLE C4.0-3 (3 of 3)

CATAWBA NUCLEAR STATION ADULT A DOSE M E E RS ait (mrem /hr per DC1/ml)

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' 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

'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 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

L L

L' i

L TABLE C4.0-3 L /" , (3 of 3) a.

g Rev. 4 W 7/18/84

m f/~'h . 'C5.0 Radiological Environmental' Monitoring

\ ):

The Radiological Environmental Monitoring Program shall be. conducted in accordance with Technical Specification, Section 3/4.12.

I

._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 JLn an intorlaboratory comparison program which has been approved by the a NRC. This program is the Environmental Protection Agency's (EPA's)

Environmental Radioactivity Laboratory Intercomparison Studies (crosscheck)

~

Program,.our participation code is CP.

The dates of the land-use census that was used to identify the controlling receptortlocations was 06/01/89 - 07/01/89.

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~(1 of.1)e CATAWBA RADIOLOGICAL MONITORING PROGRAM SAMPLING IDCATIONS (TLD LOCATIONS)

SAMPLING LOCATION DESCRIPTION SAMPLING IDCATION DESCRIPTION ~

200 SITE BOUNDARY (0.6M NNE) 232 4-5 MILE RADIUS (4.1M NE)'l' 201 SITE BOUNDARY .(0.5M NE) 233 4-5 MI E RADIUS (4.0M ENE)--

202 SITE BOUNDARY (0.6M E) Deleted 234 4-5 MILE RADIUS (4.5M E) I 203 SITE BOUNDARY (0.4M ESE) 235 4-5 MIE RADIUS (4.0M ESE)l-204 SITE BOUNDARY (0.5M SSW) 236 4-5 MILE RADIUS (4.2M SE):

205 SITE BOUNDARY (0.3M SW) 237 4-5 MILE RADIUS (4.8M SSE)l 206- SITE BOUNDARY (0.7M WNW) 238 4-5-MILE RADIUS (4.2M S)-

207 SITE BOUNDARY (0.9M NNW) 239 4-5 MILE RADIUS (4.6M SSW)l 212 SPECIAL INTEREST (3.3M E) 240 4-5 MILE RADIUS (4.1M SW).

217 CONTROL (10.0M SSE). 241 4-5. MILE RADIUS (4.7M WSW) 222 SITE BOUNDARY (0.7M N) 242 4-5 MILE RADIUS (4.6M W)-

223 SI'If EDUNDARY (0.6M E) 243 4-5 MIE RADIUS (4.6M WNW)I 224 SITE BOUNDARY (0.6M ESE) Deleted 244 4-5 MILE RADIUS (4.1M NW) 1 725 SITE BOUNDARY (0.7M SE) 245 4-5 MILE RADIUS -(4.2M NNW)I 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 (0.8M NW) 249 SPECIAL INTEREST .(8.1M S) l ,

'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) 255 SITE BOUNDARY (O.6M ENE)- 1 256 SITE BOUNDARY (0.6M SSE)- I Rev. 25' 1/1/90'

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~ CATAWBA RADI0IDGICAL ?fGNITT) RING PROGRAM ANALYSES - ': -

ANAT.YSF.S TRITIUM-  : GROSS BETA: . TIR

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SAMPLE MEDIUM- ANALYSIS: SCHEDULE GAMMA ISO'IT)PIC IDW IIVEL I-131

1. Radiolodine and ' Weekly X.

Particulates X X

2. Direct Radiation -Quarterly X:
3. Surface Water Biweekly. X Honthly Composite. X Quarterly' Composite .X

' 4. Drinking Water Biweekly. X Monthly.Ccaposite 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-(a) X (a) during harvest season ..

-Rev. 13-1/1/87 .

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