ML20072F522
| ML20072F522 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 03/21/1983 |
| From: | UNION ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20072F520 | List: |
| References | |
| PROC-830321, NUDOCS 8303240311 | |
| Download: ML20072F522 (66) | |
Text
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Rev. 0 1 .
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4 lo UNION ELECTRIC COMPANY CALLAWAY. PLANT OFFSITE DOSE CALCULATION MANUAL i
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8303240311 830321 i PDR ADOCK 05000403 PDR A
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Rav. 0 Table of Contents Page 1.0 PURPOSE AND SCOPE 1 2.0 LIQUID EFFLUENTS 2 2.1 RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATION 3.3.3.9 2 2.2 LIQUID EFFLUENT MONITORS 2 2.2.1 Coatinuous Liquid Effluent Monitors 3 2.2.2 Radioactive Liquid Batch Release C1 Effluent Monitors 5 2.3 ODCM METHODOLOGY FOR THE DETERMIN-ATION OF LIQUID EFFLUENT MONITOR SETPOINTS 6 2.3.1 Development of ODCM Methodology for the Determination of Liquid Effluent Monitor Setpoints 6 2.3.2 Summary, Setpoint Determination Methodology for Liquid Effluent Monitors 11 2.4 LIQUID EFFLUENT CONCENTRATION MEASUREMENTS 12 2.4.1 Radiological Effluent Technical Specification 3.11.1.1 12 2.4.2 Description of Liquid Effluent Concentration Measurements 12 2.5 INDIVIDUAL DOSE DUE TO LIQUID EFFLUENTS 13 2.5.1 Radiological Effluent Technical Specification 3.11.1.2 13 2.5.2 The Maximum Exposed Individual 13 2.5.3 ODCM Methodology for Determining Dose Contributions from Liquid Effluents 13
-i-
. . _ - . - . - _ _ . , , .- - , - ~ -. ~- .
Rsv. O Table of Contents (continued)
. Page 2.5.3.1 Calculation of Dose Contributions 13 '
2.5.3.2 Dose Factor Related to Liquid Effluents 15 2.5.4 Summary, Determination of Individ-ual Dose Due to Liquid Effluents 17 2.6 LIQUID RADWASTE TREATMENT SYSTEM 21 2.6.1 Radiological Effluent Technical Specification 3.11.1.3 21 2.6.2 De'scription of the Liquid Radwaste Treatment System 21 2.6.3 Operability of the L' quid i Radwaste Treatment System 21 3.0 GASEOUS EFFLUENTS 22 3.1 RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATION 3.3.3.10 22 3.2 RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATION 3.11.2.1 22 3.3 GASEOUS EFFLUENT MONITORS 22 3.3.1 Continuous Release Gaseous Effluent Monitors 23
() 3.3.2 Batch Release Gaseous Monitors 25 3.4 ODCM METHODOLOGY FOR THE DETERMIN-l ATION OF GASEOUS EFFLUENT MONITOR SETPOINTS 26 l 3.4.1 Development of ODCM Methodology l for the Determination of Gaseous Effluent Monitor Setpoints 26 3.4.1.1 Whole Body Dose Rate Setpoint Calculations 28 3.4.1.2 Skin Dose Rate Setpoint Calculation 29 3.4.1.3 Gaseous Effluent Monitors Setpoint Determination 31
-ii- i
Rrs. 0_
Table of Contents (continued)
- Page 3.4.2 Summary, Gaseous Effluent Monitors Setpoint Determination 31 3.5 ODCM METHODOLOGY FOR DETERMINING
, DOSE CONTRIBUTIONS FROM GASEOUS EFFLUENTS 31 3.5.1 Determination of Dose Rate 31 3.5.1.1 Noble Gases 31 3.5.1.2 Radionuclides Other Than Noble Gases 32 3.5.2 Individual Dose Due to Noble Gases 36
, 3.5.2.1 Radiological Effluent Technical Specification 3.11.2.2 35 3.5.2.1.1 Noble Gases 35 3.5.2.2 Radiological Effluent Technical Specification 3.11.2.3 36 3.5.2.2.1 Radionuclides Other Than Noble Gases 37 3.6 GASEOUS RADWASTE TREATMENT SYSTEM 40 3.6.1 Radiological Effluent Technical Specification 3.11.2.4 40 O 3.6.2 Description of the Geseous Radwaste Treatment System 40 l 3.6.3 Operability of the Gaseous Radwaste Treatment System 40 l
4.0 DOSE AND DOSE COMMITMENT FROM URANIUM FUEL CYCLE SOURCES 41 4.1 RADIOLOGICAL EFFLUENT TECHNICAL l SPECIFICATION 3.11.4 41 j 4.2 ODCM METHODOLOGY FOR DETERMINING i DOSE AND DOSE COMMITMENT FROM l URANIUM FUEL CYCLE SOURCES 41 '
i 4.2.1 Calculation of Total Dose 41 j l
iii -
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l y -, w- ,y- w-- w --- 9
-g-- -y- ------% - -w. -y = .-+4w<r w % -.-w.
Rw. O Table of Contents (continued) ,
Page ,
4.2.1.1 Calculation of Total Dose from .
Noble Gases 41 4.2.1.2 Calculation of Total Dose from Radionuclides Other Than Noble Gases 42 4.2.1.3 Summary, Determination of Total Dose 42 5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING 44 5.1 RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATION 3.12.1 44
5.2 DESCRIPTION
OF THE RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 44 6.0 METEOROLOGICAL DATA COLLECTION AND PROCESSING 52 6.1 METEOROLOGICAL DATA COLLECTION 52 6.2 METEOROLOGICAL DATA PROCESSING 52 7.0 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 54 8.0 IMPLEMENTATION OF ODCM METHODOLOGY 55
9.0 REFERENCES
56 l
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Rev. 0
- List of Figures Figure 5.lA Radiological Air Sampling Network Figure 5.lB Radiological Air Sampling Network i
Figure 5.2 Location of Aquatic Sampling Stations Figure 5.3 Groundwater Quality Monitoring Locations Figure 5.4 Food Products and Milk Sampling Locations
- Figure 5.5 Locations of Soil Sampling Stations O
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Li~st of Tables
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, s
');s ,
' ^ - . Page rable 1 Ingestion Dose Commitment Factor' , .. < ,
4 s (Ai ) for Adult Age Group- , 18' l v ,m . -
Table 2 Bioaccumulation Factor (BF.) Used in the Absence' of Site-Spehific V
.i' < Data 20 s
Table;3 ' " Dosh Factors for Exposure to A
.. Semi-Infinite Cloud of Noble Gases 30 Table"4 Dose Parameter (P- for Radio-
'nuclides Other Thkn) Noble Gases 34 Table 5 Pathway. Dose Factors (R- for tradiosuclides Other Thah) Noble O s
. t. oefe=
39 Table 6 Radiological Environmental
,, Monitoring F.rogram 45 N
'e - . Table 7 i N Reporting Leve'is for Radioactivity N'
\ Concentratiods in Envircnmental
't w .. Samples -
48
. m_ .
Table 8 Maximum Values for the Lowed s Limits of Det6ction ,
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R v. 0 1.0 PURPOSE AND SCOPE The Offsite Dose Calculation Manual (ODCM) describes the methodology and parameters used in the calculation of offsite doses and dose rates due to radioactive liquid and gaseous effluents and in the calculation of liquid and gaseous effluent monitoring instrumentation alarm / trip setpoints.
The ODCM also contains a list and description of the specific sample locations for the radiological environmental monitoring program.
Changes in the calculational methodologies or parameters will be incorporated into the ODCM and documented in the Semi Annual Radioactive Effluent Release Report. The ODCM does not
(~J)
L replace any station implementing procedures.
O f
Rnv. 0 2.0 LIQUID EFFLUENTS 2.1 Radiological Effluent Technical Specification 3.3.3.9 The radioactive liquid effluent monitoring instrumenta-tion channels shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Radiological Effluent Technical Specification 3.11.1.1 are not exceeded. The alarm / trip setpoint of these channels shall be determined in accordance with the methodology described in the ODCM.
2.2 Liquid Effluent Monitors Gross radioactivity monitors which provide for au-tomatic termination of liquid effluent releases are present on the liquid effluent lines. Flow rate meas-urement devices are present on the liquid effluent s lines and the discharge line (cooling tower blowdown).
Setpoints, precautions and limitations applicable to the operation of the Callaway Plant liquid effluent monitors are provided in the appropriate Plant Procedures. Setpoint values are calculated to assure that alarm and trip actions occur prior to exceeding the Maximum Permissible Concentration (MPC) limits in 10 CFR Part 20 at the release point to the unrestricted area. The calculated alarm and trip action setpoints for the liquid effluent line monitors and flow measur-ing devices must satisfy'the following equation:
(' cf S C
(_) F+f (2.1)
Where:
C= the liquid effluent concentration limit (MPC) implementing Radiological Effluent Technical Specification 3.11.1.1 for the site in (pCi/ml).
c= The setpoint, in (pCi/ml), of the radioactiv-ity monitor measuring the radioactivity concentration in the effluent line prior to dilution and subsequent release; the setpoint, which is inversely proportional to the volumetric flow of the effluent line and Rev. 0 directly proportional to the volumetric flow of the dilution stream plus the effluent stream, represents a value, which, if ex-ceeded, would result in concentrations exceed-ing the limits of 10 CFR Part 20 in the unres-tricted area.
f= The flow setpoint as measured at the radiation monitor location, in volume per unit time, but in the same units as F, below.
F= The dilution water flow setpoint as measured prior to the release point, in volume per unit time. {If (F) is large compared to (f), then F + f ~ F} .
If'no dilution is provided, then c < C.
The radioactive liquid waste stream is diluted by the Os plant discharge line prior to entry into the Missouri River. Normally, the dilution flow is obtained from the cooling tower blowdown, but should this become unavailable, the plant water treatment facility sup-plies the necessary dilution flow via a bypass line.
The batch release limiting concentration (c) which cor-responds to the liquid radwaste effluent line monitor setpoint is to be calculated using methodology from the expression above.
Thus, the expression for determining the setpoint on the liquid radwaste effluent line monitor would become:
({) c<-
C(F + f) (pCi/ml) f (2.2) 2.2.1 Continuous Liquid Effluent Monitors The radiation detection monitors associated with conti-nous liquid effluent releases are:
Monitor I.D. Description O-BM-RE-52 Steam Generator Blowdown Discharge Monitor 0-LE-RE-59 Turbine Building Drain Monitor
Rnv. 0 These effluent streams are not considered to be radi-oactive unless radioactivity has been detected by the associated effluent radiation monitor or by laboratory analysis. -The sampling frequency, minimum analysis frequency, and type of analysis performed are as per Radiological Effluent Technical Specification Table 4.11-1.
The steam generator blowdown discharge monitor conti-nuously monitors the blowdown discharge pump cutlet to detect radioactivity due to system demineralizer break through and to provide backup to the steam generator blowdown process radioactivity monitor to prevent dis-charge of radioactive fluid. The sample point is located on the discharge of the pump in order to moni-tor discharge or recycled blowdown fluid and upstream of the discharge isolation valve to permit termination of the radioactive release prior to exceeding the in-stantaneous concentration limits of 10 CFR Part 20.
c).
(. The high radioactivity alarm / trip setpoint initiates control room alarm annunciation and automatic isolation of the blowdown isolation valves and the blowdown dis-charge valve.
The turbine building drain effluent monitor is provided to monitor turbine building liquid effluents prior to release to the environs. The fixed-volume detector as-sembly continuously monitors the drain effluent line upstream of the drain line isolation valve. The high radioactivity alarm / trip setpoint initiates control room annunciation and automatic isolation of the drain line isolation valve to prevent the release of radioac-tive fluids. The sample location ensures that all potentially radioactive turbine building liquid ef-
! fluents are monitored prior to discharge.
O Each monitor channel is provided with a two level sys-tem which provides sequential alarms on increasing radioactivity levels. These setpoints are designated as alert setpoints and alarm / trip setpoints.-
The alarm / trip setpoints are determined through the use of Equation (2.2) methodology to ensure that Radiologi-cal Effluent Technical Specification 3.11.1.1 limits are not exceeded at the site unrestricted area boundary. The alert setpoints have been administra-tively established below the alarm / trip setpoints, thus providing an additional margin of safety.
The alarm / trip setpoint calculations for the steam gen-erator blowdown discharge monitor and the turbine bu-ilding drain monitor are based on the representative dilution flow rate, the~ maximum effluent stream flow
., 7 Rev. 0 rate, and the' MPC for Cs-137 as the controlling isotope. A portion of the total site allocation is provided for the value of the radiation monitor set-point,(c), based on local procedures and operating experience. In the event the alarm / trip setpoint is reached, the radiation monitor setpoint (c),.will be reevaluated using the actual dilution flow rate (F),
the actual effluent stream flow rate (f), and the ac-tual isotopic analysis. This evaluation will then be used to ensure that Radiological Effluent Technical Specification 3.11.1.1 limits were not exceeded.
2.2.2 Radioactive Liquid Batch Release Effluent Monitor
- The two radiation monitors which are associated with the liquid effluent batch release systems are: ,'
MONITOR I.D. Description O O-HB-RE-18 Liquid Radwaste Discharge Monitor 0-HF-RE-45 Secondary Liquid Waste System Monitor' 4
The liquid radwaste radiation monitor continuously monitors the discharge of the liquid radwaste process-ing system to prevent the discharge of radioactive fluid to the environs. The fixed-volume detector as-sembly continuously monitors the system discharge line upstream of the discharge valve. The high radioactiv-ity alarm / trip setpoint initiates control room alarm annunciation and automatic isolation of the liquid rad-waste system discharge valve to terminate discharge.
l O The sample point is located to ensure that all potenti-ally radioactive fluids from the liquid radwaste processing system are monitored prior to discharge.
The secondary liquid waste system discharge radioactiv-ity monitor continuously monitors secondary liquid waste system effluents prior to discharge to the environs. The fixed-volume detector assembly monitors.
the discharge line upstream of the discharge isolation valve. The high radioactivity alarm / trip setpoint ini-tiates control room alarm annunciation and automatic l
isolation of the secondary liquid waste system dis-charge valve to prevent the discharge of radioactive fluid.. The sample location ensures that all potenti-l ally radioactive sources from the system are monitored prior to discharge.
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Rev. 0 The setpoint for these monitors is determined according to the methodology described by Equation 2.2 and is a function of the dilution flow rate (F), the radioactive effluent line flow rate (f) and the tank liquid ef-fluent concentration, as determined by a pre-release isotopic analysis. Based on these factors, a setpoint is calculated for the appropriate monitor to ensure that Radiological Effluent Technical Specification 3.11.1.1 limits are not exceeded at the site unres-tricted area boundary.
2.3 ODCM Methodology for the Determination of Liquid Effluent Monitor Setpoints The dependence of the setpoint (c), on the radionuclide .
distribution, yields, calibration, and monitor paramet-ers, requires that several variables be considered in setpoint calculations.
() 2.3.1 Development of ODCM Methodology for the Determination of Liquid Effluent Monitor Setpoints
~
2.3.1.1 The isotopic concentration of the release be-ing considered must be determined. This is obtained from the sum of the measured concentrations as deter-mined by the analysis required per Radiological Ef-fluent Technical Specifications Table 4.11-1:
ICi=IC gg + C, + C s +C t (2.3)
O Where:
C. = the concentration of each radionuclide i, as 1
determined by the analysis of the waste sample.
IC 99
=
the sum of the concentrations (C ) of each measured gamma emitting nuclide 8bserved by gamma-ray spectroscopy of the waste sample.
C**
=
the measured concentrations (C of alpha em-itting nuclides observed by gr8s)s alpha analysis of the monthly composite sample.
Rnv. O C*=
s the measured concentrations of Sr-89 and Sr-90 in liquid waste as determined by analysis of the quarterly composite sample.
C*=
t the measured concentration of H-3 in liquid waste as determined by analysis of the monthly composite sample.
The C term is included in the analysis of each batch; terms9 for alpha, strontium, and tritium are included as appropriate.
- Values for these concentrations will be based on previous composite sample analyses as required by Table 4.11-1 of the Radiological Effluent Technical Specifications.
2.3.1.2 The measured radionuclide concentrations are used to calculate a Dilution Factor (Fg ), which is the O- ratio of total dilution flow rate to tank flow rate required to assure that the limiting concentrations of Radiological Effluent Technical Specification 3.11.1.1 are met at the point of discharge. This is referred to as the required Dilution Factor and is determined ac-cording to:
Fd=I1 MPC
+F s (2.4) i O Cg C C C
=I , a + s + t +F g (2.5) 9MPC MPC a MPC MPC t g s Where:
C. = measured concentrations of C , C, C and C as defined in 2.3.1.1. Term 2 C a C8 s,and C t 1
t will be included in the calculation as appropriate.
MPC.
1
= MPC MPC , MPC , and MPC are limiting con 8e,ntrations 8f the appbopriate radionuclide from 10CFR 20, Appendix B, Table II, Column 2.
Rev. O For dissolved or entrained noble gases t the concentration shall be limited to 2x10 4 pCi/ml total activity.
F = the safety factor; a conservative factor used s
to compensace for statistical fluctuations and errors of measurements. (For example, F g = 0.5 corresponds to a 100 percent variation.) Default value is F g = 0.9.
2.3.1.3 For the case Fd < 1, the monitor tank effluent concentration meets the limits of Radiological Effluent Technical Specification 3.11.1.1 without dilution and the effluent may be released at any desired flow rate.
If Fa > 1 then dilution is required to ensure com-pliaHee with Radiological Effluent Technical Specifica-tion 3.11.1.1 concentration limits. If simultaneous releases are occuring or are anticipated, a modified dilution factor (Fd ), must be determined so that
() available dilution 21ow may be apportioned among simultaneous discharge pathways.
Fdn = Fd+F a (2.6)
Where:
F" = the allocation factor which will modify the required dilution factor such that l
/~)
k-simultaneous liquid releases may be made l without exceeding the limits of Radiological Effluent Technical Specification 3.11.1.1.
2.3.1.4 The most straight-forward determination of the allocation factor is:
i I
l l
F, = f (2.7) l f
(
I
)
Rsv. 0 1
l
)
Where: l n= the number of liquid discharge pathways for which Fa > 1 and which are planned for simultaMeous release.
However, this value for F may be unnecessarily res-trictive in that all rele$se pathways are apportioned the same fraction of the available dilution stream, regardless of the relative concentrations of each of the sources.
Since the radionuclide concentration of the two conti-nuous sources is less than that of the batch release source, it is acceptable to allocate smaller portions of the dilution stream to the continuous releases and a larger portion to the batch releases.
7- Therefore, F is necessarily defined as a flexible
(,)/ quantity wit 8 a default value of 1/n. Prior to initi-ating a simultaneous release, a check must be made to assure that the sum of the allocation factors assigned to pathways for the simultaneous release is 1 1.
2.3.1.5 The calculated maximum permissable whste tank effluent flow rate, (f , is based on the modified dilution factor, (F ) main)d the effective dilution flow rate, (F ff). TheNEfectivedilutionflowrateis given by Feff = (0.9)Fe I2*0)
Where:
the cooling tower blowdown flow rate and/or F* = bypass dilution flow.
A conservative value for F would be the minimum allow-able cooling tower blowdod of 5000gpm which is used as a default value.
2.3.1.6 Having established the values of F and F thecalculatedmaximumpermissiblewastetaNRflow95fe, can be calculated by:
_9_
Rev. O f
max i *ff g
P~ *ff g (f r fp << Feff) (2.9) dn dn Where:
f p = the expected undiluted effluent flow rate.
Thus the effluent Even though flowof the value rate f is set at or below f"6Ee.
actualeffluentpumpcapacEfy,maybelargerthanit (f does represent the upper limit to the effluent fEo)w, rate whereby the s,/ requirements of Radiological Effluent Technical Specif-ication 3.11.1.1 may still be met. If Fa < 1, the ef-fluent flow rate setpoint may be assigned any value since the waste tank effluent concentration meets the limits of Radiological Effluent Technical Specification 3.11.1.1 without dilution and the release may be made without regard to the setpoints for other release pathways. For those discharge pathways selected to be secured during the release under consideration, the flow rate setpoint should be set at as low a value as practicable to detect any inadvertent release.
2.3.1.7 The liquid radiation monitor y setpoint may now be determined based on the values of . C and f whichwerespecifiedtoprovidecomplianbe, withe 85, limits of Radiological Effluent Technical Specification 3.11.1.1.
(])
The monitorthe response agia therefore, actual is primarily setpoint istobased gamma on {g g. tion, The calculated monitor setpoint concentration is determined as follows:
c=AIC gg pCi (Refer to Note) (2.10)
Following
R'av. O Where:
A= Adjustment factor which will allow the set-point to be established in a practical manner for convenience and to prevent spurious alarms.
A=Imax (Refer to Note) (2.11) f Following) p If A > 1: Calculate c and determine the maximum value
() for the actual monitor setpoint (pCi/ml).
If A < 1:No release may be made. This condition must be flagged and the operator instructed to re-evaluate 2.3.1.3 and 2.3.1.5 (i.e., reduce ef-fluent flow rate or return radwaste for reprocessing).
NOTE If Fa < 1, no further dilution is required and the releMse may be made without regard to available dilu-tion or to other releases made simultaneously. How-ever, it is necessary to estahlish a monitor setpoint which will provide alarm should the release concentra-tion inadvertently exceed Radiological Effluent Techni-(~N cal Specification 3.11.1.1 limits. This can be accom-N/ plished by establishing the adjustment factor as follows:
1 A=F d (2.12) 2.3.2 Summary, Setpoint Calculation Methodology for Liquid Effluent Monitors The methodology described in 2.3.1 is used to determine setpoints for each of the radiation monitors assigned a d
Rav. O liquid process functi-on. The limiting release concen-tration can be increased by reducing the discharge flow-rate and decreased by reducing the cooling tower blowdown flow-rate.
2.4 Liquid Effluent Concentration Measurements 2.4.1 Radiological Effluent Technical Specification 3.11.1.1 The concentration of radioactive material released from the site shall be limited to the concentrations speci-fied in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained no-ble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2.0 E-04 pC1/ml total activity.
2.4.2 Liquid Effluent Concentration Measurements
)
Liquid batch releases are discharged as a discrete volume and each release is authorized based upon the sample analysis and the dilution flow rate existing in the discharge line at time of release. To assure re-presentative campling, each liquid monitor tank will be isolated and thoroughly mixed by recirculation of tank contents prior to sample collection. The methods for mixing, sampling, and analyzing each batch are outlined in applicable plant procedures. The allowable release rate limit is be calculated for each batch based upon the pre-release analysis, dilution flow-rate, and other procedural conditions, prior to authorization for release. The radwaste liquid effluent discharge is monitored prior to entering the dilution discharge line and will automatically be terminated if the pre-()
'~
selected alarm / trip setpoint is exceeded. Concentra-tions are determined primarily from the gamma isotopic analysis of the liquid batch sample. For alpha, Sr-89, Sr-90 and H-3, the measured concentration from the previous composite analysis is used. Composite samples are collected for each batch release and monthly and quarterly analyses are performed in accordance with Table 4.11-1 of the Radiological Effluent Technical Specifications.
Dose contributions from liquids discharged as conti-nuous releases are determined by utilizing the last measured values of samples required in accordance with Radiological Effluent Technical Specifications Table 4.11-1. )
1
)
Rev. 0 2.5 Individual Dose Due to Liquid Effluents 2.5.1 Radiological Effluent Technical Specification 3.11.1.2 The dose or dose commitment to an individual from radi-oactive materials in liquid effluents released from the site shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and less than or equal to 5 mrem to any organ, and L. During any calendar year to less than or equal to 3 mrem to the whole body and to less than or equal to 10 mrem to any organ.
2.5.2 The Maximum Exposed Individual
() The cumulative dose determination considers the dose contributions from the maximum exposed individual's consumption of fish and potable water, as appropriate.
Normally, the adult is considered to be the maximum ex-posed individual.
The Callaway Plant's liquid effluents are discharged to the Missouri River. As there are no potable water in-takes within 50 miles of the discharge point, this pathway does not require routine evaluation. There-fore, the dose contribution from fish consumption is expected to account for more than 95% of the total man-rem dose from discharges to the Missouri River.
Dose from recreational activities is expected to con-tribute the additional 5%.
Thus, the maximum exposed individual is an adult, l
(]) receiving 95% of the total dose from eating fish and 5%
l of the total dose from recreational activities.
t 2.5.3 ODCM Methodology for Determining Dose Contributions
! From Liquid Effluents l
2.5.3.1 Calculation of Dose Contributions
! The dose contributions for the total time period l
! m IAtg 1=1 are calculated monthly (at least one each 31 days) and
! a cumulative summation of these total body and any or-gan dose is maintained for each calendar quarter.
'N , . . _ _ ._ . _ - -
Rev. O.
These dose contributions are calculated for all radion-uclides identified in liquid effluents released to un-restricted areas using the following expression:
m D
7 =I1 [Ai7 I Atg C ig F]
g (2.13) t=1 Where:
D r= the cumulative dose commitment to the whole
(]) body or any organ, t, from the liquid ef-fluents for the total period m
- Iaty t=1 in mrem.
Atg= the length of the Ath time period over which and F are averaged for all liquid C.$
rd eases,Ein hours.
Cig = the average concentration of radionuclide, i, in undiluted liquid effluent during time period Atg from any liquid release, in pCi/ml.
() A.
1*
= the site related ingestion dose commitment factor to the total body or any organ I for each identified principal gamma and beta emit-ter listed in Table 4.11-1, Radiological Ef-fluent Technical Specifications, (in mrem /hr) per (pCi/ml). These factors are given in l Table 1, as derived through the use of
- Equation (2.17).
the near field average dilution factor for C F1 = during any liquid effluent release. Defined iE as the ratio of the maximum undiluted liquid waste flow during release to the product of l the average flow from the site discharge
! structure to unrestricted receiving waters l
, times 89.77. (89.77 is the site specific ap-plicable factor for the mixing effect of the discharge structure.)
i
1 Rcv. O The term C is the composite undiluted concentration ofradioackkvematerialinliquidwasteatthecommon l release point determined from the Radioactive Liquid Waste Sampling and Analysis Program, Table 4.11-1 in the Radiological Effluent Technical Specifications.
All dilution factors beyond the sample point (s) are in-cluded in the F g term.
2.5.3.2 Dose Factor Related to Liquid Effluents Calculating dose contributions via Equation (2.13) requires the use of a dose factor A for each nuclide, i, which embodies the dose factors,ihathwaytransfer factors (e.g., bioaccumulation factors), pathiSay usage factors, and dilution factors for the points of pathway origin. The adult total body dose factor and the maxi-mum adult organ dose factor for each radionuclide will be used from Table E-11 of Regulatory Guide 1.109; thus r the list contains critical organ dose factors for
(_3
/ various organs. The dose factor is calculated accord-ing to:
Aiz = k 3(Uy/Dy + UFBFg )DFi (2.14)
Where:
A*=
1 composite dose parameter for the whole body or critical organ of an adult for nuclide, i, for
({) all appropriate pathways, as (mrem /hr) per (pci/ml).
kg = units conversion factor, derived according to:
1.14E05 = (1E06pCi/pCi x lE03ml/kg) + 8760 hr/yr.
U = adult fish consumption factor, equal to F
21kg/yr (Regulatory Guide 1.109, Table E-5).
BF. = Bioaccumulation factor for nuclide, i, in fish 1
(Table 2), as pCi/kg per pCi/2.
DF. = Dose conversion factor for nuclide, i, for 1
adults in pre-selected organ, t, as (mrem /pCi)
(Regulatory Guide 1-109, Table E-ll).
15 -
Rev. O i
l U" = receptor individual's water consumption by age !
group as per Regulatory Guide 1.109, Table E-5. For adults, Uy = 730kg/yr.
D" = dilution factor from the near field area within one-quarter mile of the release point to the potable water intake for the adult water consumption.
NOTE The nearest municipal potable water intake downstream from the liquid effluent discharge point into the Mis-souri River is located near the city of St. Louis, Mo.,
approximately 78 miles downstream. Therefore, it is not nycessary to e/aluate (U /Dy ) at this time, and Equation (2.14) simplies to:y O
A it
=k g (UFBFi)DF g (2.15)
Inserting the appropriate usage factors from Regulatory Guide 1.109 into Equation (2.15) yields the following expression:
(~}
s-A f7 = 1.14E05 (21BFi)DF f (2.16) or A ~(2.17) f7 = 2.39E06 x BFf x DFf I
L
Rav. 0 2.5.4 Summary, Determination of Individual Dose Due to Liquid Effluents The dose contribution for the total time period m
Iatt A=1
, is determined by calculation for each monthly . period (at least once per 31 days) and a cumulative summation
- of these total body and organ doses is maintained for i each calendar quarter. The projected dose contribution -
l from batch releases for which radionuclide concentra-tions are determined by periodic composite and grab sample analysis, as stated in Table 4.11-1 of the Radi-ological Effluent Technical Specifications, may be ap-proximated by using the last measured value. However, ,
, n for reporting purposes, the calculated dose contribu-U tien from those radionuclides is based on actual composite / grab sample analysis. Dose contributions are determined for all radionuclides identified in liquid
- effluents released to unrestricted areas. Nuclides I
which are below the LLD for the analyses are reported
- ' as "less than" the nuclide's LLD, and are not' reported as being present at the LLD level for that nuclide.
The "less than" values are not used in the required dose calculations.
O
Rnv. O TABLE 1 INGESTION EOSE CCMMITMENT FACTOR (Air) FOR ADULT AGE GROUP (mrem /hr) per (uCi/ml) l I I Total i .1 I I Nuclidel Bone l Liver ! Body i Thyroid i Kidney l Lung l GI-LLI l H-3 lNo Data l2.2EE-01 l2.26E-Oll2.26E-01 12,26E-Oll2.26E-Oll2.26E-Oll l C-14 !3,13E+0416.26E+03 16.26E+0336.26E*03 16.26E+0312.26E+0316.26E+031 i N -24 l4.07E+0214.07E+02 14.07E+0214.07E+02 14.07E+0214.07E+0214.07E+021 i P-32 i4.62E+0712.672+06 11.78E+06lNo Data lNo Data lNo Lata 15.19E+061 l CR-51 lNo Data INo Data 11. 27 E+0 0 ! 7. 62 E-01 13 81-01 11.69E+00l3.2-E+02i l HN-S4 lNo Data )4.38E+03 l8.35E+02lde Data ll.30E+03lNo Data ll.34E+041 l MN-56 luo Data ll.10E+02 11.95E+0llNe Data ll,40EtD2lNJ Data 13.52E+031 INo Data 12.53E+02l2.61E+02l
{s~wjFE-55 16.57E40214.54E+C219.34E+02lNo FE-59 11.04E+0312.44E+03 II.05E+02lNo Data Data lNo Data l6.81E+02lG.13E+C3]
l CO-58 lNo Data 18.94E+01 12.00E+02lNo Data lNo Data lNo Data ll.81E+031 1 CO-60 luo Data 12.57E+02 15.66E+02lNo Data lNo Data lNo Data 14.82E+03l l NI-63 13.llE+04!2.15E+03 ll.04E+031No Data lNo Data lNo Data l4.49E+021 l N1-65 11.26E+02l1.64E+01 l7.48E+90lNo Data JNo Data lNo' Data 14.16E+021 l CU-64 INC Data l1.00E+01 14.69E+00lNo Data 12.52E+01lNo Data 18.52E+02l l ZN-65 12.32E+0417.38E+04 13.33E+04]No Data 14.93E+04lNo Data l4.65E+041 l ZN-69 14.93E+0119.44E+01 16.56E+00lNo Data 16.13E+0llNo Data l1.42E+0ll l BR-84 lNo Data INo Data 15.26E+01lNo Data lNo Data [No Data l4.13E-04l l RB-88 lNo Data 12.90E+02 ll.54E+02lNo Data INo Data lNo Data 14.00E-091 l RB-89 lNo Data l1.92E+02 11.35E+02lNo Data lNo Data lNo Data l1.12E-lli l SR-89 l2.21E+04lNo Data 16.35E+02lNo Data lNo Data lNo Data 13.55E+031 l SR-90 l5.44E+05lNo Data 11.34E+05lNo Data lNo Data INo Data ll.57E+041 SR-91 l4.07E+02lNo Data ll.64E+0llNo Data INo Data lNo Data ll.94E+031
(')lSR-92ll.54E+02lNo Data l6.68E+00lNo Data lNo Data INo Data 13.06E+03l l Y-90 15.75E-OllNo Data l1.54E-02lNo Data INo Data lNo Data l6.10E+031 l Y-91M 15.44E-03lNo Data 12.10E-04lNo Data lNo Data INo Data ll.60E-021 l Y-91 18.43E+00lNo Data l2.25E-OllNo Data INo Data lNo Data 14.64E+03l l Y-92 15.05E-02lNo Data ll.48E-03lNo Data INo Data INo Data 18.85E+021 l ZR-95 l2.40E-Oll7.70E-02 15.21E-02lNo Data 11.21E-OllNo Data 12.44E+021 l ZR-97 l1.33E-02l2.68E-03 11.22E-03lNo Data 14.04E-03lNo Data 18.30E+021 l NB-95 14.47E+02l2.48E+02 11.34E+02lNo Data l2.46E+02lNo Data ll.51E+06l l MO-99 lNo Data 11.03E+02 l1.96E+0llNo Data 12.33E+02lNo Data 12.39E+021 l TC-99M18.87E-0312.51E-02 13.19E-OllNo Data 13.81E-Oll1.23E-0211.48E+011 l RU-10314.42E+00lNo Data ll.90E+00lNo Data ll.69E+0llNo Data 15.17E+02l l RU-105l3.68E-OllNo Data 11.45E-OllNo Data 14.76E+00lNo Data l2.25E+021 l RU-10616.57E+01lNo Data 18.32E+00lNo Data l1.27E+02lNo Data 14.25E+031 e
Rtv. O TABLE 1 (continued)
~'
l l l Total l l l I"~
Nuclidel Bone ! Liver l Body l Thyroid ! Kidney I Luna ! GI-LLI l TE-13212,41E+03tl.56E+03 l1.47E+0311.72E+03 l1.50Et04iNo Data 17.382+041 l I-130 12.71E+0ll8.01E+0] 13.16E+0il6:79E403 11.25E&O21No Data 16.89Etoil
! I-131 11.49E+0212.14E+02 11.22E+0217.00E+04 13,66E+02fNo Lata !5.64E4011 1 I-132 17.29E+00l1.95E+01 16.82E+00l6.82E+02 13.llE40llNo Data (3.66E+001 l I-133 15.10E+0118.87E+01 12. 70 E+0l i l . 30E +0 4 J1.55E+02iNo Data 17,97E+031
()I-134 13.81E+00ll.03E+01 l3.70E+0Ll1.79E+02 11.64E+0llNo Data 19.01E-03l l I-135 l1.59E+0114.16E+01 il.54E+01l2.75E+03 16.58Et0llNo Data 14.70E+011 ;
lCS-134 12.98E+0517.09E+05 15.80E+05lNo Data 12.29E+0517.62E+0411.24E+04l lCS-136 13.12E+0411.23E+05 18.26E+04lNo Data 16.85E+04l9.39Et03)l.40E+04l
-lCS-137 13.82E+0515.22E+05 13.42E+05lNo Data l1.77E+0515.89E+0411.01E+04!
lCS-138 l2.64E+0215.22E+02 12.59E+02lNo Data 13. 84 E+0 213. 7 9E +0 l l 2. 23 E-0 31 IBA-139 19.29E-0116.62E-04 12.72E-02lNo Data 16.19E-0413.76E-04l1.65E+001 lBA-140 ll.94E+02l2.44E-01 il.27E+01lNo Data 18. 31E-02 l 1. 4 0 E-Ol l 4. 00 E+0 21 ILA-140 l1.50E-Oll7.53E-02 ll.99E -02 l No Data lNo Data lNo Data l5.53E+031 ICE-141 12.24E-02l1.51E-02 ll.72E-03lNo Data l7.03E-03lNo Data 15.78E+0ll lCE-143 l3.94E-0312.92E+00 13.23E-04lNo Data ll.28E-03lNo Data ll.09E+021 lCE-144 ll.17E+00l4.88E-01 16.26E-021No Data 12.89E-01lNo Data 13.94E+02l lPR-143 15.50E-0112.21E-01 12.73E-02lNo Data ll.27E-01lNo Data 12.41E+031 lND-147 13.76E-Oll4.35E-01 12.60E-02lNo Data 12.54E-OllNo Data 12.09E+031
(])W-187 12.96E+02l2.47E+02 18.64E+01lNo Data lNo Data lNo Data l8.09E+04l lNP-239 12.84E-0212.80E-03 ll.54E-03lNo Data l8.72E-03lNo Data 15.74E+02l Rnv. O TABLE 2 BIOACCUMULATION FACTORi(BF ) M D IN M E SENCE OF SITE-SPECIFIC DATA" (pCi/kg) per-(pCi/ liter)
BF f
Element Fish (Freshwater)
H 9.0 E - 01 C 4.6 E + 03 <
Na 1.0 E + 02 P 1.0 E + 05 Cr 2.0 E + O2 Mn 4.0 E + 02 i Fe 1.0 E & 02 '
Co 5.0 E + 01 Ni Q' Cu 1.0 5.0 E
E
+
+
O2 01 2n 2.0 E + 03 Br 4.2 E + O2 Rb 2.0 E + 03 Sr 3.0 E 4 01 Y 2.3 E + 01 Zr 3.3 E + 00 Nb 3.0 E + 04 Mo 1.0 E + 01 Tc 1.5 E + 01 Ru 1.0 E + 01 Rh 1.0 E + 01
, Te 4.0 E + 02 I 1.5 E + 01 Cs 2.0 E + 03 Ba 4.0 E + 00 O- La 2.5 E + 01 Ce 1.0 E + 00 Pr 2.5 E + 01 Nd 2.5 E + 01 W l.2 E + 03 Np 1.0 E + 01 (a) Values taken from Regulatory Guide 1.109, Rev 1, Table A-1.
R v. 0 2.6 Liquid Radwaste Treatment System 2.6.1 Radiological Effluent Technical Specification 3.11.1.3 The liquid radwaste treatment system shall be OPERABLE.
The appropriate portions of the system shall be used to reduce the radioactive materials in liquid waste prior to their discharge when the projected doses due to the liquid effluent from the site when averaged over the calendar quarter, would exceed 0.18 mrem to the total body or 0.6 mrem to any organ.
2.5.2 Description of the Liquid Radwaste Treatmen't
, System 2.6.3 Operability of the Liquid Radwaste Treatment _ System i The liquid radwaste system is capable of varying treat-ment, depending on waste type and product desired. It -
is capable of concentrating, gas stripping, and distil-lation of liquid wastes through the use of the evapora-tor system. The demineralization system is capable of removing radioactive ions from solutions to be reused as makeup water. Filtration is performed on certain liquid wastes and it may, in some cases, be the only required treatment prior to release. The system has the ability to absorb halides through the use of char-coal filters prior to their release.
The design and operation requirements of the liquid radwaste treatment system provide assurance that releases of radioactive materials in liquid effluents will be kept "As Low As Reasonably Achievable" (ALARA).
() The operability of the liquid radwaste treatment system ensures this system will be available for.use when liquids require treatment prior to their release to the environment. To determine operability requirements, doses due to liquid releases are projected once each 31 l
days in accordance with the methodology described in Section 2.5.3.
l l
l l
RSv. 0 3.0 GASEOUS EFFLUENTS 3.1 Radiological Effluent Technical Specification 3.3.3.10 The radioactive gaseous effluent monitoring instrumen-tation channels shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Radiological Effluent Technical Specification 3.11.2.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the ODCM.
3.2 Radiological Effluent Technical Specification
[,11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site shall be limited to the following:
() a. For noble gases: Less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin, and
- b. For all radioiodines and for all radioactive materials in particulate form and radionuclides (other than noble gases) with half lives greater than 8 days: Less than or equal to 1500 mrem /yr to any organ from the inhalation pathway only.
3.3 Gaseous Effluent Monitors Noble gas activity monitors, iodine monitors, and par-ticulate monitors are present on the containment build-ing ventilation system, plant unit ventilation system,
, and radwaste building ventilation system.
The alarm / trip setpoint for any gaseous effluent radia-tion monitor is determined based on the instantaneous concentration limits of 10 CFR Part 20, Appendix B, Table II, Column 1, and are applied at the point at which the discharge leaves the restricted area boundary into an unrestricted area.
Each monitor channel is provided with a two level sys-tem which provides sequential alarms on increasing :
radioactivity levels. These setpoints are designated i as alert setpoints and alarm / trip setpoints.
The radiation monitor alarm / trip setpoints for each release point are based on the radioactive noble gases in gaseous effluents. It is not considered practicable to apply instantaneous alarm / trip setpoints to inte-grating radiation monitors sensitive to radioiodines, l
R5v. O radioactive materials in particulate form and radionu-clides other than noble gases. Conservative assump-tions may be necessary in establishing setpoints to ac-count for system variables, such as the measurement system efficiency and detection capabilities during normal, anticipated, and unusual operating conditions, the variability in release flow and principal radionu-clides, and the time lag between alarm / trip action and the final isolation of the radioactive effluent.
Table 4.3-13 of the Radiological Effluent Technical Specifications provides the instrument surveillance requirements, such as calibration, source checking, functional testing, and channel checking.
3.3.1 Continuogs Release Gaceous Effluent Monitors The radiation detection monitors associated with conti-nucus gaseous affluent releases are:
A) t,_ Monitor I.D. Description 0-GT-RE-21 Unit Vent 0-GH-RE-lO Radwaste Building Vent The Unit Vent reanitor continuously monitors the ef-fluent from the unit vent for particulate, iodine (halogen), and gaseous radioactivity. The unit vent, via ventilation exhaust systems, continuously purges various tanks and sumps normally containing low-level radioactive aerated liquids that can potentially gener-ate airborne activity.
The exhaust systems which supply air to the unit vent are from the fuel building, auxiliary building, the ac-gx cess control area, the containment purge, and the con-denser air discharge.
All of these systems are filtered before they exhaust to the unit vent. The unit vent monitor measures ac-tual plant effluents and not inplant concentrations.
Thus, the system continuously monitors downstream of the last point of potential radioactivity entry. The monitoring system consists of an off-line, three-way airborne radioactivity monitor. An isokinetic sampling probe is located downstream of the last point of poten-tial radioactivity entry for sample collection.
The sample extracted by the isokinetic nozzle is passed through the fixed filter (particulate), charcoal filter (iodine), and fixed-volume (gaseous) detector assem-blies and then through the pumping system for discharge back to the unit vent. Indication is provided on the
Rsv. 0 radioactivity monitoring system CRT in the control IOOm.
The Radwaste Building Ventilation effluent monitor con-tinuously monitors for particulate, halogen, and gaseous radioactivity in the effluent duct downstream of the exhaust filter and fans. The sample point is located downstream of the last possible point of radi-oactive influent, including the waste gas decay tank
- discharge line. The flow path provides ventilation ex-haust for all' parts of the building structure and com-ponents within the building and provides a discharge '
path for the waste gas decay tank release line. These components represent potentiel sources for the release of gaseous and air particulate and iodine activities in addition to the drainsge sumps, tanks, and equipmenu purged by the vaste processing system.
4
- The monitoring system consists of a fixed filter par-ticulate monitor, an iodine monitor, and gaseous activ-
, ity monitor.
t The sample is extracted through an iso %inetic nozzle to ensure that a representative sample of the air is ob-tained prior to release to the environment. After passino through the fixed filter (particulate), char- .
coal filter (halegen), and fixed-volume (noble gas) detector assemblies and the pumping system, the sample is discharged back to the exhau'st duct. Indication is provided on the radiation monitoring system CRT in the 4
control room.
This monitor will isolate the waste gas decay tank dis-
, charge line if the radioactivity release rate is above the present limit when the waste gas' discharge valve
(]) has been deliberately or inadvertently opened.
The continuous gaseous effluent monitor setpoints are established using Xe-133 as the controlling isotope.
Since there are two continuous gaseous effluent release points, a fraction of the total MPC will be allocated to each release point. Neglecting the batch releases, i the plant Unit Vent monitor has been allocated 0.7 MPC and the Radwaste Building Vent monitor has been allo-I cated 0.3 MPC. These will be changed as required, but limited to 1 MPC of Xe-133. Therefore, a particular monitor reaching the fractional MPC setpoint would not necessarily mean the MPC limit at the site boundary is being exceeded; the alarm only indicates that the spe-cific release point is contributing a greater fraction of the MPC limit than was allocated to the associated monitor and will constitute an evaluation of both systems.
f
Rpv. 0 3.3.2 Batch Release Gaseous Monitors i
The radiation monitors associated with batch release gaseous effluents are:
Monitor I.D. Description 0-GT-RE-22 Containment Purge System Monitors 0-GT-RE-33 0-GT-RE-31 Containment Atmosphere Radioactivity 0-GT-RE-32 Monitors 0-GH-RE-10 Radwasta Building Vent l
, The Containment Purge System continuously monitors the containment purge- exhaust duct during purge cperations for particulate, iodine, and gaseous radioactivity. .
w The purpose of these monitors is to isolate the con-s) tainment purge systen on high gaseous activity via the ESFAS. These monitors also serve as backup indication for personnel protection and reactor coolant pressure boundary leakage detection for the containment at- :
mosphere radioactivity monitors. 1 The sample points are located outside the containment i between the containment isolation dampers and the con-tainment purge filter adsorber unit. ;
Each monitor is provided with two isokinetic nozzles to l ensure that representative samples are obtained for both normal purge and minipurge flow rates. The sample is extracted through the selected nozzle and then passed through the selector valve, the fixed filter (particulate), charcoal filter (iodine), and fixed-volume gaseous detectors. The sample then passes
({) through the pumping system and is discharged back to the duct.
Indication is provided for each monitor on individual indicators on the radioactivity monitoring system con-trol panel and, through isolated signals, on the radio-activity monitoring system CRT in the control room.
The Containment Atmosphere Radioactivity monitors, con-tinuously monitor the containment atmosphere ~for par-ticulate, iodine, and gaseous radioactivity. They isolate the containment purge system on high gaseous activity via the ESFAS. These monitors also serve for reactor coolant pressure boundary leakage detection and for personnel protection. The containment atmosphere radioactivity monitors provide backup indication for the containment purge monitors.
Rnv. 0 Samples are extracted from the operating deck level (El 2047'-6") through sample lines which penetrate the containment. The monitors are located as close as possible to the containment penetrations to minimize the length of the sample tubing and the effects of sam-ple plate out. The sample points are located in areas which ensure that representative samples are obtained.
Each sample passes through the penetration, then through the fixed filter (particulate), charcoal filter (iodine), and fixed-volume gaseous detector assemblies.
After passing through the pumping system, the sample is discharged back to the containment through a separate penetration.
Indication is provided for each monitor on individual indicators on the radioactivity monitoring nyctea con-trol panel and, through isolated signals, on the radio- '
activity monitoring system CRT in the control room.
p)
(_ The Radwaste Building Vent nonitors are described'in Section 3.3.1.
The batch gaseous effluent monitors setpoints are nor-cally established using Xe-133 as the controlling iso-tope, with the exception of the Radvaste Building Vent monitor, for ahich Kr-85 is normally the controlling isotope.
l A pre-release isotopic analysis will be performed for each batch release to determine the identity and quan-tity of the principal radionuclides. The alarm / trip setpoint(s) will be adjusted accordingly to ensure that the limits of Radiological Effluent Technical Specifi-cation 3.11.2.1 are not exceeded.
ODCM Methodology for the Determination of Gaseous
(]) 3.4 Effluent Monitor Setpoints 3.4.1 Development of ODCM Methodology for the Determi-nation of Gaseous Effluent Monitor Setpoints The alarm / trip setpoint for gaseous effluent monitors l is determined based on the lesser of the whole body dose rate and skin dose rate, as calculated for the un-restricted site boundary.
3.4.1.1 Whole Body Dose Rate Setpoint Calculations To ensure that the limits of Radiological Effluent Technical Specification 3.11.2.1 are met, the alarm / trip setpoint based on the whole body dose rate is calculated according to:
R.v. O Swb < Dwb R wb FFs , (3.1)
Where:
Swb = the response of the gaseous effluent noble gas monitor at the alarm / trip setpoint based on the whole dose rate.
Dwb = Radiological Effluents Technical Specification 3.11.2.1 limit of 500 mrem /yr, conservatively interpreted as a continuous release over a one year period.
F = the safety factor; a conservative factor used 8
to compensate for statistical fluctuations and errors of measurement. (For example, F = 0.5 e
corresponds to a 100% variation.) Defa61t value is F, = 1.0.
F = the allocation factor which will modify the 8
required dilution factor such that simultaneous gaseous releases may be made without exceeding the limits of Radiological Effluent Technical Specification 3.11.2.1.
The default value is 1/n, where n is the num-ber of pathways planned for release.
(pCi/cc) per (mrem /yr) to the whole body, Rwb = determined according to:
({}
Ryg=C+ [(X/Q) I K Qg) g (3.2) 1 Where:
C= monitor reading of a noble gas monitor cor-responding to the sample radionuclide concen-trations for the batch to be released.
~
s
' R;v. 0 5- "
4 s i '%
Concentrations are determined in accordance with Table 4.11-2 of the Radiological Effluent Technical Specifications. The mixture of radionuclides determined via grab sampling of the effluent stream or source is correlated to a calibration factor to determine monitor response. The monitor response is based on concentrations, not release rate, and is in units of (pCi/cc) .
J the highest calculated annual average relative R79=
concentration for any area at or beyogd the unrestricted area boundary,-in (sec/m ).
K. =
1 the whole body dose factor dd'e' to gamma emis-
~
sionsforeachidentifiednogle clide, is (mrem /yr per pCi/m .) .' gas radionu-(Table 3)
() \'Q.=
1 rate of release of noble gas radionuclide, i, in (pCi/sec) t
, _us '
x 3.4.1.2ssSkin Dose Rate Setpoint Calculation st --
To ensule that the Liimits' of Radiological Effluent l
~
Technien1 Specification 3.11.2.1 are met, the alarm / trip setpoint based on the skin dose rate is cal-culated,aceording to: ,,
SgiDRFF, sss (3.3)
~
C)
Where:
s F
s and Fa.are as previously defined.
~
S = the response of the gaseous effluent noble gas g
' monitor at the. alarm / trip setpoint based on the skin dose rate.
D s
= Radiological Efiluents Technical Specification 3.11.2.1 limit of 3000 mrem /yr, conservatively int.erpreted as a continuous release over a one year period.
R:v. O R
s
= (pCi/cc) per (mrem /yr) to the skin, determined according to:
R, = C + [(X/Q) 1I (Li + 1.1M g) Qg] (3.4)
Where:
L. =
1 the skin dose factor due to beta emissions for each (mrem identified noblg)gase
/yr) per (pci/m radionuclide,
,( Table 3) in
)
1.1 = conversion factor: 1 mrad air dose = 1.1 mrem skin dose.
M. =
1 the air dose factor due to gamma emissions for in each (mrad identified noblg) gas(Table
/yr) per (pCi/m .
radionuclide, 3)
C, (X/Q) and Qi are as previously defined.
3.4.1.3 Gaseous Effluent Monitors Setpoint Determination The results of Equation (3.1) and Equation (3.3) are compared. The setpoint is then selected as the lesser
- (]) .
of the two values.
l I
O O TABLE 3 DOSE FACTORS FOR EXPOSURE TO A SEMI-INFINITE CLOUD OF NOBLE GASES" Whole Body Gamma Air Beta Air Dose Factor Skin Dose Factor Dose Factor Dose Factor K L Mi N Radionuclide (mrem /yr)iper (pci/ma ) (mrem /yr)pbr(pci/m3) (mrad /yr) per (pCi/m3) (mrad /yr)peb(pci/m3)
, Kr-83m 7.56 E-02 --- 1.93 E+01 2.88 E+02 Kr-85m 1.17 E+03 1.46 E+03 1.23 E+03 1.97 E+03 Kr-85 1.61 E+01 1.34 E+03 1.72 E+01 1.95 E+03
. Kr-87 5.92 E+03 9.73 E+03 6.17 E+03 1.03 E+04 Kr-88 1.47 E+04 2.37 E+03 1.52 E+04 2.93 E+03 Kr-89 1.66 E+04 1.01 E+04 1.73 E+04 1.06 E+04 Kr-90 1.56 E+04 7.29 E+03 1.63 E+04 7.83 E+03 Xe-131m 9.15 E+01 4.76 E+02 1.56 E+02 1.11 E+03 3
Xe-133m 2.51 E+02 9.94 E+02 3.27 E+02 1.48 E+03 i
Xe-133 2.94 E+02 3.06 E+02 3.53 E+02 1.05 E+03 i Xe-135m 3.12 E+03 7.11 E+02 3.36 E+03 7.39 E+03 w Xe-135 1.81 E+03 1.86 E+03 1.92 E+03 2.46 E+03 Xe-137 1.42 E+03 1.22 E+04 1.51 E+03 1.27 E+04
' Xe-138 8.83 E+03 4.13 E+03 9.21 E+03 4.75 E+03 Ar-41 8.84 E+03 2.69 E+03 9.30 E+03 3.28 E+03 (a) The listed dose factors are derived from Table B-1 in Reg. Guide 1.109 and are for detected in gaseous effluents.
f 4
Rav. 0 3.4.2 Summary, Gaseous Effluent Monitors Setpoint Determination The gaseous effluent monitors setpoints are calculated as described in Section 3.4. However, it should be .
noted that a batch release will alter the flow rate characteristics at the Unit Vent and therefore the concentration as sensed by the monitor. For example, in the case of a mini-purge, the setpoint for the Unit Vent monitor must be re-calculated to include both the continuous and batch sources.
3.5 ODCM Methodology for Determining Dose Contributions From Gaseous Effluents Dose rate calculations are performed for gaseous ef-fluents to ensure compliance with Radiological Effluent Technical Specification 3.11.2.1 as stated in Section 3.2.
[}
3.5.1 Determination of Dose Rate The following methodology is applicable to the location (unrestricted area or beyond) characterized by the values of the parameter (X/Q) which results in the max-imum whole body or skin dose rate. In the event that the analysis indicates a different location for the whole body and skin dose limitations, the location selected for consideration is that which minimizes the allowable release values.
The factors K , L and M. relate the radionuclide air-borneconcentbatibn,stovkriousdoserates,assuminga semi-infinite cloud model, and are tabulated in s Table 3.
3.5.1.1 Noble Gases The release rate limit for noble gases is determined according to the general relationships delineated as l follows:
l Dyg = I (Kg ((X/Q)Q )] < 500 mrem /yr i (3.5) 1 l
l l . . . _ . ._
Rsv. O D, = 1{ [(Li + 1.1 Mi )((X/Q)Q i)] < 3000 mrem /yr (3.6)
Where:
Dwb = Whole body dose rate, conservatively averaged over a period of one year.
K. =
1 Whole body dose factor due to gamma emissions for each identified goble gas radionuclide, in (mrem /yr) per (uCi/m ). (Table 3)
(R7Q) = The highest calculated annual average relative
(]) concentration for any area at or beyond the unrestricted area boundary. I Q.1 =
The release rate of radionuclides, i, in gaseous effluents, from all vent releases in (pci/sec).
D s
= Skin dose rate, conservatively averaged over a period of one year.
L. =
1 Skin dose factor due to beta emissions for each identified noble gas radionuclide, in (mrem /yr) per (pci/m3) (Table 3).
1.1 = Units conversion factor; 1 mrad air dose = 1.1 mrem skin dose.
() M.1
= Air dose factor due to gamma emissions for each identified noble gas radionuclide, in (mrad /yr) per (pci/m3) (Table 3).
3.5.1.2 Radionuclides Other Than Noble Gases
- The release rate limit for all radionuclides and radi-oactive materials in particulate form and radionuclides other than noble gases is determined according to:
Dg =
?P i[(X/Q)Q i] < 1500 mrem /yr (3.7) 1 R;v. O l Where:
Do = Dose rate to any critical organ, in (mrem /yr).
P. = Dose parameter for radionuclides other than 1
noble gases for the inhalation pathway for the child, based on the critical organ, in (mrem /yr) per (pci/m a). (Table 4)
(X/Q) and Qi are as previously defined.
The dose parameter (P ) includes the internal dosimetry of radionuclide, i,akdthereceptor'sbreathingrate, o which are functions of the receptor's age. Therefore the child age group has been selected as the limiting age group.
For the child exposure, separate values of P lated in Table 4 for the inhalation pathway.i are tabu-These values were calculated according to:
Pi = K' (BR) DFA:
(3.8)
O where: ~
K' = Units conversion factor: 1pCi = 1E06pCi.
BR= The breathing rate of the child age group, in (m3/yr). (Regulatory Guide 1.109, Table E-5).
DFA.1 = The maximum organ inhalation dose factor for the child age group for the ith radionuclide, in (mrem /pci). The whole body is considered as an organ in the selection of DFA .
(Regulatory Guide 1.109, Table E-9)g Note: All radiciodines are assumed to be released in elemental form.
l Rnv. 0 TABLE 4 Dose Parameter (Pf ) For Radionuclides Other Than Noble Gases"
- P i
b RADIONUCLIDE InhalatT6n (mrem /yr) Pathway per (pC1 /m3)
H-3 1.12 E + 03 C-14 3.59 E + 04 Cr-51 1.70 E + 04 Mn-54 1.58 E + 06 Fe-59 1.27 E + 06 Co-58 1.11 E + 06 Co-60 7.07 E + 06 Zn-65 9.95 E + 05 Sr-89 2.16 E + 06 O
v Sr-90 Zr-95 1.01 2.23 E
E
+
+
08 06 Mo-99 1.35 E + 05 I-131 1.62 E + 07 I-133 3.85 E + 06 I-135 7.92 E + 05 Cs-134 1.01 E + 06 Cs-136 1.71 E + 05 Cs-137 9.06 E + 05 Ba-140 1.74 E + 06 La-140 2.26 E + 05 Ce-141 5.44 E + 05 Ce-144 1.20 E + 07 (a) dose parameters listed are for radionuclides that may be detected in gaseous effluents. Additional h<, dose parameters not listed may be determined using methodology described in NUREG-0133.
(b) the childs age group determination; Table E-9, Reg.
Guide 1.109, Rev. 1, 1977
R;v. 0 3.5.2 Individual Dose Due To Gaseous Effluents 3.5.2.1 Radiological Effluent Technical Specification 3.11.2.2 The air dose due to noble gases released in gaseous ef-fluents from-the site shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation and,
- b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.
3.5.2.1.1 Noble Gases
() The air dose in unrestricted areas due to noble gases released from the site is determined according to the following methodology:
During any calendar quarter, for gamma radiation:
D g = 3.17 x 10 s? [Mi {(X/Q) Qi + (X/q) qi}] 15 mrad (3.9) 1 During any calendar quarter, for beta radiation:
( 8 D3 = 3.17 x 10 3 1
[Ni {(X/Q) Qi + (X/q) qi}] 1 10 mrad (3.10)
During any calendar year, for gamma radiation:-
g = 3.17 x 108 3 [Mi [(X/Q) Qi + (X/q) qi}] 1 10 mrad (3.11)
D 1
During any calendar year, for beta radiation:
Db = 3.17 x 10 s? 1
[Ni {(X/Q) Qi + (X/q) gi}] 1 20 mrad (3.12) l l l
l u
1 Rav. O Where:
D = Air dose from gamma radiation due to noble 9 gases released in gaseous effluent.
Db= Air dose from beta radiation due to noble gases released in gaseous effluents.
(X/q) = The relative concentration for areas at or beyond the unrestricted area boundary for short term releases (equal to or less than 500 hrs / year).
- q. = The average release of noble gas radionu-1 clides, i, in gaseous effluents from all vent O releases for short-term releases (equal to or less than 500 hrs / year), in (pci). Releases are cumulative over the calendar quarter or year, as appropriate.
N. = The air dose factor due to beta emissions for 1
each identified noble gas radionuclide, i, in (mrad /yr) per (pci/m3). (Table 3)
Q.1 =
The average release of noble gas radionu-clides, i, in gaseous effluents from all vent releases for long-term releases (greater than 500 hrs / year), in (pci). Releases are cumula-tive over the calendar quarter or year, as appropriate.
O (X/Q) = The highest calculated annual average relative concentration for areas at or beyond the unrestricted area boundary for long-term releases (greater than 500 hrs /yr).
3.17E-08 = The inverse of the number of seconds per year.
M i is as previously defined. (Refer to Section 3.4.1.2) 3.5.2.2 Radiological Effluent Technical Specification 3.11.2.3 The dose to an individual from radionuclides and radi-oactive materials in particulate form, and radionu-clides (other than noble gases) with half-lives greater than 8 days in gaseous effluents released from the site shall be limited to the following:
\
Rnv. O
- a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ and,
- b. During any calendar year: Less than or equal to 15 mrem to any organ.
3.5.2.2.1 Radionuclides Other Than Noble Gases The dose to an individual from radioiodines, radioac-tive materials in particulate form, and radionuclides other than noble gases with half-lives greater than 8 days in gaseous effluents released to unrestricted areas is determined by the following expressions:
O During eny ce1 ender querter:
Di =
3.17E-08 { Ri [W Qi + w qi] 1 7.5 mrem (3.13) 1 During any calendar year:
D i
=
3.17E-08 { Ri [W Qi + w qi] i 15 mrem (3.14) 1 Where:
D. = Dose to an individual from radionuclides other 1
(]) than noble gases.
Q.1 =
The releases of radionuclides, radioactive materials in particulate form, and radionu-clides other than noble gases, i, in gaseous effluents, for all vent releases for long-term releases (greater than 500 hrs /yr), in (pci).
Releases are cumulative over the calendar quarter or year as appropriate.
- q. = The releases of radionuclides, radioactive 1
materials in particulate form and radionu-clides other than noble gases, i, in gaseous effluents for all vent releases for short-term releases (equal to or less than 500 hrs /yr),
in (pCi). Releases are cumulative over the calendar quarter or year as appropriate.
R v. O i R.
1
= The dose factor for each identified radionu-clide, i, in m2(mrem /yr) per (pci/sec) or (mrem /yr) per (pci/m a). (Table 5)
W= The dispersion parameter for estimating the dose to an individual at the controlling loca-tion for long-term releases (greater than 500 hrs /yr):
W = (X7Q) for the inhalation pathway, in(sec/m a),
W = (D7Q) for the food and grounc plane
, pathways, in(meters 2),
w= The dispersion parameter for estimating the es dose to an individual at the controlling loca-(s/ tion for short-term releases (equal to or less than 500 hrs /yr):
w = (X/q) for the inhalation pathway, in(sec/m3) w = (D/q) for the food and ground plane pathway, in (meters 2),
3.17 x 108 = The inverse of the number of seconds per year.
(D/Q) = the average relative deposition of the ef-fluent at the unrestricted area boundary, con-sidering depletion of the plume during trans-port, for long term releases (greater than
(]) 500 hrs /yr), in (meters 2).
(D/q) = the relative deposition of the effluent at the unrestricted area boundary, considering deple-tion of the plume during transport, for short term releases (less than or equal to-500 hrs /yr), in (meters 2),
Note: For the direction sectors with existing pathways within 5 miles from the site, the appropriate R. values are used. If no real pathway exists within 5 miles from the center of the building complex, the cow-milk R. value is used, and it is assumed that this pathway ekists at the 4.5 to 5.0 mile distance in the limiting-case sector. If the R. for an existing pathway within l
5 miles is less than a cow-milk Rg at 4.5 to 5.0 miles, then the value of the cow-milk Ri at 4.5 to 5.0 miles is used.
(w- \ I) x_-
TABLE 5 PATHWAY DOSE FACTORS (R g ), FOR RADIONUCLIDES OTHER THAN NOBLE GASES Inhalation Ground Plane Grass-Cow-Milk Meat Vegetation Pathway Pathway Pathway Pathway Pathway (mrem /yr) 3 ("2 *#**/Y#) ("2 "# */Y#) ("2 "# "/Y#) ("2 "#*"/Y#)
per (pCi/m ) per (pCi/sec) per (pCi/sec) per (pCi/sec) per (pC1/sec)
Radionuclide Infant Child Whole Skin Infant Child Child Child Body H-3 (X /Q) 6.47E+02 1.12E+03 ---
2.38E+03 1.57E+03 2.34E+02 4.01E+03 C-14 2.65E+04 3.59E+04 --- ---
2.34E+09 1.20E+09 3.84E+08 8.91E+08 Cr-51 1.28E+04 1.70E+04 4.65E+06 5.50E+06 4.69E+06 5.39E+06 3.26E+05 6.21E+06
, Mn-54 1.00E406 1.58E+06 1. 39E+06 1.63E+09 3.89E+07 2.09E+07 8.02E+06 6.65E+08 Fe-59 1.02E+06 1.27E+06 2.72E+08 3.20E+08 3.93E+08 2.03E+08 4.46E+08 6.68E+08 Co-58 7.77E+05 1.11E+06 3.79E+08 4.44E+08 6.05E+07 7.07E+07 9.59E+07 3.77E+08 Co-60 4.51 E406 7.07E+06 2.15E+10 2.53E+10 2.10E+08 2.40E+08 3.84E+08 2.09E+09 Zn-65 6.47E+05 9.95E+05 7.47E+08 8.59E+08 1.90E+10 1.10E+10 9.99E+08 2.16E+09 Sr-89 2.03E+06 2.16E+06 2.16E+04 2.51E+04 1.26E+10 6.62E+09 4.82E+08 3.60E+10 Sr-90 4.09E+07 1.01E+08 --- ---
1.22E+11 1.12E+11 1.04E+10 1.24E+12 Zr-95 1.75E+06 2.23E+06 2.45E408 2.84E+08 8.27E+05 8.80E+05 6.12E+08 8.85E+08 I-131 1.48E+07 1.62E+07 1.72E+07 2.09E+07 1.05E+12 4.34E+11 5.58E+09 4.76E+10 I-133 3.56E+06 3.85E406 2.45E+06 2.98E+06 9.63E+09 3.78E+09 8.75E+01 8.27E+08 I-135 6.96E+05 7.92E+05 2.51E+06 2.93E+06 1.97E+07 8.49E+06 1.05E+07 Cs-134 7.03E405 1.01E406 6.86E209 8.00E+09 6.81E+10 3.72E+10 1.51E+09 2.63E+10 Cs-136 1.35E+05 1.71E+05 1.53E+08 1.74E+08 5.80E+09 2.77E+09 4.42E+07 2.24E+08 Cs-137 6.12E+05 9. 06E+05 1.03E+10 1.20E+10 6.04E+10 3.23E+10 1.34E+09 2.39E+10 Ba-140 1.6P6+06 1.74E+06 2.05E+07 2.35E+07 2.42E+08 1.18E+08 4.43E+07 2.79E+08 La-140 1.68E+05 2.26E+05 1.92E+07 2.18E+07 1.88E+05 1.90E+05 5.52E+02 3.20E+07 Ce-141 5.17E405 5.44E+05 1.37E+07 1.54E+07 1.37E+07 1.36E+07 1.38E+07 4.08E+08 Cc-144 9.84E+06 1.20E+07 6. 96E+0 7 8.04E+07 1.33E+08 1.32E+08 1.89E+08 1.04E+10
% 09 1.15F40 5 1. 3 5E I O S 3.98E406 4. 62E4 06 3.10E408 1.74E408 2.44E+05 1.65E+07 _ _ _ _ _
Mxv. 0:
The cumulative critical organ doses for a monthly, quarterly or annual evaluation are based on the calcu-lated dose contribution from each specified time period occurring during the reporting period.
3.6 Gaseous Radwaste Treatment System 3.6.1 Radiological Effluent Technical Specification 3.11.2.4 The-Gaseous Radwaste Treatment System and the Ventila-tion Exhaust Treatment System shall be OPERABLE. The appropriate portions of the gaseous radwaste treatment system shall be used to reduce radioactive materials in gaseous waste prior to their discharge when.the projected gaseous effluent air dose due to gaseous ef-fluent releases from the site when averaged over the calendar quarter, would exceed 0.6 mrad for gamma radi-r~
ation and 1.2 mrad for beta radiation. The appropriate portions of the ventilation exhaust treatment system shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases from the site when averaged over the calendar quarter would exceed 0.9 mrem to any organ.
3.6.2 Description of the Gaseous Radwaste Treatment System The gaseous radwaste treatment system and the ventila-tion exhaust system are available for use whenever
. gaseous effluents require treatment prior to being 4
released to the environment. The gaseous radwaste
' treatment system is designed to allow for the retention of all gaseous fission products to be discharged from
(]) the reactor coolant system. The retention system con-sists of eight (8) gas decay storage tanks, six (6) for
, use during normal operations and two (2) for use during shutdown conditions. These systems will provide reas-onable assurance that the releases of radioactive materials in gaseous effluents will be kept ALARA.
3.6.3 Operability of the Gaseous Radwaste Treatment System The operability of the gaseous radwaste treatment sys-tem ensures this system will be available for use when gases require treatment prior to their release to the environment. To determine operability requirements, doses due to gas release are projected once each 31 days in accordance with the methodology described in Section 3.5.2.
R3v. 0 4.0 DOSE AND DOSE COMMITMENT FROM URANIUM FUEL CYCLE SOURCES 4.1 Radiological Effluent Technical Specification 3.11.4 The dose or dose commitment to any member of the public from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrem) over 12 consecutive months.
4.2 ODCM Methodology for Determining Dose and Dese Commitment from Uranium Fuel Cycle Sources For the purpose of calculating the annual population-integrated dose within a 50 mile radius of the site, the whole body, skin and thyroid are considered the O critical organs. Units of man-rem apply to the whole body and skin dose, and units of man-thyroid-rem apply to the summation of thyroid dose.
4.2.1 Calculation of Total Dose The cumulative dose to a member of the public due to radioactive releases from the site can be determined by summing the calculated doses to critical organs (whole body, skin and thyroid) from effluent sources and methodology previously discussed. The methodology pre-sented in Equations (2.13) and (2.14) is used to deter-mine the annual dose to members of the public from liquid effluent releases. Through the use of the methodology presented in Equations (3.9), (3.10),
(3.11), (3.12), and (3.13), the annual dose to members
(') of the public from gaseous effluents is determined.
4.2.1.1 Calculation of Total Dose from Noble Gas Effluents 4.2.1.1.1 Dose to the whole body is calculated according to:
Dwb = 3.17 E-081I Ki [(X/Q)Q i + (X/q)q g] 1 25 mrem (4.1) 41 -
Rnv. O Where: '
whole body dose Dwb =
K., (X/Q), (X/q), Q , and qi are previsouly defined (refer tb Sections 3.5.1.1iand 3.5.2.1.1).
- 4.2.1.1.2 Dose to the skin is calculated according to
Dg = 3.17 E-08 1{ (L 1 + 1.1 Mi )[(X/Q)Qi + (X/q)qi] 1 25 mrem (4.2)
Where:
Ds = skin dose L., M , (X/Q), (X/q), Q and q.
(lefebtoSections3.5.k,.1and3.areaspreviouslydefined 5.2.1.1).
4.2.1.2 Calculation of Total Dose from Radionuclides Other Than Noble Gases 4.2.1.2.1 Dose to the thyroid is calculated according to:
O Dth = 3.17 E-08 { Rf [WQi + wqi] 1 75 mrem (4.3) 1 Where: '
thyroid dose Dth =
R , W, w, Q and q tb Section b,.5.2.2.k)are as previously defined (refer 4.2.1.3 Summary, Determination of Total Dose The methodology developed in Equations (2.13) and (2.14) is used to determine the annual dose to members j
R~v. O of the public from liquid effluent releases. Equations (4.1), (4.2), and (4.3) are used to determine the an-nual dose to members of the public from gaseous effluents.
No attempt has been made to include the dose reduction due to shielding provided by residential structures in the development of Equations (4.1) and (4.2). A shielding factor of 0.7 was used in the calculation of the G5ound Plane Pathway Factor (R.), presented in Table C. "he annual average relative concentration (X/Q) and the average relative deposition rate (lE97) are at the approximate receptor location in lieu of the site boundary for these calculations.
O L
R v. 0 5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING l
5.1 Radiological Effluent Technical Specification 3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12-1. (ODCM Table 6) 5.2 Description of the Radiological Environmental Monitoring Program The environmental monitoring program is intended to act as a background data base for pre-operation and to sup-plement the radiological effluent release monitoring program during plant operation. Radiation exposure to the public from the various specific pathways and direct radiation can be adequately evaluated by this
(]) program.
Some deviations from the sampling frequency may be necessary due to seasonal unavailability, hazardous conditions, or other legitimate basis. Efforts will be made to obtain all required samples within time frame outlines. Any deviation (s) in sampling frequency or location will be documented in the Annual Radiological Environmental Operating Report.
The Environmental samples are collected and analyzed at the frequency outlined in Table 6. Reporting levels and lower limits of detection (LLD) are outlined in Tables 7 and 8. Figures 5.1 through 5.5 provide loca-tions of samples required in Table 6.
O
Q TABLE 6 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations ** Collection Frequency of Analysis
- 1. AIRBORNE 5 Continuous operations Radiciodine canister.
Radioiodine of sampler with sample Analyze at least once and collection as required per 7 days for I-131.
Particulates by dust loading but at least once per 7 days. Particulate Sampler.
Analyze for gross beta radioactivity > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> <
following filter change. i Perform gamma isotopic analysis on each sample when gross beta activity is > 10 times the yearly mean of control samples.
- Perform gamma isotopic e analysis on composite (by
[ location) sample at least once per 92 days.
- 2. DIRECT RADIATION 40,(2 dosimeters for At least one per 31 Gamma Dose. At least i continuously measuring days. once per 31 days.
j and recoridng dose rate at each location.). .
v l
i i
i a s
- Sample locations are given on Figures 5.1-5.5.
TABLE 6 (Continund)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations ** Collection Frequency of Analysis
- 3. WATERBORNE
- a. Surface 2 Composite sample Gamma isotopic analysis collected over a period of each sample. Tritium of i 31 days. analysis of sample at least once per 92 days.
- b. Ground 2 At least one per Gamma isotopic and 92 days. tritium analyses of each sample,
- c. Drinking 1* Grab sample collected Gamma isotopic and Gross and composited at least Beta analyses of each
, once per 31 days. sample. Tritium
- analyses of sample at
' least once per 92 days.
- d. Sediment from 1 At least once per Gamma isotopic analysis Shoreline 184 days. of each sample.
- 4. INGESTION
- a. Milk 4 At least once per Gamma isotopic and 15 days when animals I-131 analysis of each are on Pasture; at sample, least once per 31 days at other times.
l
- No drinking water int'ake within 50 miles of plant discharge. Sample collected from the City of St. Louis, MO water intake.
i l ** Sample locations are given on Figures 5.1-5.5.
I
TABLE 6 (Centinutd)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations ** Collection Frequency of Analysis
- b. Fish and 2 One sample in season, Gamma isotopic Invertebrates or at least once per analysis on edible '
184 days if not seasonal. portions.
One sample of each of the following.
- 1. Bottom Feeder Species
- 2. Predator Species
- c. Food Products 2 At time of harvest. Gamma isotopic and Sample from each loca- I-131 analyses of tion of broad leaf each sample.
vegation.
O 1
- Sample locations are given on Figures 5.1-5.5.
O O TABLE 7 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Reporting Levels Water Airborne Particulate Fish Milk Food Product
, Analysis (pCi/1) or Gases (pCi/m3) (pCi/kg), wet (pCi/l) (pCi/kg, wet) 1 H-3 2 x 104 (a)
Mn-54 1 x 103 3 x 104 Fe-59 4 x 102 1 x 104 h Co-58 1 x 103 3 x 104 m
' Co-60 3 x 102 1 x 104 i
Zr-Nb-95 4 x 102 (b) 1-131 2 0.9 3 1 x 102 Cs-134 30 10 1 x 103 60 1 x 103 Cs-137 50 20 2 x 103 70 2 x 103 Ba-La-140 2 x 102 (b) 3 x 102 (b)
(a) For drinking water samples. Values are from 40 CFR 141.
(b) Total for parent and daughter.
J
O O TABLE 8 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)a.d Water Airborne Particulate Fish Milk Food Products Sediment Analysis (pCi/1) or Gas (pC1/m3) (pCi/kg, wet) (pCi/1) (pC1/kg, wet) (pCi/kg. dry)
Gross Beta 4b 1 x 10-2 3 3g 2000 (1000by 54Mn 15 130 59 pe 30 260 58,60Co 15 130 95Zr-Nb 15c 131 1 lb 7 x 10-2 1 60 134,137Cs 15 (10b), 18 1 x 10-2 130 15 60 150 140Ba-La 15c 15c
Rnv. O TABLE 8 (CONTINUED)
TABLE NOTATION (a) The LLD is the smallest concentration of radi-oactive material in a sample that will be detected with 95% probability with 5% proba-bility of falsely concluding that a blank ob-servation represents a "real" signal.
For a particular measurement system (which may include radiochemical separation):
LLD = 4*00 b b
O E V- 2.22 Y + exp (-Aat)
Where:
LLD = the lower limit of detection as defined above (as picocurie per unit mass or volume).
Sb= the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute).
/~s E= the counting efficiency (as counts per i
transformation).
V= the sample size (in units of mass or volume).
2.22 = the number of transformation per minute per picocurie.
Y= the fractional radiochemical yield (when applicable).
A= the radioactive decay constant for the partic-ular radionuclide and, at = the elapsed time between sample collection (or
-end of the sample collection period) and time of counting (for environmental samples, not plant effluent samples).
1 R v. O The value of S used in the calculation of the LLD for a detection syNtem is based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calcu'ating the LLD for a radionuclide determined by gamma-ray spectrometry, the background includes the typical contributions of other radionu-clides normally present in the samples (e.g.,
potassium-40 in milk samples). Typical values of E, V, Y and at shall be used in the calculations.
e (b) LLD for drinking water.
(c) Total for parent and daughter.
(d) Other peaks which are measurable and identifi-able, together with the radionuclides in Table 8, are identified and reported.
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LEGEND:
O TOWNS
-INTERMITTENT STREAMS CONTINUOUS STREAMS G AOUATIC SAMPLING STATIONS 1 0 O 1, COMPOSITE SURFACE WATER ._____.
MILES UNION ELECTRIC CO.
CALLAWAY PLANT LOC ATION OF AQU ATIC S AMPLING STATIONS FIGURE 5.2
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F-GROUND WATER MILES D-DRINKING WATER UNION ELECTRIC CO.
CALLAWAY PLANT i
GROUNDWATER QU ALITY
REFERENCE:
^ '
MISSOURI STATE HIGHWAY DEPARTMENT, GENERAL HIGHWAY MAP CALLAWAY COUNTY, MISSOURI. OCTOBER 1, 1977. FJGURE 5.3
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l Rnv. 0 l
6.0 METEOROLOGICAL DATA COLLECTION AND PROCESSING 6.1 Meterological Data Collection The Meterological System has been established to meet the requirements of NUREG-0654 and Regulatory Guide 1.23. A 90 meter primary and a 10 meter backup tower have been erected in the vicinity of the plant site.
The primary meterological tower is located in an open field approximately 1.4 miles east-northeast of the plant. The backup tower is located approximately 1 mile west of the plant, adjacent to the EOF.
Meterological instruments are mounted on the primary meterological tower at heights of 10, 60, and 90 meters and on the backup tower at 10 meters. These instru-ments monitor wind speed, wind direction, reference temperature, temperature differential, precipitation, and dew point.
)
A readout of these parameters is available in the Con-trol Room, TSC, and the EOF. Primary tower recorder equipment is housed in a shelter located near the tower and the backup tower equipment is located in the Commu-nications Equipment Room of the EOF. Meterological parameters, along with radiological monitoring and ven-tilation system flows, are input to the Radioactive Release Information System.
There is an emergency electric generator for both tower locations in the event of power failure, and the National Weather Service in Columbia, Missouri is available for backup meterological data.
6.2 Meterological Data Processing Real time windspeed, wind direction, atmospheric sta-bility, and precipitation data is processed and stored for generation of the reports required by Regulatory ;
Guide 1.21 and for calculation of the various relativa i atmospheric concentrations (X/Q) and relative deposi- I tion (D/Q). '
X/Q (undepleted and undecayed, undepleted and decayed, and depleted and decayed) is calculated for each 22.5 sector (as defined in NUREG 0654) at 21 distances. The determination of (X/Q) employs the constant mean wind direction model with building wake correction and as-sumes a ground level release. Radioactive decay and dry deposition are considered during plume transport.
The value of D/Q is determined by solving polynomial regression equations for each of the deposition curves l
l l
1
Rev. 0
.) in Regulatory Guide 1.111 as a function of height of release, stability class, and distance.
The methodology for calculating X/Q and D/Q is con-sistent with that outlined in Regulatory Guide 1.111.
O 9
- f l
l
Rnv. 0 l l
l 7.0 SEMI ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT Regulatory Guide 1.21 requires that potential doses to individuals and populations are calculated using meas-ured effluent and meteorological data. Each Semi An-nual Radioactive Effluent Release Report contains the following information:
- Whole body and significant organ doses to individu-als in unrestricted areas from receiving-water-related exposure pathways.
- Whole body and skin doses to individuals exposed at the point of maximum offsite ground-level concentra-tions of . radioactive materials in gaseous effluents.
- Organ doses to individuals in unrestricted areas from radioactive material in particulate form from all pathways of exposure.
O
- Whole body doses to individuals and populations in unrestricted areas from direct radiation from the facility.
4
- Whole body doses to the population and average doses to individuals in the population from all receiving-water-related pathways.
.
- Whole body doses to the population and average doses to individuals in the population from gaseous ef-fluents to a distance of 50 miles from the site.
2 The specific pathways to man from airborne releases are:
Plume exposure i Q
- Ground exposure Inhalation
- Cow-milk pathway
- Vegetable pathway
- Meat pathway
- Goat-milk pathway The specific pathways to man from liquid releases are:
- Aquatic food chain
- Dose from shoreline deposits
- Dose from swimming
- Dose from boating
- Drinking of potable water (if applicable; refer to Section 2.5.2)
The age groups considered are:
Rev. 0
- Adult
- Teen
- Child
- Infant The organ doses are:
- Bone
- Liver
- Total body
- Thyroid
- Kidney
- Lung
- GI-LLI Dose calculations for the Semi-Annual Radioactive Ef-fluent Release Report are performed consistent with the previously defined methodologies, utilizing the dose commitment factors obtained from Regulatory Guide O 1 1o9-O l
l Rev. 0 8.0 IMPLEMENTATION OF ODCM METHODOLOGY The ODCM provides the mathematical relationships used to implement the Radiological Effluent Technical Specifications.
For routine effluent release and dose assessment, com-puter codes have been developed which employ Regulatory Guide 1.109 calculational techniques to implement the 2
ODCM methodologies. These calculational methods in-clude the same general features as provided in the ODCM. These codes will be verified to produce results consistent with the ODCM methodologies.
O O
l Rev. O I
9.0 REFERENCES
9.1 U.S. Nuclear Regulatory Commission,
" Preparation of Radiological Effluent Techni-cal Specifications For Nuclear Power Plants", )
USNRC NUREG-0133, Washington D. C. 20555, Oc-tober 1978.
9.2 Callaway Radiological Effluent Technical Spe-cifications, Sections 3.3.3.9, 3.3.3.10, 3/4.11 and 3/4.12, as submitted to the U.S.
Nuclear Regulator Commission, May 1982.
9.3 Regulatory Guide 1.109, " Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents For The Purpose of Evaluating Com-pliance With 10 CFR Part 50, Appendix I,
" Revision 1, U.S. Nuclear Regulatory Commis-(]) sion, Washington, D.C. 20555, October 1977.
9.4 Regulatory Guide 1.111, " Methods for Estimat-ing Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases From Light-Water-Cooled Reactors, " Revision 1, U.S.
Nuclear Regulatory Commission, Washington, D.C. 20555, July 1977. '
9.5 Title 10, " Energy", Chapter 1, Code of Federal Regulations; Part 20, U.S. Government Printing Office, Washington, D.C. 20402.
9.6 Title 10, " Energy", Chapter 1, Code of Federal Regulations; Part 50, Appendix I, U.S. Govern-ment Printing Office, Washington, D.C. 20402.
9.7 Regulatory Guide 1.16, " Reporting of Operating Information - Appendix A Technical Specifications", Revision 4, U.S. Nuclear Reg-ulatory Commission, Washington, D.C. 20555, August 1975.
9.8 Regulatory Guide 1.21, " Measuring, Evaluating, I
and Reporting Radioactivity in Solid Wastes l and Releases of Radioactive Materials in i
Liquid and Gaseous Effluents From Light-Water-Cooled Nuclear Power Plants",
Revision 1, U.S. Nuclear Regulatory Commis-sion, Washington, D.C. 20555, June 1974.
9.9 U.S. Nuclear Regulatory Commission, " Criteria For Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in I
L .
Rev. O Support of Nuclear Power Plants", USNRC NUREG-0654, Revision 1, Appendix 2, Washington, D.C. 20555, November 1980.
9.10 U.S. Nuclear Regulatory Commission, " Final En-vironmental Statement Related to the Operation of Callaway Plant, Unit No.1", USNRC NUREG-0813, Section 5.9, Washington, D.C.
20555, January 1982.
9.11 Final ~ Safety Analysis Report, Standardized Nu-clear Unit Power Plant Systems (SNUPPS),
Sections 11 and 12.
9.12 Final Safety Analysis Report, Site Addendum, Sections 2, 11, and 12.
9.13 Regulatory Guide 1.23, "Onsite Meteorology Program," Revision 0, U.S. Nuclear Regulatory
.() Commission, Washington, D.C.
1972.
20555, February, (I
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