ML20082F566
| ML20082F566 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 11/18/1983 |
| From: | Graham C, Schnell D UNION ELECTRIC CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| PROC-831118, ULNRC-688, NUDOCS 8311290147 | |
| Download: ML20082F566 (121) | |
Text
G Rev. 1 UNION ELECi'RIC COMPANY CALLAWAY PLANT OFFSITE DOSE CALCULATION MANUAL
.~
//-/ T
- M
/8 [
Approve'd:
6[.w
/
w Chairman, ORC
/
Date ORC me'eting number
\\h A
I[(/'[g$
Reviewed:
Superingen~ dent, He alth Physics Date Prepared By:
C 6d44*4 ^ -
11/I 8 l83 Date 8311290147 831118 l
PDR ADOCK 05000483 p; g l
^
.n 3 (
78 1
r Rtv. 1 This document contains the following:
Pages.
1 through 99 Tables 1
through 12 Figures 4.1, 5.1A, 5.1B, 5.2A, 5.2B, 5.3
(
I p
l f
4
Rev. 1 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 Continuous Liquid Effluent Monitors 4
2.2.2 Radioactive Liquid Batch Release 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 12 2.4 LIQUID EFFLUENT CONCENTRATION MEASUREMENTS 12 2.4.1 Radiological Effluent Technical Specification 3.11.1.1 12 2.4.2 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-l i
Rtv. 1 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 Description of the Liquid Radwaste Treatment System 21 2.#.3 Operability of the Liquid Radwaste Treatment System 21 3.0 GASEOUS EFFLUENTS 22 3.1 RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATION 3.3.3.1 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-ATION OF GASEOUS EFFLUENT MONITOR SETPOINTS 26 3.4.1 Development of ODCM Methodology for the Determination of Gaseous Effluent Monitor Setpoints 26 3.4.1.1 Total Body Dose Rate Setpoint Calculations 26 3.4.1.2 Skin Dose Rate Setpoint Calculation 28 3.4.1.3 Gaseous Effluent Monitors Setpoint Determination 29
-ii-
Rsv. 1 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 37 3.5.2.1 Radiological Effluent Technical Specification 3.11.2.2 37 3.5.2.1.1 Noble Gases 37 3.5.2.2 Radiological Effluent Technical Specification 3.11.2.3 38 3.5.2.2.1 Radionuclides Other Than Noble Gases 39 3.6 GASEOUS RADWASTE TREATMENT SYSTEM 60 3.6.1 Radiological Effluent Technical Specification 3.11.2.4 60 3.6.2 Description of the Gaseous Radwaste Treatment System 60 3.6.3 Operability of the Gaseous Radwaste Treatment System 60 1
4.0 DOSE AND DOSE COMMITMENT FROM URANIUM FUEL CYCLE SOURCES 61 4.1 RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATION 3.11.4 61 4.2 ODCM METHODOLOGY FOR DETERMINING DOSE AND DOSE COMMITMENT FROM URANIUM FUEL CYCLE SOURCES 61 4.2.1 Identification of the MEMBER 62 OF THE PUBLIC
-iii-
Rev. 1 3'
Table of Contents (continued)
Page 4.2.1.1 Utilization of Areas Within the SITE BOUNDARY 62 4'.2.2 Total Dose From Gaseous Effluents.
63 4.2.3 Total Dose From Direct Radiation 63 4.2.3.1 Direct Radiation From Outside Storage Tanks 63 4.2.3.2 Direct Radiation From the Reactor 66 5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING 67 5.1 RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATION 3.12.1 67
5.2 DESCRIPTION
OF THE RADIOLOGICAL-ENVIRONMENTAL MONITORING PROGRAM 67 6.0 DETERMINAITON OF ANNUAL AVERAGE AND SHORT TERM ATMOSPHERE DISPERSION PARAMETERS 83 4
6.1 ATMOSPHERE DISPERSION PARAMETERS 83 6.1.1 Long Term Diffusion Estimates 83 6.1.1.1
-The PUFF Model 83 6.1.1.2 The Straight-Line Gaussian Diffusion Model 84 6.1.2 Short Term Diffusion Estimates 87 7.0 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 93 8.0 IMPLEMENTATION OF ODCM METHODOLOGY 95
9.0 REFERENCES
96.
-iv-
_=.. _ _ _ -.
Rsv. 1 List of Figures Figure 4.1 Site Area Closed to Public Use f
Figure 5.1A Radiological Air Sampling Network f
Figure 5.lB Radiological Air Sampling Network
. Figure 5.2A Location of Aquatic Sampling Stations Figure 5.2B Location of Aquatic Sampling Stations Figure 5.3 Milk Sampling Locations i.
l j
i a
i
-V-
R v.
1 List of Tables Page Table 1 Ingestion Dose Commitment Factor (AgT) for Adult Age Group 18 Table 2 Bioaccumulation Factor (BF.) Used in the Absence of Site-Spebific Data 20 Table 3 Dose 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 Radionuclides Other Thah Noble Gases 42 Table 6 Radiological Environmental Monitoring Program 68 Table 7 Reporting Levels for Radioactivity Concentrations in Environmental Samples 79 Table 8 Maximum Values for the Lower Limits of Detection 80 Table 9 Highest Annual Average Atmospheric Dispersion Parameters - Radwaste Building Vent 89 Table 10 Highest Annual Average Atmospheric Dispersion Parameters - Unit Vent 90 Tuble 11 Standard Deviation of Annual Average Dispersion Parameters 91 Table 12 Application of Atmospheric Dispersion Parameters 92
-vi-
Rev. 1
~
Record of Revisions 4
Revision Number Date Reason for Revision I
Rev. O March 1983 i
Rev. 1 November 1983 Revised to support the current RETS submittal i
and to incorporate NRC Staff comments i
4 i
a i
i l
5 vii-l
- P-m f
yrr www gi->;
y m s wgc-ewy ry.<
y,
...----vg-p g s y y up wg y ye.-qq 9-m%-----.-
p % ge rvvww w em'y-+ pay-+u--
eyn-y wy.3..emm
Rev. 1 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 descripticn of the specific sample locations for the radiological environmental monitoring program.
Changes in the calculational methodologies or paramet-ers will be incorporated into the ODCM and documented in the Semi Annual Radioactive Effluent Release Report.
The ODCM does not replace any station implementing procedures.
f Rev. 1 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 setpo.ints of these channels shall be determined in accordance with the methodology ~
and parameters 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 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 Proce-dures, which are a portion of the Plant Operating Manual.
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 F+f
- C (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. 1 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}.
(Ref. 9.8.1)
If no dilution is provided, then c 1 C.
The radioactive liquid waste stream is diluted by the 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 becomes:
c1 C(F + f)
(pCi/ml) f (2.2) --
Rev. 1 2.2.l Continuous Liquid Effluent Monitors The radiation detection monitors associated with conti-nuous liquid effluent releases are (Ref. 9.6.1, 9.6.2):
Monitor I.D.
Description 0-BM-RE-52 Steam Generator Blowdown Discharge Monitor 0-LE-RE-59 Turbine Building Drain Monitor 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 outlet 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 f]uid.
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.
The high radioactivity alarm / trip (alarm and trip) set-point initiates control room alarm annunciation and au-tomatic isolation of the blowdown isolation valves and the blowdown discharge 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.
Each monitor channel is provided with a two level sys-tem which provides sequential alarms on increasing i
radioactivity levels.
These setpoints are designated as alert setpoints and alarm / trip setpoints.
(Ref.
9.6.3) _,., _,. _ _ - - _, _ _,,, _ _... - _,. _, _ _. _. _. ~, _. _. _, _ _ _ _ _ _., _.., _,.
Rtv.'1 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 BOUNDARY.
The alert set-points have been administratively established below the alarm / trip setpoints, thus providing an additional mar-gin of safety.
The alarm / trip setpoint calculations are based on the minimum dilution flow rate (5000 gpm), the maximum ef-fluent stream flow rate, and the actual isotopic analysis.
Due to the possibility of a simulataneous release from more than one release pathway, a portion of the total site release limit is allocated to each pathway.
The determination and usage of the allocation factor is discussed in Section 2.3.1.
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 actual isotopic analysis.
This eval-uation 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 (Ref.
9.6.4, 9.6.5):
MONITOR I.D.
Description O-HB-RE-18 Liquid Radwaste Discharge Monitor O-HF-RE-45 Secondary Liquid Waste System Monitor 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.
The sample point is located to ensure that all potenti-ally radioactive fluids from the liquid radwaste processing system are monitored prior te discharge.
The secondary liquid waste system discharge radioactiv-ity monitor continuously monitors secondary liquid _ - - -
Rnv.
1-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 isolation of the secondary liquid waste system dis-charge valve to prevent the discharge of radioactive fluid.
The sample location ensures that all potenti-ally radioactive sources from the system are monitored prior to discharge.
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 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.
(Ref. 9.8.1) 2.3.1 Development of ODCM Methodology for the Determination of Liquid Effluent Monitor Setpoints The isotopic concentration of the release being consid-ered must be determined.
This is obtained from the sum of the measured concentrations as determined by the analysis required per Radiological Effluent Technical Specifications Table 4.11-1:
C= ({ (Cg)f) + C S+
t+
F
+
(.3)
T a
Where:
the total concentration of radionuclides as C
=
T determined by the analysis of the waste sample.
Rnv. 1
{(C9)1. = the sum of the concentrations (C ) of each measured gamma emitting nuclide 8bserved by gamma-ray spectroscopy of the waste sample.
the measured concentrations (C of alpha em-itting nuclides observed by gr8s)s alpha C"*
=
analysis of the monthly composite sample.
C*=
the measured concentrations of Sr-89 and Sr-90 s
in liquid waste as determined by analysis of the quarterly composite sample.
C*=
the measured concentration of H-3 in liquid t
waste as determined by analysis of the monthly composite sample.
C*=
the measured concentration of Fe-55 in liquid y
waste as determined by analysis of the quar-terly composite sample.
The C term is included in the analysis of each batch; 9
terms for alpha, strontium, Fe-55, and tritium are in-cluded 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.
The measured radionuclide concentrations are used to calculate a Dilution Factor (Fa), which is the ratio of total dilution flow rate to taMk 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 according to:
3
( g)1 hF C
C C
+
a
+
g t
+
F F
I d
"l S
C g
a s
t p
)
Where:
k measured concentrations of as C,
C",
C, C
C
=
9 8efibe,d [n 2.3.1.1.
Terms C, C C
and C will be included in the calc 81atfo,n $s, t
appropriate.
= are limiting concentra-MPC9, MPCtions8ftheapprohriateradionuclidesfrom F - -
Rev. 1 10CFR 20, Appendix B, Table II, Column 2.
For dissolved or entrained noble gases, the concentration shall be limited to 2x10~4 pCi/ml total activity.
the safety factor; a conservative factor used F
=
s to compensate for statistical fluctuations and errors of measurements.
(For example, F
= 0.5 corresponds to a 100 percent v5riation.)
Default value is F
= 0.9.
g For the case Fg< 1, the monitor tank effluent concen-tration meets me limits of Radiological Effluent Tech-nical Specification 3.11.1.1 without dilution and the effluent may be released at any desired flow rate.
If F
> 1 then dilution is required to ensure compliance wYth Radiological Effluent Technical Specification 3.11.1.1 concentration limits.
If simu?:aneous releases are occuring or are anticipated, a modified dilution factor (Fd ), must be determined so that available dilution flow may be apportioned among simultaneous discharge pathways.
Fdn = Fd+F (2.5) a Where:
F" =
the allocation factor which will modify the required dilution factor such that simultaneous liquid releases may be made without exceeding the limits of Radiological Effluent Technical Specification 3.11.1.1.
The most straight-forward determination of the alloca-tion factor is:
a (2.6) where:
Rnv. 1 n=
the number of liquid discharge pathways for which Fa > 1 and which are planned for simultaMeous release.
However, this value for F may be unnecessarily res-trictive in that all rele$se pathways are apportioned the same fraction of the available dilution stream, regardless of the re]ative 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.
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 < l.
The calculated maximum permissible waste tank effluent
, is based on the modified dilution flow rate, (f 8En)d the effective dilution flow rate, TNO)" effective dilution flow rate is given by:
- factor, (F
(Feff).
Feff = (0.9)F (2.7) e Where:
the cooling tower blowdown flow rate and/or F
=
e bypass dilution flow.
A conservative value for F able cooling tower blowdow$ would be the minimum allow-of 5000gpm which is used as a default value.
_g-
Rav. 1 Having established the values of F and F the cal-culatedmaximumpermissiblewasteGHnxfloeffa,tecanbe calculated by:
f 1
- ff + fp~
eff (for fp << Feff)
(2.8) max Fdn dn Where:
f
= the expected undiluted effluent flow rate.
p Thus, the effluent flow rate is set at or below f actual effluent pump capacE6y,may be larger than E86 Even though the value of f (f
it does represent the upper limit to the effluent fEo)w, rate whereby the requirements of Radiological Effluent Technical Specif-ication 3.11.1.1 may still be met.
If F
< 1, the ef-g fluent flow rate setpoint may be assignec 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.
The liquid radiation monitor setpoint may now be deter-mined based on the values of C, and f which were specified to provide compliancI with tE8* limits of Radiological Effluent Technical Specification 3.11.1.1.
The monitor response is primarily to gamma yadiation, therefore, the actual setpoint is based on 1(C )
The calculatedmonitorsetpointconcentrationisdEtkr. mined as follows: i
~
a
Rnv. 1 c = h ((Cg)fi pCi (Refer to Note (2.9)
- 1 Following)
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=
max (Refer to Note (2.10) 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 F and F (i.e., reduce effluent flow rate h9 retur$ fadwaste for f
reprocessing).
NOTE If Fg< 1, no further dilution is required and the release may be made without regard to available dilu-tion or to other releases made simultaneously.
How-ever, it is necessary to establish a monitor setpoint which will provide alarm should the release concentra-tion inadvertently exceed Radiological Effluent Techni-cal Specification 3.11.1.1 limits.
This can be accom-plished by establishing the adjustment factor as follows:
A=F (2.11) d
~, ---
v
Ray, y 2.3.2 Summary, Setpoint Calculation Methodology for Liquid Efiluent Monitors The methodology described in 2.3.1 is used to determine setpoints for each of the radiation monitors assigned a liquid process function.
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 in liquid effluents to UNRESTRICTED AREAS shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases.
For dissolved or entrained noble gases, the concentration shall be limited to 2.0 E-04 pCi/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 sampling, each liquid monitor tank is 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 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, Fe-55, 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 accord-ance with Table 4.11-1 of the Radiological Effluent Technical Specifications.
Rcv. 1 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.
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 an Individual from radioactive materials in liquid effluents released, for each unit, to UNRESTRICTED AREAS 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 b.
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.
(Ref. 9.8.3, 9.8.4)
The Callaway Plant's liquid effluents are discharged to the Missouri River.
As there are no notable water in-takes within 50 miles of the dischary point (Ref.
9.7.1, 9.6.6), this pathway does not 12 quire routine evaluation.
Therefore, 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 contribute the additional 5%, which is considered to be negligible.
(Ref. 9.6.7)
Thus, the maximum exposed individual is an adult, receiving 95% of the total dose from eating fish and 5%
of the total dose from recreational activities.
2.5.3 ODCM Methodology for Determining Dose Contributions From Liquid Effluents 2.5.3.1 Calculation of Dose Contributions The dose contributions for the total time period.- _ --.
Rsv. 1 m
IAt g
A=1 are calculated at least one each 31 days and a cumula-i tive summation of the total body and individual organ doses is maintained for each calendar quarter.
These dose contributions are calculated for all radionuclides identified in liquid effluents released to UNRESTRICTED
- AREAS using the following expression (Ref. 9.8.3 )
m D
=I [A I at C
F]
(2.12) 7 i7 g
ig g
1 A=1 Where:
Dr=
the cumulative dose commitment to the total body or any organ, t,
from the liquid ef-fluents for the total period m
IAtg A=1 in mrem.
the length of the Ath time period over which At
=
g C
and F are averaged for all liquid rk$ eases,#nhours.
i Cig =
the average measured concentration of radionu-clide, i, in undiluted liquid effluent during time' period At from any liquid release, in g
(pCi/ml).
the site related ingestion dose commitment A.
=
1*
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 Table 1, as derived through the use of Equation (2.16).
FA =
the near field average dilution factor for Cig during any liquid effluent release.
Defined.-.._
R3v. 1 as the ratio of the maximum undiluted liquid waste flow during release to the product of the average flow from the site discharge structure to unrestricted receiving waters times 89.77.
(89.77 is the site specific ap-plicable factor for the mixing effect of the discharge structure.)
(Ref. 9.5.1)
The term C is the composite undiluted concentration ofradioackkvematerialinliquidwasteatthecommon 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 term.
g 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,lhathway transfer factors (e.g., bioaccumulation factors), pathway 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 is used from Table E-11 of Regulatory Guide 1.109; thus, Table 1 contains critical organ dose factors for various organs.
The dose factor is calculated accord-ing to (Ref. 9.8.4):
A
= k (U /D
+ U BF )DF (2.13) iT g
g y
p i
i 15 -
s w---m rr.
w e,
Rev. 1 Where:
composite dose parameter for the total body or A.
=
lI critical organ of an adult for nuclide, i, for all appropriate pathways, as (mrem /hr) per (pCi/ml).
units conversion factor, derived according to:
k
=
g 1.14E05 = (lE06pCi/pCi x lE03ml/kg) + 8760 hr/yr.
adult fish consumption factor, equal to U
=
p 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/f).
Dose conversion factor for nuclide, i, for DF. =
1 adults in pre-selected organ, r, as (mrem /pCi)
(Regulatory Guide 1-109, Table E-11).
U" =
receptor individual's water consumption by age group as per Regulatory Guide 1.109, Table E-5.
For adults, U
= 730kg/yr.
p dilution factor from the near field area D
=
y 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.
As there are cur-rently no potable water intakes within 50 river miles of the discharge point, the drinking water pathway is not included in dose estimates to the maximally exposed individual, or in dose estimates to the population.
Should future water intakes be constructed within 10 river miles of the discharge point, then this manual will be revised to include this pathway in dose estimates.
(Ref. 9.6. 6 ).
Therefore, it is not neces-sary to evaluate (U p at this time, and Equation (2.13) simpli$s) to:
(2.14) o (U BF )DFi Air "
p f. - _
4 Rev. 1 Inserting the appropriate usage factors from Regulatory Guide 1.109 into Equation (2.14) yields the following 2
expression:
1 A
= 1.14E05 (21EF )DF (2.15) ft i
i or Aiz = 2.39E06 x BFi x DFi (2.16) i 2.5.4 Summary, Determination of Individual Dose Due to Liquid Effluents The dose contribution for the total time period m
IAtE A=l i
is determined by calculation at least once per 31 days and a cumulative summation of the total body and organ doses is maintained for each calendar quarter.
The projected' dose contribution from batch releases for which radionuclide concentrations are determined by periodic composite and grab sample analysis, as stated in Table 4.11-1 of the Radiological Effluent Technical Specifications, may be approximated by using the last 4
measured value.
However, for reporting purposes, the calculated dose contribution from those radionuclides j
is based on actual composite / grab sample analysis.
Dose contributions are determined for all radionuclides identified in liquid effluents released to UNRESTRICTED AREAS.
Nuclides which are below the LLD for the analyses are reported as "less than" the nuclide's Min-imum Detectable Activity (MDA) 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.
f I
I., _ -..__.
~ - -... _ ~ - ~,,, _ -
Rev. 1 TABLE 1 INGESTION DOSE COMMITMENT FACTOR (A _)
FOR ADULT AGE CROUP (mrem-hr per pciihl)
Total Nuclide Bone Liver Body Thyroid Kidney Lung GI-LLI H-3 No Data 2.26E-01 2.26E-01 2.26E-01 2.26E-01 2.26E-01 2.26E-01 C-14 3.13E+04 6.26E+03 6.26E+03 6.26E+03 6.26E+03 2.26E+03 6.26E+03 Na-24 4.07E+02 4.07E+02 4.07E+02 4.07E+02 4.07E+02 4.07E+02 4.07E+02 P-32 4.62E+07 2.87E+06 1.78E+06 No Data No Data No Data 5.19E+06 CR-51 No Data No Data 1.27E+00 7.62E-01 2.81-01 1.69E+00 3.2-E+02 MN-54 No Data 4.38E+03 8.35E+02 No Data 1.30E+03 No Data 1.34E+04 MN-56 No Data 1.10E+02 1.95E+01 No Data 1.40E+02 No Data 3.52E+03 FE-55 6.57E+02 4.54E+02 1.06E+02 No Data No Data 2.53E+02 2.61E+02 FE-59 1.04E+03 2.44E+03 9.34E+02 No Data No Data 6.81E+02 8.13E+03 CO-58 No Data 8.94E+01 2.00E+02 No Data No Data No Data 1.81E+03 C0-60 No Data 2.57E+02 5.66E+02 No Data No Data No Data 4.82E+03 NI-63 3.11E+04 2.15E+03 1.04E+03 No Data No Data No Data 4.49E+02 NI-65 1.26E+02 1.64E+01 7.48E+00 No Data No Data No Data 4.16E+02 CU-64 NO Data 1.00E+01 4.69E+00 No Data 2.52E+01 No Data 8.52E+02 ZN-65 2.32E+04 7.38E+04 3.33E+04 No Data 4.93E+04 No Data 4.65E+04 ZN-69 4.93E+01 9.44E+01 6.56E+00 No Data 6.13E+01 No Data 1.42E+01 BR-84 No Data No Data 5.26E+01 No Data No Data No Data 4.13E-04 RB-88 No Data 2.90E+02 1.54E+02 No Data No Data No Data 4.00E-09 RB-89 No Data 1.92E+02 1.35E+02 No Data No Data No Data 1.12E-ll SR-89 2.21E+04 No Data 6.35E+02 No Data No Data No Data 3.55E+03 SR-90 5.44E+05 No Data 1.34E+05 No Data No Data No Data 1.57E+04 SR-91 4.07E+02 No Data 1.64E+01 No Data No Data No Data 1.94E+03 SR-92 1.54E+02 No Data 6.68E+00 No Data No Data No Data 3.06E+03 Y-90 5.75E-01 No Data 1.54E-02 No Data No Data No Data 6.10E+03 Y-91M 5.44E-03 No Data 2.10E-04 No Data No Data No Data 1.60E-02 Y-91 8.43E+00 No Data 2.25E-01 No Data No Data No Data 4.64E+03 Y-92 5.05E-02 No Data 1.48E-03 No Data No Data No Data 8.85E+02 ZR-95 2.40E-01 7.70E-02 5.21E-02 No Data 1.21E-01 No Data 2.44E+02 l
ZR-97 1.33E-02 2.68E-03 1.22E-03 No Data 4.04E-03 No Data 8.30E+02 NB-95 4.47E+02 2.48E+02 1.34E+02 No Data 2.46E+02 No Data 1.51E+06 MO-99 No Data 1.03E+02 1.96E+01 No Data 2.33E+02 No Data 2.39E+02l TC-99M 8.87E-03 2.51E-02 3.19E-01 No Data 3.81E-01 1.23E-02 1.48E+01 RU-103 4.42E+00 No Data 1.90E+00 No Data 1.69E+01 No Data 5.17E+02 RU-105 3.68E-01 No Data 1.45E-01 No Data 4.76E+00 No Data 2.25E+02 RU-106 6.57E+01 No Data 8.32E+00 No Data 1.27E+02 No Data 4.25E+03 l
18 -
Rev. 1 TABLE 1 (continued) s Total Nuclide Bone Liver Body Thyroid Kidney Luno GI-LLI I
TE-132 2.41E+03 1.56E+03 1.47E+03 1.72E+03 1.50E+04 No Data 7.38E+04 I-130 2.71E+01 8.01E+01 3.16E+01 6.79E+03 1.25E+02 No Data 6.89E+01 I-131 1.49E+02 2.14E+02 1.22E+02 7.00E+04 3.66E+02 No Data 5.64E+01 I-132 7.29E+00 1.95E+01 6.82E+00 6.82E+02 3.11E+01 No Data 3.66E+00 I-133 5.10E+01 8.87E+01 2.70E+01 1.30E+04 1.55E+02 No Data 7.97E+01 I-134 3.81E+00 1.03E+01 3.70E+00 1.79E+02 1.64E+01 No Data 9.01E-03 I-135 1.59E+01 4.16E+01 1.54E+01 2.75E+03 6.68E+01 No Data 4.70E+01 CS-134 2.98E+05 7.09E+05 5.80E+05 No Data 2.29E+05 7.62E+04 1.24E+04 CS-136 3.12E+04 1.23E+05 8.86E+04 No Data 6.85E+04 9.39E+03 1.40E+04 CS-137 3.82E+05 5.22E+05 3.42E+05 No Data 1.77E+05 5.89E+04 1.01E+04 CS-138 2.64E+02 5.22E+02 2.59E+02 No Data 3.84E+02 3.79E+01 2.23E-03 BA-139 9.29E-01 6.62E-04 2.72E-02 No Data 6.19E-04 3.76E-04 1.65E+00 BA-140 1.94E+02 2.44E-01 1.27E+01 No Data 8.31E-02 1.40E-01 4.00E+02 LA-140 1.50E-01 7.53E-02 1.99E-02 No Data No Data No Data 5.53E+03 CE-141 2.24E-02 1.51E-02 1.72E-03 No Data 7.03E-03 No Data 5.78E+01 CE-143 3.94E-03 2.92E+00 3.23E-04 No Data 1.28E-03 No Data 1.09E+02 CE-144 1.17E+00 4.88E-01 6.26E-02 No Data 2.89E-01 No Data 3.94E+02 PR-143 5.50E-01 2.21E-01 2.73E-02 No Data 1.27E-01 No Data 2.41E+03 ND-147 3.76E-01 4.35E-01 2.60E-02 No' Data 2.54E-01 No Data 2.09E+03 W-187 2.96E+02 2.47E+02 8.64E+01 No Data No Data No Data 8.09E+04 NP-239 2.84E-02 2.80E-03 1.54E-03 No Data 8.72E-03 No Data 5.74E+02 l
t._ -
.... ~.
M Rsv. 1 TABLE 2 w
BIOACCUMULATION FACTOR (BF ) USED IN THE ABSENCE f
OF SITE-SPECIFIC DATA"
_(pCi/kg) per (pCi/literl BF
~
i Element Fish (Freshwater)
H 9.0 E - 01 C
4.6 E + 03 Na 1.0 E + O2 P
1.0 E + 05 Cr 2.0 E + O2 Mn 4.0 E + O2 Fe 1.0 E + O2 Co 5.0 E + 01 Ni 1.0 E + O2 Cu 5.0 E + 01 Zn 2.0 E + 03 Br 4.2 E + O2 Rb 2.0 E 4 02 Sr 3.0 E + 01 Y
2.5 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
~
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 3
Np 1.0 E + 01 (a)
Valued taken from Regulatory Guide 1.109, Rev 1, Table A-1.
I
~
4 5
k. _-
Rev. 1 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 and appropriate portions of the system shall be used to reduce releases of radioactivty when the projected doses due to the liquid effluent, from each reactor unit, to UYRESTRICTED AREAS when averaged over 92 days, would exceed 0.18 mrem to the total body or 0.6 mrem to any organ.
2.6.2 Description Of The LIQUID RADWASTE TREATMENT SYSTEM 2.6.3 OPERABILITY Of The LIQUID RADWASTE TREATMENT SYSTEM The LIQUID RADWASTE TREATMENT SYSTEM is capable of varying treatment, depending on waste type and product desired.
It is capable of concentrating, gas strip-ping, and distillation of liquid wastes through the use of the evaporator 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 sys-tem has the ability to absorb halides through the use of charcoal 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.
OPERABILITY is demonstrated through com-pliance with Radiological Effluent Technical Specifica-tions 3.11.1.1 and 3.11.1.2...
Dev. 1 3.0 GASEOUS EFFLUENTS 3.1 Radiological Effluent Technical Specification 3.3.3.1 The radiation monitoring instrumentation channels for plant operations shall be OPERABLE with their Alarm Trip Setpoints within the specified limits.
3.2 Radiological Effluent Technical Specification 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY 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 Iodine - 131 and 133, for tritium, and for all radionuclides in particulate form 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 (alarm & trip) setpoint for any gaseous effluent radiation 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 SITE BOUNDARY.
(Ref. 9.8.6)
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 setnoints and alarm / trip setpoints.
(Ref.
9.6.3)
The radiation monitor alarm / trip setpoints for each release point are based on the radioactive noble gases in gaseous effluents.
It is not consi(lered practicable to apply instantaneous alarm / trip setpoints to inte-grating radiation monitors sensitive to radioiodines, radioactive materials in particulate form and radionu-clides other than noble gases.
Conservative assump-tions may be necessary in establishing setpoints to ac-.-
Rav. 1 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.
(Ref.
9.8.6.)
Table 4.3-13 of the Radiological Effluent Technical Specifications provides the instrument sur-veillance requirements, such as calibration, source checking, functional testing, and channel checking.
3.3.1 Continuous Release Gaseous Effluent Monitors The radiation detection monitors associated with conti-nuous gaseous effluent releases are (Ref. 9.6.8, 9.6.9):
Monitor I.D.
Description 0-GT-RE-21 Unit Vent 0-GH-RE-10 Radwaste Building Vent The Unit Vent monitor 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-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 radioactivity monitoring system CRT in the control room. l
Rnv. 1 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 potential sources for the release of gaseous and air particulate and iodine activities in addition to the drainage sumps, tanks, and equipment purged by the waste processing system.
The monitoring system consists of a fixed filter par-ticulate monitor, an iodine monitor, and gaseous activ-ity monitor.
The sample is extracted through an isokinetic nozzle to ensure that a representative sample of the air is ob-tained prior to release to the environment.
After passing through the fixed filter (particulate), char-coal filter (halogen), and fixed-volume (noble gas) detector assemblies and the pumping system, the sample is discharged back to the exhaust duct.
Indication is provided on the radiation monitoring system CRT in the 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 the methodology described in Section 3.4.
Since there are two continuous gaseous effluent release points, a fraction of the total MPC will be al-located to each release point.
Neglecting the batch releases, the plant Unit Vent monitor has been allo-cated 0.7 MPC and the Radwaste Building Vent monitor has been allocated 0.3 MPC.
These will be changed as required, but limited to 1 MPC.
Therefore, a particu-lar 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 specific release point is contributing a greater frac-tion of the MPC limit than was allocated to the associ-ated monitor and will necessitate an evaluation of both systems.
~
.~-
R;v. 1 3.3.2 Batch Release Gaseous Monitors The radiation monitors associated with batch release gaseous effluents are (Ref. 9.6.9, 9.6.10, 9.6.11):
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 Radwaste Building Vent The Containment Purge System continuously monitors the containment purge exhaust duct during purge operations for particulate, iodine, and gaseous radioactivity.
The purpose of these monitors is to isolate the con-tainment purge system 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.
The sample points are located outside the containment between the containment isolation dampers and 'he con-t tainment purge filter adsorber unit.
Each monitor is provided with two isokinetic nozzles to 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.
Rtv. 1 Samples are extracted from the operating deck level (El. 2047'-6") through sample lines which pcnetrate 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 system con-trol panel and, through isolated signals, on the radio-activity monitoring system CRT in the control room.
The Radwaste Building Vent monitors are described in Section 3.3.1.
The batch gaseous effluent monitors setpoints are nor-mally established using the methodology described in Section 3.4.
A pre-release isotopic analysis is performed for each batch release to determine the identity and quantity of the principal radionuclidas.
The alarm / trip setpoint(s) are adjusted accordingly to ensure that the limits of Radiological Effluent Technical Specification 3.11.2.1 are not exceeded.
3.4 ODCM Methodology for the Determination of Gaseous 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 is determined based on the lesser of the total body dose rate and skin dose rate, as calculated for the SITE BOUNDARY.
3.4.1.1 Total 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 total body dose rate is calculated according to:
Rnv. 1 S
iDtb tb g a R
FF (3.1) tb Where:
Stb =
the response of the gaseous effluent noble gas monitor at the alarm / trip setpoint based on the total body dose rate (pci/cc).
Dtb =
Radiological Effluents Technical Specification 3.11.2.1 limit of 500 mrem /yr, conservatively interpreted as a continuous release over a one year period.
the safety factor; a conservative factor used F =
s to compensate for statistical fluctuations and errors of measurement.
(For example, F
= 0.5 corresponds to a 100% variation.)
Defa61t value is F
= 1.0.
g F" =
the allocation factor which will modify the 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.
Rtb =
(pCi/cc) per (mrem /yr) to the total body, determined according to:
Rtb =
+ [(X/Q) { K Qg]
(3.2) i 1
Where:
C=
monitor reading of a noble gas monitor cor-responding to the sample radionuclide concen-trations for the batch to be released.
Concentrations are determined in accordance with Table 4.11-2 of the Radiological Effluent Technical Specifications.
The mixture of L
radionuclides determined via grab sampling of the effluent stream or source is correlated to - -..
Y Rev. 1 5
a calibration factor to determine monitor response.
The monitor response is based on concentrations, not release rate, and is in units of (pCi/cc).
X/Q =
the highest calculated annual average relative j
concentration for any agea at or beyond the SITE BOUNDARY in (sec/m ).
K. =
the total body dose factor due to gamma emis-1 sions for each identified noblg) gas radionu-clide, in (mrem /yr) per (pCi/m (Table 3) a i
Q. =
rate of release of noble gas radionuclide, i,
1 in (pci/sec).
Q is calculated as the product of the appro-pbiateventilationpathdesignflowrateand the effluent stream activity concentration.
3.4.1.2 Skin Dose Rate Setpoint Calculation I
To ensure that the limits of Radiological Effluent Technical Specification 3.11.2.1 are met, the alarm / trip setpoint. based on the skin dose rate is cal-culated according to:
i S
< DRFF, (3.3) s sss i-Where:
F and F, are as previously defined in Section 3.4.1.1.
s the response of the gaseous effluent noble gas S
=
s
. monitor at the alarm / trip setpoint based on the skin dose rate.
Radiological Effluents Technical Specification D
=
s 3.11.2.1' limit of-3000 mrem /yr, conservatively interpreted as a continuous release over a one year period.
1 28 -
4
-,vr
--,,+.-y-,
y-
,y..,
r,.
y-,.,,
.y,.,.,
.,7-,-g y.
,,,,,-.,,fg,,
,,,w-m-.wm.
.-.,,,y%.,
,,~r,
- -w g,--
.m,,.m-
,--r-,,,
Rev. 1 (pCi/cc) per (mrem /yr) to the skin, determined R
=
s according to:
R
=C+
[(X/Q) I (Li + 1.1M ) Qg]
(3.4) s g
1 Where:
the skin dose factor due to beta emissions for L. =
1 in each identified noblg) gas radionuclide, (mrem /yr) per (pCi/m (Table 3) 1.1 =
conversion factor:
1 mrad air dose = 1.1 mrem skin dose.
the air dose factor due to gamma emissions for M.
=
1 each identified noblg) gas radionuclide, in (mrad /yr) per (pci/m (Table 3) are as previously defined.
C, (X/Q) and Qi 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.
29 -
R *N. 1 TABLE 3 DOSE FACTORS FOR EXPOSURE TO A SEMI-INFINITE CLOUD OF NOBLE GASES" Total Body Gamma Air Beta Air Dose Factor Skin Dose Factor Dose Factor Dose Factor K
L M
N Radionuclide (mrem /yr)iper (pCi/m3) (mrem /yr)phr(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 Xe-133m 2.51 E+02 9.94 E+02 3.27 E+02 1.48 E+03 8 Xe-133 2.94 E+02 3.06 E+02 3.53 E+02 1.05 E+03
Xc-135m 3.12 E+03 7.11 E+02 3.36 E+03 7.39 E+03 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.
. =
Rev. 1 3.4.2 Summary, Gaseous Effluent Monitors Setpoint s
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 l
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 i
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 I
(SITE BOUNDARY or beyond) characterized by the values of the parameter (X/Q) which results in the maximum total body or skin dose rate.
In the event that the analysis indicates a different location for the total body and skin dose limitations, the location selected for consideration is that which minimizes the allowable release values.
(Ref. 9.8.7) borne concentba,tibn,s to vh. relate the radionuclide air-The factors K L
and M.
rious dose rates, assuming a semi-infinite cloud model, and are tabulated in
. Table 3.
3.5.1.1-Noble Gases The release rate limit for noble gases is determined according to the following general relationships (Ref.
9.8.7):
[K ((X/Q)Q )] 1 500 mrem /yr (3.5)
Dtb = I f
i 1
t 31 -
1 3
-,-ee7.<
c
,-r--,,w+.
.rq~--y.m,h-~4----,-r 1
y-y
- g m
--A
,m
,y--9_.y,,
.,w,yyw-----ms ym-.-,_.,-,+m%.m e-e-, e-s-w v-.y--y n s
,+,.----e-=~-**r
-r
Rev. 1 D,
= { [(Li +-1.1 M )((X/Q)Q )] 1 3000 mrem /yr (3.6) g i
1 Where:
1Dtb =
Total body dose rate, conservatively averaged over a period of one year.
K. =
Total body dose factor due to gamma emissions 1
for each icentified goble gas radionuclide, in (mrem /yr) per (pci/m ).
(Table 3)
(X/Q) =
The highest calculated annual average relative concentration for any area at or beyond the i
SITE BOUNDARY.
Q. =
The release rate of radionuclides, i, in 1
gaseous effluents, from all vent releases in (pci/sec).
Skin dose rate,-conservatively averaged over a D
=
g period of one year.
4
)
L. =
Skin dose factor due to beta emissions for 1
i each identified noble gas radionuclide, in 3
(mrem /yr) per (pCi/m ) ' Table 3).
1.1 =
Units conversion factor; 1 mrad air dose = 1.1 mrem skin dose.
Air dose factor due to gamma emissions for M. =
1 each identified noble gas radionuclide, in 3
(mrad /yr) per~(pci/m ) (Table 3).
3.5.1.2 Radionuclides Other Than Noble Gases i
The release rate limit for Iodine 131 and 133, for tri-tium, and for all radioactive materials in particulate 4
form with half lives greater than 8 days is determined according to (Ref. 9.8.8):
[
= { P [(X/Q)Qg] $ 1500 mrem /yr (3.7)
D f
g 1 ~.
Rnv. 1 Where:
Dose rate to any critical organ, in (mrem /yr).
D
=
g Dose parameter for radionuclides other than P. =
1 noble gases for the inhalation pathway for the child, based on the critical organ, in (mrem /yr) per (pCi/m3).
(Table 4)
(X/Q) and Qi are as previously defined.
The dose parameter (P.) includes the internal dosimetry of radionuclide, i, ahd the receptor's breathing rate, 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 (Ref. 9.8.9):
(3.8)
Pf = K' (BR) DEAg Where:
K' =
Units conversion factor: 1pCi = lE06 pCi.
BR=
The breathing rate of the child age group, in (m3/yr).
(Regulatory Guide 1.109, Table E-5).
The maximum organ inhalation dose factor for DFA. =
1 the child age group for the ith radionuclide, in (mrem /pci).
The total body is considered as an organ in the selection of DFA.
(Regulatory Guide 1.109, Table E-9)g Note:
All radioiodines are assumed to be released in elemental form.
(Ref.9.8.8) _-
Rav. 1 TABLE 4 a
DOSE PARAMETER (P ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES g
Inhalation Pathway 8
(mrem /yr) per (pC1/m )
NUCLIDE
' BONE LIVER TOTAL BODY THYROID KIDNEY LUNG GI-5LI H-3 ND 1.12E3 1.12E3 1.12E3 1.12E3 1.12E3 1.12E3 C-14 3.59E4 6.73E3 6.73E3 6.73E3 6.73E3 6.73E3 6.73E3 Na-24 1.61E4 1.61E4 1.61E4 1.61E4 1.61E4 1.61E4 1.61E4 P-32 2.60E6 1.14E5 9.88E4 ND ND ND 4.22E4 Cr-51 ND ND 1.54E2 8.55El 2.43El 1.70E4 1.08E3 Mn-54 ND 4.29E4 9.51E3 ND 1.00E4 1.58E6 2.29E4 Mn-56 ND 1.66E0 3.12 E-1 ND 1.67E0 1.31E4 1.23E5 Fe-55 4.74E4 2.52E4 7.72E3 ND ND 1.llE5 2.87E3 Fe-59 2.07E4 3.34E4 1.67E4 ND ND 1.27E6 7.07E4 Co-58 ND 1.77E3 3.16E3 ND ND 1.llE6 3.44E4 Co-60 ND 1.31E4 2.26E4 ND ND 7.07E6 9.62E4 Ni-63 8.21E5 4.63E4 2.80E4 ND ND 2.75ES 6.33E3 Ni-65 2.99E0
- 2. 96 E-1
- 1. 64 E-1 ND ND 8.18E3 8.40E4 Cu-64 ND 1.99E0 1.07E0 ND 6.03E0 9.58E3 3.67E4 Zn-65 4.26E4 1.13E5 7.03E4 ND 7.14E4 9.95E5 1.63E4 Zn-69
- 6. 70 E-2 9.66 E-2 8.92 E-3 ND
- 5. 85 E-2 1.42E3 1.02E4 Br-83
- ND ND 4.74E2 ND ND ND 0
Br-84 ND ND 5.48E2 ND ND ND 0
Br-85 ND ND 2.53El ND ND ND 0
Rb-86 ND 1.98E5 1.14E5 ND ND ND 7.99E3 Rb-88 ND 5.62E2 3.66E2 ND ND ND 1.72El Rb-89 ND 3.45E2 2.90E2 ND ND ND 1.89E0 Sr-89 5.99E5 ND 1.72E4 ND ND 2.16E6 1.67ES Sr-90 1.01E8 ND 6.44E6 ND ND 1.48E7 3.43E5 Sr-91 1.21E2 ND 4.59E0 ND ND 5.33E4 1.74E5 TABLE 4 (Cont'd.)
1 DOSE PARAMETER (P ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES" g
i Inhalation Pathway 8
(mrem /yr) per (pCi/m )
NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG GI-LLI Sr-92 1.31El ND 5.25E-1 ND ND 2.40E4 2.42E5 Y-90 4.llE3 ND 1.llE2 ND ND 2.62E5 2.68E5 Y-91m 5.07 E-1 ND
- 1. 84 E-2 ND ND 2.61E3 1.72E3 Y-91
.9.14E5 ND 2.44E4 ND ND 2.63E6 1.84E5 Y-92 2.04El ND
- 5. 81 E-1 ND ND 2.39E4 2.39E5 Y-9 3 1.86E2 ND 5.llE0 ND ND 7.44E4 3.89E5 Zr-95
- 1. 90E5 4.18E4 3.70E4 ND 5.96E4 2.23E6 6.llE4 Zr-97 1.88E2 2.72E1 1.60El ND 3.89El 1.13E5 3.51E5 Nb-95 2.33E4 9.18E3 6.55E3 ND 8.62E3 6.14E5 3.70E4 Mo-99 ND 1.72E2 4.26El ND 3.92E2 1.35ES 1.27E5 Tc-99m
- 1. 78 E-3 3.48E-3 5.77E-2 ND 5.07E-2 9.51E2 4.81E3 Tc-101 8.10 E-5
- 8. 51E -5 1.08 E -3 ND 1.45E-3 5.85E2 1.63El Ru-103 2.79E3 ND 1.07E3 ND 7.03E3 6.62E5 4.48E4 Ru-105 1.53E0 ND 5.55E-1 ND 1.34E0 1.59E4 9.95E4 Ru-106 1.36E5 ND 1.69E4 ND 1.84E5 1.43E7 4.29ES Ag-110m 1.69E4 1.14E4 9.14E3 ND 2.12E4 5.48E6 1.00E5 Te-125m 6.73E3 2.33E3 9.14E2 1.92E3 ND 4.77E5 3.38E4 Te-127m 2.49E4 8.55E3 3.02E3 6.07E3 6.36E4 1.48E6 7.14E4 Te-127 2.77E0
- 9. 51E -1
- 6. llE -1 1.96E0 7.07E0 1.00E4 5.62E4 Te-129m 1.92E4 6.85E3 3.04E3 6.33E3 5.03E4 1.76E6 1.82E5 t
Te-129
- 9. 77E -2 3.50E-2 2.38E-2 7.14E -2 2.57E-1 2.93E3 2.55E4 Te-131m 1.34E2 5.92El 5.07El 9.77El 4.00E2 2.06E5 3.08E5 Te-131 2.17E -2 8.44E-3 6.59E-3
- 1. 70E -2 5.88E-2 2.05E3 1.33E3 Te-132 4.81E2 2.72E2 2.63E2 3.17E2 1.77E3 3.77E5 1.38E5 I-130 8.18E3 1.64E4 8.44E3 1.85E6 2.45E4 ND 5.11E3 i _ _. -
~
TABLE 4 (Cont'd'.')
a DOSE PARAMETER (P ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES 1
Inhalation Pathway 3
(mrem /yr) per (pCi/m )
NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG GI-LLI I-131 4.81E4 4.81E4 2.73E4 1.62E7 7.88E4 ND 2.84E3 I-132 2.12E3 4.07E3 1.88E3 1.94E5 6.25E3 ND 3.20E3 I-133 1.66E4 2.03E4 7.70E3 3.85E6 3.38E4 ND 5.48E3 I-134 1.17E3 2.16E3 9.95E2 5.07E4 3.30E3 ND 9.55E2 I-135 4.92E3 8.73E3 4.14E3 7.92E5 1.34E4 ND 4.44E3 Cs-134 6.51E5 1.01E6 2.25ES ND 2.30E5 1.21E5 3.85E3 Cs-136 6.51E4 1.71E5 1.16E5 ND 9.55E4 1.45E4 4.18E3 Cs-137 9.07ES 2.2SE5 1.28E5 ND 2.72E5 1.04E5 3.62E3 Cs-138 6.33E2 8.40E2 5.55E2 ND 6.22E2 6.81El 2.70E2 Ba-139 1.84E0 9.84E-4 5.37E-2 ND 8.62E-4 5.77E3 5.77E4 i
Ba-140 7.40E4 6.48El 4.33E3 ND 2.11El 1.74E6 1.02E5 Ba-141 2.19E-1 1.09E-4
- 6. 36 E-3 ND 9.47E-5 2.92E3 2.75E2 Ba-142
- 5. 00 E-2 3.60E-5
- 2. 79 E-3 ND 2.91E-5 1.64E3 2.74E0 La-140 6.44E2 2.25E2 7.55El ND ND 1.83E5 2.26E5 La-142 1.30E0 4.11E-1 1.29E-1 ND ND 8.70E3 7.59E4 Ce-141 3.92E4 1.95E4 2.90E3 ND 8.55E3 5.44E5 5.66E4 Ce-143 3.66E2 1.99E2 2.87El ND 8.36El 1.15E5 1.27E5 Ce-144 6.77E6 2.12E6 3.61E5 ND 1.17E6 1.20E7 3.89E5 Pr-143 1.85E4 5.55E3 9.14E2 ND 3.00E3 4.33E5 9.73E4 Pr-144 5.96E-2 1.85E-2 3.0CE-3 ND 9.77E-3 1.57E3 1.97E2 Nd-147 1.08E4 8.73E3 6.81E2 ND 4.81E3 3.28E5 8.21E4 W-187 1.63El 9.66E0 4.33E0 ND ND 4.11E4 9.10E4 Np-239 4.66E2 3.34El 2.35El ND 9.73El 5.81E4 6.40E4 f
l (a) The child age group determination; Table E-9 Reg. Guide 1.109 Rev. 1, 1977..
Rev. 1 4
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 each unit, to areas at and beyond the SITE BOUNDARY 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 at the SITE BOUNDARY due to noble gases released from the site is determined according to the following methodology (Ref. 9.8.10):
During any calendar quarter, for gamma radiation:
D
= 3.17 E-08
[Mg {(X/Q) Qi + (X/q) qi}] 15 mrad (3.9) g During any calendar quarter, for beta radiation:
Db = 3.17 E-08 { [Ni {(X/Q) Qi + (X/q) qi} ] 110 mrad (3.10) 1 During any calendar year, for gamma radiation:
D
= 3.17 E-08 { [Mi {(X/Q) Qi + (X/q) qi} ] i 10 mrad (3.11) g 1
During any calendar year, for beta radiation:
Db = 3.17 E-08 { [Ni {(X/Q) Qi + (X/q) qi}] 1 20 mrad (3.12) 1 -.
.~,
R*v. 1 Where:
Air dose from gamma radiation due to noble D
=
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 SITE 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 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.
The air dose factor due to beta emissions for N.
=
1 each identified noble gas radionuclide, i, in (mrad /yr) per (pci/m3). (Table 3)
Q. =
The average release of noble gas radionu-1 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.
(X/Q) =
The highest calculated annual average relative concentration for areas at or beyond the SITE BOUNDARY for long-term releases (greater than 500 hrs /yr).
3.17E-08 = The inverse of the number of seconds per year.
M is as previously defined. (Refer to Section 3.4.1.2) i 3.5.2.2 Radiological Effluent Technical Specification 3.11.2.3 The dose to an Individual from Iodine-131 and 133, tri-tium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at and beyond the SITE BOUNDARY shall be limited to the following (Ref.
9.8.10):
a.
During any calendar quarter:
Less than or equal to 7.5 mrem to any organ and, --.
R v. 1 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 Iodine-131 and 133, tri-tium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at and beyond the SITE BOUNDARY, is determined according to the following expressions:
During any calendar quarter:
Di = 3.17E-08 { Ri [W Qi + w qi] 1 7.5 mrem (3.13) 1 During any calendar year:
D -= 3.17E-08 { Rg [W Qg + w qi] i 15 mrem (3.14) f 1
Where:
Dose to an individual from radionuclides other D. =
1 than noble gases.
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 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.
The dose factor for each identified radionu-R.
=
1 in m (mrem /yr) per (pCi/sec) or a
clide, i, (mrem /yr) per (pCi/m3). (Table 5) -
Rev. 1 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= (X/Q) for the inhalation pathway, in(sec/m3).
W= (D/Q) for the food and ground plane pathways, in(meters 2),
w=
The dispersion parameter for estimating the dose to an individual at the controlling loca-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 E-08
= The inverse of the number of seconds per year.
(D/Q) =
the average relative deposition of the ef-fluent at the SITE BOUNDARY, considering depletion of the plume during transport, for long term releases (greater than 500 hrs /yr),
in (meters 2),
(D/q) =
the relative deposition of the effluent at the SITE BOUNDARY, considering depletion of the l
plume during transport, for short term l
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 mlles i
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-l case sector.
If the R.
for an existing pathway within l
l 5 miles is less than a cow-milk R at 4.5 to 5.0 miles, g
l then the value of the cow-milk Rg at 4.5 to 5.0 miles l
is used. (Rev. 9.8.10.)
No attempt has been made to include the dose reduction due to shielding provided by residential structures in l
the development of Equations (3.11), (3.12), and (3.14).
The annual average relative concentration ;
1
Rr.v. 1 (X/Q) and the average relative disposition rate (D/Q) are at the approximate receptor location in lieu of the SITE BOUNDARY for these calculations.
Rev. 1 TABLE 5 PATHWAY DOSE FACTORS (R f) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Inhalation Pathway s
(mrem /yr) per (pCi/m )
NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG GI-LLI H-3 ND 1.12E3 1.12E3 1.12E3 1.12E3 1.12E3 1.12E3 C-14 3.59E4 6.73E3 6.73E3 6.73E3 6.73E3 6.73E3 6.73E3 Na-24 1.61E4 1.61E4 1.61E4 1.61E4 1.61E4 1.61E4 1.61E4 1
P-32 2.60E6 1.14E5 9.88E4 ND ND ND 4.22E4 Cr-51 ND ND 1.54E2 8.55El 2.43El 1.70E4
.l.08E3 Mn-54 ND 4.29E4 9.51E3 ND 1.00E4 1.58E6 2.29E4 Mn-56 ND 1.66E0 3.12E-1 ND 1.67E0 1.31E4 1.23E5 Fe-55 4.74E4 2.52E4 7.72E3 ND ND 1.11E5 2.87E3 Fe-59 2.07E4 3.34E4 1.67E4 ND
!!D 1.27E6 7.07E4 Co-58 ND 1.77E3 3.16E3 ND ND 1.llE6 3.44E4 Co-60 ND 1.31E4 2.26E4 ND ND 7.07E6 9.62E4 Ni-63 8.21E5 4.63E4 2.80E4 ND ND 2.75E5 6.33E3 Ni-65 2.99E0
- 2. 96 E-1
- 1. 64 E-1 ND ND 8.18E3 8.40E4 Cu-64 ND 1.99E0 1.07E0 ND 6.03E0 9.58E3 3.67E4 Zn-65 4.26E4 1.13E5 7.03E4 ND 7.14E4 9.95E5 1.63E4 Zn-69
- 6. 70 E-2
- 9. 66 E-2 8.92E-3 ND
- 5. 85 E-2 1.42E3 1.02E4 Br-83 ND ND 4.74E2 ND ND ND 0
Be-84 ND ND 5.48E2 ND ND ND 0
Br-85 ND ND 2.53E1 ND ND ND 0
Rb-86 ND 1.98E5 1.14E5 ND ND ND 7.99E3 Rb-85 ND 5.62E2 3.66E2 ND ND ND 1.72El Rb-89 ND 3.45E2 2.90E2 ND ND ND 1.89E0 Sr-89 5.99E5 ND 1.72E4 ND ND 2.16E6 1.67ES Sr-90 1.01E8 ND 6.44E6 ND ND 1.48E7 3.43E5 Sr-91 1.21E2 ND 4.59E0 ND ND 5.33E4 1.74E5 l
f l _. - -
Rev. 1 TABLE 5 (Cont'd.)
PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES g
Inhalation Pathway 8
(mrem /yr) per (pCi/m )
J NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG GI-LLI Sr-92 1.31El ND
- 5. 25 E-1 ND ND 2.40E4 2.42E5 Y-90 4.llE3 ND 1.llE2 ND ND 2.62E5 2.68E5 Y-91m 5.07 E-1 ND
- 1. 84 E-2 ND ND 2.81E3 1.72E3 Y-91 9.14E5 ND 2.44E4 ND ND 2.63E6 1.84E5 Y-92 2.04El ND
- 5. 81 E-1 ND ND 2.39E4 2.39ES Y-93 1.86E2 ND 5.11E0 ND ND 7.44E4 3.89E5 Zr-95
- 1. 90E5 4.18E4 3.70E4 ND 5.96E4 2.23E6 6.llE4 Zr-97 1.88E2 2.72E1 1.60El ND 3.89El 1.13E5 3.51E5 Nb-95 2.33E4 9.18E3 6.55E3 ND 8.62E3 6.14E5 3.70E4 Mo-99 ND 1.72E2 4.26El ND 3.92E2 1.35E5 1.27E5 Tc-99m
- 1. 78 E-3 3.48E-3 5.77E-2 ND 5.07E-2 9.51E2 4.81E3 Tc-101-8.10 E-5
- 8. 51E -5 1.08E-3 ND 1.45E-3 5.85E2 1.63El Ru-103 2.79E3 ND 1.07E3 ND 7.03E3 6.62E5 4.48E4 Ru-105 1.53E0 ND 5.55E-1 ND 1.34E0 1.59E4 9.95E4 Ru-106 1.36ES ND 1.69E4 ND 1.84E5 1.43E7 4.29E5 Ag-110m 1.69E4 1.14E4 9.14E3 ND 2.12E4 5.48E6 1.00E5 Te-125m 6.73E3 2.33E3 9.14E2 1.92E3 ND 4.77E5 3.38E4 Te-127m 2.49E4 8.55E3 3.02E3 6.07E3 6.36E4 1.48E6 7.14E4 Te-127 2.77E0
- 9. 51E -1 6.11E-1 1.96E0 7.07E0 1.00E4 5.62E4 Te-129m 1.92E4 6.85E3 3.04E3 6.33E3 5.03E4 1.76E6 1.82E5 Te-129
- 9. 77E -2
- 3. 50E -2 2.38E-2 7.14E -2 2.57E-1 2.93E3 2.55E4 Te-131m 1.34E2 5.92El 5.07El 9.77El 4.00E2 2.06E5 3.08E5
.Te-31 2.17E -2 8.44E-3 6.59E-3
- 1. 70E -2 5.88E-2 2.05E3 1.33E3 1
Te-132 4.81E2 2.72E2 2.63E2 3.17E2 1.77E3 3.77E5 1.38E5 1-130 8.18E3 1.64E4 8.44E3 1.85E6 2.45E4 ND 5.llE3 i
Rev. 1 TABLE 5 (Cont'd.)
PATHWAY DOSE FACTORS '(R ) FOR RADIONUCLIDES OTHER THAN NOBLE CASES g
Inhalation Pathway 8
(mrem /yr) per (pCi/m )
NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG CI-LLI I-131 4.81E4 4.81E4 2.73E4 1.62E7 7.80E4 ND 2.84E3 I-132 2.12E3 4.07E3 1.88E3 1.94E5 6.25E3 ND 3.20E3 I-133 1.66E4 2.03E4 7.70E3 3.85E6 3.38E4 ND 5.48E3 I-134 1.17E3 2.16E3 9.95E2 5.07E4 3.30E3 ND 9.55E2 I-135 4.92E3 8.73E3 4.14E3 7.92E5 1.34E4 ND 4.44E3 Cs-134 6.51E5 1.01E6 2.25E5 ND 2.30E5 1.21ES 3.85E3 Cs-136 6.51E4 1.71E5 1.16E5 ND 9.55E4 1.45E4 4.18E3 Cs-137 9.07E5 2.25ES 1.28E5 ND 2.72E5 1.04E5 3.62E3 Cs-138 6.33E2 8.40E2 5.55E2 ND 6.22E2 6.81El 2.70E2 Ba-139 1.84E0
- 9. 84 E-4 5.37E-2 ND 8.62E-4 5.77E3 5.77E4 Ba-140 7.40E4 6.48E1 4.33E3 ND 2.11El 1.74E6 1.02E5 Ba-141 2.19 E-1
- 1. 09 E-4 6.36E-3 ND 9.47E-5 2.92E3 2.75E2 Ba-142
- 5. 00 E-2
- 3. 60 E-5 2.79E-3 ND 2.91E-5 1.64E3
- 2. 74 E0 La-140 6.44E2 2.25E2 7.55El ND ND 1.83E5 2.26E5 La-142 1.30E0 4.11E-1 1.29E-1 ND ND 8.70E3 7.59E4 2
Ce-141 3.92E4 1.95E4 2.90E3 ND 8.55E3 5.44E5 5.66E4 Ce-143 3.66E2 1.99E2 2.87El ND 8.36El 1.15ES 1.27ES Ce-144 6.77E6 2.12E6 3.61E5 ND 1.17E6 1.20E7 3.89E5 Pr-143 1.85E4 5.55E3 9.14E2 ND 3.00E3 4.33E5 9.73E4 4
Pr-144
- 5. 96 E-2
- 1. 85 E-2 3.0CE-3 ND 9.77E-3 1.57E3 1.97E2 Nd-147 1.08E4 8.73E3 6.81E2 ND 4.81E3 3.28E5 8.21E4 W-187 1.63El 9.66E0 4.33E0 ND ND 4.11E4 9.10E4 Np-239 4.66E2 3.34El 2.35El ND 9.73El 5.81E4 6.40E4 (a) The child age group determination; Table E-9 Reg. Guide 1.109, Rev. 1, 1977
)
Rev. 1 TABLE 5 (Contd.)
PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN NOBLE CASES y
Ground Plane Pathway 2
(m mrem /yr) per (uCi/sec)
Nuclide Total Body Skin Na-24 1.19E7 1.39E7 Cr-51 4.65E6 5.51E6 Mn-54 1.39E9 1.63E9 Mn-56 9.03E5 1.07E6 Fe-59 2.72E8 3.20E8 Co-58 3.79E8 4.44E8 Co-60 2.15E10 2.53E10 Ni-65 2.97E5 3.45E5 Cu-64 6.07E5 6.88E5 Zn-65 7.47E8 8.59E8 Br-83 4.87E3 7.08E3 Br-84 2.03E5 2.36E5 Rb-86 8.99E6 1.03E7 Rb-88 3.31E4 3.78E4 Rb-89 1.23E5 1.48E5 Sr-89 2.16E4 2.51E4 Sr-91 2.15E6 2.51E6 Sr-92 7.77E5 8.63E5 Y-90 4.49E3 5.31E3 Y-91m 1.00E5 1.16E5 Y-91 1.07E6 1.21E6 Y-92 1.80E5 2.14E5 Y-93 1.83E5 2.51E5 Zr-95 2.45E8 2.84E8 Zr-97 2.96E6 3.44E6 Nb-95 1.37E8 1.61E8 Mo-94 3.98E6 4.62E6 Te-99m 1.84E5 2.11E5 Tc-101 2.04E4 2.26E4 Ru-103 1.08E8 1.26E8 Ru-105 6.36E5 7.21ES Ru-106-4.22E8 5.07E8 Ag-110m 3.44E9 4.01E9 Te-125m 1.55E6 2.13E6 Te-127m 9.16E4 1.08E5
=.
Rev. 1 TABLE 5 (Contd.)
Ground Plane Pathway (m2 mrem /yr) per (uci/sec)
Nuclide Total Body Skin Te-127 2.98E3 3.28E3 Te-129m 1.98E7 2.31E7 Te-129 2.62E4 3.10E4 Te-131m 8.03E6 9.46E6 Te-131 2.92E4 3.45E7 Te-132 4.23E6.
4.98E6 I-130 5.51E6 6.69E6 I-131 1.72E7 2.09E7 I-132 1.23E6 1.45E6 I-133 2.45E6 2.98E6 1
1-134 4.47ES 5.30E5 j
I-135 2.51E6 2.93E6 Cs-134 6.86E9 8.00E9 Cs-136 1.53E8 1.74E8 Cs-137 1.03E10 1.20E10 Cs-138 3.59E5 4.10E5 Ba-139 1.06E5 1.19E5 Ba-140 2.05E7 2.35E7 Ba-141 4.15E4 4.73E4 Ba-142 4.44E4 5.06E4 La-140
~1.92E7 2.18E7 La-142 7.40E5 8.89E5 Ce-141 1.37E7 1.54E7 Ce-143 2.31E6 2.63E6 Ce-144 6.96E7 8.04E7 4
Pr-144 1.84E3 2.llE3 Nd-147 8.41E6 1.01E7 W-187 2.36E6 2.74E6 Np-239 1.71E6 1.98E6 n
o 4.-
Rev. 1 j
TABLE 5 (Cont'd.)
+
,_,y PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Vegetation Pathway
.(m mrem /yr) per (pCi/sec)
.NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG GI-LLI H-3 ND 4.01E3 4.01E3 4.01E3 4.0kE3 4.01E3 4.01E3 C-14 8.89E8 1.78E8 1.78E8 1.78E8 1.78E8 1.78E8 1.78E8 Na-24 3.28E5 3.28E5 3.28E5 3.28E5 3.28E5 3.28E5 3.28E5 P-32
- 3.37E9 1.57E8 1.30E8 ND ND ND 9.30E7 Cr-51 ND ND 1.17ES 6.50E4 1.78E4 1.19E5 6.21E6
.i Mn-54 ND 6.65E8 1.77E8 ND 1.86E8' ND' 5.58E8 e
~'
Mn-56 ND 1.88El 4.24E0 ND 2.27El ND 2.72E3 Fe-55 8.01E8 4.25E8 1.32E8 ND ND
_2.40E8 7.87E7
Fe-59 3.97E8 6.43E8 3.20E8 ND ND
'1 '. 86E8 6.69E8 Co-58 ND 6.44E7 1.97E8 ND ND ND 3.76E8 Co-60 ND 3.78E8 1.12E9 ND ND ND 2.10E9
. s.
Ni-63 3.95E10 2.11E9 1.34E9 ND ND ND 1.42E8 Ni-65 1.05E2 9.89E0 5.77E0 ND ND ND 1.21E3
-- Cu-64 ND 1.10E4 6.64E3 ND 2.66E4 ND 5.16E5 "2n-65 8.12E8 2.16E9 1.35E9 ND 1.36E9 ND 3.80E8
~
Zn-69 1.09E-5 1.57E-5 1.45E-6 ND 9.52E-6 ND 9.89E-4 Br-83 ND' ND 5.37E0 ND ND ND 0
Br-84 ND' ND 0
ND ND ND 0
Br-85 ND ND
, 0 ND ND ND 0
Rb-86 ND 4.58E8 2.82E8 ND ND ND 0
Rb-88 ND 0
0 ND ND ND 0
Rb-89 ND 0.
O ND ND ND 0
Sr-89 3.59E10 ND 1.03E9
-ND ND ND 1.39E9 Sr 1.24E12 ND-3.15 Ell ND ND ND 1.67E10 Sr-91 5.24E5 ND 1.98E4 ND ND ND 1.16E6 A
t-7'
=
f.
~.
s'
- Rev. 1 TABLE 5 (Cont'd.)
s PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Vegetation Pathway 2
(m mrem /yr) per (pCi/sec) 1r
'q,-.
NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG GI-LLI Sr-92 7.28E2 ND 2.92E1 ND ND ND 1.38E4 Y-90 2.'31E4 ND 6.18E2 ND ND ND 6.57E7 i.
Y-91m 8.87E-9 ND 3.23E-10 ND ND ND 1.74E-5 i'
Y-91' 1.86E7 ND 4.99ES ND ND ND 2.48E9 1.58E0 -
ND 4.53E-2 ND ND ND 4.58E4 Y-92 Y-93 3.01E2 ND 8.25E0 ND ND ND 4.48E6 Zr-95 3.86E6 8.45E5 7.55E5 ND 1.21E6 ND 8.84E8 Zr-97 5.70E2 8.24El 4.86El ND l'18E2 ND 1.25E7 Nb-95 4.10E5 1.59E5 1.14E5 ND 1.50E5 ND 2.95E8 Mo-99 ND 7.71E6
' 91E6 ND 1.65E7 ND 6.38E6
.u Tc-99m 4.71E0 9.24E0 1.53E2 ND 1.34E2 4.69E0 5.26E3 Tc-101 0
0 0
ND 0
0 0
Ru-103 1.54E7 ND 5.90E6 ND 3.87E7 ND 3.97E8 Ru-105 9.16El ND 3.32El ND 8.05E2 ND 5.98E4
[
Ru-106 7.45E8 ND 9.30E7 ND 1.01E9 ND 1.16E10 Ag-110m 3.22E7 2.17E7 1.74E7 ND 4.05E7 ND 2.58E9 Te-125m 3.51E8 9.50E7 4.67E7 9.84E7 ND ND 3.38E8 s
Te-127m 1.32E9 3.56E8 1.57E8 3.16E8 3.77E9 ND 1.07E9 i
Te-127 1.00E4 2.69E3 2.14E3 6.91E3 2.84E4
. ND 3.90E5 Te-129m 8.38E8 2.34E8 1.30E8 2.70E8 2.46E9 ND 1.02E9 Te-129 1.16 E-3
- 3. 23 E-4 2.75E-4 8.26E-4 3.39E-3 ND 7.20 -2 E
Te-131m 1.54E6 5.33E5 5.68E5 1.10E6 5.16Eb ND 2.16E7 Te-131 0
0 0
0 0
ND 0
Te-132 6.98E6 3.09E6 3.73E6 4.50E6 2.87E6 ND 3.11E7 I-130 6.16E5 1.24E6' l.28E5 1.37E8 1.86E6 ND 1.16E5 4 _
Rev. 1 TABLE 5 (Cont'd.)
PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES f
Vegetation Pathway 2
i (m mrem /yr) per (pC1/sec)
NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG C -LLI I-131.
1.43E8 1.44E8 8.17E7 4.75E10 2.36E8 ND 1.28E7 i
1-132 8.58El 1.58E2 7.25El 7.31E3 2.41E2 ND 1.86E2 I-133 3.56E6 4.40E6 1.67E6 8.18E8 7.34E6 ND 1.77E6 I-134 1.55E-4 2.88E-4 1.32E-4 6.62E-3 4.40E-4 ND 1.91E-4 I-135 6.62E4 1.13E5 5.33E4 9.97E6 1.57E5 ND 8.58E4 c
Cs-134 1.60E10 2.63E10 5.55E9 ND 8.15E9 2.93E9 1.42E8 Cs-136 8.17E7 2.25E8 1.45E8 ND 1.20E8 1.78E7 7.90E6 Cs-137 2.39E10 2.29E10 3.38E9 ND 7.46E9 2.68E9 1.43E8 Cs-138 0
0 0
ND 0
0 0
Ba-139 4.80E-21 2.56E-5 1.39E-3 ND 2.24E-5 1.51E-3 2.77E0 i
Ba-140 2.77E8 2.42E5 1.62E7 ND 7.89E4 1.45E5 1.40E8 i
Ba-141 0
0 0
-ND 0
0 0
Ba-142 0
0 0
ND 0
0 0
La-140 3.25E3 1.14E3 3.83E2 ND ND ND 3.17E7 La-142 2.50E-4 7.98E-5 2.50E-5 ND ND ND 1.58E1 Ce-141 6.56E5 3.27ES 4.86E4 ND 1.43E5 ND 4.08E8 Ce-143 1.72E3 9.31E5 1.35E2 ND 3.91E2 ND 1.36E7 Ce-144 1.27E8 3.98E7 6.78E6 ND 2.21E7 ND 1.04E10 Pr-143 1.46E5 4.38E4 7.25E3 ND 2.37E4 ND 1.58E8 Pr-144 0
0 0
ND 0
ND 0
Nd-147 7.17E4 5.81E4 4.50E3 ND 3.19E4 ND 9.20E7 W-187 6.47E4 3.83E4 1.72E4 ND ND ND 5.38E6
- Np-239 2.55E3 1.83E2 1.29E2 ND 5.30E2 ND 1.36E7 4
1 4
i 4
.t -
Rev. 1
)
TABLE 5 (Contd.)
PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN NOBLE CASES GRASS-COW-MILK PATHWAY 2
(m mrem /yr) per (pCi/sec)
NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG GI-LLI H-3 ND 2.38E3 2.38E3 2.38E3 2.38E3 2.38E3 2.38E3 C-14 2.34E9 5.00E8 5.00E8 5.00E8 5.00E8 5.00E8 5.00E8 Na-24 1.55E7 1.55E7 1.55E7 1.55E7 1.55E7 1.55E7 1.55E7 P-32 1.60 Ell 9.42E9 6.21E9 MD ND ND 2.17E9 Cr-51 ND ND 1.61E5 1.05E5 2.30E4 2.05ES 4.70E6 Mn-54 ND 3.90E7 8.84E6 ND 8.64E6 ND 1.43E7 Mn-56 ND 3.15E-2 5.43E-3 ND 2.71E-2 ND 2.86E0 Fe-55 1.35E8 8.73E7 2.33E7 ND ND 4.27E7 1.llE7 Fe-59 2.24E8 3.92E8 1.54E8 ND ND 1.16E8 1.87E8 Co-58 ND 2.42E7 6.05E7 ND ND ND 6.04E7 Co-60 ND 8.82E7 2.08E8 ND ND ND 2.10E8 Ni-63 3.49E10 2.16E9 1.21E9 ND ND ND 1.07E8 Ni-65 3.50E0 3.97E-1 1.80E-1 ND ND ND 3.02El Cu-64 ND 1.85E5 8.59E4 ND 3.14E5 ND 3.81E6 Zn-65 5.55E9 1.90E10 8.78E9 ND 9.23E9 ND 1.61E10 Zn-69 0
0 0
ND 0
ND 3.86E-9 Br-83 ND ND ND ND ND ND ND Br-84 ND ND ND ND ND ND ND Br-85 ND ND ND ND ND ND ND Rb-86 ND 2.23E10 1.10E10 ND ND ND 5.72E8 Rb-88 ND 0
0 ND ND ND 0
Rb-89 ND 0
0 ND ND ND 0
l Sr-89 1.26E10 ND 3.61E8 ND ND ND 2.59E8 Sr-90 1.22 Ell ND 3.10E10 ND ND ND 1.52E9 Sr-91 2.72E5 ND 9.83E3 ND ND ND 3.21ES l
I I
i - -......._
R:v. 1 TABLE 5 (Contd.)
PATHWAYDOSEFACTORS(RgFORRADIONUCLIDESOTHERTHANNOBLEGASES GRASS-COW-MILK PATHWAY 2
(m mrem /yr) per (pCi/sec)
NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG GI-LLI Sr-92 4.64E0 ND 1.72E-1 ND ND ND 5.00El Y-90 6.81E2 ND 1.83El ND ND ND 4.41E5 Y-91m 0
ND 0
ND ND ND 0
Y-91 7.33E4 ND 1.95E3 ND ND ND 5.25E6 Y-92 5.38E-4 ND 1.51E-5 ND ND ND 1.03El Y-93 2.25E0 ND 6.13E-2 ND ND ND 1.78E4 Zr-95 6.80E3 1.66E3 1.18E3 ND 1.79E3 ND 8.26E5 Zr-97 4.06E0 6.97E-1 3.18E-1 ND 7.03E-1 ND 4.45E4 Nb-95 5.93E5 2.44E5 1.41E5 ND 1.75ES ND 2.06E8 Mo-99 ND 2.08E8 4.06E7 ND 3.11E8 ND 6.85E7 Tc-99m 2.80El 5.78El 7.45E2 ND 6.22E2 3.02El 1.68E4 Tc-101 0
0 0
ND 0
0 0
Ru-103 8.67E3 ND 2.90E3 ND 1.80E4 ND 1.05E5 Ru-105 8.05E-3 ND 2.71E-3 ND 5.92E-2 ND 3.20E0 Ru-106 1.90E5 ND 2.38E4 ND 2.25E5 ND 1.44E6 Ag-110m 3.86E8 2.82E8 1.86E8 ND 4.03E8 ND 1.46E10 Te-125m 1.51E8 5.04E7 2.04E7 5.07E7 ND ND 7.18E7 Te-127m 4.21E8 1.40E8 5.10E7 1.22E8 1.04E9 ND 1.70E8 Te-127 6.47E3 2.17E3 1.39E3 5.27E3 1.58E4 ND 1.36ES Te-129m 5.57E8 1.91E8 8.52E7 2.14E8 1.39E9 ND 3.32E8 Te-129 2.10E-9 7.24E-10 4.90E-10 1.76E-9 5.23E-9 ND 1.68E-7 Te-131m 3.38E6 1.36E6 1.12E6 2.75E6 9.35E6 ND 2.29E7 Te-131 0
0 0
0 0
ND 0
Te-132 2.10E7 1.04E7 9.72E6 1.54E7 6.51E7 ND 3.85E7 I-130 3.55E6 7.81E6 3.13E6 8.75E8 8.58E6 ND 1.67E6 - -
Rav. 1 TABLE 5 (Contd.)
PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN NOBEL GASES GRASS-COW-MILK PATHWAY (tr.2 mrem /yr) per (pCi/sec)
NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG GI-LLI I-131 2.72E9 3.20E9 1.41E9 1.05E12 3.74E9 ND 1.14E8 I-132 1.25E0 2.53E0 9.02E-1 1.19E2 2.83E0 ND 2.05E0 I-133 3.67E7 5.34E7 1.56E7 9.72E9 6.28E7 ND 9.04E6 I-134 0
0 0
8.40E-10 0
ND 0
I-135 1.12E5 2.23E5 8.14E4 2.00E7 2.49E5 ND 8.08E4 Cs-134 3.65E10 6.80E10 6.87E9 ND 1.75E10 7.18E9 1.85E8 Cs-136 1.97E9 5.79E9 2.16E9 ND 2.31E9 4.72E8 8.80E7 Cs-137 5.15E10 6.02E10 4.27E9 ND 1.62E10 6.55E9 1.88E8 Cs-138 0
0 0
ND 0
0 0
Ba-139 4.02E-7 2.67E-10 1.16E-8 ND 1.60E-10 1.62E-10 2.55E-5
. Ba-140 2.41E8 2.41E5 1.24E7 ND 5.72E4 1.48E5 5.92E7 Ba-141 0
0' 0
ND 0
0 0
Ba-142 0
0 0
ND 0
0 0
La-140 4.07El 1.60El 4.13E0 ND ND ND 1.88E5 La-142 0
0 0
ND ND ND 6.02E-6 Ce-141 4.34E4 2.64E4 3.llE3 ND 8.15E3 ND 1.37E7 Ce-143 3.97E2 2.63E5 3.00E1 ND 7.67El ND 1.54E6 Ce-144 2.33E6 9.52E5 1.30E5 ND 3.85E5 ND 1.33E8 Pr-143 1.49E3 5.56E2 7.37El ND 2.07E2 ND 7.85E5 Pr-144 0
0 0
ND 0
ND 0
Nd-147 8.82E2 9.06E2 5.55El ND 3.49E2 ND 5.74E5 W-187 6.12E4 4.26E4 1.47E4 ND ND ND 2.50E6 Np-239 3.64El 3.25E0 1.84E0 ND 6.49E0 0
9.40E4 4 _
Rev. 1 TABLE 5 (Cont'd.)
4 PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN NOBEL GASES g
Grass-Goat-Milk Pathway (m mrem /yr) per (uCi/sec)
NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG GI-LLI H-3 ND 4.86E3 4.86E3 4.86E3 4.86E3 4.86E3 4.'86E3
'C-14 2.34E9 5.00E8 5.00E8 5.00E8 5.00E8 5.00E8 5.00E8 Na-24 1.86E6 1.86E6 1.86E6 1.86E6 1.86E6' 1.86E6 1.86E6 P-32 1.92 Ell 1.13E10 7.45E9 ND ND ND 2.60E9 Cr-51 ND.
ND 1.93E4 1.26E4 2.76E3 2.46E4 5.64E5 Mn-54 ND 4.68E6 1.06E6 ND 1.04E6 ND 1.72E6 n
Mn-56 ND
- 3. 78E-3 6.52E-4 ND 3.25E-3 ND 3.43E-1 Fe-55 1.76E6 1.13E6 3.03E5 ND ND 5.55E5 1.44E5 Fe-59 2.92E6 5.09E6 2.01E6 ND ND 1.51E6 2.43E6 Co-58 ND 2.91E6 7.26E6 ND ND ND 7.25E6 4
Co-60 ND 1.06E7 2.50E7 ND ND ND 2.52E7 Ni-63 4.19E9 2.59E8 1.45E8 ND ND ND 1.29E7 Ni-65 4.21E-1 4.76E-2 2.17E-2 ND ND ND 3.62E0
[
Cu-64 ND 2.07E4 9.57E3 ND 3.50E4 ND 4.24E5 Zn-65 6.66E8 2.28E9 1.05E9 ND 1.llE9 ND 1.93E9 Zn-69 0
0 0
ND 0
ND 4.63E-10 i
Br-83 ND ND ND ND ND ND ND
.Br-84 ND ND ND ND ND ND ND Br-85 ND ND ND ND ND ND ND Rb-86 ND 2.68E9 1.32E9-ND ND ND 6.86E7 i
Rb-88 ND 0
0 ND ND ND 0
Rb-89 ND 0
0 ND ND ND 0
Sr-89 2.64E10 ND 7.58E8 ND ND ND 5.43E8 Sr-90 2.55 Ell ND 6.50E10 ND ND ND 3.19E9 Sr-91 5.70E5 ND 2.06E4 ND ND ND 6.75E5 i
I I
f I
E 1
I.
1 i
f i
=. - - _ -.. - -, -.,.
Rev. 1 TABLE 5 (Cont'd.)
PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN NOBLE CASES f
Grass-Goat-Milk Pathway (m mrem /yr) per (pCi/sec)
NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG GI-LLI Sr-92 9.75E0 ND 3.62E-1 ND ND ND 1.05E2 Y-90 8.17El ND 2.19E0 ND ND ND 1.13E5 Y-91m 0
ND 0
ND ND ND 0
Y-91 8.79E3 ND 2.34E2 ND ND ND 6.30E5 Y-92 6.45E-5 ND 1.81E-6 ND ND ND 1.23E0 Y-93 2.70E-1 ND 7.35E-3 ND ND ND 2.13E3 Zr-95 8.16E2 1.99E2 1.41E2 ND 2.14E2 ND 9.91E4 Zr-97 4.87E-1 8.37E-2 3.82E-2 ND 8.43E-2 ND 5.34E3 Nb-95 7.12E4 2.93E4 1.69E4 ND 2.10E4 ND 2.47E7 Mo-99 ND 2.50E7 4.87E6 ND 3.73E7 ND 8.22E6 Tc-99m 3.36E0 6.94E0 8.93El ND 7.46El 3.63E0 2.01E3 Tc-101 0
0 0
ND 0
0 0
Ru-103 1.04E3 ND 3.48E2 ND 2.17E3 ND 1.27E4 Ru-105 9.66E-4 ND 3.25E-4 ND
- 7. llE-3 ND 3.84E-1 Ru-106 2.28E4 hD 2.85E3 ND 2.70E4 ND 1.73E5 Ag-110m 4.63E7 3.38E7 2.24E7 ND 4.84E7 ND 1.75E9 Te-125m 1.81E7 6.05E6 2.45E6 6.09E6 ND ND 8.62E6 Te-127m 5.05E7 1.68E7 6.12E6 1.46E7 1.24E8 ND 2.04E7 Te-127 7.77E2 2.60E2 1.67E2 6.32E2 1.90E3 ND 1.63E4 Te-129m 6.68E7 2.29E7 1.02E7 2.57E7 1.67E8 ND 3.99E7 Te-129 2.52E-10 0
0 2.llE-10 6.27E-10 ND 2.01E-8 Te-131m 4.05E5 1.63E5 1.35E5 3.31E5 1.12E6 ND 2.75E6 Te-131 0
0 0
0 0
ND 0
Te-132 2.52E6 1.25E6 1.17E6 1.84E6 7.82E6 ND 4.62E6 I-130 4.26E6 9.37E6 3.76E6 1.05E9 1.03E7 ND 2.01E6 Rev. 1 TABLE 5 (Cont'd.)
PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES g
Grass-Goat-Milk Pathway (m mrem /yr) per (pCi/sec)
NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG GI-LLI I-131 3.26E9 3.85E9 1.69E9 1.26E12 4.49E9 ND 1.37E8 I-132 1.50E0 3.04E0 1.08E0 1.43E2 3.39E0 ND 2.46E0 1-133 4.40E7 6.41E7 1.88E7 1.17E10 7.54E7 ND 1.09E7 I-134 0
0 0
1.01E-9 0
ND 0
I-135 1.35E5 2.68E5 9.77E4 2.40E7 2.99E5 ND 9.70E4 Cs-134 1.09 Ell 2.04E11 2.06E10 ND 5.25E10 2.15E10 5.54E8 Cs-136 5.91E9 1.74E10 6.49E9 ND 6.93E9 1.42E9 2.64E8 Cs-137 1.54 Ell 1.81 Ell 1.28E10 ND 4.85E10 1.96E10 5.65E8 Cs-138
-0 0
0 ND 0
0 0
'Ba-139 4.83E-8 0
1.40E-9 ND 0
0 3.06E-6 Ba-140 2.89E7 2.89E4 1.49E6 ND 6.87E3 1.78E4 7.10E6 Ba-141 0
0 0
ND 0
0 0
Ba-142 0
0 0
ND 0
0 0
La-140 4.88E0 1.92E0 4.95E-1 ND ND ND 2.26E4 La-142 0
0 0
ND ND ND 7.22E-7 Ce-141 5.20E3 3.17E3 3.74E2 ND 9.79E2 ND 1.64E6 Ce-143 4.76El 3.16E4 3.60E0 ND 9.20E0 ND 1.84E5 Ce-144 2.79ES 1.14E5 1.56E4 ND 4.62E4 ND 1.60E7 Pr-143 1.78E2 6.67El 8.84E0 ND 2.48E1 ND 9.41E4 Pr-144 0
0' 0
ND 0
ND 0
Nd-147 1.06E2 1.09E2 6.66E0 ND 4.19El ND 6.89E4 W-187
' 35E3 5.11E3 1.77E3 ND ND ND 3.00E5 Np-239 4.36E0
- 3. 90 E-1 2.20E-1 ND 7.78E-1 ND 1.13E4 Rev. 1 TABLE 5 (Contd.)
PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES f
MEAT PATHWAY 2
(m mrem /yr) per (pCi/sec)
NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG GI-LLI H-3 ND 2.34E2 2.34E2 2.34E2 2.34E2 2.34E2 2.34E2 C-14 3.83E8 7.67E7 7.67E7 7.67E7 7.67E7 7.67E7 7.67E7 Na-24 1.78E-3 1.78E-3 1.78E-3 1.78E-3 1.78E-3 1.78E-3 1.78E-3 P-32 7.41E9 3.47E8 2.86E8 ND ND ND 2.05E8 Cr-51 ND ND 8.79E3 4.88E3 1.33E3 8.91E3 4.66ES Mn-54 ND 8.01E6 2.13E6 ND 2.25E6 ND 6.72E6 Mn-56 ND 0
0 ND 0
ND 0
Fe-55 4.57E8 2.42E8 7.51E7 ND ND 1.37E8 4.49E7 Fe-59 3.76E8 6.09E8 3.03E8 ND ND 1.76E8 6.34E8 Co-58 ND 1.64E7 5.02E7 ND ND ND 9.58E7 Co-60 ND 6.93E7 2.04E8 ND ND ND 3.84E8 Ni-63 2.91E10 1.56E9 9.91E8 ND ND ND 1.05E8 Ni-65 0
0 0
ND ND ND 0
Cu-64 ND 2.97E-7 1.79E-7 ND 7.17E-7 ND 1.39E-5 Zn-65 3.75E8 1.00E9 6.22E8 ND 6.30E8 ND 1.76E8 Zn-69 0
0 0
ND 0
ND 0
Br-83 ND ND ND ND ND ND ND Br-84 ND ND ND ND ND ND ND
-Br-85 ND ND ND ND ND ND ND Rb-86 ND 5.82E8 3.58E8 ND ND ND 3.74E7 Rb-88 ND 0
0 ND ND ND 0
Rb-89 ND 0
0 ND ND ND 0
Sr-89 4.82E8 ND 1.38E7 ND ND ND 1.86E7 Sr-90 1.04E10 ND 2.64E9 ND ND ND 1.40E8 Sr-91 2.40E-10 ND 0
ND ND ND 5.29E-10
Rev. 1 TABLE 5 (Contd.)
PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES f
MEAT PATHWAY 2
(m mrem /yr) per(pC1/sec)
NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG GI-LLI Sr-92 0
-ND 0
ND ND ND 0
Y-90 1.71E2 ND 4.59E0 ND ND ND 4.88E5 Y-91m 0
ND 0
ND ND ND 0
Y-91 1.80E6 ND 4.82E4 ND ND ND
.2.40E8 Y-92 0
ND 0
ND ND ND 0
Y-93 0
ND 0
ND ND ND 1.55E-7 Zr-95 2.66E6 5.85E5 5.21ES ND 8.38E5 ND 6.llE8 Zr-97 3.20E-5 4.63E-6 2.73E-6 ND 6.65E-6 ND 7.02E-1 Nb-95 3.09E6 1.20E6 8.61E5 ND 1.13E6 ND 2.23E9 Mo-99 ND 1.15E5 2.84E4 ND 2.46E5 ND 9.51E4 Tc-99m 0
0 0
ND 0
0 0
Tc-101 0
0 0
ND 0
0 0
Ru-103 1.55E8 ND 5.96E7 ND 3.90E8 ND 4.01E9 Ru-105 0
ND 0
ND 0
ND 0
Ru-106 4.44E9 ND 5.54E8 ND 5.99E9 ND 5.90E10 Ag-110m 8.40E6 5.67E6 4.53E6 ND 1.06E7 ND 6.75E8 Te-125m 5.69E8 1.54E8 7.59E7 1.60E8 ND ND 5.49E8 Te-127m 1.77 E9 4.78E8 2.llE8 4.24E8 5.06E9 ND 1.44E9 Te-127 4.11E-10 1.llE-10 0
2.85E-10 1.17E-9 ND 1.61E-8 Te-129m 1.79E9 4.99E8 2.77E8 5.76E8 5.25E9 ND 2.18E9
'Te-129 0
0 0
0 0
ND 0
Te-131m 7.00E2 2.42E2 2.58E2 4.98E2 2.34E3 ND 9.82E3 Te-131 0
0 0
0 0
ND 0
Te-132 2.09E6 9.26E5 1.12E6 1.35E6 8.60E6 ND 9.33E6 I-130 3.04E-6 6.13E-6 3.16E-6 6.76E-4 9.17E-6 ND 2.87E-6 l
I i
l i
=-__
Rev. 1 TABLE 5 (Contd.)
J PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN NOBLE CASES 7
MEAT PATHWAY 2
(m mrem /yr) per (pCi/sec)
NUCLIDE BONE LIVER TOTAL BODY THYROID KIDNEY LUNG GI-LLI l
I-131 1.66 E7 1.66E7
-9.46E6 5.50E9 2.73E7 ND 1.48E6 i.
I-132 0
0 0
0
'0 ND 0
I-133 6.16E-1 7.61E-1 2.88E-1 1.41E2 1.27E0 ND-3.07E-1 I-134 0
0 0
0 0
ND 0
I-135 0
0 0
0 0
ND 0
l Cs-134 9.22E8
.1.51E9 3.19E8 ND 4.69E8 1.68E8 8.16E6 Cs-136 1.61E7 4.43E7 2.86E7 ND 2.36E7 3.51E6 1.56E6 Cs-137 1.33E9 1.28E9 1.88E8 ND 4.16E8 1.50E8 7.99E6 Cs-138 0
0 0
ND 0
0 0
Ba-139 0
0 0
ND
-0 0
0 Ba-140 4.38E7 3.84E4 2.56E6 ND 1.25E4 2.29E4 2.22E7 Ba-141 0
0 0
ND 0
0 0
Ba-142 0
0 0
ND 0
0 0
La-140 5.69E-2 1.99E-2 6.70E-3 ND ND ND 5.54E2~
La-142 0
0 0
ND ND ND 0
Ce-141 2.22E4-1.11E4 1.64E3 ND 4.85E3 ND 1.38E7 Ce-143 3.17E-2
.1.72El 2.49E-3 ND 7.21E-3 ND 2.52E2 Ce-144 2.32E6 7.26E5 1.24E5
.ND 4.02E5 ND 1.89E8 Pr-143 3.35E4 1.00E4 1.66E3 ND 5.44E3 ND 3.61E7 Pr-144 0
0 0
ND 0
ND 0
Nd-147 1.17E4 9.50E3 7.35E2 ND 5.21E3 ND 1.50E7 W-187 3.35E-2 1.98E-2 8.91E-3 ND ND ND 2.79E0 l
Np-239 4.20E-1 3.02E-2 2.12E-2 ND 8.72E-2 ND 2.23E3 i
l i
6 e
0 a i
~.
Rev. 1 2
TABLE 5 NOTES The values presented in Table 5 were calcu-4 lated according to the methodology and guidance provided in NUREG 0133, Rev. O.
Specific parameters utilized are:
1 Parameter Value Reference 0.7-Ref. 9.11.2 SF f
1.0 Ref. 9.8.2 fP 1.0 Ref. 9.8.2 s
H 8.0 Ref. 9.8.2 f
1.0 Ref. 9.8.5 b
f 0.76 Ref. 9.8.5 9
i i
a,
R%v. 1 The cumulative critical organ doses for a monthly, quarterly or annual evaluation are based on the calcu-lated dase 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 VENTILATION EXHAUST TREATMENT SYSTEM and the WASTE GAS HOLDUP SYSTEM shall be OPERABLE and appropriate portions of these systems shall be used to reduce releases of radioactivity when lhe projected doses in 92 days due to gaseous effluent releases, from each unit, to areas at and beyond the SITE BOUNDARY would exceed:
a.
0.6 mrad to air from gamma radiation, or b.
1.2 mrad to air from beta radiation, or c.
0.9 mrem to any organ of an Individual 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 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) waste gas decay 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.
OPERABILITY is demonstrated through com-pliance with Radiological Effluent Technical Specifica-tions 3.11.2.1, 3.11.2.2, and 3.11.2.3..
R*v. 1 4.0 DOSE AND DOSE COMMITMENT FROM URANIUM FUEL CYCLE SOURCES 4.1 Radiological Effluent Technical Specification 3.11.4 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactiv-ity and to radiation 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.
4.2 ODCM Methodology for Determining Dose and Dose Commitment from Uranium Fuel Cycle Sources The annual dose'or dose commitment to a MEMBER OF THE PUBLIC for Uranium Fuel Cycle Sources is determined as:
a)
Dose to the total body due to gamma ray expo-sure from immersion in a cloud of radioactive noble gases and direct radiation from the unit and outside storage tanks; b)
Dose to the skin due to beta radiation from immersion in a cloud of radioactive noble gases; c)
Thyroid dose due to inhalation and ingestion of radioiodines.
d)
Organ dose due to inhalation and ingestion of radioactive material.
Since the doses via liquid releases are very conserva-tively evaluated, there is reasonable assurance that no real individual will receive a significant dose from radioactive liquid release pathways (<1 mrem fyr/
reactor).
Therefore, only doses to individuals via airborne pathways and doses resulting from direct radi-ation are considered in determining compliance to 40 CFR 190.
(Ref. 9.12.3)
It should be noted that there are no other Uranium Fuel Cycle Sources within 8km of the Callaway Plant.
The annual dose or dose commitment to a MEMBER OF THE PUBLIC from Uranium Fuel Cycle Sources, is determined whenever the calculated doses from the release of radi-oactive materials in liquid or gaseous effluents exceed twice the limits of Radiological Effluent Technical Specification 3.11.1.2a, 3.11.1.2b, 3.11.2.2a, 3.11.2.2b, 3.11.2.3a, or 3.11.2.3b.
(Ref. 9.12.1 and --
R?iv. 1 9.12.2.)
For those situations where these limits are not exceeded by substantial amounts, it should be poss-ible to demonstrate continued compliance with 40 CFR 190 through reevaluation of the exceeded Appendix I design objective doce using more realistic assumptions.
(Ref. 9.12.3 and 9.12.4.)
4.2.1 Identification of the MEMBER OF THE PUBLIC The MEMBER OF THE PUBLIC is considered to be a real in-dividual, including all persons not occupationally as-sociated with the Callaway Plant, but who may use por-tions of the plant site for recreational or other pur-poses not associated with the plant.
(Ref. 9.13.1 and 9.8.11.)
Accordingly, it is necessary to characterize this individual with respect to his utilization of areas both within and at or beyond the SITE BOUNDARY and identify, as far as possible, major assumptions which can be reevaluated as previsously mentioned.
4.2.1.1 Utilization of Areas Within the SITE BOUNDARY The Union Electric Company has entered into an agreement with the State of Missouri Department of Con-servation for management of the residual lands sur-rounding the Callaway Plant, including some areas within the SITE BOUNDARY.
Considering the terms of this agreement and the pre-construction utilization of the area, it is reasonable to assume that primary util-ization of lands within the SITE BOUNDARY will be by hunters.
(Ref. 9.7.2, 9.7.4, and 9.14.)
Based on the availability of game, State of Missouri hunting regula-tions and certain assumptions regarding hunting habits, the average hunter is postulated to occupy areas within the SITE BOUNDARY for a maximum of 448 hours0.00519 days <br />0.124 hours <br />7.407407e-4 weeks <br />1.70464e-4 months <br /> each year (16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> each weekend for 28 weeks; a combination of the squirrel and rabbit hunting seasons).
This value is considered to be acceptably conservative such that the effect of minor variations in hunting regulations will be negligible.
(Ref. 9.7.3, 9.7.5, and 9.15)
Occupancy of areas within the SITE BOUNDARY is assumed to be averaged over a period of one year.
Figure 4.1 identifies the area immediately adjacent to the plant site which is restricted from public use.
(Ref. 9.14.1.)
Any reevaluation of assumptions should include a reevaluation of the occupancy period at the locations of maximum exposure (e.g. a real individual would not simultaneously exist at each point of maximum exposure; squirrel and rabbit hunters would be constan-tly on the move, most likely in the direction away from the plant site).
Rev. 1 4.2.2 Total Dose from Gaseous Effluents The annual dose to a MEMBER OF THE PUBLIC from gaseous effluents is determined through the use of the metho-dology presented by equations (3.11), (3.12), and (3.14), using the appropriate atmospheric dispersion parameters from Table 9 and Table 10 for the maximum exposed real individual.
It is assumed that ingestion pathways do not exist for areas within the SITE BOUNDARY.
4.2.3 Total Dose from Direct Radiation 4.2.3.1 Direct Radiation from Outside Storage Tanks The Refueling Water Storage Tank (RWST) has the highest potential for receiving significant amounts of radioac-tive materials, and constitutes the only potentially significant source of direct radiation dose from out-side storage tanks to a MEMBER OF THE PUBLIC. (Ref.
9.6.17, 9.6.18, 9.6.19, and 9.6.20.)
The direct radiation dose from the RWST is determined by isotopic measurement of the tank contents and calcu-lation of the direct dose.
Direct radiation dose from the RWST to a MEMBER OF THE PUBLIC is determined at the nearest point of the boun-dary of the area closed to public use which is not ob-scured by significant plant structures.
This has been determined to be approximately 1,020 meters from the RWST.
The RWST is approximately 12 meters in diameter, 14 meters in height with a capacity of approximately 1,514,000 liters.
(Ref. 9.6.20. )
The walls are of type 304 stainless steel and have an average thickness of.87 cm.
(Ref. 9.16.1.)
Assuming that the RWST approximates a point source at this distance, and neglecting attenuation provided by the walls of the tank, the exposure rate from monoen-ergetic gamma radiation is given by:
ER = BAT exp(-pen )
(4*l) d 2
d Where:
ER
= Exposure rate at distance d from a point source of strength A, (in Roentgens / hour).
B
= Buildup factor.
Rev. 1 A
= Activity of the source, (in Curies).
d
= Distance from the source, (in meters).
p
= Ligear attenuation coefficient for air, in en (cm
).
2 F
= Exposure rate constant, (in R - m /Ci -
hr.).
The exposure rate constant F is given by:
F = K f Ep (4.2) a Where:
E
= Energy of the gamma radiation, in (MeV).
p
=L{nearabsorptioncoefficientforair, (in cm
).
f
= Number of photons emitted per disintegration.
3 K
= Constant, 1.49 E04 R - cm /hr - MeV - Ci.
E
= Is as previously defined.
For nuclides emitting multiple gamma rays, equation (4.1) becomes:
ER = KA IB (f
E p i) eXP (-Peni d)
(4.3) i i
g a
2 1
d Where:
B
= Buildup factor for ith photon.
f f
= Energy of the ith photon, (in MeV).
i E.
=Linearabsorptioncoe{ficientforair, for 1
the ith photon, (in cm~ ).
p"Ul
= Linear attenuation cogfficient for air, for the ith photon, (in cm- ). -
Rnv. 1 ER, K,A, and d are as previously defined.
For photon energies in the range of 60 kev to 2 MeV, the value of the linear absorption coefficient for air is relatively constant (i 15%), therefore, equation (4.3) can be approximated as:
ER = K'A IBf (fg i) exp (-peni )
(4*4)
E d
2 1
d Where:
2 K'
= A constant, 0.48 R-m /hr - MeV - Ci.
B, and peni are as previously ER, A, d,
f,E, i
i g
defined.
Through the use of equation (4.4), the exposure rate for a particular nuclide can be determined.
The total exposure rate from the RWST is calculated as:
ERtotal
- h R3 (4.5)
ERtotal
= Total exposure rate at the location of the MEMBER OF THE PUBLIC from the RWST, (in Roentgen / hour).
ER.
= Calculated exposure rate from the jth J
nuclide, (in Roentgen /hr).
The total direct radiation dose rate from the RWST to a MEMBER OF THE PUBLIC is given by:
(4.6)
DRtotal =ERtotal Where:
DR
= Total dose rate from the RWST (in total rem /hr).
ER is as previously defined.
total The direct radiation dose to a MEMBER OF THE PUBLIC is i
then determined for a specific time period:
D
=1.
(DRtotal) ( )
(
DR l
Rnv. 1 D
= Direct radiation dose to a MEMBER OF DR THE PUBLIC for the specific time inter-val, (in rem).
1.23
= Occupancy factor (448 hours0.00519 days <br />0.124 hours <br />7.407407e-4 weeks <br />1.70464e-4 months <br /> / year +
365.25 days / year), (in hours / day).
t
= Length of specific time period, (in days).
DR is as previously defined.
total 4.2.3.2 Direct Radiation from the Reactor The maximum direct radiation dose from the Unit to a MEMBER OF THE PUBLIC has been calculated to be 0.2 mrem / calendar year.
This was based on exposure at the point of the boundary of the area closed to public use which is not obscurred by significant plant structures; a distance of approximately 1222 meters.
(Ref. 9.5.3)
The maximum direct radiation dose from the Unit to a MEMBER OF THE PUBLIC due to activities within the SITE BOUNDARY is thus approximately.01 mrem per year, as-suming a maximum occupancy of 448 hours0.00519 days <br />0.124 hours <br />7.407407e-4 weeks <br />1.70464e-4 months <br /> per year. -,
I
\\
s N
NNW NNE l
(f I
i f
O NW x
L L.i
(
W WNW
?
OUW v>
~
g
~
li j!i
- p f I
w y
W y
k
~'
s e
o i
~
O o
WSW o
+jp 0'
0 v{
\\
g SW
\\
/ 1,200 METERSsj\\
i i
NE
"'IS"#E*f
~
[
f PLANT
.s SITE 80 NDARY-*
Z NE T7 ROAD 448
/A
@ CLOSED TO PUBLIC USE sf g gg 1
E
~';e D.
.,.;,)
A1Dcl-3 ~; y j,,-
~
CARD
/
ESE N
Also Available On Aperture Card N
SE AREA CLOSED TO PUBLIC USE 831129 014 7 -O l UNION ELECTRIC CO.
CALLAWAY PLANT SITE AREA CLOSED j
30 0
TO PUBLIC USE g
FIGURE 4.1 REV 1
Rev. 1 5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING 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 Radiological Enviormental Monitoring Program is in-tended to act as a background data base for preopera-tion and to supplement the radiological effluent release monitoring program during plant operation.
Radiation exposure to the public from the various spe-cific 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 bases.
Efforts are made to obtain all required samples within time frame outlines.
Any deviation (s) in sampling frequency or location is documented in the Annual Radiological En-vironmental 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 Ta-bles 7 and 8.
mm _..
4 Rev. 1 TABLE 6 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM h
huMBER OF REPRESENTATIVE EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE SAMPLE LOCATIONS COLLECTION FREQUENCY OF ANALYSIS
- 1. Direct Radiation 40 routine monitoring stations either At least once per 92 Gamma Dose j
with two or more dosimeters or with days.
j one instrument for measuring and
}
recording dose rate continuously, placed as follows:
,i An inner ring of sixteen stations, cN one in each meteorological sector in 3
the general area of the SITE BOUNDARY.
i Station j
Code Sector Site Description Location 1
4 A
0.6 Miles East of Hwy O and CC Junction 1.9 mi. @ 354* N l
47 B
County Road 335, 0.9 Miles South of 0.9 mi. @ 20' NNE Hwy 0 48 C
Plant Security and Wildlife Management 0.5 mi. @ 47' NE Area Sign Post (Heavy Haul Road) 5 D
Primary Meteorological Tower 1.3 mi. @ 76* ENE 49 E
Callaway Electric Cooperative Utility 1.7 mi. @ 94* E Pole No. 06959 i
52 F
Light Pole Near East Plant Security 0.3 mi. @ 111* ESE Fence 51 G
Located in the "Y" of the Railroad 0.7 mi. @ 132' SE Spur, NW of Sludge Lagoon 50 H
Heavy Haul Road, Intake / Discharge 1.1 mi. @ 157* SSE i
Pipeline Marker
.~.
1 L
Rev. l' l
TABLE 6 (Continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 1
t J'
Station 3
Code Sector Site Description Location i
i 7
J Callaway Electric Cooperative Utility 1.3 mi. 0 173*.S Pole No. 18715 f
37 K
Piezometer M8 and M6 0.5 mi. 0 204* SSW 43 L
Plant Security and Wildlife Management 0.5 mi. 0 224* SW Area Sign (Heavy Haul Road) i.
.44 M
Callaway Electric Cooperative Utility 1.7 mi. @ 249' WSW Pole No. 18769 S
6 N
Akers Farm 1.8 mi. 0 277* W 45 P
NW Side of Intersection of CC anu AD 0.9 mi. 0 287* WNW
)
3 Q
Callaway Electric Cooperative Utility 1.6 mi.
- 322' NW
}
Pole No.-18450 46 R
0.3 Mile South of.the CC and 0 Junction 1.5 mi. @ 333' NNW i.
I An outer ring of sixteen stations,
}
one in each meteorological sector in the 6-to 8-km range from the site I
j 36 A
callaway Electric Cooperative Utility 4.9 mi. @ 8' N Pole No. 19137 21 B
Callaway Electric Cooperative Utility 3.8 mi. @ 28' NNE
)
Pole No. 19100 1
j 20 C
Callaway Electric Cooperative Utility 4.8 mi. 0 45' NE Pole No. 12630 16 D
Callaway Electric Cooperative Utility 4.1 mi. @ 75* ENE
]
Pole No. 12976 i
f i
Rev. 1 TABLE 6 (Continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROCRAM Station Code Sector Site Description Location 17 E
0.5 Miles East of Hwy D, 1.5 Miles 4.0 mi, @ 88* E South of Ilwy D and 0 Junction 15 F
Lamb Farm 4.2 mi. @ 117' ESE 11 C
City of Portland 5.0 mi. @ 136* SE 10 H
Callaway Electric Cooperative Utility 4.0 mi. @ 156' SE Pole No. 12179 N
C) 9 J
NW Side of the lleavy llaul Road and 3.7 mi. @ 181* S Hwy 94 Junction 30 K
City of Steedman 4.5 mi. 0 203* SSW 42 L
Callaway Electric Cooperative Utility 4.4 mi. @ 230' SW Pole No. 06326 32 M
D. Bartley Farm 5.1 mi. @ 241' WSW 41 N
Callaway Electric Cooperative Utility 4.8 mi. @ 227' W Pole No. 18239 40 P
Callaway Electric Cooperative Utility 4.2 mi. @ 291* WNW Pole No. 18145 39 Q
Callaway Electric Cooperative Utility 5.4 mi. @ 312' NW Pole No. 17516 38 R
Callaway Electric Cooperative Utility 4.5 mi. @ 334' NNW Pole No. 34708 Eight stations to be placed in special interest areas such as population centers, nearby resi-dences, schools, and in 1 or 2 areas to serve as control stations.
I
.,o m.
.a m
u e
++6=.
-e.
-,a4+
mm 4 -.-
m-.
.g_e4 1
1 Rev. 1 i
TABLE 6 (Continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM i
i Station Code Site Description Location 33 City of.Itams Prairie 7.3 mi. 0 '71* W l
31 City of Mokane 7.6 mi. @ 218' SW 4
26 Town of Americus 12.1 mi. @ 82* E 27 Town of Bluffton 9.5 mi. @ 112' ESE 5
da yd 35 City of Toledo 5.8 mi. @ 340* NNW 23 City of Yucatan 6.7 mi. @ 14* NNE j
11 City of Portland 5.0 mi @ 136' SE 20 I
j 34 (P-Control) 2.5 Miles South of 0 and C 3.5 mi. @ 291* WNW Junction I (Q-Control)
City Limits of Fulton on liwy 10.6 mi. @ 311' KJ 1
Z 2.
Airborne Radioiodine Canister:
Radioiodine and Samples from five locations Continuous operations of Ane.yze at least once Particulates sampler with sample per 7 days for I-131.
Three samples from close to the collection as required by three SITE BOUNDARY locations, dust loading, but at least Particulate Sampler:
in different sectors, of the once per 7 days.
highest calculated annual average Analyze for gross beta ground level D/Q.
radioactivity > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following filter change. Perform gamma
Rev. 1 TABLE 6 (Continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM isotopic analysis on those samples for which the gross bcta activity is > 10 times.the yearly mean of control samples. Perform d
gamma isotopic analysis on composite samples (by location) at least s,
once per 92 days.
Y Station Code Sector Site Description Location A-5 Q
Smola Farm 6.6 mi. @ 318' NW A-3 C
Bahr Brothers Farm 7.0 mi. @
50' NE A-4 B
Cregan Farm 6.7 mi @
11' N One sample from a control location, as for example 15-30 km distant and in the least prevalent wind direction.
A-7 Q
C. Bartley Farm 9.5 mi. @ 312" NW s
4
~
m Rev. 1 TABLE 6 (Continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM w
3.
Waterborne e
f h'
d a.
Surface One sample upstream.
Composite sample over a Gannb isotopic analysis period of less than or
' of each sample. Tritium equal to 31 days.
analysis of composite sample at least once per -
92 days.
Station Code Sector Site Description Location Nu e
S01 H
84 Feet Upstream of Discharge, 4.8 mi. 0 144' SE North Bank One sample downstream.
s S02 G
1.1 Miles Downstream of Dis-5.2 mi. 0 133* SE charge, North Bank b.
Drinking One sample of the nearest
' Grab sample collected Gamma isotopic and gross water supplies that could at, least once per 31 beta analyses of each be affected by its dis-days.
sample. Tritium analyses
- charge, of composite sample at least once per 92 days.
Station Code Sector Site Description Location S03 E
City of St. Louis Water intake 68 mi. @ 94* E I
l E
l One sample from a control Composite sample over J
location.
a period of less than or equal to 31 days.
Rev. 1 TABLE 6 (Continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Station Code Sector Site Description Location S01 H
84 Feet Upstream of Discharge.
4.8 mi. @ 144* SE North Bank 4.
Ingestion d
a.
Milk One sample from milking animals At least once per 15 Camma isotopic and l
aM in each of two areas between 5 days when animals I-131 analysis of each 7
to 8 km distance where doses are on pasture; at sample.
are calculated to be greater least once per 31 days 8
at other times.
than 1 mrem per yr.
Station Location C. de Sector Site Description M-3 C
Davidson Farm (Goat Milk) 3.8 mi. @ 55' NE Perotka Farm (Cow Milk) 3.1 mi. @ 136' SE M-2 G
Rev. 1 TABLE 6 (Continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM One sample from milking animals at a control location, 15-20 km distant and in the Icast pre-valent wind direction.
Station Code Sector Site Description Location M-1 M
Green Farm (Cow Milk) 12.3 mi. @ 243* ueu da tn l
8 b.
Fish One sample of each of two One sample in season, Gamma isgtopic species in the vicinity of or at least once per analysis on edible plant discharge area.
184 days if not seasonal.
portions.
One sample of each of the following:
1.
Bottom Feeder Species 2.
Predator Species Station Code Sector Site Description Location C
G 1.0 River Miles Downstream 5.1 mi. @ 135' SE of Discharge, North Bank One sample of similar typa in areas not influenced by plant discharge.
E S
S
'4 n
5 o
1 i
1 t
a c
v o
i e
L m
R 9
4 MARGO R
P G
s N
i I
D RO f
T o
I N
n m
O o
a M
i e
6 t
r L
p t
E A
i s
L T
r pk B
N c
U n A
E s
a T
M e
sB N
D e
O l h R
e it I
t M r V
i o
N S
rN E
ev,
L i e A
R g C
r
- 6. ha I
CD 0 c IO I
DA R
r o
tce S
noe id t o A
aC t
S
Rev. 1 TABLE 6 (Continued)
TABLE NOTATION (a) Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment, and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every ef fort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time.
In these instances, suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the radiological environmental monitoring program. Identify the cause of the un-availability of samples for that pathway and identify the new location (s) for obtaining replacement samples in the next Semi-Annual Radioactive Ef fluent Release Report and also include in the report a revised e
2j figure (s) and table for the ODCM reflecting the new location (s).
e (b) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. The 40 stations is not an absolute number. The number of direct radiation monitoring stations may be reduced according to geographical limitations; e.g., at an ocean site, some sectors will be over water so that the number of dosimeters may be reduced accordingly. The frequency of t..alysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.
(c) The purpose of this sample is to obtain background information..If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites that provide valid background data may be substituted.
(d) Gamma isotopic analysis is defined as the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.
Rev. 1 TABLE 6 (Continued)
TABLE NOTATION (e) The " upstream sample" shall be taken at a distance beyond significant influence of the discharge. The
" downstream" sample shs11 be taken in an area beyond, but near the mixing zone.
(f) In this program, constant volume sample aliquets are collected at time intervals that are short (e.g.
hourly) relative to the compositing period (e.g. monthly).
(g) The dose shall be calculated for the maximum ogran and age group, using the methodology and parameters in the ODCM.
b
=
a
Rev. 1 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/m )
(pCi/kg), wet (pCi/1)
(pCi/kg, wet) 3 11 - 3 20,000 (a) 2, Mn-54 1,000 30,000 l
Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zr-Nb-95 400 (b) 1-131 2
0.9 3
100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140 200 (b) 300 (b)
(a) For drinking water samples. Values are from 40 CFR 141.
If no drinking water pathway exists, a value of 30,000 pCi/l nay be used.
(b) Total for parent and daughter.
I
Rev. 1 TABLE 8 MAXIMUM VALUES FOR Tile LOWER LIMITS OF DETECTION (LLD)a.d.e Water Airborne Particulate Fish Milk Food Product Sediment 3
Analysis (pCi/1) or Gases (pCi/m )
(pC1/kg, wet)
(pCi/l)
(pCi/kg, wet)
(pCi/kg, dry) b Cross Beta 4
.01 11 - 3 2000*
Fn-54 15 130 Fe-59 30 260 Co-58,60 15 130 Zr-Nb-95 15" b
I-131 l
.07 1
60 l
Cs-134 15
.05 130 15 60 150 Cs-137 18
.06 150 18 80 180 Ba-La-140 15" 15"
- If no drinking water pathway exists, a value of 3000 pCi/l may be used, i
n Rsv. 1 TABLE 8 (CONTINUED)
TABLE NOTATION (a)
The LLD is defined for purposes of compliance with the Radiological Effluent Technical Spe-cifications as the smallest concentration of radioactive material.in a sample that will yield a net count, above system background, that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system (which may include radiochemical separation):
LLD =
4.66 Sb 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 tha counting rate of a blank sample as appropriate (as counts per minute).
E=
The counting efficiency (as counts per disintegration).
V=
The sample size (in units of mass or volume).
2.22 =
The number of disintegrations per minute per picocurie.
Y=
The fractional radiochemical yield (when applicable).
A=
The radioactive decay constant for the par-ticular radionuclide and,.
m Rnv. 1 at =
the elapsed time between sample collection (or end of the smaple collection period) and time of counting (for environmental samples, not plant effluent samples).
Typical values of E, V, Y and At shall be used in the calculations.
It should be recognized that the LLD is defined as a a
'ariori (before the fact) limit representing the capa -
Jility of a mesaurement system and not as an a posterio i (after the fact) limit for a particular measurer tnt.
Analyses are performed in such a manner that the stated LLDs are achieved under routine conditi ns.
Occassionally background fluctuations, unavoidable small sample sizes, the presence of inter-fering nuclides, or other uncontrollable circumstances may render these LLDs unachievable.
In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report.
(b)
LLD for drinking water.
(c)
Total for parent and daughter.
(d)
This list does not mean that only these nu-clides are to be considered.
Other peaks that are identifiable, together with those of the above nuclides, shall also be anlayzed and reported in the Annual Radiological Environ-mental Operating Report.
\\
//
10 MILE R ADMJS s$ 1 3
, /
NNW i
l
'.]
i..
Id NW (
vuCATAN O
ga
^
j TotgDo S
6 MILE R ADIUS
~m
?:
) ?
/
,,. m 1
wNW W
e REFORM
$ 40 P
,o~^ = ~~
.... x4
- =
vv
[c]
i
}
STEEDW AN
( WSW
(
3e N
. a., *...-
(
\\-.
,o,
K LER Mg,=;.-
F N
7
- s',/'
en4uo,s
'S' Ssw
, 6. / *;s *-
e
,s s-
/
ST.AUSERT 4+-
f
- o 33w
/,,
.V s
\\
to MILE R ADIUS
I i
v N
I N
l NNEl I
N l
i 6
NE,
I N
?RC h
A ? E R ?- la 6 lo 34,.,
CAaJ ENE l
- M
'y/
I T -1 LEOEND:
gg iNcoaPon ATEo couuuMITIES e
y e
l l
O uNimcoaPon ATEo couuuNifie s 2
7 1
A A'n s AuPLtn Loc ATsoNS I
G TLD LOC ATIONS y
25 9 3
TLD & AIR S AMPLER LOC ATIONS D
_lg E
oir
-e,
-.! 5
!!8 Ot>
Oi
)
/
58 st g
,,, E aE NcE:
.A 0
TMi$ u A P W A S PREP ARE D F ROM A PORTION y t t i.NO 1
A l
OF THE FOLLOWINO U 8.0 5. W AP:
stur TOWN
~ ~ ~ ~ ~ ~ ' '
YCOUNTY 5 7. Louts, uO.,19 6 2.
Og
'N l ko, %, s COUNTY
,T
!ESE
_f' r
?
j woaaisoN uit E S l
l Also AvailaMe Un SE l
Aperture Card 7
,5 UNION ELECTRIC CO.
li$
c CALLAWAY PLANT ll R ADIOLOGIC AL AIR SSE w80
$l$
S AMPLING NETWORK 0,0 FIG U R E 5.1 A REV 1 l
8 31129 014 7 -OL
NW NNW N
/
NNE.
NE
~ENE 4
REFORM r
N h46 LEGEND; N
O TLD, LOCATION 3G I FCf RESTRICTED AREA BOUNDARY 47 (SITE BOUNDARY FOR tN N W GASEOUS EFFLUENTS 45 5
48 3
g, g
j
___ 7 I
/
448 4
37 450 7e ESE SW COUNTY R D. 4 5 9 ~' i - R AILRO AD SPUR SE COUNTY RD.468 l
PORTLAND 9
s4 O
\\Eg CHUTE g
STEEDMAN O
2000 0
2000 UNION ELECTRIC CO.
FEET SSE C ALLAWAY PLANT S
^ '
' ^
^'
l 5 MILE RADIUS S AMPLING NETWORK FIGU RE 5.1 B REV 1
W NW NNW N
NNE NE WNW L'NE N
REFORM
\\
./
E W
SITE T
o e$
+
WSW o
ESE v
\\o,09 SSE PORTLAND SW ee LOGAN REEK O/
~
~
U:c7 m D -
o*e e
801 o
+
N 'b*
A*
%egag'k gsO*p 8
DISCHARGE POINT (SITE BOUNDARY FOR LIQUID EFFLUENTS)
HA OlS LEGEND:
O TOWNS
- -INTERMITTENT STREAMS CONTINUOUS STREAMS 9
AQUATIC SAMPLING STATIONS 1
0 1
O COMPOSITE SURFACE WATER g_________.
i MILES UNION ELECTRIC CO.
CALLAWAY PLANT LOCATION OF AQUATIC S AMPLING STATIONS FIGURE 5.2 A REV 1
I h
N NNE NE NNW 10 gLE RAD /03 5 Y 0g SIT,E BLUFFTON HER5 ANN S01
_~
NEW WSW S2 CHAMOIS
\\
SW f
REFERENCE:
U.S. ARMY CORPS OF ENGINEERS,1977 MISSOURI f
8 RIVER HYDROGRAPHIC SURVEY, RULO, NEBRASKA TO THE SOUTH, KANSAS CITY DISTRICT.
SSW I
S
\\
,l j
l
- o f'
ENE e,
t, C.
[
Q ST. CHARLES j
')
AUGUST A. BUSCH WILDLIFE AREA S03 E
W I h,,/ /'I,/
/'
ST. LOUIS CITY
& COUNTY WATER WORKS UNIVERSITY OF MISSOURI WELDON SPRING Dp EXPERIMENTAL FARMS I\\ V l]
L. i t+
CARD
~
W
' W
~h iAVEN WASHINGTON Also Available en SOUTH Aperture Card POINT I
ESE 4
0 4
8 N"
MILES UNION ELECTRIC CO.
CALLAWAY PLANT ACE WATER STATIONS S
NG STATIONS
@ AQUATIC / SURFACE WATER STATION (REFER TO FIG. 5.2A)
FIGURE 5.2B 8311290147-03
\\
,0 MIL. R ADIUS 8
NNW lN lNW TOLEDO
\\N l
3 WILE R ADIUS
/ /
WNW
/
(
l q
C REFORM x
1
(
l Q
w
'N A
l vv l
1 u
l l
x STEEDM AN
,s, ( WSw
\\
,,,. e + "",
,,,,3
['
, se d * *
" '"' f,
g;,
yj
,/ p ?;;],, / g e-sw g
/
4,., -
,4 0
,o SSW i
,7, S
,.. m Aom.
i
l l
l l
NE,
j qi N
?70 A3-A1-bA2 1
CAlJ ENE svue/
_ % v li LEoENo; l
g inconeonareo couuunities O
uninconpon Arto couuunities M
y' ui<.
V
~
i l
E
_l:
El'1 8?
ii Also lvailable On si Aperture Card u.,
52
\\
nersatuct:
I Bluf TOWN TH,S uAP W AS PREPAREo FRou A PoRisoN hou 1, s
or 1 s, otto
,~o u s.o.. um,:
'"2-40, ig,;, '-
..o '..
8 v.
-s e
!ESE
=Y o
l t
0 1
2, 3,
g i
l uites i
l SE
_g UNION ELECTRIC CO.
/
li CALLAWAY PLANT SSE 8!!
si$
MILK E.E S AMPLING LOC ATIONS FIGURE 5.3 REV 1 8 31129 014 7 -D4
Rtv. 1 2
6.0 DETERMINATION OF ANNUAL AVERAGE AND SHORT TERM ATMOSPHERIC DISPERSION PARAMETERS 6.1 Atmosphere Dispersion Parameters The values presented.in Table 9 and Table 10 were determined through the analysis of on-site meterologi-cal data collected-during the three year period of May 4,.1973 to May 5, 1975 and March 16, 1978 to March 16,
)
-1979.
The PUFF (fluctuating plume) model and the straight-line Gaussian (constant meP' wind direction) model were used for determination of the long-term atmospheric dispersion parameters.
A more detailed discussion of the methodology and data utilized-to. calculate these 4
parameters can be found elsewhere (Ref. 9.6.12).
The terrain within 80 km of the site is gently rolling with no important ranges of hills or mountains.
Thers are several small lakes and reservoirs in the region, however none are large enough to,significantly affect the ambient dispersion parameters (Ref. 9.6.13).
6.1.1 Long Term Diffusion Estimates 6.1.1.1 The PUFF Model 4
The general equation for the PUFF model is (Ref. 9.10.1):
3/2 2
-1 2-2 2 2 X/Q = 2 (2n)
H "z _
g 4 (r / H + h,/o )
(6.1) g Where:
2=
(x-ut) 2, y2 r
and oH * "y * "x h, =
Effective release height (in meters).
Q=
Effluent emission over the time interval (in i'.
Curies).
I t=
Travel time (in seconds).
u=
mean windspeed at the height of the release point (in m/sec).
x=
Distance from center of PUFF in the direction of flow (n meters). -,.~.
c R*.v.
1 y=
Distance from center of PUFF in the crossflow direction (in meters).
Plume spread along the direction of flow (in e
=
x meters).
Lateral plume spread (in meters).
o
=
y Vertical plume spread (in meters).
o
=
g X=
Atmospheric concentration of effluent in a PUFF at ground level and at distance, x, from 3
the PUFr center (in Ci/m ).
Calculations utilizing the PUFF model were performed for 22 standard distances to obtain the desired disper-sion parameters.
Dispersion parameters at the SITE BOUNDARY and at special receptor locations were esti-mated by logarithmic interpolation according to (Ref.
9.6.14):
y
( )B (6.2)
X=X 1
Where:
B=
In (X /X )/in (d /d ).
2 1
2 y
X,X2 = Atmospheric concentrations at distances dy and y
d, gespectively, from the source (in C$/m).
The distances d and d were selected such that y
2 dl <d<d '
2 6.1.1.2 The Straight-Line Gaussian Diffusion Model.
Long-term dispers' ion parameters were datermined through the use of the straight-line Gaussian diffusion model, incorporating site specific parameters, plume meander
- and building wake effects.
The general equation for the straight-line Gaussian model is (Ref. 9.10.2):
2 2
(X/Q)D=2.032 {$ ng$ (NxEi zj(x) )
exp (-h /2o j(x) )
(6.3)
I e
1 Where: -
Rtv. 1 h, =
Effective release height (in meters).
n..
=
Length of time (hours of valid data) weather 13 conditions were observed to be at a given wind direction; wind speed class, i; and at-mospheric stability class, j.
N=
Total hours of valid data.
II. =
Windspeed at the midpoint of windspeed class, 1
i, at a height representative of release (in m/sec).
zJ(x) = Vertical plume spread (in meters) without o
volumetric correction at distance, x, for sta-bility class, j.
Z3(x) = Vertical plume spread (in meters) with I
volumetric correction for a release within the building wake cavity, at a distance x, for stability class, j.
Otherwise IzjIX)-"zjIX)*
(X/Q) =
Average effluent concentration, X, normalized by source strength Q, at a distance, x in a 3
given down wind direction, D (in sec/m ).
For effluent releases from the Unit Vent, the plume was considered as an elevated release whenever the exit velocity of the plume W, was at least five times the horizontal windspead, d at the height of the release.
u, When W was less tnan u, the release was considered to be gro8nd level.
For all other cases, a mixed mode release was assumed, in which the plume was considered to be an elevated release during some fraction of the time and as a ground level release (h =0) during the remainder of the time.
An entrainment c8 efficient, E was determined for the mixed mode release case (Reff,9.10.3):
E =2.58 - 1.58 (Wo/u) for 1<(Wo/u)<l.5 (6.4) t and E =0.3 - 0.06 (Wo/u) for 1.5 <(Wo/u)<5 (6.5) t,_.
Rsv. 1 The release was considered to occur as an elevated release 100 (1-E,f the time.) % of the time and as a ground level release 100E
%5 Each case was evaluated separately aid the concentration was calculated accord-ing to the fraction of time each type of release occured.
However, subsequent to the determination of the disper-sion parameters presented in Table 10, a rain cover was installed over the Unit Vent, thus effectively elimi-nating the possibility of the plume vertical exit velo-city exceeding five times the horizontal wind speed.
It therefore became necessary to evaluate the resulting impact on the calculated dispersion parameters for the Unit Vent.
The vertical exit velocity of the plume was determined to be approximately 19.4 mph.
In order for this to be considered an elevated release point, the horizontal wind velocity would necessarily have to be less than approximately 4 mph.
Inspection of the 60 meter wind speed measurement data for the study period revealed that this situation had occurred approximately 3% of the time.
It was therefore concluded that the values presented in Table 10 contain an insignificant error and require no modification.
(Ref. 9.5.2 and 9.6.16.)
For ground-level releases (h =0) consideration was given to initial mixing of t8e effluent plume within the building wake, and equation (6.3) becomes: (Ref.
9.10.4)
(X/Q)D=2.032 ((o j(x) + cV 73)1/2 y x ) -1 (6.6) 2 2
z l
l Where:
c=
Building shape factor.
l V=
Vertical height of the highest adjacent build-ing (in meters).
"zj *)' "i, and x are as previously defined.
I (X/Q)D' 2
The wake factor, (cV /n) is restricted by the condition that 2
(cV /n ) = YT o $ (x)
(6.7) g When 2
2
+ cV jy)l/2 >
3 0
(6.8) 2 (o gj gj r
Rnv. 1 The dispersion parameters resulting from the straight-line Gaussian model were modified through the applica-tion of terrain / recirculation factors, determined as the ratio between the PUFF-advection X/Q estimates and the straight-line X/Q estimates.
(Ref. 9.6.14.)
Decayed X/Q values were determined based on half lives of 2.26 and 8.0 days.
Depleted X/Q values were deter-mined as a function of stability class, plume travel distance, and height of release.
Ground deposition rates were determined according to:
D/Q = RDEP (2 sin (11.25) x)
(6.9)
Where:
-2).
D/Q =
Ground deposition rate (m RDEP =
Relative ground deposition rate (m-3),
x=
Distance from the release oint (meters).
(Ref. 9.6.15) 6.1.2 Short Term Diffusion Estimates Airborne releases are classified as short term if they are less than or equal to 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> during a calendar year and not more than 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> in any quarter.
Short term diffusion estimates are determined by multiplying the appropriate long term diffusion estimate by a correction factor (Ref. 9.9.1):
F = (T /Ta)
(6.10) s Where:
The total number of hours of the short term T
=
g release.
The total number of hours in the data collec-T
=
a tion period from which the long term diffusion estimate was determined and S = -Log (1 + o/(X/Q) )
(6.11)
Log Ta.
R9.v. 1 Where:
3 a=
The standard deviation of (X/Q) (in sec/m ),
(X/Q) =
The undecayed, undepletgd, long term relative concentration (in sec/m ).
T is a previously defined.
a Values of a are presented in TABLE 11.
Short term diffusion estimates are applicable to short term releases which are not sufficiently random in both time of day and duration (e.g., the short term release periods are not dependent solely on atmospheric condi-tions or time of day) to be represented by the annual average dispersion conditions.
(Rev. 9.8.12.).
_.m.-__
t f
i Rev. 1 l
TABLE 9 UIGl!EST ANNUAL AVERAGE ATMOSPilERIC DISPEP.SION PARAMETERS (a) t l
RADWASTE BUII. DING VENT DISTANCE X/Q X/Q LOCATION (b)
SECTOR (METERS) y/g Decayed /Undepleted Decayed / Depleted D/Q SITE BOUNDARY S
1300 9.7E-07 9.7E-07 8.6E-07' 4.0E-09 1
Nearest Cow (c)
NNW 2736 7.4E-07 7.3E-07 6.lE-07 2.7E-09 1
i Nearest Coat NNE 3540 3.2E-07 3.2E-07 2.6E-07 1.4E-09
}
Nearest Meat Animal NNW 2736 5.9E-07 5.9E-07 5.0E-07 2.3E-09
[
4 L
)
oo Nearest Vegetable Carden NNW 2736 5.9E-07 5.9E-07 5.0E-07 2.3E-09 ua Nearest Residence N
1448 1.5E-06 1.5E-06 1.3E-06 7.7E-09 4
Boundary, Area Closed i
to Public Use (d)
SW 700 2.9E-06 2.9E-06 2.7E-06 8.8E-09 1
(a) Values given are from FSAR, Table 2.3-84, and. Table 2.2-86 4
(b) Data from 1978 survey i
4 j
(c)
Data corresponds to grazing season (Ref. 9.10.5) r (d)
Values derived from FSAR, Table 2.3-81, using the mehtodology presented in Equation (6.2).
I I
I 2
Cross Sectional Area = 650 m (Ref. 9.5.4) i
=
r 2
I l
f l
i i
- 4'
Rev. 1 TABLE 10 IIIGIIEST ANNUAL AVERAGE ATMOSPilERIC DISPERSION PARAMETERS (a)
UNIT VENT DISTANCE X/Q X/Q LOCATION (b)
SECTOR (Meters) x/Q Decayed /Undepleted Decayed / Depleted D/Q SITE BOUNDARY NW 2300 2.5E-07 2.5E-07 2.2E-07 1.5E-09 Nearest Cow (c)
NNW 2736 1.9E-07 1.9E-07 1.6E-07 9.8E-10 Nearest Goat NNE 3540 1.0E-07 1.0E-07 8.8E-08 6.4E-10 Nearest Meat Animal NNW 2736 1.7E-07 1.7E-07 1.5E-07 9.6E-10 Nearest Vegetable Garden NNW 2736 1.7E-07 1.7E-07 1.5E-07 9.6E-10 Nearest Residence N
1448 3.3E-07 3.3E-07 3.0E-07 2.9E-09 NW 1154 5.2E-07 5.2E-07 4.7E-07 3.9E-09 P 1ic s d)
(a) Values given are f rom FSAR Table 2.3-82, and Table 2.3-85.
(b)
Data from 1978 survey (c)
Data corresponds to grazing season (d) Values derived from FSAR, Table 2.3-83, using the methodology presented in Equation (6.2).
Cross Sectional Area = 2650 m (Ref. 9.5.4)
Rnv. 1 TABLE 11 STANDARD DEVIATION OF ANNUAL AVERAGE DISPERSION PARAMETERS (TBD)
Rev. 1 TABLE 12 APPLICATION OF ATMOSPilERIC DISPERSION PARAMETERS DOSE PATilWAY ODCM REFERENCE DISPERSION PARAMETER CONTROLLING AGE GROUP Noble Gas, Beta Air 3.5.2.1 X/Q, decayed /undepleted Noble Gas, Gamma Air 3.5.2.1 X/Q, decayed /undepleted Noble Gas, Total Body 3.4.1 & 3.5.1.1 X/Q, decayed /undepleted b
Y Noble Gas Skin 3.4.1 & 3.5.1.1 X/Q, decayed /undepleted Ground Plane Deposition 3.5.2.2.1 D/Q Inhalation 3.5.2.2.1 X/Q, decayed / depleted Child Vegetation 3.5.2.2.1 D/Q*
Child Milk 3.5.2.2.1 D/Q*
Infant Meat 3.5.2.2.1 D/Q*
Child
- For 11-3 and C-14, X/Q, decayed / depleted is used instead of D/Q (Reference 9.11.1).
i
Rav. 1 7.0 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT.
Routine Radioactive Effluent Release Reports covering.
the operation _of the. unit during the previous 6 months of operation.are submitted within 60 days after January l 'and July 1 of each year.
The period of the first report begins with the date of initial criticality.
The Radioactive Effluent Release Reports include a sum-mary of the quantities of radioactive liquid and-gaseous effluents and. solid waste released from the unit as outlined in Regulatory Guide.l.21,." Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases.of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June.1974, with data summarized on a quarterly basis following the format of~ Appendix B thereof.
For solid wastes, the format for Table 3 in Appendix B is supplemented with three additional categories:
class of solid waste (as defined by 10 CFR Part 61), type of container (e.g., cement, urea formaldehyde).
The Radioactive Effluent Release Report to be submitted within 60 days after. January 1 of each year includes an annual summary of hourly meteorological data collected over-the previous year which may be either in the form
'of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and pre-cipitation, or in the form of joint frequency distribu-tions of wind speed wind direction, and atmospheric stability.*
This same report includes an assessment of the radiation doses due'to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year.
This same report also includes, as required by Technical Specification 3.11.4, an assessment of the radiation doses from radi-oactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due'to their' activities inside the SITE BOUNDARY during the~' report-period.
All assumptions used in mak-ing these assessments,~i.e., specific activity, expo-sure time and location, is included in these reports.
The meteorological conditions concurrent with the time of release of radioactive materials in. gaseous ef-fluents, as determined by sampling frequency and meas-urement, is used for determining the gaseous pathway doses.
The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year also includes, as required by Technical Specification 3.11.4, an assess-ment of radition' doses to the likely most exposed MEM-BER Of-THE PUBLIC from Reactor releases and other r
Rev. 1 nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calender year to show conformance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation".
The Radioactive Effluent Release Reports include a list
. and description of unplanned releases from the site to UNRESTRICTED AREAS'of radioactive materials in gaseous and liquid effluents made during the reporting period.
The Radioactive Efkluent Release Reports include any changes made during the. reporting period to the PROCESS CONTROL PROGRAM and to the ODCM, pursuant to Specifica-tion 6.13 and 6.14, respectively, as well as any major change to Liquid, Gaseous, or Solid Radwaste Treatment System, pursuant to Specification 6.15.
It also in-cludes a listing of new locations for dose calculations
- and or environmental. monitoring identified by the Land Use Census pursuant to Specification 3.12.2.
The Radioactive Effluent Release Reports also include the following information:
An explanation as to why the inoperability"of liquid or gaseous effluent moni-p toring instrumeritation, was not corrected within the time specified in Specification 3.3.3.10 or 3.3.3.11, respectively; and description of the events leading to Y
,' liquid holdup tanks or gas enorage tanks exceeding the i
limits of Specification 3.11,.1.4 or 3.11.2.5, respectively.
,j '^
- In lieu of submission, the Union Electric Company has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.
t O
(Ref.9.4) s l'
I
.4
+
w T
4
[?
94 -
i r
\\
a
Riv. 1 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 assessrtet, com-puter codes are utilized to implement the ODCM methodologies.
These calculational methods include the same general features as provided in the ODcM.
These codes will be verified to produce results consistent with the ODCM methodologies. -.
r Rev. 1
9.0 REFERENCES
9.1 Title 10, " Energy", Chapter 1, Code of Federal Regulations, Part 20; U.S. Government Printing Office, Washington, D.C. 20402.
9.2 Title 10, " Energy", Chapter 1, Code of Federal Regulations, Part 50, Appendix I; U.S. Govern-ment Printing Office, Washington, D.C. 20402.
9.3 Title 40, " Protection of Environment", Chapter 1, Code of Federal Regulations, Part 190; U.S.
Government Printing Office, Washington, D.C.
20402.
9.4 Callaway Technical Specifications, Section 3.3.3.9, 3.3.3.10, 3/4.11, 3/4.12, and 6.9.1.7 as submitted to the U.S. Nuclear Regulatory Commisssion, August 1983.
9.5 Communications 9.5.1 Letter NEO-54, D.W. Capone to S.E. Milten-berger, dated January 5, 1983; Union Electric Company correspondence.
9.5.2 Letter BLSE 12,825, J.H. Smith (Bechtel Power Corporation) to N.A. Petrick (SNUPPS), dated October 7, 1983.
9.5,3 Letter NEO-114, D. W. Capone to G. L. Ran-dolph, dated November 8, 1983.
Union Electric Correspondence.
9.5.4 Ping Wan (Bechtel Power Corporation) to C.C.
Graham, Personal Communication, November 17, 1983.
9.6 Union Electric Company Callaway Plant, Unit 1, Final Safety Analysis Report.
9.6.1 Section 11.5.2.2.3.1 9.6.2 Section 11.5.2.2.3.4 9.6.3 Section 11.5.2.1.2 9.6.4 Section 11.5.2.2.3.2 9.6.5 Section 11.5.2.2.3.3 9.6.6 Section 11.2.3.3.4 t
-r Rev. 1 9.6.7 Section 11.2.3.4.3 9.6.8 Section 11.5.2.3.3.1 9.6.9 Section 11.5.2.3.3.2 9.6.10 Section 11.5.2.3.2.3 9.6.11 Section 11.5.2.3.2.2 9.6.12 Section 2.3.5 9.6.13 Section 2.3.5.1 9.6.14 Section 2.3.5.2.1.2 9.6.15 Section 2.3.5.2.3 9.6.16 Table 2.3-59 9.6.17 Section 9.2.6 9.6.18 Section 9.2.7.2.1 9.6.19 Section 6.3.2.2 9.6.20 Table 11.1-6 9.7 Union Electric Company Callaway Plant Environ-mental Report, Operating License Stage.
9.7.1 Table 2.1-19 9.7.2 Section 2.1.2.3 9.7.3 Section 2.1.3.2.8 9.7.4 Sectica 2.1.3.3.4 9.7.5 Section 2.1.3.1.3 9.8 U.S. Nuclear Regulatory Commission,
" Preparation of Radiological Effluent Techni-cal Specification For Nuclear Power Plants",
USNRC NUREG-0133, Washington, D.C.
20555, Oc-tober 1978.
9.8.1 Pages AA-1 through AA-3 9.8.2 Sectoin 5.3.1.3 9.8.3 Section 4.3 r-1 Rnv. 1 9.8.4 Section 4.3.1 9.8.5 Section 5.3.1.5 9.8.6 Section 5.1.1 9.8.7 Section 5.1.2 9.8.8 Section 5.2.1 9.8.9 Section 5.2.1.1 9.8.10 Section 5.3.1 9.8.11 Section 3.8 9.8.12 Section 3.3 9.9 U.S. Nuclear Regulatory Commission, "XOQDOQ, Program For the Meterological Evaluation Of Routine Effluent Releases At Nuclear Power Stations", USNRC NUREG-0324, Washington, D.C.
20555.
9.9.1 Pages 19-20 Subroutine PURGE 9.10 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.
9.10.1 Section c.1.b 9.10.2 Section c.l.c 9.10.3 Section c.2.b 1
9.10.4 Section c.2.c 9.10.5 Section c.4 9.11 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.
9.11.1 Appendix C, Section 3.a 1 _
s Rev. 1 9.11.2 Appendix E, Table E-15 9.12 U.S. Nuclear Regulatory Commission, " Methods for Demonstrating LWR Compliance with the EPA Uranium Fuel Cycle Standard (40 CFR Part 190)", USNRC NUREG-0543, Washington, D.C.
20555, January 1980.
9.12.1 Section I, Page 2 9.12.2 Section IV, Page 8 9.12.3 Section IV, Page 9 9.12.4 Section III, Page 6 9.13 U.S. Nuclear Regulatory Commission, " Standard Radiological Effluent Technical Specifications for Pressurized Water Reactors", USNRC NUREG-0472, Draft Revision 3, Washington, D.C.
20555, January 1983.
9.13.1 Definition 1.7, Page 1-2 9.14 Management Agreement for the Public Use of i
Lands, Union Electric Company and the State of Missouri Department of Conservation, December 21, 1982.
9.14.1 Exhibit A 9.15 Wildlife Code of Missouri, Rules of the Con-servation Commission, Issued January 1, 1983.
9.16 Miscellaneous References 4
9.16.1 Drawing Number M-109-0007-06, Revision 5. -. _ _ _
~ - - -
. ~
~
UNION ELECTRIC COM PANY 1901 GR ATIOT STREET ST. Louis, MISSOU RI 4
November 22, 1983 MAILING ADDRESS:
DON ALD F. SCHNELL P. O. SOM 949 vaca pnessormet ST. LOUIS, MISSOURt SS I SS Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulations U.
S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Denton:
ULNRC - 6 8 8 DOCKET NUMBER 50-483 CALLAWAY PLANT, UNIT 1 OFFSITE DOSE CALCULATION MANUAL REV. 1
References:
1)
ULNRC-613, dated March 21, 1983 2)
NRC letter dated October 26, 1983, from B.
J.
Youngblood Reference 1 transmitted the Callaway Plant Offsite Dose Calculation Manual (ODCM) to NRC for review.
The NRC staff comments on the ODCM were transmitted by reference 2.
Enclosed herewith for NRC review and approval are five copies of the ODCM Revision 1 which incorporates the NRC comments from reference 2.
By copy of this letter, we are also transmitting a copy of.the ODCM to Messrs. Joseph Holonich'and E. Branagan.
Very truly yours, 1
l Donald F.
Schnell DFS/BFH/msc Attachment - ODCM (5) e C
chs O
e
STATE OF MISSOURI )
)
Donald F. Schnell, of lawful age, being first duly sworn upon oath says that he is Vice President-Nuclear and an officer of Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the f acts therein stated are true and correct to the best of his knowledge, information and belief.
By Donald F. 'Schnell Vice President Nuclear SUBSCRIBED and sworn to before me this EdfI day of 1983 bb~0 &
BARRARMd.'PFAff # 8 NOTARY PUBLIC, STATE OF MISSOURI MY COMMISSION EXPIRES APR!L 22,1935 SL LOUIS COUNTY.
l l
cc:
Glenn L. Koester Vice President Operations Kansas Gas & Electric P.O. Box 208 Wichita, Kansas 67201 Donald T. McPhee Vice President Kansas City Power and Light Company 1330 Baltimore Avenue Kansas City, Missouri 64141 Gerald Charnof f, Esq.
Shaw, Pittman, Potts & Trowbridge 1800 M. S tr ee t, N.W.
Washington, D.C.
20036 Nicholas A. Petrick Executive Director SNUPPS 5 Choke Cherry Road Rockville, Maryland 20850 John H. Neisler Callaway Resident Office U.S. Nuclear Regulatory Commission RRil Steedman, Missouri 65077 J. Holonich (NRC)
E. Branagan (NRC)
-