ML20247D829
| ML20247D829 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 07/19/1989 |
| From: | Baxter L UNION ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20247D815 | List: |
| References | |
| APA-ZZ-01003, APA-ZZ-1003, ULNRC-2070, NUDOCS 8909150098 | |
| Download: ML20247D829 (186) | |
Text
{{#Wiki_filter:_ - - _ _ . - . . _ - _ - - _ _ - - - - - . _ _ ..;_--- Attachment 4 ULNRC- 2070
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f. OFFSITE DOSE CALCULATION MANUAL l 8909150098 890906 3 PDR ADOCK 0500 P-
s APA-ZZ-01003 Rsvision 0 July 18, 1989 L NUCLEAR FUNCTION ADMINISTRATIVE PROCEDURE APA-ZZ-01003 OFFSITE DOSE, CALCULATION MANUAL RESPONSIBLE DEPARTMENT HEALTH PHYSICS
' PREPARED BY _ hs[u . DATE -7//t//f l APPROVED BY ,
DATE DATE' ISSUED - This procedure contains the following: Pages 1 through 178 Attachments through Tables I through 12 Figures 4 through 5 Appendices through Checkoff Lists through _. 1
c- .- APA-ZZ-01003 Rev. O i l 1 Table of Contents Section Page Number
1.0 Purpose and Scope
1 2.0 Liquid Effluents 1 f 2.1 Radioactive Effluent Controls (REC) Section i 9.1.1.1 1 { Liquid Effluent Monitors 2.2 2 2.3 Calculation of Liquid Effluent Monitor Setpoints 5 2.4 Liquid Effluent Concentration Measurements 11 2.4.1 REC Section 9.3.1.1 11 2.4.2 Liquid Effluent Concentration Measurements 11 2.5 Dose Due to Liquid Effluents 12 2.5.1 REC Section 9.4.1.1 12 2.5.2 The Maximum Exposed Individual 12 2.5.3 Calculation of Dose from Liquid Effluents 13 ; 2.5.4 Summary, Calculation of Dose Due to Liquid Effluents 15 2.6 Liquid Radwaste Treatment System 19 2.6.1 REC Section 9.5.1.1 19 2.6.2 Operability of the Liquid Radwaste Treatment System 20 3.0 Gaseous Effluents 21 3.1 REC Section 9.2.1.1 21 3.2 REC Section 9.6.1.1 21 l 3.3 Gaseous Effluent Monitors 21 3.4 Calculation of Gaseous Effluent Monitor Setpoints 25 3.4.1 Total Body Dose Rate Setpoint Calculations 26 3.4.2 Skin Dose Rate Setpoint Calculations 28 3.4.3 Gaseous Effluent Monitors Setpoint Determination 30 3.4.4 Summary, Gaseous Effluent Monitors Setpoint Determination 32 3.5 Calculation of Dose from Gaseous Effluents 32 3.5.1 Calculation of Dose Rate 32 { ' 3.5.1.1 Noble Gases 32 3.5.1.2 Radionuclides Other Than Noble Gases 34 3.5.2 Individual Dose Due to Noble Gasec 39 3.5.2.1 REC Section 9.7.1.1 39 3.5.2.1.1 Noble Gases 40 l 3.5.2.2 REC Section 9.8.1.1 42 3.5.2.2.1 Radionuclides Other Than Noble Gases 42 3.6 Gaseous Radwaste Treatment System 64 3.6.1 REC Section 9.9.1.1 64 I I l i m._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
1 APA-ZZ-01003 ! Rnv. O j l i Table of Contents
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0 Section Page Number j 3'. 6. 2 Description of the Gaseous Radwaste Treatment System 64 1 3.6.3 operability of the Gaseous Radwaste l Treatment System 65
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4.0 Dose and Dose Commitment from Uranium Fuel Cycle Sources 65 4.1 REC Section 9.10.1.1 65 4.2 Calculation of Dose and Dose Commitment from Uranium Fuel Cycle Sources 65 1 4.2.1 Identification of the MEMBER OF THE PUBLIC 67 f 4.2.2 Total Dose to the Nearest Resident 67 l 4.2.3 . Total Dose.to the Critical Receptor Within 68 the SITE BOUNDARY 5.0 Radiological Environmental Monitoring 73 1 5.1 REC Section 9.11.1.1 73 5.2 Description of the Radiological Environmental Monitoring Program 73 5.3 Performance Testing of Environmental Thermoluminescence Desimeters 74 . 6.0 Determination of Annual Average and Short Term Atmospheric Dispersion Parameters 92 6.1 Atmospheric Dispersion Parameters 92 6.1.1 Long-Term Dispersion Estimates 92 6.1.2 Determination of Long-Term Dispersion Estimates for Special Receptor Locations 93 6.1.3 Short-Term Dispersion Estimates 93 7.0 Semi-Annual Radioactive Effluent Release Report 101 8.0 Implementation of ODCM Methodology 107 9.0 Radioactive Effluent Controls (REC) 108 9.1 Radioactive Liquid Effluent Monitoring Instrumentation 109 9.2 Radioactive Gaseous Effluent Monitoring Instrumentation 117 9.3 Liquid Effluents Concentration 125 9.4 Dose 131 9.5 Liquid Radwaste Treatment System 133 9.6 Gaseous Effluents Dose Rate 135 9.7 Dose-Noble Gas 141
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APA-ZZ-01003 Rav. O ! Table of Contents Section Page Number 9.8 Dose-Iodine-131 and 133, Tritium, and Radioactive Material in Particulate Form 143 - 9.9 Gaseous Radwaste Treatment System 145 9.10 Total Dose 147 9.11 Radiological Environmental Monitoring Program 151 9.12 Radiological Environmental Monitoring Land Use Census 167 9.13 Radiological Environmental Monitoring Inter-laboratory Comparison Program 170 10.0 Administrative Control 171 10.1 Major Changes to Liquid and Gaseous Radwaste ' Treatment Systems 171 10.2 Changes to the Offsite Dose Calculation Manual (ODCM) 172 , 11.0 References 173 Figure 4.1 Site Area Closed to Public Use 72 Figure 5.1A Airborne & TLD Sampling Network 87 Figure 5.1B Airborne & TLD Sampling Network 88 Figure 5.2A Location of Aquatic Sampling Stations 89 Figure 5.2B Location of Aquatic Sampling Stations 90 l Figure 5.3 Food Products Sampling Locations 91 4
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t-APA-ZZ-01003 [ R5v. O I Table of Contents ) Section Page Number ! i Table 1 Ingestion Dose Commitment Factor ( Ag ,) for Adult Age Group 16-17 Table 2 Bioaccumulation Factor (BFg ) Used in the Mosence of Site-Specific Data 18' ] Table 3 Dose Factors for Exposure to A Semi-Infinite i Cloud of Noble Gases 31 Table 4 Dose Parameter (P g ) for Radionuclides Other Than Noble Gases . 36-38 Table 5 Pathway Dose Factors (R g ) for Radionuclides i Other Than Noble Cases 46-63 L . Table 6 Radiological Environmental Monitoring Program 75-82 Table 7 Reporting Levels for Radioactivity Concentrations in Environmental Samples 83 , I Table 8 Maximum Values for the Lower Limits of Detection 84 Table 9 Highest Annual Average Atmospheric Dispersion Parameters - Radwaste Building Vent 97 Table 9.1-A Radioactive Liquid Effluent Monitoring
. Instrumentation 112-114 Table 9.1-B Radioactive Liquid Effluent Monitoring i Instrumentation Surveillance Requirements 115-116 l Table 9.2-A Radioactive Gaseous Effluent Monitoring i Instrumentation 119-121 Table 9.2-B Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 122-124 Table 9.3-A Radioactive Liquid Waste Sampling and Analysis Program 127-130 Table 9.6-A Radioactive Gaseous Waste Sampling and Anslysis Program 137-140 Table 9.11-A Radiological Environmental Monitoring Program 155-161 Table 9.11-B Reporting Levels for Radioactivity Concentrations In Environmental Samples 162 ,
Table 9.11-C Detection Capabilities for Environmental ! Sample Analysis 163-1f6 Table 10 Highest Annual Average Atmospheric Dispersion Parameters - Unit Vent 98 Table 11 Short Term Dispersion Parameters 99 . Table 12 Application of Atmospheric Dispersion I Parameters 100
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APA-ZZ-01003 I R;v. 0 1 Record of Revisions 1 1 R; vision Reason j Number Date for Revision j l R v. O March 1983 I
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Rsv. 1 November 1983 Revised to support the current RETS i submittal and to incorporate NRC ! Staff comments Rev. 2 March 1984 Revised to incorporate NRC Staff j comments l 1 R3v. 3 June 1985 Revised to incorporate errata identified by ULNRC-803 and changes to the Environmental Monitoring Program. Incorporate results of 1984 Land use Census. ; I Rsv. 4 February 1987 Minor clarifications, incorporated )' 31-day projected dose methodology. Change in the utilization of areas within the Site Boundary. Rsv. 5 January 1988 Minor clarifications, revised descriptions of liquid and gaseous ) rad monitors, revised liquid setpoint methodology to incorporate monitor background, revised dose j calculations for 40CFR190 requirements, Revised Table 6 and Figures 5.1A and 5.1B to refine I descriptions of environmental TLD stations, incorporated description of environmental TLD testing required by Reg. Guide 4.13, revised Tables 1, 2, 4, and 5 to add additional nuclides, deleted redundant material from Chapter 6.
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APA-ZZ-01003 Rav. O f . Record of Revisions Rev. 6 May 1989 Revised methodology for calculating maximum permissible liquid effluent
' discharge rates land. liquid effluent monitor setpoints,_provided . .
methodology for' calculating liquid effluent monitors response > correction factors, provided an enhanced description of controls on liquid monitor background limits,' provided additional liquid'&- gaseous dose conversion factors and' bioaccumulation factors (Tables 1, 2, 4 & 5), provided description of the uselof the~setpoint required by Technical Specification.4.9.4.2-during Core Alterations, added discussion of gaseous & liquid monitor setpoint selection in the event that the' sample.contains no-detectable activity, added minimum holdup requirements for Waste Gas Decay Tanks, revised dispersion parameters & accompanying description per FSAR Change Notice 88-42. APA-ZZ-01003 August 1989 Radiological Effluent Technical Rev. O Specifications were moved from.the callaway Plant Standard Technical Specifications to Section 9.0, Radioactive Effluent Controls, of the ODCM as per NRC Generic Letter 89-01. At the same time, in order to formalize control of the entire ODCM, it was converted to APA-ZZ-01003, OFFSITE DOSE CALCULATION MANUAL.
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[l; APA-ZZ-01003 Rsv. O b p OFFSITE DOSE CALCULATION MANUAL 1.0 PURPO,SE AND SCOPE 1.1 The OFFSITE' DOSE CALCULATION' MANUAL.(ODCM) . ' describes the methodology and parameters used'in the calculation of.offsite doses resulting from radioactive gaseous and liquid. effluents, in the . calculation of gaseous and liquid effluent
' monitoring' Alarm / Trip.Setpoints, and in'the conduct of.the Environmental Radiological Monitoring Program. The ODCM also contains'the
. Radioactive Effluent Controls and Radiological-Environmental Monitoring Programs required by Technical Specification 6.8.4, and descriptions of the~information that should be included in thes Annual Radiological Environmental Operating and Semi-annual Radioactive Effluent-Release Reports required by Technical Specifications 6.9.1.6 and 6.9.1.7. -The ODCM also contains a list and description of the specific sample locations for the radiological environmental monitoring program.
'2.0 LIQUID EFFLUENTS 2.1 RADIOACTIVE EFFLUENT CONTROLS (REC) SECTION 9.1.1.1 41838 The radioactive liquid effluent monitoring instrumentation channels shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Section-9.3.1.I are not exceeded. The 41840 alarm / trip setpoints of these channels shall be adjusted to the values determined in accordance with the methodology and parameters in the ODCM.
l- APA-ZZ-01003 I Rsv. O 2.2 Liquid Effluent Monitors
, Gross radioactivity monitors which provide for automatic termination of liquid effluent' releases are present on the liquid effluent lines. Flow rate measurement 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. .
l monitors are provided in the appropriate Plant Procedures. Setpoint values are calculated to assure that alarm and-trip actions occur prior to exceeding the Maximum Permissible Concentration (MPC) limits in 10 CFR Part 20 at the release point to the UNRESTRICTED AREA. The calculated alarm and trip action setpoints for the liquid-effluent line monitors and flow measuring devices must satisfy the following equation: cf -
< C F+f (2.1)
Where: C= the liquid effluent concentration limit (MPC) implementing Section 9.3.1.1 for the site in (pci/ml). c= The setpoint, in (uCi/ml), of the radioactivity monitor measuring the radioactivity concentration in the effluent line prior to dilution and subsequent release; the setpoint, which is inversely related to the volumetric flow of the effluent line and directly related to the volumetric flow of the dilution stream plus the effluent stream, represents a value, which, if exceeded, would result in l concentrations exceeding the limits of 10 CFR Part 20 in the UNRESTRICTED AREA. i l 4 i l i i 1
h L APA-ZZ-01003 Rsv. O 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. 11.8.1) If no-' dilution is provided, then c 5 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 supplies the necessary. dilution flow via a bypass line. The batch release limiting concentration-(c) which corresponds ~to the liquid radwaste effluent line monitor setpoint is to be calculated using methodology from the expression above.
= _ _ _ . - _ - _____ _ _ - _ -
'M a
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l' l l H Thus, the expression for determining the.setpoint. on'the. liquid radwaste effluent line monitor.- ,I becomes: I c$ C(F + f) (yCi/ml) .! f (2.2) The alarm / trip setpoint calculations are based on
. the minimum dilution flow rate (cooling tower blowdown, 5000 gpm), the maximum effluent stream
- flow rate, and the actual' isotopic analysis. Due to the possibility of a simultaneous 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.
In the. event the alarm / trip setpoint is reached, an evaluation will be performed-using actual dilution and effluent flow values and actual isotopic analysis to ensure that'Section 9.3;1.1<
' limits were not exceeded.
2.2.1- Continuous Liquid Effluent Monitors The radiation detection monitors associated with j continuous liquid effluent releases are (Ref. -' 11.6.1, 11.6.2): l Monitor I.D. Description BM-RE-52 Steam Generator Blowdown Discharge Monitor LE-RE-59 Turbine Building Drain Monitor 3 l; l - -_ ---___ . ___ i-
H ue : APA-ZZ-01003. Rsv. O 15
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These effluent streams are not considered to be radioactive unless radioactivity has'been detected by-the associated effluent radiation monitor or by laboratory analysis. The sampling b frequency, minimum analysis frequency, and type l of analysis performed-are as per Table.9.3-A. Since the. Turbine Building Drain is not a designated' liquid radioactive affluent discharge point, its setpoint is established as.a function of background. 2.2.2 Radioactive Liquid Batch Release Effluent Monitors The two radiation monitors which are associated with the liquid effluent batch release systems are (Ref. 11.6.4, 11.6.5): MONITOR I.D. Description HB-RE-18 Liquid Radwaste Discharge Monitor
-HF-RE-45 Secondary Liquid Waste System Monitor These effluent streams are normally considered to be radioactive. The sampling frequency,' minimum analysis' frequency,.and the type of analysis performed are-as per Table 9.3-A.
2.3 Calculation of Liquid Effluent Monitor Setpoints The dependence of the setpoint (c), on the radionuclides distribution, yields, calibration, and monitor parameters, requires that several variables be considered in setpoint calculations. (Ref. 11.8.1)
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1 APA-ZZ-01003 1 Rsv. 0 1 2.3.1 Calculation of the MPC Sum j l The isotopic concentration of the release (s) . being considered must be determined.. This'is' ; obtained from the analyses required per Table i 9.3-A, and is used to calculate an MPC sum , (MPCSUM): ] 1 c MPCSUM = (I(C g)g/(MPCg)g)+(C,/MPC,)+(C,/MPC,)+(Ct / t I*I f / C) f (2.3) i Where: (Cg )g = the concentration (C g) of each measured gamma emitting nuclide, i, observed by gamma-ray spectroscopy of the waste sample. C,* = the' measured concentration (C,) of alpha emitting nuclides observed by gross alpha analysis. C,* = the measured concentrations of Sr-89 and Sr-90 as determined by analysis of the quarterly composite sample. C t* = the measured concentration of H-3.in liquid effluents. C f* = the measured concentration of Fe-55 in liquid waste as determined by analysis of the quarterly composite sample.
- Values for these concentrations will be based on previous conposite sample analyses as required by Table 9.3-A.
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'P MPC 9 , MFC,, MPC,, MPC f , MPC t ; = are the limiting
' concentrations'of'the appropriate. radionuclides
.from 10CFR 20, Appendix B, Table II, Column'2.
For dissolved or entrained noble gases,.the concentration shall be limited .to. 2x107' pCi/mi total activity. SF = the safety factor; an administrative factor used to compensate for. statistical. fluctuations and errors of measurements. This factor also provides a margin of' safety in the calculation of.the maximum liquid affluent discharge flowrate (f,,x). The value.of SF should be fl. For the case MPCSUM $ 1, the monitor tank effluent' concentration meets the limits of Section 9.3.1.1 without dilution and the affluent: may.be released.at any desired flow rate.- .If . MPCSUM > 1 then dilution is required to ensure compliance with.Section 9.3.1.1 concentration
. limits. If simultaneous releases are occuring or are anticipated,-an allocation fraction, N, must.
be applied so that available dilution flow may,be ' apportioned among simultaneous discharge pathways.. The.value of N may be any value between 0 and-1 for a particular discharge point, provided that the sum total allocation fractions l for all discharge points must be $ 1. L L L___________ __
APA-ZZ-01003 Rav. O 2.3.2 calculation of the Maximum Permissible Liquid Effluent Discharge Flowrate The maximum permissible liquid effluent discharge . flowrate is calculated by: f,,x $ (F + f )-(SF) p (N) + (MPCSUM) (2.4) Where:
= Maximum permissible liquid effluent f,,3 discharge flowrate, in (gallons / minute);
f = the expected undiluted liquid effluent P. flowrate, in gpm. N= the allocation fraction which apportions dilution flow among simultaneous-discharge pathways (see discussion above) F, SF, & MPCSUM, are as previously defined. The minimum value of F is 5000 gpm, which is used as a default value. The dilution water supply is furnished with a flow monitor which' isolates the liquid effluent discharge if the dilution flow rate falls below the 5000 gpm minimum value. In the event that f,,x is less than f ,p then the value of fmu is substituted into the equation for f and a new value of f,,x is calculated. p This substitution is performed for three
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iterations in order to calculate the correct value of f,,x. l
APA-ZZ-01003 g .Rev. 0 1 i
.2.3.3 Calculation of Liquid Effluent Monitor Setpoint The liquid effluent monitors are NaI(Tl) based =
systems and respond primarily to gamma radiation. j Accordingly, their setpoint is based on the total i concentration of gamma emitting nuclides in the effluent: i c = BKG + (I(Cg)g + SF) pCi/ml (2.5) i i Where: e= the monitor setpoint as previously defined, in (pCi/ml); BKG = the monitor background prior to discharge, in (yCi/ml); g and SF are as previously defined. C The monitor's background is controlled at an appropriate limit to ensure adequate sensitivity. Utilizing the methodology of ANSI.N13.10-1974 (Ref. 11.21), the background must be maintained at a value of less than or equal to 2.23E-6 pCi/ml (relative to Cs-137) in order to detect a change of 1E-7 pCi/ml of I-134 (the most restrictive nuclide in Table 1 of reference 11.21). In the event that there is no detectable gamma activity in the effluent or if the value of (I(Cg)g+SF) is less than the background of the monitor, then the monitor setpoint will be set at twice the current background of the monitor. 9 _m._ _._. ___-_.._. _
APA-ZZ-01003 ; Rsv. O t-As previously stated, .idue monitor's response is dependent on the gamma emitting radionuclides J distribution of the effluent. Accordingly, a new ' database conversion factor is calculated for each release based upon.the.results of the gamma spectrometric analysis'of the' effluent sample and the measured response of the monitor to the
. National Bureau of Standards (NBS) traceable calibration sources:
i DBCF c = (I(C g)f)+(CMR) x (ECF) (2.6) Where: DBCF, = the monitor data base conversion factor which converts count rate into concentration (VCi/ml); CMR = the calculated response of the radiation monitor to the liquid effluent; ECF = the conversion factor for Cs-137, which converts count rate into concentration (pCi/ml). C g is as previously defined. The new value of the DBCE is calculated and c entered into the monitor data base prior to each discharge. A more complete discussion of the derivation and calculation of the CMR.is given in reference 11.14.7.
APA-ZZ-OlOO3 Rsv.. O L i 2.4 Liquid Effluent Concentration Measurements . 41846 ;2. 4.1 REC Section 9.3.1.1 4160 The concentration of radioactive material l released in liquid effluents to UNRESTRICTED 1 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 - i
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dissolved or entrained noble-gases, the concentration shall be limited to 2.0 E-04 yCi/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 representative 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. Concentrations are determined primarily from the gamma isotopic, H-3, & gross alpha analyses of the liquid batch sample. For Sr-89, Sr-90, & Fe-55, the measured l concentration from the previous composite j analysis is used. Composite samples are , collected for each batch release and quarterly i analyses are performed in accordance with Table 9.3-A. j l I
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P :: I n Doses from liquida discharged as continuous releases are calculated by utilizing the last L
- measured values of samples required in accordance with Table 9.3-A.-
i 2.5: Dose Due to Liquid Effluents "41849 2.5.1 REC Section 9.4.1.1. 4160 The dose'or dose commitment to a MEMBER OF THE
- PUBLIC'from radioactive materials in liquid effluents released, to UNRESTRICTED AREAS shall-be: limited:
- a. During any calendar quarter to less than or equal to 1.5 mrem to the whole 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-total 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 exposed individual. (Ref. 11.8.3) The Callaway Plant's' liquid effluents are discharged to the Missouri River. As there are ' no potable water intakes within 50 miles of the discharge point (Ref. 11.7.1, 11.6.6), this pathway does not require 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. 11.6.7) aa --__.___m________.u____ . . _ _ _ ._____ _-
L APA-ZZ-OlOO3 -{ RSv. O j y i 2.5.3 Calculation of Dose From Liquid Effluents. 2.5.3.1 Calculation of Dose contributions 1
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The dose contributions for the total' time period m IAt g t=1 . are calculated at least'once each 31' days and a
-cumulative summation of the total body'and. ,
individual. organ doses is maintained for-each. ;
. calendar quarter. These dose contributions are l calculated for all radionuclides identified in '
liquid' effluents released to UNRESTRICTED AREAS
.using the following expression (Ref. 11".8.3)
I l m D = I [A it I Atg C gg F] g (2.12)- i g=y Where: D t = the cumulative dose commitment to the total body or any organ, t, from the liquid effluents for the total period m IAt g t=1 in mrem.
o 7" , APA-ZZ-01003 Rav. 0 j
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.I Atg = the11ength of the Ath time period over
)
which.c gg and Fg are averaged for all i liquid releases, in hours. Atg corresponds to the actual duration.of the release (s). Ogg = the average measured concentration of radionuclides, i, in undiluted liquid effluent during time period At g from any liquid release, in (yci/ml). Ag , = the site related ingestion dose commitment factor to the total body or any organ t for each identified principal gamma and beta emitter. listed in Table 9.3-A,.(in mrem /hr) per-(yci/ml). The calculation of the Ag ,
-values is detailed in Ref. 11.14.5 and are given.in Table 1.
F1 = the near field average dilution factor for C gg during any liquid effluent release. I f ~ t= max _(F.+ f,,x) 89.77 Where: f = maximum undiluted affluent flow rate max during the release. F= average dilution flow 89.77.= site specific applicable factor for the mixing effect of the discharge structure. (Ref 11.5.1) The term C yg is the. undiluted concentration of radioactive material in liquid waste at the common release point determined in accordance with Section 9.3.1.1, Table 9.3-A, " Radioactive Liquid Waste Sampling and Analysis Program". All dilution factors beyond the sample point (s) are H included in the Fg term. l _=_____=___--
mo x ' 4
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The nearest municipal potable water intake o downstream from the liquid. effluent discharge
- point into the Missouri River is located'near-the g city of St.. Louis,'Mo., approximately 78' miles p
downstream. As there--are currently no" potable-water intakes within 501 river miles of the discharge point, the drinking. water pathway is-
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r > not included in dose estimates lto.the maximally exposed individual, or in dose estimates to.the population..fShould future-water intakes be-constructed within 10 river miles' downstream of the discharge point, then.this manual will be
- revised to~ include dais pathway 1n dose i
estimates. '(Ref. 11.6.6). 2.5.4 Summary, calculation of Dose Due to Liquid Effluents. The-dose contribution for the total time period m IAt g tal L is determined ~by calculation at least once per 31 deys.and.a cumulative summation of the total body and organ' doses is maintained for each calendar quarter. The projected' dose' contribution from liquid effluents for which radionuclides
- concentrations are determined by periodic composite and~ grab sample analysis, may be approximated by using the last measured value.
Dose contributions are determined for all radionuclides identified in liquid effluents released to UNRESTRICTED AREAS. Nuclides which are not detected in the analyses are reported as "less than" the nuclide's Minimum Detectable Activity (MDA) and are not reported as being present at the Lower Level of Detection (LLD) level for that nuclide. The "less than" values. are not used in the required dose calculations. I j l 1 ________ _ A
APA-ZZ-01003 Rov. O TABLE 1 INGESTION DOSE COMMITMENT FACTOR (Ah) FOR ADULT AGE GROUP (mres/hr) per (sci /ml) l l .j l Total l l l l l lNuclidel Bone l Liver l Body l Thyroid i Kidney I Lunt i GI-LLI l
, H-3 lNo Data l2.26E-01 l2.26E-01l2.26E-01 l2.26E-01l2.26E-01l2.26E-01l Be-7 11.30E-02l2.98E 02 11.45E-02lNo Data l3.15E-02lNo Data l5.16E+00 C-14 13.13E+04l6.26E+03 16.26E+03l6.26E+03 l6.26E+03l6.26E+0316.26E+03
, Na-24 l4.07E+02l4.07E+02 14.07E+0214.07E+02 l4.07E+0214.07E+02l4.07E+02l l P-32 14.62E+0712.87E+06 l1.78E+06lNo Data (No Data lNo Data l5.19E+06l L Cr-51 lNo Data lNo Data .1.27E+00l7.62E-01 l2.81-01 ll.69E+00l3.20E+02l l Mn-54 lNo Data l4.38E+03 8.35E+02]No Data l1.30E+03lNo Data 11.34E+04l .
Mn-56 lNo Data 11.10E+02 ;1.95E+01[No Data l1.40E+02lNo Data 13.52E+031 Fe-55 16.57E+0214.54E+02 l1.06E+02]No Data lNo Data l2.53E+02 2.61E+02l Fe-59 l1.04E+03l2.44E+03 l9.34E+02lNo Data [No Data l6.81E+02 8.13E+03l l Co-57 lNo Data 12.09E+01 l3.48E+01lNo Data lNo Data lNo Data 15.31E+02l l Co-58 lNo Data 18.94E+01 12.00E+02lNo Data lNo Data (No Data l1.81E+031 l Co-60 lNo Data l2.57E+02 15.66E+02lNo Data lNo Data lNo Data ]4.82E+03l l Ni-63 13.11E+0412.15E+03 l1.04E+03lNo Data INo Data lNo Data 14.49E+02] l Ni-65 11.26E+0211.64E+01 l7.45E+00lNo Data lNo Data (No Data 14.16E+02l l Cu-64 lNO Data 11.00E+01 l4.69E+00lNo Data 12.52E+01lNo Data 18.52E+02l l Zn-65 12.32E+0417.38E+04 l3.33E+04lNo Data l4.93E+04lNo Data 14.65E+04l l Zn-69 l4.93E+01]9.44E+01 16.56E+00lNo Data 16.13E+01lNo Data l1.42E+01l l Br-82 lNo Data INo Data l2.27E+03]No Data INo Data [No Data 12.60E+03l l Br-83 lNo Data [No Data l4.04E+01lNo Data (No Data INo Data l5.81E+01] l Br-84 lNo Data lNo Data 15.26E+01lNo Data lNo Data INo Data l4.13E-041 l Br-85 lNo Data lNo Data l2.15E+00lNo Data (No Data lNo Data 1 0 l l Rb-86 (No Data 11.01E+05 14.71E+04lNo Data lNo Data [No Data l1.99E+04) l Rb-88 lNo Data 12.90E+02 l1.54E+02lNo Data lNo Data lNo Data ]4.00E-09l l Rb-89 lNo Data l1.92E+02 l1.35E+02lNo Data [No Data (No Data l1.12E-111 l Sr-89 12.21E+04lNo Data 6.35E+02lNo Data INo Data lNo Data 3.55E+031 l Sr-90 15.44E+05lNo Data 1.34E+05lNo Data (No Data (No Data 1.57E+04) l Sr-91 14.07E+02lNo Data 1.64E+01lNo Data (No Data {No Data 11.94E+03l l Sr-92 11.54E+02lNo Data l6.68E+00lNo Data lNo Data lNo Data 13.06E+031 l Y-90 l5.75E-01lNo Data 11.54E-02lNo Data jNo Data lNo Data (6.10E+03l l Y-91M 15.44E-03lMo Data 12,10E-04lNo Data lNo Data lNo Data l1.60E-02l l Y-91 IB.43E+00lNo Data l2.25E-011No Data lNo Data lNo Data l4.64E+03l l Y-92 15.05E-02lNo Data 11.48E-03lNo Data (No Data lNo Data l8.85E+02l l Y-93 11.60E-01lNo Data 14.42E-03lNo Data {No Data INo Data 15.08E+03l l Zr-95 l2.40E-01l7.70E-02 15.21E-02lNo Data ll.21E-01lNo Data 12.44E+021 l Zr-97 11.33E-0212.68E-03 11.22E-03tNo Data 14.04E-03lNo Data 18.30E+02l l Nb 95 14.47E+02l2.48E+02 l1.34E+02lNo Data 12.46E+02]No Data 11.51E+06l l Mo-99 lNo Data l1.03E+02 11.96E+01lNo Data 12.33E+02lNo Data 12.39E+02l l Tc-99Ml8.87E-03l2.51E-02 l3.19E-01lNo Data l3.81E-01l1.23E-02l1.48E+011 l Tc-101l9.11E-03]1.31E-02 11.29E-01[No Data 12.36E-01l6.70E-031 0 l l Ru-103l4.42E+00lNo Data 11.90E+00lNo Data l1.69E+01]No Data 15.17E+021 l Ru-105l3.68E-01lNo Data l1.45E-01lNo Data 14.76E+00lNo Data 12.25E+02l l Ru 10el6.57E+011No Dats l8.32E+00lNo Data l1.27E+021No Data l4.25E+03l l Cd 1091No Data 15.54E+02 11.94E+01lNo Data l5.31E+021No Data l5.59E+031 i Sn-!!315.66E4 l1.61E3 13.26E3 19.18E2 lNo Data lNo Data 11.69E5 l l Sb- Relo.69E+00l1.26E-01 12.65E+0011.62E-02 lNo Data l5.21E+00ll.90E+02l
. - ~ _ .. . - __.
e
' APA-ZZ-01003-1; RQv. 0 TABLE 1 (Continued) l INGESTION DOSE COMMITMENT FACTOR (Ah) FOR ADULT AGE GROUP I
- (ares /hr) per (uci/al)
N
- l. l l l Total l l l lNuclidel Bone l Liver I Body l Thyroid ! Kidney Luna l GI-LLI
.lSb-125 l4.28E+00 4.78E-02 11.02E+00l4.35E-03 lNo Data l3.30E+0014.71E+01H lTe-125Ml2.57E+03 9.30E+02 l3.44E+02l7.72E+02 l1.04E+04lNo Data l1.02E+04 lTe-127Ml6.47E+0312.32E+03 17.90E+0211.66E+03 ;2.63E+04lNo Data 12.17E+04 ;
lTe-127 l1.03E+0213.78E+01 l2.28E+01l7.80E+01 4.29E+02lNo Data l8.30E+03'l lTe-129Mll.10E+04l4.11E+03 l1.74E+03[3.78E+03 4.60E+04tNo Data l5.54E+04l lTe-129 l3.01E+01l1.13E+01 17.33E+00l2.31E+01 l1.26E+02lNo Data l 2.27E+01l lTe-131Mll.66E+03l8.09E+02 16.75E+02l1.28E+03 l8.21E+03lNo Data ; 8.03E+04l lTe-131 l1.89E+0117.88E+00 5.96E+00]1.55E+01 18.25E+01!No Data 2.67E+00] Te-132 l2.41E+03l1.56E+03 1.47E+0311.72E+03 11.50E+04lNo Data 7.38E+04l I-130 12.71E+41l8.01E+01 ;13.16E+01l6.79E+03 l1.25E+02lNo Data h 6.89E+01l
' i I-131 '1.49E+02l2.14E+02 l1.22E+02l7.00E+04 l*i.66E+02lNo Data 15.64E+01l I-132 7.29E+00ll.95E+01' 6.82E+0016.82E+02 l3.11E+01lNo Data l3.66E+00l lI-133 5.10E+01l8.87E+01 2.70E+01l1.30E+04 l1.55E+02lNo Data l7.97E+01l lI-134 13.81E+0011.03E+01 :l3.70E+00l1.79E+02 l1.64E+01lNo Data l9.01E-03) lI-135 l1.59E+0114.16E+01 11.54E+01l2.75E+03 l6.68E+01[No Data (4.70E+01]
lCs-134 12.98E+05l7.09E+05 l5.80E+0$lNo Data l2.29E+0517.62E+04]1.24E+041 lCs-136 13.12E+04l1.23E+05 8.86E+04lNo Data l6.85E+04l9.39E+03tl.40E+04] lCa-137 l3.82E+05l5.22E+05 3.42E+05lNo Data :l 1. 77E+0515. 89E+0411. 01E+04 l lCs-138 12.64E+02l5.22E+02 l2.59E+02lNo Data 3.84E+0213.79E+01l2.23E-03l lBa-139 19.29E-01l6.62E-04 l2.72E-02tNo Data 6.19E-0413.76E-0411.65E+00l lBa-140 l1.94E+02l2.44E-01 lJ.27E+01lNo Data 18.31E-02l1.40E-01l4.00E+021 lBa-141 14.50E-0113.40E-04 l1.52E-02lNo Data l3.'16E-04 1.93E-04l2.12E-10l lBa-142 l2.04E-01l2.09E-04 l1.28E-02lNo Data l1.77E-04 1.19E-04l 0 l lLa-140 l1.50E-0117.53E-02 .1.99E-02lNo Data lNo Data lNo Data l5.53E+03l lLa-142 l7.65E-0313.48E-03 ;8.66E-04lNo Data lNo Data lNo Data 12.54E+01l lCo-141 l,1.24E-02l1.51E-02 l1.72E-03lNo Data 17.03E-03lNo Data 15.78E+01l {Co-143 l3.94E-0312.92E+00 l3.23E-04lNo Data 1.28E-03lNo Data 'll.09E+02l lCo-144 11.17E+0014.88E-01 l6.26E-02]No Data 2.89E-01lNo Data 3.94E+021 lPr-143 15.50E-01l2.21E-01 l2.73E-02lNo Data l1.27E-01lNo Data 2.41E+03l lNd-147 l3.76E-0114.35E-01 l2.60E-02lNo Data l2.54E-011No Data 12.09E+03l lEu-154 l3.67El l4.5 E0 13.21E0 lNo Data l2.16El lNo Data l3.27E3 l lHf-181 l3.99E-02 1.94E-01 11.80E-02lNo Data l4.17E-02lNo Data l2.21E+02] lW-187 l2.96E+02 2.47E+02 18.64E+01lNo Data INo Data lNo Data l8.09E+04l lNp-239 l2.84E-02l2.80E-03 11.54E-03lNo Data l8.72E-03lNo Data 15.74E+021
=_L_-_.__ - _ . . _ _ _ _ . . _ _ _ _
I-t. APA-ZZ-01003 I Rev. O j 1 h TABLE 2 BIOACCUMULATION-FACTOR (BF g ) USED IN THE ABSENCE OF SITE-SPECIFIC DATA" .j (pCi/kg) per (pCi/ liter)
.BF i
. Element Fish (Frashwater) ;
1 H. 9.0 E - 01
- Be f 2.0 E + 00
'3 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 + 03 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 Cd 2.0 E + O2 Sn 3.0 E + 03 Sb 1.0 E + 00 l Te 4.0 E + O2 I 1.5 E + 01 Cs 2.0 E + 03 Ba 4.0 E + 00 La 2.5 E + 31 Ce 1.0 E + 00 Pr 2.5 E + 01 Nd 2.5 E + 01 Eu 2.5 E + 01 Hf 3.3 E + 00 W 1.2 E + 03 Np 1.0 E + 01 (a) Values from Regulatory Guide 1.109, Rev 1, Table A-1 and References 11.14.4 and 11.14.8.
___.m____.___ . . _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _
APA-ZZ-01003 Rsv.-0 2.6 LIQUID RADWASTE TREATMENT SYSTEM 4160. 2.6.1 REC Section 9.5.1.1 41851 The LIQUID RADWASTE TREATMENT SYSTEM shall be-OPERABLE and appropriate portions of the system shall be used to reduce releases of radioactivity
-when the projected doses due-to the liquid effluent, to UNRESTRICTED AREAS, would exceed 0.06 mrem-to the total body or 0.2 mram to any organ in a 31 day period.
1 f f
l
, . 1 APA-ZZ-01003 )
Rsv. 0 , l 2.6.2' OPERABILITY Of The LIQUID RADWASTE TREATMENT SYSTEM The LIQUID'RADWASTE TREATMENT SYSTEM is capable ) H of varying treatment, depending on waste type and { product desired. It is capable of. concentrating, j gas. stripping, and distillation of liquid wastes- .) through the use of the evaporator system. The .i 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 j system 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 compliance with Sections 9.3.1.1 and 9.4.1.1. Projected doses due to liquid releases to UNRESTRICTED AREAS are determined each 31 days by dividing the cummulative annual total by the number of elapsed months.
)
b
s- _ _ _ . - -. _ _ _ . - _ _ _ L .APA-ZZ-01003 h Rsv. 0 1 L: 1 L 2902 3.0 GASEOUS EFFLUENTS 41842 3.1 REC Section 9.2.1.1
< 4160-The radioactive gaseous-effluent monitoring instrumentation channels shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Section 9.6'.1.1 are not exceeded. The i Alarm / Trip Setpoints of these channels shall be I adjusted to the values determined in accordance i with the methodology and parameters in the ODCM.
L- 41853 3.2 REC Section 9.6.1.1 4160 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 then or equal to 500 i 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 are present on the containment building 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 i based on the instantaneous noble gas total body and skin dose rate limits of Section 9.6.1.1, at the SITE BOUNDARY location with the highest annual average X/Q value. (Figure 5.1B)
y ,- F . APA-ZZ-01003 Rsv. 0 Each monitor channel is provided with a two level L system which provides sequential' alarms on
-increasing radioactivity levels. These setpoints are designated as alert setpoints and alarm / trip setpoints. (Ref. 11.6.3)
The radiation monitor alarm / trip setpoints for each release point-are based on the radioactive noble gases in gaseous effluents. It is not ' considered practicable to apply instantaneous alarm / trip setpoints to integrating radiation monitors sensitive to radioiodines, radioactive materials in particulate form and radionuclides other than noble gases. Conservative assumptions may be necessary in establishing setpoints to account 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 radionuclides, and the time lag between alarm / trip action and the final isolation of the radioactive effluent. (Ref.
-11.8.5.) Table 9.2-B provides the instrument surveillance 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 continuous gaseous effluent releases are (Ref. 11.6.8, 11.6.9): Monitor I.D. Description GT-RE-21 Unit Vent GH-RE-10 Radwaste Building Vent Each of the above continuously monitors gaseous radioactivity concentrations downstream of the last point of potential influent, and therefore - measures effluents and not inplant concentrations. l
- h. _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - - - - _ -
APA-ZZ-01003 ! R2v. O The Unit Vent monitor continuously monitors the effluent from the unit vent for. 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 generate airborne activity. 1 The exhaust systems which supply air to the unit vent are from the fuel building, auxiliary buildin'g, the access control area, the containment purge, and the condenser air discharge. The Unit Vent monitor provides alarm functions only, and does not terminate releases from the Unit Vent. The Radwaste Building Ventilation effluent monitor continuously monitors for gaseous radioactivity in the effluent duct downstream of the exhaust filter and fans. The flow path provides ventilation exhaust for all parts of the building structure and components 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, j tanks, and equipment purged by the waste ! processing system. This monitor will isolate the waste gas decay tank discharge line upon a high gaseous radioactivity alarm. i l l 1 i 1 l i - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ d
1 I APA-ZZ-01003 l Rav. O l
,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 I the total dose rate limit (DRL) will be allocated d to each release point. Neglecting the batch. I releases, the plant Unit Vent monitor has been J allocated 0.7 DRL and the Radwaste Building Vent monitor has been allocated 0.3 DRL. These . j allocation factors may be changed as required to ) support plant operational needs, but shall not be allowed to exceed unity (i.e. 1.0). Therefore, a particular monitor reaching the setpoint would not necessarily mean the dose rate limit at the l SITE BOUNDARY is being exceeded; the e,larm only I indicates that the specific release point is contributing a greater fraction of the dose rate limit than was allocated to the associated monitor, and will necessitate an evaluation of both systems. 3.3.2 Batch Release Gaseous Monitors The radiation monitors associated with batch release gaseous effluents are (Ref. 11.6.9, 11.6.10, 11.6.11): Monitor I.D. Description GT-RE-22 Containment Purge System GT-RE-33 GH-RE-10 Radwaste Building Vent The Containment Purge System monitors continuously monitor the containment purge exhaust duct during purge operations for gaseous radioa;tivity. The primary purpose of these j monitors is to isolate the containment purge system on high gaseous activity via the ESFAS. The sample points are located outside the containment between the containment isolation
APA-ZZ-01003 Rsv. O dampers and the containment purge filter adsorber unit. > The Radwaste Building Vent monitor was previously described in Section 3.3.1. Setpoints for the batch gaseous effluent monitors are calculated 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 radionuclides. The alarm / trip setpoint(s) is adjusted accordingly to ensure that the limits of Section 9.6.1.1 are not' exceeded. 3.4 Determination 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. During core alterations, the setpoint for the Containment ~ Purge Monitors, GT-RE-22 and GT-RE-33 is set at a value-of less.than or equal to SE-3 yC1/ce, as required by Technical Specification. ' 4.9.4.2. The actual setpoint value will be l reduced according to the Instrument Loop Uncertainty Estimate (ILUE). This value will also be. utilized in the event that there is no l detectable noble gas activity in the containment atmosphere sample analyzed in accordance with Section 9.6.1.1. The full derivation of this value is discussed in reference 11.14.6. APA-ZZ-01003 [; R:v. 0 1 L 3.4.1 Total Body Dose Rate Setpoint Calculations To ensure that the limits of Section 9.6.1.1 are i. met, the alarm / trip setpoint based on the total l body dose rate is calculated according to: 1 ! Stb < D tbR tb F ,F, (3.1) Where: l Stb = the alarm / trip setpoint based on the total body dose rate (pci/cc) . Dtb = Secti n 9.6.1.1 limit of_500 mrem /yr, conservatively interpreted as a continuous release over a one year period. F,_= the safety factor; a conservative factor-used to compensate for statistical fluctuations and errors of measurement. (For example, F, = 0.5 corresponds to a 100% variation.) Default value is F, = 1.0. F, = the allocation factor which will modify the a required dilution factor such that' simultaneous gaseous releases may be made without exceeding the limits of Section 9.6.1.1. The default value is 1/n, where n is the number of pathways planned for release. 1 1 i 1 i
t APA-ZZ-01003 Rsv. O Rtb = factor used to convert dose rate to the effluent concentration as measured by the effluent monitor, in (pCi/cc) per (mrem /yr) to the total body,. determined according to:
, Rtb =
+
[(X/Q) I K Qg] g (3.2) Where: C= monitor reading of a noble gas monitor corresponding to the sample radionuclides concentrations for the batch to be released. Concentrations are determined 1n accordance with Table 9.6-A. The mixture of radionuclides determined via grab sampling'of the effluent stream or source is correlated to a calibration factor to' determine monitor response. The monitor response is based on concentrations, not release rate, and is in units of (pCi/cc). X/Q = the highest calculated annual average relative concentration for any area at or beyond the SITE BOUNDARY in (sec/m'). Refer to Tables 9, 10, and 12. Kg= the total body dose factor due to gamma emissions for each identified noble gas radionuclides, in (mrem /yr) per (pci/m 8 ). (Table 3) Qg= rate of release of noble gas radionuclides, 1, in (pC1/sec). ; 1 1
APA-ZZ-01003 Rsv. O Qg is' calculated as the product of the ventilucion path design flow rate and the measured activity of the effluent stream as determined by grab sampling. Flow rates for the ventilation pathways can be found in references 11.6.18, 11.6.19, 11.6.20, f and 11.6.21. l 3 . 4 '. 2 Skin Dose Rate Setpoint Calculation To ensure that the limits of Section 9.6.1.1 are met, the alarm / trip setpoint based on the. skin dose rate is calculated according to: 1 S,.5 D,R,F,F, (3.3) Where: F, and F, are as previously. defined in Section 3.4.1.1. - l S, = the alarm / trip setpoint based on the skin q dose rate. D, = Section 9.6.1.1 limit of 3000' mrem /yr, l conservatively interpreted as a continuous release over a one year period. j I i j 4 i i l i i
e APA-ZZ-01003 Rtv. O' g. R, = factor used to convert dose rate'to the effluent concentration as measured by the effluent monitor, in (pci/cc) per (mram/yr) to the skin,~ determined according to: R, = C + [(X/Q).I-(Lg + 1.1M g) Qg] (3.4)
'i Where:
Lg= the skin dose factor due to beta emissions for each identified noble' gas radionuclides, 3 in (mrem /yr) per (pCi/m ). (Table 3) 1.1 = conversion factor: 1 mrad air dose = 1.1 mrem skin dose. Mg= de air dose factor due to gamma emissions for each identified noble gas radionuclides, in (mrad /yr) per (pCi/m ). (Table 3) C, (X/Q) and Q g are as previously defined. 1
- 'r
[: ' 1 . APA-ZZ-01003 Rcv. ' O.. I-
, 3 '. 4. 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 lof the'.two values. t , W
m. 8
- N E
7 NO k N w e e b La bo b
=& 9 Nmm#m#mmmmNmemm 000000000000000 45 (oe CL +++++++++++++++
WWWWWWWWWWWWWWW eW ==w e u -b- eD N e m m @ m m eD m ChC P= m c to &CEh
.W gD e N ED. O.s Ch. O. th. 0 40. *=. M. o. m. e. N. N. N.
sn O v N w * *= N *= b w w e= P= N e= M M ( .O e O b
- E a W w P=
.J P=
CD Ch h m E - 6 % - O =
.k .
9 J b b ' ** m e m a e s N N N m ei m m m .E O -b e 000000000000000 40- CL +++++++++++++++
- W W == WWWWWWWWWWWWWWW e t= eUEa b ED
== ga . m m N N N m ei W P= m @ N * ** O e6 h E= e N Ch. N. N. .ea. th.
. P= @ sh. N. ah. m. th. @. N. r1 .S=
6 Q@ *O t= ** e= @ e= *= c= e= m m M e= e= Ch Dg A m E m e e
- O b 9-W 8 'O E
.J = w a 40
( K a*J Ch O e 9 N . to a m m e. 8 4 L E *= 0 N O e = @ N OO "O
# "L =
W 6 w rimmM#MNNNNmemm 3 E C0000000000000 0
- D D L ++++++++++++++
to e 9 WWWWWWWWWWWWWW
- O O == CL 6 On A= 0I @ M m P= = Ch@ a @ v= @ N m Ch @
K a e X m W m O N P= Ch o w e N == @ E W C b +**** *** * * * * *
===
h 4 *= *= Ch N w P= 4 Os m N e= *= # N E JC N O to E 6 g 6 m E u E w & O >
>= .=
0 a b 4 et 9 6 E D W N.= S to b
~
Ab w NmemMMXwNNmmmeim DO C00000000000000 # Os b o++++++++++++++ b (D D - 9 WWWWWWWWWWWWWWW D e CL - M
.= 6 == W P= *= N P= @ @ &h *= 3 N e= N m # - O e Ma e we b mee=e@ . D.s#. c. ah. ee-em O.w. C.4. c. sD. 6 Om h k w e= sh w w e Os N N m e w e e No %
n O E # l 9 O
== D E
, w e
- l G l e e e t =
= EE E -
- EE *= m m e nh N e .
O m eA4h P= C Ch C M m M m m m m e= @ j-3 C EO C to e sO Ch w *= w e= *= w *= # .C C e ee ei B e e e e ee8 8 e *= 0 bbbbbbbO@@CC##b
= MMMMMMMKXXXXXK4 =
5 e e E w
APA-ZZ-01003 R v. 0 3.4.4 Summary, Gaseous Effluent Monitors Setpoint Determination The gaseous effluent monitors setpoints are f calculated according to equations (3.1) and 1 (3.3), as described in Section 3.4. However, it should be not.ed that a batch release will alter the flow rate characteristics at the Unit Vent 3 and therefore the concentration as sensed by the l 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. ) l 3.5 Calculation of Dose From Gaseous Effluents ] l Dose rate calculations are performed for gaseous I effluents to ensure compliance with Section i 9.6.1.1. I 3.5.1 Calculation of Dose Rate The following methodology is applicable to the location (SITE BOUNDARY or beyond) characterized by ths 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. 11.8.6) The factors K g, L, g and M g relate the radionuclides airborne concentrations to various dose rates, assuming a semi-infinite cloud model, , and are tabulated in Table 3. 3.5.1.1 Noble Gases ) i The release rate lir 't for noble gases is i determined accordice co the following general relationships (Ref. 11.8.6): , i l l 1
APA-ZZ-01003 Rsv. O u D tb'" I IK ((X/Q)Qg)] i $ 500 mrem /yr (3.5). i. D, = I [(L g + 1.1 M g)((Y./Q)Qg)] $ 3000 mrem /yr (3.6) i 7 .. Where: Dtb = Total body dose rate, conservatively averaged over a period of one year. K=g Total body dose. factor due to gamma emissions for each identified noble gas 3 radionuclides, in (mrem /yr) per (pCi/m ), (Table 3)
.(X/Q) = The highest calculated annual average relative concentration for any area at or beyond the SITE BOUNDARY. Refer to Tables 9, 10, and 12.
Qg = The release rate of. noble gas radionuclides, i, in gaseous effluents, from all vent releases in (pci/sec). is calculated as the product of the Q1= ventilation path design flow' rate and the measured activity of the effluent stream as determined by grab sampling. Flow rates for the ventilation pathways can be found in references 11.6.18, 11.6.19, 11.6.20, and 11.6.21. D, = Skin dose rate, conservatively averaged over a period of one year.
/
(-
APA-ZZ-01003 Rev. O Lg= Skin dose factor due to beta emissions for each identified noble gas radionuclides, in (mrem /yr) per (pCi/m8 ) (Table 3). 1.1 = Units conversion factor; 1 mrad air dose = 1.1 mrem skin dose. Mg= Air dose factor due to gamma emissions for each identified noble gas radionuclides, in (mrad /yr) per (pCi/m') (Table 3). 3.5.1.2 Radionuclides Other Than Noble Gases The release rate limit for Iodine-131 and-133, for tritium, and for all radioactive materials in particulate form with half lives greater than 8 days is determined according to (Ref. 11.8.7): D, = I P [(X/Q)Qg] f 5 1500 mrem /yr (3.7) i Where: D, = Dose rate to any critical organ, in (mrem /yr). Pg= Dose parameter for radionuclides other than noble gases for the inhalation pathway for the child, based on the critical organ, in (mrem /yr) per (uCi/m'). (Table 4) Q= 1 The release rate of radionuclides, i, in gaseous effluents, from all vent releases, in (pCi/sec). Q g is calculated as the product of the ventilation path design flow rate and the measured activity of the effluent streim as determined by grab sampling. Flow rates for the ventilation pathways can be found in references 11.6.18, 11.6.19, 11.6.20, and 11.6.21.
APA-ZZ-01003 R v. 0 (X/Q) is as previously defined. The dose parameter (P g ) includes the internal dosimetry of radionucif de, i, and 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 expcsure, separate values of P g are tabulated in Table 4 for the inhalation pathway. These values were calculated according to (Ref. 11.8.8): Pg = K' (BR) DEA g (3.8) Where: K' = Units conversion factor: 1pCi = 1E06 pCi. BR= The breathing rate of the maximum exposed child age group, 3700 m*/yr. (Regulatory Guide 1.109, Table E-5). ! DEAg= The maximum organ inhalation dose factor for the child age group for the ith radionuclides, in (mrem /pC1). The total body is considered as an organ in the selection of DFA g. (Ref. 11.11.5 and 11.14.4) Note: All radioiodines are assumed to be released in elemental form. (Ref.11.8.7) i i l l 1
4 APA-ZZ-01003 Rsv. 0
. TABLE 4 a
DOSE PARAMETER g (P ) FOR RADIOWWDES NR M NOBE GASES Inhalation Pathway (arem/yr) per (pCi/m') l l l l Total l l l l l lNuclidel Bone i Liver l Body l Thyroid l Kidney I Lung l GI-LLI I l H-3 l ND l1.12E3 l1.12E3 l1.12E3 l1.12E3 l1.12E3 l1.12E3 l l Be-7 l8.47E2 l1.44E3 l9.25E2 i ND l ND l6.48E4 l2.55E3 l l C-14 l3.59E4 l6.73E3' l6.73E3 l6.73E3 l6.73E3 l6.73E3 l6.73E3. l l Na-24 l1.61E4 l1.61E4 l1.61E4 l1.61E4 l1.61E4 l1.61E4 l1.61E4 l l P-32 l2.60E6 l1.14ES. l9.88E4 l ND l ND l ND 14.22E4 l l Cr-51 l ND l ND l1.54E2 l8.55El l2.43E1 11.70E4 l1.08E3 l l Mn-54 j ND 14.29E4 l9.51E3 l ND l1.00E4 l1.58E6' l2.29E4 l l Mn-56 l ND l 1.66E0 l3.12E-1 l ND l1.67E0 l1.31E4 l1.23E5 l l Fe-55 l4.74E4 l2.52E4 17.72E3 l ND l ND l1.11E5 l2.87E3 l l'Fe-59 l2.07E4 l3.34E4 l1.67E4 l ND l ND l1.27E6 l7.07E4 l l Co-57 l ND l9.03E2 l1.07E3 l ND l ND l5.07ES l1.32E4 l l Co-58 l ND l1.77E3 l3.16E3 l ND l ND l1.11E6 l3.44E4 l l Co-60 l ND' l1.31E4 l2.26E4 l ND l ND l7.07E6 l9.26E4 l l Ni-63 l8.21E5 l4.63E4 l2.80E4 l ND l ND l2.75E5 l6.33E3 l l Ni-65 12.99E0 l2.96E-1 l1.64E-1 l ND l ND l8.18E3 l8.40E4 l l Cu-64 l ND l1.99E0 l1.07E0 l ND l6.03E0 l9.58E3 13.67E4 l l Zn-65 l4.26E4 l1.13E5 17.03E4 l ND l7.14E4 l9.95E5 l1.63E4 l l Zn-69 l6.70E-2 l9.66E-2 l8.92E-3 l ND l5.85E-2 l1.42E3 l1.02E4 l l Br-82 l ND l ND l?.09E4 l ND l ND l ND l ND l l Br-83 l ND l ND l4.74E2 l ND l ND l ND l 0 l l Br-84 l- ND l ND l5.48E2 l ND l ND l ND l 0 l l Br-85 l ND l ND l2.53E1 l ND l ND l ND l 0 l l Rb-86 l ND l1.98E5 l1.14E5 l ND l ND l ND l7.99E3 l _l Rb-88 l ND l5.62E2 {3.66E2 l ND l ND l ND l1.72E1 l l Rb-89 l 'ND l3.45E2 l2.90E2 l ND l ND l ND l1.89E0 l l Sr-89 l5.99E5 l ND l1.72E4 l ND l ND l2.16E6 l1.67E5 l l Sr-90 l1.01E8 l ND l6.44E6 l ND l ND l1.48E7 l3.43E5 l l Sr-91 l1.21E2 l ND 14.59E0 l ND l ND l5.33E4 l1.74E5 l l Sr-92 l1.31El l ND l5.25E-1 l ND l ND l2.40E4 l2.42E5 l l Y-90 14.11E3 l ND l1.11E2 l ND l ND l2.62E5 l2.68E5 l
APA-ZZ-01003 Rsv. O TABLE 4 (Cont'd.) a DOSE PARAMETER (Pf ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Inhalation Pathway (mrem /yr) per (pCi/m') l l l l Total l l l l l lNuclidel Bone l Liver l Body l Thyroid l Kidney l Lung l GI-LLI l l Y-91m l5.07E-1 l ND l1.84E-2 l ND l ND l2.81E3 l1.72E3 l l Y-91 l9.14E5 l ND l2.44E4 l ND l ND l2.63E6 l1.84E5 l l Y-92 j2.04E1 l ND l5.81E-1 l ND l ND l2.39E4 l2.39E5 l l Y-93 l1.86E2 l ND l5.11E0 l ND l ND l7.44E4 l3.89E5 l l Zr-95 l1.90E5 l4.18E4 l3.70E4 l ND l5.96E4 l2.23E6 l6.11E4 l l Zr-97 l1.88E2 l2.72E1 l1.60E1 l ND l3.89El l1.13E5 l3.51E5 l l Nb-95 l2.33E4 l9.18E3 l6.55E3 l ND l8.62E3 l6.14E5 l3.70E4 l l Mo-99 l ND l1.72E2 l4.26El l ND l3.92E2 11.35E5 l1.27E5- l l Tc-99ml1.78E-3 l3.48E-3 l5.77E-2 l ND l5.07E-2 l9.51E2 l4.81E3 l l Tc-101l8.10E-5 [8.51E-5 l1.08E-3 l ND l1.45E-3 l5.85E2 l1.63E1 l l Ru-103l2.79E3 l ND l1.07E3 l ND l7.03E3 l6.62E5 14.48E4 l l Ru-105l1.53E0 l ND l5.55E-1 l ND l1.34E0 l1.59E4 l9.95E4 l l Ru-106l1.36E5 l ND l1.69E4 l ND l1.84E5 l1.43E7 l4.29E5 l lAg-110ml1.69E4 l1.14E4 l9.14E3 l ND l2.12E4 l5.48E6 l1.00E5 l l Cd-109l ND l5.48E5 l2.59E4 l ND l4.96E5 l1.05E6 l2.78E4 l l Sn-113l1.13E5 l3.12E3 l8.62E3 l2.33E3 l ND l1.46E6 l2.26E5 l l Sb-124l5.74E4 l7.40E2 l2.00E4 l1.26E2 l ND l3.24E6 l1.64E5 l l Sb-123l9.84E4 l7.59E2 l2.07E4 l9.10E1 l ND l2.32E6 l4.03E4 l lTe-12Sml6.73E3 l2.33E3 l9.14E2 l1.92E3 l ND l4.77ES l3.38E4 l lTe-127ml2.49E4 l8.55E3 l3.02E3 l6.07E3 l6.36E4 l1.48E6 l7.14E4 l lTe-127 l2.77E0 l9.51E-1 l6.11E-1 l1.96E0 l7.07E0 l1.00E4 l5.62E4 l lTe-129ml1.92E4 l6.85E3 l3.04E3 l6.33E3 l5.03E4 l1.76E6 l1.82E5 l lTe-129 l9.77E-2 l3.50E-2 l2.38E-2 l7.14E-2 l2.57E-1 l2.93E3 l2.55E4 l lTe-131ml1.34E2 l5.92E1 l5.07El l9.77El l4.00E2 l2.06E5 l3.08E5 l lTe-131 l2.17E-2 l8.44E-3 l6.59E-3 l1.70E-2 l5.88E-2 l2.05E3 l1.33E3 l lTe-132 l4.81E2 l2.72E2 l2.63E2 l3.17E2 l1.77E3 l3.77ES l1.38E5 l lI-130 l8.18E3 l1.64E4 l8.44Ep l1.85E6 l2.45E4 l ND l5.11E3 l lI-131 l4.81E4 l4.81E4 l2.73E4 l1.62E7 l7.88E4 l ND l2.84E3 l lI-132 l2.12E3 l4.07E3 11.88E3 l1.94E5 l6.25E3 l ND l3.20E3 l lI-133 l1.66E4 l2.03E4 l7.70E3 l3.85E6 l3.38E4 l ND l5.48E3 l lI-134 l1.17E3 l2.16E3 l9.95E2 l5.07E4 l3.30E3 l ND l9.55E2 l lI-135 l4.92E3 l8.73E3 14.14E3 l7.92E5 l1.34L4 l ND l4.44E3 l
1 APA-ZZ-01003 I Rev. O i i
.i TABLE 4 (Cont'd.) ]
DOSE PARAMETER (Pg ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES" Inhalation Pathway 1 (mrem /yr) per (liC1/m') l l l l Total l l l l l lNuclidel Bone l Liver l Body l Thyroid 1 Kidney l Lung l GI-LLI l lCs-134 l6.51ES l1.01E6 l2.25E5 l ND l3.03E5 l1.21E5 13.85E3 l ! lCs-136 l6.51E4 l1.71ES l1.16ES l ND l9.55E4 l1.45E4 l4.18E3 l lCs-137 l9.07E5 l8.25ES l1.28E5 l ND l2.72E5 l1.04E5 l3.62E3 l lCs-138 l6.33E2 l8.40E2 l5.55E2 l ND l6.22E2 l6.81El l2.70E2 l lBa-139 l1.84E0 l9.84E-4 l5.37E-2 i ND l8.62E-4 l5.77E3 l5.77E4 l lBa-140 l7.40E4 l6.48E1 14.33E3 l ND l2.11El l1.74E6 l1.02E5 l lBa-141 l2.19E-1 l1.09E-4 l6.36E-3 l ND l9.47E-5 l2.92E3 l2.75E2 l lBa-142 l5.00E-2 l3.60E-5 l2.79E-3 l ND l2.91E-5 l1.64E3 l2.74E0 l lLa-140 l6.44E2 l2.25E2 l7.55El l ND l ND l1.83E5 l2.26E5 l lLa-142 l1.30E0 l4.11E-1 l1.29E-1 l ND l ND l8.70E3 l7.59E4 l lCe-141 l3.92E4 l1.95E4 l2.90E3 l ND 18.55E3 l5.44E5 l5.66E4 l lCe-143 l3.66E2 l1.99E2 l2.87El l ND 18.36El l1.15E5 l1.27E5 l lCe-144 l6.77E6 l2.12E6 l3.61E5 l ND l1.17EG l1.20E7 l3.89E5 l lPr-143 l1.85E4 l5.55E3 l9.14E2 l ND l3.00E3 l4.33E5 l9.73E4 l lPr-144 l5.96E-2 l1.85E-2 l3.00E-3 l ND l9.77E-3 l1.57E3 l1.97E2 l lNd-147 l1.08E4 l8.73E3 l6.81E2 l hT l4.81E3 l3.28E5 l8.21E4 l lEu-154 l1.01E7 19.21E5 l8.40E5 l ND l4.03E6 l6.14E6 l1.10E5 l lHf-181 l2.78E4 l1.01E5 l1.25E4 l ND l2.05E4 l1.06E6 l6.62E4 l lW-187 l1.63E1 l9.66E0 l4.33E0 l ND l ND l4.11E4 l9.10E4 l lNp-239 l4.66E2 l3.34E1 l2.35El l ND l9.73E1 :l5.81E4 l6.40E4 l (a) The child age group; refer to reference 11.14.5.
APA-ZZ-01003 Rt.v. O 3.5.2 Dose Due To Gaseous Effluents .41858 3.5.2.1 REC Section 9.7.1.1 4160 The air dose due to noble gases released in. gaseous effluente, to area at and beyond the SITE BOUNDARY shall be limited to the following:
- a. During.any'calondar 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.
l 1
APA-ZZ-01003 Rev. O I 3.5.2.1.1 Noble Cases
)
The air dose at the SITE BOUNDARY due to noble gases released from the site is calculated according to the following methodology (Ref. 11.8.9): During any calendar quarter, for gamma radiation: l Dg = 3.17E-08 I (Mg {(X/Q) Qg + (X/q) q1)] $5 mrad (3.9) i ; i During any calendar quarter, for beta radiation: Db = 3.17E-08 I (Ng ((X/Q) Q 1 + (X/q) q g}] 5 10 mrad (3.10) i During any calendar year, for gamma radiation: Dg = 3.17E-08 I (My {(X/Q) Qg + (X/q) qf}] 5 10 mrad (3.11) i During any calendar year, for beta radiation: Db = 3.17E-08 I (Ng {(X/Q) Qg + (X/q) q1)] 5 20 mrad (3.12) i l
APA-ZZ-01003 R;v. O Where: Dg = Air dose from gamma radiation due to noble 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). Refer to Tables 9, 10, 11, and 12. qg= The average release'of noble gas radionuclides, 1, in gaseous effluents from all vent releases for short-term releases (equal to or less than 500 hrs / year), in (pC1). Releases are cumulative over the calendar quarter or year, as appropriate. Ng= The air dose factor due to beta emissions for each identified noble gas radionuclides, 1, in (mrad /yr) per (pCi/m'). (Table 3) Qg= The average release of noble gas radionuclides, i, in gaseous effluents from all vent releases for long-term releases (greater than 500 hrs / year), in (vC1). I Releases are cumulative over the calendar quarter or year, as appropriate. (X/Q) = The highest calculated annual average relative concentration for areas at or beyond the I SITE BOUNDARY for long-term releases (greater than 500 hrs /yr). Refer to Tables 9, 10, and 12. 3.17E-08 = The inverse of the number of seconds per year. M g is as previously defined. (Refer to Section 3.4.1.2) .
APA-ZZ-01003 l Rcv. 0 l l I I i 41860 3.5.2.2 REC Section 9.8.1.1 l 4160 The dose to a MEMBER OF THE PUBLIC from 4 Iodine-131 and 133, tritium, and all j radionuclides in particulate form with half-lives i greater than 8 days in gaseous effluents ! released, to areas at and beyond the SITE j BOUNDARY shall be limited to the following (Ref. j 11.8.9): l l
- a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ and, l
- b. During any calendar year: Less than or 1 equal to 15 mrem to any organ.
3.5.2.2.1 Radionuclides Other Than Noble Gases The dose to a MEMBER OF THE PUBLIC from Iodine-131 and 133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, to areas at and beyond the SITE BOUNDARY, is calculated according to the following expressions: During any calendar quarter: D1 = 3.17E-08 I Rg [W Q 1 + w gg) 5 7.5 mrem (3.13) i During any calendar year: D1 = 3.17E-08 I Rg [W Q1 + w qf] 5 15 mrem (3.14) i Where: D1= Dose to a MEMBER OF THE PUBLIC from radionuclides other than noble gases.
APA-ZZ-01003' Rsv. O
<Qg . = ' The releases'of radiciodines, radioactive
, materials in particulate form, and
- radionuclides other than noble gases, i, in
. gaseous effluents, for all long-term vent
. releases (greater than 500. hrs /yr),-in
-(pC1). Releases are cumulative over the calendar quarter-or year as appropriate.
qg= ~ The releases of radiciodines, radioactive
-materials in particulate form and radionuclides odher than noble gases, i, in '
. gaseous' effluents for all short-term vent releases (equal to or:less than 500 hrs /yr), in (pCi). Releases are' cumulative over the calendar quarter or year.as appropriate.
Rg= The dose-factor for each' identified , radionuclides, i, in m*(mrem /yr) per
. (pci/sec) or (mrem /yr) per (pci/m 8). (Table.
5) W= The dispersion parameter for estimating the dose to an individual at the controlling location for.long-term releases (greater than 500 hrs /yr): W =: (X/Q) for the inhalation and tritium pathways, in(sec/m ) . 8 W = (D/Q) for the food and ground plane pathways, in(meters"8). Refer to Tables'9, 10, and 12. APA-ZZ-01003. Rsv. O i w= The dispersion parameter for estimating the dose to an individual at the controlling location for short-term releases (equal to or less than 500 hrs /yr): w= (X/q) for the inhalation pathway, i in(sec/m )8 w= (D/q) for the food and ground plane pathway, in (meters ~8). Refer to Tables 9, i 10, 11, and 12. 3.17 E-08 = The inverse of the number ' of seconds per year. (D/Q) = the average relative deposition of the effluent at or beyond the SITE BOUNDARY, considering depletion of the plume during transport, for long term releases
~
(greater than 500 hrs /yr), in (meters 8). (D/q) = the relative deposition of the effluent at or beyond the SITE BOUNDARY, considering depletion of the plume during transport, for short term releases (less than or equal to 500 hrs /yr), in (meters ~2). i l 1 _ _ _ _ _ _ i
if s .
, APA-ZZ-01003 Rav. O c
. Note: For the direction sectors with existing pathways within 5 miles from,the site, the appropriate R g values are used. If no real pathway exists within 5 miles from the center of the building complex, the cow-milk R value is used, and it is assumed 1
that this pathway exists at the 4.5 to 5.0 mile distance in the limiting-case sector. If the R g for an existing pathway within 5 miles is less than a cow-milk Rf at 4.5 to 5.0 miles, then the value of the cow-milk R g at 4.5 to 5.0 mlles is used. (Rev. 9.8.10.) Although the annual average relative concentration (X/Q) and the average relative deposition rate (D/Q) are genarally considered to be at the approximate receptor location in lieu of the SITE BOUNDARY for these calculations, it is acceptable to consider the ingestion,
-inhalation, and ground plane pathways to coexist at the location of the nearest residence with the highest value of (X/Q). (Ref. 11.8.9) The Total Body dose from ground plane deposition is added to the dose for each individual organ. (Ref.
11.11.3) The cumulative critical organ doses for a monthly, quarterly or annual evaluation are based on the calculated dose contribution from each specified time period occurring during the reporting period.
~_____ -
e
- l I
APA-ZZ-01003 Rt.v. O TABLE 5 PATHWAY DOSE FACTORS g(R ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES" I Inhalation Pathway (mrem /yr) per (11Ci/m8 ) l l l l Total l l l l l lNuclidel Bone l Liver 1 Body l Thyroid i Kidney l Lung l GI-LLI l l H-3 l ND l1.12E3 l1.12E3 l1.12E3 11.12E3 l1.12E3 l1.12E3 l l Be-7 l8.47E2 l1.44E3 l9.25E2 l ND l ND l6.48E4 l2.55E3 l l C-14 l3.59E4 l6.73E3 l6.73E3 l6.73E3 l6.73E3 l6.73E3 l6.73E3 l l Na-24 l1.61E4 l1.61E4 l1.61E4 l1.61E4 l1.61E4 l1.61E4 l1.61E4 l l P-32 l2.60E6 l1.14E5 l9.88E4 l ND l ND l ND l4.22E4 l l Cr-51 l ND l ND l1.54E2 l8.55El l2.43E1 l1.70E4 l1.08E3 l l Mn-54 l ND l4.29E4 l9.51E3 l ND l1.00E4 l1.58E6 l2.29E4 l l Mn-56 l ND l1.66E0 l3.12E-1 l ND l1.67E0 l1.31E4 l1.23E5 l l Fe-55 l4.74E4 l2.52E4 l7.72E3 l ND l ND l1.11E5 l2.87E3 l l Fe-59 l2.07E4 l3.34E4 l1.67E4 l ND l ND l1.27E6 l7.07E4 l l Co-57 l ND l9.03E2 l1.07E3 l ND l ND l5.07E5 l1.32E4 l l Co-58 l ND l1.77E3 l3.16E3 l ND l ND l1.11E6 l3.44E4 l l Co-60 l ND l1.31E4 l2.26E4 l ND l ND l7.07E6 l9.26E4 l l Ni-63 l8.21ES l4.63E4 l2.80E4 l ND l ND l2.75E5 16.33E3 l l Ni-65 l2.99E0 l2.96E-1 l1.64E-1 l ND l ND l8.18E3 l8.40E4 l l Cu-64 l ND l1.99E0 l1.07E0 l ND l6.03E0 19.58E3 l3.67E4 l l Zn-65 l4.26E4 l1.13E5 l7.03E4 l ND l7.14E4 19.95E5 l1.63E4 l 1 Zn-69 l6.70E-2 l9.66E-2 l8.92E-3 l ND l5.85E-2 11.42E3 11.02E4 l l Br-82 l ND l ND l2.09E4 l ND l ND l ND l ND l l Br-83 l ND l ND l4.74E2 l ND l ND l ND l 0 l l Br-84 l ND l ND l5.48E2 l ND l ND l ND l 0 l l Br-85 l .ND l ND l2.53E1 l ND l ND l ND l 0 l [ Rb-86 l ND l1.98E5 l1.14E5 l ND l ND l ND [7.99E3 l l Rb-88 l ND l5.62E2 l3.66E2 l ND l ND l ND l1.72E1 l l Rb-89 l ND 13.45E2 l2.90E2 l ND l ND l ND l1.89E0 l l Sr-89 l5.99ES l ND l1.72E4 l ND l ND l2.16E6 l1.67E5 l l Sr-90 l1.01E8 l ND l6.44E6 l ND l ND l1.48E7 l3.43E5 l l Sr-91 l1.21E2 l ND l4.59E0 l ND l ND l5.33E4 l1.74E5 l l Sr-92 l1.31El l ND [5.25E-1 l ND l ND l2.40E4 l2.42E5 l l Y-90 l4.11E3 l ND l1.11E2 l ND l ND l2.62E5 l2.68E5 l _t - _ - - - - - - - . - _ - - - - - _ - - - . - - _ - - -
APA-ZZ-01003 Rsv. O TABLE 5 (Cont'd. ) PATHWAY DOSE FACTORS (R g ) FOR RADIONUCLIDES OTHER THAN NOSLE GASES" Inhalation Pathway (mrem /yr) per (vCi/m8 ) l l l l Total l l l l l lNuclidel Bone l Liver l Body l Thyroid i Kidney l Lung l GI-LLI i l Y-91m l5.07E-1 l ND l1.84E-2 l ND l ND 12.81E3 l1.72E3 l J l Y-91 l9.14E5 l ND l2.44E4 l ND l ND j2.63E6 l1.84E5 l ; l l Y-92 l2.04E1 l ND l5.81E-1 l ND l ND l2.39E4 l2.39E5 l l Y-93 l1.86E2 l ND l5.11E0 l ND l ND l7.44E4 l3.89E5 l l Zr-95 l1.90E5 'l4.18E4 l3.70E4 l ND l5.96E4 l2.23E6 l6.11E4 l l Zr-97 l1.88E2 l2.72E1 l1.60E1 l ND l3.89El l1.13E5 l3.51E5 l l Nb-95 l2.33E4 l9.18E3 l6.55E3 l ND 18.62E3 l6.14E5 l 3. 70E4- l l Mo-99 l ND l1.72E2 l4.26El l ND l3.92E2 l1.35E5 l1.27E5 l l Tc-99 mil.78E-3 l3.48E-3 l5.77E-2 l KD l5.07E-2 l9.51E2 l4.81E3 l l Tc-101l8.10E-5 l8.51E-5 l1.08E-3 l ND l1.45E-3 l5.85E2 11.63E1 l l Ru-103l2.79E3 l ND l1.07E3 l ND l7.03E3 l6.62E5 14.48E4 l l Ru-105l1.53E0 l ND l5.55E-1 l ND l1.34E0 l1.59E4 19.95E4 l l Ru-106l1.36ES l ND l1.69E4 l ND l1.84E5 l1.43E7 l4.29E5 l lAg-110ml1.69E4 l1.14E4 l9.14E3 l ND l2.12E4 l5.48E6 l1.00E5 l l Cd-109l ND l5.48E5 l2.59E4 l ND l4.96E5 l1.05E6 l2.78E4 l l Sn-113l1.13E5 l3.12E3 l8.62E3 l2.33E3 l ND l1.46E6 l2.26E5 l l Sb-124l5.74E4 l7.40E2 l2.00E4 l1.26E2 l ND l3.24E6 l 1.'64E5 l l Sb-125l9.84E4 l7.59E2 12.07E4 l9.10E1 l ND l2.32E6 l4.03E4 l lTe-125ml6.73E3 l2.33E3 l9.14E2 l1.92E3 l ND l4.77ES l3.38E4 l lTe-127ml2.49E4 l8.55E3 l3.02E3 l6.07E3 16.36E4 l1.48E6 17.14E4 l , lTe-127 l2.77E0 l9.51E-1 l6.11E-1 l1.96E0 l7.07E0 l1.00E4 l5.62E4 l l lTe-129ml1.92E4 l6.85E3 l3.04E3 16.33E3 l5.03E4 l1.76E6 l1.82E5 l , lTe-129 l9.77E-2 l3.50E-2 l2.38E-2 l7.14E-2 l2.57E-1 l2.93E3 12.55E4 l J lTe-131ml1.34E2 l5.92E1 l5.07El l9.77El l4.00E2 12.06E5 l3.08E5 l 1 lTe-131 l2.17E-2 l8.44E-3 l6.59E-3 l1.70E-2 l5.88E-2 l2.05E3 l1.33E3 l I lTe-132 l4.81E2 l2.72E2 l2.63E2 l3.17E2 l1.77E3 l3.77E5 l1.38E5 l lI-130 l8.18E3 l1.64E4 l8.44E3 l1.85E6 l2.45E4 l ND 15.11E3 l lI-131 l4.81E4 l4.81E4 l2.73E4 l1.62E7 [7.88E4 l ND l2.84E3 l l lI-132 l2.12E3 l4.07E3 l1.88E3 11.94E5 16.25E3 l ND l3.20E3 l 1 lI-133 l1.66E4 l2.03E4 l7.70E3 l3.85E6 13.38E4 l ND l5.48E3 l lI-134 11.17E3 l2.16E3 l9.95E2 l5.07E4 l3.30E3 l ND 19.55E2 l I i
APA-ZZ-01003 ) Rsv. O I TABLE 5 (Cont'd.) PATHWAY DOSE FACTORS (R ),FOR RADIONUCLIDES OTHER THAN NOBLE GASES" f Inhalation Pathway (mrem /yr) per (11Ci/m8 ) l l l l Total l l l l l lNuclide1 Bone l Liver l Body l Thyroid l Kidney l- Lung' l GI-LLI l
.lI-135 l4.92E3 l8.73E3 l4.14E3 l7.92E5 l1.34E4 l' ND l4.44E3 l lCs-134 l6.51E5 l1,01E6 l2.25E5 l ND l3.03E5 l1.21ES l3.85E3 l lCs-136 l6.51E4 l1.71E5 l1.16ES l ND l9.55E4 l1.45E4 l4.18E3 l
-lCs-137 l9.07E5 l8.25ES l1.28E5 l ND l2.72E5 l1.04E5 l3.62E3 l
-]Cs-138 l6.33E2 l8.40E2 l5.55E2 l ND l6.22E2 l6.81El l2.70E2 l I
lBa-139 l1.84E0 l9.84E-4 ~l5.37E-2 l ND l8.62E-4 l5.77E3 l5.77E4- l lBa-140 l7.40E4 l6.48E1 l4.33E3 l ND l2.11El l1.74E6 l1.02E5 l lBa-141 l2.19E-1 l1.09E-4 l6.36E-3 l ND l9.47E-5 l2.92E3 l2.75E2 l lBa-142-l5.00E-2 l3.60E-5 l2.79E-3 l ND l2.91E-5 l1.64E3 l2.74E0- l
~lLa-140 l6.44E2 l2.25E2 l7.55El l ND l ND l1.83E5 l2.26E5 l lLa-142 l1.30E0 l4.11E-1 l1.29E-1 l ND l ND l8.70E3 l7.59E4 l lCe-141 l3.92E4 l1.95E4 l2.90E3 l ND l8.55E3 l5.44E5. l5.66E4 l lCe-143 l3.66E2 -l1.99E2 l2.87El l ND l8.36El l1.15ES l1.27E5 l lCe-144 l6.77E6 l2.12E6 l3.61E5 l ND l1.17E6 l1.20E7 l3.89E5 l lPr-143 l1.85E4 l5.55E3 l9.14E2 l ND l3.00E3 l4.33E5 l9.73E4 l lPr-144 l5.96E-2 l1.85E-2 l3.00E-3 l ND l9.77E-3 l1.57E3 l1.97E2 l lNd-147 l1.08E4 l8.73E3 l6.81E2 l ND l4.81E3 l3.28E5 18.21E4 l lEu-154 l1.01E7 l9.21ES l8.40E5 l ND l4.03E6 l6.14E6 l1.10E5 l lHf-181 l2.78E4 11.01ES l1.25E4 l ND l2.05E4 l1.06E6 l6.62E4 l lW-187 l1.63E1 l9.66E0 l4.33E0 l ND l ND l4.11E4 l9.10E4 l lNp-239 l4.66E2 l3.34E1 l2.35El l ND [9.73E1 l 5. 81E4 l6.40E4 l (a) The child age group; refer to reference 11.14.5.
APA-ZZ-01003 l R:v. O l TABII 5 (Cont'd.) PATWAY DOSE FACTORS (Rg) FOR RADIONUCLIDES OTHER THAN NO3LE GASES" Ground Plane Pathway (m2 mrem /yr) per (pCI/sec). Nuclide Total Body Skin Be-7 2.24E7 3.21E7 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-57 2.98E8 4.37E8 Co-58 3.79E8 4.44E8 Co-60 2.15E10 2.53E10 Ni-65 2.97ES 3.45E5 Cu-64 6.07ES 6.88E5 Zn-65 7.47E8 8.59E8 Br-82 3.14E7 4.49E7 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-99 3.98E6 4.62E6
.o ...
APA-ZZ-01003 Rsv. O TABLE 5 (Cont'd.) PATHWAY. DOSE FACTORSg (R ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES" Ground Plane Pathway (m8 mrem /yr) per (pC1/sec) Nuclide Total Body Skin Tc-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 Cd-109 3.76E7 1.54E8 Sn-113 1.43E7 4.09E7 Sb-124 8.74E8 1.23E9 Sb-125 3.57E9 5.19E9 Te-125m 1.55E6 2.13E6 Te-127m 9.16E4 1.08E5 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 3.45E4 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 I-134 4.47E5 5.30E5 I-135 2.51E6 2.9?E6 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 1.06ES 1.19E5 Ba-140- 2.05E7 2.35E7
l., l,
- I, 2 APA-ZZ-01003 Rev. 0 ]
---TABLE 5 (Cont'd.F j PATHWAY DOSE FACTORS (R g ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES"'
Ground Plane Pathway (m* mrcm/yr) per (pCi/sec) i L Nuclide Total Body Skin 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 Pr-144 1.84E3 '2.11E3 Nd-147 8.41E6' 1.01E7 En-154 2.21E10 3.15E10 Hf-181 1.97E8 2.82E8 W-187 2.36E6- 2.74E6 Np-239 1.71E6 1.98E6 (a) Refer to reference 11.14.5 for calculational details.
APA-ZZ-01003 Rsv. O TABLE 5 (Cont'd.) PATHWAY DOSE FACTORS (R g ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES" Meat Pathway (m* mree/yr) per (pCi/sec) l l l l Total l l l l l lN_uclidel Bone i Liver 1 Body l Thyroid l Kidney l Lung l GI-LLI l l H-3 l ND l2.34E2 l2.34E2 l2.34E2 l2.34E2 l2.34E2 l2.34E2 l l Be-7 l7.37E3 l1.26E4 l8.06E3 l ND l1.23E4 l ND l7.00E5 l l C-14 l3.83E8 l7.67E7 l7.67E7 l7.67E7 l7.67E7 l7.67E7 l7.67E7 l l Na-24 l1.78E-3 l1.78E-3 l1.78E-3 l1.78E-3 l1.78E-3 l1.78E-3 l1.78E-3 l l P-32 l7.41E9 l3.47E8 l2.86E8 l ND l ND l ND l2.05E8 l l Cr-51 l ND l ND l8.79E3 l4.88E3 l1.33E3 l8.91E3 j4.66ES l l Ma-54 l ND l8.01E6 l2.13E6 l ND l2.25E6 l ND l6.72E6 l l Mn-56 l ND l 0 l 0 l ND l 0 l ND l 0 l l Fe-55 l4.57E8 l2.42E8 l7.51E7 l ND l ND l1.37E8 l4.49E7 l l Fe-59 l3.76E8 l6.09E8 l3.03E8 l ND l ND l1.76E8 l6.34E8 l l Co-57 l ND 15.92E6 l1.20E7 l ND l ND l ND l4.85E7 l l Co-58 l ND l1.64E7 l5.02E7 l ND l ND l ND l9.58E7 l l Co-60 .l ND l6.93E7 l2.04E8 l ND l ND l ND l3.84E8 l l Ni-63 12.91E10 l1.56E9 l9.91E8 l ND l ND l ND l1.05E8 l l Ni-65 l 0 l 0 l 0 l ND l ND l ND l 0 l l l Cu-64 l ND l2.97E-7 l1.79E-7 l ND l7.17E-7 l ND l1.39E-5 l l Zn-65 l3.75E8 l1.00E9 l6.22E8 l ND l6.30E8 l ND l1.76E8 l l Zn-69 l 0 l 0 l 0 l ND l 0 l ND l 0 l l Br-82 l ND l ND l1.52E3 l ND l ND l ND l ND l l Br-83 l ND l ND l ND l ND l ND l ND l ND l l Br-84 l ND l ND l ND l ND l ND l ND l ND l l Br-85 l ND l ND l ND l ND l ND l ND l ND l l Rb-86 l ND 15.82E8 l3.58E8 l ND l ND l ND [3.74E7 l l Rb-88 l ND l 0 l 0 l ND l ND l ND l 0 l l Rb-89 l ND l 0 l 0 l ND l ND l ND l 0 l j 1 l Sr-89 l4.82E8 l ND l1.38E7 l ND l ND l ND l1.86E7 l l l Sr-90 l1.04E10 l ND l2.64E9 l ND l ND l ND l1.40E8 l l Sr-91 l2.40E-10l ND l 0 l ND l ND l ND l5.29E-10l l Sr-92 l 0 l ND l 0 l ND l ND l ND l 0 l l Y-90 l1.71E2 l ND l4.59E0 l ND l ND l ND l4.88E5 l 1 l
e APA-ZZ-01003 Rw. O TABLE 5 (Cont'd.) PATINAY DOSE FACTORS (Rg) FOR RADIONUCLIDES OTHER THAN NOBLE GASES" Meat Pathway (m8 mrem /yr) per (pC1/sec) l l l l Total l l l l l lNucifdel Bone I Liver 1 Body l Thyroid i Kidney l Lung l GI-LLI J, l Y-91m l 0 l ND l 0 l ND l ND l ND l 0 1 l Y-91 l1.80E6 l ND l4.82E4 l ND l ND l ND l2.40E8 l l Y-92 l 0 l ND l 0 l ND l ND l ND l 0 l l Y-93 l 0 l ND l 0 l ND l ND l ND l1.55E-7 l l Zr-95 l2.66E6 l5.85E5 l5.21E5 l ND l8.38E5 l ND l6.11E8 l l Zr-97 l3.20E-5 l4.63E-6 l2.73E-6 l ND l6.65E-6 l ND l7.02E-1 l l Nb-95 l3.09E6 l1.20E6 l8.61E5 l ND l1.13E6 l ND l2.23E9 l l Mo-99 l ND l1.15E5 l2.84E4 l ND l2.46ES l ND l9.51E4 l l Tc-99ml 0 l 0 l 0 l ND l 0 l 0 l 0 l l Tc-101l 0 l 0 l 0 l ND l 0 l 0 l 0 l l Ru-10311.55ES l ND l5.96E7 l ND l3.90E8 l ND l4.01E9 l l Ru-105l 0 l ND l 0 l ND l 0 l ND l 0 l l Ru-106l4.44E9 l ND l5.54E8 l ND 15.99E9 l ND l6.90E10 l lAg-110ml8.40E6 l5.67E6 l4.53E6 l ND l1.06E7 l ND l6.75E8 l lCd-109 l ND l1.90E6 18.83E4 l ND l1.70E6 l ND l6.18E6 l lSn-113 l2.18E9 l4.48E7 l1.24E8 l3.31E9 l ND l ND l1.54E9 l lSb-124 l2.93E7 l3.79E5 l1.02E7 l6.45E4 l ND l1.62E7 l1.83E8 l lSb-125 l2.85E7 l2.20E5 l5.97E6 l2.64E4 l ND l 1.59E7 l6.80E7 l lTe-125ml5.69E8 l1.54E8 17.59E7 l 1. 60s.o l ND l ND l5.49E8 l lTe-127ml1.77E9 l4.78E8 l2.11E8 l4.24E8 l5.06E9 l ND l1.44E9 l lTe-127 l4.11E-10l1.11E-10 l 0 l2.85E-10 l1.17E-9 l ND l1.61E-8 l lTe-129ml1.79E9 l4.99E8 l2.77E8 l5.76E8 l5.25E9 l ND l2.18E9 l lTe-129 l 0 l 0 l 0 l 0 l 0 l ND l 0 l lTe-131ml7.00E2 l2.42E2 l2.58E2 l4.98E2 l2.34E3 l ND l9.82E3 l lTe-131 l 0 l 0 l 0 l 0 l 0 l ND l 0 l lTe-132 l2.09E6 l9.26E5 l1.12E6 l1.35E6 18.60E6 l ND l9.33E6 l lI-130 l3.04E-6 l6.13E-6 l3.16E-6 l6.76E-4 l9.17E-6 l ND l2.87E-6 l lI-131 l1.66E7 l1.66E7 l9.46E6 l5.50E9 l2.73E7 l ND ll.48E6 l lI-132 l 0 l 0 l 0 l 0 l 0 l ND l 0 l lI-133 16.16E-1 l7.61E-1 l2.88E-1 l1.41E2 l1.27E0 l ND l3.07E-1 l lI-134 l 0 l 0 l 0 l 0 l 0 l ND l 0 l l 1 1
-- )
j APA-ZZ-01003 I Rsv. O i i TABLE 5 (Cont'd. ) PATHWAY DOSE FACTORS g (R ) FOR MOWMES MR N NOBE GASES" Meat Pathway (m8 mrem /yr) per (pCi/sec) l l l l l Total l l l l l f lNuclidel Bone l Liver l Body l Thyroid i Kidney l Lung l GI-LLI l ! lI-135 l 0 l 0 l 0 l 0 l 0 l ND l 0 l lCs-134 l9.22E8 l1.51E9 l3.19E8 l ND l4.69E8 l1.68E8 l8.16E6 l lCs-136 l1.61E7 l4.43E7 l2.86E7 l ND l2.36E7 l3.51E6 l1.56E6 l lCs-137 l1.33E9 l1.28E9 l1.88E8 l ND l4.16ES l1.50E8 l7.99E6 l k lCo-138 l 0 l. 0 l 0 l ND l 0 l 0 l 0 l l lBa-139 l 0 l 0 l 0 l ND l 0 ] O l 0 l j lBa-140 l4.38E7 l3.84E4 l2.56E6 l ND l1.25E4 l2.29E4 l2.22E7 l lBa-141 l 0 l 0 l 0 l ND l 0 l 0 l 0 l lBa-142 l 0 l 0 l 0 l ND l 0 l 0 l 0 l lLa-140 l5.69E-2 l1.99E-2 l6.70E-3 l ND l ND l ND l5.54E2 l lLa-142 l 0 l 0 l 0 l ND l ND l ND l 0 l lCo-141 l2.22E4- l1.11E4 l1.64E3 l ND l4.85E3 l ND l1.38E7 l lCe-143 l3.17E-2 l1.72E1 l2.49E-3 l ND l7.21E-3 l ND l2.52E2 l ; lCe-144 l2.32E6 l7.26E5 l1.24E5 l ND l4.02E5 l ND l1.89E8 l 1 lPr-143 l3.35E4 l1.00E4 l1.66E3 l ND l5.44E3 l ND l3.61E7 l lPr-144 l 0 l 0 l 0 l ND l 0 l ND l 0 l lNd-147 l1.17E4 l9.50E3 17.35E2 l ND l5.21E3 l ND l1.50E7 l ) lEu-154 l1.12E7 l1.01E6 l9.20E5 l ND l4.42E6 l ND l2.34E8 l 1 ' lHf-181 l4.76E6 l1.73E7 l2.15E6 l ND l3.52E6 l ND l6.40E9 l lW-187 l3.35E-2 l1.98E-2 18.91E-3 l ND l ND l ND l2.79E0 l lNp-239 l4.20E-1 l3.02E-2 l2.12E-2 l ND 18.72E-2 l ND l2.23E3 l (a) The child age group; refer to reference 11.14.5. I
APA-ZZ-01003 i Rsv. O s > TABLE 5 (Cont'd.) a PATHWAY DOSE FACTORS (Rf ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Grass-Cow-Milk Pathway 1 (m8 mrem /yr) per (pCi/sec) l l l l Total l l l l l lNuclidel Bone I Liver l Body l Thyroid l Kidney 1 Lung l GI-LLI l l l.H-3 l ND l1.57E3 l1.57E3 l1.57E3 l1.57E3 l1.57E3 l1.57E3 l L l Be-7 l7.49E3 l1.28E4 l8.19E3 l ND l1.25E4 l ND l7.11E5 l 4 l l C-14 l1.19E9 l2.39E8 l2.39E8 l2.39E8 12.39E8 l2.39E8 l2.39E8 l~ l Na-24 l8.89E6 l8.89E6 l8.89E6 l8.89E6 18.89E6 l8.89E6 18.89E6 l l P-32 l7.77E10 l3.64E9 l3.00E9 l ND l ND l ND l2.15E9 l l Cr-51 l ND l ND l1.03E5 l5.65E4 l1.56E4 l1.04E5 l5.40E6 l l Mn-54 l ND l2.10E7 15.59E6 l ND l5.88E6 l ND l1.76E7 l l Mn-56 l ND l1.29E-2 l2.90E-3 l ND l1.56E-2 l ND l1.86E0 l l Fe-55 l1.12E8 l5.93E7 l1.84E7 l ND l ND l3.35E7 l1.10E7 l l Fe-59 l1.20E8 l1.94E8 l9.69E7 l ND l ND l5.64E7 [2.02E8 l l Co-57 l ND l3.84E6 l7.76E6 l ND l ND l ND l3.15E7 l l Co-58 l ND l1.21E7 l3.71E7 l ND -l ND l ND l7.07E7 l l Co-60 l ND l4.32E7 l1.27E8 l ND l ND l ND l2.39E8 l l Ni-63 l2.96E10 l1.59E9 l1.01E9 l ND l ND l ND l1.07E8 l l Ni-65 l1.66E0 l1.56E-1 l9.01E-2 l ND l ND l ND {1.91El l l Cu-64 l ND l7.46E4 l4.51E4 l ND l1.80E5 l ND l3.50E6 l l Zn-65 l4.13E9 l1.10E10 l6.85E9 l ND 16.94E9 l ND l1.93E9 l l Zn-69 l 0 l 0 l 0 l ND l 0 l ND l1.12E-9 l l Br-82 l ND l ND l1.15E8 l ND l ND l ND l ND l l Br-83 l ND l ND l ND l ND l ND l ND l ND l ; l Br-84 l ND l ND l ND l ND l ND l ND l ND l l Br-85 l ND l ND l ND l ND l ND l ND l ND l l Rb-86 l ND l8.80E9 l5.41E9 l ND l ND l ND l5.66E8 l l Rb-88 l ND l 0 l 0 l ND l ND l ND l 0 l l Rb-89 l ND l 0 l 0 l ND l ND l ND l 0 l l Sr-89 l6.62E9 l ND l1.89E8 l ND l ND l ND l2.56E8 l l Sr-90 l1.12E11 l ND l2.83E10 l ND l ND l ND l1.51E9 l l Sr-91 l1.30E5 l ND l4.92E3 l ND l ND l ND l2.88E5 l l Sr-92 l2.18E0 l ND l8.75E-2 l ND l ND l ND l4.13E1 l l Y-90 l3.22E2 l ND [8.62E0 l ND l ND l ND l9.17E5 l
l l APA-ZZ-01003 { R:v. O j TABLE 5 (Cont'd.) l l a PATHVAY DOSE FACTORS (Rg) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Grass-Cow-Milk Pathway j l (m8 mrem /yr) per (11Ci/sec) l l l l Total l l l l l lNuclidel Bone l Liver i Body l Thyroid l Kidney i Lung l GI-LLI l l l Y-91m l 0 l ND l 0 l ND l ND l ND l 0 l l Y-91 l3.90E4 l ND l1.04E3 l ND l ND l ND l5.20E6 l l Y-92 l2.53E-4 l ND l7.24E-6 l ND l ND l ND 17.31E0 l l Y-93 l1.05E0 l ND l2.90E-2 l ND j ND l ND l1.57E4 l l Zr-95 l3.83E3 l8.42E2 l7.50E2 l ND l1.21E3 l ND l8.79ES l l Zr-97 l1.92E0 l2.77E-1 l1.64E-1 l ND l3.98E-1 l ND l4.20E4 l l Nb-95 l3.18E5 l1.24E5 l8.84E4 l ND l1.16E5 l ND l2.29E8 l l Mo-99 l ND l8.14E7 l2.01E7 l ND l1.74E8 l ND l6.73E7 l l Tc-99ml1.32E1 l2.59El l4.29E2 l ND j3.76E2 l1.32E1 l1.47E4 l l Tc-101l 0 l 0 l 0 l ND l 0 l 0 l 0 l l Ru-103l4.28E3 l ND l1.65E3 l ND l1.08E4 ( ND l1.11ES l l Ru-105l3.82E-3 l ND l1.39E-3 l ND l3.36E-2 l ND l2.49E0 l l Ru-106l9.24E4 l ND l1.15E4 l ND l1.25ES l ND l1.44E6 l lAg-110ml2.09E8 l1.41E8 l1.13E8 l ND l2.63E8 l ND l1.68E10 l lCd-109 l ND l3.86E6 l1.79ES l ND l3.45E6 l ND l1.25E7 l lSn-113 l6.10E8 l1.25E7 l3.48E7 19.27E8 l ND l ND l4.31E8 l lSb-124 l1.08E8 l1.41E6 l3.81E7 l2.40E5 l ND l6.03E7 16.79E8 l lSb-125 l8.70E7 l6.71E5 l1.83E7 l8.06E4 l ND l4.85E7 l2.08E8 l lTe-125ml7.38E7 l2.00E7 l9.84E6 l2.07E7 l ND l ND l7.12E7 l lTe-127ml2.08E8 l5.60E7 l2.47E7 l4.97E7 l5.93E8 l NT l1.68E8 l lTe-127 l3.05E3 l8.22E2 l6.54E2 l2.11E3 18.67E3 l ND l1.19E5 l lTe-129ml2.71E8 l7.57E7 14.21E7 18.74E7 l7.96E8 l ND l3.31E8 l lTe-129 l 0 l 0 l 0 l 0 l2.90E-9 l ND l6.17E-8 l lTe-131ml1.60E6 l5.53E5 l5.89E5 l1.14E6 l5.35E6 l ND l2.24E7 l lTe-131 l 0 l 0 l 0 l 0 l 0 l ND l 0 l lTe-132 l1.02E7 l4.52E6 l5.46E6 l6.58E6 l4.20E7 l ND l4.55E7 l 1 l lI-130 l1.73E6 l3.49E6 l1.80E6 l3.84E8 l5.22E6 l ND l1.63E6 l lI-131 l1.30E9 l1.31E9 l7.45E6 l4.33E11 l2.15E9 l ND l1.17E8 l lI-132 l6.02E-1 l1.11E0 l5.08E-1 l5.13E1 l1.69E0 l ND l1.30E0 l lI-133 l1.74E7 l2.15E7 l8.13E6 13.99E9 l3.58E7 l ND l8.66E6 l lI-134 l 0 l 0 l 0 l 0 l 0 l ND l 0 l l
j I APA-ZZ-01003 Rsv. O TABII 5 (Cont'd.) PATHWAY DOSE FACTORS g (R ) FOR RADIONUCLIDES 0THER THAN NOBLE GASES" Grass-Cow-Hilk Pathway (m8 mrem /yr) per (uCi/sec) l l l l Total l l l l l lNuclidel Bone l Liver l Body l Thyroid l Kidney l Lung i GI-LLI l lI-135 l5.40E4 l9.72E4 l4.60E4 l8.61E6 l1.49ES l ND l7.40E4 l lCs-134 l2.26E10 l3.72E10 l7.84E9 l ND l1.15E10 l4.13E9 l2.00E8 l lCs-136 l1.01E9 l2.77E9 l1.79E9 l ND l1.48E9 l2.20E8 l9.74E7 l lCs-137 l3.22E10 l3.09E10 l4.56E9 l ND l1.01E10 l3.62E9 l1.93E8 l lCs-138 l 0 l 0 l 0 l ND l 0 l 0 l 0 l lBa-139 l1.89E-7 l 0 l5.48E-9 l ND l 0 l 0 l1.09E-5 l lBa-140 11.17E8 l1.03E5 l6.84E6 l ND l3.34E4 l6.12E4 l5.93E7 l lBa-141 l 0 l 0 l 0 l ND l 0 l 0 l 0 l lBa-142 l 0 l 0 l 0 l ND l 0 l 0 l 0 l lLa-140 l1.95El l6.80E0 l2.29E0 l ND l ND l ND l1.90E5 l lLa-142 l 0 l 0 l 0 l ND l ND l ND l2.90E-6 l lCe-141 l2.19E4 l1.09E4 l1.62E3 l ND l4.78E3 l ND l1.36E7 l lCe-143 l1.87E2 l1.02E5 l1.47El l ND l4.26E1 l ND ll.49E6 l lCe-144 l1.62E6 l5.09E5 l8.66E4 l ND l2.82E5 i ND l1.33E8 l lPr-143 l7.19E2 l2.16E2 l3.57El l ND l1.17E2 l ND l7.75ES l lPr-144 l 0 l 0 l 0 l ND l 0 l ND l 0 l lNd-147 l4.45E2 l3.61E2 l2.79El l ND l1.98E2 l ND l5.71E5 l lEu-154 l9.41E4 l8.47E3 l7.73E3 l ND 13.72E4 l ND l1.97E6 l lHf-181 l6.43E2 l2.35E3 l2.90E2 l ND l4.75E2 l ND l8.65E5 l lW-187 l2.91E4 l1.73E4 l7.73E3 l ND l ND l ND l2.42E6 l lNp-239 l1.72E1 l1.23E0 l8.68E-1 l ND l3.57E0 l ND l9.14E4 l (a) The child age group; refer to reference 11.14.5. l l l
l APA-ZZ-01003 L Rw. 0 l 1 I l TABLE 5 (Contd.) l l PATHWAY DOSE FACTORS g (R ) FOR RADIOWNDES NR M NOBE GASES
- Grass-Goat-Milk Pathway
)
(m8 mrem /yr) per (ttci/sec) ] I l \ l l l l Total l l l l l j lNuclidel Bone l Liver i Body l Thyroid 1 Kidney 1 Lung l GI-LLI l l l H-3 l ND l3.20E3 l3.20E3 .l3.20E3 l3.20E3 l3.20E3 l3.20E3 l ! l Be-7 l8.98E2 l1.53E3 l9.82E2 l ND l1.50E3 l ND 18.53E4 l l l C-14 l1.19E9 12.39E8 l2.39E8 l2.39E8 l2.39E8 l2.39E8 l2.39E8 l l Na-24 l1.07E6 l1.07E6 l1.07E6 l1.07E6 l1.07E6 l1.07E6 l1.07E6 l l P-32 l9.33E10 l4.37E9 l3.60E9 l ND l ND l ND l2.58E9 l l Cr-51 l ND l ND l1.23E4 l6.78E3 l1.87E3 l1.25E4 l6.48E5 l l l Mn-54 l ND l2.52E6 l6.70E5 l ND l7.06E5 l ND l2.11E6 l l Mn-56 l ND l1.54E-3 j3.49E-4 l ND l1.87E-3 l ND l2.24E-1 l l Fe-55 l1.45E6 l7.71E5 l2.39E5 l ND { ND l4.36E5 l1.43E5 l l Fe-59 l1.56E6 l2.53E6 l1.26E6 l ND l ND l7.33E5 l2.63E6 l l Co-57 l ND l4.60E5 l9.31E5 l ND l ND l ND l3.77E6 l l Co-58 l ND l1.45E6 l4.45E6 l ND l ND l ND l8.49E6 l l Co-60 l ND l5.18E6 l1.53E7 l ND l ND l ND l2.87E7 l l Ni-63 [3.56E9 l1.90E8 l1.21E8 l ND l ND l ND l1.28E7 l l Ni-65 11.99E-1 l1.87E-2 l1.09E-2 l ND l ND l ND l2.29E0 l l Cu-64 l ND l8.31E3 l5.02E3 l ND l2.01E4 l ND l3.90E5 l l Zn-65 14.96E8 l1.32E9 l8.22E8 l ND l8.33E8 l ND l2.32E8 l l Zn-69 l 0 l 0 l 0 l ND l 0 l ND l1.35E-10] l Br-82 l ND l ND l1.38E7 l ND l ND l ND l ND l l Br-83 l ND l ND l ND l ND l ND l ND l ND l l Br-84 l ND l ND l ND l ND l ND l ND l ND l l Br-85 l ND l ND l ND l ND l ND l ND l ND l l Rb-86 l ND l1.06E9 l6.50E8 l ND l ND l ND l6.80E7 l l Rb-88 l ND l 0 l 0 l ND l ND l ND l 0 l l Rb-89 l ND l 0 l 0 l ND l ND l ND l 0 l l Sr-89 l1.39E10 l ND l3.97E8 l ND l ND l ND l5.38E8 l l Sr-90 l2.35E11 l ND l5.95E10 l ND l ND l ND l3.16E9 l l Sr-91 l2.74E5 l ND l1.03E4 l ND l ND l ND l6.04E5 l
]
I
l APA-ZZ-01003 Rav. 0 TABLE 5 (Contd.) a
' PATHWAY DOSE FACTORS (Rg ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Grass-Goat-Milk Pcthway-(m8 mrem /yr) per (pCi/sec) l l l l Tatal l l l l l lNuclidel Bone l Liver i Body l Thyroid l Kidney l Lung l GI-LLI l l Sr-92 14.58E0 l NO l1.84E-1 l ND l ND l ND l8.68E1 l l Y l3.87El l ND l1.03E0 l ND l ND l ND l1.10E5 l ,
l Y-91m l 0 l ND l 0 l ND l ND l ND l 0 l l Y-91 l4.68E3 l ND l1.25E2 l ND l ND l ND l6.24E-5 l l Y-92 l3.04E-5 l ND l8.69E-7 l ND l ND l ND l8.77E-1 l l Y-93 l1.27E-1 ] ND l3.48E-3 l ND l ND l ND l1.89E3 l l Zr-95 l4.60E2 l1.01E2 l9.00E1 l ND l1.45E2 l ND l1.05E5 l l Zr-97 l2.30E-1 l3.33E-2 l1.96E-2 l ND l4.78E-2 l ND l5.04E3 l l Nb-95 l3.81E4 l1 48E4 l1.06E4 l ND l1.39E4 l !!D l2.75E7 l l Mo-99 l ND l9.76E6 l2.42E6 l ND l2.09E7 l ND l8.08E6 l l Tc-99ml1.59E0 l3.11E0 l5.15El l ND l4.52E1 l1.58E0 l1.77E3 l
-l Tc-101l 0 l 0 l 0 l ND l 0 l 0 l 0 l l Ru-103l5.14E2 l ND l1.98E2 l ND l1.29E3 l ND l1.33E4 l l Ru-105l4.58E-4 l ND l1.66E-4 l ND l4.03E-3 l ND l2.99E-1 l l Ru-106l1.11E4 l ND l1.38E3 l ND l1.50E4 l ND l1.72E5 l lAg-110ml2.51E7 l1.69E7 l1.35E7 l ND l 3.15E7 l ND l2.01E9 l lCd-109 l ND l4.63E5 l2.15E4 l ND l4.13E5 l ND l1.50E6 l lSn-113 l7.32E7 l1.5CE6 l4.17E6 l1.11E8 l ND l ND l5.17E7 l lSb-124 l1.30E7 l1.69E5 l4.56E6 l2.87E4 l ND 17.22E6 l8.14E7 l lSb-125 l1.04E7 l8.04E4 l2.19E6 l9.66E3 l ND [5.81E6 l2.49E7 l !
lTe-125ml8.85E6 12.40E6 l1.18E6 l2.48E6 l ND l ND [8.54E6 l lTe-127ml2.50E7 l6.72E6 l2.96E6 l5.97E6 l7.12E7 l ND l2.02E7 l lTe-127 l3.66E2 l9.86El l7.85El l2.53E2 l3.04E3 l ND l1.43E4 l lTe-129ml3.25E7 l9.09E6 l5.05E6 l1.05E7 l9.55E7 l ND l3.97E7 l lTe-129 l 0 l 0 l 0 l 0 l 0 l ND l7.40E-9 l lTe-131ml1.92E5 l6.64E4 l7.07E4 l1.37E5 l6.43E5 l ND l2.69E6 l lTe-131 l 0 l 0 l 0 l 0 l 0 l ND l 0 l J lTe-132 l1.23E6 15.42E5 l6.55E5 l7.90E5 l5.04E6 l ND l5.46E6 l J lI-130 l2.07E6 l4.19E6 l2.16E6 l4.61E8 l6.26E6 l ND l1.96E6 l ] l 1
\
APA-ZZ-01003 R2v. O TABLE 5 (Contd.) PATHWAY DOSE FACTORS (R g ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES" Grass-Goat-Milk Pathway (m8 mrem /yr) per (yCi/sec) l l l l Total l l l l l lNuclidel Bone i Liver l Body l Thyroid l Kidney l Lung l GI-LLI l lI-131 l1.56E9 l1.57E9 l8.94E8 l5.20E11 l2.58E9 l ND l1.40E8 l lI-132 l7.22E-1 l1.33E0 l6.10E-1 l6.15El l2.03E0 l ND l1.56E0 l lI-133 l2.09E7 l2.58E7 19.76E6 14.79E9 l4.30E7 l ND l1.04E7 l lI-134 l 0 l 0 l 0 l 0 l 0 l ND l 0 l lI-135 l6.48E4 l1.17ES l5.52E4 l1.03E7 l1.79E5 l ND l8.88E4 l lCs-134 l6.79E10 !1.11E11 l2.35E10 l ND l3.45E10 l1.24E10 l6.01E8 l lCs-136 l3.03E9 l8.32E9 l5.38E9 l ND l4.43E9 l6.61E8 l2.92E8 l lCs-137 l9.67E10 19.26E10 l1.37E10 l ND l3.02E10 l1.09E10 l5.80E8 l lCs-138 l 0 l 0 l 0 l ND l 0 l 0 l 0 l lBs-139 l2.27E-8 l 0 l 0 l ND l 0 l 0 l1.31E-6 l lBa-140 l1.41E7 l1.23E4 18.20E5 l ND 14.01E3 17.34E3 l7.12E6 l lBa-141 l 0 l 0 l 0 l ND l 0 l 0 l 0 l lBa-142 l 0 l 0 l 0 l ND l 0 l 0 l 0 l lLa-140 l2.34E0 l8.17E-1 l2.75E-1 l ND l ND l ND l2.28E4 l lLa-142 l l ND l ND l ND 0 l 0 l 0 l3.49E-7 l lCe-141 l2.62E3 l1.31E3 l1.94E2 l ND 15.74E2 l ND l1.63E6 l lCo-143 l2.25El l1.22E4 l1.77E0 l ND l5.12E0 l ND l1.79E5 l lCe-144 l1.95E5 l6.11E4 l1.04E4 l ND l3.38E4 l ND l1.59E7 l lPr-143 l8.62E1 12.59El l4.28E0 l ND l1.40E1 l ND 19.30E4 l lPr-144 l 0 l 0 l 0 l ND l 0 l ND l 0 l lNd-147 l5.34E1 l4.33E1 l3.35E0 l ND l2.37El l ND l6.85E4 l lEu-254 l1.13E4 l1.02E3 l9.27E2 l ND l4.46E3 l ND l2.36ES l lHf-181 l7.71El l2.81E2 l3.48E1 l ND l5.70E1 l ND l1.04E5 l lW-187 l3.49E3 l2.07E3 l9.27E2 l ND l ND l ND l2.90E5 l lNp-239 l2.06E0 l1.48E-1 l1.04E-1 l ND l4.28E-1 l ND l1.10E4 l (a) The child age group; refer to reference 11.14.5. l l - _ _ _
APA-ZZ-01003 Riv. O TABLE 5 (Contd.) PATINAY DOSE FACTORS (Rg) FOR RADIONUCLIDES OTHER THAN NOBLE GASES" Vegetation Pathway (m8 mrem /yr) per (11Ci/sec) l l l l Total l l l l l lNuclidei Bone I Liver i Body l Thyroid l Kidney l Lung l GI-LLI l l H-3 l ND l4.01E3 l4.01E3 l4.01E3 l4.01E3 l4.01E3 l4.01E3 l l Be-7 13.38E5 l5.76ES l3.70E5 l ND l5.64E5 l ND l3.21E7 l l C-14 l8.89E8 l1.78E8 l1.78E8 l1.78E8 l1.78E8 l1.78E8 l1.78E8 l l Na-24 l3.75ES l3.75ES l3.75ES l3.75ES l3.75ES l3.75E5 l3.75E5 l l P-32 l3.37E9 l1.57E8 l1.30E8 l ND l ND l ND l9.30E7 l l Cr-51 l ND l ND l1.17E5 l6.50E4 l1.78E4 l1.19E5 l6.21E6 l l Mn-54 l ND l6.65E8 l1.77E8 l ND l1.86E8 l ND l5.58E8 l l Mn-56 l ND l1.88E1 l4.24E0 l ND l2.27El l ND l2.72E3 l l Fe-55 l8.01E8 l4.25E8 l1.32E8 l ND l ND l2.40E8 l7.87E7 l l Fe-59 l3.97E8 l6.43E8 l3.20E8 l ND l ND l1.86E8 l6.69E8 l l Co-57 l ND l2.98E7 l6.04E7 l ND l ND l ND l2.45E8 l l Co-58 l ND l6.44E7 l1.97E8 l ND l ND l ND l3.76E8 l l Co-60 l ND l3.78E8 l1.12E9 l ND l ND l ND l2.10E9 l l Ni-63 l3.95E10 l2.11E9 l1.34E9 l ND l ND l ND l1.42E8 l l Ni-65 l1.05E2 l9.89E0 l5.77E0 l .ND l ND l ND l1.21E3 l l Cu-64 l ND l1.10E4 l6.64E3 l ND l2.66E4 l ND l5.16ES l l Zn-65 l8.12E8 l2.16E9 l1.35E9 l ND l1.36E9 l ND l3.80E8 l l Zn-69 l1.09E-5 l1.57E-5 l1.45E-6 l ND l9.52E-6 l ND l9.11E-4 l l Br-82 l ND l ND l2.04E6 l ND l ND l ND l ND l l Br-83 l ND l ND l5.37E0 l ND l ND l ND l 0 l l Br-84 l ND l ND l 0 l ND l ND l ND l 0 l l Br-85 l ND l ND l 0 l ND l ND l ND l 0 l l Rb-86 l ND l4.58E8 l2.82E8 l ND l' ND l ND l2.94E7 l l Rb-88 l ND l 0 l 0 l ND ND l ND l 0 l l Rb-89 l ND l 0 l 0 l ND l ND l ND l 0 l l Sr-89 l3.59E10 l ND l1.03E9 l ND l ND l ND l1.39E9 l 1 Sr-90 l1.24E12 l ND l3.15E11 l ND l ND l ND l1.67E10 l l Sr-91 l5.24E5 l ND l1.98E4 l ND l ND l ND l1.16E6 l l Sr-92 l7.28E2 l ND l2.92E1 l ND l ND l ND l.'.38E4 l l Y-90 l2.31E4 l ND l6.18E2 l ND l ND l ND l6.37E7 l
APA-ZZ-01003 Rsv. 0 TABLE 5 (Contd.) PATHWAY DOSE FACTORS g(R ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES" Vegetation Pathway (m8 arem/yr) per (lici/sec) l l l Total l l l l l lNuclide- Bone l ' Liver i Body l Thyroid l Kidney l Lung 1 GI-LLI l l Y-91m l8.87E-9 l ND l3.23E-10l ND l ND l ND l1.74E-5 l l Y-91 l1.86E7 l ND ;4.99ES- l ND l ND l ND l2.48E9 l l Y-92 l1.58E0 l ND l4.53E-2 l ND l ND l ND l4.58E4 l l Y-93 l3.01E2 l ND l8.25E0 l ND l ND l ND l4.48E6 l l Zr-95 l3.86E6 l8.45ES l7.55E5 l ' ND l1.21E6 l ND l8.84E8 l l Zr-97 l5.70E2 l8.24E1 l4.86El l ND' l1.18E2 l ND l1.25E7 l l Nb-95 l4.10E5 ll.59E5 l1.14E5' l ND l1.50E5 l ND l2.95E8- l l Mo-99 l -ND l7.71E6 .l1.91E6 l ND l1.65E7 l ND l6.38E6 l l Tc-99ml4.71E0 19.24EO. l1.53E2 l ND l1.34E2 l4.69E0 l5.26E3 l l Tc-101l 0 l 0 l 0 l ND l 0 l 0 l 0 l
'l Ru-103l1.54E7 l ND l5.90E6 l ND l3.87E7- l ND l3.97E8 l l Ru-105l9.16El l ND l3.32E1 l ND l8.05E2 l ND l5.98E4 1 l Ru-106l7.45E8 l ND l9.30E7 l ND l1.01E9 l ND l1.16E10 l lAg-110ml3.22E7 l2.17E7 l1.74E7 l ND 14.05E7 l ND l2.58E9 l lCd-109 l ND. l2.45E8 l1.13E7 l ND l2.18E8 l~ND l7.94E8 l lSn-113 l1.58E9 j3.25E7 .l9.00E7 l2.40E9 l ND l ND l1.12E9 l lSb-124 l3.52E8 l4.56E6 l1.23E8 l7.76E5 l ND ll.95E8 l2.20E9 l lSb-125 l4.99E8 l3.85E6 l1.05E8 l4.62E5 l ND l2.78E8 l1.19E9 l lTe-125ml3.51E8 l9.50E7 l4.67E7 19.84E7 l ND l ND l3.38E8 l lTe-127ml1.32E9 l3.56E8 11.57E8 l3.16E8 l3.77E9 l ND l1.07E9 l
-lTe-127-l1.00E4 l2.69E3 l2.14E3 l6.91E3 l2.84E4 l ND l3.90E5 l lTe-129ml8.38E8 l2.34E8 11.30E8 l2.70E8 l2.46E9 l ND l1.02E9 l lTe-129 l1.16E-3 l3.23E-4 l2.75E-4 !8.26E-4 l3.39E-3 l ND l7.20E-2 l lTc-131ml1.54E6 l5.33E5 i 5.68E5 l1.10E6 l5.16E6 l ND l2.16E7 l lTe-131 l 0 1 0 1 0 l 0 l 0 l ND l 0 l lTe-132 l6.9.iE6 l3.09E6 l3.73E6 l4.50E6 l2.87E7 l ND l3.11E7 l lI-130 l6.16ES l1.24E6 l6.38E5 l1.37E8 l1.86E6 l ND l5.79E5 l lI-131 l1.43E8 l1.44E8 l8.17E7 l4.75E10 l2.36E8 l ND l1.28E7 l lI-132 18.58E1 11.58E2 l7.25El 17.31E3 l2.41E2 l ND l1.86E2 l lI-133 l3.56E6 l4.40E6 l1.67E6 l8.18E8 l7.34E6 l ND l1.77E6 l
'lI-134 l1.55E-4 l2.88E-4 l1.32E-4 16.62E-3 l4.40E-4 l ND l1.91E-4 l lI-135 l6.62E4 l1.13E5 l5.33E4 l9.97E6 l1.70E5 l ND l8.58E4 l APA-ZZ-01003 R2v. O TABLE 5 (Contd.)
PATHWAY DOSE FACTORS (R g ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES" Vegetation Pathway (m8 mrem /yr) per (pCi/sec) l l l l Total l l l l l lNuclidel Bone l Liver l Body l Thyroid l Kidney l Lung i GI-LLI l lCs-134 l1.60E10 l2.63E10 l5.55E9 l ND l8.15E9 l2.93E9 l1.42E8 l lCs-136 l8.17E7 l2.25E8 l1.45E8 l ND l1.20E8 l1.78E7 l7.90E6 l lCs-137 l2.39E10 l2.29E10 l3.38E9 l ND l7.46E9 l2.68E9 l1.43E8 l lCs-138 l 0 l 0 l 0 l ND l 0 l 0 l 0 l lBa-139 l4.80E-2 l2.56E-5 l1.39E-3 l ND l2.24E-5 l1.51E-5 l2.77E0 l lBa-140 l2.77E8 l2.42E5 l1.62E7 l ND l7.89E4 l1.45E5 l1.40E8 l lBa-141 1 0 l 0 l 0 l ND l 0 l 0 l 0 l lBa-142 l 0 l 0 l 0 l ND l 0 l 0 l 0 l lLa-140 13.25E3- l1.14E3 l3.83E2 l ND l ND l ND l3.17E7 l lLa-142 l2.50E-4 l7.98E-5 l2.50E-5 l ND l ND l ND l1.58E1 l lCe-141 l6.56E5 l3.27E5 l4.86E4 l ND l1.43E5 l ND l4.08E8 l lCe-143 l1.72E3 19.31E5 l1.35E2 l ND l3.91E2 l ND l1.36E7 l lCo-144 l1.27E8 l3.98E7 l6.78E6 l ND l2.21E7 l ND l1.04E10 l lPr-143 l1.46E5 l4.38E4 17.25E3 l ND l2.37E4 l ND l1.58E8 l lPr-144 l 0 l 0 l 0 l ND l 0 l ND l 0 l lNd-147 l7.17E4 e5.81E4 l4.50E3 l ND l3.19E4 l ND l9.20E7 l lEu-154 l1.66E8 l1.50E7 l1.37E7 l ND l6.57E7 l ND l3.48E9 l lHf-181 l4.90E5 l1.79E6 l2.21E5 l ND l3.62E5 l ND l6.59E8 l lW-187 l6.47E4 l3.83E4 l1.72E4 l ND l ND l ND l5.38E6 l lNp-239 l2.55E3 l1.83E2 l1.29E2 l ND l5.30E2 l ND l1.36E7 l (a) The child age group; refer to reference 11.14.5. w____.-_-
't APA-ZZ-01003 Rav. O 3.6 Gaseous Radwaste Treatment System
' 41862 3.6.1 REC 9.9.1.1
- 4160.
'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 the projected doses in 31 days due to gaseous effluent releases, from each unit, to areas'at and beyond the SITE BOUNDARY would exceed:
- a. O.2 mrad to air from gamma radiation, or
- b. O.4 mrad to air from beta radiation, or
- c. O.3 mrem to any organ of an Individual 3.6.2 Description of the Gaseous Radwaste Treatment System j The gaseous radwaste treatment system and the ventilation 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 consists of eight (8) waste gas decay tanks, six (6) for use during normal operations and two (2) for use during shutdown conditions. Normally, waste gases will be retained for at least 60 days prior to discharge. These systems will provide reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept ALARA. 1 i
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l APA-ZZ-01003 R;v. O 3.6.3 Operability of the Gaseous Radwaste Treatment Syster! , The OPERABILITY of the gaseous radwaste treatment system ensures this system will be available for use when gases require treatment prior to their release to the environment. OPERABILITY is demonstrated through compliance with Sections 9.6.1.1, 9.7.1.1, and 9.8.1.1. Projected doses (gamma air, beta air, and organ dose) due to gaseous effluents at or beyond the SITE BOUNDARY are determined each 31 days by dividing the cumulative annual total by the number of elapsed months. 4.0 DOSE AND DOSE COMMITMENT FROM URANIUM FUEL CYCLE SOURCES 41864 4.1 REC Section 9.10.1.1 4160 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity 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 Calculation of 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 and internal organs due to gamma ray exposure from submersion in a cloud of radioactive noble gases, ground plane j exposure, and direct radiation from the Unit and outside storage tanks; I f 1 l 1
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iAPA-ZZ-01003 L ' Rsv 0-L.s y p u h). Dose to the skin due to beta radiation from submersion in a cloud of radioactive noble t gases, and ground plane exposure; c). Thyroid dose due to inhalation and ingestion
~
of radioiodines; and. d): Organ dose due to inhalation and ingestion of radioactive material. It is assumed that total body dose from sources of. gamma radiation irradiates internal body P organs at the same numerical. rate. (Ref. 11.12.5) The dose'from gaseous effluents is' considered to
~be the summation of the dose at the individual's residence and.the dose to the individual from activities within the SITE BOUNDARY.
Since the doses via liquid releases are very-
. conservativelycavaluated,.there is reasonable assurance that no.real individual will'~ receive.a significant dose from radioactive liquid release pathways. Therefore, only-doses to individuals via airborne pathways and doses resulting from
" direct radiation are considered in determining compliance to 40 CFR 190. (Ref. 11.12.3)
Itlshould.be noted that there'are no other Uranium Fuel Cycle Sources withf.n 8km of the-p Callaway~ Plant.
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4.2.1 Identification of the MEMBER OF THE PUBLI_C-The MEMBER OF THE PUBLIC is considered to be.a. ) real individual, including all persons.not i occupationally associated with the Callaway Plant, but who may use portions.of'the plant' site n for. recreational or other_ purposes ~not associated with the' plant. (Ref. 11.4 and 11.8.10.') 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 could be reevaluated.if necessary to. demonstrate continued compliance with 40 CFR 190 through the use of more realistic assumptions. (Ref. 11.12.3 and 11.12.4) The evaluation of Total Dose from the Uranium Fuel Cycle should consider the dose to two Critical Receptors: a)~The Nearest Resident, and b) The Critical Receptor within the SITE BOUNDARY. 4.2.2 Total Dose to the Nearest Resident The dose to.the Nearest Resident is.due to plume-exposure from noble gases,. ground plane exposure, and inhalation and ingestion pathways. It is conservatively assumed that each ingestion pathway (meat, milk,. and vegetation) exists at the location of the Nearest Resident. It is assumed that direct radiation dose from operation of the Unit and outside storage tanks, and dose from gaseous effluents due to activities within the SITE BOUNDARY, is negligible for the Nearest Resident. The total Dose from the Uranium Fuel Cycle to the Nearest Resident is calculated using the methodology discussed in Section 3, using concurrent meteorological data for the location of the Nearent Resident with the highest value of X/Q. L u L
= _ _ _ _ _ _ = _ _ _ _ ._
j APA-ZZ-01003 ! Rsv. 0- J
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i i The location-of the Nearest Resident in each meteorological sector is determined from the l Annual Land Use Census conducted in accordance j with the Requirements of Section 9.12.1.1. j 1 4.2.3 Total Dose to the Critical Receptor Within the SITE BOUNDARY l The Union Electric Company has entered into an agreement with the State of Missouri Department of.' Conservation for management of the residual 1 ands surrounding the Callaway Plant, including some areas within the SITE BOUNDARY. Under the terms of this agreement, certain areas have been Opened to the public for low intensity recreational uses (hunting, hiking, sightseeing, etc. ) but recreational use is excluded in sus area immediately surrounding the plant site (Refer to Figure 4.1). Much of the residual lands within the SITE BOUNDARY are leased to area farmers by the Department of Conservation to provide income to support management and development costs. Activities conducted under these leases are primarily comprised of farming (animal feed), grazing, and forestry. (Ref 11.7.2, 11.7.3, 11.13, 11.13.1). Based on the utilization of areas within the SITE BOUNDARY, it is reasonable to assume that the critical receptor within the SITE BOUNDARY is a farmer, and that his dose from activities within the SITE BOUNDARY is due to exposure incurred while conducting his farming activities. The current tenant has estimated that he spends approximately 1100 hours per year working in this area (Ref 11.5.5). Occupancy of areas within the SITE BOUNDARY is assumed to be averaged over a period of one year. Any reevaluation of assumptions should include a reevaluation of the occupancy period at the locations of real exposure (e.g. a real individual would not simultaneously exist at each point of maximum exposure). i i -- -__ -- 1
APA-ZZ-01003 Rsv. 0 l 1 4.2.3.1 Total Dosa to the Farmer from Gaseous Effluents The Total Dose to the farmer from gaseous effluents is calculated using the methodology discussed in Section 3, utilizing concurrent meteorological data at the farmer's residence and historical meteorological data from Table 10 for i activities within the SITE BOUNDARY. These ' dispersion parameters were calculated by assuming that the farmer's time is equally distributed over the areas farmed within the SITE BOUNDARY, and already have the total occupancy of 1100 hours / year factored into their value (Ref. j 11.5.6). l The residence of the current tenant is located at a distance of 3830 meters in the SE sector. No meat or milk animals or vegetable gardens were q identified by the 1987 Land Use Census for this location, therefore, the gaseous effluents dose at the farmer's residence is due to plume exposure from Noble Gases and the ground plane and inhalation pathways. It is assumed that food ingestion pathways do not
, exist within the SITE BOUNDARY, therefore the gaseous effluents dose within the SITE BOUNDARY is due to plume exposure from Noble Gases and the ground plane and inhalation pathways.
4.2.3.2 Total Dose from Direct Radiation , 4.2.3.2.1 Direct Radiation Dose from outside Storage Tanks The Refueling Water Storage Tank (RWST) has the highest potential for receiving significant amounts of radioactive materials, and constitutes the only potentially significant source of direct radiation dose from outside storage tanks to a MEMBER OF THE PUBLIC. (Ref. 11.6.14, 11.6.15, 11.6.16, and 11.6.17.) I
[p-APA-ZZ-01003~
'Rev.. 0.
Dir
- radiation dose from the RWST to a MEMBER
. 1 OF d PUBLIC-is. determined at the nearest point ;
of the owner Controlled Area fence which is not obscured by significant plant structures. This has been determined to be 450 meters from.the RWST.- 1 The RWST.is a right circular cylinder approximately 12 meters in. diameter,.14 meters in height with a capacity.of-approximately 1,514,000 liters. (Ref. 11.6.17.) The walls are of type 304 stainless steel and have an average thickness oof .87 cm. (Ref. 11.14.1.) The' direct radiation dose from the RWST is calculated. based on the tank's average isotopic content and the' parameters discussed above,
- considering buildup'and. attenuation within the volume source. Appropriate methodology for calculating the dose rate from a volume source is-given in TID-7004, " Reactor Shielding Design Manual" (Ref. 11.17). The computer program ISOSHLD-(Ref.'11.18, 11.19, 11.20) will normally be utilized to perform this calculation.
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' APA-ZZ-01003- 1 "l i .
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. 4.2.3.2.2 Direct Radiation Dose from the-Reactor;
"- The maximum direct radiation dose from'the Unit to a MEMBER OF THE.PUBLIC.has been determined to~
be 7E-2 mrads/ calendar year, based on a point source of primary coolant N-16 in the steam generators._ This' source; term.was then projected onto the inside surface of the" containment dome, taking credit ~for shielding provided by the containment dome and for distance attenuation. No credit was allowed fer-shielding by other-structures or. components. The number of. gammas per second was generated and then converted to a dose: rate at the given distance by use of ANSI /ANS-6.6.1,_" Calculation and Measurement of Directiand' Scattered Gamma Radiation from LWR Nuclear Power Plant 1979", which considers attenuation and buildup in air.- The final value is based ~on one unit operating at 100%' Power.. The distance was determined to be 367 meters, which is approximately the closest point of the boundary _of the owner Controlled Area fence which1 is not;obscurred by significant plant structures.
-(Ref. 11.14.3.).
.The maximum direct radiation dose from the Unit.-
.to a MEMBER OF THE PUBLIC due to-activities
-within the SITE BOUNDARY is thus approximately 9E-3 mrads per year, assuming a maximum occupancy of 1100 hours per year.
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I 41835 5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING' 3535 41866 5.1 REC Section 9.11.1.1
~
i The radiological environmental monitoring program i shall-be conducted as spacified in Table 9.11-A. q 5.2 Description of the Radiological Environmental ' Monitoring Program The Radiological Environmental Monitoring Program is intended to act as a background data base for reoperation.and to supplement the radiological effluent release monitoring program during plant operation. Radiation exposure to the public from the various specific pathways and direct radiation'can be. adequately evaluated by this program. Some deviations from the sampling frequency may be necessary:due to seasonal unavailability,. hazardous conditions, or other legitimate 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 Environmental Operating Report. The Environmental samples are collected and analyzed at the frequency outlined in Table 6. Reporting levels and lower limits of detection (LLD) are given in' Tables-7 and 8.
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.. c Airborne, waterbere, and ingestion samples E F' collected under the monitoring program are b
A analyzed by an independent,' third-party laboratory. This laboratory is required to participate in the Environmental Protection Agency's (EPA) Environmental Radioactivity
, Laboratory.Intercomparison Studies (Crosscheck) a Program or an equivalent program. Participation includes all of the determinations (sample medium ,
- radionuclides combination) that are offered by the EPA and that are also included in the monitoring program.
5.3 Performance Testing of Environmental Thermoluminescence Dosimeters Thermoluminescence Detectors (TLD's) used in the Environmental Monitoring Program are tested for accuracy and precision to demonstrate compliance with Regulatory Guide 4.13. (Ref. 11.16). Energy dependence is tested at several energies between 30 kev and 3MeV corresponding to the approximate' energies of the predominant Noble Gases (80, 160, 200 kev), Cs-137-(662 kev), Co-60 (1225 kev), and at least one energy less than 80 kev. Other testing is performed relative to either Cs-137 or~Co-60. I 4 1 i i i L - _ I
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. TABLE 8 (CONTINUED _)_
' TABLE NOTATION-
-(a)- The LLD is defined'for purposes of
, , compliance with the Radioactive Effluent
' Controls as the' smallest concentration of e: 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 S b
, E
- V
- 2.22 'Y
- exp (-lat)
Where: LLD = The lower limit of detection as defined above (as picoeurie per unit mass or
. volume).
Sb= The standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute). 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 picoeurie. Y= The fractional radiochemical yield (when applicable).
APA-ZZ-OlOO3 Rzv.. O 1= The radioactive decay constant for the particular radionuclides and, ht = 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 priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. Analyses are performed in such a manner that the-stated LLDs are achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering 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) This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report. (c) Required detection capabilities for thermoluminescent dosimeters used for 1 environmental measurements shall be in acccrdance with the recommendations of Regulatory Guide 4.13, Revision 1, July 1977. (Refer to Section 5.3) I l l
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LE!!D UNION ELECTRIC CO. A IIIIL!!LilLOCli1033 CALLAWAY FLANT
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APA-ZZ-01003 R;v. O 6.0 DETERMINATION OF ANNUAL AVERAGE AND'SHORT TERM ATMOSPHERIC DISPERSION PARAMETERS 6.1 Atmospheric Dispersion Parameters The values presented in Table 9 and Table 10 were determined through the analysis of on-site meteorological data collected during the three year period of May 4, 1973 to May 5, 1975 and March 16, 1978 to March 16, 1979. 6.1.1 Long-Term Dispersion Estimates The variable trajectory plume segment atmospheric transport model MESODIF-II (NUREG/CR-0523) and the straight-line Gaussian dispersion model XOQDOQ (NUREG/CR2919) were used for determination of the long-term atmospheric dispersion parameters. A more detailed discussion of the methodology and data utilized to calculate these parameters can be found elsewhere (Ref. 11.6.12). The Unit Vent and Radwaste Building Vent releases are at elevations 66.5 meters and 20 meters above grade, respectively. Both release points are within the building wake of the structures on which they are located, and the Unit Vent is equipped with a rain cover which effectively eliminates the possibility of the exit velocity exceeding five times the horizontal wind speed. All gaseous releases are thus considered to be ground-level releases, and therefore no mixed mode or elevated release dispersion parameters were determined. (Ref. 11.5.2) l 1
)
i j 1 l l l l L___________ )
i APA-ZZ-01003 ) R;v., O ' l l 6.1.2 Determination of Long-Term Dispersion Estimates i for Special Receptor Locations Calculations utilizing the PUFF model were l performed for 22 standard distances to obtain the desired dispersion parameters. Dispersion ]! parameters at the SITE BOUNDARY and at special receptor locations were estimated by logarithmic interpolation according to (Ref. 11.6.13): X=X y (6.1) (d)B dy Where: B= In (X2 /X y)/in (d 2/d y). X,X2 y = Atm spheric concentrations at distances d y .and d , respectively, from the source 2 3 (in Ci/m ), The distances d y and~d2 were selected such that dy <d<dg . 6.1.3 Short Term Dispersion Estimates Airborne releases are classified as short term if they are less than or equal to 500 hours during a i calendar year and not more than 150 hours in any quarter. Short term dispersion estimates are determined by multiplying the appropriate long term di'spersion estimate by a correction factor (Ref. 11.9.1 and 11.15.1): F = (T,/T,) (6.2) J 1 l l l I
> a N
ll ( APA-ZZ-01003 R;v. 0 l-E Where: l ~ j Tg= The total number of hours of the short term l- release. ! i
- T, = The total number of hours in the data collection period from which the long term
! diffusion estimate was determined (Refer to Section 6.1). Values of the slope factor (S), are presented in TABLE 11. Short term dispersion 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 conditions or time of day) to be represented by the annual average dispersion conditions. .(Ref. 11.8.11.) 6.1.3.1' The Determination of the Slope Factor (S). The general approach employed by subroutine PURGE of XOQD0Q (Ref. 11.15.1) was utilized to produce
, values of the slope of the (X/Q) curves (Slope Factor-(S)) for both the Radwaste Building Vent and the Unit Vent. However, instead of using approximation procedures to produce the 15 precentile (X/Q) values, the 15 percentile (X/Q) value for each release and at each location was determined by ranking all the 1-hour (X/Q)3 values for that release and at the location in r descending order. The (X/Q)y value which corresponded to the 15 percentile of all the calculated (X/Q) values within a sector was extracted for use in the intermittent release (X/Q) calculation.
/
7 -^ - I APA-ZZ-01003 ! Rov; O .
- g i !
The' intermittent' release (X/Q) curve was, constructed using the calculated 1-hour 15
. percentile (X/Q)1 and its corresponding annual average (X/Q),. A graphic representation,.of how .
j the computational procedure works is illustrated by Figure 4.8.of-reference-11.15.1. The straight line connecting these points represents (Y/Q)y values for intermittent releases, ranging in duration from one (1) hour to 8760 hours. The slope (S) of the curve is expressed as:
\
I
-log ('(X/Q)1/(X/Q),1 (6.3) 'l S=
log (Ta/T y) i or
-(log (X/Q)y - log (X/0),1 (6.4)
S= log Ta - log T y 6.2 Atmospheric Dispersion Parameters for Farming Areas Within The SITE BOUNDARY The dispersion parameters for farming areas within the SITE BOUNDARY are intended for a narrow scope application: That of calculating the dose to the current tennant farmer from gaseous effluents while he conducts farming activities within the SITE BOUNDARY. 1 l _ _ - - _ - _ _ - _ _ _ - _ _ - - - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ - _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - __ __ _ __s
>- APA-ZZ-01003-Rsv. O-1
~
For the purpose of these calculations, it.was
- assumed that all of the farmer's time, approximately 1100 hours per year, is spent on croplands within the SITE-BOUNDARY, and that his time is divided evenly qver all of the croplands.
Fractional acreage / time -' weighted dispersion parameters were calculated for each plot as described in reference 11.5.6. The weighted dispersion parameters for each plot were then , summed (according to type) in order to produce a composite value of'the dispersion parameters. The dispersion parameters presented in Tables 9 & 10 therefore represent the distributed activities of the farmer within the SITE BOUNDARY and his estimated' occupancy period. I i i; l i _ _ _ _ _ _ _ _ _ _ . - _ _ . J
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( elf
) ) n r aen t l c c i) a8d ewi eh a ( ( he 8e)m n pg m e i)t( si
)
)
c n i e c n Wc( n9v6 e1i v r5 tem ai he Sl l ime.sur l c ( A b e sy el e Y ( a d god 1 rat gl l R t t t i ar r 1 ave na s ea b ( A w a D N C G H V o o e e a g e rd R An sfs e uauf sn e.ene d ic di l t N U u l tl eeee ir O O t t t t t go aaaRhhh ue I B s s s sn s nB VDV(1Tw BV T A C E T e rer r e ee a a a ar a rd e r ime rt ))) )) O t e e e ea e ai abc de L S N N N NC N FS ((( (( i
y-APA-ZZ-01003 R3v. O TABLE 11 SHORT TERM DISPERSION PARAMETERS (a) (c) Slope Factor (S) Location (b) Sector Distance Unit Vent Radwaste (meters) Building Vent Site Boundary S 1300 .328 .3'20 Nearest Cow NW 5053 .263 .266 Nearest Goat NW 5053 .263 .266 Nearest Meat NNW 2736 .262 .268 Animal Nearest Vegetable NNW 2865 .264 .268 Garden Nearest Residence NNW 2865 .264 .268 (a) Reference 11.5.3 (b) Data from 1987 Land Use Census (c) Recirculation Factor = 1.0 i
- ss -
l 4 l
t t t t t n n n n n _ 3 y y y y e e e e e 0 d d 0 G r r r r i i s s ds sidsd i i 1 NN a a a a 3 IO d d d d e e e e e
- LI n n n n R R R R R Z0 LT u u u u Z- OA o o o o t t t t t
- . RC B B B B s s s s s
'AV TO e r r e e e e PI NL e e e e r r r AR O t t t t a a a a a C i i i i e e e e e S S S S N N N N N G
S NP R IU E LO - T LR d d . E OG - - - - - l dl dl l ) M R - - - - - i i i i 1 A TE - h h h h . R NG C C C C 1 A OA 1 P C . 1 N 1 O I e S c R n E e P . r S ) ) -
) ) e I
de de de de f 2 D er er er er . e 1 tt tt tt tt d) R C lei lei ei lei ee ( E I r r lr r tf L R pi pi pi pi Q B E ea ea ea ea lei l / A H dh dh dh dh pr D T P n uy uynuy n n uy ei S N da * *
- f O OR /a /a /a /a Q /h Q Q Q o M IE dd dd dd dd / d / / /
T ST e e e e D ey D D D d ya A RE y6 y6 y6 y6 ' a EM a2 a2 a2 a2 ad e F PA c. c. c. c. c t O SR e2 e2 e2 e2 e8 s IA d( d( d( d( d( n N DP I O - I Q Q Q Q Q d T / / / / / e A X X X X X s C I u L s P i P A E d C 1 1 e N . . t E 1 1 e R . . l E 5 5 p F e _ E 1 1 3 3 2 2 2 2 2 d R . / 2 2 a t
& 2 2 2 2 2 d M e C 5 5 1 2 5 5 5 5 5 y 0 . a 0 3 3 4 4 3 3 3 3 3 c e
3 3 d n , y o Q i / r d t X r i o i i A B s , A o p 4 a l 1 a m a n e - t m t i D C e a o k Y B G T S e d A n n W , , , , a n n o a f s s s s l o Tf a a a a P i i 3 A G C G G t t - P d a a H e e e e n l t E l l l l u a e k t r S b b b b o h g l a o O o o o o r n e i e F D N H N N C I V M H * , i
k/E APA-ZZ-01003 i Rcv.-0 ; i 1 1 7.0 REPORTING REQUIREMENTS 1 7.1 ANNUAL RADIOLOGIC 3L ENVIRONMENTAL OPERATING ) REPORT l I Routine Annual Radiological Environmental Operating. Reports covering the-operation of the j unit during the previous calendar year shall be J submitted prior to May 1 of each year. The ,) initial report shall be submitted prior to May 1 1 of the year following initial criticality, i The Annual Radiological Environmental Operating Reports'shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance - activities for the report period, including a comparison with preoperational studies, with operational' controls and with previous l environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of Land Use Censuses required by Section 9.12. The Annual Radiological Environmental Operating Reports'shall include the results of~ analysis of all radiological environmental samples and of all' environmental radiation measurements taken during the-period pursuant to the locations specified in the Table and Figures in the ODCM, as well as summarized the tabulated results of these analyses and measurements in the format of the table in.the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.
- 101 -
l
7 4 APA-ZZ-01003 P R3v. O The reports shall also include the following:- a. f summary description of the . radiological: U environmental monitoring program; at'least two legible. maps
- covering all sampling locations keyed to a table giving distances and directions from the. centerline of one reactor;ithe results.
of licensee participation in the InterlaboratoryL Comparison Program and the' corrective action-being taken if the specified program is not being performed as required by Section-9.13.1; reasons for not. conducting the Radiological Environmental
' Monitoring Program as required'by-Section 9.11.1 and discussion of all deviations-from the-sampling schedule of-Table 9.11-A, discussion of environmental. sample measurements that exceed the reporting levels of Table 9.11-B,' but are not the result.of the plant effluents, pursuant to Section 9.11.1; and discussion of all analyses in.
which the LLD required by Table 9.11-C was not achievable.
*0ne map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.
-102-l l
'I k----_ - - . - - . - - - , , - - - - - - . . . - - - - - - - - . - - -
m7 y, -
;d ( g - '
APA-ZZ-OlOO3 1 j Rov. O E 7.2- SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT a Routine Semiannual: Radioactive Effluent' Release Reports coveringfthe operation of the unit during.
-the previous 6 months of operation-shall be .
submitted within 60 days after January 1 and July
~1 of each year. The-period of.the first report shall begin with the date of initial criticality.-
The Semiannual'Radioact'ive~ Effluent Release Reports shall include a summary of.the quantities of radioactive liquidLand' gaseous effluents and solid waste released from the. unit as outlined in Regulatory Guide'1.21; ." 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 shall be supplemented with three additional categories:
class of solid waste (as defined by 10 CFR Part
'60),. type of container (e.g.', LSA,- Type A, Type l
B, Large Quantity), and SOLIDIFICATION agent or absorbent (e.g.,. cement, urea formaldehyde). 4
- 103 -
I
APA-ZZ-01003 R2v. O The Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed,' wind direction, and atmospheric stability *. This same report shall include 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 shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluants to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Technical Specifications, Figures 5.1-3 and 5.1-4) during the report period using historical average atmospheric conditions. All assumptions used in making these assessments, i.e., specific activity, exposure time and location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in~ gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM). 1
*In lieu of submission with the Semiannual Radioactive Effluent Release Report, Union Electric has the option of retaining this summary of required meteorological data on site in a file that shall be provided 1 to the NRC upon request.
-104-l l 1 l l
APA-ZZ-01003 R;v. O The Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from Reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977. The Semiannual Radioactive Effluent Release Reports shall 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 Semiannual Radioactive Effluent Release Reports shall include any major changes made during the reporting period to any Liquid or Gaseous Treatment Systems, pursuant to Section 10.1. It shall also include a listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to Section 9.12.1. The Semiannual Radioactive Effluent Release Reports shall also include the following ; information: An explanation as to why the j inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Section 9.1.1 or 9.2.1, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Technical Specification 3.11.1.4 or 3.11.2.6, respectively. l
-105-I i
l I
-______-__-__a
l ~ m. , APA-ZZ-OlOO3 Rsv. 0
- ,c e2815 The Samiannual Radioactive Effluent Release'-
. Reports shall' also include as.a part of or.
submitted concurrent with,-a~-complete and legible copy of'all revisions of the ODCM that-occurred-during the reporting period pursuant-to-
. Specification 6.14.2.
{ 4 9
-106-
sc , :APA-ZZ-01003~ 4 Rnv. 0 8.0' IMPLEMENTATION OF ODCM METHODOLOGY The ODCM provides the mathematical relationships used to implement the Radioactive Effluent
. Controls..'
For routine effluent release:and dose assessment, .1 - . computer codes are utilized to implement the ODCM methodologies. These codes have been evaluated , by a qualified independent reviewer.to ensure that they. produce-results. consistent with the
-^
methodologies presented in the ODCM. (Ref.
- 11. 5 . 4.' )
- 107 - {
APA-ZZ-01003 R v. O
'9.0 RADIOACTIVE EFFLUENT CONTROLS (REC)
NOTE 1. The terms in this section that appear in CAPITALIZED TYPE are defined in Technical Specifications.
- 2. All frequency notations are per Table 1.1 of Technical Specifications.
9.O.1 Compliance with the Controls contained in the succeeding Controls is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Control, the associated ACTION requirements shall be met. 9.O.2 Noncompliance with a Control shall exist when the I requirements of the Control and associated ACTION requirements are not met within the specified time intervals. If the Control is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required. 9.O.3 When a Control is not met, except as provided in the associated ACTION requirements, within 1 hour ACTION shall be initiated to place the unit in a ~ MODE in which the Control does not apply by placing it, as applicable, in:
- a. At least HOT STANDBY within the next 6 hours,
- b. At least HOT SHUTDOWN within the following 6 hours, and
- c. At least COLD SHUTDOWN within the subsequent l 24 hours. l Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Control. Exceptions to these requirements are stated in the individual Controls.
This Control is not applicable in MODE 5 or 6. I
-108-i
APA-ZZ-01003 R2v. D 9.O.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions for the Control are met without reliance on provisions contained in the ACTION requirements. This provision shall not prevent. passage through or to OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirerrnts are stated in the individual Controls 9.0.5 Operability of equipment included in Section 9.0 must be tracked in the Equipment Out-Of-Service Log (EOSL)-as per ODP-ZZ-00002, Equipment Status Control.
-109-
APA-ZZ-01003 Rsv. O 9.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 41838 9.1.1 Controls 9.1.1.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table 9.1-A shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Section 9.3.1.1 ' are not exceeded. The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM). APPLICABILITY: At all times. ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above Control, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable.
- b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 9.1-A. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the next Semiannual Radioactive Effluent Release Report, pursuant to Section 7.2, why this inoperability was not corrected within the time specified.
- c. The provisions of Sections 9.O.3 and 9.O.4 are not applicable.
41839 9.1.2 Surveillance Recu$rements 9.1.2.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 9.1-B.
-110-l l
_. . _ _ . _ _ _ - -- -__-_--_-_-___-_-__-_--__-A
APA-ZZ-01003 Rcv. O j, i 1
,41840 9.1.3 Bases l
4 9.1.3.1 Radioactive Liquid Effluent Monitoring ! Instrumentation i The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in' liquid effluents during actual or potential releases of liquid effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure L that the alarm / trip will occur prior to exceeding
- the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to l 10 CFR Part 50.
i 1 l l
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l l l l l l
-111-l l
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3 0 0 1 0 ZO Z'.
, Av Pe i l
u AR O I T 1 2 2 3 4 4 4 4 4 C 3 3 3 3 3 3 3 3 3 A ~ , N SE O MLL I UE8 1 1 1 1 1 1 1 1 I T MNA A INR T NAE N IHP E MCO M U R T S N I G N I R O ) A T 2
- I 5
1 N -
. O E 9 M k I ) ;
- 5 w E T M 4 o L N B - l B E ) ( E e F A U 8 R n T L 1 r - i s F - o ) F L s F E t 9 H -
a E R i 5 ( e e e e g yp d - n - - n n n D n B o E r i i r B I a H M R o iL L L a U ( - t h d Q m e E i e e e c ns r r g L n g r rg r g i a I L a o r ( o l t a M a a a D n E A i h r e h h h w V ge o s n c o tm n c c c m o I e i s s s e d T ns M i i t L i i i t w C ia D n s o y s e D D D s o A T de e y l D N il g n M S e g A B n S B I D E ve r w M oR a o n e i a k k c r w o et r e A U r' h d i t v h n n d _ R R Pf c w a s e c a a Tw s w T o s o r a D s T o a o S N s rn iD B l DW t i ro or B D l W T oo d n g e r n g i e et d I ti t t r i n it t o i u m i i o u i na s t d q e s n n t iq l on e a l i r e o o a o Mi w r i L u w M M r L o e u e m d s d C e e n y yr e n B y a e te R e r e R t t e r d iT G e a M s s G a e v d n d e a d n ic ti iu ma b i don eti W W m n i u a o b _, ct q e r c a q e c m aa t u e R i ) ) t e o om iL S T S L 1 2 S S C io w _ dt o _ au . . . . l . . . . RA a b c d F a b c d 1 2 Ll1
APA-ZZ-01003~ Rsv.~ 0 Table 9.1-A (Continued) ACTION STATEMENTS ACTION 31.- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent. releases via this pathway may continue for up to '14 days provided that prior to initiating a release:
- a. At least two independent samples are analyzed in accordance with Section 9.3.2.1, and
- b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge'line valving.
Otherwise, suspend release of radioactive effluents via this pathway. ACTION 32 - With the number of channels OPERABLE Less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are analyzed for principal gn.nma emitters-and I-131 at a lower limit of detection as specified in Table 9.3-A:
- a. At least once per 12 hours when the specific activity of the secondary coolant is greater than 0.01 microcurie / gram DOSE EQUIVALENT I-131, or
- b. At least once per 24 hours when the specific activity of the secondary coolant is less than or equal to 0.01 microcurie / gram DOSE EQUIVALENT I-131.
-113-
. _ 1.,
APA-ZZ-01003 R2v.-O-Table 9.1-A (Continued) ACTION STATEMENTS ACTION 33 - With the number of channels OPERABLE Less than l' required by the Minimum Channels OPERABLE requirement, effluent releases via this' pathway I may continue for up'to 30 days provided that ' i prior to initiating a releasei
- a. At-least two independent samples are analyzed in accordance with Section 9.3.2.1, and
- b. At least'two technically qualified members of' the facility staff independently verify the'
-release rate calculations and discharge line valving.
Otherwise, vnpend release of radioactive affluents via this pathway. ACTION 34 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated'at'least ence per 4. hours during' actual releases. Pump performance curves generated in place may be used to estimate-flow. 5
-114-
3 0 0 L 1 0 L
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Z; GEO
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Av LNTS 1 1 1 1 Pe AAAE ( ( ( ( A. A. A. A. AR NHRT Q Q Q Q N N N N ACE . P O S T N E M N E O R LI I ET ~ - U NA ) ) ) ) Q NR 2 2 2 2 E AB ( ( ( ( R HI R R R R R R R R CL - E A C C N A L L I E V . . . . R EK U CC A. A. A. A. S RE P M M P M N N N UH N OC O S I - T A T N E B M
- U 1 R
. T L ) ) ) )
9 S EK 3 3 3 3 N NC , ( ( ( ( , t E I NE D D D 'D D 0 D D L AH . B G HC A N C T I R O - T ) e I 8 n N 1 r i O - o ) L M E t 9 R i 5 e eg T d - n - n N n B o E r i r E a H M R o L a U ( - t h d L m e E i e c n F r r g L n g s a F a o r ( o r i E l t a M a D n A i h r e h w D ge o s n c o me n c m o I t i s e d U ns M i i t L i t w Q ia D n sy D s o I de e o s eg y l L il g n M S e n S B E ve oR a o r w n et i c ra w o e r V r h d i v h d t e I T Pf c wo s a r s a D e c w s s o e o w C s o i l DW i l W T A rn D B t D B D oo g d n d g I ti e r n i e e r i n D it t o i u m t o u i A na s t d iq e se t q l R on e a l r a i o Mi w r i u L u w r L ow m d e) ) s d e Co yr e n2 B y5 a a n y l te R e5 r4 e R e r dF iT G- e a- M G a e v d E n d d ns ic ti iu mR a-ib dE nR e i m n o- t u a o ba is eM r a q e c mp ct aa iq tB u eH R i cF t e oy T L S( T S( L S S CB N E iom dt o w o M au . . . . l . . . . U RA a b c d F s b c d R
. T S
N . . I 1 2
! f
< 4 APA-ZZ-01003' Rsv.-O.
' TABLE 9.1-B (Continued)'
TABLE NOTATIONS q
- 41841 .
(1) - The ANALOG CHANNEL OPERATIONAL TEST'shall'also demonstrate'that automatic isolation of this
- pathway and control. room alarm annunciation occur as appropriate if any of the following conditions '
exists:' a.' Instrument. indicates measured levels above the
~
Alarm / Trip Setpoint (isolation and' alarm), or y b. Circuit failure (alarm only), or.
- c. Instrument indicates a downscale failure (alarm'only), or
- d. Instrument controls not set in operate mode (alarm only).
(2) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference (gas l or liquid and solid) standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers - - - - l that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its' intended E range of energy, measurement range, and establish mcnitor response to a solid calibration source. For subsequent CHANNEL CALIBRATION, NBS traceable standard (gas, liquid, or solid) may be used; or a gas, liquid, or solid source that has been calibrated by relating it to equipment that was-previously (within 30 days) calibrated by the same. geometry and type of source standard traceable to NBS. (3) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days on which continuous, periodic, or batch releases are made.
-116-L l
I t . _ _ ._ _ _ -_ _ _- __ - _ _ __ ___ _ _
APA-ZZ-01003 R;v. 0 9.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 41872 9.2.1 Controls 9.2.1.1 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 9.2-A shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Section 9.6.1.1 and Technical Specification 3.11.2.5 are not exceeded. The Alarm / Trip Setpoints of these channels meeting Section 9.6.1.1 shall be determined and adjusted in accordance with the methodology and parameters in the ODCM. APPLICABILITY: As shown in Table 9.2-A. ACTION:
- a. With a radioactive gaseous effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable ,
- b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 9.2-A. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the next Semiannual Radioactive Effluent Release Report, pursuant to Technical Specification 6.9.1.7, why this inoperability was not corrected within the time specified.
- c. The provisions of Sections 9.0.3 and 9.0.4 are not applicable.
41843 9.2.2 Surveillance Requirements 9.2.2.1 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 9.2-B.
-117-
7 APA-ZZ-01003-Rsv. 0'
.9.2.3 Bases 41844 9.2.3.1 Radioactive' Gaseous Effluent Monitoring
- Instrumentation The radioactive gaseous affluent. instrumentation ~
is provided to monitor and control, as applicable, the releases of-radioactive materials in gaseous affluents during actual or potential releases of gaseous effluents. The Alarm / Trip Setpoints for these instruments shall' be calculated,and adjusted in accordance with the methodology and parameters in the ODCM to
. ensure that the alarm / trip will occur prior to exceeding'the limits of 10.CFR Part 20. The OPERABILITY and use of this instrumentation is consistent.with the requirements of General
- Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The sensitivity of any noble.
gas activity monitor used to show compliance with the gaseous effluent ~ release requirements of Section 9.7.1.1 shall be such that concentrations as low as 1 x 10 ' pCi/cc are measurable.
-118-L
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E M U R T S N I S G L N E I N R NE O AL T iB l A I CA
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M UP 9 MO ' - ' - A. ' . A. T I 1 l 1 l 1 2 1 1 N l 1 1 1 N 1 E N N L E I B U M A L T F F E e s S a U e O gl E ne S iR A d G ir vo ) E o 0 V rn 1 I T Po n-i
.oE C - - t iR A a '
t-O r) r rn rah r I o1 oi r m omC D t2 o tm o e to( o A i- t ir t t it t R nE i ne i s nue i oR n oT n y oAs n M- o m M ) o S M a o T r M e c3 r M de r M yG e t yi3 e t ynl e t( l p e s tt- l p e n tae l p e i t y iaE t e R t vm a V ivm m a S vmR m m a ir r a R io- r a R irr r a R ta e S e ttT e S g tao e S w din AAn _ m cl l w g cuG l cl l w e AA p e o r AA p e o p e o _ T t m t l u , m t l o m t l _ N s sg a a e F P sd2 a a e F l s gi a a e F E y an S _ l t an2 S l t i ant S l t _ M S Gi u a r t Ga- u a r u Gia u a r U d e c R e n E e c R e B dn en itc R e R t ei l e emR n i l eii l T n l v i n i w t p m l r- i t r o m w p e l vm i r o m w p S N I e bo d V or o a NP t P r o m n bat l F S a i a NA( olG d I o aP F l S a t s e bor d ore NPT o a t P l F S a _ t t w i n d - n . . . . . o . . . . . a . . . . . U a b c d e C a b c d e R a b c d e 1 2 3
a r APA-ZZ Rsv. O J TABLE 9.2-A (Continued)
. TABLE NOTATIONS
- At all times.
ACTION STATEMENTS i ACTION 38 - With the number of channels OPERABLE less than 4 required by the Minimum Channels OPERABLE ! requirement, the contents of the tank (s) may be
-released to the environment for up to 14 days provided that prior to initiating the release:
- a. At least two independent samples of the tank's contents are analyzed, and I
- b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup.
Otherwise, suspend release of radioactive effluents via this pathway. ACTION 39 - With the number of~ channels OPERABLE lens'than-required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated based on fan status and operating curves or actual measurements at least once per 4 hours. ACTION 40 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 12 hours and these samples are analyzed for radioactivity within 24 hours. ACTION 41 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway. l
-120- ,
l
-APA-ZZ-01003 R:v. 0 TABLE 9.2-A (Continued) 4 TABLE NOTATIONS -
ACTION 43 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary-sampling equipment as required in Table 9.6-A. ACTION 45 - Flow rate for this system shall be based on fan status and operating curves or. actual measurements. 1 1
-121-
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L A LN . GEO~
) . . ) . . . . ) . . . .
S ONIT 2 1 1 T LNTS ( A. A. ( A. A. A. A. ( A. A. A. A. N AAAE Q M N Q Q Q N N N N Q N M N N E NHRT M ACE E P R I O
- U Q
E R E C N A N L O L LI I ET ) . . ) ) . . ) ) . . ) E NA 3 4 3 4 3 4 V NR ( A. A. ( ( A. A. ( ( A. A. ( R AB R N N R R R M N R R R N N R R U Hl ' S Cl A N C O I T A T N E M B U
- R EK . . . . . . . . P . . . .
2 T CC S RE A. A. A. A. A. A. A. A. , A. A. A. A. 9 N Ul l M N N N N P N N N N M N N M N I OC ; ' ' . . E S L G B N A I T R O T I N O M L . . P . EK T NC A. A. , A. N NE D W W N D D W W N D D W W N D E AH U HC L C F F E ) . 0 S 1 - U - O c cE _ E i iR S t t - _ A r) ra rai l r G o1 r om r m omG _ t2 o to o e to( t o E t i- t it t it V nE i nue i s nue i I oR n oAs n y oAs n T M- o m M a) o S M a o _ C T r M e de3 r M de r M A yG e t ynl3 e t ynl e O t( i e sy l e n tae l e I i p t itae- RE p t e i R p t _ D vm m a S vm R r a m a V vm r a m a A ir r a R irf- R irf g tao e S R _ R ta e S e taoT e S n cl _ t m e AA cl l p e m t l w o g r u cl AAn. o G l p m t e w di AAno lo m l p e t l w o s sg a a et F P s gi 2 a a et F l s gi a a e F y an S ant 2 S l l t _ S Gi l u a r e n t Gia- u a r iu antGia S u c R a r e e B dn p ni d e c R c R t n l v i e bo ei d n i w t l p r o m n bort d e eiiR m dnE e ir w l vm- i n t o m l p e t eii l vm bor d i t r o m w l p _ V or o a l a i oreG to a l a s ore o a l a _ T NP l P F S a NPT( P F S a HPT B P F S N t t w - E i n d M n . . . . . o . . . . a . . . . . _ U U a b c d e C a b. c d e R a b c d e R T S _ N . . . I 1 2 3
APA-ZZ-01003-Rsv. O TABLE 9.2-B (Continued) TABLE NOTATIONS 41845-
- LAt'all' times.
(1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of-this pathway and control room alarm annunciation occur as appropriate if any of the following conditions exists:
- a. Instrument indicates measured'l'evels above the Alarm / Trip Setpoint (isolation and alarm), or
- b. Circuit failure (alarm only), or
- c. Instrument indicates a downscale failure (alarm only), or
- d. Instrument controls not set in operate mode (alarm only).
(2) The ANALOG CHANNEL OPERATIONAL TEST shall'also
-~
demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
- a. Instrument indicates measured levels above the Alarm Setpoint, or
- b. Circuit failure, or
- c. Instrument indicates a downscale failure, or
- d. Instrument controls not set in operate mode.
l
-123- ,
' ' -_u-----__-_--___--a--
APA-ZZ-01003. Rsv. 01 TABLE 9.2-B (Continued) TABLE NOTATIONS-(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference (gas or liquid and solid) standards certified by the National' Bureau of Standards (NBS) orlusing standards that have been obtained from' suppliers that participate in measurement asaurance. activities with NBS. These standards shall- . permit calibrating the system over its intended range of energy,' measurement range, and establish monitor response to:a-solid calibration source. For subsequent CHANNEL CALIBRATION, NBS traceable standard'(gas, liquid,'or solid) may be used; or: a: gas, ' liquid, or solid source that has.been calibrated.by relating it to equipment that was previously (within 30 days) by the same geometry. and type of source traceable to NBS. (4) If flow rate is determined by exhaust fan status and fan performance curves, the.following surveillance operations shall be performed at-least once per 18 months:
~a. The specific vent flows by direct measurement, - -
or
- b. The differential pressure across the exhaust fan and vent flow established by the fan's
" flow-AP" curve, or
- c. The fan motor horsepower measured and vent flow established by the fan's
" flow-horsepower" curve.
-124-
APA-ZZ-01003' ; , R v. 0 i i i i 2881 9.3 LIQUID EFFLUENTS CONCENTRATION
\
41846 9.3.1 Controls 4160 9.3.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Technical Specification's Figure 5.1-4) 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 x 10-
- microcurie /ml total activity.
APPLICABILITY: At all times. ACTION:
- a. With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, immediately restore the concentration to within the above limits.
~
- b. The provisions of Sections 9.0.3 and 9.O.4 are not applicable.
41847 9.3.2 Surveillance Requirements 2895 9.3.2.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 9.3-A. 9.3.2.2 The results of the radioactivity analysis shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Section 9.3.1.1.
-125-
i ,, APA-ZZ-01003 Rsv. 0 9.3.3: Bases l 9.3.3.1- =This~section is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the' concentration levels' a specified in 10 CFR Part 20,, Appendix B, '
*/'-
Table II, Column 2.- This limitation provides additional assurance'that~the: levels of ,
. radioactive' materials in bodies:of. water in UNRESTRICTED AREAS will-result-in> exposures within: ..(1) the Section II. A design objectives V of Appendix I,-10 CFR Part.50, to a MEMBER OF THE' PUBLIC, and (2) the limits.of.
10 CFR Part 20.106(e) to-the population. The concentration limit.for dissolved or' entrained-noble gases is based upon the' assumption that
-Xe-135,is.the controlling radioisotope and its
'MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International' Commission.on Radiological Protection-(ICRP) Publication 2.
9.3.3.2 The: required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs-). Detailed' discussion of-the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-3OO (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry", Anal. Chem. 40, 586-93 (1968), and Hartwelle J. K., " Detection Limits for Radioanalytical Counting Techniques", Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). l
-126-1
APA-ZZ-01003 R,v. O TABLE 9.3-A RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM l l l l LOWER. LIMIT l'
-l l MINIMUM .l l OF DETECTION l l LIQUID RELEASE l SAMPLING L ANALYSIS' l TYPE OF ACTIVITY l (LLD)(1) l TYPE I FREQUENCY FREQUENCY l ANALYSIS l (vCi/ml) l l l l l l~1. Batch Waste l P l P l l l
~
-l Release lEach Batch lEach Batch l Principal' Gamma l 5x10 l Tanks I) l l l l Emitters (3) l l l l l l l l
-6 l l l l I-131 l 1x10 g l a. Waste l- l l l
-l Monitor l l l- l Tank
-5 l l P l M l Dissolved and l 1x10 g l lOne Batch /Ml l Entrained Gases l l-l l l l(Gamma Emitters) l l l b. Secondary l l l
l Liquid l l l l
~
l Waste l P l M l H-3 l 1x10 l l Monitor leach Batch l Composite (')l 1 l l Tank l l l Gross Alpha l 1x10" l l l l l I l
- c. Discharge l lSr-89, Sr-90 -8 l P l Q 1 5x10 g
-Monitor -lEach Batch l Composite (')l Fe-55 -6 l ~l' 1x10 g l Tank I l 1 I l l l l l l l l 2. Continuous l l W l Principal Gamma l 5x10" l l Releases ( ) l Daily ( ) l Composite )l Emitters (3) l l l l Grab Sample l l l l
-6 l l l l I-131 l 1x10 g l Steam l l l l l Generator l l l l l Blowdown l M l M l Dissolved and l 1x10' l l l Grab Sample l l Entrained Gases l l l l l l(Gamma Emitters) l l .,
l l I I I l 1
-5 l l l M l H-3 l 1x10 g l [ Daily (0) l Composite )l l l
~7 l l Grab Sample l l Gross Alpha l 1x10 l l l l l l l
~0 l l l Q ISr-89, Sr-90 l 5x10 l l l Daily (6) l Composite )l l l j
-6 l l Grab Sample l lFe-55 l 1x10 ;
I I i l I l l
-127- ;
APA-ZZ-01003 R;v. O TABLE 9.3-A (Continued) TABLE NOTATIONS (1) The LLD is defined, for purposes of these controls, 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 only 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 s b E
- V = 2.22 x 108
- Y
- exp (-lat)
Where: LLD = the "a priori" lower limit of detection (microcuries per unit mass of volume), sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute), E = the counting efficiency (counts per disintegration), V = the sample size (units of mass or volume), 2.22 x 10' = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, X = the radioactive decay constant for the particular radionuclides (s 1), and at = the elapsed time between the midpoint of l sample collection and the time of counting (s). Typical values of E, V, Y, and At should be used in the calculation. 1
-128-
M APA-ZZ-OlOO3 Rsv. 0 TABLE 9.3-A (Continued) TABLE NOTATIONS 41848 It.should be recognized that the LLD is defined as a a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori_(after the-fact) limit for a particular measurement. (2) A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed by a method described in the ODCM to assure representative sampling. (3) The principal gamma emitters for which the LLD control applies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Seminannual Radioactive Effluent Release Report pursuant to Technical Specification 6.9.1.7, in the format outlined in
-Regulatory Guide 1.21, Appendix B, Revision 1, June 1974.
(4) A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released. Prior to analysis, all samples taken for the composite shall be thoroughly mixed in order for the composite samples to be representative of the effluent release. ; l (5) A continuous release is the discharge of liquid f wastes of a nondiscrete volume, e.g., from a 2 volume of a system that has an input flow during the continuous release. I 1 I l
-129- i
< r, APA-ZZ-01003' Rsv.=0 TABLE 9.3-A icontinued)
.i TABLE NOTATIONS
'(6) - Samples shall be taken at the initiation of l
effluent flow and at.least once-per 24 hours thereafter while the release is r e: urring. To be
- representative of'the.' liquid effluent, the sample volume'shall be proportioned to the effluent ~
stream discharge volume . The ratio of sample
. volume to effluent discharge volume shall be' '
maintained constant for all samples taken for the composite. sample. c I
-130-
r __ , ! APA-ZZ-01003 R;v. 0
-9.4 DOSE L 41849 9.4.1 Controls l
j 4160 9.4.1.1 The' dose or dose commitment to a MEMBER OF THE I PUBLIC from radioactive materials in liquid i effluents released, from each unit, to UNRESTRICTED AREAS (see Technical Specification's j Figure 5.1-4) shall be limited: l 1 i
- a. During any calendar quarter to less than or !
equal to 1.5 mrems to the whole body and to j less than or equal to 5 mrems to any organ, ' and
- b. During any calendar year to less than or equal to 3 mrems to the whole body and to less than or equal to 10 mrems to any organ.
APPLICABILITY: At all times. ACTION:
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. This Special Report shall also include: (1) the results of radiological analyses of the drinking water source, and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR Part 141, Clean Drinking Water Act.*
I
- b. The provisions of Sections 9.0.3 and 9.0.4 are not applicable.
*The requirements of ACTION a.(1) and (2) are applicable only if drinking water supply is taken from the receiving water body within 3 miles of the plant discharge. In the case of river-sited plants this is 3 miles downstream only. l
-131- I l
APA-ZZ-01003: RIv. 0 9.4.2 Surveillance Require nents 9.4.2.1 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days. 41850 9.4.3. Bases 9.4.3.1 This section is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix I, 10 CFR Part 50.. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the' required operating flexibility and at the same time implement the
-guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable". Also, for fresh water sites with drinking water supplies that can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclides concentrations in the finished drinking water that are in excess of the requirements of 40 CFR part 141. The dose calculation methodology and parameters in the ODCM implement the requirements in Section III.A of Appendix I which specify that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I",
Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I", April 1977. I
-132- 1 1
s APA-ZZ-01003 R;v 0 9.5 LIQUID RADWASTE TREATMENT SYSTEM 41851 9.5.1 Controls 4160 9.5.1.1 The Liquid Radwaste Treatment System shall be OPERABLE and appropriate portions of the system shall be used to reduce releases of radioactivity when the projected doses due to the liquid effluent, from each unit, to UNRESTRICTED AREAS (see Technical Specification's Figure 5.1-4) would exceed 0.06 mrem to the whole body or O.2 mrem to any organ in a 31 day period. APPLICABILITY: At all times. ACTION:
- a. With radioactive liquid waste being discharged without treatment and in excess of the above limits and any portion of the Liquid Radwaste Treatment System not in operation, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that includes the following information:
- 1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the' reason for the inoperability, 1
- 2. Action (s) taken to restore the inoperable l equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Sections 9.0.3 and 9.O.4 are not applicable.
41852 9.5.2 Surveillance Requirements j 9.5.2.1 Doses due to liquid releases from each unit to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Liquid Radwaste Treatment Systems are not being fully utilized.
-133-
'APA-ZZ-01003 Rsv. O 9.5.2.2- The installed Liquid Radwaste Treatment System shall~be considered OPERABLE by meeting
. Sections 9.3.1.1 and 9.4.1.1.
9.5.3' Bases 9.5.3.1- The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require- , treatment prior to release to the environment. The requirement that the appropriate portions of this. system be_used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is'reasonabley achievable".. This section implements the. requirements of ._ 10 CFR Part 50.36a, General Design Criterion 60 of: Appendix A to 10 CFR Part 50 and the design. objective given in Section II.D of Appendix I to 10 CFR Part 50. ~The specified limits governing the use of appropriate portions of the Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives-set forth in Section 11.A of Appendix I, 10 CFR Part 50,-for liquid effluents.
-134-
= ___-_ ___ _ _ --__ - -- _ . _ _ _ _ _
73 6'."I L'^
~APA-ZZ-01003 Rsv. 0-4.
9.6- GASEOUS EFFLUENTS DOSE RATE
'1853, 4 9.6.1 Controls
- 4160 9.6.1.1 The dose rate due to radioactive materials
~~ released in gaseous effluents from the site to areas at and.beyond the SITE BOUNDARY'(see Technical Specification's Figure 5.1-3) shall be-limited-to the following: ,
- a. For noble gases: Less than or equal to 500 mrems/yr to the whole body and less than oor equal to 3000 mrems/yr to the skin, and l
- 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.mrems/yr to any organ.
APPLICABILITY: At.all times.
. ACTION:
- a. With the dose rate (s) exceeding the above limits, immediately restore the release rate to within the above limit (s).
- b. The provisions of Sections 9.0.3 and 9.O.4 are not applicable.
41854 9.6.2 Surveillance Requirements 9.6.2.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM. 9.6.2.2 The dose rate due to Iodine-131 and 133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 9.6-A. 1
-135-
o APA-ZZ-01003 R;v. O l .9.6.3 Bases
; 41855 9.6.3.l' This section is provided to ensure that the dose-at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose. limits of
-10 CFR Part 20 to UNRESTRICTED AREAS. The annual
- l. dose limits are the doses associated with the l' concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits previde reasonable assurance that radioactive material discharged in gaseous effluents will not result in the~ exposure of a MEMBER'OF THE PUBLIC in'an
UNRESTRICTED AREA, either within or outside the SITE' BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR 20.lO6(b)). For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the ocrupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODCM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond SITE BOUNDARY to less than or equal to 500 mrems/ year to the whole body or to less than or equal to 3000 mrems/ year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/ year. 9.6.3.2 The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry", Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques", Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
-136-
ii1 lj1 ; q11ll -
~
, lIIlll lell ll
)
- 8
( 3 0 - ) 0 FO-1 ' OLl 0 Li _
- T(m 2 0 1 1 1 Z0 I
/4 4 6 4 6 4 6 4 1 1 1 1 1 MNl - - - - - - - - - - - -
~
Z. Av IOC f LII - 1
~
0 0 1 0 1 1 0 0 1 0 1 0 1 0 1 1 0 0 1 0 1 1 0 0 1 Pe T( x x x x x x x x x x x x x 0 AR RC 1 1 - 1 1 1 1 1 1 1 1 1 1 1 EE WT OE LD ll -
) ) ) ) ) )
2' 2 2 2 2 2 ( ( ( ( ( ( s s s s s s r r r r r r M e e e e e e A t t t t t t R t t t t t t G i i i i i i O m m m m m m R S E E E E E E P I S e a e e a a S Y m m m m m m I L m m m m m m S A a a a a a o 0 Y N G G n C n G n G C a 9 L l'A e e e h p r A O l l d l d l d l l N Y a a i a i a i a a l S A ET p o x p x o x p p A PI i l o i o l o i i , D YV c n c l c f c f c 1 3 n c s 9 N TI n n n n 3 3 s 8 A T 3 3 3 o - ir 1 i i i i 1 i C r - r - r - r - - r r r G A P P H P H P H P I 1 P G S A I I l l
- I 6 L P
9 E' M A S
)
3 ( e e e m L k E t t t B E Y n C a ea ea A T SC a R l l tl tl T S MIN T U ) ) a ) u i u i u A USE P ) ) .5 7 oe7 ce sce sce W MYU ILQPc h P h 3 ( 4 ( M ( H M ( Wrp cl( Wtp il Moil ptp 0 oil ptp ' S NAE E a c M M M am rm aa mrm oaa mrm oaa U INR a ha O MAF E CS PS CPS CPS E lllllIlIll l l lI S A C ) ) ) ) 8 8 8 8 E ( ( ( ( V I ) ) ) ) ) T e 3 6 6 6 6 C l ( e e e( ( ( s ( A ko E l l l s s s O Y nm C p p p u u u u l GC aa R ) m m m o o o o 0 NN TS U 4 a a a u u u u A E P e( S ) S S n n n n R ILU PQPcaP hbp) hb b _l , 5 i t' i i t i t ( b b t ME ar cam 3 a Mar Mar n n n n AH EG ara ( r -o o o o SF llIllilIl EGS MG I G G C C (llIlllIllI C C g n E P e' g i g n s e, d Y y r l i p .d T a u i d yI n c P u l T a E e B i n, S D t u ei A n) l B s . E s e3 t e ad4 m( n u e lee,e L a E G n t e Ft t t v R i n V s s es.o e a e tu e Ri3b s u tk sn t n V i t na eh dn wt l l
,a o ea o r n px ae l s ..
E WT C o U SE RV Aa25 s A . . . . . . C 1 2 3 4 5 6 llIl1ii11llllll1lll11l11llllllllllIllI
! fljilil llil lll[lillj(llilll lt 1
i APA-ZZ-01003 Rev. O TABLE 9.6-A (Continued) TABLE NOTATIONS (Continued) (1) The LLD is defined, for purposes of these controls, 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 only 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 s b E
- V
- 2.22 x 10'
- Y
- exp (-lat)
Where: LLD = the "a priori" lower limit of detection (microCuries per unit mass of volume), sb = the standard deviation of the background counting rate or of the counting rate of a _ blank sample as appropriate (counts per minute), E = the counting efficiency (counts per disintegration), V = the sample size (units of mass or volume), 2.22 x 10' = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, 1 = the radioactive decay constant for the particular radionuclides (s 2), and At = the elapsed time between the midpoint of sample collection and the time of counting (s). Typical values of E, V, Y, and At should be used in the calculation.
-138-
7 APA-ZZ-01003 RGv. O TABLE 9.6-A (Continued) TABLE NOTATIONS (Continued) 41857 It should be recognized that the LLD is defined as a a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. (2) The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, I-131, Cs-134, Cs-137, Ce-141, and Ce-144 in iodine and particulate releases. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Eff3uent Release Report pursuant to Section 7.2, in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974. (3) Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within I hour period. (4) Tritium grab samples shall be taken and analyzed at least once per 24 hours when the refueling canal is flooded. (5) Tritium grab samples shall be taken and analyzed at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool. Grab samples need to be taken only when spent fuel is in the spent fuel pool. (6) The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Sections 9.6.1.1, 9.7.1.1, and 9.8.1.1.
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APA-ZZ-01003-R v. 0 TABLE 9.6-A (Continued) TABLE NOTATIONS (Continued) (7) Samples shall be changed at least once per.7 days and analyses.shall be completed within 48 hours after changing, or-after removal from sampler. For unit vent, sampling shall also be performed at least once per 2 hours for at least 7 days following each shutdown, STARTUP or THERMAL-POWER change exceeding 15% of RATED THEM5AL POWER within a 1-hour period and analyses shall be completed.within 48 hours of changing. When samples collected for 24 hours are analyzed, the corresponding LLDs may be increased by a factor of 10.- This requirement does not apply if: (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the reactor coolant has not
' increased more than.a factor of 3, and (2) the noble gas monitor shows that effluent activity has not. increased more than a factor of 3.
(8) Continuous sampling of the spent fuel building exhaust needs to be performed only when spent fuel is in the spent fuel pool.
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f APA-ZZ-01003 7 : R v. 0 - 9.7 DOSE - NOBLE ' GASi'S
;41858- 9.7.1 Controls
- 4160 9.7.1.1 The. air dose due to noble gases released in gaseous affluente,-from each unit, to areas at and beyond the SITE BOUNDARY (see Technical Specification's Figure 5.1-3) shall be limited to the following: ,
- a. During.any calendar quarter: Less than or equal to 5 mrads for. gamma radiation and'less than or equal ~to 10 mrads for. beta-radiation, and-
- b. During any calendar year: Less than or equal to 10 mrads for gamma radiation and.less than or equal to 20 mrads for beta radiation.
. APPLICABILITY: At all times.
ACTION:
- a. With the calculated air dose from radioactive noble gases in gaseous effluetns. exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification-6.9.2, a Special.
Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to _ reduce the releases and the proposed l corrective actions to be taken to' assure that subsequent releases will be in compliance with the above limits.
- b. The provisions of Sections 9.0.3 and 9.O.4 are not applicable.
41859 9.7.2 Surveillance Requirements 9.7.2.1 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days. 1
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APA-ZZ-01003 R;v. O 9.7.3 Bases 9.7.3.1 This section is provided to implement the requirements of Sections II.B, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Apoendix I to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable". The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors", Revision 1, July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE l BOUNDARY are based upon the historical average atmospheric conditions.
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l APA-ZZ-01003 R3v. 0 , 9.8 DOSE - IODINE-131 AND 133, TRITIUM, AND RADIOACTIVE MATERIAL IN PARTICULATE FORM 41860 9.8.1 Controls 4160 9.8.1.1 The dose to a MEMBER OF THE PUBLIC from Iodine-131 and 133, tritium, 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 (see Technical Specification's Figure 5.1-3) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 7.5 mrems to any organ, and
- b. During any calendar year: Less than or equal to 15 mrems to any organ.
APPLICABILITY: At all times. ACTION:
- a. With the calculated dose from the release of Iodine-131 and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Ccmmission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limits and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
i
- b. 12 provisions of Sections 9.O.3 and 9.O.4 are
.ot applicable.
41861 9.8.2 Surveillance Requirements 9.8.2.1 Cumulative dose contributions for the current ! calendar quarter and current calendar year for Iodine-131 and 133, tritium, and radionuclides in particulate form with hcif-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at least l once per 31 days. j l 1
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I l
- r. .
APA-ZZ-01003 Rsv. 0 9.8.3 Bases 9.8.3.1 This section is provided to implement the
-requirements of Sections II.C, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix.I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable".
The Surveillance Requirements. implement in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely-to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses'to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating-Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors", Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate controls for Iodine-131 and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclides pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were: (1) individual inhalation of airborne radionuclides, i (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) disposition of radionuclides onto grassy areas where milk animals and meat-producing i animals graze with consumption of the milk and meat by man, and (4) deposition on the ground with subsequent exposure of man.
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)
APA-ZZ-01003 R;v. O i 9.9 GASEOUS RADWASTE TREATMENT SYSTEM 41862 9.9.1 Controls 4160 9.9.1.1 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 the projected doses in 31 days due to gaseous effluent releases, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure Technical Specification's 5.1-3) would exceed:
- a. 0.2 mrad to air from gamma radiation, or
- b. O.4 mrad to air from beta radiation, or
- c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.
APPLICABILITY: At all times. ACTION:
- a. With radioactive gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that includes the following information:
- 1. Identification of any inoperable equipment or subsystems, and the reason for the inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Sections 9.0.3 and 9.0.4 are not applicable.
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APA-ZZ-01003 R2v. O
- 9. 9. 2. Surveillance Requirements 41863 9.9.2.1 Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Gaseous Radwaste Treatment Systems are not being fully utilized.
9.9.2.2 The installed VENTILATION EXHAUST TREATMENT SYSTEM and the WASTE GAS HOLDUP SYSTEMS shall be considered OPERABLE by meeting Sections 9.6.1.1 and 9.7.1.1 or 9.8.1.1. 9.9.3 Bases 9.9.3.1 The OPERABILITY of the WASTE GAS HOLDUP SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM
. ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used,-when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable".
This control implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.
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_ _ _ _ _ _ - - - _ - _ _ _ _ _ _ _ _ _ - - _ - _ - - _ . _ _ ___ .- ]
- t. ,
APA-ZZ-01003 Rsv. O l g 9.10 TOTAL DOSE p, 41864 9.10.1 Controls 9.10.1.1 The annual (calendar year) dose or dose commitment to any MEMBER 0F THE PUBLIC due to i releases of radioactivity and to radiation from uranium fuel cycle sources shall.be limited to less than or equal to 25 mrems-to the whole body or any organ, except the thyroid, which shall be - limited to less than or equal to 75 mrems. APPLICABILITY: At all times.
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- APA-ZZ-01003
- Rsv. 0 ACTION:
'a. With the calculated doses from the-release of s
. radioactive materials in-liquid-or gasenus-
' effluents exceeding twice the limits oi Section 9.4.1.la., 9.4.1.lb, 9.7.1.la.,
9.7.1.lb.,.9.8.1.la'., or 9.8.1.lb., calculations-should be made including direct
. radiation contributions from the units and<
from outside storage tanks to determine - whether the above limits of Section 9.10.1.1 ? have-been exceeded. If such'is-the case, prepare and submit to the Commission within-30 days,' pursuant to Technical Specification 16.9.2, a Special_ Report that, defines the corrective actionL.to be taken to reduce subsequent releases _to prevent-
- recurrence.of exceeding-the above limits and includes the schedule for achieving-conformance with the above limits. This.
Special Report, as. defined in 10 CFR 20.405c, shall include an' analysis that estimates the radiation exposure (dose) to a' MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct. radiation,'for the calendar year that includes-the~ release (s) covered by this report. It-shall also describe levels of radiation and concentrations of' radioactive material involved, and'the cause of the' exposure levels or concentrations. If the estimated dose (s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR' Part'190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete,
- b. The provisions of Sections 9.0.3 and 9.0.4 are not applicalbe.
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APA-ZZ-01003 Rav. O t l 41865 9.10.2 - Surveillance Requirements 9.10.2.1 Cumulative dose contributions from liquid and gaseous effluents shall be deternined in accordance with Sections 9.4.2.1, 9.7.2.1 and 9.8.2.1, and in accordance with the methodology and parameters in the ODCM. 9.10.2.2 Cumulative dose contributions from direct radiation from the units and from radwaste storage tanks shall be determined in accordance with the methodology and parameters in the ODCM. This requirement is applicable only under conditions set forth in ACTION a. of Section 9.10.1.1 O
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ . . i
APA-ZZ-01003 R;v. 0 9.10.3 Bases 9.10.3.1 This section is provided to meet the dose limitations of 40 CFR Part 190 that have been. incorporated into 10 CFR Part 20 by 46 ER 18525. The control requires the preparation and submittal of a Special-Report whenever the calculated doses due to releases of radioactivity and the radiation from uranium fuel cycle sources exceed 25 mrema to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the reactor units and from outside storage tanks are kept , l small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contribution from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Sections 9.3.1.1 and 9.6.1.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.
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,._ 7.---_--__---___-_--. s
'APA-ZZ-01003 Rev. 0-9.11 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 141866 9.11.1 Controls 9.11.1.1 The Radiological' Environment MoSitoring Program shall be conducted as specified in Table 9.11-A.
APPLICABILITY: At all times.
- ACTION:
- a. With the Radiological Environmental Monitoring Program not being conductedLas specified in Table 9.4-A, prepare and submit to the Commission, in the Annual' Radiological Environmental Operating Report required by.
Section 7.1, a description of the reasons.for not conducting the program as required and the plans for preventing a recurrence. I.
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APA-ZZ-01003 i R:v. O ] I
- b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 9.11-A ,
when averaged over any calendar quarter, { prepare and submit to the Comminsion within 1 30 days, pursuant to Technical l Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive offluents so that the potential annual dose
- to a MEMBER OF THE PUBLIC is less than the calendar year ,
limits of Sections 9.4.1.1, 9.7.1.1, or 9.8.1.1. When more than one of the radionuclides in Table-9 11-B are detected in the sampling medium, this report shall be submitted if: concentration (1) concentration (2) reporting level (1) + reporting level (2) + ...> 1.0 When radionuclides other than those in Table 9.11-B are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose
- to A MEMBER OF THE PUBLIC from all radionuclides is equal to or greater than the calendar year limits of Sections 9.4.1.1, 9.7.1.1 or 9.8.1.1. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report, required by Section 7.1.
*The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.
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y_- _____ h APA-ZZ-01003 l R;v. O I ,
- i I
- c. With milk or fresh leafy vegetable samples
) ; unavailable from one or more of the sample locations required by the Table 9.11-A, ;
l identify specific locations for obtaining replacement samples and add them within 30 days to the Radiological Environmental Monitoring Program given in the ODCM.** The
]
specific locations from which samples were 1 unavailable may then be deleted from the monitoring program. Pursuant to Technical Specification 6.14, submit as part of, or concurrent with, the next Semiannual 1:adioactive Effluent Release Report a complete j and legible copy of the entire ODCM, including the revised figure (s) and table reflecting the new location (s) with supporting information identifying the cause of the unavailability of . samples and justifying the selection of new { ' location (s) for obtaining samples.
- d. The provisions of Sections 9.0.3 and 9.0.4 are not applicable.
41867 9.11.2 Surveillance Requirements 9.11.2.1 The radiological environmental monitoring samples shall be collected pursuant to Table 9.11-A from the specific locations given in the table and figure (s) in the ODCM, and shall be analyzed pursuant to the requirements of Table 9.11-A and the detection capabilities required by Table 9.11-c. 9.11.3 Bases 9.11.3.1 The Radiological Enviornmental Monitoring Program required by this section provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest pctsntial radiation exposures of MEMBERS OF TFi PUBLIC resulting from the station operation. This monitoring program implements Section IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the Radiological Effluent Monitoring Program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the
** Excluding short term or temporary unavailability.
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APA-ZZ-01003 R0v. O modeling.of the environmental exposure pathways. Guidance for this monitoring program is provided by.the Radiological Assessment Branch Technical Position on Environmental. Monitoring, Revision 1, November 1979. The initially specified monitoring program will be effective for at'least' the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience. , The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection.(LLDs).
' The LLDs required by Table'9.11-C are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that.the LLD is defined as a priori (before the fact) limit representing the capability of.a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.
Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually),
- Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to i Radiochemistry", Anal. Chem. 40, 586-93 (1986), 4 and Hartwell, J. K., " Atlantic Richfield Hanford j Company Report ARH-SA-215 (June 1975).
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)
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i t n M o L C A h Q . f T nc d/ d A N oa eD n t e
- E i e t o s m 1 M t cl i arr 1 N tat ie t eio
. O nts dv ea l f 9 R eeeeee rgrtrl t hto nr I
teteoe E V e e ai p - o g hip L N rves dn fe0tt B E r nrtn ov1 ct A ) ir rsus (neo T L denoeoi hr cami rn 1 A ( ew hr i C S el otggg aektds I N e rni n el n i G E O rd geh el 0ad - O V I ha r gi rd3tng L I T toeer ap oa sin O I D TDA ANC fTAO rbbr rl rorm. aisvsd ea l ertd b ooiwi tp l A oN L f idna e p 5)nm R ES sol o km mr1 eea RSEE e aoill r aa EELL BRPP l svwtaio pd a t a u m f sl nial s iwmvk l MPMM mn cn r emo el UEAA airfonfe nir0ri NRSS Skiol aI p O s g2 p m
)
_ d e u i n _ t ) e n d _ o e a _ C su Y ( tn _ AE ci WL n ut fP i dn io _ TM doo AA PS t s orc oP( _ e F _ ER g RO n . U/ I c _ SD ON _ PA X . E 4
k APA-ZZ-01003 f , Rsv. 0 p f TABLE 9.11-A (Continued) TABLE NOTATIONS 41868 (1) Specific parameters of distance and-direction sector from the centerline of one unit, and additional description where pertinent, shall be provided for.each and every sample location in Table 9.11-A in a table and figure (s) in the ODCM. Deviations are. permitted from the required p 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 effort 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 pursuant to Section 7.1. 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 specific 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 given in the ODCM. Pursuant to Technical Specification 6.14, submit as part of, or concurrent with, the next Semiannual Radioactive Effluent Release Report a' complete and legible copy of the entire ODCM including the revised figure (s) and table reflecting the new location (s) with supporting information identifying the cause of the unavailability of . samples for that pathway and justifying the selection of the new location (s) for obtaining samples. i
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L APA-ZZ LO1003L ' L. R0v. 0-p,
' TABLE 9.11-A'(Continued)
TABLE NOTATIONS
.(2) 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 analysis 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.
.(3) 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. (4) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours or more after sampling to allow for radon and thoron daughter decey. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples. (5) Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility. j (6) The " upstream sample" shall be taken at a distance beyond significant influence of the discharge. The " downstream" sample shall be taken in an area beyond but near the mixing zone.
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_ _ = - _ _ _ . _ _ _ -
E I( ' APA-ZZ-01003' Rsv. O
-TABLE 9.11-A (Continued)
TABLE NOTATIONS (7) ~In this program composite: sample aliquots shall be collected at time intervals that are'very short (e.g.,. hourly): relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample. (8) Groundwater samples shall be taken when this source is tapped for drinking or irrigation: 1 ' purposes in' areas where the hydraulic gradient cnr l' recharge properties are suitable for contamination. (9) The dose shall be calculated for the maximum organ and age ^ group, using the methodology and parameters in the ODCM. l (10) If harvest occurs more than once a year, sampling shall be performed during each discrete harvest. If harvest occurs continuously, sampling shall be monthly. Attention shall be paid to including samples of tuberous and root food products.
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'APA-ZZ-01003 R2v. O TABLE 9.11-C'(Continued)
TABLE NOTATIONS
'41869 (1) This list does not mean that only these nuclides are to be considered. Other peaks'that are identifiable, together with those of the above nuclides, shall'also be analyzed.and reported in.
. the Annual Radiological Environmental Operating
' Report pursuant to Section-7.1.
(2) Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13, Revision 1, July:1977. I 1 1
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APA-ZZ-OlOO3 Rsv. O TABLE 9.11-C (Continued) TABLE NOTATIONS (3) The LLD-is defined, for purposes of these controls, 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 only 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system, which may include radiochemical separation: gg , 4.66 s b E
- V
- 2.22
- Y
- exp (- AAt)
Where: LLD = the "a priori" lower limit of detection (microcuries per unit mass or volume), sb = the standard deviation of the background , counting rate or of the counting rate of a blank sample as appropriate.(counts per minute), E = the counting efficiency (counts per disintegration), V = the sample size (units of mass or volume), 2.22 = the number of disintegrations per minute per picocurie, Y = the fractional radiochemical yield, when applicable, 1 = the radioactive decay constant for the ptrticular radionuclides (s 2), and at = the elapsed time between the sample col'ection, or end of the sample collection peri.4, and the time of counting (s). Typical values of E, V, Y, and At should be used in the calculation. 1 1
-165- )
i t'
APA-ZZ-01003'
- l. Rsv. O I
l. 1:
- l. . TABLE 9.11-C (Cont'inued) l'L TABLE NOTATIONS It should be recognized that the LLD is defined as a a priori (before the fact) limit l representing the capability of a measurement l system and.not as an a posteriori (after the fact) limit for a particular measurement.
Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering 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 pursuant to Section 7.1. (4) LLD for drinking water samples. For surface water samples, the LLD of gamma isotopic analysis may be used.
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APA-ZZ-01003 RSv. O i 9.12 RADIOLOGICAL ENVIRONMENTAL MONITORING LAND USE CENSUS 41870 9.12.1 Controls 9.12.1.1 A Land Use Census shall be conducted and shall identify within a distance of 8 km (5 miles) the lecition in each of the 16 meteorological sectors of 5.1?e nearest milk animal, the nearest residence and the nearest garden
- of greater than 50 m 2 (500 ft8 ) producing broad leaf vegetation.
APPLICABILITY: At all times. ACTION:
- a. With a Land Use Census identifying a location (s) that yields a calculated dose or dose commitment greater than the values carrently being calculated in Section 9.8.2.1, identify the new location (s) in the next Semiannual Radioactive Effluent Release Report, pursuant to Section 7.2.
- Broad leaf vegetr.'; don sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census. Specifications for broad leaf vegetation sampling in Table 9.11-A, Part 4.c. shall be followed, including analysis of control samples.
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4 APA-ZZ-01003 Rcv. 0
- b. With a Land Use Census identifying a i location (s) that ydelds a calculated dose or I dose commitment (via the same exposure pathway) 20% greater than at a location from which samplec are currently being obtained in accordance with Section 9.11.1.1, add the new location (s) within 30 days to the Radiological Environmental Monitoring Program given in the ODCM. The sampling location (s), excluding the control station location, having the lowest calculated dose or dose commitment (s), via the same exposure pathway, may be deleted from this monitoring program after October 31 of the year in which this Land Use Census was conducted. Pursuant to Technical Specification 6.14, submit as part of, or concurrent with, the next Semiannual Radioactive Effluent Release Report a complete and legible copy of the entire ODCM, including the revised figure (s) and table (s) reflecting the new location (s) with information supporting the change in sampling locations.
- c. The provisions of Sections 9.0.3 and 9.0.4 are not applicable.
41871 9.12.2 Surveillance Requirements 9.12.2.1 The Land Use Census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities. The results of the Land Use Census shall be included in the Annual Radiological Environmental Operating Report pursuant to Section 7.1. 1
- -168-1 l
l L_ __ - _
I t
,v APA-ZZ-01003 p~ Rcv. O l' :&
-41872 9.12.3 Bases i i
9.12.3.1 This section is provided to ensure that changes in the use of areas at and beyond'the SITE
' BOUNDARY are identified and that modifications to the Radiological Environmental Monitoring Program given in the ODCM are made if required by-the results of.this census.- Information that.will provide the-best results, wuch as door-to-door survey, aerial survey, or consulting with local agricultural authorities, shall be used. This census satisfies the requirements of
.Section IV.B.3 of. Appendix.I to 10 CFR Part 50.
Restricting the census to gardens of greater than 50 m2 provides assurance'that significant exposure. pathways via leafy vegetables will be-
' identified and monitored since a garden of this
- size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: (1) 20% of the. garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m 2.
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9 APA-ZZ-01003 R;v. 0 4 I 9.13 RADIOLOGICAL ENVIRONMENTAL MONITORING INTERLABORATORY COMPARISON PROGRAM 41873 9.13.1 Controls 9.13.1.1 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission. APPLICABILITY: At all times. ACTION:
- a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual-Radiological Environmental Operating Report pursuant to Section 7.1.
- b. The provisions of Sections 9.0.3 and 9.0.4 are not applicable.
41874 9.13.2 Surveillance Requirements 9.13.2.1 The Interlaboratory Comparison Program shall be . described in this procedure. A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Section 7.1. 9.13.3 Bases, 9.13.3.1 The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.
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n . - _-_-___ e APA-ZZ-01003 Rev. O 10.0 ADMINISTRATIVE CONTROLS 10.1 MAJOR CHANGES TO LIQUID AND GASEOUS RADWASTE TREATMENT SYSTEMS
- 10.1.1 Licensee-initiated major changes to the Radwaste Treatment Systems (liquid and gaseous):
- a. Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the On-Site Review Committee (ORC). The discussion of each change shall contain:
- 1) A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59;
- 2) Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;
- 3) A detailed description of the equipment, components and processes involved and the interfaces with other plant systems;
- 4) An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the License application and amendments thereto;
- 5) An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and i to the general pcpulation that differ from those previously estimated in the License application and amendments thereto;
- 6) A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;
- Union Electric Co. may choose to submit the information called for in this specification as part of the annual FSAR update.
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APA-ZZ-01003 R0v. 0
- 7) An estimate of the exposure to plant operating personnel as a result of the change; and
- 8) Documentation of the fact that the change was reviewed and found acceptable by the ORC.
- b. Shall become effective upon review and epproval by the ORC an in accordance with Technical Specification 6.5.3.1.
10.2 CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL (ODCM) 2815 10.2.1 All changes in the ODCM shall be completed pursuant to Technical Specification 6.14.2 and approved as per APA-ZZ-00101, Preparation, Review, Approval And Control Of Procedures. 2815 10.2.1.1 All changes shall be approved by the ORC PRIOR to implementation. 10.2.2 Cross Disciplinary Review for each revision of the ODCM must include, as a minimum, Health
- Physics, Quality Assurance, and Radiological Engineering.
2815 10.2.3 A complete and legible copy of each revision of the ODCM that became effective during the last semiannual period shall be submitted as a part of, or concurrent with that periods Semiannual Radioactive Effluent Release Report pursuant to Technical Specification 6.14.2.
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l' APA-ZZ-01003 Rsv. 0
11.0 REFERENCES
11.1 Title 10, " Energy", Chapter 1, Code of Federal Regulations, Part 20; U.S. Government Printing Office, Washington, D.C. 20402. 11.2 Title 10, " Energy", Chapter 1, Code of Federal Regulations, Part 50, Appendix I; U.S. Government Printing Office, Washington, D.C. 20402. 11.3 Title 40, " Protection of Environment", Chapter 1, Code of Federal Regulations, Part 190; U.S. Government Printing Office, Washington, D.C. 20402. 11.4 U.S. Nuclear Regulatory Commission, " Technical Specifications Callaway Plant, Unit NO. 1", NUREG-1058 (Rev. 1), October 1984. 11.4.1 Section 6.8.1 (2791) 11.4.2 Section 6.8.4f (41834) 11.5 Communications 11.5.1 Letter NEO-54, D.W. Capone to S.E. Miltenberger, dated January 5, 1983; Union Electric Company correspondence. 11.5.2 Letter BLUE 1285, "Callaway Annual Average X/Q and D/Q Values", J. H. Smith (Bechtel Power Corporation), to D. W. Capone (Union Electric Co. ), dated February 27, 1984. 11.5.3 Letter BLUE 1232, "Callaway Annual Average X/Q Values and "S" Values", J. H. Smith (Bechtel Power Corporation) to D. W. Capone (Union Elecetric Co.), dated February 9, 1984. 11.5.4 Letter BLUE 1358, " Comparison of Callaway Plant Offsite Dose Calculations for Routine Effluents", J.H. Smith (Bechtel Power Corporation) to D.W. Capone (Union Electric Company), dated March 22, 1984. 11.5.5 Private Communication, H.C. Lindeman & B.F. Holderness, August 6, 1986 11.5.6 Calculation ZZ-67, " Annual Average Atmospheric , Dispersion Parameters", April 1989. j l
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_ _ _ _ _ _ _ . _ _ __ _ - - - ____ d
-e APA-ZZ-01003 Rsv.- O
-:-11.6 . Union Electric Company Callaway Plant, Unit:1, FinalLSafety Analysis Report.
11'.6;1 Section 11.5.2.2.3.1
- 11.6.2 .Section 11.5.2.2.3.4 11.6.3' LSection. 11.5.2.1.2
' ~
11.6.4 Section 11.5.2.2.3.2 11.6.5 .Section 11.5.2.2.3.3 11.6.6 'Section 11.2.3.3.4 11.6.7- Section' 11.2.3.4.3 11.6.8 Section 11.5.2.3.3.1
' 11.6.9 .Section 11.5.2.3.3.2 11.6.10 Section 11.5.2.3.2.3 11.6.11 Section 11.5.2.3.2.2 11.6.12 Section'2.3.5
-11.6.13 Section 2.3.5.2.1 2 11.6.14 Section 9.2.6 l
11.6.15 Section 9.2.7.2.1 11.6.16 Section 6.3.2.2 11.6.17 Table 11.1'6 11.6.18 Table 9.4-6 11.6.19 Table-9.4-8
' 11.6.20 Table 9.4-11 11.6.21 JTable 9.4-12 11.6.22 Table 2.3-68 1
1
-174-
p, APA-ZZ-01003
'Rev. O l
11.7 Union Electric Company Callaway Plant Environmental Report, Operating License Stage.
'11.7.1 Table'2.1-19 11.7.2 Section 2.1.2.3 11.7.3. Section 2.1.3.3.4 11.7.4 Section 5.'2.4.1 11.7.5 Table 2.1-19 11.8 HU.S. Nuclear: Regulatory Commission, " Preparation of Radiological Effluent Technical Specification For' Nuclear Power Plants", USNRC NUREG-0133, Washington,.D.C. 20555, October 1978.
11.8.1 Pages AA-1 through AA-3 11.8.2 Section 5.3.1.3 11.8.3 Section 4.3 11.8.4 Section 5.3.1.5 11.8.5 Section'5.1.1 11.8.6 Section 5.1.2
~
11.8.7 Section 5.2.1 11.8.8 Section 5.2.1.1 11.8.9 Section 5.3.1 11.8.10 Section 3.8 11.8.11 Section 3.3 71.9 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. 11.9.1 Pages 19-20 Subroutine PURGE
-175-
_ _ _ _ _ _ _ _ _ . - - - _ - . _ - - __ _ .)
APA-ZZ-01003 Rcv. 0 11.10 Regulatory Guide 1.111, " Methods For Estimating 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. 11.10.1 Section c.1.b 11.10.2 Figures 7 through 10 11.10.3 Section c.4 11.11 Regulatory Guide 1.109, " Calculation of Annual Doses to Man From Routine Releases Of Reactor Effluents For the Purpose Of Evaluating Compliance With 10 CFR Part 50, Appendix I", Revision 1, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, October 1977. 11.11.1 Appendix C, Section 3.a 11.11.2 Appendix E, Table E-15 11.11.3 Appendix C, Section 1 11.11.4 Appendix E, Table E-11 11.11.5 Appendix E, Table E-9 11.12 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. 11.12.1 Section I, Page 2 1 11.12.2 Section IV, Page 8 ' 11.12.3 Section IV, Page 9 11.12.4 Section III, Page 6 l 11.12.5 Section III, Page 8 l
-176-m________ /
k / APA-ZZ-01003 - Rsv. O i) 1 11.13 Management Agreement for the Public Use of Lands, Union Electric Company and the State of Missouri Department of Conservation, December 21, 1982. 11.13.1 Exhibit A 11.14 Miscellaneous References 11.14.1 Drawing Number M-109-0007-06, Revision 5. 11.14.2 Callaway Plant Annual Environmental Operating Report (updated annually). 11.14.3 UE Safety Analysis Calculation 87-001-00. 11.14.4 Calculation ZZ-48, " Calculation of Inhalation and Ingestion Dose Commitment Factors for the. Adult and Child", January, 1988. 11.14.5 HPCI 89-02, " Calculation of ODCM Dose Commitment Factors", March, 1989. 11.14.6 HPCI 87-04, " Calculation of the Limiting Setpoint for the Containment Purge Exhaust Monitors, GT-RE-22 and GT-RE-33", March, 1987. 11.14.7 HPCI 58-10, " Methodology for Calculating the Response of Gross NaI(TA) Monitors to Liquid Effluent Streams", June, 1988. 11.14.8 Calculation'ZZ-57, " Dose Factors for Eu-154", January, 1989. 11.15 U.S. Nuclear Regulatory Commission, "XOQDOQ: Computer Program for the Meterological Evaluation of Routine Effluent Releases at Nuclear Power Stations", USNRC NUREG/CR-2929, September, 1982, Washington, D.C. 20555. 11.15.1 Section 4, " Subroutine PURGE", pages 27 and 28. 11.16 Regulatory Guide 4.13, " Performance, Testing, and procedural specifications for Thermoluminescence Dosimetry: Environmental Applications
"(Revision 1), July 1977; USNRC, Washington, D.C. 20555 11.17 TID-7004, " Reactor Shielding Design Manual",
Rockwell, Theodore, ed; March 1956.
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W-APA-ZZ-01003 ih Rav. O 1 11.18 BNWL-236, "ISOSHLD - A' computer code for General Purpose Isotope Shielding Analysis", Engel, R.C., Greenberg, J., Hendrichson, M.M.; June 1966. 11.19 BNWL-236, Supplement 1, "ISOSHLD-II: Code Revision.to include calculation of Dose Rate from Shielded Bremsstrahlung Sources", Simmons, G.L., et al; March 1967. 11.20 BNWL-236, Supplement 2, "A Revised Photon Probability Library for use with ISOSHLD-III", Mansius, C.A.; April 1969. 11.21 ANSI N13.10-1974, " Specification & Performance of On-Site Instrumentation for Continuously Monitoring Radioactivity in Effluents"; September, 1974. 11.22 Nuclear Regulatory Commission Generic Letter 89-01, " Guidance for the Implementation of Programmatic Controls for RETS in.the Administrative Controls Section of Technical Specifications and the Relocation of Procedural Details of Current RETS to the Offsite Dose Calculation Manual or Process Con'. col Program", January 1989. 11.23 ODP-ZZ-00002, Equipment Status control. l l 1
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