ML20108E371

Rev 5 to Administrative Procedure APA-ZZ-01003, Odcm

ML20108E371
Person / Time
Site:	Callaway
Issue date:	02/16/1996
From:	UNION ELECTRIC CO.
To:
Shared Package
ML20108E364	List: ML20108E363 ML20108E366 ML20108E371
References
APA-ZZ-01003, APA-ZZ-1003, NUDOCS 9605100210
Download: ML20108E371 (112)

Text

bNION Etzcraic n Revision 5 m

CALLAWAY PLANT OFF-SITE DOSE CALCULATION MANUAL February,1995 l

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APA-ZZ-01003 Reviston 5 February 1,1995 NUCLEAR FUNCTION ADMINISTRATIVE PROCEDURE APA-ZZA1003 OFT-SITE DOSE CALCULATION MANUAL RESPONSIBLE DEPARTMENT f/hCru 6464 C T PREPARED BY bbNb ~

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i DATE ISSUED -2 l-W e q\b This procedure contains the fo!!owing:

Pages I through 85 Attachments I through 2 Tables through Figures I through 1 Appendices A through A CheckofrLists through i

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APA.ZZ.01003 i Rev. 5 i

TABLE OF CONTERTS Section Pace Number l

PURPOSE AND SCOPE. __ . .. . 1 LIQUID EFFLUENTS . - -......2 j Liquid Emuent Monitors . .... .... . .. .. . .. .. 2 Calculation Of Liquid EfIluent Monitor Setpoints.. . . . . . . . . .3 Liquid EfDuent Concentration Measurements.. .. . .. . . . a.. . . . . . . . .5 i Dose Due To Liquid Emuents... . .. ... . . .... . . . . . . . . . ... . . . . . ..5 l The Maximum Exposed Individual.. . . . . . . . . . . . . .5 Calculation Of Dose From Liquid EfDuents . . . . . . . . . . . . ... . . . . . ..6 ,

Summary, Calculation Of Dose Due To Liquid EfDuents : . .... .. . 7  !

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Liquid Radwaste Treatment System.. . . . . . .7 l Dose Factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .7 ..

GASEOUS EFFLUENTS _ . . . 11 Gaseous Emuent Monitors.. . . . . . . . . . .. . . . . . .11 1 Gaseous EfDuent Monitor Serpoints.. ... . . . . . 12 1 Total Body Dose Rate Serpoint Calculations .... .. . .. . . . . . 13 Skin Dose Rate Setpoint Calculation.. .. . . .. .. . . . . . . . . 14 Calculation Of Dose And Dose Rate From Gaseous EfIluents.. . . .14

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, NOBtt GASES .........................................................a RADIONUCLIDES OTliER THAN NOBLE GASES . . . .... . . ............15 DOSE DUE TO GASEOUS EFFLUENTS - _ ~. . . . 16 NOBLE GASES .................._..16 RAD 1ONUCLIDES OTIIER TIIAN NOBLE GASES. ... .... .....................................16 Gaseous Radwaste Treatment System.. . ..' . .17 D O S E FA CTO RS ....... .......... .. ..... ..... ...... ..... . ... ... .... ... .... ....... ......... .. I 7 DOSE AND DOSE COMMITMENT FROM URANIUM FUEL CYCLE SOURCES ... ......... 41 Calculation Of Dose And Dosc Commitment From Uranium Fuel Cycle Sources.. . .41 ID ENTIFICATION OF TIIE M EM BER O F TIIE PUBLIC...... .. ...... . ..................... ..... 41 TO TAL D O S E TO Til E N E AREST RES ID E NT .. . ........ . ..... ... ........... .. . ......... .... .. . 41 TOTAL DOSE TO TIIE CRITICAL RECEPTOR WITHIN TIIE SITE BOUNDARY.... ....... 42 RADIOLOGICAL ENVIRONMENTAL MONITORING. .~......... . . . . . . . . . . . . . 45 Description Of The Radiological Environmental Monitoring Program.. .45 Performance Testing Of Environmental Thermoluminescence Dosimeters.. .45 DETERMINATION OF ANNUAL AVERAGE AND SilORT TERM ATMOSPIIERIC D IS P E RS I O N P A RA M ETE RS ................ ... ... ........... . .. ..... .. . . . . . .. 46 Atmospheric Dispersion Parameters.. . .46

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APA.ZZ.01003 Rev. 5 TABLE OF CONTENTS lI l hp.cIl Pace Number

. . . . .. .. . .. . ........... 46 LONG-TERM DISPERSION ES$1ATES i DETERMINATION OF LONG-TERM DISPERSION ESTIMATES FOR SPECIAL

... 46 RECEPTOR LOCATIONS. ..

SIIORT-TERM DISPERSION ESTIMATES .....................46 Annual Meteorological Data Processing.. ..t.. . . .47 REPORTING REQUIREMENTS. . _.. . . .. ... . 5 3 l

. 53 =

Annual Radiological Environmental Operating Report (CTSN 2804) . . .. . .

Annual Radioactive Emuent Release Report (CT. SN 2805) . . 53 IMPLEMENTATION OF ODCM METHODOLOGY (CTSN 2791) _ ._ . . .. 55 RADIOACTIVE EFFLUENT CONTROLS (REC) .. . . .... .. ... ..... 56 Radioactive Liquid Emuetn Monitoring Instrumentation.. . . .. . 57 Radioactive Gaseous Emuent Monitoring Instrumentation.. . .57 Liquid Emuents Concentration.. .. . . 58 g

Dose From Liquid Emuents.. . .. . . .61 Liquid Radwaste Treatment System. . . . . . .62 5 Gascous Emuents Dose Rate.. . . . . .63

-) Dose . Noble Gases.. . .. .

Dose -Iodinc.131 And 133. Tritium. And Radioactive Material In Particulate Form .

. .66

.67

.68 Gaseous Radwaste Treatment System.. . .. . . . .

TotalDose ... . . . . . .. . . . . . . . . . . . . . . .. . . . .-69 Radiological Environmental Monitoring Program - . . .. . .70 l Radiological Emironmental Monitoring Land Use Census . .. .79 m Radiological Emironmental Monitoring Interlaboratory Comparison Program.. . 80 AD MINISTRATIVE CONTROLS . ....... ... .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 81 Major Changes To Liquid And Gascous Radwaste Treatment Systems.. . 81 Changes To The Offsite Dose Calculation Manual (ODCM) (CTSN 2815). . 81 R E FE RE N C E S .. .... .... . ......... .. . .. ..... . . .... ..... ... .. ... .... .... . ... .. .......... .. .. 8 2 E, CURES g

FIG URE 4.1... 44 I

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APA.ZZ.01003 Rev. 5 l l

1 TABLE OF CONTENTS 1

1 Section Pace Number l i

TABLES '

1 1

TABLE 2.1 _. .. .. .. 8 l TABLE 2.2 .. . ... . . 10 l TABLE 3.1- . _ _ . . _ - . _ 18 TABLE 3.2 -

._ .. . . 19 TABLE 3.3 -INHALATION PATHWAY ~ _ .

. 21 TABLE 3.3 - MEAT PATHWAY . 23 TABLE 3.3 - GRASS-COW-MILK PATHWAY . . . 25 .

l TABLE 3.3 - GRASS-GOAT-MILK PATHWAY -. .. .. . .. . 2 7 l TABLE 3.3 - VEGETATION PATHWAY -. - ._ ~. . 29 TABLE 3.4 -INHALATION PATHWAY _ .._ . . 31 TABLE 3.4 - MEAT PATHWAY .. . ~ . .

33 TABLE 3.4 - GRASS-COW-MILK PATHWAY . . 35 I TABLE 3.4 - GRASS-GOAT-MILK PATHWAY . 37 TABLE 3.4 - VEGETATION PATHWAY. . . ... . . . . . . . 39 TABLE 6.1..............................................~......~.........................48 TABLE 6.2...............................................................................49 TAB LE 6. 3 ....... ....... ....... . . . . .... ... . .... ... ... .... ... 5 0 TABLE 6.4...................................................................51 TABLE 6.5.......................................................................52 TABLE 9.3-A.......................................................................59

) TABLE 9.3-A....................................~........................................................59 i

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TAB LE 9.6.A .. . . ... .. . ... .. . . ..... . 64 TABLE 9.11-A ... . . .. 7 2 TAB LE 9.11-B ... ... .. . . ... . . . . .... 76 TABLE 9.11-C .. .. ... . . ... 77 Attachment 1 -

Lower Limit of Detection (LLD). . .. .. .1 Page Attachment 2 - Bases for Radiological Efiluent Controls. - .,7 Pages Appendix A -

Summary Review Of Radiological EfIluent Tech Spec Potentially Affected By The Implementation Of The Resised 10 CFR 20., .12 Pages t

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I APA ZZ-01003 Rev. 5 RECORD OF REVISIONS Rev.No. O Date: March 1983 1

Rev. No. Date: November,198 1

EI Revised to support the current RETS submittal and to incorporate NRC Staff comments. Ei Rev. No. 2 Date: March,1984 Revised to incorporate NRC StafTcomments Rev. No. 3 Date: June,1985 Revised to incorporate errata identified by ULNRC-803 and changes to the Environmental Monitoring Program. Incorporate results of 1984 L.and Use Census.

Rev. No. 4 Date: February,1987 l'

M Minor clarifications, incorporated 31 day projected dose methodology. Change in the utilization of areas within the Site Boundary.

Rev.No. 5 Date: January,1988 Minor clarifications, revised descriptions ofliquid and gaseous rad monitors, revised liquid setpoint methodology to incorporate monitor background, revised dose calculations for 40CFR190 requirements, Revised g Table 6 and Figures 5.l A and 5.lB to refine descriptions of environmental TLD stations, incorporated description g of emironmental 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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Rev.No. 6 Date: May,1989 5 Revised methodology for calculating maximum permissible liquid emuent discharge rates and liquid emuent discharge rates and liquid emuent monitor setpoints, provided methodology for calculating liquid emuent l monitors response correction factors, provided an enhanced description of controls on liquid monitor background E limits, provided additional liquid and gaseous dose conversion factors and bioaccumulation factors (Tables 1. 2. 4

& 5), provided description of the use of the serpoint required by Technical Specification 4.9.4.2 during Core g Alterations, added discussion of gaseous and 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 5

parameters and accompanying description per FSAR Change Notice 88-42.

APA-Z.Z41003 Rev. No. O Date: August,1989 Radiological Emuent Technical Specifications were moved from the Callaway Plant Tect nical Specifications to Section 9.0, Radioactive Emuent Controls, of the ODCM as per NRC Generic Letter 89-01. At l

5 the same time, in order to formalize control of the entire ODCM, it was converted to APA-ZZ-01003, OFF SITE DOSE CALCULATION MANUAL.

Rev.No. 1 Date: October,1990 Revise Action 41 of Table 9.2-A to allow continued purging for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as per Amendment 20 to operating license, issued 4/10/87.

Rev.No. 2 Date: May,1991 Section 2.4.2 - Changed gross alpha analysis frequency from "cach batch" to a monthly composite as per l Table 93-A, and the Callaway Plant NPDES permit (reissued March 15, 1991). W I

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APA-ZZ-01003

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RECORD OF REVISIONS Rev.No. 3 Date: June,1993 Deleted HF-RE-45 and LE-RE 59 as effluent monitors. Revised table numbering for consistency with those in Section 9.0, deleted redundant material, incorporated 1992 Land Use Census results, moved LLD description to Attachment 1, moved REC Bases to Attachment 2. Deleted reporting requirements for solid radwaste, which are described in APA-ZZ-01011, PROCESS CONTROL PROGRAM. Addressed compliance with 10 CFR 20.1301. Revised the dilution flow rate to allow values other than 5000 gpm, based on dilution flow monitor setpoint, Revised "MPC" terminology to "ECV". Added Action 46 to REC 9.2 to clarify actions for inoperable mid and high range WRGM Channels. Revised references to be consistent with the revised 10 CFR 20.

Added Appendix A. Revised Action 41 of Rec 9.2 and the operability requirements of GT-RE 22/33. Incorporated the revised Ri values in Tables 3.2 and 3.3. Added Section 6.2 and Table 6.5.

Rev. No. 4 Date: September,1994 Increased the minimum channels OPERABLE requirement of REC 9.2 for GT-RE-22 & 23 from I channel to 2 channels. Revised Action 41 and the Bases for REC 9.2 accordingly. Incorporated the operability requirements from Tcch Spec 3.9.9 into the Action statement for clarity. (Refer to SOS 94-1176).

Rev.No. 5 Date: February,1995

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Removed the REMP station locations. Removed particulate nuclides with a half-life ofless than 8 days from Tables 3.2-3.4 and removed C", P'2, Ni, Tc*'", and from Tables 2.1,2.2,3.2,3.3, and 3.4. Changed the reporting frequency of the Semiannual Efiluent Release Report from semiannual to annual. Removed the meat.

milk and vegetable pathway dispersion parameters from Tables 6.1,6.2, and 6.3, and clarified the applicability of

, the dispersion parameters end dose locations in Table 6.4. Relocated REC 9.1 and 9.2 to the FSAR. Resised footnotes 3 and 7 of Tab!c 9.6 A to require additional sampling of the Unit Vent in the event of a reactor power transient, only if the Unit Vent noble gas activity increases by a factor of 3 or greater. Added Section 4.1.3.1.3 for

) determination of dose due to the on-site storage of low level radioactive waste.

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APA ZZ41003 I

Rev. 5 E g,

OFF-SITE DOSE CALCULATION M ANUAL

1. E.URPOSE AND SCOPE The OFF SITE DOSE CALCULATION MANUAL (ODCM) describes the methodology and parameters used in the calculation of off site doses resulting from radioactive gaseous and liquid g emuents, in the calculation of gaseous and liquid emuent monitoring Alarm / Trip Setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM also contains the g

Radios:tive Emuent Controls and Radiological Environmental Monitoring Program required by Technical Specification 6.8.4, and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Emuent Release Reports l

un required by Technical Specifications 6.9.1.6 and 6.9.1.7.

Compliance with Radiological Emuent Controls limits demonstrates compliance with the limits of 10 CFR 20.1301. (Ref. I1.1.1,11.2.1,11.23.3) (CTSN 4121)

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APA-ZZ-01003 Rev. $

2. LIOUID EFFLUENTS I

2.1 LIOUID EFFLUENT MONITORS Gross radioactivity monitors which provide for automatic termination ofliquid emuent releases are present on the liquid emuent lines. Flow rate measurement devices are present on the liquid emuent lines and the discharge line (cooling tower blowdown). Setpoints, precautions, and limitations applicable to the operation of the Callaway Plant liquid emuent monitors are provided in the appropriate Plant Procedures. Setpoint values are calculated to assure that alarm and trip actions occur prior to exceedmg the Emuent Concentration Values (ECV) limits in 10 CFR Pan 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 sC (2.1)

F+f Where:

C = The liquid effluent concentration value (ECV) implementing REC i.3.1.1 for the site in

( Ci/ml).

e = The setpoint, in ( Ci/ml), of the radioactivity monitor measuring the radioactivity concentration in the emuent line prior to dilution and subsequent release; the setpoint, which is inversely related to the volumetric flow of the emuent li se and directly related l

to the volumetric flow of the dilution stream plus the emuent ster m, represents a value,  ;

which, if exceeded, would result in concentrations exceeding the values of 10 CFR Part 20 Appendix B. Table II, Column 2, in the UNRESTRICTED AREA.

f = The flow setpoint as measured at the radiation monitor location, in volume per unit time, but in the same units as F, below.  !

F = The dilution water flow rate setpoint as measured prior to the release point, in volume per unit time. {1f(F)is large compared to (f), then F + f a F).

(Ref.11.8.1) l If no dilution is provided, then c < 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 blowdonti, but should this become unavailable, the plant water treatment facility supplies the necessaq dilution flow via a bypass line. The limiting concentration which corresponds to the liquid radwaste efIluent monitor setpoint is to be calculated using methodology from the expression above.

Thus, the expression for determining the setpoint of the liquid radwaste emuent line monitor becomes:

Cs C(Ff + f) ( Ci/ml) (2.2) I The alarm / trip setpoint calculations are based on the minimum dilution flow rate (corresponding to the dilution flow rate setpoint), the maximum emuent stream flow rate, and the actual isotopic anal) sis. 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 cach pathway. The determination and usage of the allocation factor is discussed in Section 2.2. In the event the alarnVtrip setpoint is reached, an evaluation will be performed using actual dilution and emuent flow values and actual isotopic

analysis to ensure that REC 9.3.1.1 timits were not exceeded l

l APA-ZZ-01003 Rev. 5 I

2.1.1 Continuous Liauid Emoent Monitors The radiation detection monitor associated with continuous liquid emuent relmses is (Ref. !!.6.1, g i 1.6.2): g Monitor i D. Description BM-RE 52 Steam Generator Blowdown Discharge l Monitor 5 The Steam Generator Blowdown discharge is not considered to be radioactive unless radioactivity has been detected by the associated eduent radiation monitor or by laboratory analysis. The sampling frequency, minimum analysis frequency, and type of analysis performed are as per Table 9.3 A.

2.1.2 Radioactive Liauid Batch Release Emuent Monitors The radiation monitor which is associated with the liquid emuent batch release system is (Ref.

I1.6.4):

Monitor I D. Descriotion HB RE 18 Liquid Radwaste Discharge Monitor This emuent stream is normally considered to be radioactive. The sampling frequency, minimum analysis frequency, and the type of analysis performed are as per Table 9.3-A.

2.2 CALCULATION OF LIOUID EFFLUEKr MONITOR SETPOINTS The dependence of the setpoint (c), on the radionuclide distribution, yields, calibration, and monitor parameters, requires that several vanables be considered in setpoint calculations. (Ref. I 1.8.1)

. 2.2.1 Calculation of the ECV Sum The isotopic concentration of the release (s) being considered must be determined. This is obtained from the analyses ruluired per Table 9.3-A, and is used to calculate an ECV sum (ECVSUM):

ECVSUM = [(C,)/(ECV,))

i = g, e , s, t, f Where:

Cg = the concentration of each measured gamma emitting nuclide observed by g gamma-ray spectroscopy of the waste sample. g C*

a

= the rneasured concentration of alpha emitting nuclides measured by gross alpha analysis of the monthly composite sample, g Cs = the measured concentrations of Sr-89 and Sr-90 as determined by analysis 5 of the quarterly composite sample.

Ct = the measured concentration of H-3 in the waste sample.

Cr*

= the measured concentration of Fe-55 as determined by analysis of the quarterly composite sample.

ECVg, ECV ,3 ECV ,aECVr, ECV =t are the limiting concentrations of the appropriate radionuclides from 10 CFR 20, Appendix B. Table II, Column 2. For dissolved or entrained noble gases, the 4 Ci/ml total activity.

I concentration shall be limited to 2s10 I

• Values fur these concentrations are tased on previous composite sample analyses as requued by Table 9 3.A

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1-I APA ZZ-01003 Rev. 5 L For the case ECVSUM 51, the monitor tank emuent concentration meets the limits of REC 9.3,l.1 without dilution and the effluent may be released at any desired flow rate. If ECVSUM > 1 then dilution is required to ensure compliance with REC 9.3.1.1 concentration limits. If simultaneous ,

releases are occurring 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 I for a particular discharge point, provided that the sum total allocation fractions for all discharge points must be 51, 2.2.2 Calculation of the Maximum Permissible Liauid Emuent Discharce FlowTgg j_ The manmum permissible liquid emuent discharge flowrate is calculated by:

fmax s (F + fp)(SF)(N) + (ECVSUM) (2.4)

Where:

fmax = maximum permissible liquid emoent discharge flowrate, in l-(gallons / minute) fp = the expected undiluted liquid emuent flowTate, in gpm.

L N = the allocation fraction which apportions dilution flow among simultaneous discharge pathways (see discussion above)

SF = the safety factor; an administrative factor used to compensate for statistical fluctuations and errors of measurements. This factor also l provides a margin of safety in the calculation of the maximum liquid effluent discharge flowrate (fmax). The value of SF should be51.

F & ECVSUM, are previously defined.'

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The dilution water supply is furnished with a flow monitor which isolates the Jiquid efIluent discharge if the dilution flow rate falls below its setpoint value.

In the event that m f ax is less than pf , then the valu' c of fmax is substituted into the equation for fp. )

l and a new value of fmax is calculated. This substitution is performed for three iterations in order to l

calculate the correct value of fmax-2.2.3 Calculation Of Liauid Emuent Monitor Sett,oint l The liquid emuent monitors are Nal(TI) based systems and respond primarily to gamma radiation.

Accordingly, their setpoint is based on the total concentration of gamma emitting nuclides in the effluent:

c = BKG + (I(C g) + SF) = CUml (2.5)

Where:

e = the monitor setpoint as previously defined, in (pCUml),

BKG = the monitor background prior to discharge, in ( Cvml),

C and SF are as previously defined. 1 E

The monitor's background is controlled at an appropriate limit to ensure adequate sensitivity.

Utilizing the methodology of ANSI N13.101974 (Ref. I1.21), the background must be maintained I at a value orless than or equal to 2.23E-6 cum! (rclative to Cs-137) in order to detcet a change of l l I E-7 CUml ofI-134 (the most restrictive nuclide in Table 1 of reference 11.21).

! In the event that there is no detectable gamma activity in the emuent or if the value of(E(Cg) + SF) is less than the background of the monitor, then the monitor setpoint will be set at twice the current l background of the monitor.

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APA Z?.-01003 Rev. 5 As previously stated, the monitor's response is dependent on the gamma enutting radionuclide distribution of the emuent. Accordingly, a new database conversion factor is calculated for each release based upon the results of the gamma spectrometric analysis of the emuent sample and the 3 measured response of the monitor to National Institute of Standards and Technology (NIST) g traceable calibration sources:

DBCF e, = C + (CMR) x(ECF) (2.6)

DBCFe = the monitor data base conversion factor which converts count rate into concentration (pCi/ml);

g CMR = the calculated response of the radiation monitor to the liquid emuent; E ECF = the conversion factor for Cs 137, which converts count rate into concentration ( Ci/ml).

Cgis as previously defined.

The new value of the DBCFcis calculated and entered into the monitor data base prior to each g discharge. A more complete discussion of the derivation and calculation of the CMR is given in g reference 11.14.7.

2.3 LIOUID EFFLUENT CONCENTRATION MEASUREMENTS Liquid batch releases are discharged as a discrete volume and each release is authorized based upon the sample analysis and the d21ution flow rate existing in the discharge line at the 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 liquid emuent discharge is monitored prior to entering the diletion discharge line and will automatically be terminated if the pre-selected "

alarm / trip setpoint is exceeded. Concentrations are determined primarily from the gamma isotopic and H 3 analyses of the liquid batch sample. For gross alpha, St-89, Sr-90, & Fe 55, the measured g concentration from the previous composite analysis is used. Composite samples are collected for each 3 batch release. Monthly analysis for gross cipha and quarterly analyses for St-89, Sr-90. and Fe 55 are performed in accordance with Table 9.3 A. Doses from liquids discharged as continuous releases are calculated by utilizing the last measured values of samples reached in accordance with Tabic 9.3-A.

2.4 DOSE DUE TO LIOUID EFFLUENTS 2.4.1 The Maximum Emosed 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. I1.8.3) l m

The Callaway Plant's liquid emuents are discharged to the Missouri River. As there are no potable water intakes within 50 miles of the discharge point (Ref. I1.7.1,11.6.6), this pathway does not l require routine evaluation Therefore, the dose contribution from fish consumption is expected to 5 account for more than 95% of the total man-rem dose from discharges to the Missouri River. Dose from recreational ahities is expected to contribute the additional 5% which is considered to be negligible. (Ref. I1.6.7)

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APA-ZZ-01003 I: Rev. S 2.4.2 Calculation Of Dose From Liouid Emuents l The dose contributions for the total ame period.

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[At, 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, Dose is calculated for all radionuclides identified in liquid emuents released to UNRESTRICTED AREAS using the following expression (Ref. !!.8.3)!

D, = A,3 At, C, F, (2.12)

,. r.,

Where:  ;

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D, = the cumulative dose commitment to the total body or any organ, t, from l' the liquid emuents for the total period i

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[At, ts,

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in mrem. j

=

At, the length of the t th time period over which Cg, and F, are averaged j for all liquid releases, in hours. 4t, corresponds to the actual duration of

) the release (s).

= the average measured concentration of radionuclide, i, in undiluted liquid Cg, emuent during time period At, from any liquid release, in ( Ci/ml).

l Aj, =

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 L Table 9.3 A,(in mrem /hr) per (pCi/m!). The calculation of the Aj, values is detailed in Ref,11.14.5 and are given in Table 2.1.

F, =

the near field average dilution factor for C;, during any liquid emuent release:

l F, = (F + f_) 89.77 Where:

f = maximum undiluted emuent flow rate during the release max F = average dilution flow 89.77 =

site specific applicable factor for the mixing effect of the discharge structure. (Ref. I1.5.1)

The term Ci , is the undiluted concentration of radioactisc material in liquid waste at the conunon release point determined in accordance with REC 9.3.1.1, Table 9.3 A, " Radioactive Liquid Waste Sampling and Analysis Program". All dilution factors beyond the sample point (s) are included in the F, term.

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APA ZZ-01003 Rev. 5 The nearest murdeipal potable water intake downstream from the liquid efIluent discharge point into l the Missouri River is located near the city of St. Louis, Missouri, approximately 78 miles dowitstream. As there are currendy no potable water intakes within 50 river miles of the discharge g point, the drinking water pathway is not included in dose estimates to the maximally exposed gl individual, or in dose estimates to the population. Should future potable water intakes be constructed 1 within 10 river miles downstream of the discharge point, then this manual will be revised to include I this pathway in dose estimates. (Ref. I1.6.6).

2.4.3 Summary. Calculation Of Dose Due To Liould Efiluents The dose contribution for the total time period m

ble tai is determined by calculation at Icast once per 31 days and a cumulative summation of the total body E and organ doses is maintained for each calendar quarter. The projected dose contribution from liquid effluents for which radionuclide concentrations are determined by periodic composite and grab g sample analysis, may be approximated by using the last measured value. Dose contributions are g determined for all radionuclides identified in liquid eHluents 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 D:tection l W

(LLD) level for that nuclide. The "less than" values are not used in the dose calculations.

2.5 LIOUID RADWASTE TREATMENT SYSTEM The LIQUID RADWASTE TREATMENT SYSTEM is capable of varying treatment, depending on waste type and product desired. It is capable of concentrating, gas stripping, and distillation ofliquid

) wastes through the use of the evaporator system. The demineralization system is capable of removing radioactive ions from solutions to be reused as makeup water. Filtration is performed on certain liquid wastes and it may, in some cases, be the only required treatment prior to release. The system l

has the ability to absort 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 w hen liquids require treatment prior to their release to the environment.

OPERABILITY is demonstrated through compliance with REC 9.3.1.1. and 9.4.1.1.

Projected doses due to liquid releases to UNRESTRICTED APEAS are detemuned each 31 days by dividing the cumulative annual total by the number of elapsed months.

2.6 DOSE FACTORS E

The dose conversion factors provided in Table 2.1 were derived from the appropriate dose conversion E factors of Regulatory Guide 1.109 and other sources as necessary (Ref: 11.14.5 and 11.14.12)

Non-gamma emitting nuclides not listed in Table 9.3-A are not considered.

l l

I I

7 l

l

APA ZZ-01003 Rev. 5 1

TABLE 2.1 i

INGESTION DOSE COMMITMENT FACTOR (Ajs) FOR ADULT AGE GROUP l

a (mrem /hr)per( Ci/ml) j i

Nuclides Bone Liver Total Thyroid Kidney Lung GILLI bL p m m m - w w w m m u szu h ,.m m m e s g e s e dggq? @'t* q q p gg g I H3 No Data 2.26E-01 2.26E-01 2.26E 01 2.26E-01 2.26E-01 2.26E-01 l

- Be 7 1.30E-02 2.98E-02 1.45E-02 No Data 3.15E 02 No Data 5.16E+00 )

I 1 Na 24 4.07E+02 4.07E+02 4.07E+02 4.07E+02 4.07E+02 4.07E+02 4.07E+02 I

Cr 51 No Data No Data 1.27E+00 7.62E-01 2.81E 01 1.69E+00 3.20E+02 -

Mn-54 No Data 4.38E+03 8.35EM2 No Data 1.30E+03 No Data 1.34 E+04

)

4 Mn 56 No Data 1.10E+02 1.95E+01 No Data 1.40E+02 No Data 3.52E+03 Fe-55 6.57E+02 4.54E+02 1.06E+02 No Data No Data 2.53E+02 2.61E+02 l Fe-59 1.04E+03 2.44E+03 9.34E+02 No Data No Data 6.81E+02 8.13E+03 l 'l l Co 57 No Data

• 2.09E@l 3.48E+01 No Data No Data No Data 5.31E+02 Co-58 No Data 8.94E+01 2.00E+02 No Data No Data No Data 1.81E+03 Co-60 No Data 2.57E+02 5.66E+02 No Data No Data No Data 4.82E+03 l

! Ni-65 1.26EM2 1.64E+01 7.48E+00 No Data No Data No Data 4.16E+02 Cu-64 No Data 1.00E+01 4.69E@0 No Data 2.52E+0! No Data 8 52E+02 l Zn-65 2.32E+04 7.38E@4 3.33E+04 No Data 4.93E+04 No Data 4.65E+04 Zn-69 4.93 E+01 9.44E+01 6.56E+00 No Data 6.13E+01 No Data 1.42E+0 i l Br 82 No Data No Data 2.27E+03 No Data No Data No Data 2.60E+03  !

l Br 83 No Data No Data 4.04E401 No Data No Data No Data 5.81E+01 Br-84 No Data No Data '5.26E+01 No Data No Data No Data 4.13E-04 Br 85 No Data No Data 2.15E+00 No Data No Data No Data 0 l Rb. 86 No Data 1.01 E+05 4.71 E+04 No Data No Data No Data 1.99E+04 i

Rb-88 No Data 2.90EM2 1.54E+02 No Data No Data No Data 4.00E-09 l Rb-89 No Data 1.92E+02 1.35E+02 No Data No Data No Data o j i

St-89 2.21E+04 No Data 6.35E+02 No Data No Data No Data 3.55E+03 Sr 90 5.44E+05 No Data 1.34E+05 No Data No Data No Data 1.57E+04 Sr 91 4.07E+02 No Data 1.64 E+01 No Data No Data No Data 1.94 E+03

! St-92 1.54E+02 No Data 6.68E+00 No Data No Data No Data 3.06E+03

Y 90 5.75E-01 No Data 1.54E-02 No Data No Data No Data 6.10E+03 l l l Y 91M 5.44E-03 No Data 2.10E-04 No Data No Data No Data 1.60E-02 l Y 91 8.43 E+00 No Data 2.25E 01 No Data No Data No Data 4.64E+03 l

! Y 92 5.0$E-02 No Data 1.48E-03 No Data No Data No Data 8.85E+02 Y 93 1.60E-01 No Data 4.42E-03 No Data No Data No Data 5.08E+03 Zr-95 2.40E-01 7.70E-02 5.21E-02 No Data 1.21E-01 No Data 2.44 E+02 Zr 97 1.33 E-02 2.68E-03 1.22E-03 No Data 4 04E-03 No Data 8 30E+02 s Nb-95 4.47E+02 2.48E+02 1.34 E+02 No Data 2.46E+02 No Data 1.51Et06

.! Mo 99 No Data 1.03 E+02 1.96E+0 ! No Data 2.33E+02 No Data 2.39E+02 Tc-99M 8 87E-03 2.51E-02 3.19E-Ol No Data 3 81E-01 1.23 E-02 148E+01 Tc-101 9. l lE-03 1.31 E-02 1.29E-01 No Data 236E-01 6.70E-03 0 i

8-l l _ _. . _ _ _ _ _ . _

APA-ZZ-01003 Rev. 5 TABLE 2.1 (Cont'd)

I INGESTION DOSE COMMITMENT FACTOR (Ajs) FOR ADULT AGE GROUP (mrem /hr)per (pCi/ml)

Nuclide Bone Liver Total Body Thyroid Kindey Lung GILLI I y n

- < y.9  : mgw.gy;y ~ qu.sM;";7"*TW x. p u y ;a mwp Ru 103 4.42E+00 No Data 1.90E+00 No Data 1.69E+01 No Data 5.17E+02 Ru-105 3.68E 01 No Data 1.45E-01 No Data 4.76E+00 No Data 2.25E+02 g Ru-106 Cd 109 6.57E+01 No Data No Data 5.54E+02 8.32E+00 1.94E+01 No Data No Data 1.27E+02 5.3IE+02 No Data No Data 4.25E+03 5.59E+03 g

Sn ll3 5.66E+04 1.61E+03 3.26E+03 9.18E+02 No Data No Data 1.69E+05 Sb-124 6.69E+00 1.26E-01 2.65E+00 1.62E 02 No Data 5.21E+00 1.90E+02 Sb-125 4.28E+00 4.78E-02 1.02E+00 4.35E-03 No Data 3.30EMO 4.71E+01 l Te 127m 6.47E+03 2.32E+03 7.90E+02 1.66E+03 2.63E+04 No Data 2.174E+04 g Te-127 1.05E+02 3.78E+01 2.28E+01 7.80E+01 4.29E+02 No Data 8.30E403 g Te-129M . 1.10E+04 4. l lE+03 1.74E+03 3.78E+03 4.60E+04 No Data 5.54E+04 Te 129 1.13E+01 7.33E+00 2.31E+01 1.26E+02 No Data 2.27E+01 Te-131M 3.01E+0) 1.66E+03 8.09E+02 6.75E+02 1.28E+03 8.21E+03 No Data 8.03 E+04 l

m Te-131 1.89E+01 7.88E+00 5.96E+00 1.55E+01 8.25E+01 No Data 2.67E+00 Te-132 2.41E+03 1.56E+03 1.47E+03 1.72E+03 1.50E+04 No Data 7.38E+04

) 1130 2.71E+01 8.01E+01 3.16E+01 6.79E+03 1.25E+02 No Data 6.89E+01 1131 1.49E+02 2.14E+02 1.22E+02 7.00E+04 3.66E+02 No Data 5.64E+01 1 132 7.29E+00 1.95E+0! 6.82E+00 6.82E+02 3.llE+01 No Data 3.66E+00 I-133 5.10E+01 8.87E+01 2.70E+01 1.30E+04 1.55E+02 No Data 7.97E+0!

l

=

I134 3.81E+00 1.03E+01 3.70E+00 1.79E402 1.64E+01 No Data 9.01E-03 1135 1.59E+01 4.16E+01 1.54E+01 2.75E+03 6.68E+01 No Data 4.70E+0!

E Cs-134 2.98E+05 7.09E+05 5.80E+05 No Data 2.29E+05 7.62E+04 1.24E+04 g Cs 136 3.12 E+04 1.23 E+05 8.86E+04 No Data 6.85E+04 9.39E+03 1.40E+04 Cs-137 3.82E+05 5.22E+05 3.42E+05 No Data 1.77E+05 5.89E+04 1.01E+04 Cs-138 Ba 139 2.64 E+02 9.29E-01 5.22E+02 6.62E-04 2.59E+02 2.72E-02 No Data No Data 3.84 E+02 6.19E-04 3.79E+01 3.76E-04 2.23E-03 1.65E+00 h

Ba 140 1.94E+02 2.44E-01 1.27E+01 No Data 8.31E-02 1.40E-01 4.00E+02 l'

Ba 141 4.50E-01 3.40E-04 1.52E 02 No Data 3.16E-04 1.93E-04 2.12E 10 g Ba 142 2.04E-01 2.09E-04 1.28E-02 No Data 1.77E-04 1.19E-04 0 La 140 1.50E-01 7.53E-02 1.99E-02 No Data No Data No Data 5.53E+03 La 142 7.65E 03 3.48E-03 8.66E-04 No Data No Data No Data 2.54 E+0!

Cc 141 2.24E-02 1.51E-02 1.72E 03 No Data 7.03E-03 No Data 5.78E401 Cc 143 3.94 E-03 2 92E+00 3.23E-04 No Data 1.28E-03 No Data 1.09E+02 Cc 144 1.17E+00 4.88E-01 6.26E-02 No Data 2 89E-01 No Data 3 94E+02 l

W Pr 143 5.50E-01 2 21E-01 2.73E-02 No Data 1.27E 01 No Data 2 4 I E+03 Nd-147 3.76E-01 4.3 5 E-01 2.60E-02 No Data 2.54E-01 No Data 2.09E+03 Eu 154 3.67E+01 4.52E+00 3 2 ] E+00 No Data 2.16E+0 ! No Data 3.27E+03 Hf 181 3.99E-02 1.94E-01 1.80E 02 No Data 4.17E-02 No Data 2.21E+02 W 187 2 96E+02 2.47E+02 8 64E+01 No Data No Data No Data 8.09E+04 Np-239 2.84E-02 2 80E-03 1.54E 03 No Data 8.72E-03 No Data 5.74E+02 l

m

APA ZZ-01003 Res . 3 TABLE 2.2 l BIOACCUMULATION FACTOR (Bfg (pCi/l:c) oer (oCi/ liter)

Bfi

. Element Fish (Freshwater)

H 9.0 E 01 Be 2.0 E + 00 l -

Na 1.0 E + 02 l

Cr 2.0 E + O2 Mn 4.0 E + O2 Fe 1.0 E + 02 Co 5.0 E + 01 Ni 1.0 E + 02 Cu 5.0 E + 01 Zn 2.0 E + 03 Br 4.2 E + 02 Rb 2.01E + 03 Sr 3.0 E + 01 Y 2.5 E + 0!

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 Te 4.0 E + O2 1 ,1.5 E + 01 Cs 2.0 E + 03 Ba 4.0 E + 00 La 2.5 E + 01 Ce 1.0 E + 00 Pr 2.5 E + 01 Nd 2.5 E + 01 Eu 2.5 E + 01 Hf 3.3 E + 00 W l.2 E + 03 Np 1.0 E + 01 i

"' Values from Regulatory Guide 1.109 Rev. I, Table A 1 and References 11 14.4 and 11.14 8 10-l... - _ _ _ - . . _ . _ .

APA ZZ-01003 Rev. 5

3. GASEOUS EFFLUENTS 3.1 GASEOUS EFFLUENT MONTTORS ,

1 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 gr:cous emuent radiation monitor is determined based l 5

on the instantaneous nob!c gas total body and skin dose rate limits of REC 9.6.1.1, at the SITE  ;

BOUNDARY locadon with the highest annual average X/Q value.

Each monitor channel is provided with a two level system which provides sequential alarms on increasing radioactivity levels. These setpoints are designated as alert setpoints and alarm / trip l l

setpoints. (Ref. I1.6.3)

The radiation monitor alarm / trip setpoints for each release point are based on the radioacdve noble l 5

gases in gaseous emuents. It is not considered practicable to apply instantaneous alarm / trip setpoints to integrating radiation monitors sensitive to radiciodines, radioactive materials in particulate form and radionuclides other than noble gases. Conservative assumpdons may be necessary in establishing setpoints to account for sy stem vasiables, such as the measurement system emeiency and detection capabilities during normal, anticipated, and unusual operadng condidons, the variability in release flow and principal radionuclides, and the time lag between alarm / trip action and the final isolation of the radioactive emuent. (Ref. I1.8.5) Table 9.2-B provides the instrument surveillance requirements, such as calibradon, source checking, functional testing, and channel checking.

3.1.1 Continuous Release Gaseous Emuent Monitors The radiation detection monitors associated with continuous gaseous emuent releases are (Ref.

I1.6.8,11.6.9):

. Monitor I.D. Description g

GT-RE 21 Unit Vent M GH-RE 10 Radwaste Building Vent Each of the above continuously monitors gaseous radioacdvity concentrations downstream of the last point of potential influent, and therefore measures emuents and not inplant concentrations.

The unit vent monitor continuously monitors the emuent from the unit vent for gaseous g radioactivity. The unit vent, via ventilation exhaust systems, continuously purges vanous tanks and g sumps normally containing low-level radioactive aerated liquids that can potentially generate airborne activity. The exhaust systems which supply air to the unit vent are from the fuel building, g auxiliary building, the access control area, the containment purge, and the condenser air discharge. g The unit vent monitor provides alarm functions only, and does not terminate releases from the unit vent.

The Radwaste Building ventilation emuent monitor continuously monitors for gaseous radioactivity 5 in the emuent 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, 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 5

I l

APA-ZZ-01003 l Rev 5 i I

l )

i The continuous gaseous emuent monitor setpoints are established using the methodology described l in Section 3.2. Since there are two continuous gaseous emuent release points, a fracdon of the total dose rate limit (DRL) will be allocated to each release point. Neglecting the batch releases, the plant Unit Vent monitor has been allocated 0.7 DRL and the Radwaste Building Vent monitor has been a!!ocated 0.3 DRL. These allocation factors may be changed as required to support plant operadonal l needs, but shall not be allowed to exceed unity (i.e., l.0). Therefore, a particular monitor reaching  !

the setpoint would not necessarily mean the dose rate limit at the SITE BOUNDARY is being j exceeded; the alarm only 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.

3.1.2 Batch Release Gaseous Monitors l

The radiation mordtors associated with batch release gaseous emuents are (Ref. I1.6.9,11.6.10, 11.6.I1):

Monitor I D. Description GT-RE-22 Containment Purge System GT-RE-33 l GT-RE 10 Radwaste Building Vent The Containment Purge System continuously monitors the containment purge exhaust duct during purge operations for gaseous radioactivity. The primary purpose of these monitors is to isolate the l containment purge system on high gaseous activity via the ESFAS.

l' The sample points are located outside the containment between the containment isolation dampers and the containment purge filter adsorber unit. l

\

- The Radwaste Building Vent monitor was previously described.

l 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 REC 9.6.1.1 are not exceeded.

3.2 G ASEOUS EFFLUENT MONITOR SETPOINTS The alarm / trip setpoint for gaseous emuent monitors is determined based on the lesser of the total body dose rate (equadon 3.1) and skin dose rate (equation 3.3), 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 ofless than or equal to SE-3 pCi/ce, as required by Technical Specification 4.9.4.2.

The actual setpoint value will be reduced according to the Instrument Loop Uncertainty Estimate (ILUE). This value will also be utilized in the event that there is no detectable noble gas actisity in the containment atmosphere sample analyzed in accordance with REC 9.6.1.1. The full derivadon of th:s value is discussed in reference 11.14.6.

1 I

l

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l 12 I

l

l APA ZZ-01003 W Res. 5 i l

i I ,

=

3.2.1 Total Body Dose Rate Setnoint Calculations i

To ensure that the limits of REC 9.6.1.1 are met, the alarm / trip setpoint based on the total body dose g rate is calculated according to: g, S. 5 D.R.F,F, (3.1)

Where:

= the alarm / trip setpoint based on the total body dose rate ( Ci/cc).

Stb

= REC 9.6.1.1 timit of 500 mrem /yr, conservatively interpreted as a continuous release Dtb over a one year period.

l W ,

Fs = the safety factor, a conservative factor used to compensate for statistical fluctuations and errors of measurement. (For example, Fs = 0.5 corresponds to a 100% variation.) l Default value is Fs = 1.0. E ,

Fa = the allocation factor which will modify the required dilution factor such that simultaneous gaseous releases may be made without exceeding the limits of REC 9.6.1.1 The default value is 1/n, where n is the number of pathways planned for release.

l m

= factor used to convert dose rate to the emuent concentration as measured by the emuent Rtb monitor, in ( Ci/cc) per (mrem /yr) to the total body, determined according to:

R. = C + (X/Q) [ K Q, 3 (3.2) ~

. i .

~

Where:

-^'

) C = monitor reading of a noble gas monitor corresponding to the sample radionuclide g concentrations for the batch to be released. Concentrations are determined in g accordance with Table 9.6-A. The mixture of radionuclides determined sia grab sampling of the emuent 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( Ci/cc).

X/Q = the highest calculated annual average relative concentration for any area at or beyond g the SITE BOUNDARY in (sec/m 3). Refer to Tables 6.1,6.2 and 6 4. 3 Kj = the total body dose factor d'uc to gamma emissions for each identified noble gas radionuclide, in (mrem /yr) per ( Ci/m3 ). (Table 3.1)

]

Q, = rate of release of noble gas radionuclide, i, in ( Ci/sec).

l Qi si calculated as the product of the ventilation path flow rate and the measured activity of the g, emuent stream as determined by grab samphng. gi l 1

I 1

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I I

I i l

.. .. . . _ . _ . _ _ . _ . _ _ . _ . . - - _ _ ~ - . _ . . - - - - . - -. ._-_

l APA-ZZ-01003 Rev. 5 l

l 3.2.2 Skin Dose lute Setooint Calculation To ensure that the limits of REC 9.6.1.1 are met, the alarm / trip setpoint based on the skin dose rate

, is calculated according to:

l S, s; D,R,F,F, (3.3)

Where:

Fsand F aare as previously defined.

=

S, the alarm / trip setpoint based on the skin dose rate.

j Ds = REC 9.6;1.1 limit of 3000 mrem /yr, conservatively interpreted as a continuous release

! over a one year petiod.

=

R, factor used to convert dose rate to the effluent concentration as measured by the effluent monitor, in (pCi/cc) per (mrem /yr) to the skin, determined according to.

L f R, = C + (X/Q) [ (L +i 1.1M,) Q i (3.4) i . i- .

Where:

= the skin dose factor due to beta emissions for each identified noble gas radionuclide in Li l- (mrem /yr)per( Ci/m3).

1.1 = conversion factor: I mrad air dose = 1.1 mrem skin dose.

, Mj =

the air dose factor due to gamma emissions for each identified noble gas radionuclide. in (mrad /p)per(pCi/m3).

l C, (X / Q) and Q iare previously defined.

3.3 CALCULATION OF DOSE AND DOSE RATE FROM GASEOUS EFFLUENTS 3.3.1 Calculation of Dose Rate l

The following methodology is applicable to the location (SITE BOUNDARY or beyond) characterized by the values of the parameter (X/Q) which results in the maximum total body or skin dose rate. In the event that the analysis indicates a different location for the total body and skin dose limitations, the location selected for consideration is that which minimizes the allowable release values. (Ref. I1.8.6)

The factors Kj, L , and M relate i the radionuclide airborne concentrations to various dose rates, assuming a semi-infinite cloud model.

3.3.1.1 Noble Gases l The release rate limit for noble gases is determined according to the following general relationships (Ref. I1.8.6):

D. = [K,((X/Q)Q,) s 500 mrem / yr (3.5) i D, = (L, +1.1 M,)((X / Q)Q,) s 3000 mrem / yr (3.6) i l \

i 14

i APA ZZ-Ul003 Rev. 5 l

Where:

= Total body dose rate, conservatively averaged over a period of one year.

Did Kj = Total body dose factor due to gamma emissions for each identified noble gas '

radionuclide, in (mrem /yr) per (pCi/m3 ). )

X/Q = The highest calculated annual average relative concentration for any area at or l

beyond the SITE BOUNDARY. Refer to Tables 6.1,6.2, and 6.4 for applicability. W Qj = The release rate of noble gas radionuclides, i, in gaseous effluents, from all vent releases in (pCi/sec). Qj is calculated as the product of the ventilation path flow rate and the measured activity of the effluent stream as determined by grab sampling.

Ds = Skin dose rate, conservatively averaged over a period of one year.

Lj = Skin dose factor due to beta emissions for each identified noble gas radionuclide, in (mrem /yr) per (pCi/m3 ).

1.1 = Units conversion factor; I mrad air dose = 1.1 mrem skin dose.

Mj = Air dose factor due to gamma emissions for each identified noble gas radionuclide, in -  ;

(mrad /yr) per (pCi/m3 ).

3.3.1.2 Radionuclides Other Than Noble Gases The release rate limit for lodine-131 and lodine-133, for tdtium, and for all radioactive materials in particulate form with halflives greater than 8 days is determined according to (Ref. I 1.8 7):

D = [ Rj X / Oj O, 5 1500 mrem / yr (3.7) =

- Where:

Do = Dose rate to any critical organ, in (mrem /yr).

Rj = Dose parameter for radionuclides other than noble gases for the inhalation pathway for E the child, based on the critical organ, in (mrem />T) per ( Ci/m3 ). l Qi and (X / Q) are as previously defined.

The dose parameter (R ) includes the in/ 'rnal dosimetry of radionuclide. i, and the receptor's breathing rate, which are functions of the receptor's age. The child age group has been selected as the limiting age group. All radiodines are assumed to be released in elemental form (ref. I 1.8.7).

Rj values were calculated according to (Ref. I1.8.8):

R, = K' (BR) DFA, (3.8)

K' = Units conversion factor: lE06 pCi/ Ci BR = The breathing rate. (Regulatory Guide 1.109, Table E-5).

=

DFAj The maximum organ inhalation dose factor for the ith radionuclide, in (mrem /pCi). The total body is considered as an organ in the selection of DFAj. (Ref. I1.11.5 and i1.14 4)

I I,

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l APA-ZZ-01003 Rev 5 l

3.3.2 Dose Due to Gaseous Emuents

! 3.3.2.1 Noble Gases i The air dose at the SITE BOUNDARY due to noble gases is calculated according to the following i methodology (Ref. I1.8.9):

Dhring any calendar quarter, for gamma radiation:

D, = 3.17E 08 [ {M,(X/Q) q,) s 5 mrad (3.9) 1 During any calenaar quarter, for beta radiation:

D, = 3.17E-08 [ (N,(X /Q) q,l s10 mrad (3.10) l 1 During any calendar year, for gamma radiation:

D, = 3.~17E-08 [ (M,(X/ Q) q,) s 10 mrad (3.11) i +

During any calendar year, for beta radiation:

D, = 3.17E-08 [ (N, (X/ Q) q, ) s 20 mrad (3.12)

Where:

Dg = Air dose from gamma radiation due to noble gases released in gaseous emuent.

) Db

=

=

Air dose from beta radiation due to noble gases released in gaseous emuents.

Ni The air dose factor due to beta emissions for each identified noble gas radionuclide, i, in (mradlyr)per(pCi/m3). j

= The releases of radioiodines, radioactive materials in particulate form, and radionuclides I qi other than noble gases, i, in gaseous emuents, for all gaseous releases in ( Ci). Releases are cumulative over the calendar quarter or year as appropriate.

1 3.17E-08 = The inverse of the number of seconds per year.

X / G & M are i as previously defined. '

3.3.2.2 Radionuclides Other Than Noble Gases The dose to a MEMBER OF THE PUBLIC from lodine-131 and 133, tritium, and all radionuclides in particulate form with half lives greater than 8 days in gaseous emuents released, to areas at and beyond the SITE BOUNDARY, is calculated according to the following expressions:

During any calendar quarter:

I D, = 3.17E-08 [Ri [W q,] s 7.5 mrem (3.13) 1 During any calendar year: I D, = 3.17E-08 [Ri (W q,) s 15 mrem (3.14) l 1

1 4

a

APA-ZZ-01003 Rev. 5 Where:

I Dj = Dose to a MEMBER Ol' THE PUBLIC from radionuclides other than noble gases.

= The dose factor for each identified radionuclide, i, in m2(mrem /yr) per ( Ci/sec) or R{

(mrem /yr)per( Ci/m3).

W = (X / Q) for the inhalation and tritium pathways, in (sec/m'). Refer to Tables 6.1, 6.2, and 6.4 for applicability.

W = (D / G) for the food and ground plane pathways, in g 4

(meters ). Refer to Tables 6.1,6.2 and 6.4 for applicability. W (D/Q) = the average relative deposition of the efDuent at or beyond the SITE BOUND ARY, E considering depletion of the plume during transport. g 3.17 E-08 = The inverse of the number of seconds per year.

q is as previously dermed.

For the direction sectors with existing pathways within 5 miles from the site, the appropriate Rj values are used. If no real pathway exists within 5 miles from the center of the building complex, the cow-milk R; value is used, and it is assumed that this pathway exists at the 4.5 to 5.0 mile l 5

distance in the limiting-case sector, If the R; for an existing pathway within 5 miles is less than a cow-milk Rj at 4.5 to 5.0 miles, then the value of the cow-milk R; at 4.5 to 5.0 miles is used. (Ref.

9.8.10)

Although the annual average relative concentration (X/Q) and the average relative deposiuon rate us (D/Q) are generally considered to be at the approximate receptor location in lieu of the SITE

) BOUNDARY for these calculations, it is acceptab!c to consider the ingestion, inhalation, and ground B plane pathways to coexist at the location of the nearest residence with the highest value of(X/Q). g (Ref. I1.8.9) The Total Body dose from ground planc deposition is added to the dose for each individual organ. (Ref. I1.11.3) 3.4 GASEOUS RADWASTE TREATMENT SYSTEM The gaseous radwaste treatment system and the ventilation exhaust system are available for use whenever gaseous efDuents require treatment prior to being released to the environment. The g gaseous radwaste treatment system is designed to allow for the retention of all gaseous fission g products to be discharged from the reactorcoolant systern. The retention system consists of eight (89 waste gas decay tanks. Normally, waste gases will be retained for at least 60 days prior to discharp These systems will provide reasonable assurance that the releases of radioactive materials in gascons efDuents will be kept ALARA.

The OPERABILITY of the gaseous radwaste treatment system ensures this system will be available E for use when gases require treatment prior to their release to the emironment. OPERABILITY is demonstrated through compliance with REC 9.6.1.1,9.7.1.1, and 9.8.1.1.

I Projected doses (gamma air, beta air, and organ dose) due to gaseous efDuents at or beyond the SITE g BOUNDARY are determined each 31 days by dividing the cumulative annual total by the number of g elapsed months.

3.5 DOSE FACTORS The dose conversion factors provided in the following tables were denved from the appropnate dos: M conversion factors in Regulatory Guide 1.109 and other sources as necessary (Ref.11.14 9 and 11.14.11). Per USNRC guidance, particulate nuclides with a halDtic ofless than 8 day s are not g considered (Ref: 11.24). Y-90. Nb-95, La 140, and Pr-144 are included because the parent halDife is g

) greater than 8 day s, and secular equilibrium is assumed Non-gamma emitting nuclides not listed in Table 9.6 A are also not considered. (CTSN 43121)

I 12 3

r

v. .

'[

APA-ZZ-01003 .j Rev.5' 4

TABLE 3.1 '

.r l DOSE FACTOR FOR EXPOSURE TO A SEMI-INFINITE CLOUD OF NOBLE GASES t

i Total Body Gamma Air Beta Air i Dose Factor Skin Dose Factor Dose Factor Dose Factor ,

l Ki L M Ni Radionuclide (mrem /yr)per( Ci/m3 ) (mrad /yr)per(pCi/m3 ) (mradlyr) per (pCi/m3 ) (mradlyrper(pCi/m3 )

! ._ C u Wii AL.. Eulaahadsiacc.cdate&WMnMnud3;EssekasilEs!EG#sissBi2iLFA-Nkyta!sN2RMPE%93Qdl t

Kr-83m 7.56 E-02 -

1.93 E+01 2.88 E+02 l Kr-85m 1.17E+03 1.46E+03 1.23 E+03 1.97 E+03  !

l Kr-85 1.61 E+01 ~

1.34 E+03 1.72 E+01 1.95 E+03 i Kr-87 5.92 E+03 9.73 E+03 6.17 E+03 1.03 E+04 Kr-88 1.47 E+04 2.37 E+03 1.52 E+04 2.93 E+03 Kr-89 1.66 E+04 1.01 E+04 1.73 E+04 1.06 E+04 [

Kr-90 1.56 E+04 7.29 E+03 1.63 E+04 7.83 E+03  !

Xc-131m 9.15 E+01 4.76 E+02 1.56 E+02 1.11 E+03 Xc-133m 2.51 E402 9.94 E+02 3.27 E+02 1.48 E+03 Xc-133 2.94 E+02 3.06 E+02. 3.53 E+02 1.05 E403 Xe-135m 3.12 E+03 7.11 E402 3.36 E+03 7.39 E+02

. Xc-135 1.81 E+03 1.86 E+03 1.92 E+03 2.46 E+03 [

4 Xe-137 1.42 E+03 1.22 E+04 1.51 E+03 1.27 E+04 Xc-138 8.83 E+03 4.13 E+03 9.21 E+03 4.75 E+03 t Ar-4 I 8.84 E+03 2.69 E+03 9.30 E+03 3.28 E+03 I l [

r b

i l

l

._- = _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _

APA-ZZ-01003 Rev.5 TABLE 3.2 PATHWAY DOSE FACTORS (Ri ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Ground Plane Pathway I

(m2mrem /yr)per( Ci/sec)

' NUCLIDE TOTAL BODY SKIN Be 7 2.24E+07 3.21E+07 i

Cr-51 4.66E+06 5.51E+06 Mn 54 1.39E+09 1.63E+09

.I =

Fe-59 2.73E+08 3.21E+08 Co-57 2.98E+08 4.37E+08 Co 58 3.79E+08 4.44E+08 g Co40 2.15E+10 2.53 E+ 10 g I

)

Zn-65 7.47E+08 8.59E+08 l "

Rb-86 8.99E+06 1.03 E+07 I

St-89 2.16E+04 2.51E+04 5.36E+06 6.32E+06 l

W Y-90 l Y 91m 1.00E+05 1.16E+05 Y-91 1.07E+06 1.21E+06 l

Zr 95 2.45E+08 2.84 Et08 i

Nb 95 2.50E+08 2.94E+08 Ru 103 1.08E+08 1.26E+08 Ru 106 4.22E+08 5.07E+08 Ag-110m 3.44E+09 4.01 E+09 Cd-109 3.76E+07 1.54E+08 Sn-113 1.43E47 4.09E+07 l Sb-124 8.74 E+08 1.23E 49 EE Sb-125 3.57E+09 5.19E49 l Tc-127m 9.17E+04 108E+05 Te 129m 1.98E+07 2 31E+07 I

l.'30 5.5 i E+06 6 69E+06 l ;31 1.72E+07 2 09E+07 I

l APA-ZZ-01003 i Rev 5 l

l l TABLE 3.2 l'

- PATHWAY DOSE FACTORS (Rj) FOR RADIONUCLIDES l OTHER THAN NOBLE GASES l Ground Plane Pathway l

(m2mrem /yr)per(pCi/sec)

' NUCLIDE TOTAL BODY SKIN l

?

l l 1132 1.25E+06 1.47E+06 1

! l 133 2.45E+06 2.98E+06 I I 134 4.47E+05 5.31E+05 I 135 2.53E+06 2.95E+06 Cs 134 6.85E+09 8.00E+09 l Cs 136 1.51E+08 1.71E+08 l Cs 137 1.03 E+10 ' l.20E+10 1~

Ba 140 2.05E+07 2.35E+07 l La 140 1.47E+08 1.66E+08 l l

I

\ l I_/ Cc-141 'I.37E+01 ' l .54 E+07 l

l Cc 144 6.96E+07 8.04 E+07 Pr-144 4.35E+07 5.00E+07 Nd 147 8.39E+06 1.0lE+07 l

l Eu-154 2.21E+10 3.15E+10

, Hf 181 1.97E+08 2.82E+08

! l l

l i-t I

i I

APA-ZZ-01003 Rev. 5 TABLE 3.3 l ClllLD PATIIWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTilER TilAN NOBLE GASES Inhalation Pathway (mrrm/yr) per (pCi/m')

TOTAL I

NUCLTDE BONE LIVER BODY THYROID KIDNEY LUNG GI LLI H-3 ND 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 Be-7 8.47E+02 1.44E+03 9.25E+02 ND ND 6.47E+04 - 2.55E+03 Cr-51 ND ND 1.54E+02 8.55E+01 2.43E+01 1.70E+04 1.08E+03 Mn-54 ND 4.29E404 9.51E+03 ND 1.00E+04 1.58E+06 2.29E+04 Fe-55 4.74E+04 2.52E+04 7.77E+03 ND ND 1llE+05 2.87E+03 Fe-59 2.07EM4 3.34E+04 1.67E+04 ND ND 1.27E+06 7.07E+04 Co-57 ND 9.03E+02 1.07E+03 ND ND 5.07E+05 1.32E+04 l Co-58 ND 1.77E+03 3.16E403 ND ND 1. l l E+06 3.44 E+04 W Co-60 ND 1.31E+04 2.26E+04 ND ND 7.07E+06 9.62E+04 Zn-65 4.25E+04 1.13E+05 7.03E+04 ND 7.14 E+04 9.95E+05 1.63 E+04 Rb-86 ND 1.98E+05 1.14E+05 ND ND ND 7.99E+03 l l Sr89 5.99E+05 ND 1.72E M4 ND ND 2.16E+06 1.67E+05 W Sr90 1.01E+08 ND 6.44E+06 ND ND 1.48E+07 3.43E+05 Y 90 4. l lE+03 ND 1.llE+02 ND ND 2.62E+05 2.6SE+05 l Y-91m 5.07E-01 ND 1.84E-02 -

ND ND 2.81E+03 4.72E+03 Y 91 9.14E+05 ND 2.44E+04 ND ND 2.63E+06 1.84E+05 l

I Zr 95 1.90E+05 4.18E+04 3.70E+04 ND 5.96E+04 2.23E+06 6. l lE+04 l Nb-95 2.35E+04 9.18E+03 6.55E+03 ND 8.62E+03 6.14E+05 3.70E+04 E I

Ru-103 2.79E443 ND 1.07E+03 ND' 7.03E+03 6.62E+05 4.48E+04 3

l Ru-106 1.36E+05 ND 1.69E+04 ND 1.84E+05 1.43E+07 4.29E+05 Ag-110m 1.69E+04 1.14E+04 9.14 E+03 ND 2.12E+04 5.48E+06 1.00E +05 Cd 109 ND 5.48E+05 2.59E+04 ND 4.96E+05 1.05E+06 2.78E+04 Sn-113 1.13E+05 3.12E+03 8.62E+03 2.33E+03 ND 1.46E+06 2 26E+05 Sb-124 5.74 E+04 7.40E+02 2.00E+04 1.26E+02 ND 3.24 E +06 1.64E+05 Sb-125 9.84 E404 7.59E+02 2.07E+04 9.10E+01 ND 2.32 E+06 4 03E+04 Tc-127m 2 49E+04 8.55E+03 3.02E+03 6.07E+03 6.36E+04 1.48E+06 7.14E+04 Tc-129m 1.92E+04 6.85E+03 3.04 E +03 6.33E+03 5.03E+04 176E406 182E+05 1-130 8.18E+03 164E +04 8.44E+03 1.85E+06 2.4 5 E +04 ND 5IIE+03 i

i APA ZZ-01003 Rev 5 TABLE 3.3 (Con'O l C111LD PAT 11WAY DOSE FACTORS (Rj) FOR RADIONUCLIDES OTilER TilAN NOBLE GASES Inhalation Pathway (mrem /yr)per( C1/m')

i 1

{ TOTAL l NUCLIDE BONE LIVER BODY THYROID KIDNEY LUNG GILLI i 1131 4.81E+04 4.81E+04 2.73E+04 1.62E+07 7.88E+04 ND 2.84E+03 l I132 2.12E+03 4.07E+03 1.88E+03 1.94E+05 6.25E+03 ND 3.20E+03 i 1133 1.66E+04 2.03 E+04 7.70E+03 3.85E+06 3.38E+04 ND 5.48E+03 1-134 1.17E+03 2.16E+03 - 9.95E+02 5.07E+04 3.30E+03 ND 9.55E+02 1135 4.92E+03 8.73E+03 4.14E+03 7.92E+05 1.34E+04 ND 4.44E+03 Cs-134 6.51E+05 1.01E+06 2.25E+05 ND 3.30E+05 1.21E+05 3.85E+03 Cs 136 6.51E+04 1.71E+05 1.16E+05 ND 9.55E+04 1.45E+04 4.18E+03 Cs 137 9.07E+05 8.25E+05 1.28E+05 ND 2.82E+05 1.04E+05 3.62E+03 Ba 140 7.40E+04 6.48E+01 4.33E+03 ND 2.llE+01 1.74E+06 1.02E+05 La 140 6.44E+02 2.25E+02 7.55E+01 ND ND 1.83E+05 2.26E+05 Cc 141 3.92E+04 1.95E+04 2.90E+03 ND 8.55E+03 5.44 E+05 5.66E+04 Cc 144 6.77E+06 2.12E+06 3.61E+05 ND 1.17E+06 1.20E M7 3.89E+05 Pr 143 1.85E+04 5.55E+03 9.14 E+02 ND 3.00E+03 4.33 E+05 9.73 E+04 l Pr 144 5.96E-02 1.85 E-02 3.00E-03 ND 9.77E-03 1.57E+03 1.97E+02 Nd 147 1.08E+04 8.73E+03 6.81E+02 ND 4.81E+03 3.28E+05 8.21 E+04 Eu 154 1.01E+07 9.21E405 8.40E+05 ND 4.03E+06 6.14 E+06 1.10E+05 Hf 181 2.78E+04 1.01E+05 1.25E+04 ND 2.05E+04 1.06E +06 E62E+04 l

l l i

l l

t

)

{

22-l l

l

APA ZZ-01003 I

, TABLE 3.3 (Cont'd) l CIIILD PATIIWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTIIER TilAN NOBLE G ASES Meat Pathway (m* mrem /yr) per ( Ci/sec)

TOTAL NUCLIDE BONE LIVER BODY TIIYROID KIDNEY LUNG GI-LLI H3 ND 2.34E+02 2.34E+02 2.34 E+02 2.34E+02 2.34E+02 2.34 E+02 -

Be 7 7.38E+03 1.26E+04 8.07E+03 0.00E+00 1.23E+04 0.00E+00 7.00E+05 I

Cr 51 0.00E+00 0.00E+00 8.80E403 4.88E+03 1.33E+03 8.92E+03 4.67E+05 Mn-54 0.00E+00 8.02E+06 2.14E+06 0.00E+00 2.25E+06 0.00E+00 6.73E46 l

Fe-55 4.58E+08 2.43E+08 7.52E+07 0.00E+00 0.00E+00 1.37E+08 4.50E+07 Fe-59 3.77E408 6.10E+08 3.04E+08 0.00E+00 0.00E+00 1.77E+08 6.35E+08 Co-57 0.00E+00 5.92E+06 1.20E+07 0.00E+00 0.00E+00 0.00E+00 4.85E+07 Co-58 0.00E+00 1.64E+07 5.03E+07 0.00E+00 0.00E+00 0.00E+00 9.59E+07 Co-60 0.00E40 6.94E+07 2.0$E+08 0.00E400 0.00E+00 0.00E+00 3.84E+08 Zn-65 3.76E+08 1.00E+09 6.23E+08 0.00E+00 6.31E+08 0.00E+00 1.76E+08 Rb-86 0.00E+00 S.77E+08 3.55E+08 0.00E+00 0.00E+00 0.00E+00 3.71 E+07 l

St-89 4 82E&O8 0.00E+00 1.38E+07 0.00E+00 0.00E400 0.00E+00 1.87E+07 St 90 1.04E+10 0.00E+00 2.64E+09 0.00E+00 0.00E+00 0.00E+00 1.40E+08 Y 90 1.93E+05 0.00E+00 5.16E+03 0.00E+00 0.00E+00 0.00E400 5.49E+08 l Y-91 m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Y 91 1.80E+06 0.00E+00 4.82E+04 0.00E+00 0.00E+00 0.00E40 2.40E+08 g l E Zr 95 2.67E+06 5.86E+05 5.22E+05 . 0.00E+00 8.39E+05 0.00E+00 6 llE+08 l

Nb 95 4.25E+06 1.66E+06 1.18E+06 0.00E+00 1.56E+06 0.00E+00 3.07E+09 <

Ru 103 1.55E48 0.00E+00 5.96E+07 0.00E+00 3.90E H)8 0.00E+00 4.01 E +09 I

Ru 106 4.44E+09 0.00E+00 5.54E+08 0.00E+00 6.00E+09 0.00E+00 6.91E+10 Ag 110m 8.40E+06 5.67E+06 4.53 E+06 0.00S+00 1.06E+07 0.00E+00 6.75E+0S Cd 109 0.00E+00 1.91E+06 8.84 E+04 0.00E+00 1.70E+06 0.00E+00 6.18E+06 Snll3 2.18E+09 4.48E+07 1.24E+0S 3.31E+09 0.00E+00 0.00E+00 1.54 E+09 Sb 124 2.93 E+07 3.80E+05 1.03 E+07 6.46E+04 0.00E+00 1.62E+07 1.83 E+08 l Sb-125 2.85E+07 2.20E+05 5.97E+06 2.64E+04 0.00E +00 1.59E+07 6 81E+07 Te 127m 1.78E+09 4.78E+08 2. l l E+08 4.25E+08 5.07E 49 0.00E+00 1.44 E+09 l Tc 129m 1.79E+09 5.00E403 2 78E+08 5.78E408 5.26E+09 0.00E+00 2.19E+09 I l130 3.06E-06 6.18E-06 3.18E-06 6.80E 04 9.23 E-06 0.00E40 2.89E-06 l 2

1 D1 1.66E+07 1.67E+07 9.47E46 5.51E+09 2.74E +07 0.00E+00 1.4 SE +06 l l ' 132 0.00E+00 0.00E400 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+ou Il

-23

APA ZZ-Otto) l Rev. 5 j TABLE 3.3 (Cont'd) l Cff LLD PATIIWAY DOSE FACTORS (Rj) FOR RADIONUCLIDES OTilER TilAN NOBLE GASES Meat Pathway (m8 mremlyr)per(pCi/sec)

TOTAL NUCl.IDE BONE LIVER BODY THYROID KIDNEY LUNG GI.LLI I133 5.70E 01 7.05E 01 2.67E 01 1.31E+02 1.17E+00 0.00E+00 2.84E-01 1134 0.00E400 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00  ;

l I-135 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00  !

Cs-134 9.23E+08 1.51E+09 3.20E+08 0.00E+00 4.69E+08 1.68E+08 8.17E+06  ;

Cs 136 1.62E+07 4.46E+07 2.89E+07 0.00E+00 2.38E+07 3.54E+06 1.57E+06 i Cs 137 1.33E+09 1.28E+09 1.89E+08 0.00E+00 4.16E+08 1.50E+08 8.00E+06 Ba-140 4.39E+07 3.85E+04 2.56E+06 0.00E+00 1.25E+04 2.29E+04 2.22E+07 La 140 3.33E+02 1.17E+02 3.93 E+01 0.00E+00 0.00E+00 0.00E+00 3.25E+06 Cc-141 2.22E+04 1.llE+04 1.65E+03 0.00E+00 4.86E+03 0.00E+00 II38E+07 I

Cc 144 2.32E+06 7.27E+05 1.24E+05 0.00E+00 402E+05 0.00E+00 1.89E+08

. Pr 143 3.34E+04 1.00E+04 1.66E+03 0.00E+00 5.43E+03 0.00E+00 3.61 E+07

.) Pr-144 5.63E+02 1.74E+02 2.83E+01 0.00E +00 9.21E+0! 0.00E+00 3.75E+05 Nd-147 1.17E+04 9.48E+03 7.34E+02 0.00E+00 5.20E+03 0.00E+00 1.50E+07 Eu 154 1.12E+07 1.01E+06 9.20E+05 0.00E+00 4.43E+06 0.00E+00 2.34E+08 Hf 181 4.77E+06 1.74E+07 2.15E+06 0.00E+00 3.53E+06 0.00E+00 6.41E+09 l

l j

l ,

l i

l r

24-l

lll APA-ZZ-01003 Rev. 5 TAHLE 3.3 (Cont'd)

CIIILD PATHWAY DOSE FACTORS (R;)FOR RADIONUCLIDES OTHER TIIAN '

l NOBLE GASES Grass-Cow-Milk Pathway (m8 mrem /yr)per(pCi/sec)

TOTAL NUCLIDE BONE LIVER BODY THYROID KIDNEY LUNG GI-LLI H3 0.00E+00 1.57EM3 1.57E+03 1.57E+03 1.57E+03 1.57E+03 1.57E+03 Be-7 7.50E+03 1.28E+04 8.20E+03 0.00E+00 1.25E+04 0.00EM0 7.12E+05

.I Cr-51 0.00E+00 0.00E+00 1.02E+05 5.66E+04 1.55E+04 1.03E+05 5.40E+06 Mn 54 0.00E+00 2.10E+07 5.59E M6 0.00E+00 5.89E+06 0.00E+00 1.76E+07 Fe 55 1.12E+08 5.94E+07 1.84E M7 0.00E+00 0.00E+00 3.36,E+07 1.10E+07 Fe-59 1.20E+08 1.95E+08 9.70E+07 0.00E+00 0.00E+00 5.64 E+07 2.03E+08 Co-57 0.00E+00 3.84E+06 7.78E+06 0.00E+00 0.00E+00 0.00E+00 3.15E+07

, Co-58 0.00E40 1.21E+07 3.72E+07 0.00E40 0.00E40 0.00E+00 7.08E+07

m. Co-60 0.00E+00 4.32E+07 1.27E+08 0.00E+00 0.00E+00 0.00E+00 2.39E+08 Zn-65 4.14E+09 1.10E+10 6.86E+09 0.00E+00 6.95E+09 0.00E+00 1.94 E+09 I E Rb-86 0.00E+00 8.78E+09 5.40E+09 0.00E+00 0.00E40 0.00E+00 5.65E+08 5 i

St-89 6.63 E+09 0.00E+00 1.89EM8 0.00E+00 0.00E+00 0.00E+00 2.57E+08 g Sr-90 1.12E+ 11 0.00E+00 2.84 E + 10 0.00E+00 0.00E+00 0.00E+00 1.51E+09 g Y-90 3.38E+03 0.00E+00 9.05E+01 0.00E+00 0.00E+00 0.00E+00 9.62E+06 l Y 91m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Y 91 3.91E+04 0.00E+00 1.04E+03 - 0.00E+00 0.00E+00 0.00E+00 5.20E+06 Zr-95 3.84E+03 8.43 E+02 7.51E+02 0.00EM0 1.21EM3 0.00E+00 8.80Et05 Nb-95 3.72E+05 1.45E+05 1.03E+05 0.00E+00 1.36E+05 0.00E+00 2.68E+08 I

Ru 103 4.29E+03 0.00E+00 1.65E+03 0.00E400 1.08E+04 0.00E+00 1.llE+05 i

Ru 106 9.25E+04 0.00E+00 1.15E+04 0.00E+00 1.25E+05 0.00E400 1.44E+06 Ag-110m 2.09E+08 1.4]E+08 1.13 E+08 0.00E+00 2.63 E+08 0.00E+00 1.6SE'10 Cd 109 0.00E+00 3.86E+06 1.79E+05 0.00E+00 3.45E+06 0.00E+00 1.25E+07 Sn-ll3 6. l lE+08 1.26E+07 3.48E+07 9.29E+08 0.00E+00 0.00E+00 4.32E+08 Sb-124 1.09E-08 1.41E+06 3.81 E+07 2.40E+05 0.00E+00 6.03 E+07 6.80E-08 Sb-125 8 71E-07 6.72E+05 1.83E+07 8.07E+04 0.00E+00 4.86E+07 2.08E+08 l Tc-127m 2 08E+0x 561E+07 2.47E+07 4.98E+07 5.94E+0S 0.00E+00 1.69E+08 -

Te129m 2.72E+08 7.59E+07 4.22E+07 8.76E+07 7.98E+08 0 00E+00 3 31E+0S i II

APA 2.Z-01003 Rev 5 TABLE 3.3 (Cont'd)

CHILD PATIIWAY DOSE FACTORS (R;)FOR RADIONUCLIDES OTIIER TilAN l NOBLE GASES Grass-Cow-Milk Pathway (m' artm/yr) per (pCl/sec)

TOTAL '

NUCLIDE BONE LIVER BODY TIIYROID KIDNEY LUNG Cl-LLI l

l130 1.73E+06 3.50E+06 1.80E+06 3.85E+08 5.23E+06 0.00E+00 1.64E+06 1131 1.30E+09 1.31E+09 7.46E+08 4.34E+11 2.15E+09

)

0.00E400 1.17E+08 '

I 132 6.92E-01 1.27E400 5.85E-01 5.90E+01 1.95E+00 0.00E+00 1.50E+00 1-133 1.72E+07 2.13E+07 8.05E+06 3.95E+09 3.54E+07 0.00E+00 8.57E+06

-l I134 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1135 5.4 ]E+04 9.74E+04 4.61E+04 8.63E+06 1.49E+05 0.00E+00 7.42E+04 Cs-134 2.27E+10 3.72E+10 7.84E+09 0.00E+00 1.15E+10 4.14E+09 2.00E+08 Cs 136 1.01E+09 2.78E+09 1.80E+09 0.00E+00 1.48E+09 2.21E+08 9.78E+07 Cs-137 3.23E+10 3.09E+10 4.56E+09 0.00E+00 1 ole +10 3.62E+09 I93E+08 l

) Ba 140 1.17E+08 1.03E+05 6.84E+06 0.00E+00 3.34E+04 6.12E+04 5.94E+07 l

Cc 141 2.19E+04 1.09E+04 1.62E+03 0.00E+00 4.79E+03 0.00E+00 1.36E+07 La-140 1.78E+02 6.23E+01 2.10E+01 0.00E+00 0.00E+00 0.00E+00 1.74E+06 Cc-144 1.62E+06 5.09E+05 8.67E404 0.00E+00 2.82E+05 0.00E+00 1.33E+08 Pr 143 7.19E+02 - 2.16E+02 3.57E+01 0.00E+00 1.17E+02 0.00E+00 7.76E+05 Pr 144 5.04E+00 1.56E+00 2.53E-01 0.00E+00 8.24E-01 0.00E+00 3.35E+03 Nd-147 4.45E+02 3.61E+02 2.79E+01 0.00E+00 1.98E*02 0.00E+00 5.71Et05 Eu-154 9.43E+04 8.48E403 7.75E+03 0.00E+00 3.73E+04 0.00E+00 1.97Et06 Hf-181 6 44E+02 2.35E+03 2.91E+02 0.00E+00 4.76E+02 0.00E+00 8.66E+05 l

1 h

a 26-

I APA-ZZ-01003 Rev. 5 TABLE 3.3 (Cont'd)

CHILD PATHWAY DOSE FACTORS (Ri ) FOR RADIONUCLIDES OTHER THAN l NOBLE GASES Grass-Goat-Milk Pathway (m8 mrem /yr)per(pCUsec)

TOTAL NUCLTDE BONE LIVER BODY' THYROID KIDNEY LUNG GI-LLI H-3 0.00E+00 3.20E43 3.20E+03 3.20E+03 3.20E+03 3.20E+03 3.20E+03 I

Be-7 9.00E+02 1.53E+03 9.84E+02 0.00E+00 1.50E+03 0.00E+00 8.55E+04 Cr-51 0.00E+00 0.00E+00 1.22E+04 6.79E+03 1.85E+03 1.24E+04 6.48E+05 Mn-54 0.00E+00 2.52E+06 6.71E+05 0.00E+00 7.06E+05 0.00E+00 2. l lE+06 m I l Fe 55 1.45E+06 7.72E+05 2.39E+05 0.00E+00 0.00E+00 4.36E+05 1.43E+05 Fe-59 1.56E+06 2.53E+06 1.26E+06 0.00E+00 0.00E+00 7.34E+05 2.64E+06 Co-57 0.00E+00 4.61E+05 9.33 E+05 0.00E+00 0.00E+00 0.00E+00 3.78E+06 Co-58 0.00E+00 1.46E+06 4.46E+06 0.00E+00 0.00E+00 0.00E+00 8.50E+06 Co-60 0.00E+00 5.19E+06 1.53E+07 0.00E+00 0.00E+00 0.00E40 2.87E+07 l

Zn-65 4.97E+08 1.32E+09 8.23E+08 0.00E+00 8.34E+08 0.00E+00 2.32E+08 l

Rb-86 0.00E+00 1.05E+09 6.48E+08 0.00E+00 0.00E+00 0.00E+00 6.78E+07 St-89 1.39E+10 0.00E+00 3.97E+08 0.00E+00 0.00E+00 0.00E+00 5.39E+08 Sr 90 2.35E+11 0.00E+00 5.95E+10 0.00E+00 0.00E+00 0.00E+00 3.16E+09 Y-90 4.06E+02 0.00E+00 1.09E+01 0.00E40 0.00E+00 0.00E+00 1.15E+06 l Y 91m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E40 g Y-91 4.69E+03 0. 00E+00 1.25E+02 0.00E+00 0.00E+00 0.00E+00 6.25E+05 g l

• Zr-95 4.60E+02 1.01E+02 9.01E+01 0.00E <00 1.45E+02 0.00Et00 1.06E+05 I

l Nb-95 4.46E+04 1.74E+04 1.24 E+04 0.00E+00 1.63E+04 0.00E+00 3.21 E+07 l Ru-103 5.14 E+02 0.00E+00 1.98E+02 0.00E+00 1.29E+03 0.00E+00 1.33E+04 l

Ru-106 1.11E+04 0.00E+00 1.38E+03 0.00E+00 1.50E+04 0.00E+00 1.73 E+05 Ag-110m 2.5 ] E+07 1.69E+07 1.35E+07 0.00E+00 3.15E+07 0.00E+00 2.01E+09 Cd-109 0.00E+00 4.64E+05 2.15 E+04 0.00E+00 4.14E+05 0.00E+00 1.50E+06 Snll3 7.33 E+07 1.51E+06 4.18E+06 1. l l E+08 0.00E+00 0.00E+00 5.18E+07 Sb-124 1.30E+07 1.69E+05 4.57E+06 2.88E+04 0.00E+00 7.24 E+06 8.16E+07 Sb-125 1.05E+07 8.06E+04 2.19E+06 9.68E+03 0.00E+00 5.83E+06 2.50E+07 g l Te 127m 2.50E+07 6 73E+06 2.97E+06 5.98E+06 7.13 E+07 0.00E+00 2.02 E+07 g Te 129m 3.26E47 9.10E+06 5.06E+06 1.05E+07 9.57E+07 0.00E+00 3.98E+07 1-130 2.08E+06 4.20E+06 2.16E+06 4.62E+08 6.27E+06 0 00E+00 1.96E +06 1131 1.57E+09 1.57E+09 8.95E+08 5.21 E+ 11 2 58E+09 1.40E+08 I132 8.30E-01 1.53E+00 7.02E-01 7.08E+01 2.34E+00 l 0.00E+00 0.00E40 1.80E+00 I

-2 7-

APA ZZ-01003 Rev. 5 I

TABLE 3.3 (Cont'd)

CHILD PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTIIER THAN l NOBLE GASES Grass-Goat-Milk Pathway (m8 mrcm/yr)per(pCi/sec)

TOTAL NUCLIDE BONE LTVER BODY THYROID KIDNEY LUNG GI-LLI l133 2.06E+07 2.55E+07 9.66E+06 4.74E+09 4.25E+07 0.00E+00 1.03E+07 l l 134 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1 135 6.49E+04 1.17E+05 5.53E+04 1.04E+07 1.79E+05 0.00E+00 8.90E+04 Cs 134 6.80E+ 10 1.12E+11 2.35E+10 0.00E+00 3.46E+10 1.24E+10 6.01E+08 Cs 136 3.04E+09 8.35E+09 5.40E+09 0.00EM0 4.45E+09 6.63E+08 2.93E+08 Cs-137 9.68E+10 9.27E+10 1.37E+10 0.00E+00 3.02E+10 1.09E+ 10 5.80E+08 I

Ba 140 1.41E+07 1.23E+04 8.21E+05 0.00E+00 4.01E+03 7.35E+03 7.13E+06 La 140 2.14E+01 7.47E400 2.52E+00 0.00E+00 0.00E+00 0.00E+00 2.08E+05 Ce 141 2.63E+03 1.31E+03 1.95E+02 0.00E+00 5.75E+02 0.00E+00 1.63 E+06 I

6.llE+04 0.00E+00 3.38E404 0.00E+00 1.59E+07

) Cc 144 Pr 143 1.95E405 8.63E+01 2.59E+01 1.04E+04 4.28E+00 0.00E+00 1.40E+01 0.00E+00 9.31 E+04 Pr 144 6.05E-01 1.87E-01 3.04E-02 0.00E+00 9.89E 02 0.00E+00 4.03E+02 Nd 147 5.34E+0! 4.33E+01 3.35E+00 0.00E+00 2.37E+01 0.00E+00 6.85E+04 Eu-154 1.13E+04 1.02E+03 9.29E+02 0.00E+00 4.47E+03 0.00E+00 2.37E+05 Hf 181 7.73E+01 2.81E+02 3.49E+01 0.00E+00 5.72E+01 0.00E+00 1.04E+05 I

4 l

-2 8-1 I

APA-ZZ 01003 Rev. 5 TABLE 3.3 (Cont'd)

CHILD PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER TIIAN l NOBLE GASES Vegetation Pathway (m mrem /yr)per(pCi/sec) 8 TOTAL BONE LIVER BODY TITYROID KIDNEY LUNG GI-LLI I

NUCLTDE H3 ND 4.01E+03 4.01E+03 4.01E+03 4.01E+03 4.01E+03 4.01E+03 Be 7 3.38E+05 5.76E+05 3.70EM5 0.00E+00 5.65E+05 0.00E+00 3.21E+07 l

Cr 51 0.00E+00 0.00E+00 1.17E+05 6.50E+04 1.78E+04 1.19E+05 6.21E+06 0.00E+00 1.86E+08 0.00E+00 5.58E+0L I

Mn-54 0.00E+00 6.65E+08 1.77E+08 i

Fe-55 8.01E+08 4.25E408 1.32E+08 0.00E+00 0.00E+00 2.40E+08 7.87E+07 Fe-59 3.98E+08 6.43E+08- 3.20E+08 0.00E+00 0.00E+00 1.87E+08 6.70E+08 Co-57 0.00E+00 2.99E+07 6.04E+07 0.00E+00 0.00E+00 0.00E+00 2.45E+08 Co-58 0.00E+00 6.44E+07 1.97E+08 0.00E+00 0.00E+00 0.00E+00 3.76E+08 Co-60 0.00E+00 3.78E+08 1.12E+09 0.00E+00 0.00E+00 0.00E+00 2.10E+09 Zn-65 8.13E+08 2.17E+09 1.35E+09 0.00E+00 1.36E+09 0.00E+00 3.80E+08 Rb-86 0.00E+00 4.5?E+08 2.78E+08 0.00E+00 0.00E+00 0.00E+00 2.91E+07 Sr-89 3.60E+10 0.00E400 1.03E+09 0.00E+00 0.00E+00 0.00E@0 1.39E+09 Sr-90 1.24E+12 0.00E+00 3.15E+11 0.00E+00 0.00E+00 0.00E+00 1.67E+ 10 Y-90 3.01E+06 0.00E+00 8.04E+04 0.00E+00 0.00E+00 0.00E@0 8.56E+09 l Y-91m 8.95E-09 0.00E+00 3.26E-10 0.00E+00 0.00E+00 0.00E+00 1.75E-05 g Y 91 1.86E+07 0.00E+00 4.99E+05 0.00E+00 0.00E+00 0.00E+00 2.4 8E+09 g l

Zr-95 3.86E+06 8.48E405 7.55E+05 0.00E+00 1.21E+06 0.00E+00 8.85E48 l

i Nb-95 7.48E+05 2.91E+05 2.08E+05 0.00E+00 2.74E+05 0.00E+00 5.39E+08 l Ru 103 1.53E+07 0.00E+00 5.90E+06 0.00E+00 3.86E+07 0.00E+00 3.97E+08 g l 3 Ru-106 7.45E108 0.00E+00 9.30E+07 0.00E400 1.01E+09 0.00E+00 1.16E+ 10 Ag-110m 3.2 ] E+07 2.17E+07 1.73E+07 0.00E+00 4.04E+07 0.00E+00 2.58E+09 Cd-109 0.00E+00 2.45E408 1.14 E+07 0.00E+00 2.18E+08 0.00E+00 7.94E+08 Sn-113 1.58E+09 3.25E407 9.00E+07 2.40E+09 0.00E+00 0.00E+00 1.12E+09 Sb-124 3.52E+08 4.57E+06 1.23 E+08 7.77E+05 0.00E+00 1.95E+08 2.20E+09 Sb-125 4.99E+08 3.85E+06 1.05E+08 4.63 E+05 0.00E+00 2.78E+08 1.19E+09 l Tc-127m 1.32E+09 3.56E+08 1.57E+08 3.16E+08 3.77E+09 0.00E+00 107E+09 Tc-129m 8 41E+08 2.35E+08 1.31E+08 2.71E+08 2.47E+09 0 00E+00 103 E+09 l

l-130 1-131 6.16E+05 1.43 E+08 1.24 E+06 1.44E+08 6.41 E+05 8.17E+07 1.37E+08 4.76E+10 1.86E+06 2.36E+0S 0 00E+00 000E+00 5 82E+05 1.28E+07 l

I

-2 9 -

Il

r-r l

APA-ZZ-01003 Rev. 5 TABLE 3.3 (Cont'd) l l- CIIILD PATIIWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTIIER TIlAN l NOBLE GASES Vegetation Pathway (m' mrem /yr) per (pCi/sec) j TOTAL NUCLiDE BONE LIVER BODY ,T, TIYROID K_IDNEY LUNG GI-LLI I132 9.23E+01 1,70E+02 7.80E+01 7.87E+03 2.60E+02 0.00E+00 2.00E+02 1 1133 3.53E+06 4.37E+06 1.65E+06 8.12E+08 7.28E+06 0.00E+00 1.76E+06 l 1-134 1.56E-04 2.89E-04 1.33E-04 6.65E-03 4.42E-04 0.00E+00 1.92E-04 I135 6.26E+04 1.13E+05 5.33E+04 9.98E+06 1.73E+05 0.00E+00 8.59E+04 Cs 134 1.60E+10 2.63E+10 5.55E+09 0.00E+00 8.15E+09 2.93E+09 1.42E+08 Cs 136 8.24E+07 2.27E+03 1.47E+08 0.00E+00 1.21E+08 1.80E+07 7.96E+06 Cs-137 2.39E+10 2.29E+10 3.38E+09 0.00E+00 7.46E+09 2.68E+09 1.43E+08 I

Ba-140 2.77E+08 2.43E+05 1.62E+07 0.00E+00 7.90E+04 1.45E+05 1.40E+0S l La-140 3.36E+04 1.18E+04 3.96E+03 0.00E+00 0.00E+00 0.00E+00 3.28E+08 'l Cc 141 6.56E+05 3.27E+05 4.86E+04 0.00E+00 1.43E+05 0.00E+00 4.08E+08 l

\

, ,.e Cc 144 1.27E+08 3.98E407 6.78E+06 0.00E+00 2.21E+07 0.00E+00 1.04E+ 10 Pr 143 1.46E+05 4.37E+04 7.23E+03 0.00EM0 2.37E+04 0.00E+00 . 1.57E+08 Pr144 7.88E+03 2.44E+03 3.97E+02 0.00E+00 1.29E+03 0.00E+00 5.25E+06 Nd 147 7.15E+04 5.79E+04 4.48E+03 0.00E+00 3.18E+04 0.00E+00 9.17E+07 Eu 154 1.66E+08 1.50E+07 1.37E+07 0.00E+00 6.57E+07 0.00E+00 3.48E+09 Hf 181 4.90E+05 1.79E+06 2.21E+05 0.00E+00 3.63E+05 0.00E+00 6.59E+08 I

i 30-

APA-ZZ-01003 Rev. 5 TABLE 3.4 ADULT PATHWAY DOSE FACTORS (Rj) FOR RADIONUCLIDES OTHER THAN l NOBLE GASES Inhalation Pathway (mrem /yr) per (pCi/m")

TOTAL NUCLIDE BONE LIVER BODY TIIYROID KIDNEY LUNG GI-LLI H3 ND 1.26E+03 1.26E+03 1.26E+03 1.26E+03 1.26E+03 1.26E+03 Be 7 4.27E+02 9.68E+02 4.70E+02 ND ND 4.21 E+04 5.35E+03 I

Cr 51 ND ND 1.00E+02 5.95E+01 2.28E+01 1.44E+04 3.32E+03 Mn 54 ND 3.96E+04 6.30E+03 ND 9.84E+03 1.40E+05 7.74E+04 l

Fe-55 2.46E+04 1.70E+04 3.94E+03 ND ND 7.21E+04 6.03 E+03 Fe 59 1.18E+04 2.78E+04 1.06E+04 ND ND 1.02E+06 1.88E+05 Co 57 ND 6.92E+02 6.71E+02 ND ND 3.70E+05 3.14E+04 Co-58 ND 1.58E+03 2.07E+03 ND ND 9.28E+05 1.06E+05

=

Co-60 ND 1.15E+04 1.48E+04 ND ND 5.97E+06 2.85E+05

~

Zn 65 3.24 E+04 1.03 E+05 l4 66E+04 ND 6.90E+04 8.64E+05 5.34E+04 l

Rb-86 ND 1.35E+05 5.90E+04 ND ND ND 1.66E+04 St89 3.04 E+05 ND 8.72E+03 ND ND 1.40E+06 3.50E+05 Sr 90 9.92E+07 ND 6.10E+06 ND ND 9.60E+06 7.22E+05 Y 90 2.09E+03 ND 5.61E+01 ND ND 170E+05 5.06E+05 l Y-91m 2.61E-01 ND 1.02E 02 ND ND i92E+03 1.33E+00 Y-91 4.62E+05 ND 1.24E+04 ND ND . 70E+06 3.85E+05 l

Zr-95 1.07E+05 3.44E+04 2.33E+04 ND 5.42E+04 1.77E+06 1.50E+05 l i bb95 1.41E+04 7.82E+03 4.21E+03 ND 7.74E+03 '. 5.05E+05 1.04E+05 l

Ru-103 1.53E+03 ND 6.58E+02 ND 5.83E+03 5.0$E405 1.10E+05 Ru-106 6.91 E+04 ND 8.72E+03 ND 1.34E+05 9.36E+06 9.12E+05 Ag 110m 1.08E+04 1.00E+04 5.94 E+03 ND 1.97E+04 4.63E+06 3.02E+05 Cd-109 ND 3.67E+05 1.31 E+04 ND 3.57E+05 6.83E+05 5.82E+04 Sn-ll3 5.72E+04 2.18E+03 4.39E+03 1.24E403 ND 9.44E+05 1.18E+05 Sb-124 3.12E+04 5.89E+02 1.24 E+04 7.55E+01 ND 2.48E+06 4.06E+05 Sb-125 5.3 4 E+04 5.95E+02 1.26E+04 5.40E+01 ND 1.74E+06 1.01E+05 l

l Te-127m 1.26E+04 5.77E+03 1.57E+03 3.29E+03 4.58E+04 9.60E+05 1.50E+05 =

Tc 129m 9.76E4 03 4.67E+03 1.58E+03 3.44E+03 3.66E+04 1.16E+06 3 83E+05 I-130 4.58E+03 1.34 E+04 5.2ME+03 1.14E+06 2.09E+04 ND 7.69E+03 1-131 2.52E+04 3.58E+04 2.05E+04 1.19E+07 6.13E+04 ND 6 28E+03 I

I l

APA ZZ-01003  ;

Rev. 5 i i

TABLE 3.4 (Cont'd) ,

1 ADULT PATIIWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTIIER TilAN I l NOBLE GASES Inhalation Pathway (mrem /yr) per (pCi/m*) l TOTAL I NUCLIDE BONE LIVER BODY TRYROTD KIDNEY LUNG GI-LLI, j I132 1.16E+03 3.26E+03 1.16E+03 1.14E+05 5.18E+03 ND 4.06E+02 1-133 8.64E+03 1.48E+04 4.52E+03 2.15E+06 2.58E+04 ND 8.88E+03 I134 6.44E+02 1.73E+03 6.15E+02 2.98E+04 2.75E+03 ND 1.01E+00 I135 2.68E+03 6.98E+03 2.57E+03 4.48E+05 1.!!E+04 ND 5.25E+03 Cs 134 3.73E+05 8.48E+05 7.28E+05 ND 2.87E+05 9.76E+04 1.04 E+04 Cs 136 3.90E+04 1.46E+05 1.10E+05 ND 8.56E+04 1.20E+04 1.17E+04 Cs 137 4.78E+05 6.21E+05 4.28E+05 ND 2.22E+05 7.52E+04 8.40E+03 I '

l Ba-140 3.90E+04 4.90E+01 2.57E+03 ND 1.67E+01 1.27E+06 2.18E+05 La-140 3.44E+02 1.74E+02 4.58E+01 ND ND 1.36E+05 4.58E+05 '

1 Cc-141 1.99E+04 1.35E+04 1.53E+03 ND 6.26E+03 3.62E+05 1.20E+05 l

}

1.

Cc-144 3.43 E+06 1.43E+06 1.84E+05 ND 8.48E+05 7.78E+06 8.16E+05

! Pr 143 9.36E+03 3.75E+03 4.64 E+02 ND 2.16E+03 2.81E+05 2.00E+05 l

Pr 144 3.01E-02 1.25E-02 1.53E-03 ND 7.05E-03 1.02E+03 2.15E-08 Nd-147 5.27E+03 6.10E+03 3.65E+02 ND 3.56E+03 2.21E+05 1.73E+05 l Eu 154 5.92E+06 7.28E+05 5.18E+05 ND 3.49E+06 4.67E+06 2.72E+05 Hf 181 1.41E+04 6.82E+04 6.32E+03 ND 1.48E+04 6.85E+05 1.39E+05 I .

l l

I f

32-l l

l APA ZZ-01003 l Rev. 5

)

TABLE 3.4 (Cont'd)

ADULT PATIIWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTIIER THAN l NOBLE GASES Meat Pathway I (m' mrem /yr) per (pCi/sec)

TOTAL NUCLIDE BONE LIVER BODY TTIYROID KIDNEY LUNG GI-LLI H3 ND 3.25E+02 3.25E+02 3.25E+02 3.25E+02 3.25E+02 3.25E+02 Be-7 4.57F+03 1.04E+04 5.07E+03 ND 1.10E+04 ND 1.81E+06 Cr 51 ND ND 7.04E+03 4.21 E+03 1.55E+03 9.34E+03 1.77E+06 Mn-54 ND 9.17E+06 1.75E+06 ND 2.73E+06 ND 2.81E+07 m i B Fe-55 2.93E+08 2.02E+08 4.72E+07 ND ND 1.13E+08 1.16E+08 Fe 59 2.65E+08 6.24E+08 2.39E+08 ND ND 1.74E+08 2 08E+09 Co-57 ND 5.63E 46 9.36E+06 ND ND ND 1.43E+08 Co-58 ND 1 '32E+07 4.08E+07 ND ND ND 3.69E+08 Co-60 ND 7.51E+07 1.66E+08 ND ND ND 1.41E+09

) l 2n-65 3.56E+08 1.33E+09 5.llE+08 ND 7.57E+08 ND 7.13 E+08 l

Rb-86 ND 4.87E+08 2.27E+08 ND ND ND 9.60E+07 l

Sr-89 3.01E+08 ND 8.65E+06 ND ND ND 4.83E+07 St-90 1.24E+10 ND 3.05E+09 ND ND ND 3.59E+08 Y-90 1.21E+05 ND 3.24E+03 ND ND ND 1.28E+09 l Y-91m 0.00E+00 ND 0.00E+00 ND ND ND 000E+00 Y-91 1.13 E+06 ND 3.02E+04 - ND ND ND 6 23E+08 '

Zr 95 1.87E+06 6.00E+05 4.06E+05 ND 9.42E+05 ND 1.90E+09 Nb-95 3.15E+06 1.75E+06 9.43E+05 ND 1.73E+06 ND 1.06E+10 .

Tc-99m 2.87E+02 8.10E+02 1.03 E+04 ND 1.23 E+04 3.97E+02 4.79E+05 l Ru-103 1.05E+08 ND 4.53 E+07 ND 4.01E+08 ND 1.23E+10 W !

l l Ru 106 2.80E+09 ND 3.54 E+08 ND 5.40E+09 ND 1.8 ] E+ 11 Agllom 6.68E+06 6.18E+06 3.67E+06 ND 1.21E+07 ND 2.52E+09 Cd-109 ND 1.59E+06 5.55E+04 ND 1.52E+06 ND 1.60E+07 Sn 113 1.37E+09 3 88E+07 7.86E+07 2.22E+07 ND ND 4.09E+09 Sb-124 1.98E+07 3.74E+05 7.84 E +06 4.79E+04 ND 1.54E+07 5.61 E+08 i Sb-125 1.91E+07 2.13E+05 4.54E+06 1.94 E+04 ND 1.47E+07 2.10E+08 l l Tc-127m 1.1 iE+09 3.98E+08 1.36E+08 2.85E+08 4.53E+09 ND 3.74E+09 l Tc-129m 1.13E+09 4.23E+08 1.79E+08 3.89E408 4.73 E+09 ND 5.71 E+09 3 1

1130 2.12E-06 6.27E 06 2.47E-06 5 31E-04 9.78E-06 ND 5.40E-06 3

1-131 1.08E+07 1.54E+07 8.82E+06 5.04E+09 2.64 E+07 ND 4 06E+06 q

APA-ZZ-01003 Rev. 5 '

l TABLE 3.4 (Cont'd)

ADULT PATHWAY DOSE FACTORS (Rj) FOR RADIONUCLIDES OTHER TilAN l l NOBLE GASES .

Meat Pathway (m8 mrem /yr) per (pCi/sec)

TOTAL NUCLTDE BONE LTVER BODY THYROID KIDNEY LUNG GI-LLI l I-132 0.00E+00 0.00E+00 0.00E+00 0.00E+00 l 0.00E+00 ND 0.00E+00 l I133 3.67E-01 6.39E 1.95E-01 9.38E+01 1.11 E+00 ND 5.74E-01 j 1134 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 ND 0.00E+00 )

1 l l-135 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 ND 0.00E+00 1 Cs 134 6.57E+08 1.56E+09 1.28E+09 ND 5.06E+08 1.68E+08 2.74E47 Cs 136 1.20E+07 4.76E+07 3.42E407 ND 2.65E+07 3.63E+06 5.40E+06 Cs-137 8.71E+08 1.19E+09 7.81E+08 ND 4.04E+08 1.34E+08 2.31E+07 l

Ba 140 2.87E+07 3.61E+04 1.88E+06 ND 1.23E+04 2.07E+04 5.91E+07 La 140 2.21E+02 1.llE+02 2.94E+01 ND ND ND 8.18E+06 '

Cc 141 1.40E+04 9.49E+03 1.08E+03 ND 4.41E+03 ND 3.63E+07 s l l

I Cc 144 1.46E+06 6.09E+05 7.82E+04 ND 3.61E+05 ND 4.92E+08 Pr143 2.10E+04 8.40E+03 1.04E+03 ND 4.85E+03 ND 9.18E+07 Pr-144 3.52E+02 1.46E+02 1.79E+01 ND 8.24E+01 ND 5.06E-05 I l

Nd-147 7.07E+03 8.17E403 4.89E+02 ND 4.77E+03 .ND 3.92E+07 Eu 154 8.02E+06 9.86E+05 7.01E+05 ND 4.72E+06 ND 7.14 E+08 Hf 181 3.01E+06 1.46E+07 1.35E+06 ND 3.14E+06 ND 1.66E+10 l

l l

l 34 l

APA ZZ-01003 l Rev.5 l TABLE 3.4 (Cont'd)

ADULT PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES IIi l

Grass-Cow-Milk Pathway (m mrcm/yr)per(pCihec) 8 TOTAL NUCLTDE BONE LIVER BODY THYROID KIDNEY LUNG GI-LLI i l

H3 ND 7.63 E+02 7.63E+02 7.63E+02 7.63 E+02 7.63E+02 7.63E402 Be 7 1.63E+03 3.72E+03 1.81E+03 ND 3.93E+03 ND 6.45E+05 I

Cr 51 ND ND 2.86E+04 1.71E+04 6.30E+03 3.79E+04 7.19E+06 Mn-54 ND 8.42E406 1.61E+06 ND 2.50E+06 ND 2.58E47 l

Fe 55 2.51E+07 1.74E+07 4.0$E+06 ND ND 9.68E+06 9.96E+06 i Fe-59 2.97E+07 6.98E+07 2.68E+07 ND ND 1.95E+07 2.33 E+08 Co-57 ND 1.28E+06 2.13E+06 ,ND ND ND 3.25E+07 l Co-58 ND 4.72E+06 1.06E+07 ND ND ND 9.56E+0,7 Co-60 ND 1.64E+07 3.62E+07 ND ND ND 3.08E+08 l

Zn-65 1.37E+09 4.37E+09 1.97E+09 ND 2.92E+09 ND 2.75E+09

~

l l Rb-86 ND 2.60E+09 1.21E+09 ND .ND ND 5.12E+08 )

I i Sr 89 1.45E+09 ND 4.17E+07 ND ND ND 2.33E+08 Sr 90 4.68E+10 ND 1.15E+10 ND ND ND 1.35E+09 Y-90 7.43E+02 ND 1.99E+01 ND ND ND 7.87E+06 l Y-91m 0 00E+00 ND 0.00E+00 ND ND ND 0.00E+00 Y-91 8 59E+03 ND 2.30E+02 , ND ND ND 4.73 E+06

! i Zr-95 9.44E+02 3.03 E+02 2.05E+02 ND 4.75E+02 ND 9.59E+05 l l

Nb-95 9.65E+04 5.37E+04 2.89E+04 ND 5.31E+04 ND 3.26E+08 l

Ru 103 1.02E+03 ND 4.39E+02 ND 3.89E+03 ND 1.19E+05 l

Ru 106 2.04 E+04 ND 2.58E+03 ND 3.94E+04 ND 1.32E+06 i Ag-110m 5.82E +07 5.39E+07 3.20E407 ND 1.06E+08 ND Cd 109 ND 1.13E+06 3.95E+04 ND 1.08E+06 ND 2.20E+10 1.14 E+07 ll E

Sn-ll3 1.34 E+08 3.81 E+06 7.73E+06 2.18E+06 ND ND 4.02 E+08 Sb-124 2.57E+07 4 86E405 1.02E+07 6.24 E+04 ND 2.00E+07 7.31E+08 g Sb-125 2.04E+07 2 2EE+05 4.87E+06 2.08E+04 ND 1.58E+07 2 25E+0S g l Tc-127m 4.58E+07 1.64E+07 5.58E +06 1.17E+07 1.86E+08 ND I$4E+08 Tc-129m 6.02E+07 2.25E+07 9.53 E+06 2.07E+07 2.51 E+08 ND 3.03E+08 l

I-130 4.21 E+05 1.24E 46 4.91 E+ 05 1.05E+08 1.94E+06 ND 1.07E+06 I-131 2.97E+08 4.25E+08 2.43 E+08 1.39E+ 11 7.28E+08 ND 1.12E+0S

-3 5-

APA-ZZ-01003 Rev.5 l

I TABLE 3.4 (Cont'd)

ADULT PATIIWAY DOSE FACTORS (Rj) FOR RADIONUCLIDES OTIIER TIIAN l NOBLE GASES l

l Grass-Cow-Milk Pathway (m8 mrem /yr)per(pC1/sec) i i TOTAL NUCLIDE BONE LIVER BODY TIIYROID KIDNEY LUNG GI-LLI I

l-132 1.65E-01 4.42E-01 1.55E 01 1.55E+01 7.04E-01 ND 8.30E-02 1133 3.88E+06 6.75E+06 2.06E+06 9.92E+08 1.18E+07 ND 6.07E+06 i l l134 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 ND 0.00E+00 l

l135 1.29E+04 3.37E+04 1.25E+04 2.23E+06 5.41E+04 ND 3.81E+04 Cs-134 5.65E+09 1.35E+10 1.10E+10 ND 4.35E+09 1.45E+09 2.35E+08 Cs-136 2.63E+08 1.04E+09 7.48E+08 ND 5.79E+08 7.93E407 1.18E+08 Cs 137 7.38E+09 1.01E+10 6.61E+09 ND 3.43E+09 1.14E+09 1.95E+08 l

l Ba 140 2.69E+07 3.38E+04 1.76E+06 ND 1.15E+04 1.93E+04 5.54E+07

l. La 140 4.14E+01 2.09E+01 5.51E+00 ND ND ND 1.53E+06 l Cc-141 4.85E+03 3.28E+03 3.72E+02 ND 1.52E+03 ND 1.25E+07 Cc 144 3.58E+05 1.50E+05 1.92E+04 ND 8.87E+04 ND 1.21E+08 l " 7.83E+00 ND 3.66E+01 ND 6 92E+05 Pr 143 1.58E+02 6.34E+01 i

1 l Pr 144 1.10E+00 4.58E-01 5.61E-02 ND 2.58E-01 ND 1.59E-07 i Nd 147 9.42E+01 1.09E+02 6.51E+00 ND 6.36E+01 ND 5.23E+05 l Eu-154 2.37E+04 2.91E+03 2.07E+03 ND 1.39E+04 ND 2. l l E+06 l Hf 181 1.42E+02 6.92E+02 6.41E+01 ND 1.49E+02 ND 7.87E+05 l  !

l l

I

APA ZZ-01003 Rev. 5 TABLE 3.4 (Cont'd)

ADULT PATHWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTHER THAN l NOBLE GASES Grass-Goat-Milk Pathway (m* mrem /yr) per (pCi/see)

TOTAL NUCLIDE BONE LIVER BODY THYROID KIDNEY LUNG GI-LLI H3 ND 1.56E+03 1.56E+03 1.56E+03 1.56E+03 1.56E+03 1.56E+03 Be 7 1.96E+02 4.47E+02 2.17E+02 ND 4.72E+02 ND 7.74E+04 l

Cr-51 ND ND 3.43E+03 2.05E+03 7.56E+02 4.56E+03 8.63E+05 Mn 54 ND 1.01E+06 1.93E+05 ND 3.01E+05 ND 3.10E+06 l l

Fe-55 3.27E+05 2.26E+05 5.26E+04 ND ND 1.26E+05 1.30E+05 Fe-59 3.87E+05 9.08E+05 3.48E+05 ND ND 2.54E+05 3.03 E+06 Co-57 ND 1.54E+05 2.56E+05 ND ND ND 3.90E+06 Co-58 ND 5.66E+05 1.27E+06 ND ND ND 1.15E+07 Co40 ND 1.97E+06 4.35E+06 ND ND ND 3.70E+07 k, 2n-65 1.65E+08 5.24E+08 2.37E+08 ND 3.51E+08 ND 3.30E+08 l

Rb-86 ND 3.12E+08 1.45E+08 ND ND ND 6.15E+07 St-89 3.05E+09 ND 8.75E+07 ND ND ND 4.89E+08 Sr-90 9.84E+10 ND 2.41E+10 ND ND ND 2.84E+09 Y-90 8.92E+01 ND 2.39E+00 ND ND ND 9.46E+05 l Y 91m 0.00E+00 ND 0.00E+00 ND ND ND 0.00EM)0 l Y-91 1.03 E+03 ND 2.76E+01 ND ND ND 5.68E+05 5 I .

Zr-95 1.13 E+02 3.63E+01 2.46E+01 ND 5.70E+01 ND 1.15E+05 l

Nb-95 1.16E+04 6.45E+03 3.47E+03 ND 6.37E+03 ND 3.91E+07 l

Ru-103 1.22E+02 ND 5.27E+01 ND 4.67E+02 ND 1.43E+04 l

Ru 106 2.45E+03 ND 3.10E+02 ND 4.73E+03 ND 1.59E+05 Ag-110m 6.99E+06 6.47E+06 3.84E+06 ND 1.27E+07 ND 2.64E+09 Cd 109 ND 1.36E+05 4.74 E+03 ND 1.30E+05 ND 1.37E+06 Snll3 1.61E+07 4.58E+05 9.28E+05 2.62E+05 ND ND 4.83E+07 Sb-124 3.09E+06 5.84E+04 1.23E+06 7.53E+03 ND 2.41 E+06 8.78E+07 Sb 125 2.46E+06 2.74E+04 5.84E+05 2.50E+03 ND 1.89E+06 2.70E+07 l Te-127m 5.50E+06 1.97E+06 6.70E+05 1.41 E+06 2.23E+07 ND 1.84E+07 Te-129m 7.23 E+06 2.70E+06 1.14E+06 2.48E+06 3.02E+07 ND 3.64E+07 I130 5.05E+05 1.49E+06 5.88E405 1.26E+08 2.32E+06 ND 1.28E +06 l 131 3.56E+08 5.09E+08 2.92E+08 1.67E+ 11 8.72E+0S ND 1.34E+08 I132 1.98E41 5.29E-01 1.85E-01 1.85E+0! 8 43E-01 ND 9 95E-02

-37

I APA-ZZ-01003 Rev. 5 t

TABLE 3.4 (Cont'd)

ADULT PATHWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTHER THAN f l NOBLE GASES Grass-Goat-Milk Pathway (m8 mrem /yr)per(pCi/sec)

TOTAL NUCLIDE BONE LIVER BODY THYROID KIDNEY LUNG GILLI l-133 4.65E+06 8.09E+06 2.47EM6 1.19E+09 1.41E+07 ND 7.27E+06 f l I l?4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 ND 0.00E+00 1 135 1.54E+04 4.04E+04 1.49E+04 2.67E+06 6.48E+04 ND 4.57E+04 Cs 134 1.70E+10 4.04E+10 3.30E+10 ND 1.31E+10 4.34E+09 7.07E+08 Cs 136 7.91E+08 3.12E+09 2.25E+09 ND 1.74E+09 2.38E+08 3.55E+08 Cs-137 2.22E+10 3.03E+10 1.99E+10 ND 1.03E+10 3.42E+09 5.87E+08 f Ba 140 3.23E+06 4.06E+03 2.12E+05 ND 1.38E+03 2.32E+03 6.65E+06 La 140 4.97E+00 2.51E+00 6.62E-01 ND ND ND 1.84E+05 Ce 141 5.82E+02 3.94E+02 4.46E+01 ND 1.83E+02 ND 1.50E+06 l

') Cc 144 4.30E+04 1.80E+04 2.31E+03 ND 1.07E+04 ND 1.45E+07 Pr-143 1.90E+01 7.61E+00 9.40E-01 ND 4.39E+00 ND 8.31E+04 Pr 144 1.33E-01 5.50E 02 6.74E-03 ND 3.10E-02 ND 1.91E-08 Nd 147 1.13E+01 1.31E+01 7.82E-01 ND 7.64E+00 ND 6.28E+04 Eu-154 2.84E+03 3.49E+02 2.49E+02 ND 1.67E+03 ND 2.53E+05 Hf181 1.71E+01 8.31E+01 7.70E+00 ND 1.79E+01 ND 9.46E+04 l

38-l

APA-ZZ 01003 I

Rev. 5 i

TABLE 3.4 (Cont'd)

ADULT PATHWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTHER THAN l NOBLE GASES Vegetation Pathway (m* mrem /yr) per (pCi/sec)

TOTAL NUCLTDE BONE LIVER BODY THYROID KIDNEY LUNG GI-LLI H-3 ND 2.26E+03 2.26E+03 2.26E+03 2.26E+03 2.26EM3 2.26E+03 Be-7 9.24E+04 2.llE+05 1.03E+05 ND 2.23E+05 ND 3.66E+07 l

Cr-51 ND ND 4.64E44 2.78E+04 1.02E+04 6.16E+04 1.17E+07 Mn-54 ND 3.13E+08 5.97E+07 ND 9.31E+07 ND 9.59E+08 I

Fe-55 2.10E+08 1.45E+08 3.38E+07 ND ND 8.08E+07 8.31E+07 Fe-59 1.26E+08 2.96E+08 1.14E+08 ND ND 8.28E+07 9.88E+08 Co-57 ND 1.17E+07 1.95E+07 ND ND ND 2.97E+08 h

Co-58 ND 3.07E+07 6.89E+07 ND ND ND 6.23E+08 m Co-60 ND 1.67E+08 3.69E+08 ND ND ND 3.14E+09

- Zn-65 3.17E+08 1.01E+09 4.56E+08 ND 6.75E+08 ND 6.36E+08 l

Rb-86 ND 2.19E+08 1.02E+08 ND ND ND 4.33 E+07 l

St 89 9.97E+09 ND 2.86E+08 ND ND ND 1.60E+09 St-90 6.05E+11 ND 1.48E+11 ND ND ND 1.75E+10 Y-90 7.67E+05 ND 2.06E+04 ND ND ND 8.14 E+09 g l Y-91m Y-91 5.24E-09

5. l lE+06 ND ND 2.03 E-10 1.37E+05 -

ND ND ND ND ND ND 1.54 E-08 2 81E+09 5

l Zr 95 1.17E+06 3.77E+05 2.55E+05 ND 5.91E+05 ND 1.19E+09 Nb-95 2.40E+05 1.34E+05 7.19E+04 ND 1.32E+05 ND 8. l lE+08 l l l Ru 103 4.77E+06 ND 2.06E+06 ND 1.82E+07 ND 5.57E+08 i

Ru-106 1.93E+08 ND 2.44E+07 ND 3.72E+08 ND 1.25E+10 Ag-110m 1.05E+07 9.75E+06 5.79E+06 ND 1.92E+07 ND 3.98E+09 l Cd-109 0.00E40 8.36E+07 2.92E+06 ND 8.00E+07 ND 8 43E+08 =

Sn 113 4.16E+08 1.!8E+07 2.40E+07 6.75E+06 ND ND 1.25E+09 Sb-124 1.04 E+08 196E+06 4. l l E+07 2.51E+05 ND 8 07E+07 2.94E+09 l Sb-125 1.37E+08 1.53E+06 3.25 E+07 1.39E+05 ND 1.0$E+08 1.50E+09 E!

l Tc-127m 3 49E+08 1.25E+08 4.26E+07 8.92E+07 1.42E+09 ND 1.17E+09 Te-129m 2.51E+08 9.38 E+07 3.98E+07 8 64E+07 1.05E+09 ND l.27Et09 l

I-130 3.93 E+05 1.16E+06 4.57E+05 9.81E+07 1.81E+06 ND 9.97E+05 1-131 8 08E+07 1.16E+48 6.62E+07 3.79E+ 10 1.98E408 ND 3 05E+07

-3 9-

APA ZZ-01003 Rev. 5 TAHLE 3.4 (Cont'd)

ADULT PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN l NOBLE GASES Vegetation Pathway (m8 mrem /yr)per(pCihec)

TOTAL NUCLIDE BONE LIVER BODY THYROID KIDNEY LUNG GI-LLI I132 5.77E+01 1.54E+02 5.40E+01 5.40E+03 2.46E+02 ND 2.90E+01 1133 2.09E+06 3.63E+06 1.llE+06 5.33E+08 6.33E+06 ND 3.26E+06 I134 9.69E-05 2.63E-04 9.42E-05 4.56E-03 4.19E-04 ND 2.30E-07 I-135 3.90E+04 1.02E+05 3.77E+04 6.74E+06 1.64E+05 ND 1.15E+05 Cs-134 4.67E+09 1.llE+10 9.08E+09 ND 3.59E+09 1.19E+09 1.94E+08 Cs-136 4.27E+07 1.69E+08 1.21E+08 ND 9.38E+07 1.29E 47 1.91E+07 l

Cs 137 6.36E+09 8.70E+09 5.70E+09 ND 2.95E+09 9.81 E+08 1.68E+08 I i i Ba 140 1.29E+08 1.61E+05 8.42E+06 ND 5.49Et04 9.24E+04 2.65E+08

! La 140 1.58E+04 7.98E+03 2. l lE+03 ND ND ND 5.86E+08 l Cc-141 1.97E+05 1.33E+05 1.51E+04 ND 6.19E+04 ND 5.10E+08 1 I l # Ce.144 3.29E+07 1.38E+07 1.77E+06 ND 8.16E+06 ND 1.1 IE+10 Pr 143 6.26E+04 2.51E+04 3.10E+03 ND 1.45E+04 ND 2.74E+08 l

l Pr144 2.03E+03 8.43E+02 1.03E+02 ND 4.75E+02 ND 2.92E-04 l Nd 147 3.33E+04 3.85E+04 2.31E+03 ND 2.25E+04 ND 1.85E+08 Eu-154 4.85E+07 5.97E+06 4.25E+06 ND 2.86E+07 ND 4.32E+09 Hf 181 1.40E+05 6.82E+05 6.32E+04 ND 1.47E+05 ND 7.76E408 i

i l

T

-4 0-

11 APA ZZ-01003 -l Rev. 5 i

4. DOSE AND DOSE COMMITMENT FROM URANIUM FUEL CYCLE SOURCES 4.1 CALCULATION OF DOSE AND DOSE COMMITMENT FROM URANIUM FUEL CYCLE SOURCES The annual dose or dose commitment to a MEMBER OF TIIE PUBLIC for Uranium Fuel Cycle Sources is determined as: l 5

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 exposure, and direct radiadon from the Unit and outside storage tanks;

b. Dose to skin due to beta radiation from submersion in a ciud of radioactive noble gases, and ground plane exposure;

c. Thyroid dose due to inhalation and ingestion of radioiodines; and

d. Organ dose due to inhalation and ingestin of radioactive material. i 1

It is assumed that total body dose from sources of gamma radiation irradiates internal body ,

organs at the same numerical rate. (Ref.11.12.5) l The dose from gaseous efIluents 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 BOUND ARY.

l m

Since the doses via liquid releases are very conservatively evaluated, there is reasonable assurance that no real individual will receive a significant dose from radioactive liquid release pathways.

Therefore, only doses to individuals via airbome pathways and doses resulting from direct radiation  ;

are considered in determining compliance to 40 CFR 190 (Ref. I1.12.3). l There are no other Uranium Fuel Cycle Sources within 8km of the Callaway Plant.

r 4.1.1 Identification of the MEMBER OF THE PUBLIC The MEMBER OF THE PUBLIC is considered to be a real individual, including all persons not  !

occupationally associated with the Callaway Plant, but who n.;y use portions of the plant site for l recreational or other purposes not associated with the plant (Rt C 11.4 and 11.8.10). Accordingly, it l is necessary to characterize this individual with respect to his udlization of areas both within and at l

or beyond the SITE BOUNDARY and identify, as far as possibin major assumpdons which could be i reevaluated if necessary to demonstrate continued compliance with 40 CFR 190 through the use of l more realistic assumptions (Ref. I1.12.3 and 11.12.4). i The evaluation of Total Dose from the Uranium Fuel Cycle should consider the dose to Iwc Critical Receptors: a) The Nearest Resident, and b) The Critical Receptor wuhin the SITE BOUNDARY-4.1.2 Total Dose to the Nearest Resident The dose to the Nearest Resident is due to plume exposure from noble guses, ground plane exposure. E and inhalation and ingestion pathways. It is conservatively assumed that each ingestion pathway l (meat, milk, and vegetation) exists at the locadon of the Nearest Resident E;

!! is assumed that direct radiation dose from operation of the Unit and outside storage tanks. and E 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 Nearest Resident with the highest value of X/Q. l Tne location of the Nearest Resident in each meteorological sector is determined from the Annual gi Land Usc Census conducted in accordance with the Requirements of REC 9.12.1.1.

g,

)

I 4,. I

r.

APA ZZ-olou3 Rev. 5 l

4.1.3 Total Dose to the Critical Receptor Within the SITE BOUNDARY The Union Electric Company has entered into an agreement with the State of Missouri Depanment of Conservation for management of the residual lands surrounding the Callaway Plant, includmg some areas within the SITE BOUNDARY. Under the terms of this agreement, certain areas have -

' been opened to the public for low intensity recreadonal uses (hundng, hiking, sightseeing, etc.) but recreational use is excluded in an area imniediately 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. 1 i Activities conducted under these leases are primarily comprised of farming (animal feed), grazing, and forestry (Ref. I1.7.2,11.7.3,11.13, and 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 j the SITE BOUNDARY is due to exposure incurred while conducting his farming activities. The '

l current tenant has estimated that he spends approximately 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> per year working in this area (Ref.11.5.5). Occupancy of areas within the SITE BOUNDARY is assumed to be averaged wer a  !

period ofone year.

Any reevaluation of assumptions should include a reevaluation of the occupancy period at the l locations of real exposure (e.g. a real individual would not simultaneously exist at each point of I maximum exposure). q 4.1.3.1 Total Dose to the Farmer from Gaseous Effluents The Total Dose to the farmer from gaseous effluents is calculated for the adult age group using the methodology discussed in Section 3. utilizing concurrent meteorological data at the farmer's ,

residence and historical metecrological data from Table 6.1 for activities within the SITE l

) 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

)

l occupancy of 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> / year factored into their value (Ref. I1.5.6).

The residence of the current tenant is located at a distance of 3830 meters in the SE s ,ctor. No meat l or milk animals or vegetable gardens were identined by the latest Land Use Census for this location. 1 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. For conservatism, it is acceptable to assume j that the ingestion pathways exist at this location. i l

it is assumed that food ingestion pathways do not exist within the SITE BOUND ARY. therefore the j gaseous effluents dose within the SITE BOUNDARY is due to pinme exposure from Noble Gases and the ground plane and inhalation pathways.

4.1.3.1.1 Direct Radiation Dose from Outside Storace 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 j radiation dose from outside storage tanks to a MEMBER OF THE PUBLIC (Ref. I1.6.14. I1.6.15 I1.6.16 and 11.6.17).

Direct radiation dose from the RWST to a MEMBER OF THE PUBLIC is determined at the nearest point of the Owner Controlled Area fence which is not obscured by significant plant structures, w hich is 450 meters from the RWST. i The RWST is a right circular cylinder approximately 12 meters in diameter,14 meters in height with a capacity of approxima :ly 1,514,000 liters (Ref. I1.6.17). The walls are of type 304 stainicss

.etect and have an average thi:kness of.87 cm (Ref. I1.14.1)

The direct radiation dose from the RWST is calculated based on the tank's average isotopic content and the parameters discussed above, considenng buildup and attenuation within the volume source.

l Appropriate methodology for calculating the dose rate from a volume source is given m TID-7004

" Reactor Shiciding Design Manual" (Ref. I1.17). The computer program ISOSHLD (Ref i1.18 I1.19 and 11.20) will normally be utihzed to perfonn this calculation.

l

MA ZZ-OlOO3 "

Res. 5 4.1.3.1.2 Direct Radiation Dose from the Reactor The maximum direct radiation dose from the Unit to a MEMBER OF THE PUBLIC has been g determined to be 7E 2 mrads/ calendar year, based on a point source of primary coolant N 16 in the 3 steam generators. This source term was then projected onto the inside stuface of the containment j dome, taking credit for shielding provided by the containment dome and for distance attenuadon. g )

No credit was allowed for shiciding by other structures or components within the Containment gI Building. The number of gammas per second was generated and then converted to a dose rate at the 4 given distance by use of ANSI /ANS-6.6.1, " Calculation and Measurement of Direct and 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 ll 5

to be 367 meters, which is approximately the closest point of the boundary of the Owner Controlled '

Area fence which is not obscured by significant plant structures (Ref. I1.14.3).

The maximum direct radiation dose from the Unit to the farmer is thus approximately 9E 3 mrads 5 per year, assuming a maximum occupancy of 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> per year.

4.1.3.1.3 Direct Radiation Dose From On-Site Storace Of Low Level Radioactive Waste The on-site storage area for radioactive wastes is located Plant Southwest of the radwaste building and consists of a concrete pad enclosed by a fence. The storage area is bounded on two sides by the radwaste building. The area is also partially bounded on a third side by the Discharge monitonng tanks dike system. The radioactive wastes are stored in this area using high integrity containers h

M (HIC) inside Onsite Storage Containers (OSC) and LSA type storage containers. The HIC has the highest potential for containing signiDeant amounts of radioactive material, and constitutes the only

- potentially significant source of direct radiation from on-site radioactive waste storage.

g Direct radiation dose from the HICs to a MEMBER OF THE PUBLIC is determined at the nearest j point of the Owner Controlled Area fence wbich is not obscured by significant plant structures.

g The HICs typically are right circular cylinders approximately 1.7 meters in diameter and 1.8 meters M in height. The HICs are stored inside OSCs which typically are constructed of concrete with additional shielding as necessary to minimize external doses. The individual parameters (e.g.. 3 dimensions, shielding material, etc.) for each OSC will be accounted for in the calculations. g The direct radiation dose from the On-Site Storage area is the summation of the individual calculated HIC doses based on the HIC isotopic contents and the OSC design parameters.

considering buildup, attenuation, and shielding. Appropriate methodology for calculating the dose rate is given in Safety Analysis Calculations ZZ-293 and ZZ-310. The computer program MICROSHIELD (Ref. I1.24) will normally be utilized to perform this calculation.

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5. RADIOLOGICAL ENVIRONMENTAL MONITORING l 5.1 DESCRWTION OF THE RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM The Radiological Emironmental Monitoring Program is intended to act as a background data base i for preoperation and to suppicment the radiological effluent release monitoring program during plant operation. Radiation exposure to the public from the various specific pathways and direct radiation l 5

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 reasons. Efforts are made to obtain all required samples within the required time frame. Any deviation (s) in sampling frequency or location is documented in ti e Annual Radiological Environmental Operating Report.

Sampling, reporting, and analytical requirements are given in Tables 9.11 A,9.11 B. and 9.1 l C.

Airborne, waterborne, and ingestion samples collected under the monitoring program are analyzed by an independent, third-party laboratory. This laboratory is required to participate in the l Emironmental Protection Agency's (EPA) Environmental Radioactivity Laboratory intercomparison Studies (Crosscheck) Program or an equivalent program. Participation includes all of the determinations (sample medium - radionuclide combination) that are offered by the EPA and that are also included in the monitoring program.

5.2 PERFORMANCE TESTfNG OF ENVIRONMENTAL THERMOLUMINESCENCE DOS! METERS j Thermoluminescence Detectors (TLD's) used in the Emironmental Monitoring Program are tested l for accuracy and precision to demonstrate compliance with Regulatory Guide 4.13 (Ref. I l.16) i 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 Co40. (Ref. I1.14.10)

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6. DETERMINATION OF ANNUAL AVERAGE AND SHORTTERM ATMOSPHERIC I DISPERSION PARAMETERS 6.1 ATMOSPHERIC DISPERSION PARAMETERS The values presented in Table 6.1 and Table 6.2 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.

l 6.1.1 Lone-Term Disnersion Estimates l The variable trajectory plume segment atmospheric transport model MESODIF !! (NUREG/-CR-l 0523) and the straight-line Gaussian dispersion model XOQDOQ (NUREG/CR2919) were used for l 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.

I1.6.12).

The Unit Vent and Radwaste Building Vent releases are at elevations of 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 climinates

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. I1.5.2).

6.1.2 Determination of Lone-Term Disocrsion Estimates for Soccial Recentor Locations

! Calculations utilizing the PUFF model were performed for 22 standard distances to obtain the l desired dispersion parameters. Dispersion parameters at the SITE BOUNDARY and at special l receptor locations were estimated by logarithmic interpolation according to (Ref. I 1.6.13):

o s

.) X = X, (d/di )" (6.1)

Where

1 B =In (X2 /Xi )/In (d2 /d)-

i X,X=

I 2 Atmospheric dispersion parameters at distance di and d2, respectively, from the sottrCe.

l The distances di and d 2were selected such that they satisfy the relationship.

! dg < d < d2 l 6.1.3 Short Term Dispersion Estimates Airborne releases are classified as short term if they are less than or equal to 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> during a calendar year and not more than 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> in any quarter. Short tenn dispersion estimates are determined by multiplying the appropriate long term dispersion estimate by a correction factor (Ref.

I1.9.1 and 11.15.1): i F = {T, / T, } * (6.2)

Where:

= The total number of hours of the short term release.

T3 Ta = The total number of hours in the data collection period from which the long term diffusion estimate was determined (Refer to Section 6.1).

j Values of the slope factor (S), are presented in Table 6.3.

1 Short term dispersion estimates are not applicable to short term releases which are sufnciently k

random in bodi time of day and duration (e.g., the short term release periods are not dependent solely on atmospheric conditions or time of day) to bc represented by the annual average dispersion condidons (Ref. I1.8.1).

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APA-ZZ-01003 Rev. 5 6.1.3.1 The Determination of the Slooe Factor (S)

The general approach employed by subroutine PURGE of XOQDOQ (Ref. I1.15.1) was utihzed to produce values of the slope of the (X/Q) curves for both the Radwaste Building Vent and the Unit Vent. However, instead of using approximation procedures to produce the 15 percentile (X/Q) values, the 15 percentile (X/Q) value for each release and at each location was determined by ranking all the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ((X/Q)3) values for that release and at that location in descending order. The (X/Q)1 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.

The intermittent release (X/Q) curve was constructed using the calculated 15 percentile (X/Q)j and h 5

its corresponding annual average (X/Q)a. A graphic representation of how the computational procedure works is illustrated by Figure 4.8 of reference 11.15.1. The straight line connecting these points represents (X/Q)g values for intermittent releases, ranging in duration from one hour to 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br />. The slope (S) of the curve is expressed as:

-log ((X / Q), /(X / Q),)

S= (6.3) log (T, / T,)

or

-(log (X / Q), - log (X / Q),)

S= (6.4) log T, - log T, 6.1.4 Atmospheric Disnersion 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 farmer from gaseous effluents while he conducts farming activities within the SITE BOUNDARY.

For the purpose of these calculations, it was assumed that all of the farmer's time, approximately 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> per year, is spent on croplands within the SITE BOUNDARY, and that his time is divided evenly over 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 which are presented in Tables 6.1 and 6.2. These dispersion parameters therefore represent the distributed activities of the farmer within the SITE BOUNDARY and h.s estimated occupancy period.

6.2 ANNUAL METEOROLOGICAL D ATA PROCESSING The annual atmospheric dispersion parameters utilized in the ci !ation of doses for demonstration of compliance with the numerical dose objectives of 10 CFR SC xmdix 1, are determined using computer codes and models consistent with XOQDOQ (Ref.1 These codes have been validated and verified by a qualified meteorologist prior to implementad z. Multiple sensors are udlized to ensure 90% valid data recovery for the wind speed, wind direction, and ambient air temperature .

parameters as required by Regulatory Guide 1.23. The selection hierarchy is presented in Table 6.5.

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' Rev 5 TABLE 6.1 IIIGIIEST ANNUAL AVERAGE ATMOSPIIERIC DISPERSION PARAhEERS (a)

UNIT VENT DISTANCE l OCATION(b) SECTOR XQ X)Q (METERS) X/Q DECAYEDr DECAYEDr UNDEPLETED DQ

'J DEPliTED

. ' ~ .%.246 d2i& A::!%2?

- Aasens;4asabhidnijMhsi46%sigg{)emEMELM2sCgtnmagn,,aq(m- )n&AsJhd -

(sechn ) (sechu (sechn )

SITE DOUNDARY NNW -2200 1.0E4 9.9E 7 8.5E-7 43E-9 Nearnt Residence (c)(d) NNW 2864 6.RE-7 6.8E-7 3.7E 7 2.6L9 l Farmer's Resklence SE 3830 2.3E-7 2.3E-7 2.lE 7 1.159 Tarming Areas within the N/A ,

N/A 2.lEr7 2.lE 7 1.9E-7 Site Doundary(c)(e) 1.I E-9 (a) Values given are from FSAR Table 2.3-82 i 1

l (h) Data frcm 1993 land Use Census (c) Values derived frorn FSAR Table 23-83. esing tFe methodology presemed in Equation (6.1)(Ref.11.3.6) I i

l (d) All pathways are assumed to exist at the location of the nearest residenL (e)

These values were derived for a narrow scope apphcatiort Extreme caution should be exercised when determining their suitability for use in other applicatiers.

Iludding Shape Parameter (C) = 0.5 (Ref. I1.3.3)

Vertical lleight of liighed Adjacem Bui! dang (V) = 66 45 meters (Ref. I1.5.3)

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APA-ZZ-01003 Rev.5 TABLE 6.2 IIIGIIEST ANNUAL _ AVERAGE ATMOSPIIERIC DISPERSION PARAMETERS (a)

RADWASTE BUILDING VEW DISTANCE X9 X/Q l OCATION (b) SECTOR (METERS) X/Q DECAYEDI DFCAYEDI Da UNDEPLETED DEFLETED

~ ~ ~ - - - ~ , - . . m. .___

3 (sec/m )

3 (sectm )

3_

(sec&n ) (m-2)

SITE IV)t'NDARY NNW 2200 1.3E4 1.3E4 1.1E4 4.3 E-9 Nearest Residence (c)(d) NNW 2864 8.7E-7 8.7E-7 7.2L7 2.6E-9 l Farmer's Residence SE 3830 3.0E-7 3.007 2.4L7 1. l E-9 Farnung Areas Within N/A N/A 2.907 2.9E-7 2.6L7 1. 8 E-9 Site ikwndary (c)(c)

(a) Values given are frorn FS AR Table 2.3-84

~

l (b) Data from 1993 Imd Use Census

(:) Values derived from FS AR Table 2.3-81, using the enethodology preser.ted in Equation (6.1)(Ref. I1.5.6) l (d) All pathways are assurned to exist at the location of the nearest resident.

(c) These values were derived for a narrow secte application Extreme caution should te exercised when determining their suitability for use in other applications.

Duildmg Shape Farameter (C) = 0 5 (Ref. I l.3.3)

Vertical licig!d of flighest Adjacent Duilding (V) = 19.96 meters (Ref 11.5.3) 4

-4 9-W W W M E% M' W W W W M M M MM M M M' W

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APA-ZZ-01003 Rev. 5 IABLE 6,3 SHORT TERM DISPERSION PARAhETERS fa)(q)

Slope Factor (s)

Location (b) Sector Distance Unit Vent Radwaste Building Vent .-

Site Boundary S 1300 .328 .320 Nearest Residence (d) NNW 2865 .264 . 268 (a) Reference 11.5.3 l (b) Data from 1993 Land Use Census

-(c) Recirculation Factor - 1.0

- l (d) All pathways are assumed to exist at the location of the nearest resident.

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APA-ZZ-01003 Rev.5 TABLE 6.4 APPLICATION OF ATMOSPIIERIC DISPERSION PARAhETERS l O D_OSE PATilWAY DISPERSION PARAMETER CONTROLLING AGE GROUP REC CONTROLLING LOCATION l Noble Gas, Beta Air x/Q, decayed /undepleted -

9.7 Site Boundary (2.26 day halflife) l Noble Gas, Gamnu Air x/Q, decayed /undepleted -

9.7 Site Boundary (2.26 day halflife) l Noble Gas Total Body x/Q, decayed /undepleted -

9.6 Site Boundary (2.26 day halflife) l Noble Gas, Skin x/Q, decayed /undepleted -

9.6 Site Boundary (2.26 day halflife) l Ground Plane Deposition D/Q -

9.8 Nearest Resident l Inhalation x/Q, decayed / depleted Child 9.8 Nearest Resident (8 day halflife) 9.6 Site Boundary l Vegetation D/Q* Child 9.8 Nearest Resident l Milk D/Q* Child 9.8 Nearest Resident l Meat D/Q* Child 9.8 Nearest Resident l

• For 11-3. VO. decayed / depleted is used instead of D/Q (Ref I 1.11.1).

m. - _ - - - - - - . -

m m m emm m m m M W W M M M M M M

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l TABLE 6.5 METEOROLOGICAL DATA SELECTION HIEARCHY l

i l Parameter Primary First Second Third >

Alternate Alternate Alternate l

, Wind Speed - 10m Pri 10m Sec 60m Pri 90m Pri Wind Direction 10m Pri 10m Sec 60m Pri 90m Pri l

l l Air Temperature 10m Pri 10m Sec Wind Variability 10m Pri 10m Sec 60m Pri 90m Pri Temp Different 6010m Pri 90-10m Pri 90-60 Pri

. Dew Point 10m Pri l

l Precipitation im Pri g (a) Priindicates primary tower

} (b) Sec indicates secondary tower i

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Rev. 5 t

7. REPORTING REOUTREMENTS 7.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (CTSN 2804)

Routine Annual Radiological Emironmental Operating Repon covering the operation of the urut during the previous calendar year shall be submitted prior to May I of each year.

The Annual Radiological Emironmental Operating Repon shall include summaries, interpretations.

and an analysis of trends of the results of the radiological environmental surveillance activities for the repon period, including a comparison with preoperational studies, with operational controls and with previous environmental surveillance repons, and an assessment of the observed impacts of the l

plant operation on the environment.

The reports shall include the results of Land Use Census required by REC 9.12. It shall also include a listing of new locations for environmental monitoring identified by the Land Use Census pursuant to REC 9.12.1.

The Annual Radiological Environmerital Operating Repon shall include the results of analysis of all radiological environmental samples and of all emironmental radiation measurements taken during the period pursuant to the ODCM, as well as summarized tabulated results of these analyses and l 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 repon, the repon 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 repon.

The repons shall also include the following: a summary description of the radiological emironmental monitoring program; at least two legible maps

• covering all sampling locadons keyed l

5 to a table giving distances and directions from the centerline of one reactor; the results oflicensee

}

, y participation in the Interlaboratory Comparison Program and the corrective action being taken if the specified program is not being performed as required by 9.13.1; reasons for not conducting the Radiological Emironmental Monitoring Program as required by 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.ll-B, but are not the result of the plant emuents, pursuant to 9.11.1; and discussion of all analyses in which the LLD required by Table 9.11-C was not achievable.

7.2 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (CTSN 2805)

Routine Annual Radioactive Emuent Release Reports covering the operation of the unit dunng the previous calendar year shall be submitted prior to May I of each year.

The Annual Radioactise Emuent Release Report shall include a sumrnary of the quanudes of radioactive liquid and gaseous emuents released from the unit as outlined in Regulatory Guide 1.21, "Measunng, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases Gaseous Emuents from Light-Water-Cooled Nuclear Power Plants, " Revision 1, June 1974, with data summanzed on a quanerly basis following the format of Appendix B thereof.

The Annual Radioactive Emuent Release Report shall include an annual summary of hourly meteorological data collected over the previous calender 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 ofjoint frequency distnbuuon of wind speed.

wind direction, and atmospheric stability l

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• One map shall cover stauons near the SrfE DOUNDARY, a wcond shall include the more disuun stauons

" In lieu of subnussion mth the Annual Radioacuve Emuent Release Repon, Union Electnc has the opuon of retauung Llus summary of required meteorological data on site in a file that shall be provided to the NRC upon request

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APA ZZ-01003..

Rev 5 l

l' This report sha!! also include an assessment of the radiation doses due to the radioactive liquid and gaseous emuents released from the unit during the previous calendar year. This report shall also include an assessment of the radiation doses from radioactive liquid and gaseous ef!1uents to l MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Technical l Specifications, Figures 5.1-3 and 5.1-4) during the report period using historical average atmospheric conditions. All assumpdons used in making these assessments, i.e., specific activity, exposure time and location, shall be included in these seports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous emuents, as determined by

.l sampling frequency and measurement, shall be used for determuung the gaseous pathway doses.

( Assessment of radiation doses shall be performed in accordance with the methodology and pr==1 :e in the OFFSITE DOSE CALCULATION MANUAL (ODCM). -

( The Annual Radioactive Emuent Release Report shall include an assessment of radiation doses to j . the most likely exposed MEMBER OF 'IEE PUBLIC from reactor releases and other nearby i uranium fuel cycle sources, including doses from primary efIluent pathways and direct radiation, for

- the previous calendar year to show conformance with 40 CFR Pan 190, " Environmental Radiation l Protection Standards for Nuclear Power Operation " Doses to the MEMBER OF THE PUBLIC shall l be calculated using the methodology and parameters of the ODCM.

, l The Annual Radioactive Emuent Release Reports shall include a list and description of unplanned / l l releases from the site to UNRESTRICTED AREAS of radioacuve matenals in gaseous and liquid i emuents made during the reporting period.

]

l The Annual Radioactive Efiluent Release Reports shall include a summary description of any major l changes made during the year to any Liquid or Gaseous Treatment Systems,' pursuant to Section 10.1, It shall also include a listing of new locations for dose calculations identified by the Land Use Census pursuant to REC 9.12.1.

Reporting requirements for changes to Solid Waste Treatment Systems is addressed in

. APA-ZZ-01011, PROCESS CONTROL PROGRAM (PCP).

l The Annual Radioactive Emuent Release Reports shall also include the following information: An explanation as to why the inoperability ofliquid or gaseous emuent monitoring instrumentation was

'l not corrected within the time specified, and a description of the events leading the liquid holdup tanks or gas storage tanks exceeding the limits of Technical Specification 3.11.1.4 or 3.11.2.6.

The Annual Radioactive Effluent Release Reports shall include as part of or submitted concurrent with, a complete and legible copy of all revisions of the ODCM that occurred during the year pursuant of Technical Specification 6.14. -

Solid Waste reporting is addressed in APA-ZZ-01011, PROCESS COKfROL PROGRAM (PCP).

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APA ZZ-01003 Rev 5

8. IMPLEMENTATION OF ODCM METHODOLOGY (CTSN 2791)

The ODCM provides the mathematical relationsidps used to implement the Radioactive EfIluent Controls. For routine efIluent release and dose assessment, computer codes are utilie.ed to implement the ODCM methodologies. These codes are evaluated in accordance with the requirements of HDP-ZZ-04500," Rad / Chem Computer Systems Conduct of Operations", to ensure that they produce results consistent with the methodologies presented in the ODCM. Procedures which implement the ODCM methodology are contained in the Plant Operating Manual.

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9. RADIOACTIVE EFFLUENT CO?iTROLS (REC)

a. The terms in this secdon that appear in capitalized type are defined in Technical f Specificadons.

b. All frequency notations are per Table 1.1 of Technical Specificadons.

( 9.0.1 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.

9.0.2 Noncompliance with a Control shall exist when the requirements of the Control and associated j 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.

)

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APA ZZ-Ol003 Rev. 5 9.1 RADIOACTIVE LIOUID EFFLUETN MONITORING INSTRUMENTATION I

REC 9.1 has been relocated to Section 16.3.3.6 of the FSAR.

The following should be used to cross-reference REC 9.1 surveillances to the appropriate section of the FSAR:

Monitor REC Number FSAR Section Tvne IIB-RE-18 9.1.2.1-1.a 16.3.3.6.1-1.a Liquid rad monitor g

BM-RE-52 9.1.2.1-1.b 16.3.3.6.1 1.b Liquid rad monitor W HB-FE-2017 9.1.2.1 2.a 16.3.3.6.1 2.a Flow element BM-FE-0054 9.1.2.1-2.b 16.3.3.6.1-2.b Flow element FE-DB 1006,1101 9.1.2.1 2.c 16.3.3.6.1-2.c Flow element 9.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION REC 9.2 has been relocated to Section 16.3.3.7B of the FSAR.

The following should be used to cross-reference REC 9.2 surveillances to the appropriate section of the FSAR:

Monitor REC Number FSAR Section Type Unit Vent GT-RE-21B 9.2.2.1 1.a 16.3.3.7b.1 1.a Gas g

< GT-RE-21 A & B 9.2.2.1-1.c 16.3.3.7b.1-1.c lodine sampler g GT-RE-21A & B 9.2.2.1-1.b 16.3.3.7b.1-1.b Pa2ticulate sampler GT-RE-21 A & B 9.2.2.1-1.d 16.3.3.7b.1-1.d Unit Vent flow rate g GT-RE-21 A & B 9.2.2.1-1.e 16.3.3.7b.1 1.c Particulate and Radioiodine e Sampler flow rate Monitor Containment Purge g GT-RE-22 9.2.2.1-2.a 16.3.3.7b.1 2.a Gas 3 GT-RE-33 .

GT-RE-22 9.2.2.1-2.c 16.3.3.7b.1 2.c lodine sampler g GT-RE-33 g GT-RE-22 9.2.2.1 2.b 16.3.3.7b.1 2.b Particulate sampler GT-RE 33 g

GT-RE 22 9.2.2.1 2.d 16.3.3.7b.1-2.d Containment purge flow 3 GT-RE-33 rate GT-RE-22 9.2.2.1-2.c 16.3.3.7b.1 -2.e Particulate and Radioactise GT-RE-33 Sampler flow rate Monitor j Radwase Building Ventillation GH-RE-10B 9.2.2.1 3.a 16.3.3.7b.1-3.a Gas GH-RE-10A & B 9.2.2.1-3.c 16.3.3.7b.1 -3.c lodine sampler GH.RE 10A & B 9.2.2.1-3.b 16.3.3.7b.1 3.b Particulate sampler GH-RE 10A & B 9.2.2.1 3.d 16.3.3.7b. ! 3.d Radwaste Building Vent E

Flow rate g

GH RE-10A & B 9.2.2.1 3.c 16 3.3.7b.1-3.c Particulate and Radioactive Sampler flow rate Monitor

APA-ZZ-o luo3 Res 5

')

9.3 LIOUID EFFLUENTS CONCENTRATION 9.3.1 Controls (CTSN 41834) 9.3.1.1 The concentration of radioactive material released in liquid efIluents to UNRESTRICTED AREAS (see Technical Specifications, Figure 5.1-4) shall be limited to the concentration specified in 10 CFR Part 20.120.601, 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 10-4 microcurie /ml total activity.) (CTSN 4160)

APPLICABILITY At all times.

. 1 l' ACTION: l

a. With the concentration of radioactive material released in liquid effluents to  ;

UNRESTRICIED AREAS exceeding the above limits, immediately restore the' l

concentration to within the above limits.  ;

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

9.3.2 Surveillance Reauirements 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 REC 9.3.1.1.

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APA-ZZ-01003 Res. 5 TABLE 9 3 A RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM l

l

^ .,s. . . .j : - . -

i .:.M% . a . <. ...,;...,....., .-.>.,.........:..<g.g:

.. .,.4 ,

I'LlQUID REMASEf.  ?' ? S AMP 11NO .l . . -

':-MIND 4UMi 's.TYFE OF J . LOWER LIMTT.

. 0FDETECTION l a" ~ iTYPS :s ' 'l Gj;TREQUENCY(7)$

L . ?. ANALY315.I... .' ACTIVITY

l. -  ;-3 pgsg;g %p@;t f- gg- <

j FREQUENCQ 5 ANALYSIS ? l (LLD)(1) ' ' l gy <

e.p + ej .wf f

ik ~ ~ ' ; -@ .' (peg)

P w

1. Batch Wasta Prirsipalg)amma 4 Release Tanks (2) Each Bae:h Emitters 5x10 1 x 10 4

1 131

a. Discharge Dissolved and Ia10 5 Monitor Entruned Cases Tank (Gartma Enuttes)

P Each Batch H-3 1x10 5 M Gross Alpha I x10'7 Composite (4)

Q Sr-89,Sr90 5 x 10-8

.p Cornposite(4)

Fe-55 lx10 4 Daily Principal Garnma 5x10'7

2. Continug( s '*

Releases ) Eminers(3) l 131 1x10 4

Steam Dissolved and lx10'3 Generator Entrained Gases l

Blowdown (Gamma Enumers)

Daily 0)

Grab Sarnple H3 1 x10*3 l

i

)

M Grosa Alpha I x10-7 i

Composite (4)

Q St-89, Sr 90 5 x 10-8 Comp mte(4)

Fe 55 lx10 4 I

I APA ZZ-01003

. Rev. 5 TABLE 9.3-A (Cont'd)

. TABLE NOTATIONS (1) The LLD is described in Attachment 1.

(2) A batch release is the discharge ofliquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed a method described in the ODCM to assure representative sampling.

(3) ' The principal gamme emitters for which the LLD control applies include the fol' lowing radionuclides: Mn 54, Fe 59, Co-58, Zn45, Mo-99, Cs-134, Cs 137 Cc-141, and Cc 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 Annual Radioactive Emuent 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 ofliquid sampled is proportional to the quantity ofliquid waste discharged and in which the method of sampling employed results in a specium 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 emuent release.

(5) A continuous release is the discharge ofliquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release.

(6) Samples shall be taken at the initiation of emuent flow and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

.) thereafler while the release is occurring. To be representative of the liquid emuent, the sampic volume shall be proportioned to the emuent stream discharge volume. The ratio of  ;

sample volume to emuent discharge volume shall be maintained constant for all samples taken for the composite sample.

{

(7) Samples shall be representative of the emuent release.

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APA-ZZ-01003 Rev. 5 9.4 DOSE FROM LIOUID EFFLUENTS 9.4.1 Controls (CTSN 41834) 9.4.1.1 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Technical Specifications, Figure 5.1-4) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 miems to the whole body and to 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: (CTSN 1161) w

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, 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: (2) the results of radiological impact on finished drinking water supplies

• with regard to the requirements of 40 CFR ? art 141, Clean Drinking Water Act.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

9.4.2 Surveillance Reauirements 9.4.2.1 Cumulative dose contributions from liquid effluents for the current calendar quarter and the carrent calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

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• The requuements of ACTION a (1) and (2) are apphcable only if dnnLing water supply is taken from the recemng water body within 3 miles of the plant discharge. In the case of rner. sited plants this is 3 nules downstream only

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APA-ZZ-01003 Rev. 5 l

9.5 LIOUID RADWASTE TREATMENT SYSTEM 9.5.1 Controls (CTSN 41834) 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 Specifications. Figure 5.1-4) would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31 day period.

APPLICABILITf. At all times.

AMION: (CTSN 1161)

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 e 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. j

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

9.5.2 Suneillance Reauirements

}

, 5 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.

9.5.2.2 The installed Liquid Radwaste Treatment System shall be considered OPERABLE by meeting REC 9.3.1.1 and 9.4.1.1.

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1 APA ZZ-01003 Rev. 5 9.6 G ASEOUS EFFLUEKrS DOSE RATE 9.6.1 Controls (CTSN 41834) 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 Specifications, 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 or .

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- g g

lives greater than 8 days: le.ss than or equal to 1500 mremstyr to any organ.

APPLICABILITY: At all times.

4 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 Technical Specifications 3.0.3 and 3.0.4 are not applicable.

9.6.2 Surveillance Recuirements 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.

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9.6.2.2 The dose rate due to lodine-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 j and performing analyses in accordance with the sampling and analysis program specified in Table s' 9.6- A.

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APA-ZZ-01003 Rev.5 MINIMUM ANALY315 .

. IDWERLIMITOF CA5FOUS R.E15ASE TYPE SAKG' LING FREQUENCY (9) . FREQUENCY TYl40F ACTIVITY ANAI.YSIS :.

- ,,, DElllCTION(LLD)(1)

, :4 W l Wace Gas (Nesy P P PrmeipalGenma EmuumUI Iale d Tar

• Ed Tank Oreb samrte Each Tank l 2 Centa:nmen hrse cm Vens P P Principal Genens Ernicars(2) lated Each PURGE Each PURGE Grab sample M Il-3(oxsde) I 10 4 l 3 p , ym Ob k[M*) kdII PnneipalOenann Emumm(2) 1:10 4

Grab sample k/4) H-1(oule) 1x104 4 Sp Furl Doddirs Exhaust Ef5) M PrincipalGemme EnumeruU) Int 0 Grab $ ample W W45) H-3 (aule) lated Grab Sampic 5 Pedweste Budoma Vent M M Principal Genuna EmismU) 4 Im:0 G oh sample i 6 A!! Retcaw Tmes as lated m 1. 2,3. 4 Cernmuous(6XI) g(7) g,g 3, . Iml0-12 and 5 ahne Chartcal Sample I-133 1m10-10 Centmucus(6X 8) H4I) Prmeipal Gamma Enunm(2) g,go-Il Pameulsee semple Cenimuous(6Xt) M Gross Alpha ImIO~II Componne Partw.utme Sample CentmunmNI) Q Sr-89. St-90 ' lat0'II Componde Particulaic Sempte

= - . .. _ _ _ . ._ - _ - . ____ __-__________-__________ - _________________ _______ __ - ____ - - _____ -___-- -________-__ _ _ _ _

, APA-ZZ411003 I

Rev. 5 TABLE 9.6-A (Cont'd)

TABLE NOTAT10NS (1) The LLD is described in Attachment 1.

(2) The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xc 133, Xe 133m, Xc.135, and Xe 138 in noble gas releases h

W l and Mn 54, Fe 59, Co-58, Co-60,2n-65,1 131, Cs 134, Cs 137, Cc 141, and Cc 144 in iodine and particulate releases. This list does not mean that only these nuclides are to be g considered. Any nuclide which is identified in the sample and which is also listed in the g ODCM gaseous emuents dose factor tables, shall be analyzed and reported in the Annual Ef!!uent Release Report.

(3) If the Unit Vent noble gas monitor (GT-RE-21B) shows that the effluent aethity has increased (relative to the pre-transient activity) by more than a factor of 3 following a reactor shutdown, startup, or a thermal power change which exceeds 15% of the rated thermal power within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, samples shall be obtained and analyzed for noble gas, particulates and

! iodines. This sampling shall continue to be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for a period l

=

of 7 days or until the Unit Vent noble gas monitor no longer indicates a factor of 3 increase in Unit Vent noble gas activity, whichever comes first.

g (4) Tritium grab samples shall be taken and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the W refueling canal is flooded.

(5) Grab samples need to be taken only when spent fuel is in the spent fuel pool.

l (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 REC 9.6.1.1, l ) 9.7,1,1, and 9.8.1.1.

(7) Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing, or removal from the sampler. When sampling is performed in accordance with footnote 3 (above), then the LLD may be increased by a factor of 10.

(8) Continuous sampling of the spent fuel building exhaust needs to be performed only when spent fuel is in the spent fuel pool.

(9) Samples shall be representative of the efDuent release. 3 I

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APA-ZZ-01003 Rev. 5

- DOSE - NOBLE GASES 9.7 9.7.1 Controls (CTSN 41834)

I 9.7.1.1 The air dose due to noble gases released in gaseous efIluents, from each unit, to areas at and beyond the SITE BOUNDARY (see Technical Specifications Figure 5.1 3) shall be limited to the following.

l a. During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less 5 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: (CTSN 1161)

I a. With the calculated air dose from radioactive noble gases in gaseous efIluents 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 l 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.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

9.7.2 Su_rveillance Reauirements I. 9.7.2.1 Cumulative dose contributions for the current calendar quarter and cunent calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

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APA-ZZ-o loo 3 Ret 5 9.8 DOSE -IODINE 131 AND 133. TRIT!Uht AND RADIOACTIVE MATERIAL IN PARTICULATE FORM 9.8.1 Controls (CTSN 41834) 9.8.1.1 The dose to a MEMBER OF THE PUBLIC from lodine 131 and 133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, a to areas at and beyond the SITE BOUNDARY (see Technical Specifications, Figure 5.1-3) shall be g limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrems to any organ, and g

b. During any calendar year: Less than or equal to 15 mrems to any organ. W APPLICABILITY: At all times.

ACTION: (CTSN 1161)

a. With the calculated dose from the release ofIodine-131 and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of g the above limits, prepare and submit to L:lc Commission within 30 days, pursuant to g 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 l W

compliance with the above limits.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

9.8.2 Surveillance Recuirements 9.8.2.1 Cumulative dose contributions for the current calendar quarter and current calendar year for lodine-131 and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall 3 be deter nined in accordance with the methodology and parameters in the ODCM at least once per g 31 days.

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APA-ZZ-01003 Rev 5 9.9 GASEOUS RADWASTE TREATMENT SYSTEM {

9,9,1 Controls (CTSN 41834) 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

/

cach 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. 0.4 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 Specifications 6.9.2, a Special Report that includes the following infomiation:

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 provision of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

9.9.2 Surveillance Reauirements 9.9.2.1 Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall bc 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 REC 9.6.1.1 and 9 7.1.1 or 9.8.1.1.

)

4

APA-ZZ41003 Rev. 5 I

9.10 TOTAL DOSE 9.10.1 Controls (CTSN 41834) 9.10.1.1 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 mrems to the whole body or any organ, except the thyroid, which shall be lirtuted to less than or equal to 75 mrem.

APPLICABILITY: At all times.

ACTION:

a. With the calculated doses from the release of radioactive materials in liquid or gaseous efIluents exceeding twice the limits of REC 9.4.1. la, 9.4.1.1b. 9.7.1 la. 9.7.1. I b. 9.8 1 1a.

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 REC 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 6.9.2, a Special Report that defines the corrective action to be taken to reduce subsequent release to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as def'med in 10 CFR 20.2203, shall include an analysis that estimates the g radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle g sources, including all efiluent 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 resuldng 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 E

. CFR Part 190. Submittal of the report is considered a timely request, and a variance is 3 granted until staff action on the request is complete.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

9.10.2 Surveillance Recuirements 9.10.2.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with REC 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 requirements is applicable only under conditions set forth in ACTION a. of REC 9.10.1.1.

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APA ZZ-01003 Res. 5 9.11 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 9.11.1 Controls (CTSN 41834) 9.11.1.1 The Radiological Environment Monitoring Program shall be conducted as specified in Table 9. I l- A.

APPLICABILITY: At all times.

ACTION:

a. With the Radiological Environmental Monitoring Program not being conducted as speciDed l in Table 9.11 A, prepare and submit to the Commission, in the Annual Radiological j Environmental Operating Report required by Technical Specification 6.9.1.6, a descripdon of the reasons for not conducting the program as required and the plans for preventing a recurrence.

b. With the level of radioactivity as the result of plant emuents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 9.11-B w hen averaged over any calendar quarter, prepare and submit to the Commission within 30 days.

pursuant to Technical Specification 6.9.2, a Special Report that identiEes the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive emuents so that the potential annual dose' to a MEMBER OF THE PUBLIC is less than the calendar year limits of REC 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 repon shall be submitted if:

concentration (1) , concentration (2) + . 2 1.0 reporting level (1) reporting (2)

)' When radionuclides other than those in Table 9.11-B are detected and are the result of plant emuents, this seport 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 REC 9.4.1.1,9.7.1.1 or 9.8.1.1. This repon is not required if the measured level of radioactivity was not the result of plant emuents; however, in such an event, the condition f shall be reported and described in the Annual Radiological Environmental Operating Report, required by Technical Specification 6.9.1.6.

c. With milk or fresh leafy vegetab!c samples unavailable from one or more of the sample locations required by Table 9.ll ,A, identify specific locations for obtaining replacement samples and add them within 30 days to the Radiological Environmental Momtonng Program".The specine locations from which samples were unavailable may then be deleted from the monitoring program. In the next Annual Radiological Environmental Operating Repon include the revised figure (s) and tables reflecting the new sample locadon(s) with supporting information identifying the cause of the unavailability of samples and justifying the selection of new location (s) for obtaining samples.

d. When LLDs specified in Table 9.11-C are unachievable due to uncontrollable circumstances, (such as background fluctuations, unavailable small sample sizes the presence ofinterfering nuclides, etc.) the contributing factors shall N identified and described in the Annual Radiological Emironmental Operating Report.

c. The prosisions of Technical Specincations 3.0.3 and 3.0.4 are not applicable.

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• The methodology and parameters used to estimate the potential annual dose to a MEMBER OF Tl[E PUBLIC shall be indsated in this report

! " Excludmg short term or temporary unavailability.

70-1

APA Z2-Ol003 E m

Rev. 3 9.11.2 Surveillance Reauirements 9.11.2.1 The radiological environmental monitoring samples shall be collected pursuant to Table 9.11 A and l shall be analyzed pursuant to the requirements of Table 9.11-A and the detcetion capabilitics required by Table 9.11 C.

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APA-ZZ-01003 Rev.5 TABLE 9.11-A RADIOLOGICAL ENVIRONMEhrTAL MONITORING PROGRAM NUMilER OF REPRESENTATIVE SAMPLES SAMPilNo AND EATOSURE PATI!WAY AND5AM COLLECTION 1TPE AND FREQUENCY AND OR SAMPLE tOCATION ) FREOUENCY OF ANALYSIS

1. Drea Radiation (2) Forty routine monitoring rtations either with two or more Quarterly Gamma done quarterly dosimeters or with one instrument for measuring and recording dose rate continuously, placed as follows:

An inner ring of sixteen stations, one in each meteorological sector in the general area of the SITE DOUNDARY:

An outer ring of etations, one in each meteorological sector in the 6 to 8-km (3 to 5 mile) range from the site; and Eight stations to be placed in special interest areas such as population centers, nearby residences, schools, and in one or two areas to serve as contrel stations.

2. Auborne Radioiodine and Sarrples frorn five locations; Continuous sangler operation with sample Radionadme Canister: I-131 analysis meckly.

Particulates collection weekly, or more frequersly if reqmrod Three samples fan close to the three SITE BOUNDARY ~ by dust loading Par'* Saneler Gross beta radioactivity locations, in difTerent sectors, of the highest calculated annual analyuss follom filter change: d) and garnma average ground level D/Q.

isotopic analysis ( ) of conposite (by location) quarte $

One sample from the vicinity of a - ~yhaving the highest calculated annual average ground level D/Q.

One sample from a control location, as for er mple !S to 30 km (j)0

( . to 20 mile) distare and in the least prevai1nt wind di,iction

i

%J APA-ZZ-01003 Rev.5 TABLE 9.11-A (Cont'd)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGM AM NUMITER OF REPRESENTATIVE SAMPtIS SAMPLING AND EN1'Ost*RE PATilWAY AND SAMPL COLIICT!ON TYPE AND FREQUENCY AND OR S AMPLE LOCATIONS ({) FREOUENCY OF ANALYSIS

3. Waterhome a Surface (6) One sample upstream Canuna isotopi[) and tritium One sample downstream Composite I-month periodsamply)over

(. analysis morddy

b. Drinking One sample of each ofone to three of the nearest water supplies I-131 analysis on each composite when the dose within 10 mi!cs downstream that could be affected by its Composite

2. week period g( )leI-13I when over analysis is calctlated for the w__.. of the water is discharge. performed, monthly composite otherwise. greater than I mrem per yeart ). Composite for gross beta and gamma isotopic analyses (I)

One sample from a controllocatiert marshly. Composite for tritium analysis 81uarterly.

As there are no dricking =ater intakes wnhin 10 miles dovmstream of the discharge point,the drinking water pathway is currevaly not included as part of the Callaway Ptars Radiological Emironmental Monitoring Propam. Should future water intakes be constructed within 10 river miles downstream of the discharge point, the gwogram will be revised to include this pathway (Ref Il.6 6)

c. Sediment from One sample from downsta am area with existing or potential Semiannually Gamma isotopic analysis (5) semiannually shorehne recreational value 4 Ingestion a Mdk Samples from milking animals in three different meteorological Semimonthly when animals are on pa_ture, Gamrna isotopic (5) and I.131 analysis sectors within 5 km (3 mile) distance having the highest dose monthly at other times semunonthly when animals are on pasture:

potential. If there are none, then one sample from milking snonthly at other times animals in each of three different meteorological sectors between 5 to 8 km (3 to 5 mile) distance where doses are calculated to be grester than I rnrem per }v.

One sarnple frorn milking animals at a controllocation,15 to 30 km (10 to 20 mile) distance and in the least gwevalent wind diressiort Due to the lack of milking animals which satisfy these requirements, the milk pathway is currernly not included as part of the Callaway Plant Radiological Environmental Monitoring Program Should the Annual land Use Census identify the existence of milking animals in locations which satisfy these requirements, then the program will be revised to include this pathway.

m W W . _ _ _ _ _

W W _

W___ _

M M M M M M M M M W . _ _ . _ _ _ _ - _______________

W W W

(

gy APA-ZZ-01003 Rev.5 TABLE 9.11-A (Continued)

TABLE NOTATIONS NUhtIlER OF REPRESENTATIVE SAlfPLES SAhfPilNO AND ENI'OSURE PATilWAY AND SAMPl.E COI.LECTION TYPE AND FREQUENCY AXI) OR S utP!.E LOCATIONSill BEp_f U M OF ANALYSM 4 Ingenion (Cont'd) b rnh One sample ofeach connercially and recreationally important Sample in season, or semiannually if they are Genna inatopic analpis(3) on edible partiam species in deinity of plant dis &arge area. not seasonal One sample of same species M areas not influenced by plant disharge.

c Food Pneus One sample of each principal class of food products from any area At time ofharvest (9)(10) Gamma isotopic analpis (5) en edible portion that is imgated by water in which liquid plant wastes have been discharged As there are no areas irrigated by nter in which liquid plant wastes have been discharged within 50 miles damnstream of the discharge point, this serrple type is not currently included as part of the Callawmy Plant R adiological Emironmental Morutoring Program. Should future irrigation water intakes be constructed within to river miles downstream of the discharge poirs, the program will be revned to mdade thn sample type (Ref. I l.7.4 and II.7.5).

j samples of three different kinds of broad leaf vegetation if Monthly when available Gamma isotopic (5) and I-131 analpis available grown nearest each of two difTerent offsite locations of {

highest predicted annual average ground level D/Q ifmilk I sampling is not performed One sample of each of the similar broad leaf vegetation grown 15 Monthly when available Oansna isotopic (5) and I-13 I analysis to 30 km (10 to 20 mile) distant in the least prentent wind direction if milk sampling is not performed

APA-ZZ-o l003 Res 5 TABLE 9 Il A (Condnued)

TABLE NOTATIONS (1) Specific parameters of distance and direction sector from the centerline of one unit, and additional desenption where pertinent, shall be provided for each and every sample location in Table 9.11.A in a table and figure (s) g in the ODCM. Deviations are permitted from the required sampling schedule if specimens are unobtainable g, 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 penod. All deviauons from the sampling schedule shall be documented in the Annual Radiological Emironmental Operating Report pursuant to Technical Specification 6.9.1.6, (CTSN 2804)

It is recognized that, at times,it may not be possible or practicable to condnue 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 l 30 days in the Radiological Emironmental Monitoring Program. Submit in the next Annual Radiological l W

Environmental Operating Report documentation for a change 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. g (2) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate W 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 g are considered as two or rnore dosimeters. Film badges shall not be used as dosimeters for measuring direct g radiation. 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 l characteristics of the specific system used and should be selected to obtain optimum dose information with W minimal fading.

(3) The purpose of this sample is to obtain background information. Ifit is not pracucal to establish control locations in accordance with the distance and wind direedon criteria, other sites that provide valid background data may be substituted.

(4) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more after sampling to allow for radon and tharon daughter decay. If gross beta activity in air pamculate samples is W greater than 10 times the yearly mean of centrol 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 efDuents from the facility.

(6) The " upstream sample" shall be taken at a distance beyond significant influence of the discharge. The "dowTistream" sample shall be taken in an area beyond but near the mixing zone.

(7) In this program, composite sample aliquots shall be collected at time intervals that are very short (e.g., hourly ) g relative to the compositing period (e g , monthly) in order to assure obtaining a representative sample.

E (8) Groundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas i

where the hydraulic gradient or recharge properties are suitable for contamination.

f (9) The dose shall be calculated for the minimum organ and age group, using the methodology and parameters in the ODCM.

(10) If harvest occurs more than once a year, sampling shall be performed during each discrete har' est if harvest l E

occurs continuously, sampling shall be monthly. Attention shall be paid to including samples of tuberous and root food products.

! I

.> 5 g

%.w APA-ZZ-01003 Rev.5 TABLE 9.11-B REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES REPORTING LEVELS ANAINSIS WATER

• FISH MHX (pCy[ }e AIRDORNE OR OASES (pCi'm PART1Ct.p) TE FOOD (pCag, wet)b

{ )a (pCWg,PROy) wet H-3 20,000'

--cym m np ..y m m mry:m m m m y:pp~xven m~.gyw w wpwmm n i s. n n .,.:...

hin-54 ->% MKu :%CiH " e+%.42i:, %m:%2 d kC.M:wiX i:m WN.%:i d$5%bH$.k;hMt.:i,PWSM5inih.d%W Mk ' ' ' '.m gmypynm-e yn -I,0,00 30 0 gnm;emwy.n:ym.aew;;g:ymp3p wamwmp. ,m-ng.g.7 00 -

%.,gg~m.dk;n,%:ik.

. . . . .  ;.ba LM.Qw: .G+ .. u.h .a. ' ; .

> '..dk nMa,.sa,h%:ks~b% a %&Aw.:9..ymg.wo g~,e l'e.59 u e x y~en;

.~

400 10,000

..m... , ,

c...m.>-~,em% :dtM.,g.yg , . g.,.y.,g.m g.g.g,..y.77.g.

aEL .:h3W.:thdX Mgh'ioM.S dC M C ;:#:4>bh mpg: 5 Co-58 ,1,000 y y y .gceMskd y:hM!TEJi4%m*iiv!.mg. ny,WIWE' -

., . . p-mm&yw*!

2Mik x::

. , 00 memep. nrm ~..wmwww~v~.).,.30,0.nw>wve.ww:s:gg...,y, , ;. g:.. gg. h

,n ~

~.m.. .. p m. .gnn. ,+- . ,r.m e, g.g

gg'

jgg cm n

g , .

g.

g ,g.;-gg mmmerw yvm~g ggg g~ww .

Co.40 .,-y .

.. c 3.m00 10,000

.: ,.-, o.e 7 .

. ' . 2:: . , E.Q$.. J..3Si$$dskqi'h!.:.dhi:p.7, w,h: _, ..-3 * .m.m ~ . . .

' ' d b ijd -

. Zr.hh93 -

400 -

NNO hy ,I:.Ndhhs[Q'iii$dNhO)Ms.$$khb.. ..h. ... _ ._ -,h.s- ~hd Mkd de'.$N5bdb . ww:rwm dN.Ni FO5}Ndb.O:

v,, m , w3 ,. .g. g. .

.ye yy -

e

- . I-13.1. m .,ep g, - 0.9 3 100

~ > p.2y.yp.g.g7y.

.. 4:4.enM wu x.

30 yy,.ypyyby:urce:yype.mg.yupsGhrmm:,r.u_;em.

x ca:.w.sAvrs>.my.au e p

_ .o :.:sm :. ann.hh.s:ww y g .,, y madh. 4. .:m.s may -w

- Cs-13 4..yy,7~. 10

-~x ,.

,.,g.~n ,,,, , ,,y wnm 1

.eg,000 60 1

_. 3 y p ,,,,v - am.r. L . A h *:+ m gg<ewmmmehdhh _,__.

. Ja.aes . < u mb;is umw:2 . us

_gp._. _

. . , . nyy

m;?M %yEnd%;;;-;s g.gy

, . . - Cs-137.. . . ,s,_.g,,,.

. $0 0 2,0 70

. py.,.g.g.,; ,.y.- 2,000

+ . Mi:ts 97 2.g%:5:@5.

s.y ,,,m&  ::m RL, hen y ,.y? y y ,.g. g :p.g 00, g" it: et9f' :. .Wsi:s:a<&usC, .MMsWWidsb%;*GEF@i  %. . ., .y.7 2 s7 my.,.,

.m 4 .

Ba.b-I40 200. . . . NW. U 300 (a) ktuhiply the values in this table by I E-9 to convert to units orpCi/mt.

(b) h1uhip!y the values in this table by i E-9 to convert to units of pCirg.

. For drinking water samples This is 40 CFR Part 141 value For surface water sam lp es, a value of 30,000 pCi/t may be used.

Total activity, parent plus daughter activity.

-9 APA-ZZ-01003 Rev.5 TABLE 9.I1-C DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS t

LOWER LIMIT OF DETECTION (LLD) (I),(2), (3)

ANAIJSIS SURFACE DRINKING AIRDORNE ,

FISII MILK SEDIhtENT WATER WATER PARTICULATE (p 'ag. wet)b FOOD PRODL*b"TS pf ya (pCag. wet) (pCag. dry)D (pCy[ )* (pCi/l )n ORGASES(pCan3)

Gmss Deta 4

4. ., , 1

~ .m ,;,7 y.g.r.-.

+,7m,+myyy00.g a:$s.u.,

y n.w.,y m ,>e gi.~eynwy g y ; m g >p. .,

.eu.a..za Amb.x %,.+<waasv hg a.':+a. y y;ssh epp prv.w.g 7 II-3 _3000 sy,vmygy.,000...

> x a.+ 4 2 - <

ppgmygg 7 3.ywggy.,r-, .w. rpm.gp Mn-34 15

deh2N4%tsa w - w^i'a.< :mu:.m.:4:W huana d m ,A ,..qvyyyh cs h , " wM*;MM' g w gkcx4.g:enb^7.~

C:s . g;v * -efm m -

15 130 o m~mme.n . 7.e ~ pyg?Adfr2E" ym.

t Fe-59

. ..  :.MW:

, , . . . , . . MDi . men.pn..m yg y> p~ p pyq:G u ssa. N z.% p t.~p'm3tdiWM rew;A e m m hSA4 s

prpge.+...n

~ d$tN:NS A W g &--%

- ..g~s

'~: x smvr --r-

, .~ n. v ..,., nv..30.~ ,.ym y^(,'; 3,.0,, 260

- J ;is?. . . . . . . ,,gy,yy;42

' 2.:idiz;c.:.%fias .

.,wg., ,,p r.,<g', ;eU 7gg;;. mm.y,pwpemm; . , , . . . , . , ,

> y,M;l ' ^ y!. MGM9uhEu%?sc;Qx;#4E Atr4swduk, aid" 'pym&8Mzu:-R KM g.mm.m:.

Co-3 R.60 >:E '

- y15 ;py y.,pg. ,15.,

.---.v.v 130

, a;.p;yy.y,.mp,.g.p. .,y n -  %.a L x ,>..-4

? MDs.vDP 1+Lw pyge y _

.;. , g . _.,.. , pgy.p.w.:,.

Zr-Nb-93 15. . .

s Xk'.;4.~t -:' X. r;;;.*K:cL%Ei% > ' GR&MX6MX:;.k.+i.%:2.< s..Mkhn '

1. >% . ++*n4 e i~ >

15

.. , mywwsv-i j

. a .s. .~w ~~ mypy~ggwm:-wrw~~;qc;7.x.

.2 sca:Wh u .m. . ,

ex -- u4 t ,

. ;p x:x.<4ch,.. .: .

p +;sm eu w&pwzu%.,3:m :b e.,

a q%s., ~ yx.

gyym.yp?:~ mrv:-

ne<:c+

. l I31 - 1000 1 0.07

..m. ^

u.~ r r~;p gr7gy - :m - 1 60

' m'.

I

. .~ .cymp&pw --6>.Kl' } .i ;;u . ~ ~.x m~>mymm'

'Dz  %+ 4:m~.msm~%E

%EktCl 362nev~ps:msw.>>7 .

..x.- g.;.x-wrgms~vmw' "> :.m:mw 9::

1 Cs134 15 IS + < cxD %:.d:% mum A . k- MMM: m--

00 130 ymme:7;- - qfg ppp,y .3. . s w-  :;

ms m p;;;:g . .. ,g 15 60 150 m ecup.

, n. ~

.. . w.n u. g .:. . m y . .or m . w .:u.#:g .. e yg,9. 3.m g.a.rh.:..,,- .u.7.-u y-~.omy+>pesm~ w :-

Cs-137 18 A x . . , c+x x < m ~;myy.-, . 18,.w . : + <my 130 18 30 180 db:pgz:.0.,.06 :  ::.m ~n~.n.,.,

?;.~.my.py emu .y#+wswnwrymm~; .

4.4 Gb.  %:o acAk9 pwm~m>< ' ' eam'iWb m m .M gusp p ygw~~r

~- . . -  ;

Ba-tr140 I3 I3 x +J.5 &+ >

I3 L

(a) Muhip!y the values in this table by I E-9 to convert to units orpCi/ml.  !

(b) Multirly the values in this table by I E-9 to convert to units of pCirg. I t

t

.. Total activity. parent plus daughter activity.

l i

?

i

APA ZZ-01003 Rev. 5 -

TABLE 9. Il-C (Continued)

TABLE NOTATIONS l

l (1) This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, tc3 ether with those of the listed nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report.

(2) Required detection capabilities for thermoluminescent dosimeters used for envirotunental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13, Resision 1, July 1977.

(3) The LLD is described in Attachment 1. .

)

i b

78

APA ZZ-01003 Rev. 5 I,

9.12 RADIOLOGICAL ENVIRONMENTAL MONITORING L AND USE CENSUS I l

9.12.1 Controls (CTSN 41835) i 9.12.1.1 A Land Use Census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence and the nearest garden

• of greater than 50m 2 (500 ft2) producing broad leaf vegetation. g APPLICABILITY: At all times. W ACTION:

a. With a Land Use Census identifying a locadon(s) that yields a calculated dose or dose commitment greater than the values currently being calculated in REC 9.8.2. I, identify the l new location (s) in the next Annual Radioactive Efiluent Release Report, pursuant to Technical Specification 6.9,1.7.

b. With a Land Use Census identifying a locadon(s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with REC 9.11.1.1, add the new location (s) within 30 days to the Radiological Environmental Monitoring Program except for vegetation samples which shall be added to the program before the next growing season.

The sampling location (s), excluding the control station location, having the lowest l 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. In the next Annual Radiological Environmental Operating Report include the revised figure (s) and tables reflecting the new sample location (s) with information supporting the change in sample location.

') c. The provisions of Technical Specifications 3.0.3 and 3.0,4 are not applicable.

9.12.2 Surveillance Recuirements 3 9.12.2.1 The Land Use Census shall be conducted during the growing season at least once per 12 months using that information which will provide the best results, such as, but not limited to, door to-door survey, aerial survey, or by consulting local agriculture authorities and/or residents. The results of l the Land Use Census shall be included in the Annual Radiological Environmental Operating Report l pursuant to Technical Specification 6.9.1.6.

1 E I

I I

I

• Broad leaf vegetation sampling of at least three ditTerent kmds of vegetation may be performed at the SffE BOUNDARY m each '

to two difTerent direction sectors with the highest predicted D/Q's in lieu of the garden census Specifications for broad leaf vegetation sampimg m Table 9.11 A. Part 4.c shall be followed, includmg analysis of control samples 79- l

APA-ZZ-01003 Rev. 5 9.13 RADIOLOGICAL ENVIRONMElfrAL MONITOPJNG INTERLABORATORY COMPARISON PROGRAM 9.13.1 Controls (CTSN 41835) 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 USNRC.

APPLICABILITY: Atalltimes.

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 Technical Specification 6.9.1.6.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

9.13.2 Surveillance Reauirements l 9.13.2.1 The Interlaboratory Comparison Program shall be described in the plant procedures. A sununary of l the results obtained as part of the above required Interlaboratory Comparison Program shall be l included in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 6.9.1.6.

l l

\

i i 1 .'

1 I.

l t i

I APA Z2-01003 I1 l Rev. 3

10. ADMINISTRATIVE CONTROLS Il; l

10.1 MAJOR CHANGES TO LIOUID AND GASEOUS RADWASTE TREATMEKr SYSTEMS l 10.1.1 Licensee-initiated major changes to the Radwaste Treatment Systems (liquid and gaseous);

l a. Shall be reported to the Commission in the Annual Radioactive Emuent Release Report for j 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 g ,

made in accordance with 10 CFR 50.59; gl

2) Sumcient detailed information to totally support the reason for the change without i benefit of additional or supplemental information; (

3) A detailed description of the equipment, components and process involved and the g interfaces with other plant systems;

4) An evaluation of the change, which shows the predicted releases of radioactive l M

materials in liquid and gaseous emuents 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 to the general l

W l

population 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 W gaseous emuents, to the actual releases for the period prior to when the changes J

are to be made; g

7) An estimate of the exposure to plant operating personnel as a result of the change; e!

and Documentation of the fact that the change was reviewed and found acceptable by 8) the ORC.

I,

b. Shall become effecdve upon review and approval by the ORC and in accordance with g Technical Specification 6.5.3.1. E 10.2 CHANGES TO THE OFFSITE DOSE CALCUL ATION MANUAL (ODCM)(CTSN 2815) ,

10.2.1 All changes to the ODCM shall be completed pursuant to Technical Specification 6.14 and approved as per APA-ZZ-00101, ' Preparation, Review, Approval and Control of Procedures".

10.2.1.1 All changes shall be approved by the ORC prior to implementation.

10.2.2 Cross Disciplinary Review for each rnision of the ODCM must include, as a minimum, the licalth 3 Physics, Quality Assurance, and Licensing and Fuels Radiological Engineering Departments.

10.2.3 A complete and legible copy of each revision of the ODCM that became effective during the last annual period shall be submitted as a part of, or concurrent with that years Annual Radioactive l

E l

Emuent Release Report pursuant to Technical Specification 6.14.

I l I i

I

.sl.

3

APA-ZZ-01003 Rev. 5

)

11. REFERENCE _S 11.1 Title 10, ' Energy", Chapter 1, Code of Federal Regulations, Part 20; U.S. Government Printing Office, Washington, D.C. 20402.

11.1.1 Statements of Consideration, Federal Register, Vol. 56, No. 98, Tuesday, May 21,1991 Subpart D, page 23374.

11.2 Title 10. " Energy", Chapter 1 Code of Federal Regulations, Part 50, Appendix 1; U.S. Government Printing OfDee, Washington, D.C. 20402.

11.2.1 10 CFR 50.36 a (b) 11.3 Title 40, " Protection of Environment", Chapter 1, Code of F-deral Regulations, Part 190; U.S.

Government Print 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

.11.4.2 Section 6.8.4f 11.5 COMMUNICATIONS l

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 Febmary 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 Electric Co.), dated February 9,1984.

l 11.5.4 Reference Deleted 11.5.5 Private Communication, H. C. Lindeman & B.F. Holderness, August 6,1986 I

11.5.6 Calculation ZZ-67, " Annual Average Atmospheric Dispersion Parameters". April 1989.

11.6 Union Electric Company Callaway Plant, Unit 1, Final Safety Analysis Report 11.6.1 Section i1.5.2.2.3.1 ,

11.6.2 Section 11.5.2.2.3.4 l

11.6.3 Section 11.5.2.1.2 l 11.6.4 Section 11.5.2.2.3.2 11.6.5 Section 11.5.2.2.3.3 11.66 Section 11.2.3.3.4 1

11.6.7 Section i1.2.3.4.3

, 11.6.8 Section 11.5.2.3.3.1 l

l 11.6.9 Section 11.5.2.3.3.2 l 11.6.10 Section 11.5.2.3.2.3 11.6.I1 Section 11.5.2.3.2.2 11.6.12 Section 2.3.5 3

11.6.13 Section 2.3.5.2.1.2 i

l 11.6.14 Section 9.2.6 i 11.6.15 Section 9.2.7.2.1

-8 2-

APA ZZ-01003 Rev. 5 l

11.6.16 Section 6.3.2.2 11.6.17 Table 11.1-6 11.6.18 Deleted 11.6.19 Deleted l 11.6.20 Deleted 11.6.21 Deleted 11.6.22 Table 2.3-68 g

11.7 Union Electric Company Callaway Plant Environmental Report, Operating License Stage. W 11.7.1 Table 2.1-19 i 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 U.S. Nuclear Regulatory Ccmmission, Preparation of Radiological Emuent Technical Specification g for Nuclear Power Plants", USNRC NUREG-0133, Washington, D. C. 20555, October 1978.

5 11.8.1 Pages AA-1 through AA-3 l 11.8.2 Section 5.3.1.3 l g 11.8.3 Section 4.3

)

l _. e 11.8.4 Section 5.3.1.5 l

i 11.8.5 Section 5.1.1 11.8.6 Section 5.1.2 l 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 11.9 U.S. Nuclear Regulatory Commission, "XOQDOQ, Program For the Meteorological Evaluation of Routine Emuent Releases at Nuclear Power Stations". USNRC NUREG4324, Washington, D. C. 20555.

I1.9.1 Pages 19-20 Subt.,utine PURGE 11.10 Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous 3 Emuents in Routine Releases from Light-Water-Cooled Reactors", Revision 1. U. S. Nuclear g Regulatory Commission, Washington, D. D. 20555, July,1977.

I1.10.1 Section c.l.b i1.10.2 Figures 7 through 10 m i1.10.3 Section c.4

', 11.11 Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Emuents for the Purposes of Evaluating Compliance with 10 CFR Part 50, Appendis 1". Revision 1.

U. S. Nu:! car Regulatory Commission Washington, D. C. 20555 October 1977.

I1.11.1 Appendix C Section 3.a t

APA-ZZ-01003 Rev. 5 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.

I1.12.1 Section I, Page 2 11.12.2 Section IV, Page 8 ,

11.12.3 Section IV, Page 9 11.12.4 Section III, Page 6 11.12.5 Section III, Page G 11.13 Management Agreement for the Public Use of Lands, Union Electric Company and the State of Missouri Department of Conservation, December 21,1982.

I1.13.1 Exhibit A l1.14 MISCELLANEOUS REFERENCES

{

11.14.1 Drawing Number M-109 0007-06, Revision 5 I1.14.2 Callaway Plant Annual Emironmental Operating Report (updated annually) 11.14.3 UE Safety Analysis Calculation 87 001-00 11.14.4 Calculation ZZ-48, " Calculation ofinhalation and Ingestion Dose Conunitment 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 88 10, " Methodology for Calculating the Response of Gross Nal(TI) Monitors to Liquid Efiluent Streams", June,1988 11.14.8 Calculation ZZ-57, " Dose Factors for Eu-154", January,1989 l 11.14.9 Calculation ZZ-78, Rev. 2, "ODCM Gascous Pathway Dose Factors for Adult Age Group",

July,1992.

I1.14.10 HPCI 88-08, " Performance Testing of the Erwironment TLD System at Callaway Plant" August, 1989.

I1.14.11 Calculation ZZ 250, Rev. O, "ODCM Gaseous Pathway Dose Factors for Child Age Group and Ground Plane Dose Factors", September,1992.

11.14.12 UOTH 83 58," Documentation of ODCM Dose Factors and Parameters". February.1983.

I1.15 U. S. Nuclear Regulatory Conunission, "XOQDOQ: Computer Program for the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations", USNRC NUREG/CR-2919 September 1982, Washington, D. C. 20555 11.15.1 Section 4, " Subroutine PURGE", pages 27 and 28 4 11.16 Regulatory Guide 4.13. " Performance, Testing, and procedural specifications for f Thermoluminiscence Dosimetry: Environmental Applications *(Revision 1) July 1977 USNRC.

Washington, D. C. 20555

/ I1.17 TID-7004, " Reactor Shielding Design Manual", Rockwell, Theodore, ed, March 1956

APA ZZ-01003 I

Rev. 5 11.18 BNWL 236, "lSOSHLD - 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, "lSOSHLD- 11: Code Revision to include calculation of Dose Rate from Shielded Bremstrahlung Sources", Simmons, G. L., et al; March 1967 11.20 BNWL-236, Supplement 2, "A Revised Photon Probability Library for use with ISOSHLD- Ill",

i Mansius, C. A.; April 1969 I1.21 ANSI N13.10 1974 , " Specification & Performance of On-Site Instrumentation for Continuously 3 Monitoring Radioactivity in Efiluents"; September,1974 g 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 SpecificatJons g and the Relocation of Procedural Details of Current RETS to the Offsite Dose Calculation Manual or g Process Control Program", January 1989 11.23 NRC Answers to 10 CFR 20 Implementation Questions 11.23.1 Letter, F. J. Congel to J. F. Schmidt, dated December 9,1991.

E W

l1.23.2 Internal USNRC memo, F. J. Congel to V. L. Miller, et al, dated April 17,1992.

I1.23.3 Letter, F. J. Congel to J. F. Schmidt, dated April 23,1992.

I1.23.4 Letter, F. J. Congel to J. F. Schmidt, dated September 14,1992.

I1.23.5 Letter, F. J. Congel to J. F. Schmidt, dated June 8,1993.

I1.24 USNRC Inspection Report 50-483/92002(DRSS) Section 5, page 5.

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APA ZZ-01003 Ret. 5 l

LOWER LIMIT OF DETECTION (LLD)

A detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual,

('-

HASL-300 (revised annually), Curie, L. A. " Limits for Qualitative Detection and Qualitative Determination -

Application to Radiochemistry", Anal. Chem. 40. 586-93 (1986), and Hartwell, J. K., " Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

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:

b

(- LLD =

E x V x 2.22E6 x Y x 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 sampic ;ize (units of mass or volume),

2.22E6 = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the panicular radionuclide (sec-I), and at =

the elapsed time between the midpoint of the sample collection period, and the time of counting (sec), for effluent samples, or At =

the elapsed time between the end of the sample collection period, and the time of counting (sec), for emironmental samples.

Typical values of E, V, Y, and at should be used in the calculation.

It should be recognized that the LLD is defined as a apriori (before the fact) limit representing the capability of a rneasurement system and not as an aposteriori (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLD's will be achieved under routine conditions.

The definition of At applies only to the calculation of the LLD. A more rigorous treatment of the buildup and decay during the sample collection and/or counting period (s) may be applied to actual sample analysis if desired.

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L Page1of1 ATTACHMENT 1

APA-ZZ-01003 Rev. 5 i

B ASES FOR RADIOLOGICAL EFr1UENT CONTROLS I

The BASES presented below surr.marize the reasons for the specified Radiological Efiluent Control, but in accordance with 10 CFR 50.36 are not part of these controls.

REC 9.1 RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUNEPTTATION Refer to FSAR CN #94-51 REC 9.2 RADIOACTIVE G ASEOUS EFFLUEPTT MONITORING INSTRUMENTATION Refer to FSAR CN #94-51 REC 9.3 L10UID EFFLUENTS CONCENTRATION This section is provided to ensure that the concentration of radioactive materials released in liquid waste ef!1uents to UNRESTRICTED AREAS will be less than the concentration levels specified in Appendix B Table II, Column 2 to 10 CFR 20.120.601. 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 E objectives of Appendix I,10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the E limits of 10 CFR Part 20.1301 to the population. The concentstion limit for dissolsed 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.

The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (L1 D's).

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APA-22-01003

[ Rev. 5 L-s B ASES FOR RADIOLOGICAL EFFLUENT CONTROLS REC 9.4 DOSE FROM LIOUID EFFLUENTS This section is provided to implement the requirements of Sections ll. A and IV. A of Appendix 1,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix 1. The ACTION statements provide the required

{ operatmg flexibility and at the same time implement the guides set forth in Section IV.A of Appendix 1 to assure that the releases of radioactive material in liquid effluents to UNRESTRICfED AREAS will be kept "as low as is reasonably achievable",

[ e 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 radionuclide concentrat i ons 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 1 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 ODC.M for calculating the doses due to the actual release rates of radioactive materials in liquid emuents are consistent with the methodology provided in Regulatory Guide 1.109,

[. " Calculations of Annual Doses to Man from Routine Releases of Reactor Emuents with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.113.

" Estimating Aquatic and Dispersion of Emuents from accidental and Routine Reactor s Releases for the Purpose ofimplementing Appendix I", April 1977.

REC 9.5 LIOUID RADWASTE TREATMENT SYSTEM 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 emironment. The requirement that the appropriate portions of this system be used when specified prmides assurance that the releases of radioactive materials in liquid emuents will be kept "as low as is reasonably 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 ! 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 ll.A of Appendix 1,10 CFR Part 50, for liquid emuents.

i e i Page 2 of 7 ATTACHMENT 2

gi APA ZZ-01003 Wl Rev. 5 I B ASES FOR RADIOLOGICAL EFFLUENT CONTROLS REC 9.6 GASEOUS EFFLUENTS DOSE RATE I,

This section is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous emuents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS. The dose rate limits are the doses associated with the concentrations of 10 CFR Part 20.1-20.601 Appendix B, Table !!,

Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous emuents 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 dose limits specified in 10 CFR Part 2010 CFR 20.1301. For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF TIIE 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 the SITE BOUNDARY to less than or equal to 500 mrem / year to g the whole body or to less than or equal to 3000 mrems/ year to the skin. These release rate g 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.

The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLD's).

3 The requirement for additional sampling of the Unit Vent following a reactor power transient is provided to ensure that the licensee is aware of and properly accounts for any increases in the release of gaseous efIluents due to spiking which may occur as a result of the power transient. Monitoring the Unit Vent for increased noble gas activity is appropriate because it is the release point for any increased activity which may result from the power transient.

Since the escape rate coefficients for the noble gas nuclides is equal to or greater than the escape rate coefficient for iodine and the particulate nuclides' 5 , it is reasonable to assume that the RCS spiking behavior of the noble gas nuclides is similar to that of the particulate and iodine nuclides. Considering the effects ofiodine and particulate partitioning, p!reout on plant and ventilation system surfaces, and the 99% efliciency of the Unit Vent HEPA filters and charcoal absorbers, it is reasonable to assume that the relative concentrations of the noble gas nuclides will be much greater than those of the iodine and particulate nuclides. '

Therefore, an increase in the iodine and particulate RCS activity is not an appropriate indicator of an increase in the Unit Vent activity, and it is appropriate to monitor the Unit Vent effluent activity as opposed to the RCS activity as an indicator of the need to perform post-transient sampling.. In addition, it is appropriate to monitor the noble gas actisity due to its relatively greater concentration in the Unit Vent.

I

• Cohen, Paul, Water Coolant Technotocv of Power Reactors Table 5.19, page 198. American Nuclear I l Society. 1980. gl 5

NUREG-0772," Technical Bases for Estimating Fission Product Behavior During LWR Accidents", gl Silberberg, M., editor, USNRC, Figure 4.3, page 4.22. June,1981.

Page 3 of 7 A'ITACHMENT 2

. APA.ZZ-01003 Rev. 5

)

BASES FOR RADIOLOGICAL EFFLUENT CONTROLS REC 9.7 DOSE - NOBLE GASES

. This section is provided to implement the requirements of Sections II.B. Ill. 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 1. 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 the requirements in Section Ill.A of Appendix !

- that conformance with the guidec of Appendix ! be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways in unlikely to be substantially underestimated. The dose calculation l 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 on Reactor Efiluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1", 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 BOUNDARY are based upon l the historical average atmospheric conditions.

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APA ZZ-01003 Rev. 5 BASES FOR RADIOLOGICAL EFFLUENT CONTROLS j REC 9.8 DOSE -IODINE-131. & 133. TRITIUM. AND RADIOACTIVE MATERI AL IN PARTICULATE FORM 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 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix !

to assure that the release of radioactive material in gaseous emuents to UNRESTRICTED AREAS will be kept "as low as reasonably achievable". The ODCM calculational methods spectfied in the Surveillance Requirements implement the requirements in Section Ill. A of g Appendix ! that conformance with the guides of Appendix I be shown by calculational g 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 ODCM calculational methodology and parameters for calculating the doses due to the actual l release rates of the subject materials are consistent with the methodology provided in W Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor EfDuents for the Purpose of Evaluating Compliance with 10 CFR Part 50 Appendix g I", Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating g Atmospheric Transport and Dispersion of Gaseous Emuent sin 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 radionuclidei in particulate form with half-lives l

u

'j greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the g 3

development of these calculations were: (1) individual inhalation of airborne radionuclides, E (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition of radionuclides onto grassy areas where milk animals and meat.

producing animals graze with consumption of the milk and meat by man, and (4) deposition l on the ground with subsequent exposure of man. =

REC 9.9 G ASEOUS RADWASTE TREATMENT SYSTEM The OPERABILITY of the WASTE GAS HOLDUP SYSTEM and the VENTILATION EXHAUST TREAThENT SYSTEM ensures that the system will be available for use whenever gaseous emuents require treatment prior to release to the emironment. The requirement that the appropriate portions of these systems be used, when specified, pro $ ides reasonable assurance that the releases of radioactive naterials in gaseous efDuents will be i 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, arid the i design objectives given in Section II.D of Appendix ! to 10 CFR Part 50. The speciEcd limits governing the use of appropriate portions of the systems were specified as a suitable fraction g of the dose design objectives set forth in Sections ll.B and II.C of Appendix 1,10 CFR Part g 50, for gaseous emuents.

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l' Pagc 5,7 mACms-o g

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1 l APA-Z7.-01003 1 l Rev. 5 l B ASES FOR RADIOLOGICAL EFFLUENT CONTROLS l

REC 9.10 TOTAL DOSE This REC is provided to meet the dose limitations of 40 CFR Pan 190 that have been incorporated into 10 CFR Part 20.1301. The control requires the preparation and submittal of a Special Report whenever the calculated doses due to releases of radioectivity and the radiation from uranium fuel cycle sources exceed 25 mrems 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 fHE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix 1, and if direct radiation doses from the reactor units and from outside storage tanks are kept small. The Special Repon will describe a course of action that should result in the limitation of the annual dose to a l MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. j 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 l

exception that dose contributions 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 .j 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 Part 190.11 and 10 ' l CFR 20.2203, is considered to be a timely request and fulfills the requirements of 40 CFR )

Pan 190 until NRC staff action is completed. The variance only relates to 40 CFR Part 190. I l

_.. and does not apply in any way to the other requirements for dose limitation of 10 CFR Pan 20, ns addressed in REC 9.3.1.1 and 9.6.1.1. An individual is not considered a MEMBER OF l THE PUBLIC during any period in which he/she is engaged in carrying out any operation j that is part of the nuclear fuel cycle.

REC.9.11 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM The Radiological Emironmental Monitoring Program required by this REC provides j representative measurements of radiation and of radioactive materials in those exposure l pathways and for those radionuclides that lead to the highest potential radiation exposures of I MEMBERS OF THE PUBLIC resulting from the station operation. This monitoring program l implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the  ;

Radiological EfDuent Monitoring Program by verifying that the measurabic concentrations of j radioactive materials and levels of radiation are not higher than expected on the basis of the efDuent measurements and the modeling of the emironmental exposure pathways. Guidance for this monitoring prognm is provided by the Radiological Assessment Branch Technical Position on Emironmental Monitoring. Revision 1. November 1979. The initially specified monitoring program will be c!Icctive 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 emironmental sample analyses are tabulated in terms of the lower limits of detection (LLD's). The LLD's required by Tabic 9.ll-C are considered optimum for routine emironmental measurements in industrial laboratories.

I Page 6 of 7 ATTACHMENT 2

l APA-ZZ-01003 Rev. 5 B ASES FOR RADIOLOGICAL EFFLUENT CONTROLS I

REC 9.12 RADIOLOGICAL ENVIRONMENTAL MONITORING LAND USE CENSUS This REC 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 l

W Monitoring Program given in the ODCM are made if required by the results of this census.

Information that will provide the best results, such 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 m2provides 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) ofleafy 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 3 (i.e., similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m2, g REC 9.13 RADIOLOGICAL ENVIRONMENTAL MONITORING INTERLABORATORY COMPARISON PROGRAM 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 emironmental sample matrices are performed as part -

of the quality assurance program for emironmental monitoring in order to demonstrate that the results are valid for the purpose of Section IV.B.2 of Appendix ! to 10 CFR Part 50.

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APA-ZZ-01003 Rev. 5 l

SUMMARY

REVIEW OF RADIOLOGICAL EFFLUENT TECII SPECS POTENTIALLY AFFECTED BY THE IMPLEMENTATION OF TIIE REVISED 10CFR20 The following is a summary review of the current Tech Specs that are potentially affected by the l implementation of the revised 10CFR20. In general, the potential impact is due to changes in the Efiluent Concentration Values (ECV's) in 10CFR20, Appendix B, Table 2, Coh;mns 1 and 2 (formerly MPC's), and  ;

10CFR20.1601. l This summary is not intended to review those changes that may be necessary as a result of the eventual issuance  ;

of the Generic Letter, I The NRR staff has stated that the current level of efBuent controls is sufIlcient to protect the health and safety ,

I of the public, and further restrictions resulting from the revision to Appendix B, Table 2, were unintentional.

They are currently preparing a Generic Letter that will provide guidance for submitting Tech Spec changes that will return to the current level of control. This is currently anticipated during late 1993. Those who implement the revised rule prior to January 1,1994, will have to do so under the requirements of 10CFR20.1008, widch basically requires that the more restrictive requirement (Tech Specs or 10CFR20) be implemented.

DEFINITIONS OF RESTRICTED AREA & MEMBER OF TIIE PUBLIC, AND TECII SPEC 5.1.2, SITE BOUNDARY FOR GASEOUS EFFLUENTS The definition of Restricted Area has not changed significantly from that in the former rule. The definition of )

the Member of the Public in the revised rule is significantly different from that in the Callaway Plant Technical l

...- Specifications (TS 1,17). There is no corresponding definition of Controlled Area in the former rule.

The Callaway Plant was licensed to operate with a Restricted Area as defined in the FSAR and shown on the figures in TS 5.1.4 and in the ODCM. Since the requirements have not been revised, there is no compelling ]

reason to change the Callaway Plant Restricted Area from its current boundaries. I 1

In addition, the NRC's backfit analysis ,t performed pursuant to 10 CFR 50.109, concludes that the resisions to .

10CFR20 apply primarily to operational procedures and should cause no modifications in facility design. Since the plant siting and the location and size of the Restricted Area are considered to be a part of the facility design, it is clearly not the intent of the NRC that revisions to 10CFR20 would require changes to the Restricted Arca i for currently licensed facilities.

1 2

There is also no requirement for the existence of a Controlled Area as defined in the revised rule , therefore it is not necesary that one be created at Callaway.

l 8 " Final Backfit Analysis for the Revision of 10CFR20, " Standards for Protection Against Radiation *",

USNRC, office of Nuc! car Regulatory Research, Division of Regulatory Applications. August.1990. (Available USNRC Public Documents Review.)

2 Refer to Question 26(a) (4th set).

Page 1 of 12 APPENDIX A 1

l APA ZZ-01003 Rev. 5 1 I

l The definition of the Member of the Public is significantly different in the revised rule reladve to that provided I in TS 1.17 and in 40 CFR 190. The revised rule defines the Member of the Public as anyone who is not in the Restricted Area. The Tech Specs and 40 CFR 190 generally define the Member of the Public as anyone who is not occupationally associated with plant operations, and also recognizes that the Member of the Public may, at i times, be within the Restricted Area. The major difference is that pursuant to the revised rule, the Member of I the Public receives dose against the occupational dose limits of 10 CFR 20.1201 once inside the Restricted Area, but the Tech Spec definition we' u ld limit the dose within the Restricted Area to the limits of 10 CFR 20.1301, Since the limit provided in 20.1301 is much lower than that of 20.1201, the continued use of the more recrictive 40 CFR 190 and Tech Spec 1.17 definitions for the Member of the Public is appropriate and is reouired pursuant to 10 CFR 20.1008(c).

A more thorough and detailed analysis of the definitions of the Member of the Public found in 10 CFR 20, g 40 CFR 190, and Tech Spec 1.17, focusing on the applicability of Occupational Vs. Non-occupational dose g limits, indicates a confusing and inconsistent anay of definitions and dose limit applicability. For conservadsm and simplicity, Union Electric has defined occupational dose as dose received while working with or around radioactive materials. This definition is more restrictive than the definition in 10 CFR 20 in that the more restrictive dose limits of 10 CFR 20.1301 are applied to Members of the Public within the Restricted Area, instead of the less restrictive limits of 10 CFR 20.1201. It is more restrictive than the Tech Spec definition in that delivery persons, service technicians, and others who may enter the site to perform non-radiological work g activities are also limited to the more restrictive dose limits of 10 CFR 20.1301. 3 There are no changes recommended for those definidons and maps relative to the Restricted Area, Site Boundary, and dose to the Member of the Public.

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TECII SPEC 6.8.4,F.2, LIQUID EFFLUENT RELEASE RATE LIMITS (REC 9.3)

On December 1,1992, Union Electric Co. provided notification3 ofintent to implement the revised 10 CFR 20, Parts 20.1001 20.2401 and associated appendices, pursuant to 10 CFR 20.1008(a). The revised rule was fully implemented on January 1,1993. The following provides clarification with respect to compliance to 10 CFR 20.1001- 20.2401 and Callaway Plant Technical Specifications 6.8.4.f(2) and 6.8.4.f(7).

Union Electric implemented the use of the revised Appendix B, Table 2 values concurrent with the implementation of the revised rule. Technical Specification 6.8.4 f(2) requires that the concentration of radioactive material in liquid discharges not exceed the values of 10 CFR 20, Appendix B. Table 11, Column 2.

The NRC had indicated via the revision to 10 CFR 50.72 that the concentration values have nominally decreased by a factor of 10, and the NRC staff had stated on numerous occasions that they considered the values in the revised rule to be more restrictive than the those in the old rule. This was frequently referred to as an

" implicit" change to the Technical Specifications.

10 CFR 20.1008 (a) requires that if the revised rule is implemented prior to January 1,1994, then "the licensee shall implement all provisions of these sections,. . and shall provide written notification.. . that the licensee is g sdopting early implementation (of the revised rule) and associated appendices." 10 CFR 20.1008 (b) requires 3 that once implemented, "the applicable section of (the revised rule) shall be used in lieu of any section (of the old rule) that is cited in license conditions or technical specifications." It further states, "if the requirements of (the revised rule) are more restrictive than the existing license condition, then the licensee shall comply with (the revised rule).

3 ULNRC 92-2729, D. F. Schnell to A. Bert Davis, dated December 1,1992.

Page 2 of 12 APPENDIX A I

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APA ZZ 01003 Rev, 5 i

l Addiuonally, the NRC had clarified the applicability of the revised Appendix B values to the Technical .J Specification instantaneous release rate limits via their formal response to three separate licensee quesdons. l Question # 18 states4 that the Tech Spec instantaneous release rate limit is based on the old Part 20 concentrations, and asks if changes are required in the Tech Specs and ODCM as a result of the revised rule.

De NRC replies

• the instantaneous release rates for liquid emuents, to the extent that they directly reference Appendix B concentration values, will need to be changed.- The corresponding bases and certain alarm set-points will have to be changed by license amendment,"

Question # 23 asks5 if computer data bases that use the old Appendix B values must be revised to the use the new values. The NRC simply answers, "Yes",

• i o

Question # 22 states 6that many alarm set-points are based on 10 CFR 20 Appendix B concentrations, and asks if they will have to be changed. The NRC answers thet the alarm set-points ofliquid emuent monitors are likely to require change, since they are based on 10 CFR 20 Appendix B concentrations, as required by Tech Specs. Because Appendix B concentration values differ for many radio nuclides between the old and new l

versions of Part 20, these set points may have to be changed. His is analogous to a restriction in flow rate, and j i the NRC cites the reduction in Appendix B concentrations as the root cause of the change. l l

Based on the preceding information, Union Electric implemented the use of the revised Appendix B values concurrent with the implementation of the resised rule on January 1,1993, Because there were no values in the revised Appendix B for dissolved and entrained noble gases in liquid emuents, the old value of 2E-4 uCi/mi was used pending regulatory guidance.

The Callaway Plant Technical Specifications contain, in Section 6.8.4.f several specifications which provide

} appropriate limits on the maximum quarterly and annual whole body and organ dose to the Member of the '

l Public from the discharge ofliquid and gaseous radioactive emuents. Compliance with these specifications. I demonstrates compliance with the limits of 10 CFR 50, Appendix 1, and 40 CFR 190 and, as stated in the supplemental information to the revised rule 7, demonstrates compliance with the 100 mrem /yr dose limit of 10 CFR '.1.1301.

j However, compliance with the dose rate limits of Specifications 6.8.4.f items (2) and (7) with respect to the implementation of the revised rule is less clear, as there is no longer a regulatory basis for these Specifications.

These Specifications formerly implemented the requirements of 10 CFR 20.106, which prosided annual average concentration limits on liquid and gaseous emuents, and specifically referenced the limits of Appendix B, Table II, Columns 1 and 2.

I i 4 Letter, F. J. Conjel (USNRC) to J. F. Schmitt (NUMARC), dated December 9,1991. page 16 of Enclosure 1.

5 ibid, page 14 of Enclosure 1.

' USNRC Memorandum, F. J. Conjel to V. L. Miller, et al, dated April 17,1992. page 13 of Enclosure 1.

7 Federal Register, Vol. 58, No. 98, Tuesday, May 21,1991. pages 23360-23474.

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Unlike the former rule, the values in the revised Appendix B, Table 2 Columns I and 2 do not of themselves 3 constitute a limit on the release rate of radioactive effluents, but rather, as discussed in 10 CFR 20.1302 E (b)(2)(i), merely provide one means of demonstrating compliance with the annual dose limit of 10 CFR 20.1301. Since there is no release rate limit provided in the revised rule, the subject Specifications are therefore license conditions. 10 CFR 20.1008 (c) requires that any existing license condition that is more l W

restrictive than the revised rule remain in force until there is a technical specification change. Additionally, since the values in the revised Appendix B. Table 2 are not limits as was the case with 20.106, there is no correspondmg provision in the new rule to 20.106. 10 CFR 20.1008(e) requires that if a license condition cites 3 a provision in the old rule for which there is no corresponding provision in the new rule, then the license l condition remains in force until there is a technical specification change.

The values of Appendix B, Table 2, Columns I and 2 of the revised rule did not change in a uniform fashion, i.e., certain nuclides numerically decreased in value whereas others numerically increased in value.

Furthermore, the values did not change by a consistent amount, varying by as much as a factor of 20 with respect to the corresponding nuclide in the former rule. This inconsistency is clearly evident for those nuclides which are commonly associated with nuclear power plant efIluents. In addition, the bases for the revised values is the dosimetry system ofICRP 26 8 and ICRP 30 .9 This is inconsistent with the bases for the dose limits of l 10 CFR 50, Appendix I and 40 CFR 190, and the dose calculational methodologies of Regulatory Guide 1.109, which are largely based on the dosimetry system ofICRP 2 10 Since the values of the revised Appendix B, Table 2, Columns 1 and 2 did not uniformly increase or decrease in value, it is not possible to determine whether Appendix B Table 11 of the former rule or Appendix B Table 2 of g the revised rule provides, in toto, the more conservative values for implementation of the subject license ~ 3I conditions. It is clear, however, that the bases for the revised Appendix B, Table 2 values are inconsistent with

) the bases of 10 CFR 50, Appendix ! and 40 CFR 190, and Regulatory Guide 1.109. Furthermore, the a

.v operational history of the Callaway Plant demonstrates that the use of the 10 CFR 20.1 20.601, Appendix B. g Table Il values is appropriate to maintain compliance with the requirements of 10 CFR 50, Appendix I and 40 CFR 190, which, in turn, demonstrates compliance with the 100 mrem /yr dose limit of 10 CFR 20.1301.

The concentration limits of the old Appendix B, Table 11 were based on a dose of 500 mrem /yr, which, when l expressed as a dose rate, is equal to .057 mrem /hr. Compliance with the requirements of Technical B Specifications 6.4.8.f(2) and (7) using 10CFR 20.106, Appendix B, Table 11 values is conservative with respect to the 2 mren'lbr limit of 10CFR20.1301(a)(2). Additionally, Technical Specifications 6.4.8.f(2) and (7) specifically require the use of Appendix B, Table 11 to 10CFR20.1- 20.601, since there is no corresponding provision in the revised rule. ,

I Thus,10 CFR 20.1008 (c) and (c) require the continued use of the values provided in Appendix B Table 11 to l' E

10 CFR 20.1 20.601 for the implementation of Technical Specifications 6.8.4.f, items (2) and (7).

Although the 2 mrem /hr limit of 10 CFR 20.1301(a)(2) was referenced in the preceding discussion, it is g important to note that the reguladon specifically states that this limit is applicable to external sources. Since, g for the Callaway Plant, the only dose pathway to man from the discharge ofliquid radioactive efIluent is through the consumption of fish, there are no external dose pathways, and therefore the requirements of  ;

10 CFR 20.1301(a)(2) are satisfied apriori, i

8 International Conunission on Radiation Protection, Publication 26. " Recommendations of the International l

Commission on Radiation Protection" , Annals of the ICRP, Volume 1, No. 2,1977.  !

' International Commission on Radiation Protection, Publication 30, " Limits for Intakes of Radionuclides by Workers", Annals of the ICRP, Volume 2, No. 3/4,1979.

10 International Commission on Radiation Protection, Publication 2, " Report of Committee 11 on Pemussible Dose for Internal Radiation",1960.

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Page 4 of 12 APPENDIX A

l APA ZZ-01003 Rev. 5 Union Electric re instituted the use of the values in Appendix B, Table !!, Columns I and 2, to 10 CFR 20.1 20.601 for Technical Specifications 6.8.4.f Items (2) and (7) pursuant to the requirements of -

10 CFR 20.1008(c) and (c), on May 4,1993.

This position was amrmed by the USNRC on June 30,1993 31 EFFLUEffT CONCENTRATION VALUE FOR Gross ALFHA IN LIQUID EFFLUENTS There are two values in the revised Appendix B for unknown mixtures in liquid emuents: 2E-9 and IE 6 uCi/ml. 'Ile less restrictive value is appropriate ifit is known that certain nuclides are "not present". The appropriate value for gross alpha in liquid emuents at the Callaway Plant from Appendix B, Table 2, Column 2 f is 1E-6 uCi/ml.

The value of IE 6 uci/ml in Appendix B Table 2, Column 2 only applies to an unknown mixture of nuclides where those listed opposite the value are known to be "not present". These nuclides are Fe-60, Sr-90, Cd 113m, Cd-ll3, in-115,1 129, Cs 134, Sm-147, Gd 148, Gd 152, Hg 194 (organic), Bi-210m, Ra-223, Ra-224, Ra-225, Ac 225, Th-228, Th-230, U-233, U-234, U-235, U-236, U-238, U-nat., Cm-242, Cf 248, Es 254, Fm-257, and Md-258. The other nuclides listed in the immediately preceding values for unknown mixtures in gaseous emuents do not apply, since they specincally apply to gaseous emuents as indicated by the designation of applicable lung clearance classifications for each of the nuclides listed. The NRC's response to Question # 71 reiterates that ingestion ALl's do not have lung c!carance classifications, which is also consistent with ICRP 30 and all other industry standards. Additionally, several of those listed in the list for liquid

{' emuents also appear in the list of nuclides given for airborne activity, which indicates that only those s

) specifically listed with the liquid emuent value apply, f Of those nuclides listed for unknown auxtures in liquid emuents, only Ra-224. Th-228, U 234 U-235, U-236, U 238, and Cm-242 are LWR produced alpha emitting nuclides. Sr 90, Cd l13m,1 129, and Cs 134 are also LWR produced, but are beta or beta / gamma emitters, and are not determined via a gross alpha analysis. The remainder of the nuclides in the list are not LWR produced.

{

The phrase 'not present" is not defined in the resised 10 CFR 20, however there is a large body ofinformation l

which can be applied to determine the meaning of"not present". The former rule,in footnote 5 to Appendix B,

[ stated that a nuclide may be considered to be "not present" ifit constitutes less than 10% of the total actisity, provided that the aggregate of all such "not present" nuclides does not exceed 25% of the total activity. The use of the " ten percent rule" is consistent with the basis of the revised rule, the NRC's response to questions regarding the meaning of "not present", and the current ICRP guidance as shown below; f

a. The revised rule is based on the dosimetry and methodology ofICRP 30 12, which in paragraph

[

3.1.3, describes the use of the current ten percent mle.

t b. The NRC's respnse to Question # 14613clearly indicates that the ten-percent rule is applicable to Appendix B.

Il Ixtter, Thomas E. Murley, Director, NRR, USNRC, to Thomas E. Tipton, NUMARC, dated June 30,1993.

12 ICRP Publication 30, " Limits for Intakes of Radionuclides by Workers", in Annals of the ICRP, Volume 2 Number 3/4,1979.

13 Ixtter, Frank J. Congel Director, DRPEP, USNRC, to John F. Schmidt, NUMARC, dated September 14, 1992 (commonly referred to as the 4th set of Q&A)

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c. The current ICRP recommendations on the release of radioactive materials to the environmenti d, and the updated recommendationsis to ICRP 30 continue to propagate the ten-percent rule, and  ;

apply it to offsite dose as well as dose to radiation workers.

It is therefore clear that the ten-percent rule continues to apply to the values in Appendix B of the revised rule.

Callaway Plant liquid efnuents have been analyzed for trar.suranic nuclides (TRU) on two separate occasions, during the second and third quarters of 1987, in each instance, TRU nuclides were not detectable, with an MDA of IE-8, uCi/ml, which is a factor of 10 below the gross alpha LLD of IE-7 uCi/ml.

The concentration of the TRU nuclides can be inferred through the use of a tracer nuclide, such as Cc 144.

Cc-144 is particularly well suited for this purpose in that it is a fission product, can be measured by gamma rar E spectroscopy, and is chemically similar to the TRU nuclides. Based on published ORIGEN code calculations is g of a representative LWR, and assuming a 90 day decay, the ratio of the nuclides ofinterest to Cc 144 is:

Ra 224/Cc-144 1.45E-9 "Ih-228/Cc-144 1.45C-9 U 234/Ce-144 1.14E-6 U-235/Cc-144 1.75E-8 ~l U-236/Cc-144 2.58E-7 j U 238/Cc 144 3.24E-7 l Cm 242/Cc-144 2.66E-2 Vollique, et al 17, found the Cm 242/Cc 144 ratio to be 6.5E-3, which is consistent with the above value.

,) Based on the above, it can be seen that Cm-242 is the only nuclide with a significant Cc-144 ratio.

- 1 Based on the data contained in the Semiannual EfDuent Release Reports for the period January,1989- W July,1992, Ce 144 accounted for less than 0.3% of the total fission and activation product activity in liquid efDuents, and less than SE-6% of the total activity discharged in liquid efDuents during the same period. 3 Therefore, the maximum activity that could have been discharged of each of the above listed nuclides is much g less than 10% Accordingly, these nuclides are "not present", i l

l TECII SPEC 6.8.4.F.4, DOSE FROM LIQUID EFFLUENTS (REC 9.4), & TECII SPEC 6.8.4.F.5, LIQUID RADWASTE TREATMENT SYSTEM (REC 9.5)

These specifications are derived from 10CFR50, Appendix 1, and are not affected by the resised rule. Doses are calculated in accordance with Regulatory Guide 1.109 which has not been revised. No changes are anticipated for these specificadons.

14 ICRP Publication 56, " Age-dependent doses to Members of the Public from the Intake of Radionuclides:

Part 1", Annals of the ICRP, Volume 20, Number 2,1989.

IS ICRP Publication 61, " Annual Limits on intake of Radionuclides by Workers Based on the 1990 Recommendations", Annals of the ICRP, volume 21 Number 4,1991.

\

16 Licht Water Reactor Nuclear Fuel Cycle. Wymer, Raymond G. and Vondra Benedict L., editors. Table 6, I pages 70- 71 and Table 7, page 72. CRC Press,1981.

17 Vollique, P. G., et al, " Solubility of Transuranic Nuclides in Acrosols in Two Ginna Steam Generator Work Emironments" Proceedings of the Twenty-First Midyear Topical Meeting of the Health Physics Society, Pages ll l 251-260. 1987 ,

Page 6 of 12 APPENDIX A

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TECH SPEC 3.11.1.4, CURIE CONTENT OF OUTDOOR LIQUID STORAGE TANKS The purpose of this speciScation is to limit the activity in the nearest receiving waters, excluding tritium and )

entrained noble gases, to the concentrations in 10CFR20, Appendix B Table 2, Column 2. l 4

l The effect of accidental contamination of the nearest ground water discharge locations due to accidental rupture )

of tanks containing radioactive liquids was perfonned as detailed in FSAR Section 2.4.13.3. It was assumed i that the liquid cuntents of a ruptured tank would immediately merge with the ground water 5 feet below plant ~

grade and travel directly from the tank to the nearest down-gradient well (Well 23). The results of the calculation show that, with the exception of H 3 and Sr-90, the radio nuclide concentrations found in ground water af ter a tank rupture will be below the original 10CFR20, Appendix B. Table II, Cohunn 2 values by the time the contaminated ground water reaches the nearest stream tributaries. The dilution capability of the streams is sufDefent to reduce the concentration of H 3 and Sr 90 below the original Appendix B values. All computed concentrations at Well 23 were below the Appendix B limits for unrestricted areas.

Tables I and II list the curie contents of the primary spent resin storage tank and refueling water storage tank used in the FSAR calculations. These values were adjusted to reflect a total tank curie content of 150 Curies, the limit identified in Tech Spec 3.11.1.4. (Even though the spent resin storage tank is not an outdoor tank, the data was used for this calculation since it is expected to have the highest curie contents for Sr-90, Cs-137 and Co-60 and the postulated accident assumes that all liquid released immediately merges with the ground water.) l

} The resultant peak concentrations at the discharge point at Logan Creek were calculated using the normalized

,/ values then compared to the revised Appendix D efnuent concentration values (ECV). All calculated concentrations at the discharge point were less than the applicable ECV.

Based on the above calculation, the existing Tech Spec limit of 150 Curies is conservative in comparison to the revised 10CFR20, Appendix B values and is therefore still applicable.

l Page 7 of 12 APPENDIX A

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APA-ZZ-01003 Rev. 5 TABLEI A.

Curie Content of Radionuclides in the Primary Spent Resin Storage Tank Ci (normalized to 150 Ci.

Ci' (in tank)

NUCLIDE total) 8.17E-01 2.91E+01 Mn-54 1.71E+01 6.10E+02 Co-58 7.19E+00 2.56E+02 Co-60 Sr-89 9.80E+00 2.75E-01 3.79E 02 g

St-90 1.35E+00 3 8.42E-02 3.00E+00 Nb-95 5.95E 02 2.12E+00 Zr-95 3.28E+01 1.17E+03 I.131 5.00E+01 1.78E+03 Cs 134 4.15E+01 1.48E+03 Cs-137 4.58E-02 1.63E+00 Ba-140 149.91 5.343E+03 TOTAL

• Values are from FSARTable 2.4 28.

)

B.

Peak Concentrations of Radionuclides at the Logan Creek Discharge Point ECV %ECV pCi/ml* (original Ci/ml (based on NUCLIDE 150 Ci total) calc) 3E-05 3E 17%

8.7E-24 Mn-54 3.lE 22 3E-06 3E 20%

3.6E 23 1.0E 24 Co-60 SE 07 67.4 %

3.4 E-07 St 90 1.2E-05 IE-06 15.4 %

5.5E 06 1.5E-07 Cs 137

• Values are from FSAR Table 2.4 30.

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APPENDIX A Page 8 of 12

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TABLE II A. Curie Content of Radionuclides in Refueling Water Storage Tank NUCLIDE Ci* (in tank) CI (normalized to 150 Ci,

~

total) i Mn-54 6.99E-06 2.19E-02 Co-58 3.36E-04 1.05E+00 Co-60 4.58E-05 1.43E-01 St-89 5.92E-05 1.85E-01

Sr-90 1.92E-06 6.02E-03 l Nb-95 1.31E-06 4.10E-03 j Zr 95 1.25E-06 3.92E 03 I-131 2.34E-02 7.33E+01 l Cs 134 1.39E-02 4.35E+0!

Cs-137 1.01E-02 3.16E+01 Ba 140 2.56E-05 8.02E-02 TOTAL 4.788E-02 149.9 l#

' Values are from FSAR Table 2.4-28.

l B. Peak Concentrations of Radionuclides at Logan Creek Discharge Point l l

l l NUCLIDE Ci/ml* (original Ci/ml (based on ECV %ECV calc) 150 Ci total)

Co40 1.lE 30 3.4E-27 3E-06 IE 19%

Sr-90 2.5E 13 7.8E 10 SE-07 0.16% l Cs 137 8.4E-13 2.6E-09 IE-06 0.26 %

l l

l

• Values are from FSAR Tabic 2.4-30 l

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Page 9 of12 APPENDIX A

APA-ZZ-01003 I

Rev. 5 TEC11 SPEC 6.8.4.F.7, DOSE RATE LIMIT FOR GASEOUS EFFLUENTS (REC 9.6)

This specification provides a gaseous emuent dose rate limit mnforming to the ECVs in 10CFR20, Appendix B. Table 2, Column 1. For the nuclides ofinterest to Callaway, the resised ECVs are nuncrically i greater, therefore the current REC is more restrictive than the dose rates conforming to the revised Appendix B values.10CFR20.1008 requires the implementation of the more restrictive of the requirements of 10CFR20, technical specifications, or any special license conditions. The current REC represents the more restrictive requirement and will be implemented without revision.

The former rule, in 20.106(a), limited the amount of radioactivity released in emuents to the concentrations spectfied in Appendix B, Table 2, averaged over a period of one year. Although not specified as a limit, this corresponded to an annual whole body dose limit of 500 mrem to the Member of the Public. The former rule did not specify a dose rate limit.

The revised rule, in 20.1301, specifies two limits on radioactivity in emuents: An annual dose limit of 100 mrem, TEDE (20.1301(a)(1)) and a dose rate limit of 2 mrem /h TEDE (20.1301(a)(2)). Note that the revised rule does not specify limits on concentration as did the former rule but does a!!ow licensees to utihze the concentration values in Appendix B, Table 2 to demonstrate compliance with the limits of 20.1301 lll (20.1302(b)(2)). Note that 20.1302(b)(2)(i) describes these as " annual average concentrations" as opposed to 3 i

instantaneous limits. Measurements and calculation means are also allowed (20.1302(b)(1)).

l l

Radiological Emuent Control (REC) 9.6 is required by Technical Specification 6.8.4.f.7 to contain:

" Limitations on the dose rate resulting from radioactive material released in gaseous emuents to areas

} beyond the SITE BOUNDARY conforming to the dos:s associated with 10CFR20, Appendix B. l Table 2, Column 1." 5 The bases for this Control state that its purpose is to en'sure that the dose at any time from gaseous emuents is g within the annual dose limit of 10CFR20, which is the dose associated with the concentrations of 10CFR20, g Appendix B, Table 2, Column 1. Additionally, this Control prosides assurance that the release of gaseous emuents will not result in the exposure of a Member of the Public to annual everage concentrations in excess of the values of 10CFR20, Appendix B, Table 2, Column 1. Note that in each case, the bases references an annual dose limit but makes no reference to a dose rate limit.

The REC establishes a release rate limit of 500 mrem /y that is equal to approximately 0.06 mrem /h. well below the dose rate limit of 2 mrem /h specified in 20.1301(a)(2), and is therefore more restnctive.

The preamble to the revised rule states that demonstration of compliance with the limits of 40CFR190 and with 10CFR50, Appendix I is sumcient to demonstrate compliance with the 100 mrem dose limit of 20.1301(a)(1).

Other Controls are provided as required Technical Specification 6.8.4.f(items 8,9, and 10) which ensure that l the limits of 40CFR190 and 10CFR50, Appendix I are not exceeded.

)

The Bases for this Control reference the concentration values of 10CFR20, Appendix B, Table 2, Column I as a l

basis for the specined dose rate limits. These values were derived using ICRP 30 calculation methodology and l the dose and dose rate values they represent are the Total Effective Dose Equivalent (TEDE) which is the f summation of the external and internal dose components. Compliance with the Control is demonstrated through calculation methodologies and parameters as established in Regulatory Guide 1.109 and NUREG 0133.

which are based on the ICRP 2 maximum organ methodology, and thus cannot be used to calculate emuent doses and dose rates that correspond to the concentration values specified in the revised 10CFR20. Appendix B.

- Table 2, Column 1.

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l The table below compares the numerical value of the former and revised Appendix B values for those nuclides most commonly reported in the Callaway Plant's gaseous efDuents:

)

10CFR20 APPENDIX B CONCENTRATION VALUES Nuclide Fonner Rule Revised Rule New/Old Kr-85 3E 7 Ci/ml 7E-7 Ci/ml 2.3 Xc-133 3E-7 SE 7 1.7 Xe-135 IE-7 7E-8 0.7 l131 IE-10 2E-10 2.0 1-133 4E 10 IE-9 2.5 Co-58 2E 9 IE-9 0.5 Co40 3E-10 SE 11 0.2 Of these, Xe 133 accounts for greater than 90% of the total activity released from the Callaway Plant in  !

gaseous efnuents for the past three years (1989-1991). The concentration value for Xe-133 actually increased I in the revise rule, as did that for Kr-85 and both iodine nuclides. Although the Co-58 and Co40 values did decrease in the revised rule, they are relatively insignificant contributors to the whole body and organ dose from gaseous efIluents discharged from the Callaway Plant as summarized below.

GASEOUS EFFLUENT ACTIVITY PROFILE 1989 - 1991

, Fraction of Total Ratio of Appendix B Nuclide Activity Released Concentration Values

)

Noble Gases:

Xe-133 0.92 1.7 Xc-135 0.04 0.7 Xc-133m 0.01 2.0 Kr-85m 0.01 1.0 Kr-85 0.01 2.3 Particulates and lodines: ,

1131 0.72 2.0 1-133 0.11 2.5 Co-58 0.03 0.5 Co40 0.14 0.2 The NRC states, in their response to Question 19, that until 10CFR50, Appendix ! is changed, licensees must continue to show compliance with Tech Specs in terms of organ and whole body doses as per Regulatory Guide 1.109. "Ile response to Question 21 states that Regulatory Guide 1.109 will not be revised at this time, thus Regulatory Guide 1.109 methodology continues to be utilized to show compliance with Tech Specs. Since the dose calculation methodology has not been revised, it would be more conservative to continue to utilize the current REC values vice dose rate limits calculated from the rnised 10CFR20, Appendix B values.

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Refer to the discussion of T/S 6.8.4.f.2 (REC 9.3) for additional details.

TECH SPEC 6.8.4.F.6, GASEOUS RADWASTE TREATMENT SYSTEM OPERABILITY (REC 9.9)

TECH SPEC 6.8.4.F.8, DOSE FROM NOBLE GASES (REC 9.7)

TECH SPEC 6.8.4.F.9, DOSE FROM IODINES AND PARTICULATES IN GASEOUS EFFLUENTS (REC 9.8)

TECH SPEC 6.8.4.F.10, TOTAL DOSE FROM THE URANIUM FUEL CYCLE (REC 9.10) l These specifications are derived from 10CFR50, Appendix ! and 40CFR190 and are not af'...d L; the revised rule. Doses continue to be calculated in accordance with Regulatory Guide 1.109 which has not beca revised.

No changes are anticipated for these specifications.

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