ML20140G233

Rev 6 to Callaway Plant Offsite Dose Calculation Manual

ML20140G233
Person / Time
Site:	Callaway
Issue date:	10/31/1996
From:	UNION ELECTRIC CO.
To:
Shared Package
ML20140G224	List: ML20140G218 ML20140G227 ML20140G233
References
PROC-961031, NUDOCS 9705080154
Download: ML20140G233 (116)

Text

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', APA-ZZ-01003 Rev. 6 RECORD OF REVISIONS Ws Rev. No. 2 Date: May,1991 Section 2.4.2 - Changed gross alpha analysis frequency from "cach batch" to a monthly composite as per Table 93-A, and the Callaway Plant NPDES pennit (reissued March 15,1991).

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 des',ribed in APA-ZZ-01011, PROCESS CONTROL PROGRAM. Addressed compliance with 10 CFR 20.1301. Ret ised 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 ofGT-RE-22/33. Incorporated the revised Rj values in Tables 3.2 and 33. Added Section 62 and Table 6.5.

Rev. No. 4 Date: September,1994 Increased the minimum channels OPERABLE requirement ofREC 9.2 for GT-RE-22 & 23 from I channel to 2 channels. Revised Action 41 and the Bases for REC 92 accordingly. Incorporated the operability requirements from Tech Spec 3.9.9 into the Action statement for clarity. (Refer to SOS 94-1176).

Rev. No. 5 Date: February,1995 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, Ni", Te'*, and from Tables 2.1,2.2,32,3J, and 3.4. Changed the reporting frequency of the Semiannual EfTluent Release Report from semiannual to ant..:al. Removed the meat, milk and vegetable pathway dispersion parameters from Tables 6.1,6.2, and 63, and clarified the applicability of the dispersion parameters and dose locations in Table 6.4. Relocated REC 9.1 and 92 to the FSAR. R,evised footnotes 3 and 7 of Table 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 oflow level radioactive waste.

Rev. No. 6 Date September,1996 2 2 Section 2: Added dose factors (A )for Ag" ", Np , Pu, Pu '*, Pp'", Am'", Cm, and Cm" to 5

Table 2.1, and Bioaccumulation Factors (Bri) for Ag, Pu, Am, and Cm to Table 22 due to a change in the liquid radwaste treatment process. Revised the description of the methodology for performing the 31 day dose projection in Section 2.5. Revised the maximum allowable background for HB-RE-18.

Section 3: Eliminated Y* and Tc" from Table 3.4 (Meat Pathway) due to a half-life of < 8 days. Substituted the phrase "more restrictive" in lieu of" lesser"in Section 3.2. Revised the definition of F. in equation 3.1. Added description of use of samples to verify dose rates in Section 33.1.2. Augmented the definition of qi in Section 33.2.1. Edited equations 3.13 and 3.14 and added equation 3.15 to clarify dose calculations. Revised the methodology for performing the 31 day dose projection in Section 3.4.

Section 4: Strengthened the discussion of the reevaluation of assumptions in Section 4.1.3.

Section 6: Added new table 6.6 to describe the selection and use of dispersion parameters during the preparation of the Annual Efiluent Release Report. Updated Tables 6.1 and 6.2 to reference the 1995 Land Use Census. There were no changes in the receptor locations.

,)- Section 8: Replaced the reference to HDP-ZZ-04500 to a more generic reference to the plant operating procedures, due to change in organizational structure and responsibilities.

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. APA-ZZ-01003 g Rev. 6 E RECORD OF REVISIONS Section 9: (1) Eliminated 9.0.1 and 9.0.2 due to redundancy with Technical Specifications 3.0.1 and 3.0.2; (2) l Revised Table 9.3 A to incorporate sampling and analysis requirements for TRU nuclides in liquid effluents; (3) E Eliminated sampling of Fuel Building Exhaust from Table 9.6 A and the associated footnotes due to redundancy with Unit Vent sampling; revised the continuous sampling requirements for the gaseous batch release points 3 consistent with plant design; revised the H' analysis frequency for Purges from weekly to " prior to each purge"; g and,(4) Revised the air sampling station location criteria on Table 9.11-A and footnote # 1, and eliminated footnote #3 in order to be less generic and more descriptive of the parameters used in determining the station locations (see SOS 95-2280). Revised the location requirements for milk and vegetables. Revised description of l use of baseline samples to trigger gamma isotopic analysis in footnote #4, revised requirement for location of downstream sample station in footnote #6. Revised Surveillance Requi ement 9.10.2.1 to eliminate liquid effluents from the surveillance. (5) Revised REC 9.5 and REC 9.9 to eliminate exceptions for partially tested effluents being released in excess of the respective limit.

Section 11: Added reference 11.14.13.

Attachment 2: Revised the Bases for REC 9.10 to support the elimination ofliquid effluents from Surveillance 9.10.2.1.

The remaining changes are editorial in nature and have no technical impact.

(This revision implements SOS's 95-2055,96-0167,96-0961,95-2280, and 96-0986).

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' ' APA-ZZ-01003 Revision 6 w

October 4,1996

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CALLAWAY PLANT ADMINISTRATIVE PROCEDURE APA-ZZ-01003 OFF-SITE DOSE CALCULATION MANUAL RESPONSIBLE DEPARTMENT HEALTH PHYSICS WRITTEN BY  ![

PREPARED BY MNA APPROVED BY

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This procedure contains the following:

Pages I through 88

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Attachments 1 through 2 s Tables through Figures 1 through 1 Appendices A through A CheckoffLists through This procedure has checkofflist(s) maintained in the mainframe computer.

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N. TABLE OF CONTENTS Section Pace Number PURPOSE AND SCOPE. . . . . . . - .1 LIQUID EFFLUENTS . . . ..- . - -~ ~ .. . 2 Li qu i d Effl u ent Monitors ... . ... .. .... ... ......... . .. ....... ... ....... . ... ....... .... ......... ...... .. ... . 2 Calculation Of Liquid Effluent Monitor Serpoints. ......................................3 Liquid Effluent Concentration Measurements ...... . . . . . . . . . . . . . . . . ..... 5 Dose Due To Liquid Effluents- .............. = . . . . . . . . . .......... . 5 The Maximum Exposed Individual.... ..... . ....................5 Calculation Of Dose From Liquid Effluents... . . . . . . . .. . 6 Summary, Calculation Of Dose Due To Liquid Effluents . . . . . . .. .. .. . 7 Liquid Radwaste Treatment System . . . . . . . . . . - . . . . . . . . . . . . . . . .7 Dose Factors ... . . ......... . . . . . . . . . . . . . . . . . . 7 GASEOUS EFFLUENTS _ _ - - 12 Gaseous Effluent Monitors . .. . .... .. .. ... . ...... ....... . ... . . . . . . . . . . . . . . . . . . 12 Gaseous Effluent Monitor Setpoints.. ... . . . . . . . . . . . . . . ~ . . . . . . . . . . . . ................13 Total Body Dose Rate Setpoint Calculations... ...... ...... .. . . . . . . . . ~ . . . . . .... ... 14 Skin Dose Rate Setpoint Calculation ...... . ... ........... .. ..... . . . . . . . . ~ . . . . . . . . . . . . . . . . . . . 15 Calculation Of Dose And Dose Rate From Gaseous Effluents.... ...... . ............... .... .. . .. . . ......... .15 NOBLE GASES 15 RADIONUCLIDES OTHER THAN NOBLE GASES.- _ _ 16 l

DOSE DUE TO GASEOUS EFFLUENTS ._ . . 17 NOBLE GASES 17 RADIONUCLIDES OTHER THAN NOBLE GASES . 17 Gaseous Rndwaste Treatment System... . . .. .. ...... .... ..... .. . . .... .. ..... ......... .. . . ..... . .... . 18 DOSE FACTORS _ . - ..~ . 19 DOSE AND DOSE COMMITMENT FROM URANIUM FUEL CYCLE SOURCES ... . 43 J Calculation Of Dose And Dose Commitment From Uranium Fuel Cycle Sources ..., ... . ..... .. . . 4 3 IDENTIFICATION OF THE MEMBER OF THE PUBLIC . . 43 TOTAL DOSE TO THE NEAREST RESIDENT ~ . 43 i

TOTAL DOSE TO THE CRITICAL RECEPTOR WITHIN THE SITE BOUNDARY. . 44 RADIOLOG1 CAL ENVIRONMENTAL MONITORING . . . . . .. 47 Description Of The Radiological Environmental Monitoring Program.. .... ... . .. . .. . . 4 7 Performance Testing Of Environmental Thermoluminescence Dosimeters - 47

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TABLE OF CONTENTS Section Pace Number DETERMINATION OF ANNUAL AVERAGE AND SHORT TERM ATMOSPHERIC DISPERSION PARAMETERS ~ .. . .. .. . . ~......... 48 l Atmospheric Dispersion Parameters . . ... ...................48 um LONG. TERM DISPERSION ESTIMATES.~ . 48 DETERMINATION OF LONG-TERM DISPERSION ESTIMATES FOR SPECIAL 48 I

RECEPTOR LOCATIONS . . . - . .. .

SHORT-TERM DISPERSION ESTIMATES ... . .. . . . .._ . 48 Annual Meteorological Data Processing = . . . . . . . . . . . . .49 REPORTING REQUIREMENTS. . . . . . ~ .._ ~ . 56 Annual Radiological Environmental Operating Report (CTSN 2804) . . . ...,. . 56 Annual Radioactive Emuent Release Repon (CTSN 2805)... . .. . . 56 IMPLEMENTATION OF ODCM METHODOLOGY (CTSN2791) - . 58 RADIOACTIVE EFFLUENT CONTROLS (REC) ~ _ . 59 Radioactive Liquid Emuent Monitoring Instrumentation.... .. . . .... . .~. ....... .... .... ... . 60 Radioactive Gaseous Emuent Monitoring Instrumentation- . . 60 Liquid Emusats Concentration. . . . . . . .

. 61 Dose From Liquid Emuents..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 64 Liquid Radwaste Treatment System ; - . . . . . . . . . . . . . . . . . -65 Gaseous Emuents Dose Rate.. ..-.. ... . . . . . . . . . . . . . . . . . . . 66 g Dose - Noble Gases ... ... . . . . . . . . . . . . .. 69 g Dose -lodine-131 And 133, Tritium, And Radioactive MaterialIn Paniculate Form .... . 70 Gaseous Radwaste Treatment System.. ... .. ... . . . , . . . . . . . . . . . ... ... 71 Total Dose ................................................ ... . . . 72 Radiological Environmental Monitoring Program . . . . . . . . . . . . . . . . . . _ . . . . . . . . . . . . =73 Radiological Environmental Monitoring Land Use Census- ....................... .. . 82 Radiological Environmental Monitoring Interlaboratory Comparison Program . . ... ... .._. . .. 83 ADMINISTRATIVE CONTROLS.. . . .. .. 84 Major Changes To Liquid And Gaseous Radwaste Treatment Systems.. . 84 Changes To The OfTsite Dose Calculation Manual (ODCM)(CTSN 2815)- . .84 REFERENCES 85 FIGURES I

FIGURE 4.1.. . . . . . . . . . . .. . .44 I

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I APA-ZZ-01003 Rev. 6 TABLE OF CONTENTS Section Pace Number TABLES TABLE 2.1 ~ .~ ~ -- ..-~ 8 TABLE 2.2 ... ~. . ~ . ~~ ~ . .. 11 TABLE 3.1 . .- .. ~ . .-.. ~ ~ . . . ..~. I9 TA B LE 3.2 ..~. ~ . . .- ~... ... . . .. ... . 2 0 TABLE 33 -INHALATION PATHWAY... ~. . . .. . . ... 22 TABLE 33 - MEAT PATHWAY . . .. . .. .... .. 24 TABLE 33 - GRASS-COW-MILK PATHWAY -. ... ... .. .... .. .... ... 26 TABLE 3.3 - GRASS-GO AT-MILK PATHWAY ...~.~ ..~....... . . ...... . ... . .... 28 TABLE 33 - VEGETATION PATHWAY . . .. . .. .... . . . . . 30 TABLE 3.4-INHALATION PATHWAY ~. . . . . 32 TABLE 3.4 - MEAT PATHWAY. . 34 TABLE 3.4 - GRASS-COW-MILK PATHWAY .. . . 36 TABLE 3.4 - GRASS-GOAT-MILK PATHWAY . . . ~.. . - ... 38 TABLE 3.4-VEGETATION PATHWAY . . . . . 40 TABLE 6.1 ~ . _ - 49 TABLE 6.2 . . 50 TABLE 63 51 TABLE 6 -. . . ... . ... . . 52 TABLE 6.5. ... . 54 TABLE 93 - A. ~ .. ~ . . -~.. . ..~ 5 5 TABLE 93 - A - -

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65 TABLE 9.11 - A -.~ . - . 68 TABLE 9.11-B. . . 68 TABLE 9.11-C . .... . . . . . . . - . . . . . - . . . . .. 68 Attachment 1 - Lower Limit of Detection ~~ .. . . . . . . . . . . . . 1 Page Attachment 2 - Bases for Radiological Efiluent Controls .. 7 pages Appendix A - Summary Review Of Radiological Efiluent Tech Spec Potentially Affected By De Implementation Of he Revised 10 CFR 20 . . . . 12 Pages

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Rev. 6 5i RECORD OF REVISIONS 4

Rev. No. O Date: March 1983 Rev. No. I Date: November,1983 i Revised to support the current RETS submittal and to incorporate NRC Staff comments.

Rev. No. 2 Date: March,1984 1

Revised to incorporate NRC Staff comments 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 Land Use Census.

Rev. No. 4 Date: February,1987 Minor clarifications, incorporated 31-day projected dose methodology Change in the utilization of areas within the Site Boundary.

Rev. No. 5 Date: January,1988 1 Minor clarifications, revised descriptions ofliquid and gaseous rad monitors, revised liquid setpoint methodology to incorporate monitor background, revised dose calculations for 40CFR190 requirements, Revised Table 6 and Figures 5.l A and 5.lB to refine descriptions of environmental TLD stations, incorporated description g; of environmental TLD testing required by Reg. Guide 4.13, revised Tables 1,2,4 and 5 to add additional nuclides, E deleted redundant material from Chapter 6.

Rev. No. 6 Date: May,1989 Revised methodology for calculating maximum permissible liquid emuent discharge rates and liquid effluent discharge rates and liquid effluent monitor setpoints, provided methodology for calculating liquid effluent g 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 setpoint required by Technical Specification 4.9.4.2 during Core 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 parameters and accompanying description per FSAR Change Notice 88-42.

APA-ZZ-01003 Rev. No. O Date: August,1989 Radiological Effluent Technical Specifications were moved from the Callaway Plant Technical Specifications to Section 9.0, Radioactive Effluent Controls, of the ODCM as per NRC Generic Letter 89-01. At the same time, in order to formalize control of the entire ODCM, it was converted to APA-ZZ-01003, OFF SITE DOSE CALCULATION MANUAL.

Rev. No. I 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 l E

operating license, issued 4/10/87.

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l. s OFF SITE DOSE CALCULATION M ANUAL

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! 1. PURPOSE 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 effluents, in the calculation of gaseous and liquid effluent monitoring Alarmffrip Setpoints, and in

the conduct of the Radiological Environmental Monitoring Program. The ODCM also contains the

. Radioactive Effluent 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 Effluent Release Reports required by Technical Specifications 6.9.1.6 and 6.9.1.7.

Compliance with the Radiological Effluent Controls limits demonstrates compliance with the limits of 10 CFR 20.1301. (Ref.11,1.1,11.2.1,11.23.3) l 1

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2. L10UID EFFLUENTS 2.1 LIOUID EFFLUENT MONITORS E

Gross radioactivity monitors which provide for automatic termination ofliquid effluent releases are B present on the liquid effluent lines. Flow rate measurement devices are present on the liquid effluent lines and the discharge line (cooling tower blowdown). Setpoints, precautions, and limitations 3 applicable to the operation of the Callaway Plant liquid effluent monitors are provided in the g appropriate Plant Procedures. Setpoint values are calculated to assure that alarm and trip actions occur prior to exceeding the Effluent Concentration Values (ECV) limits in 10 CFR Part 20 at the release point to the UNRESTRICTED AREA. The calculated alarra and trip action setpoints for the liquid effluent line mcaitors and flow measuring devices must satisfy the following equation:  !

cf sC F + f- i l

Where:

C - The liquid effluent concentration value (ECV) implementing REC 9.3.1.1 for the site in (pCi/ml). I c = The setpoint, in ( Ci/ml), of the radioactivity monitor measuring the radioactivity concentration in the effluent line prior to dilution and subsequent release; the setpoint, which is inversely related to the volumetric flow of the effluent line and directly related to the volumetric flow of the dilution stream plus the effluent steam, represents a value, ll which, if exceeded, would result in concentrations exceeding the values of 10 CFR Part 20 y l 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 = ne dilution water flow rate setpoint as measured prior to the release point, in volume per g l unit time. (If(F) is large compared to (f), then F + f a F). g (Ref. I1.8.1)

If no dilution is provided, then c 5 C.

The radioactive liquid waste stream is diluted by the plant discharge line prior to entry into the Missouri River. Normally, the dilution flow is obtained from the cooling tower blowdown, but should this become unavailable, the plant water treatment facility supplies the necessary dilution flow via a bypass line. The limiting concentration which corresponds to the liquid radwaste effluent monitor setpoint is to be calculated using methodology from the expression above.

Rus, the expression for determining the setpoint of the liquid radwaste effluent line monitor becomes:

cs C(F + f) ( Ci / ml) (2.2) ne alarm / trip setpoint calculations are based on the minimum dilution flow rate (corresponding to the dilution flow rate setpoint), the maximum effluent stream flow rate, and the actual isotopic g analysis. Due to the possibility of a simultaneous release from more than one release pathway, a y portion of the total site release limit is allocated to each pathway. ne determination and usage of the allocation factor is discussed in Section 2.2. In the event the alarm / trip setpoint is reached, an g evaluation will be performed using actual dilution and effluent flow values and actual isotopic g analysis to ensure that REC 9.3.1.1 limits were not exceeded.

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t. I e 2.1.1 Continuous Liauid Effluent Monitors

' - De radiation detection monitor associated with continuous liquid emuent releases is (Ref.11.6.1,

-11.6.2):

1 Monitor I.D. Descriptio(

BM RE 52 Steam Generator Blowdown Discharge Monitor ne Steam Generator Blowdown discharge is not considered to be radioact.i,ve unless radioactivity has been detected by the associated eff uent radiation monitor or by laboratory analysis. The i sampling frequency, minimum analysi. frequency, and type of analysis performed are as per Table 9.3 A.

. 2.1.2 Radioactive Liould Batch Release Emuent Monitors 4

! The radiation monitor which is associated with the liquid emuent batch release system is (Ref.

11.6A):

l Monitor 1.D. Descriotion i

i HB-RE-18 Liquid Radwaste Discharge Monitor f' This emuent stream is normally considered to be radioactive. De sampling frequency, minimum j; analysis frequency, and the type of analysis performed are as per Table 9.3-A.

I 2.2 CALCULATION OF LIOUID EFFLUENT MONITOR SETPOINTS f The dependence of the setpoint (c), on the radionuclide distribution, yields, calibration, and monitor i parameters, requires that several variables be considered in setpoint calculations. (Ref.11.8.1) 2.2.1 Calculation of the ECV Sum

- The isotopic concentration oithe reles;e(s) being considered must be determined his is obtained j from the analyses required per Table 9.3-A, and is used to calculate an ECV sum (ECVSUM):

I ECVSUM = ([(C,)/(E.L 6))

i = g , s , s , t. f (2.3)

Whm: 1 i c =- the concentration of each measured gamma emitting nuclide observed by gamma-ray l spectr%y of the waste sample.

! Ca= the measured concentration of alpha emining nuclides measured by gross alpha analysis

[ of the monthly composite sample.

C*= the measured concentrations of St-89 and Sr-90 as determined by analysis of the quarterly j f 3 composite sample.

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

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the measured concentration of Fe 55 as determined by analysis of the quarterly f

l composite sample, i ECV g, ECV ,s ECV,, ECV , fECV: = are the limiting concentrations of the appropriate radionuclides from 10 CFR 20, Appendix B, Table 11 Column 2. For dissolved or entrained noble gases, the concentration shall be limited to 2x10 pCi/ml total activity.

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- For the case ECVSUM si, the monitor tank effluent concentration meets the limits of REC 93.1.1

- 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 93.1.1 concentration link is sicultaneous releases are occurring or are anticipated, an allocation fraction, N, must be t pplied so that available dilution flow may be apportioned among simultaneous discharge pathways. The value of N may be any value between 0 and 1 for a particular discharge point, provided that the sum total allocation fractions for all discharge points must be sl.

2.2.2 Calculation of the Maximum Permissible Liauid Effluent Discharce Flowrate The maximum permissible liquid effluent discharge flowrate is calculated by:

f max 5 (F + f p)(SF)(N) + (ECVSUM) (2.4)

Where:

fmax = maximum permissible liquid effluent discharge flowrate, in (gn!!ons/ minute);

fp = the expected undiluted liquid effluent flowrate, in gpm.

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 g fluctuations and errors of measurements. This factor also provides a margin of safety g in the calculation of the maximum liquid effluent discharge flowraie (fmax). He value of SF should be 51.

F & ECVSUM, are previously defined.

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

/ In the event that fmax is less than fp, then the value of fmax is substituted into the equation for fp, and a new value of fmax is calculated. This substitution is performed for three iterations in order to calculate the correct value of fmax-l T

2.23 Calculation Of Liould Effluent Monitor Setpoint The liquid effluent monitors are Nal(TI) based systems and respond primarily to gamma radiation. 3 Accordingly, their setpoint is based on the total concentration of gamma emitting nuclides in the g effluent:

c = BKG + ( E (Cg) + SF ) = pCi/ml (2.5)

Where:

c = the monitor setpoint as previously defined, in (pCi/ml); g BKG = the monitor background prior to discharge, in (pCi/ml); E Cgand SF are as previously defined.

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

Utilizing the methodology of ANSI N13.10-1974 (Ref. I1.21), the background must be maintained at a value of less than or equal to 9E-6 pCi/ml(relative to Cs-137) in order to detect a change of 4E-7 pCi/mi of Cs.137. (Ref. I1.25).

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

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. . As previously stated, the monitor's response is dependent on the gamma emitting radionuclide distribution of the effluent. Accordingly, a new database conversion factor is calculated for each l t i release based upon the results of the gamma spectrometric analysis of the effluent sample and the

! measured response of the monitor to National Institute of Standards and Techr. ology (NIST)  :

traceable calibration sources: [

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f . DBCFc = h,f + (CMR) x (ECF)'. (2.6)

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DBCFc = the monitor data base conversion factor which converts count rate into concentration -  ;

,' (pCi/ml); )

CMR '= the calculated response of the radiation m anitor to the liquid effluent;

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!~ ECF = the conversion factor for Cs-137, which converts count rate into concentration ( l l Ci/ml). l Cgis as previously defined. l

' i The new value of the DBCF eis calculated and entered into the monitor data base prior to each i

. discharge. A more corrplete discussion of the derivation and calculation of the CMR is given in i reference 11.14.7.

i' 2.3 LIOUID EFFLUENT CONCENTRATION MEASUREMENTS

) - Liquid bstch releases are discharged as a discrete volume and each release is authorized based upon

! the sample analysis and the dilution flow rate existing in the discharge line at 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

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) procedural conditions, prior to authorization for release. The liquid effluent discharge is monitored prior to entering the dilution discharge line and will automatically be terminated if the pre-selected -

. alarm / trip setpoint is exceeded. Concentrations are determined primarily from the gamma isotopic  ;

and H-3 analyses of the liquid batch sample. For gross alpha, Sr-89, Sr-90, & Fe-55, the measured i i concentration from the p.evious composite analysis is used. Composite samples are collected for each batch release. Monthly analysis for gross alpha and quarterly analyses for Sr-89, Sr 90, and 4L Fe-55 are performed in accordance with Table 9.3 A. Doses from liquids discharged as continuous i j releases are calculated by utilizing the last measured va!=s of samples required in accordance with j l Table 9.3 A.

I 2.4 DOSE DUE TO LIOUID EFFLUENTS '

b 2.4.1 he Mavimum Exoosed Individual  !

i i 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 i to be the maximum exposed individual. (Ref. I1.8.3)'

ne Callaway Plant's liquid effluents are discharged to the Missouri River. As there are no potable water intakes within 50 miles of the discharge point (Ref. I1.7.1,11.6.6), this pathway does not j require routine evaluation. Therefore, the dose contribution from fish consumption is expected to account for more than 95% of the total man-rem dose from discharges to the Missouri River. Dose

from recreational activities is expected to contribute the additional 5%, which is considered to be negligible. (Ref. I1.6.7)

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. APA-ZZ-01003 Rev. 6 2.4.2 Calculation Of Dose From Liould eft 1uents The dose contributions h.- the total time period.

m

[At, t=1 are calculated at least once each 31 days and a cumulative summation of the total body and individual organ doses is maintained for each calendar quaner. Dose is calculated for all radionuclides identified in liquid effluents released to UNRESTRICTED AREAS using the following expression (Ref. I1.8.3):

D,= A,, Ar, C , F, (2.12)

i. t.1 .

Where:

D, = the cumulative dose commitmer. co the total body or any organ, t, from the liquid efiluents for the total period m

{At,

,.1 in mrem.

l 4t, = the length of the l th time period over which Cj, and F, are averaged for all liquid releases, in hours. At, corresponds to the actual duration of the release (s).

Cj, = the average measured concentration of rrdionuclide, i, in undiluted liquid effluent during time period At, from any liquid telease,in (pCi/ml).

Aj, = the site related ingestion dose commitment faai to the total body or any organ i for each identified principal gamma and beta emitter lived in Table 9.3-A, (in mrem /hr) per (pCi/ml). The calculation of the Aj, values is detaihd in Ref. I1.14.5 and are given in Table 2.1.

F, = the near field average dilution factor for Cj, during any liquid :ffluent release:

f.,

a F, = (F + f ,) 89.77 Where:

I fmax = maximum undiluted efnuent flow rate during the release F = average dilution slow 89.77 = site specific r.pplicable factor for the mixing effect of the discharge structure. (Ref.

I1.5.1)

The term Cj, is the undiluted concentration of radioactive material in liquid waste at the common 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. 6

- The nearest municipal potable water intake downstream from the liquid effluent discharge point into i the Missouri River is I cated near the city of St. Louis, Missouri, approximately 78 miles downstream. As there are currently no potable water intakes within 50 river miles of the discharge point, the drinking water pathway is not included in dose estimates to the maximally exposed individual, or in dose estimates to the population lShould future potable water intakes be constructed within 10 river miles downstream of the discharge point, then this manual will be revised to include this pathway in dose estimates. (Ref. I1.6.6).

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

[ht,

,.t is d:terinined by calculation at least once per 31 days and a cumulative summation of the total body and organ doses is maintained for each alendar quaner. The projected dose contribution from liquid effluents for which radionuclide concentrations are determined by periodic composite and grab sample analysis, may be approximated by using the last measured value. Dose contributions are determined for all radionuclides identified in liquid effluents released to UNRESTRICTED AREAS. Nuclides which are not detected in the analyses are reported as "less than" the nuclide's Minimum Detectable Activity (MDA) and are not reported as being present at the Lower Level of Detection (LLD) level for that nuclide. The "less than" values are not used m the dose calculations.

2.5 L10UID 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 of

• liquid wastes through the use of the evaporator system. The demineralization system is capable of removing radioactive ions from solutions to be reused as makeup water. Filtration is performed on cenain liquid wastes and it may, in some cases, be the only required treatment prior to release. The system has the ability to absorb halides through the use of charcoal filters prior to their release.

The design and operation requirements of the LIQUID RADWASTE TREATMENT SYSTEM provide assurance that releases of radioactive materials in liquid effluents will be kept "As Low As Reasonably Achievable"(ALARA).

The OPERABILITY of the LIQUID RADWASTE TREATMENT SYSTEM ensures this system will be available for use when liquids require treatment prior to their release to the environment.

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

Projected doses due to liquid releases to UNRESTRICTED AREAS are determined each 31 days and are equal to the average of the previous 12 months. This may be modified as appropriate to account for changes in radwaste treatment which may have a significant effect on the projected doses.

2.6 DOSE FACTORS The dose conversion factors provided in Table 2.1 were derived from the appropriate dose conversion factors of Regulatory Guide 1.109 and other sources as necessary (Ren 11.14.5 and 11.14.12) Non gamma emitting nuclides not listed in Table 9.3-A are not considered.

~

, APA-Z2-01003 Rev. 6

- TABLE 2.1 INGESTION DOSE COMMITMENT FACTOR (A g) FOR ADULT AGE GROUP (mrem /hr)per(pCi/ml)

I Nuclide Bone Liver Total Thyroid Kidney Lung GI-LLI ,

Body H-3 No Data 2.26E-01 2.26E-01 2.26E-01 2.26E-01 2.26E-01 2.26E-01 Be-7 1.30E-02 2.98E-02 1.45E-02 No Data 3.?5E-02 No Data 5.16E+00 Na-24 4.07E+02 4.07E+02 4.07E+02 4.07E+02 4.0 7E+0'! 4.07E+02 4.07E+02 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 8.35E+02 No Data 1.30E+03 No Data 1.34E+04 Mn-56 No Data 1.10E+02 1.95E+01 No Data 1.40E+02 No Data 3.52E+03 Fe-55 6.57E+02 4.54E+02 1.06E+02 No Data . No Data 2.53E+02 2.61E+02 Fe-59 1.04E+03 2.44E+03 9.34E+02 No Data No Data 6.81E+02 8.13E+03 Co-57 No Data 2.09E+01 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 Ni-65 1.26E+02 1.64E+01 7.48E+00 No Data No Data No Data 4.16E+02 Cu-64 No Data 1.00E+01 4.69E+00 No Data 2.52E+01 No Data 8.52E+02 Zn-65 2.32E+04 7.38E+04 3.33E+04 No Data 4.93E+04 No Data 4.65E+04 g

) Zn-69 Br-82 4.93E+01 No Data 9.44E+01 No Data 6.56E+00 2.27E+03 No Data No Data 6.13E+01 No Data No Data No Data 1.42E+01 2.60E+03 m '

Br-83 No Data No Data 4.04E+01 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 O Rb-86 No Data 1.01E+05 4.71E+04 No Data No Data No Data 1.99E+04 )

Rb-88 No Data 2.90E402 1.54E+02 No Data No Data No Data 4.00E-09 Rb-89 No Data 1.92E+02 1.35E+02 No Data No Data No Data O Sr-89 2.21E+04 No Data 6.35E+02 No Data No Data No Data 3.55E+03 ,

Sr-90 5.44E+05 No Data 1.34E+05 No Data No Data No Data 1.57E+04 l Sr-91 4.07E+02 No Data 1.64E+01 No Data No Data No Data 1.94E+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 Y-91M 5.44E-03 No Data 2.10E-04 No Data No Data No Data 1.60E-02 i l

Y-91 8.43E+00 No Data 2.25E-01 No Data No Data No Data 4.64E+03 Y-92 5.05E-02 No Data 1.48E-03 No Data No Data No Data 8.85E+02 Y-93 1.60E-01 No Data 4.42E-03 No Data No Data No Data 5.08E+03 lj Zr-95 2.40E-01 7.70E-02 5.21E-02 No Data 1.21E-01 No Data 2.44 E+02 B s

Zr-97 1.33E-02 2.68E-03 1.22E-03 No Data 4.04E-03 No Data 8.30E+02 g!

Nb-95 4.47E+02 2.4 SE+02 1.34E+02 No Data 2.46E+02 No Data 1.51E+06 g!

Mo-99 No Data 1.03E+02 1.95E+01 No Data 2.33E+02 No Data 2.39E+02 l

) Tc-99M 8.87E 03 2.51E-02 3.19E-01 No Data 3.81E-01 1.23E-02 1.48E+01

_/ Tc-101 9.11E-03 1.31E-02 1.29E-01 No Data 2.36E-01 6.70E-03 0 i

. APA-ZZ-01003 Rev. 6 TABLE 2.1 (Cont'd)

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

Nuclide Bone Liver Total Thyroid Kidney Lung GI-LLI Ru 103 4.42E+0C No Data 1.90E+00 No Data 1.69E+01 No Data 5.17E+02 Ru-105 3.68E.01 No Data 1.45E-01 No Data 4.76E+00 No Data 2.25E+02 Ru 106 6.57E4 01 No Data 8.32E+00 No Data 1.27E+02 No Data 4.25E+03 Cd-l09 No Data 5.54E402 1.94E+01 No Data 5.31E+02 No Data 5.59E+03 l Ag-110m 8.83E-01 8.17E-01 4.6 E-01 No Data 1.61E+00 No Data 3.33E402 y Sn-ll3 5.66E+04 1.61E+03 3.16E+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 3.30E+00 4.71E+01 Te-127m 6.47E+03 2.32E+03 7.90E+02 1.66E+03 2.63E+04 No Data 2.17E+04 Te-127 1.05E+02 3.78E+01 2.28E+01 7.80E+01 4.29E+02 No Data 8.30E+03 Te-129M 1.10E+04 4.llE+03 1.74E+03 3.78E+03 4.60E+04 No Data 5.54E+04 Te-129 3.01E+01 1.13E+01 7.33E+00 2.31E+0! 1.26E+02 No Data 2.27E+01 Te-131M 1.66E+03 8.09E+02 6.75E+02 1.28E+03 8.21E+03 No Data 8.03E+04 Te-131 1.89E+01 7.88E+00 5.96E+00 1.55E+01 8.25E401 No Data 2.67E+00

., Te-132 2.41E+03 1.56E403 1.47E+03 1.72E+03 1.50E+04 No Data 7.38E+04 1-130 2.71E+01 8.01E+01 3.16E+01 6.79E+03 1.2SE+02 No Data 6.89E+01 I-131 1.49E+02 2.14E+02 1.22E+02 7.00E+04 3.66E+02 No Data 5.64E+01 1-132 7.29E+00 1.95E+01 6.82E400 6.82E+02 3.11E+01 No Data 3.66E+00 1-133 5.10E+01 8.87E+01 2.70E+01 1.30E+04 1.55E+02 No Data 7.97E+01 1-134 3.81E+00 1.03E+01 3.70E+00 _ l.79E+02 1.64E+01 No Data 9.01E-03 1-135 1.59E+01 4.16E+01 1.54E+01 2.75E+03 6.68E+01 No Data 4.70E+01 Cs-134 2.98E+05 7.092+05 5.80E+05 No Data 2.29E+05 7.62E+04 1.24E+04 Cs-136 3.12E+04 1.23E+05 8.86E+04 No Data 6.85E+04 9.39E+03 1.40E404 Cs-137 3.82E+05 5.22E+05 3.42E+05 L Data 1.77E+05 5.89E+04 1.01E+04 Cs-138 2.64E+02 5.22E+02 2.59E+02 No Data 3.84E+02 3.79E+01 2.23E-03 Ba-139 9.29E-01 6.62E-04 2.72E-02 No Data 6.19E-04 3.76E-04 1.65E+00 Ba-140 1.94E+02 2.44E-01 1.27E+01 No Data 8.31E-02 1.40E-01 4.00E+02 Ba-141 4.50E-01 3.40E-04 1.52E-02 No Data 3.16E-04 1.93E-04 2.12E-10 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 2.54E+01 i

Ce-141 2.24E-02 1.51E-02 1.72E-03 No Data 7.03 E-03 No Data 5.78E+01 Ce-143 3.94E-03 2.92E+00 3.23E-04 No Data 1.28E-03 No Data 1.09E+02 Ce-144 1.17E+00 4.88E-01 6.26E-02 No Data 2.89E-01 No Data 3.94E+02 Pr-143 5.50E-01 2.21E-01 2.73E-02 No Data 1.27E-01 No Data 2.41 E+03 Nd-147 3.76E-01 4.35E-01 2.60E-02 No Data 2.54E-01 No Data 2.09E+03 Eu-154 3.67E+01 4.52E+P'J 3.21E+00 No Data 2.16E+01 No Data 3.27E+03

-- ) Hf-l81 W 187 3.99E-02 2.96E+02 1.94E-01 2.47E+02 1.80E-02 8.64E+01 No Data No Data 4.17E 02 No Data No Data No Data 2.21E+02 8.09E+04

. APA-ZZ-01003 Rev. 6 '

TABLE 2.1 (Cont'd) l INGESTION DOSE COMMITMENT FACTOR (Ait) FOR ADULT AGE GROUP (mrem /hr) per(pCi/ml) l Nuclide Bone Liver Total Thyroid Kidney Lung GI-LLI d

IL9_I l l Np-237 3.27E+04 2.84E+03 1.32E+03 No Data 9.85E+03 No Data 1.90E+03 Np-239 2.84E-02 2.80E-03 1.54E-03 No Data 8.72E-03 No Data 5.74E+02 Pu-238 5.69E+03 8.01E+02 1.43E+02 No Data 6.12E+02 No Date 6.11E+02 Pu-239' 6.58E+03 8.87E+02 1.60E+02 No Data 6.78E+02 No Data 5.67E+02 Pu 241 138E+02 7.06E+00 2.78E+00 No Data 1.28E+01 No Data 1.17E+01 l Am 241 4.89E+04 1.72E404 3.23E+03 No Data 2.43E+04 No Data 4.43E+03 g Cm-242 1.23E+03 1.25E+03 8.19E+01 No Data 3.72E+02 No Data 4.73E+03 g Cm-243** 3.82E+04 1.44E+04 2.24E+03 No Data 1.05E+04 No Data 4.67E+03

• Includer Pu-240 contribution

" Includes Cm-244 contribution I

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. 1 APA-ZZ-01003 -

Rev. 6 l

TABLE 2.2

\

2 BIOACCUMULATION FACTOR (Bfj),,W (oCi/ke) Der (oCi/ liter) ,

Bfj -

Element Fish (Freshwater)

H 9.0 E - 01 ]

Be 2.0 E + 00 l

'Na 1.0 E + 02 Cr 2.0 E + 02 Mn 4.0 E + 02 Fe 1.0 E + O2 ,

Co 5.0 E + 01 Ni 1.0 E + 02 Cu 5.0 E + 01 Zn 2.0 E + 03 Br 4.2 E + O2 Rb 2.0 E + 03 Sr 3.0 E + 01 Y 2.5 E + 01 ,

Zr 3.3 E + 00 Nb 3.0 E + 04 Mo 1.0 E + 01 Tc 1.5 E + 01 Ru 1.0 E + 01 Rh 1.0 E + 01

-Ag -

2.3 E + 00 -

l Cd 2.0 E + 02 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 i La 2.5 E + 01 Ce 1.0 E + 00 Pr 2.5 E + 01 Nd 2.5 E + 01 Eu 23 E + 01 Hf 3.3 E + 00 W 1.2 E + 03 Np 1.0 E + 01 Pu 3.5 E + 00 Am 2.5 E + 01 Cm 2.5 E + 01 1l

"' Values from Regulatory Guide 1.109. Rev.1. Table A 1 and Referen.es 11.14.4,11.14.8, and 11.14.13.

. APA-ZZ-01003 Rev. 6 I

3. G ASEOUS EFFLUENTS 3.1 GASEOUS EFFLUENT MONITORS Noble gas activity monitors are present on the containment building ventilation system, plant unit ventilation system, and radwaste building ventilation system.

He alarm / trip (alarm & trip) setpoint for any gaseous effluent radiation monitor is determined g based on the instantaneous noble gas total body and skin dose rate limits of REC 9.6.1.1, at the SITE BOUNDARY location with the highest annual average X/Q value.

E Each monitor channel is provided with a two level system which provides sequential alarms on g increasing radioactivity levels. These setpoints are designated as alert setpoints and alarm / trip 5 setpoints. (Ref. I 1.6.3) ne radiation monitor alarm / trip setpoints for each release point are based on the radioactive noble g gases in gaseous effluents. it is not considered practicable to apply instantaneous alarm / trip 5 setpoints to integrating radiation monitors sensitive to radioiodines, radioactive materials in particulate form and radionuclides other than noble gases. Conservative assumptions may be necessary in establishing setpoints to account for system variables, such as the measurement system efficiency and detection capabilities during normal, anticipated, and unusual operating conditions, the variability in release flow and principal radionuclides, and the time lag between alarm / trip action and the final isolation of the radioactive effluent. (Ref. I1.8.5) Table 9.2-B provides the instmment surveillance requirements, such as calibration, source checking, functional testing, and l

channel checking.

3.1.1 Continuous Release Gaseous Ef0uent Monitors The radiation detection monitors associated with continuous gaseous effluent releases are (Ref.

I1.6.8,11.6.9): j Monitor I.D. Description j GT-RE-21 Unit Vent GH-RE-10 Radwaste Building Vent Each of the above continuously monitors gaseous radioactivity concentrations downstream of the i last point of potential influent, and therefore measures effluents and not inplant concentrations.

l The unit vent monitor continuously monitors the effluent from the unit vent for gaseous i radioactivity. De unit vent. via ventilation exhaust systems, continuously purges various tanks and sumps normally containing low-level radioactive aerated liquids that can potentially generate i airbome activity. The exhaust systems which supply air to the unit vent are from the fuel building, auxiliary building, the access control area, the containment purge, and the condenser air dischar,ge. i ne unit vent monitor provides alarm functions only, and does not terminate releases from the unit vent. l The Radwaste Building ventilation effluent monitor continuously monitors for gaseous radioactivity in the effluent duct downstream of the exhaust filter and fans. The flow path provides ventilation exhaust for all parts of the building structure and components within the building and provides a discharge path for the waste gas decay tank release line. nese 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.

_ i APA-ZZ-01003 i

Rev. 6 He continuous gaseous effluent monitor setpoints are established using the methodology described in Section 3.2. Since there are two continuous gaseous efiluent release points, a fraction 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 allocated 0.3 DRL. Rese allocation factors may be changed as required to support plant operational needs, but shall not be allowed to exceed unity (i.e.,1.0). Therefore, a particular monitor reaching the setpoint would not necessarily mean the dose rate limit at the SITE BOUNDARY is being 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 The radiation monitors associated with batch release gaseous effluents are (Ref. I1.6.9,11.6.10, 11.6.11):

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

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

The Radwaste Building Vent monitor was previously described.

s 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 GASEOUS EFFLUENT MONITOR SETPOINTS_

l, The alarm / trip setpoint for gaseous effluent monitors is determined based on the more restrictive of the total body dose rate (equation 3.1) and skin dose rate (equation 3.3), as calculated for the SITE BOUNDARY.

During core alterations, the setpoint for the Containment P.irge 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 activity in the containment atmosphere sample analyzed in accordance with REC 9.6.1.1. The full derivation of this value is discussed in reference 11.14.6.

. APA-ZZ-01003 g Rev. 6 5

- 3.2.1 Total Body Dose Rate Setpoint Calculations 3

.I To ensure that the limits of REC 9.6.1.1 are met, the alarm / trip setpoint based on the total body dose rate is calculated according to:

S. s D.R.F,F, (3.1)

Where:

Stb

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

Dtb

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

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.)

Default value is Fs = 1.0.

Fa = the allocation factor which will modify the required dilutior, factor such that simultaneous gaseous releases may be made without exceeding the limits of REC 9.6.1.1.

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

P ,, = 0 + (X/Q) [ K 0 3 (3.2)

. l .

Where:

C = monitor reading of a noble gas monitor corresponding to the sample radionuclide concentrations for the batch to be released. Concentrations are determined in g

) accordance with Table 9.6-A. The mixture of radionuclides determined via grab g sampling of the effluent stream or source is correlated to a calibration factor to determine monitor response. The monitor response is bcsed on concentrations, not E release rate, and is in units of(pCi/cc). g X/Q = the highest calculated annual average relative concentration for any area at or beyond the SITE BOUNDARY in (sec/m 3). Refer to Tables 6.1,6.2 and 6.4.

Kg - the total body dose factor due to gamma emissions for each identified noble gas I radionuclide,in (mrem /yr)per(pCi/m3),(Table 3.1) l Qj = rate of release of noble gas radionuclide, i, in ( Ci/sec).

Q is calculated as the product of the ventilation path flow rate and the measured activity of the effluent stream as determined by sampling.

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. APA-ZZ-01003 Rev. 6

- 3.2.2 Skin Dose Rate Semoint Calculation I To ensure that the limits of REC 9.6.1.1 are met, the alann/ trip setpoint based on the skin dose rate is calculated according to:

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

Where:

Fsand F aare as previously defined.

Ss = the alarm / trip setpoint based on the skin dose rate.

Ds = REC 9.6.1.1 limit of 3000 mrem /yr, conservatively interpreted as a continuous release over a one year period.

R3 = 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:

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

Where:

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

1.1 = conversion factor: 1 mrad air dose = 1.1 mrem skin dose.

Mi

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

C, (X / Q) and Qj are previously defined.

3.3 CALCULATION OF DOSE AND DOSE RATE FROM G ASEOUS EFFLUENTS 3.3.1 C_alculation of Dose Rate 7h.: 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, Lj, and Mj relate the radionuclide airbome concentrations to various dose rates, assuming a semi-infinite cloud model.

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

D,=[K((X/OJQ) e i i s 500 mrem / yr (3.5)

D, = [(L +1.1 i M,)((X/ Q)O )< i s 3000 mrem /yr (3.6) 4 APA-ZZ-01003

!' Rev. 6 Where:

Qj = The release rate of noble gas radionuclides, i, in gaseous effluents, from all vent releases in (pCi/sec).

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

I 1.1 Lj, M;, K;, (X / Q), De & D, are as previously identified.

3.3.1.2 Radionuclides Other Than Noble Gases The release rate limit for lodine-131 and lodine-133, for tritium, and for all radioactive materials in particulate form with halflives greater than 8 days is determined according to (Ref. I1.8.7):

=

D o = [ R [ iX/ QJ Q, s 1500 mrem /yr (3.7) l 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 the child, based on the critical organ, in (mrem /yr) per (pCi/m3 ),

Q = The release rate of radionuclides other than noble gases, i, in gaseous effluents, from all vent releases in (pCi/sec).

(X/ Q) is as previously defined.

3 The dose parameter (Rj) includes the intemal 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 J

the limiting age group. All radiodines are assumed to be released in elemental form (ref. I1.8.7).

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

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

Where:

K' = Units conversion factor: 1E06 pCi/pCi BR = The breathing rate. (Regulatory Guide 1.109, Table E-5).

DFAj = Re 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 11.14.4) ne results of periodic tritium, iodine and particulate samples of the Unit Vent and Radwaste Vent are used to verify the dose rate limit was not exceeded for the period during which the samples or composite samples were obtained.

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)

. APA-7_Z-01003 Rev. 6

- , 3.3.2 Dose Due to Gaseous Effluents 4 ) 3.3.2.1 Noble Gases The air dose at the SITE BOUNDARY due to noble gases is calculated according to the following methodology (Ref.11.8.9):

During any calendar quarter, for gamma radiation:

D, = 3.17E-08 [ [M, (X/QJ q,) $ 5 mrad (3.9) i During any calendar quarter, for beta radiation: l f D3 = 3.17E-08 [ 'N, (X / Q) q, s 10 mrad (3.10) i During any calendar year, for gamma radiation:

4 D, = 3.17E-08 [ (M, (X/Q) q( s 10 mrad (3.11) i During any calendar year, for beta radiation:

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

I Where:

l Dg = Air dose in mrad, from gamma radiation due to noble gases released in gaseous effluent.

Db

= Air dose in mrad, from beta radiation due to noble gases released in gaseous effluents.

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

qi = The releases of noble gas radionuclides, i, in gaseous effluents, for all gaseous releases in (pCi). Releases are cumulative over the calendar quarter or year as appropriate. qi is calculated as the product of the ventilation flow rate and the measured activity of the effluent stream as determined by sampling.

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

X / Q & Mj are 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 effluents released to areas at and beyond the SITE BOUNDARY, is calculated according to the following expressions:

During any calendar quaner:

I; Dy 5 7.5 mrem (3.13)

During any calendar year:

I;D y515 mrem (3.14)

For each pathway, j, (i.e., for inhalation, ground plane, meat, cow- milk, goat- milk, and vegetation) Dyis calculated according to the expression:-

Du= 3.17E-8 I, RuJ {W) qj (3.15)

APA-ZZ-01003 g Rev. 6 5 r

m Where:

) = Dose in mrem, to a MEMBER OF THE PUBLIC from radionuclides other than D u) noble gases, from pathway 3, received by organ 1 (including total body).

= The dose factor for each identifici! radionuclide, i, in m2(mrem /yr) per (pCi/sec) or R u,;

(mrem /yr) per (pCi/m 3) as appropriate, for the pathway 3, and exposed organ 1, appropriate to the age group of the critical MEMBER OF THE PUBLIC receptor.

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/ Q)"for the food and ground plane pathways, in (meters ). Refer to Tables 6.1,6.2 and 6.4 for applicability.

(D/Q) = the average relative deposition of the effluent at or beyond the SITE BOUNDARY, considering depletion of the plume during transport.

q; = The releases of radiciodines, radioactive materials in particulate form, and radionuclides other than noble gases,1,in gaseous effluents, for all gaseous releases in ( Ci). Releases are cumulative over the calendar quarter or year as appropriate, q, is calculated as the product of ventilation flow rate and the measured activity of the effluent stream as determined by sampling.

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

X / Q is as previously defined.

For the direction sectors with existing pathways within 5 miles from the site, the appropriate Ri ,ij values are used.. If no real pathway exists within 5 miles from the center of the building complex, the cow-milk Rgj value is used, and it is assumed that this pathway exists at the 4.5 to 5.0 mile distance in the limiting-case sector. If the R g,ij for an existing pathway within 5 miles is less than a E

cow-milk R ,ij i at 4.5 to 5.0 miles, then the value of the cow milk R i,ij at 4.5 to 5.0 miles is used. 3 (Ref. I1.8.9)

Although the annual average relative concentration (X / Q) and the average relative deposition rate (D/ Q)are generally considered to be at the approximate receptor location in lieu of the SITE BOUNDARY for these calculations, it is acceptable to consider the ingestion, inhalation, and g ground plane pathways to coexist at the location of the nearest residence with the highest value of g, (X / Q) . (Ref.11.8.9) The Total Body dose from ground plane deposition is added to the dose for each individual organ. (Ref. I1.11.3) 5; 3.4 GASEOUS RADWASTE TREATMENT SYSTEM El The gaseous radwaste treatment system and the ventilation exhaust system are available for use whenever gaseous effluents require treatment prior to being released to the environment. The gaseous radwaste treatment system is designed to allow for the retention of all gaseous fission products to be discharged from the reactor coolant system. The retention system consists of eight (8) waste gas decay tanks. Normally, waste gases will be retained for at least 60 days prior to g discharge. These systems will provide reasonable assurance that the releases of radioactive materials E in gaseous effluents will be kept ALARA.

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

ll W

-l8-

_. _ ___ , _ .._ _ ~._. _ . - . _ _ . _ . . . _ . . - _ _ _ _ - _ _ _ _ . _ . - . _

. APA-ZZ-01003

Rev. 6 r .

i m Projected doses (gamma air, beta air, and organ dose) due to gaseous effluents at or beyond the

. I

~

SITE BOUNDARY are determined each 31 days and are equal to the average of the previous 12 months. This may be modified as appropriate to account for changes in radwaste treatment which may have a significant effect on the projected doses.

}

! 3.5 DOSE FACTORS The dose conversion factors provided in the following tables were derived from the appropriate dose 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 half life ofless than 8 days are not considered (Ref: 11.24). Y 90, Nb-95, La 140, and Pr-144 are included because the parent half-life is greater than 8 days, and secular equilibrium is assumed. Non-gamma emitting nuclides not listed j in Table 9.6-A are also not considered. (CTSN 43121)

J APA-ZZ-01003 Rev.6 TABLE 3.1 DOSE FACTOR FOR EXPOSURE TO A SEMI-INFINITE CLOUD OF NOBLE G ASES Total Body Gamma Air Beta Air Dose Factor Skin Dose Factor Dose Factor Dose Factor K; Li Mi Ni Radionuclide (mrem /yr)per(pCi/m3 ) (mrad /yr)per(pCi/m3 ) (mrad /yr)per( Ci/m3 ) (mradlyrper(pCi/m3 )

ECE lamawaheacm%hn adisaanEine5Alsaussudjh:ndsiJb:eaidhnaaewi auwSakJ:asas:isnZMfhiW4ssnsiECML:41Eid353 Kr-83m 7.56 E-02 --

1.93 E+01 2.88 E+02 Kr-85m 1.17E403 1.46E+ 03 1.23 E+03 1.97 E403 Kr-85 1.61 E+0i 1.34 E+03 1.72 E401 1.95 E+03 Kr-87 5.92 E403 9.73 E+03 6.17 E403 1.03 E+04 Kr-88 1.47 E+04 2.37 E403 1.52 E+04 2.93 E+03 Kr-89 1.66 E+04 1.01 E404 1.73 E+04 1.06 E404 Kr-90 1.56 E+04 7.29 E403 1.63 E+04 7.83 E+03 Xe-131m 9.15 E401 4.76 E402 1.56 E402 1.11 E403 Xe-133m 2.51 E402 9.94 E+02 3.27 E402 1.48 E403 Xc-133 2.94 E+02 3.06 E402 3.53 E+02 1.05 E403 Xe-135m 3.12 E403 7.11 E402 3.36 E+03 7.39 E402 Xc-135 1.81 E403 1.86 E+03 I.92 E403 2.46 E+03 Xc-137 1.42 E403 1.22 E+04 1.51 E403 1.27 E+04 Xe-138 8.83 E+03 4.13 E+03 9.21 E403 4.75 E403 ,

Ar-4 I 8.84 E403 - 2.69 E403 9.30 E+03 3.28 E403 M M M M M M M M M M M M W W W W m m

APA-ZZ-01003 Rev.6

~

TABLE 3.2 l

PATHWAY DOSE FACTORS (R;) FOR RADIONUCLIDES l OTHER THAN NOBLE GASES i

Ground Plane Pathway (m2mrem /yr)per(pCi/sec)

NUCLIDE TOTAL BODY , SKIN Be-7 2.24E+07 3.21 E+07 Cr 51 4.66E+06 5.51E+06 Mn-54 1.39E+09 1.63 E+09 Fe-59 2.73E+08 3.21 E+08 4 Co-57 2.98E+08 4.37E+08 Co-58 3.79E+08 4.44 E+08 Co-60 2.15E+10 2.53E+ 10 Zn-65 7.47E+08 8.59E+08 i

Rb-86 8.99E+06 1.03 E+07 Sr 89 2.16E+04 2.51 E+04 Y-90 5.36E+06 6.32E+06 l

Y-91 1.07E+06 1.21E+06 Zr-95 2.45E+08 2.84E+08 Nb-95 2.50E+08 2.94 E+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 1.54 E+08 Sn-113 1.43E+07 4.09E+07 Sb-124 8.74E+08 1.23E+09 Sb-125 3.57E+09 5.19E+09 Te-127m 9.17E+04 1.08E+05 Te-129m 1.98E+07 2.31 E+07 l I-130 5.51E+06 6.69E+06 l-131 1.72E+07 2.09E+07 l

]

e s- --- a s ..aaa a --a*-+-M -

APA-ZZ-01003 ,

Rev.6 l

~

TABLE 3.2 i

PATHWAY DOSE FACTORS (R;) FOR RADIONUCLIDES I

l OTHER THAN NOBLE GASES l

Ground Plane Pathway 1

i (m2mrem /yr)per(pCi/sec) l NUCLTDE TOTiLBODY SKIN 1132 1.25E+06 1.47E+06 1-133 2.45E+06 2.98E+06 1-134 4.47E+05 5.31E+05 g I-135 2.53E+06 2.95E+06 m Cs-134 6.85E+09 8.00E+09 Cs 136 1.51E+08 1.71E+08 Cs 137 1.03E+10 1.20E+ 10 Ba140 2.05E+07 2.35E+07 l

=

La 140 1.47E+08 1.66E+08 1.54E+07

] Cc-141 1.37E+07 E J B1 ~

Ce-144 6.96E+07 8.04E+07 Pr-144 4.35E+07 5.00E+07 g Nd-147 8.39E+06 1.01E+07 g Eu-154 2.21E+10 3.15E+10 1.97E+08 2.82E+08

. Hf-181 I'

I I

I I

) g APA-ZZ-01003 i- Ray. 6

- TABLE 3.3 CHILD PATHWA.-' .SE FACTORS (Rj) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Inhalation Pathway (mrem /yr) per (pCi/m*)

TOTAL NUCLIDE BONE LTVER 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 4.29E+04 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 1.11E+05 2.87E+03 Fe-59 2.07E+04 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 Co-58 ND 1.77E+03 3.16E+03 ND ND 1.llE+06 3.44E+04 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 7.14E+04 9.95E+05 1.63E+04 Rb-86 ND 1.98E+05 1.14E+05 ND ND ND 7.99E+03 Sr 89 5.99E+05 ND 1.72E+04 ND ND 2.16E+06 1.67E+05 Sr 90 1.01E+03 ND 6.44E+06 ND ND 1.48E407 3.43E+05 I

Y-90 4.11E+03 ND 1.11E+02 ND ND 2.62E+05 2.68E+05 Y-91 9.14E+05 ND 2.44E+04 ND ND 2.63E+06 1.84E+05 Zr 95 1.90E+05 4.1SE+04 3.70E+04 ND 5.96E+04 2.23E+06 6.11E+04 Nb-95 2.35E+04 9.18E+03 6.55E+03 ND 8.62E+03 6.14E+05 3.70E+04 Ru 103 2.79E+03 ND 1.07E+03 ND 7.03E+03 6.62E+05 4.48E+04 Ru-l06 1.3(E+05 ND 1.69E+04 ND 1.84E+05 1.43E+07 4.29E+05 Ag-110m 1.69E+04 1.14E+04 9.14E+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 5.74E+04 7.40E+02 2.00E+04 1.26E+02 ND 3.24E+06 1.64E+05 Sb-125 9.84E+04 7.59E+02 2.07E+04 9.10E+0! ND 2.32E+06 4.03E+04 Te-127m 2.49E+04 8.55E+03 3.02E+03 6.07E+03 6.36E+04 1.48E+06 7.14E+04 Te-129m 1.92E+04 6.85E+03 3.04E+03 6.33E+03 5.03E+04 1.76E+06 1.82E+05

\

_l 1130 8.1SE+03 1.64E+04 8.44E+03 1.85E+06 2.45E+04 ND 5.11E+03 23-

APA Z.Z-01003 i Rev. 6 TABLE 3.3 (Con't)

CHILD PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER TH AN NOBLE G ASE3 Inhalation Pathway I,1l i

(mrem /yr) per (pCi/m')

TOTAL NUCLIDE BONE LIVER. HODY THYROID KIDNEY LUNG GI-LLI 1

l131 4.81E+04 4.81E@4 2.73E+04 1.62E+07 7.88E+04 ND 2.84E+03 f I-132 2.12E+03 4.07E+03 1.88E+03 1.94E+05 6.25E+03 ND 3.20E+03 I I133 1.66E+04 2.03E+04 7.70E+03 3.85E+06 3.38E+04 ND 5.48E+03 I-134 1.17E+03 2.16E+03 9.95E+02 5.07E+04 3.30E+03 ND 9.55E+02 i

1-135 4.92E+03 8.73E+03 4.14E+03 7.92E+05 1.34E+04 ND 4.44 E+03 Cs-134 6.51E+05 1.01E+06 , 2.25 E+05 ND 3.30E+05 1.21E+05 3.85E+03 g, Cs-136 6.51E+04 1.71E+05 8.2SE+05 1.16E+05 1.28E+05 ND ND 9.55E+04 2.82E+05 1.4.1E+04 1.04E+05 4.18E+0a 3.62E+03 54 !

Cs-137 9.07E+05 Ba-140 7.40E+04 6.48E+01 4.33E+03 ND 2.llE+01 1.74E+06 1.02E+05 l W

La 140 6.44E+02 2.2SE+02 7.55E+01 ND ND 153E+05 2.26E+05 Cc-141 3.92E+04 1.95E+04 2.90E+03 ND 8.55E+03 5.44E+05 5.66E+04 Ce-144 6.77E+06 2.12E+06 3.61E+05 ND 1.17E+06 1.20E+07 3.89E+05 Pr-143 1.85E+04 5.55E+03 9.14E+02 ND 3.00E+03 4.33E+05 9.73E+04 Pr 144 5.96E-02 1.85E-02 3.00E-03 ND 9.77E-03 1.57E+03 1.97E+02 1.08E+04 8.73E+03 6.81E+02 ND 4.81E+03 3.28E+05 8.21 E+i,4 Nd-147 Eu-154 1.01E+07 9.21E+05 8.40E+05 ND 4.03E+06 6.14E+06 1.10E+05 Hf-181 2.78E+04 1.01E+05 1.25E+04 ND 2.05E+04 1.06E+06 6.62E+04 I

I, I

I I

I

)

g I

APA ZZ-01003 Rev. 6 4

,, TABLE 3.3 (Cont'd)

CHILD PATHWAY DOSE FACTORS (Rj) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Meat Pathway (m* mrem /yr) per (pCi/sec)

TOTAL NUCLIDE BONE LIVER BODY THYROID JilDNEY LUNG GI-LLI H-3 ND 2.34E+02 2.34E+02 2.34E+02 2.34E+02 2.34E+02 2.34E+02 Be-7 7.38E+03 1.26E+04 8.07E+03 0.00E+00 1.23E+04 0.00E+00 7.00E+05 Cr-51 0.00E+00 0.00E+00 8.80E+03 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 6.73E+06 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 3.77E+08 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 0.00E+00 6.94E+07 2.05E+08 0.00E+00 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 i

Rb-86 0.00E+00 5.77E+08 3.55E+08 0.00E+00 0.00E+00 0.00E+00 3.7]E+07 Sr-89 4.82E+08 0.00E+00 1.38E+07 0.00E+00 0.00E+00 0.00E+00 1.87E+07 Sr-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 0.00E+00 5.49E+08 ,

I \

Y-91 1.80E+06 0.00E+00 4.82E+04 0.00E+00 0.00E+00 0.00E+00 2.40E+08 Zr-95 2.67E+06 5.86E+05 5.22E+05 0.00E+00 8.39E+05 0.00E+00 6.llE+08 Nb-95 4.26E+06 1.66E+06 1.18E+06 0.00E+00 1.56E+06 0.00E+00 3.07E+09 l Ru-103 1.55E+08 0.00E+00 5.96E+07 0.00E+00 3.90E+08 0.00E+00 4.01E+09 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 4.53E+06 0.00E+00 1.06E+07 0.00E+00 6.75E+08 Cd 109 0.00E+00 1.91E+06 8.84E+04 0.00E+00 1.70E+06 0.00E+00 6.18E+06 Snll3 2.18E+09 4.48E+07 1.24E+08 3.31E+09 0.00E+00 0.00E+00 1.54E+09 Sb-124 2.93E+07 3.80E+05 1.03E+07 6.46E+04 0.00E+00 1.62E+07 1.83E+08 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.llE+08 4.25E+08 5.07E+09 0.00E+00 1.44E+09 Te-129m 1.79E+09 5.00E+08 2.78E+08 5.78E+08 5.26E+09 0.00E+00 2.19E+09 l-130 3.06E-06 6.18E-06 3.18E-06 6.80E-04 9.23E-06 0.00E+00 2.89E-06 1-131 1.66E+07 1.67E+07 9.47E+06 5.51E+09 2.74E+07 0.00E+00 1.48E+06

) 1132 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 APA-ZZ-01003 '

Rev. 6 TABLE 3.3 (Cont'd)

CHILD PATHWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Meat Pathway E1 4 I (m2mrem /yr) per (pCi/sec) j 1

TOTAL i NUCL,IDE BONE LIVER BODY THYROID KIDNEY LUNG GI-LLI l-133 5.70E-01 7.05E-01 2.67E-01 1.31E+02 1.17E+00 0.00E+00 2.84E-01 l 1134 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 l l135 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.20E+08 0.00E+00 4.69E+08 1.68E+08 8.17E+06 Cs-134 9.23E+08 1.51E+09 0.00E+00 2.38E+07 3.54E+06 1.57E+06 gI g

Cs-136 1.62E+07 4.46E+07 2.89E+07 Cs-137 1.33E+09 1.28E+09 1.89E+08 0.00E+00 4.16E+08 1.50E+08 8.00E+06 j 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 3.93E+01 0.00E+00 0.00E+00 0.00E+00 3.25E+06 Ce-141 2.22E+04 1.llE+04 1.65E+03 0.00E+00 4.86E+03 0.00E+00 1.38E+07 Cc-144 2.32E+06 7.27E+05 1.24E+05 0.00E+00 4.02E+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 3.61E+07 Pr-144 5.63E+02 1.74E+02 2.83E+01 0.00E+00 9.21E+01 0.00E+00 3.75E+05 Nd-147 1.17E+04 0.48E+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 0.00E400 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 I

I I

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I

_)

-2 6-

l APA ZZ-01003 Rsv. 6

- TABLE 3.3 (Cont'd)

CHILD PATHWAY DOSE FACTORS (R;)FOR RADIONUCLIDES OTHER THAN NOBLE GASES.

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

TOTAL NUCLIDE BONE LIVER BODY THYROID KIDNEY LUNG GI-LLI H-3 0.00E+00 1.57E+03 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 0.00E+00 7.12E+05 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+06 0.00E+00 5.89E+06 0.00E+00 1.76E+07 Fe 55 1.12E+08 5.94E+07 1.84E+07 0.00E+00 0.00E+00 3.36E+07 1.10E+07 Fe 59 1.20E+08 1.95E+08 9.70E+07 0.00E+00 0.00E+00 5.64E+07 2.03E+08 Co-57 0.00E+00 3.84E+06 7.78E406 0.00E+00 0.00E+00 0.00E+00 3.15E+07 Co 58 0.00E+00 1.21E+07 3.72E+07 0.00E+00 0.00E+00 0.00E+00 7.08E+07 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 414E+09 1.10E+10 6.86E+09 0.00E+00 6.95E+09 0.00E+00 1.94E+09 l

Rb-86 0.00E+00 8.78E+09 5.40E+09 0.00E+00 0.00E+00 0.00E+00 56.5E+08 Sr-89 6.63E+09 0.00E+00 1.89E+08 0.00E+00 0.00E+00 0.00E+00 2.57E+08 Sr-90 1.12E+11 0.00E+00 2.84E+10 0.00E+00 0.00E+00 0.00E+00 1.51E+09 Y-90 3.38E+03 0.00E+00 9.05E401 0.00E+00 0.00E+00 0.00E+00 9.62E+06 l

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 8.43E+02 7.51E+02 0.00E+00 1.21E+03 0.00E+00 8.800+05 Nb-95 3.72E+05 1.45E+05 1.03E+05 0.00E+00 1.36E+05 0.00E+0L 2.68E+08 Ru-103 4.29E+03 0.00E+00 1.65E+03 0.00E+00 1.08E+04 0.00E+00 1.llE+05 Ru 106 9.25E+04 0.00E+00 1.15E+04 0.00E+00 1.25E+05 0.00E+00 1.44E+06 Ag 110m 2.09E+08 1.41E+08 1.13E+08 0.00E+00 2.63E+08 0.00E+00 1.68E+10 Cd-109 0.00E+00 3.86E+06 1.79E+05 0.00E+00 3.45E+06 0.00E+00 1.25E+07 Snll3 6.11E+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 3.81E+07 2.40E+05 0.00E+00 6.03E+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 Te-127m 2.08E+08 5.61E+07 2.47E+07 4.98E+07 5.94E+08 0.00E+00 1.69E+08 Te 129m 2.72E+08 7.59E+07 4.22E+07 8.76E+07 7.98E+08 0.00E+00 3.31E+08 I-27-

APA-ZZ 01003 Rev. 6 TABLE 3.3 (Cont'd)

CHILD PATHWAY DOSE FACTORS (R;)FOR RADIONUCLIDES OTHER TIIAN NOBLE GASES.

Grass-Cow-Milk Pathway (m2 mrem /yr) per (pCi/sec)

TOTAL I

NUCLTDE BONE LIVER BODY THYROID KIDNEY LUNG GI-LLI l-130 1.73E+06 3.50E+06 1.80E+06 3.85E+08 5.23E+06 0.00E+00 1.64E+06 I-131 1.30E+09 1.31E+09 7.46E+08 4.34E+11 2.15E+09 0.00E+00 1.17E+08 1132 6.92E-01 1.27E+00 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 I-134 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1135 5.41E+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 g Cs-136 Cs-137 1.01E+09 3.23E+10 2.78E+09 3.09E+10 1.80E+09 4.56E+09 0.00E+00 0.00E+00 1.48E+09 1.01E+10 2.2]E+08 3.62E+09 9.78E+07 1.93E+08 5

Ba-140 1.17E+08 1.03E+05 6.84E+06 0.00E+00 3.34E+04 6.12E+04 5.94E+07 Ce-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 g Ce-144 1.62E+06 5.09E+05 8.67E+04 0.00E+00 2.82E+05 0.00E+00 1.33E+08 E 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 5.71E+05 Eu-154 9.43E+04 8.48E+03 7.75E+03 0.00E+00 3.73E+04 0.00E+00 1.97E+06 Hf-181 6.44E+02 2.35E+03 2.91E+02 0.00E+00 4.76E+02 0.00E+00 8.66E+05 I

I I

I

)

28 g

APA ZZ 01003 Rev. 6 TABLE 3.3 (Cont'd)

CIIILD PATIBVAY DOSE FACTORS (R I) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Grass-Goat-M:"t Pathway (m* mrem /yr) per (pCi/sec)

TOTAL NUCLIDE BONE LIVER BODY THYROID KIDNEY LUNG GI-LLI H-3 0.00E+00 3.20E+03 3.20E+03 3.20E+03 3.20E+03 3.20E+03 3.20E+03 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 2.11E+06 Fe-55 1.45E+06 7.72E+05 239E+05 0.00E+00 0.00E+00 436E405 1.43E+05 Fe 59 1.56E+06 2.53E+06 1.26E+06 0.00E+00 0.00E+00 734E+05 2.64E+06 Co-57 0.00E+00 4.61E+05 933E+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 0.00E+00 2.87E+07 Zn-65 4.97E+08 132E+09 8.23E+08 0.00E+00 834E+08 0.00E+00 232E+08 Rb-86 0.00E+00 1.05E+09 6.48E+08 0.00E+00 0.00E+00 0.00E+00 6.78E+07 i Sr-89 1.39E+10 0.00E400 3.97E+08 0.00E+00 0.00E+00 0.00E+00 5.39E+08 Sr-90 235E+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 1.09E401 0.00E+00 0.00E+00 0.00E+00 1.15E+06 l

Y-91 4.69E+03 0.00E+00 1.25E+02 0.00E+00 0.00E+00 0.00E+00 6.25E+05 Zr-95 4.60E+02 1.01E+02 9.01E+01 0.00E+00 1.45E+02 0.00E+00 1.06E+05 Nb-95 4.46E+04 1.74E+04 1.24E+04 0.00E+00 1.63E+04 0.00E+00 3.21E+07 Ru-103 5.14E+02 0.00E+00 1.98E+02 0.00E+00 1.29E+03 0.00E+00 133E+04 Ru-106 1.llE+04 0.00E+00 138E+03 0.00E+00 1.50E+04 .0.00E+00 1.73E+05 Ag-110m 2.5]E+07 1.69E+07 135E+07 0.00E+00 3.15E+07 0.00E+00 2.01E+09 f Cd-109 0.00E+00 4.64E+05 2.15E+04 0.00E+00 4.14E+05 0.00E+00 1.50E+06 Sn-ll3 733E+07 1.51E+06 4.18E+06 1.llE+08 0.00E+00 0.00E+00 5.18E+07 Sb-124 130E+07 1.69E+05 4.57E+06 2.88E+04 0.00E+00 7.24E+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 f Te-127m 2.50E+07 6.73E+06 2.97E+06 5.98E+06 7.13E+07 0.00E+00 2.02E+07 Te-129m 3.26E+07 9.10E+06 5.06E+06 1.05E+07 9.57E+07 0.00E+00 3.98E+07 1130 2.08E+06 4.20E+06 2.16E+06 4.62E+08 6.27E+06 0.00E+00 1.96E+06 1-131 1.57E+09 1.57E+09 8.95E+08 5.21E+11 2.58E+09 0.00E+00 1.40E+08 1132 830E-01 1.53E+00 7.02E-01 7.08E+01 234E+00 0.00E+00 1.80E+00

~

(

APA ZZ-01003 E W

Rev. 6 TABLE 3.3 (Cont'd) l CHILD PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES h Grass-Goat-Milk Pathway (m2 mrem /yr)per(pCi/sec)

TOTAL NUCLIDE BONE LIVER BODY TITYROID KIDNEY LUNG GI-LLI I-133 2.06E+07 2.55E+07 0.00E+00 9.66E+06 0.00E+00 4.74E+09 0.00E+00 4.25E+07 0.00E+00 0.00E+00 0.00E+00 1.03E+07 0.00E+00 l

m l-134 0.00E+00 I 135 6.49E+04 1.17E+05 5.53E+04 1.04E407 1.79E+05 0.00E+00 8.90E+04 g Cs-134 6.80E+10 1.12E+11 2.35E+10 0.00E+00 3.46E+10 1.24E+10 6.01E+08 g Cs-136 3.04E+09 8.35E+09 5.40E+09 0.00E+00 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 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 7.47E+00 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 0.00E400 5.75E+02 0.00E+00 1.63E+06 Cc-144 1.95E+05 6.11E+04 1.04E+04 0.00E+00 3.38E+04 0.00E+00 1.59E+07

% Pr-143 8.63E+01 2.59E+01 4.28E+00 0.00E+00 1.40E+01 0.00E+00 9.31E+04 g j Pr-144 6.05E-01 1.87E-01 3.04E-02 0.00E+00 9.89E-02 2.37E+01 0.00E+00 0.00E+00 4.03E+02 6.85E+04 5

Nd 147 5.34E+01 4.33E+01 3.35E+00 0.00E+00 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

I I

I I

I

-) I

~

l APA ZZ-0100",  !

Rev. 6 f TABLE 3.3 (Cont'd) l CHILD PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES I

Vegetation Pathway l l

(m8mrem /yr)per(pCi/sec) l I

TOTAL l NUCLIDE BONE LTVER BODY THYROID KIDNEY LUNG GT-LLT i

H3 ND 4.01E+03 4.01E+03 4.01E+03 4.01E+03 4.01E+03 4.01E+03 J

Be-7 3.38E+05 5.76E+05 3.70E+05 0.00E+00 5.65E+05 0.00E+00 3.21E+07 -

)

Cr 51 0.00E+00 0.00E+00 1.17E+05 6.50E+04 1.78E+04 1.19E+05 6.21E+06 Mn-54 0.00E+00 6.65E+08 1.77E+08 0.00E+00 1.86E+08 0.00E+00 5.58E+08 Fe-55 8.01E+08 4.2SE+08 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 l 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 4.52E+08 2.78E+08 0.00E+00 0.00E+00 0.00E+00 2.91E+07 I

Sr-89 3.60E+10 0.00E+00 1.03E+09 0.00E+00 0.00E+00 0.00E+00 1.39E+09 Sr-90 1.24E+12 0.00E+00 3.15E+11 0.00E400 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+00 8.56E+09 1

Y-91 1.86E+07 0.00E+00 4.99E+05 0.00E+00 0.00E+00 0.00E+00 2.48E+09 Zr-95 3.86E+06 8.48E+05 7.55E+05 0.00E+00 1.21E+06 0.00E+00 8.85E+08  :

Nb-95 7.48E+05 2.91E+05 2.08E+05 0.00E A) 2.74E+05 0.00E+00 5.39E+08 Ru-103 1.53E+07 0.00E+00 5.90E+06 0.00E+00 3.86E+07 0.00E+00 3.97E+08 Ru 106 7.45E+08 0.00E+00 9.30E407 0.00E+00 1.01E+09 0.00E+00 1.16E+10 Ag-110m 3.21E+07 2.17E+07 1.73E+07 0.00E+00 4.04E+07 0.00E+00 2.58E+09  ;

Cd 109 0.00E+00 2.45E+08 1.14E+07 0.00E+00 2.18E+08 0.00E+00 7.94E+08 l Sn-113 1.58E+09 3.25E+07 9.00E+07 2.40E+09 0.00E+00 0.00E+00 1.12E+09 l Sb-124 3.52E+08 4.57E+06 1.23E+08 7.77E+05 0.00E+00 1.95E+08 2.20E+09 Sb-125 4.99E+08 3.85E+06 1.05E+08 4.63E+05 0.00E+00 2.78E+08 1.19E+09 Te-127m 1.32E+09 3.56E+08 1.57E+08 3.16E+08 3.77E+09 0.00E+00 1.07E+09 )

Te-129m 8.41E+08 2.35E+08 1.31E+08 2.71E+08 2.47E+09 0.00E+00 1.03E+09  !

l130 6.16E+05 1.24E+06 6.41E+05 1.37E+08 1.86E+06 0.00E+00 5.82E+05 1-131 1.43E+08 1.44 E+08 8.17E+07 4.76E+10 2.36E+08 0.00E+00 1.28E+07

APA ZZ-01003 Rev. 6 TABLE 3.3 (Cont'd)

CHILD PATHWAY DOSE FACTORS (Rj) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Vegetation Pathway (m8 mrem /yr)per(pCi/sec)

TOTAL NUCLIDE BONE LIVER BODY THYROID KIDNEY LUNG Cl-LLI l132 9.23E+01 1.70E+02 7.80E+0) 7.87E+03 2.60E+02 0.00E+00 2.00E+02 1-133 3.53E+06 4.37E+06 1.65E+06 8.12E+08 7.2EE+06 0.00E+00 1.76E+06 1134 1.56E-04 2.89E-04 1.33E-04 6.65E-03 4.42E-04 0.00E+00 1.92E-04 1135 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+08 1.47E+08 0.00E+00 1.21E+08 1.80E 47 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 Ba-140 2.77E+08 2.43E+05 1.62E+07 0.00E+00 7.90E+04 1.45E+05 1.40E+08 La-140 3.36E+04 1.18E+04 3.96E+03 0.00E+00 0.00E+00 0.00E+00 3.28E+08 Cc-141 6.56E+05 3.27E+05 4.86E+04 0.00E+00 1.43E+05 0.00E+00 4.08E+08 0.00E+00 Ce-144 1.27E+08 3.98E47 6.78E+06 0.00E+00 2.21E+07 1.04E+10 l

~

) Pr-143 Pr-144 1.46E+05 7.88E+03 4.37E+04 2.44E+03 7.23E+03 3.97E+02 0.00E+00 0.00E+00 2.37E+04 1.29E+03 0.00E+00 0.00E+00 1.57E+08 5.25E+06 up Nd 147 7.15E+04 5.79E+04 4.48E+03 0.00E+00 3.18E+04 0.00E+00 9.17E+07 g Eu-154 1.66E+0S 1.50E+07 1.37E+07 0.00E+00 6.57E+07 0.00E+00 3.48E+09 g Hf-181 4.90E+05 1.79E+06 2.21E45 0.00E+00 3.63E+05 0.00E+00 6.59E'+08 I

I I

I I

I

)

I

APA-ZZ-01003 Rev. 6

_. TABLE 3.4 ADULT PATHWAY DOSE FACTORS (R i) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Inhah Lion Paibway (mrem /yr) per(pCi/m 3)

TOTAL NUCLIDE BONE LIVER BODY THYROID KIDNEY LUNG GI-LLI H-3 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 4.21E+04 5.35E+03 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+06 ~7.74E+04 Fe-55 2.46E+04 1.70E+04 3.94E+03 ND ND 7.21E+04 6.03E+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 3.24E+04 1.03E+05 4.66E+04 ND 6.90E+04 8.64E+05 5.34E+04 Rb-86 ND 1.35E+05 5.90E+04 ND ND ND 1.66E+04 Sr89 3.04E45 ND 8.72E+03 ND ND 1.40E+05 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 1.70E+05 5.06E+05 I

Y-91 4.62E+05 ND 1.24E+04 ND ND 1.70E+06 3.85E+05 Zr-95 1.07E+05 3.44E+04 2.33E+04 ND 5.42E+04 1.77E+06 1.50E+05 Nb-95 1.41E+04 7.82E+03 4.21E+03 ND 7.74E+ 03 5.05E+05 1.04E+05 Ru-103 1.53E+03 ND 6.58E+02 ND 5.83E+03 5.05E+05 1.10E+05 1

Ru-106 - 6.91E+04 ND 8.72E+03 ND 1.34E+05 9.36E+06 9.12E+05 Ag-110m 1.08E+04 1.00E+04 5.94E+03 ND 1.97E+04 4.63E+06 3.02E+05 Cd-109 - ND 3.67E+05 1.31E44 ND 3.ME+td 6.83E+05 5.82E+04 Sn-113 5.72E+04 2.18E+03 4.39E+03 1.24E+03 ND 9.44E+05 1.18E+05 Sb 124 3.12E+04 5.89E+02 1.24E+04 7.55E+01 ND 2.48E+06 4.06E+05 Sb-125 5.34E+04 5.95E+02 1.26E+04 5.40E+01 ND 1.74E+06 1.01E+05 Te 127m 1.26E+04 5.77E+03 1.57E+03 3.29E+03 4.58E+04 9.60E+05 1.50E+05 Te-129m 9.76E+03 4.67E+03 1.58E+03 3.44E+03 3.66E+04 1.16E+06 3.83E+05 l

l-130 4.58E+03 1.34E+04 5.28E+03 1.14E+06 2.09E+04 ND 7.69E+03 l-131 2.52E+04 3.58E+04 2.0$E+04 1.19E+07 6.13E+04 ND 6.28E+03 I

APA 22-01003 Rev. 6 l

TABLE 3.4 (Cont'd) i ADULT PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN '

NOBLE GASES Inhalation Pathway 1 (mrem /yr) per (pCl/m')

TOTAL GI-LLI l

=

NUCLIDE BONE LIVER BODY THYRO 1D KIDNEY LUNG l132 1.16E+03 3.26E+03 1.16E+03 1.14E+05 5.18E+03 ND 4.06E+02 g, I-133 8.64E+03 1.48E+04 4.52E+03 2.15E+06 2.58E+04 ND 8.88E+03 E I-134 6.44E+02 1.73E+03 6.15E+02 2.98E+04 2.75E+03 ND 1.0lE+00 1-135 2.68E+03 6.98E+03 2.57E+03 4.48E+05 1.11 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 1.17 E+04 4.78E+05 6.21E+05 4.28E+05 ND 2.22E+05 7. 2E+04 8.40E+03 Cs 137 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 4.58 E+05 1.99E+04 1.35E+04 1.53E+03 ND 6.26E+03 3.62E+05 1.20E+05 Ce-141 3.43E+06 1.43E+06 1.84E+05 ND 8.48E+05 7.78E+06 8.16E+05 Cc-144 Pr-143 9.36E+03 3.75E+03 4.64E+02 ND 2.16E+03 2.81E+05 2.00E+05 3.01E-02 1.25E-02 1.53E-03 ND 7.05E-03 1.02E+03 2.15E-08 Pr-144 5.27E+03 6.10E+03 3.65E+02 ND 3.56E+03 2.21E+05 1.73 E+05 Nd 147 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 H 2E+04 6.32E+03 ND 1.4EE+61 6.85E+05 1.39E+05 I

I I

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J g I

-34

APA-ZZ-01003 Rev. 6

. , . TABLE 3.4 (Cont'd)

ADULT PATHWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTHER THAN NOBLE GASES.

Meat Pathway ]

1 (m8 mrem /yr) per(pCi/sec)

TOTAL 1 NUCLlDE BONE LIVER BOpY THYROID KIDNEY LUNG GI-LLI l H-3 ND 3.25E+02 3.2SE+02 3.25E+02 3.25E+02 3.25E+02 3.25E+02 Be-7 4.57E+03 1.04E+04 5.07E+03 ND 1.10E+04 ND 1.81EiO6 Cr 51 ND ND 7.04E+03 4.21E+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.SIE+07 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+06 9.36E+06 ND ND ND 1.43E+08 Co-58 ND 1.82E+07 4.08E%7 ND ND ND 3.69E+08 Co-60 ND 7.51E+07 1.66E+08 ND ND ND~ 1.41E+09 Zn-65 3.56E+08 1.13E+09 5.11E+08 ND 7.57E+08 ND 7.13E+08 Rb-86 ND 4.87E+08 2.27E+08 ND ND ND 9.60E+07 Sr-89 3.01E+08 ND 8.65E+06 ND ND ND 4.83E+07 Sr-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+99 I

Y-91 1.13E+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 l

Ru-103 1.05E+08 ND 4.53E+07 ND 4.01E+08 ND 1.23h+10 Ru-106 2.80E+09 ND 3.54E+08 ND 5.40E+09 ND 1.81E+11 Ag-110m 6.68E+06 6.18E+06 3.67E+06 ND 1.21E+07 ND 2.52E+09 Cd-10'.C 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 7.84E+06 4.79E+04 ND 1.54E+07 5.61E+08 Sb-125 1.91E+07 2.13E+05 4.54E+06 1.94E+04 ND 1.47E+07 2.10E+08 Te-127m 1.11E+09 3.98E+08 1.36E+08 2.85E+08 4.53E+09 ND 3.74E+09 Te-129m 1.13E+09 4.23E+08 1.79E+08 3.89E+08 4.73E+09 ND 5.71E+09 l130 2.12E 6.27E-06 2.47E-06 5.31E-04 9.78E-06 ND 5.40E-06 I-131 1.0SE+07 1.54E+07 8.82E+06 5.04E+09 2.64E+07 -ND 4.06E+06 35-

APA ZZ-01003 Rev. 6 TABLE 3.4 (Cont'd)

ADULT l'ATHWAY DOSE FACTORS (Ri ) FOR RADIONUCLIDES OTHER THAN ,

NOBLE GASES.

Meat Pathway (m8mrem /yr) per(pCi/sec)

'OTAL NUCLTDE BONE LIVER PJDY THYROID KIDNEY LUNG GI-LLI l-132 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 ND 0.00E+00 1-133 3.67E-01 6.39E-01 1.95E-01 9.38E+01 1.llE+00 ND 5.74E-01 1-134 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 ND 0.00E+00 1-135 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 ND 0.00E+00 Cs-134 6.57E+0S 1.56E+09 1.28E+09 ND 5.06E+08 1.68E+08 2.74E+07 Cs-136 1.20E+07 4.76E+07 3.42E+07 ND 2.65E+07 3.63E+06 5.40E+06 Cs-137 8.71E+08 1.19E+09 7.31E+08 ND 4.04E+08 1.34E+08 2.31E+07 Ba-140 2.87E+07 3.61E+04 1.88E46 ND 1.23E+04 2.07E+04 5.91E+07 La-140 2.21E+02 1.llE+02 9.49E+03 2.94E+01 1.0SE+03 ND ND ND 4.41E+03 ND ND 8.18E+06 3.63E+07 h

Ce-141 1.40E+04 Ce-144 1.46E+06 6.09E+05 7.82E+04 ND 3.61E+05 ND 4.92E+08

1. ) Pr-143 Pr-144 2.10E+04 3.52E+02 8.40E+03 1.46E+02 1.04E+03 1.79E+01 ND ND 4.85E+03 8.24E+01 ND ND 9.18E+07 5.06E-05 Nd-147 7.07E+03 8.17E+03 4.89E+02 ND 4.77E+03 ND 3.92E+07 Eu-154 8.02E46 9.86E+05 7.01E+05 ND 4.72E+06 ND 7.14b . 98 Hf-181 3.01E^06 1.46E+07 1.35E+06 ND 3.14E+06 ND 1.66E+10 I

I I

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I J E 36-E

APA ZZ-01003 )

R v.6 q TABLE 3.4 (Cont'd)

ADULT PATHWAY DOSE FACTORS (Ri ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES -

Grass-Cow-Milk Pathway (m' mrem /yr) per (pCi/sec)

TOTAL NUCLTDE BONE LIVER BODY THYROID KIDNEY LUNG GILLI H-3 ND 7.63E+02 7.63E+02 7.63E+02 7.63E+02 7.63E+02 7.63E+02 Be-7 1.63E+03 3.72E+03 1.81E+03 ND 3.93E+03 ND 6.45E+05 Cr-51 ND ND 2.86E+04 1.71E+04 630E+03 3.79E+04 7.19E+06 Mn 54 ND 8.42E+06 1.61E+06 ND 2.50E+06 ND 2.58E+07 Fe-55 2.51E+07 1.74E+07 4.05E+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 233E+08 Co-57 ND 1.28E+06 2.13E+06 ND ND ND 3.25E+07 Co-58 ND 4.72E+06 1.06E+07 ND ND ND 9.56E+07 Co-60 ND 1.64E+07 3.62E+07 ND ND ND 3.08E+08 j

% Zn-65 137E+09 437E+09 1.97E+09 ND 2.92E+09 ND 2.75E+09 i

Rb-86 ND 2.60E+09 1.21E+09 ND ND ND 5.12E+08 St-89 1.45E+09 ND 4.17E+07 ND ND ND 2.33E+08 St-90 4.68E+10 ND 1.15E+10 ND ND ND 135E+09 Y-90 7.43E+02 ND 1.99E+01 ND ND ND 7.87E+06 l

Y-91 8.59E+03 ND 230E+02 ND ND ND 4.73E+06 Zr-95 9.44E+02 3.03E+02 2.05E+02 ND 4.75E+02 ND 9.59E+05 Nb-95 9.65E+04 537E+04 2.89E+04 ND 531E+04 ND 3.26E+08 Ru-103 1.02E+03 ND 439E+02 ND 3.89E+03 ND 1.19E+05 Ru-106 2.04E+04 ND 2.58E+03 ND 3.94E+04 ND 132E+06 Ag-110m 5.82E+07 539E+07 3.20E+07 ND 1.06E+08 ND 2.20E+10 Cd-109 ND 1.13E+06 3.95E+04 ND 1.08E+06 ND 1.14E+07 Sn-ll3 134E+08 1.81E+06 7.73E+06 2.18E+06 ND ND 4.02E+08 Sb-124 2.57E+07 4.86E+05 1.02E+07 6.24E+04 ND 2.00E+07 73]E+08 Sb-125 2.04E+07 2.28E+05 4.87E+06 2.08E+04 ND 1.58E+07 2.25E+08 Te-127m 4.58E+07 1.64E+07 5.58E+06 1.17E+07 1.86E+08 ND 1.54E+08 Te-129m 6.02E+07 2.25E+07 9.53E+06 2.07E+07 2.51E+08 ND 3.03 E+08 1-130 4.21E+05 1.24E+06 4.91E+05 1.05E+08 1.94E+06 ND 1.07E+06 1-131 2.97E+08 4.25E+08 2.43E+08 139E+11 7.28E+08 ND 1.12E+08 APA-ZZ-01003 Rev.6 TABLE 3.4 (Cont'd)

ADULT PATHWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Grass-Cow-Milk Pathway g (m' mrem /yr) per (pCi/sec)

TOTAL NUCLIDE BONE LIVER BODY THYROID KIDNEY LUNG GTLLI I132 1.65E-01 4.42E-01 1.55E-01 1.55E+01 7.04E-01 ND 8.30E-02 1 133 3.88E+06 6.75E+06 2.06E+06 9.92E+08 1.18E+07 ND 6.07E+06 I-134 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 ND 0.00E+00 I-135 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 7.93E+07 1.18E+08 Cs-137 7.38E+09 1.01E+10 6.61E+09 ND 3.43E+09 1.14E+09 1.95E+08 Ba 140 2.69E+07 3.38E+04 1.76E+06 ND 1.15E+04 1.93E+04 5.54E+07 g La-140 4.14E+01 2.09E+01 5.51E+00 ND ND ND 1.53E+06 E

Ce-141 4.85E+03 3.28E+03 3.72E+02 ND 1.52E+03 ND 1.25E+07  ;

Ce-144 3.58E+05 1.50E+05 1.92E+04 ND 8.87E+04 ND 1.21E+08 l!

Pr-143 1.58E+02 6.34E+01 7.83E+00 ND 3.66E+01 ND 6.92E+05 W Pr 144 1.10E+00 4.58E-01 5.61E-02 ND 2.58E-01 ND 1.59E-07 g Nd-147 9.42E+01 1.09E+02 6.51E+00 ND 6.36E+01 ND 5.23E+05 g Eu 154 2.37E+04 2.91E+03 2.07E+03 ND 1.39E+04 ND 2.llE+06 Hf-181 1.42E+02 6.92E+02 6.41E401 ND 1.49E+02 ND 7.87E+05 I,

I I

l' i l

) I:

-3 8-

). APA-ZZ-01003

- Rev. 6 ,

TABLE 3.4 (Cont'd)

ADULT PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN

[ NOBLE GASES . l Grass-Goat-Milk Pathway (m' mrem /yr) per (pCi/sec)

TOTAL NUCLIDE BONE LTVER BODY THYROID KIDNEY LUNG GI-LLI I H-3 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 a

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

! Fe 55 3.27E+05 2.26E+05 5.26E+04 ND ND 1.26E+05 130E+05 ,

Fe-59 3.87E+05 9.08E+05 1.48E+05 ND ND 2.54E+05 3.03E+06 )

Co 57 ND 1.54E+05 2.56E+05 ND ND ND 3.90E+06 Co-58 ND 5.66E+05 1.:l7E+06 ND ND ND 1.15E+07 Co-60 ND 1.97E+06 4.35E+06 ND ND ND 3.70E+07 j l

Zn-65 1.65E+08 5.24E+08 237E+08 ND 3.51E+08 ND 330E+08

]

Rb-86 ND 3.12E+08 1.45E+08 ND ND ND 6.15E+07 )

i Sr-89 3.05E+09 ND 8.75E+07 ND ND ND 4.89E+08 l Sr-90 9.84E+10 ND 2.41E+10 ND ND ND 2.84E+09 Y-90 8.92E+01 ND 239E+00 ND ND ND 9.46E+05 I

- Y-91 1.03E+03 ND 2.76E+01 ND ND ND 5.68E+05 Zr 95 1.13E+02 3.63C ')1 2.46E+01 ND 5.70E+01 ND 1.15E+05  ;

s Nb-95 1.16E+04 6.45E+03 3.47E+03 ND 637E+03 ND 3.91E+07 i l Ru-103 1.22E+02 ND 5.27E+01 ND 4.67E+02 ND 1.43E+04 i i 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 3.84EM6 ND 1.27E+07 ND 2.64E+09 Cd 109 ND IJ6E+05 4.74E+03 ND 1.30E+05 ND 137E+06 Sn-ll3 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 7.50E403 ND 2.41E+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 Te 127m 5.50E+06 1.97E+06 6.70E+05 1.41E+06 2.23E+07 ND 1.84E+07 Te 129m 7.23E+06 2.70E+06 1.14E+06 2.48E+06 3.02E+07 ND 3.64 E+07 j

4 1-130 5.05E+05 1.49E+06 5.88E+05 1.26E+08 232E+06 ND 1.28E+06 1-131 3.56E+08 5.09E+08 2.92E+08 1.67E+11 8.72E+08 ND 1.34E+0S I-132 1.98E-01 5.29E-01 1.85E-01 1.85E+01 8.43E-01 ND 9.95E-02

]

APA ZZ-01003 Rev. 6 TABLE 3.4 (Cont'd)

ADULT PATHWAY DOSE FACTORS (Rj) FOR RADIONUCLIDES OTHER THAN g NOBLE GASES . 5 Grass-Goat-Milk Pathway (m2mrem /yr)per(pCi/sec)

TOTAL NUCLTDE BONE LTVER bop _Y' THYROID }<iDNEY LUNG GI-LLI l133 4.65E+06 8.09E+06 2.47E+06 1.19E+09 1.41E+07 ND 7.27E+06 1-134 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 ND 0.00E+00 1135 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 330E+10 ND 131E+10 434E+09 7.07E+08 ='

Cs-136 7.91E+08 3.12E+09 2.25E+09 ND 1.74E+09 238E+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 Ba-140 3.23E+06 4.06E+03 2.12E+05 ND 138E+03 232E+03 6.65E+06 La-140 4.97E+00 2.51E+00 6.62E-01 ND ND ND 1.84E+05 Cc-141 5.82E+02 3.94E+02 4.46E+01 ND 1.83E+02 ND 1.50E+06 Cc-144 430E+04 1.80E+04 231E+03 ND 1.07E+04 ND 1.45E+07

\ Pr-143 1.90E+01 7.61E+00 9.40E-01 ND- 439E+00 ND 8.31E+04 l j Pr-144 Nd-147 1.33E-01 5.50E-02 6.74E-03 ND 3.10E-02 ND 1.91E-08 W 1.13E+01 131E+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 Hf-181 1.71E+01 831E+01 7.70E+00 ND 1.79E+01 ND 9.46E+04 I

I R:

I I

I J t g

APA-ZZ-01003 Rzv. 6

^

TABLE 3.4 (Cont'd)

(

4 ADULT PATHWAY DOSE FACTORS (Ri ) FOR RADIONUCLIDES OTHER THAN NOBLE GASES -

Vegetation Pathway i (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 2.26E+03 2.26E+03 l Be-7 9.24E+04 2.11E+05 1.03E+05 ND 2.23E+05 ND 3.66E+07  ;

1 Cr 51 ND ND 4.64E+04 2.78E+04 1.02E+04 6.16E+04 1.17E+07 Mn-54 ND 3.13E+08 5.97E+07 ND 931E+07 ND 9.59E+08 1 Fe-55 2.10E+08 1.45E+08 338E+07 ND ND 8.08E+07 '831E+07 l Fe-59 1.26E+08 2.96E+08 1.14E+08 ND ND 8.28E+07 9.88E+08 i Co-57 ND 1.17E+07 1.95E+07 ND ND ND 2.97E+08 Co ND 3.07E+07 6.89E+07 ND ND ND 6.23E+08 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 636E+08 Rb-86 ND 2.19E+08 1.02E408 ND ND ND 433E+07 -)

Sr-89 9.97E+09 ND 2.86E+08 ND KD ND 1,60E+09 Sr-90 6.0SE+11 ND 1.48E+11 ND ND ND 1.75E+10 I Y-90 7.67E+05 ND 2.06E+04 ND ND l ND 8.14E+09 i

l.

Y-91 5.11E+06 ND 137E+05 ND ND ,

ND 2.81E+09 Zr-95 1.17E+06 3.77E+05 2.55E+05 ND 5.91E+05 ND 1.19E+09 Nb-95 2.40E+05 134E+05 7.19E+04 ND 132E+05 ND 8.llE+08 Ru-103 4.77E+06 ND 2.06E+06 ND 1.82E+07 ND 5.57E+08 Ru-106 1.93E+0? ND 2.44E+07 ND 3.72E+08 ND 1.25E+10 Ag 110m LC'5L C 9.75E+06 5.79E+06 ND 1.92E+07 ND 3.98E+09 Cd-109 0.00E+00 836E+07 2.92E+06 ND 8.00E+07 ND 8.43E+08 Sn-113 4.16E+08 1.18E+07 2.40E+07 6.75E+06 ND ND 1.25E+09 Sb-124 1.04E+08 1.96E+06 4.l lE+07 2.51E+05 ND 8.07E+07 2.94E+09 Sb-125 137E+08 1.53E+06 3.25E+07 139E+05 ND 1.05E+08 1.50E+09 Te 127m 3.49E+08 1.25E+08 4.26E+07 8.92E+07 1.42E+09 ND -1.17E+M Te-129m 2.51E+08 938E+07 3.98E+07 8.64E+07 1.05E+09 ND 1.27E+09 l-130 3.93E+05 1.16E+06 4.57E+05 9.81E+07 1.81E+06 ND 9.97E+05 1-131 8.0SE+07 1.16E+08 6.62E+07 3.79E+10 1.98E+08 ND 3.05E+07

APA ZZ-01003 h WF Rev. 6 TAllLE 3.4 (Cont'd)

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

Vegetation Pathway (m* mrem /yr)per( Ci/sec)

TOTAL NUCLIDE DONE LIVER BODY TIIYROID KIDNEY LUNG GI-LLI l-132 5.77E+01 1.54E+02 5.40E+01 5.40E+03 2.46E+02 ND 2.90E+01 1-133 2.09E+06 3.63E+06 1.llE+06 5.33E+08 633E+06 ND 3.26E+06 1-134 9.69E-05 2.63E-04 9.42E-05 4.56E-03 4.19E-04 ND 230E-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+0S Cs 136 4.27E+07 1.69E+08 1.21E+08 ND 9.38E+07 1.29E+07 1.91E+07 Cs-137 636E+09 8.70E+09 5.70E+09 ND 2.95E+09 9.81E+08 1.68E+08 Ba-140 1.29E+08 1.61E+05 8.42E+06 ND 5.49E+04 9.24 E+04 2.65E+08 g La-140 1.58E+04 7.98E+03 2.llE+03 ND ND ND 5.86E+08 Ce-141 1.97E+05 1.33E+05 1.51E+04 ND 6.19E404 ND 5.10E+08 E

Cc-144 3.29E+07 1.38E+07 1.77E+06 ND 8.16E+06 ND 1.llE+10 l Pr-143 6.26E+04 2.51E404 3.10E+03 ND 1.45E+04 ND 2.74E+08 up Pr-144 2.03E+03 8.43E+02 1.03E+02 ND 4.75E+02 ND 2.92E-04 g Nd-147 333E+04 3.85E+04 231E+03 ND 2.25E+04 ND 1.85E+08 g Eu-154 4.85E+07 5.97E+06 4.25E+06 ND 2.86E+07 ND 432E+09 Hf-181 1.40E+05 6.82E+05 632E+04. ND 1.47E+05 ND 7.76E+08 I

I I

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i ) 3 2

g

APA-ZZ-01003 i Rev. 6 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 THE PUBLIC for Uranium Fuel Cycle Sources is determined as:

a) Dose to the total body and internal organs due to gamma ray exposure from submersion in a cloud of radioactive noble gases, ground plane exposure, and direct radiation from the Unit and outside storage tanks; b) Dose to skin due to beta radiation from submersion in a could 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 ingestion of radioactive material.

It is assumed that total body dose from sources of gamma radiation irradiates internal body organs at the same numerical rate. (Ref. I1.12.5)

The dose from gaseous effluents is considered to be the summation of the dose at the individual's residence and the dose to the individual fr',m activities within the SITE BOUNDARY.

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

Rerefore, only doses to individuals via airborne pathways and doses resulting from direct radiation are considered in determining compliance to 40 CFR 190 (Ref. I1.12.3).

There are no other Uranium Fuel Cycle Scarce: within 8' m of the Callaway Plant.

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

occupationcily associated with the Callaway Plant, but who may use portions of the plant site for recreational or other pumoses not associated with the plant (Ref. I1.4 and 11.8.10). Accordingly, it is necessary to ch aracterize this individual with respect to his utilization of areas both within and at or beyond the SITE BOUNDARY and identify, as far as possible, major assumptions which could be reevalu?(W ;f necessary to demonstrate continued compliance with 40 CFR 190 through the use of more realk.ie assumptions -(Ref. I1.12.3 and 11.12.4).

The evaluation of Tool Dose from the Uranium Fuel Cycle should consider the dose to two Critical Receptors: a) Re Nearest Resident, and b) The Critical Receptor within 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 gases, ground plane exposure, and inhalation and ingestion pathways. It is conservatively assumed that each ingestion pathway (meat, milk, and vegetation) exists at the location of the Nearest Resident.

It is assumed that direct radiation dose from operation of the Unit and outside storage tanks, and dose from gaseous effluents due to activities within the SITE BOUNDARY, is negligible for the Nearest Resident. The total Dose from the Uranium Fuel Cycle to the Nearest Resident is calculated using the methodology discussed in Section 3, using concurrent meteorological data for the location of the Nearest Resident with the highest value of X/Q.

The location of the Nearest Resident in each meteorological sector is determined from the Annual Land Use Census conducted in accordance with the Requirements of REC 9.12.1.1.

t 1

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APA-ZZ-01003 Rev. 6 4.1.3 Total Dose to the Critical Recetstor Within the SITE BOUNDARY ne Union Electric Company has entered into an agreement with the State of Missouri Department of Conservation for management of the residual lands surrounding the Callaway Plant, including g some areas within the SITE BOUNDARY. Under the tenns of this agreement, certain areas have 5 been opened to the public for low intensity recreational uses (hunting, hiking, sightseeing, etc.) but recreational use is excluded in an area immediately surrounding the plant site (re^r to Figure 4.1). g Much of the residual lands within the SITE BOUNDARY are leased to area farn s by the g Department of Conservation to provide income to support management and development costs.

Activities conducted under these leases are primarily comprised of farming (animal feed), grazing, and forestry (Ref. I 1.7.2,11.7.3,11.13, and 11.13.1).

Based on the utilization of nreas within the SITE BOUNDARY, it is reasonable to assume that the critical receptor within the SITE BOUNDARY is a farmer, and that his dose from activities within the SITE BOUNDARY is due to exposure incurred while conducting his farming activities. The current tenant has estimated that he spends approximately 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> per year working in this area (Ref. I1.5.5). Occupancy of areas within the SITE BOUNDARY is assumed to be averaged over a period of one year.

Any reevaluation of assumptions should consider only real receptors and real pathways using realistic assumption, and should include a reevaluation of the occupancy period at the locations of real exposure (e.g. a real individual would not simultaneously exist at each point of maximum exposure).

4.1.3.1 Total Dose to the Farmer from Gaseous Effluents g

The Total Dose to the farmer from gaseous effluents is calculated for the adult age group using the M methodology discussed in Section 3, utilizing concurrent meteorological data at the farmer's residence and historical meteorological data from Table 6.1 for activities within the SITE g BOUNDARY. Dese dispersion parameters were calculated by assuming that the farmer's time is g equally distributed over the areas farmed within the SITE BOUNDARY, and already have the total 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 sector. The gaseous efiluents dose at the farmer's residence is due to plume exposure from Noble Gases and the ground plane, inhalation, and ingestion pathways. For conservatism, it is acceptable to assume that g all of the ingestion pathways exist at this location.

5 It is assumed that food ingestion pathways do not exist within the SITE BOUNDARY, therefore the gaseous effluents dose within the SITE BOUNDARY is due to plume exposure from Noble Gases 3 and the ground plane and inhalation pathways. g 4.1.3.1.1 Direct Radiation Dose from Outside Storace Tanks ne Refueling Water Storage Tank (RWST) has the highest potential for receiving significant amounts of radioactive materials, and constitutes the only potentially significant source of direct radiation dose from outside storage tanks to a MEMBER OF THE PUBLIC (Ref. I1.6.14,11.6.15, 11.6.16 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, which is 450 meters from the RWST.

He RWST is a right circular cylinder approximately 12 meters in diameter,14 meters in height wi1 a capacity of approximately 1,514,000 liters (Ref. I1.6.17). He walls are of type 304 stainless steel and have an average thickness of.87 cm. (Ref.11.14.1).

) 3,

. APAoZZ-01003 Rev. 6

- The direct radiation dose from the RWST is calculated based on the tank's average isotopic content

- and the parameters discussed above, considering buildup and attenuation within the volume source.

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

" Reactor Shielding Design Manual" (Ref. I1.17). The computer program ISOSHLD (Ref. I 1.18, 11.19 and 11.20) will normally be utilized to perform this calculation.

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 determined to be 7E-2 mrads/ calendar year, based on a point source of primary coolant N-16 in the steam generators. This source term was then projected onto the inside surface of the containment dome, taking credit for shielding provided by the containment dome and for distance attenuation.

No credit was allowed for shielding by other structures or components within the Containment Building. The number of gammas per second was generated and then converted to a dose rate at the given distance by use of ANSI /ANS-6.6.1, " Calculation and Measurement of 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. He distance was determined 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 per year, assuming a maximum occupr.ncy 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. He area is also partially bounded on a third side by the Discharge monitoring tanks dike system. The radioactive wastes are stored in this area using high integrity containers (HIC) inside Onsite Storage Containers (OSC) and LSA type storage containers. The HIC has the s highest potential for containing significant amounts of radioactive material, and constitutes the only potentially significant source of direct radiation from on-site radioactive waste storage.

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

The HICs typically are right circular cylinders approximately 1.7 meters in diameter and 1.8 meters in height. He 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.,

dimensions, shielding material, etc.) for each OSC will be accounted for in the calculations.

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. He computer program MICROSHIELD (Ref.11.24) will normally be utilized to perform this calculation.

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. 5. RADIOLOGICAL ENVIRONMENTAL MONITORING i

5.1 DESCRIPTION

OF THE RADIOLOGICAL ENVIRONMEhTAL MON 1TORING PROGR AM The Radiological Environmental Monitoring Program is intend ed to act as a background data base i for preoperation and to supplement the radiological efiluent reh ase monitoring pragram during  ;

plant operation. Radiation exposure to the public from the variou; specific pathways and direct radiation can be adequately evaluated by this program.

Some deviations from the sampling frequency may be necessary due to seasonal unavailability, hazardous conditions, or other legitimate reasons. Efforts are made to obtain all required samples within the required time frame. Any deviation (s) in sampling frequency or location is documented l in the Annual Radiological Environmental Operating Report. I Sampling, reporting, and analytical requirements are given in Tables 9.11-A,9.11-B, and 9.11-C. l Airbome, waterbome, 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 Environmental Protection Agency's (EPA) Environmental Radioactivity Laboratory l 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 TESTING OF ENVIRONMENTAL THERMOLUMINESCENCE DOSIMETERS ,

Thermoluminescence Detectors (TLD's) used in the Environmental Monitoring Program are tested for accuracy and precision to demonstrate compliance with Regulatory Guide 4.13 (Ref. I1.16).

Energy dependence is tested at several energies between 30 kev and 3MeV corresponding to the approximate energies of the predominant Noble Gases (80,160,200 kev), Cs 137 (662 kev), Co-60 (1225 kev), and at least one energy less than 80 kev. Other testing is performed relative to i cither Cs-137 or Co-60. (Ref. I1.14.10) 1 1

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t APA-Z2-01003 5 Rev. 6 3 l

6. DETERMINATION OF ANNUAL AVERAGE AND SHORTTERM ATMOSPHERIC I ' DISPERSION PARAMETERS l 6.1 ATMOSPHERIC DISPFRSION PARAMETERS The values presented in Table 6.1 and Table 6.2 were detennined through the analysis of on-site meteorological data collected during the three year period of May 4,1973 to May 5,1975 and March 16,1978 to March 16,1979. '

6.1.1 Lonc-Term Dispersion Estimates j I

The variable trajectory plume segment atmospheric transpon model MESODIF II (NUREG/-CR-0523) and the straight-line Gaussian dispersion model XOQDOQ (NUREG/CR2919) were used for l determination of the long-term atmospheric dispersion para neters. A more detailed discussion of I the methodology and data utilized to calculate these parameters can be found elsewhere (Ref.  ;

11.6.12). l 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 I

which they are located, and the unit Vent is equipped with a rain cover which effectively eliminates the possibility of the exit velocity exceeding five times the horizontal wind speed. All gaseous I

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).

Determination of Lona-Term Dispersion Estimates for Special Receptor Locations 6.1.2 Calculations utilizing the PUFF model were performed for 22 standard distances to obtain the g desired dispersion parameters. Dispersion parameters at the SITE BOUNDARY and at special 3<

receptor locations were estimated by logarithmic interpolation according to (Ref. I1.6.13):

X = X, (dId,)* (6.1)

I Where:

B =in (X2 /X,)/In (d2/di )-

X1 , X2= Atmospheric dispersion parameters at distance d and i d ,2respectively, froro the source.

The distances d g and d2 were selected such that they satisfy the relationship. ,

d<d<d2 i

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 term dispersion estimates are determined by multiplying the appropriate long term dispersion estimate by a correction factor (Ref.

I1.9.1 and 11.15.1):

F = (T, / T,)* (6.2)

Where:

TS

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

Ta = Thr. total number of hours in the data collection period from which the long term .

diffusion estimate was determined (Refer to Section 6.1).

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

Short term dispersion estimates are not applicable to short term releases which are sufficiently random in both time of day and duration (e.g., the short term release periods are not dependent solely on atmospheric conditions or time of day) to be represented by the annual average dispersion conditions (Ref.11.8.1).

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- APA-ZZ-01003 Rev. 6

.~ 6.1.3.1 The Determination of the Slope Factor (S)

I ne general approach employed by subroutine PURGE of XOQDOQ (Ref. I1.15.1) was utilized to produce values of the s! ope 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) I 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)i) values for that release and at that location in descending order. The (X/Q): 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.

He intermittent release (X/Q) curve was constructed using the calculated 15 percentile (X/Q); and 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); 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 ) 3 or

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

S= (6.4) log T, - log T3 6.1.4 Atmospheric Dispersion Parameters for Farmine Areas within the SITE BOUNDARY l

The dispersion parameters for farming arcas within the SITE BOUNDARY are intended for a narrow scope application: nat of calculating the dose to the current fanner 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

1. 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 '.veighted 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 BOUNCARY and his estimated occupancy period.

6.2 ANNUAL METEOROLOGICAL DATA PROCESSING ne annual atmospheric dispersion parameters utilized in the calculation of doses for demonstration of compliance with the numerical dose objectives of 10 CFR 50, Appendix 1, are determined using computer codes and models consistent with XOQDOQ (Ref.11.15). These codes have been validated and verified by a qualified meteorologist prior to implementation. Multiple sensors are utilized 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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%./ V APA-ZZ-01003 Rev.6 TABLE 6.1 l IHGIIEST ANNUAL AVERAGE ATMOSPIIERIC DISPERSION PARAMETERS.

' UNIT VENT DISTANCE X/Q X/Q SECTOR (METERS) X,Q DECAYED / DECAYED / D!Q LOCATION (b)

UNDEPLETED DEPLETED fin thedss@#E3 Aid 5!sirissEnnississfinissENVihRhdENAnAM5AMOs&f)SnAPIsssWN40srsjMI5ihWM4niidsOnjMMMdiis#55dd:My-(sec/m (sec/m ) (sec/m ) (m )ssts&

2200 1.0E-6 9.9E-7 8.5E-7 4.3 E-9 l SITE DOUNDARY(a) NNW NNW 2864 6.8E-7 6.8E-7 5.7E-7 2.6E-9 Nearest Residence (c)(d)

SE 3830 2.5E-7 2.5E-7 2.lE-7 1.1 E-9 l Farmer's Residence (c)

N/A N/A 2IE-7 2.IE-7 f .9E-7 I .l E-9 Farming Areas within the Site Boundary (c)(c)

(a) Values given are from I SAR Table 23-82 l (b) Data from 1995 Land itse Census (c)- Values derived from FSAR Table 23-83. using the methodology presented in Equation (6.1) (Ref. Il.5.6)

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

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

Building Shape Parameter (C) = 0.5 (Ref. II.53)

Vertical IIcight ofIlighest Adjacent Dailding (V)- 66.45 meters (Ref. Il.53)

E M MM M M M M M M M M M M M e a __

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APA-ZZ-01003 Rev.6 TABLE 6.2 ,

l IIIGIIEST ANNUAL AVERAGE ATMOSPilERIC DISPERSION PARAMETERS RADWASTE BUILDING VENT DISTANCE X/Q X/Q LOCATION (b) SECTOR (METERS) XJQ DECAYED / DECAYED / D/Q

. UNDEPLETED DEPLETED isWS35tGAs&R@tA!W W Mts&&Ennd!MMisAN8sMEWh-- -

eM5Fe#Ei=Janr- EiE&Mi4M!f Fw P M*STK4A%E & MKA@&T # isfNn! ,

NNW' 2200 13E-6 I.3E-6 1. E-6 43E-9 l SITE DOUNDARY(a)

NNW 2864 B 7E-7 8.7E-7 7.2E-7 2.6E-9 Nearest Residence (c)(d)

SE 3830 3.0E-7 3.0E-7 2.4E-7 1.lE-9 l Farmer's Residence (c)

Farming Areas Within N/A N/A 2.9E-7 2.9E-7 2.6E-7 f.I E-9 Site Boundary (c)(e)

(a) Vahres given are from FSAR Table 23-84 ,

l (b) Data from 1995 Land Use Census .

(c) Values derived from FSAR Table 23-81, using the methodology presented in Equation (6.1) (Ref. I1.5.6) ,

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

(c) Dese values were derived for a narrow scope application. Extreme caution should be exercised when deterraining their suitability for use in other applications Building Shape Parameter (C) = 0.5 (Ref. I1.53)

Vertical IIcight ofIfighest Adjacent Building (V) = 19 % meters (Ref. I1.53) i

.._m _ _ . . _ ___ ___._.._____..m_ . _ _ _ _ _ _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _ _ . _ _ . . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _

APA ZZ-01003 i Rev.6 i-TABLE 6.3 SHORT TERM DISPERSION PARAMETERS (a)(c) l Slope Factor (s)

Location (b) Sector Distance Unit Vent Radwaste l Building Vent 1

Site Boundary S 1300 .328 .320

}

Nearest Residence (d) NNW 2865 .264 .268 8 1 I!

(a) Reference 11.5.3 l (b) Data from 1995 Land Use Census (c) Recirculation Factor = 1.0 (d) All pathways are assumed to exist at the location of the nearest resident.  ;

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APA-ZZ-01003 Rev.6 TABLE 6.4 APPLICATION OF ATMOSPHERIC DISPERSION PARAMETERS Dispersion Parameter Controlling Ane Group Rec Controllina Location l Dose Pathway x/Q, decayed /undepleted N/A 9.7 Site Boundary Noble Gas, Beta Air & Gamma Air (2.26 day halflife) x/Q, decayed /undepleted N/A 9.6 Site Boundary Noble Gas, Total Body & Skin (2.26 day halflife) x/Q, decayed / depleted Child 9.6 Nearest Resident l Inhalation (8 day halflife) 9.8 Site Boundary N/A 9.8 Nearest Resident l Ground Plane Deposition D/Q Child 9.8 Nearest, Resident l Ingestion pathways D/Q*

l

• For 11-3, x/Q, decayed / depleted is used instead of D/Q (Ref. I1.11.1).

I i

I

APA ZZ-01003 El Rev.6

) TABLE 6.5 METEOROLOGICAL DATA SELECTION HIEARCHY Parameter Primary First Second Third Alternate Alternate Alternate Wind Speed 10m Pri 10m Sec 60m Pri 90m Pri j Wind Direction 10m Pri 10m Sec 60m Pri 90m Pri Air Temperature 10m Pri 10m Sec Wind Variability 10m Pri 10m Sec 60m Pri 90m Pri Temp Different 60-10m Pri 90-10m Pri 90-60 Pri Dew Point 10m Pri Precipitation im Pri (a) Priindicates primary tower I

(b) Sec indicates secondary tower J '

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APA-ZZ- 1 V Rev.6

< Table 6.6 '

s

' ApplickNon or Atateschefid iiispessidnharametersi Annuni Nffinent Relesse ReporO 4 Controlling # >' Dis 3 Controlling

• Location, ;;,

> 'NEM

- ~ ppy&fisgNgh hDose Pathway 4DP*fg gp @$ gDis M $p$ ertlos

$ PParan6eterg EM$jgy@ $$difspf g y? '; ' Values

$,persion s 's; wc c:-

5 C ' ' NJ , - (

Noble Gas, x/Q, decayed /undepleted N/A Concurrent Site Boundary Beta Air & Gamma Air Dose (2.26 day halflife) Nearest Resident Noble Gas, x/Q, decayed /undepleted N/A Concurrent Site Boundary Total Body & Skin Dose (2.26 day halflife) Nearest Resident Concurrent Farmer's Residence Ilistorical Inside Site Boundary Ground Plane Deposition Dose D/Q N/A Concurrent Site Boundary Nearest Resident Concurrent Farmer's Residence IIistorical Inside Site Boundary inhalation Dose x/Q, decayed / depleted Child Concurrent Site Boundary (8 day halllife) Nearest Resident Adult Concurrent Farmer's Residence Ilistorical Inside Site Boundary Ingestion Dose Pathways D/Q* Child Concurrent Site Boundary Nearest Resident Adult Concurrent Farmer's Residence 11istorical Inside Site Boundary

• For 11-3, x/Q, dccayed/ depleted is used instead of D/Q (Ref. I 1.11.1).

t

I APA-ZZ-01003 l Rev. 6

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

Routine Annual Radiological Environmental Operating Report covering the operation of the nnit during the previous calendar year shall be submitted prior to May 1 of each year.

The Annual Radiological Environmental Operating Report shall include summaries, interpretations, j and an analysis of trends of the results of the radiological environmental surveillance activities for '

the report period, including a comparison with preoperational studies, with operational controls and with previous environmental surveillance reports, and an assessment of the observed impacts of the 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 3I to REC 9.12.1. gI De Annual Radiological Environmental Operating Report shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during gl the period pursuant to the ODCM, as well as summarized tabulated results of these analyses and gI measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979, in the event that some individual results are not available for ,

inclusion with the report, the repon shall be submitted noting and explaining the reasons for the l missing results. The missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the following: a summary description of the radiological 3 environmental monitoring program; at least two legible maps' covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of g,

licensee 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 1 conducting the Radiological Environmental Mcnitoring Program as required by 9.11.1 and l discussion of all deviations from the sampling schedule of Table 9.11-A, discussion of l environmental sample measurements that exceed the reporting levels of Table 9.11-B, but are not  ;

the result of the plant effluents, 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 Effluent Release Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.

The Annual Radioactive Efiluent Release Report shall include a summary of the quantities of radioactive liquid and gaseous effluents released from the unit as outliried in Regulatory Guide 1.21,

{

" Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases Gaseous i Effluents from Light-Water-Cooled Nuclear Power Plants, " Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof. ~

The Annual Radioactive Effluent Release Report shall include an annual summary of hourly a, meteorological data collected over the previous calendar year. This annual summary may be either gj in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric ~i stability, and precipitation (i.Pmeasured), or in *the form ofjoint frequency distribution of wind ,

speed, wind direction, and atmospheric stability .

Il

• One map shall cover s:ations near the SITE BOUNDARY; a second shall include the more distant stations.

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• In lieu of submission with the Annual Radioactive Effluent Release Report, Union Electric has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

APA-ZZ-01003 Rev. 6  ;

1 i

l This repon shall also include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents 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 emuents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Technical Specifications, Figures 5.1-3 and 5.1-4) during tiie repon period using historical average atmospheric conditions. All assumptions used in making these assessments, i.e., specific activity, exposure time and location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and rnessurement, shall be used for determining the gaseous pathway doses.

Assessment of radiation doses shall be performed in accordance with the methodology and

)

[ parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

The Annual Radioactive Effluent Release Report shall include an assessment of radiation doses to the most likely exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation." Doses to the MEMBER OF THE PUBLIC shall be calculated using the methodology and parameters of the ODCM.

The Annual Radioactive Effluent Release Repons shallinclude a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reponing period.

The Annual Radioactive Effluent Release Repons shall include a summary description of any major 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.

3 Reponing requirements for changes to Solid Waste Treatment Systems is addressed in J APA-ZZ-01011, PROCESS CONTROL PROGRAM (PCP).

'Ihe Annual Radioactive Effluent Release Reports shall also include the following information: An explanation as to why the inoperability ofliquid or gaseous effluent monitoring instrumentation was not corrected within the time specified, and a description of the events leading to the liquid holdup tanks or gas storage tanks exceeding the limits of FSAR Section 16.11.1.1 OR 16.11.3.1.

The Annual Radioactive Effluent Release Repons 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 CONTROL PROGRAM (PCP).

r

, APA-ZZ-01003 Rev. 6

8. IMPLEMENTATION OF ODCM METHODOLOGY (CTSN 2791)

The ODCM provides the mathematical relationships used to implement the Radioactive Effluent Controls. For routine effluent release and dose assessment, computer codes are utilized to 3 implement the ODCM methodologies. These codes are evalused in accordance with the g requirements of plant operating procedures to ensure that the;, , roduce 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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. APA-ZZ-01003 i Rev. 6 ,

9. _RADIOACTTVE EFFLUENT CONTROLS REC) ,

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i m. The terms in this section that appear in capitalized type are defined in Technical  ;

i t Specifications.

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

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APA-ZZ-01003 g Rev. 6 5l

- 9.1 RAD 10 ACTIVE LIOUID EFFLUENT MONITORING TMSTRUMENTATION REC 9.1 has been relocated to Section 1633.6 of the FSAR.

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

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Monitor REC Number FSAR Section Type i HB-RE-18 9.1.2.1-1.a 163.3.6.1-1.a Liquid rad monitor BM RE 52 9.1.2.1-1.b 1633.6.1-1.b Liquid rad monitor HB FE 2017 9.1.2.1-2.a 1633.6.1-2.a Flow element BM-FE-0054 9.1.2.1-2.b 1633.6.1 2.b Flow element FE DB-1006,1101 9.1.2.1-2.c 1633.6.1-2.c Flow element j

9.2 RADIOACTIVE G ASEOUS EFFLUENT MONITORING INSTRUMENTATION REC 9.2 has been relocated to Section 1633.7B of the FSAR.

The following should be used to cross-rderence REC 9.2 surveillances to the appropriate section of l the FSAR: l Monitor REC Number FS AR Section Type Unit Vent j GT-RE 21B 9.2.2.1-1.a 1633.7b.1-1.a Gas j 3 GT-RE 21 A & B 9.2.2.1-1.c 1633.7b.1-1.c lodine sampler  !

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GT-RE-21 A & B 9.2.2.1-1.b 1633.7b.1 1.b Particulate sampler GT-RE-21 A & B 9.2.2.1-1.d 1633.7b.1-1.d Unit Vent flow rate GT-RE-21 A & B 9.2.2.1-1.e 1633.7b.1-1.e Particulate and Radioiodine l T

Sampler flow rate Monitor Containment Purge GT RE-22 9.2.2.1-2.a 1633.7b.1 2.a Gas GT RE-33 GT-RE-22 9.2.2.1-2.c 1633.7b.1-2.c lodine sampler GT RE 33 GT-RE-22 9.2.2.1-2.b 16.3.3.7b.1-2.b Particulate sampler GT RE-33 GT RE-22 9.2.2.1 -2.d 1633.7b.1-2.d Containment purge flow rate GT-RE 33 l GT-RE-22 9.2.2.1-2.c 1633.7b.1-2.e Particulate and Radioactive GT-RE-33 Sampler flow rate Monitor Radwaste Building Ventilation GH-RE-10B 9.2.2.1-3.a 1633.7b.1-3.a Gas g GH-RE-10A & B 9.2.2.1-3.c 1633.7b.1-3.c lodine sampler g GH-RE-10A & B 9.2.2.1-3.b 1633.7b.1-3.b Particulate sampler GH RE-10A & B 9.2.2.1-3.d 1633.7b.1 3.d Radwaste Building Vent Flow rate GH RE-10A & B 9.2.2.1-3.c 1633.7b.1-3.e Particulate and Radioactive

) Sampler flow rate Monitor l

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

- - APA-ZZ-01003 Rev. 6

.- y 9.3 LlOUID EFFLUENTS CONCENTRATION ,

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/ 9.3.1 Controls (CTSN 41834) l 9.3.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Technical Specifications, Figure 5.1 4) shall be limited to the concentration specified in 10 )

CFR Part 20.1 20.601, Appendix B, Table 11, Column 2, for radionuclides other than dissolved or l entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 104 microcurie /ml total activity.)

' APPLICABILITY: At alltimes. i ACTION:

2. With the concentration of radioactive material released in liquid tffluents to UNRESTRICTED AREAS exceeding the above limits, immediately restore the 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, l

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APA ZZ-01003 Rev.6 TABLE 9.3-A

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RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM I

eMontforgrakshn qaggg h wj ^Qidsi kN 9gy[M7dflez df#Mg N$@(B.2tc

?$v[gil?Discha rP GeltesehdQMW;ih.ggr gsg(jng%

ggy SAMPLING MINIMUM TYPE OF 12 D (1)

FREQUENCY (7) ANALYSIS ACTDTFY (pCl/ml)

FREQUENCY ANALYSIS Pnor to Pnor to Pnncipal Gamma Emmers (3) SE 7 Each Batch Each Batch 1131 1E4 Dissolved and Entrained Gases (Gamma 1E 5 Emitters)

H-3 IE-5 Monthly Gross Alpha IE-7 Composite (4)

Quanerly Sr-89, St-90 SE4 Composite (4) Fe-55 IE4 Np 237 5E-9 Pu-238 SE-9 Pu-239/240 SE 9 Pu-241 SE-7 Am-24i SE-9 Cm-242 SE-9 Cm-243/244 $E-9

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>l^42.' Steam Generator Blowdown J'~I: :

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SAMPLING MINIMUM TYPE OF LLD (1) J FREQUENCY (7) ANALYSIS ACT. nTn' (pC1/ml) ,

FREQUENCY ANALYSIS ,

Daily Pnncipal Gamma Emitters (3) SE-7 I Daily Grab Sample (6) 1-131 IE4 i Dissolved and Entrained Gases (Gamma 1E-5 I Emitters)

H-3 IE 5 I Monthly Gross Alpha 1E-7 j Composite (4)

Quanerly Sr49. Sr-90 SE4 um Composite (4) Fe-55 1E4 I

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- TABLE 9.3-A (Cont'd)

TABLE NOTATlONS

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(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) 'Ile principal gamma emitters for which the LLD control applies include the following radionuclides: Mn-54, Fe 59, Co-58,2n-65, Mo-99, Cs-134, Cs-137, Cc-141, and Ce-144. This list does not mean that only these nuclides are to be cons *dered. 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 whici. the quantity of liquid sampled is proportional to the quantity ofliquid waste discharged and n, 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 effluent 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 /> thereafter while the release is occurring. To be representative of the liquid emuent, the

) sample volume shall be proportioned to the effluent stream discharge volume. The ratio "f of sample volume to emuent discharge vohune shall be maintained constant for all samples taken for the composite sample.

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

' APA-ZZ-01003 Ra. 6 gg l

9.4 DOSE FROM L10UID EFFLUENTS 9.4.1 Controls (CTSN 41834) l 9.4.1.1 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in j liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Technical B l Specifications, Figure 5.1-4) shall be limited:

l a. During any calendar quarter to less than or equal to 1.5 mrems to the whole body and to h y

less than or equal to 5 mrems to pay organ, and l b. During any calendar year to bss than or equal to 3 mrems to the whole body and to less than or equal to 10 mrems '.o any organ.

APPLICABILITY: At all times.

ACTION: (CTSN 1161)

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,  ;

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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 l, ,

results of radiological impact on finished drinking water supp* lies with regard to the Bl requirements of 40 CFR Part 141, Clean Drinking Water Act. {

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable. ,{ l 9.4.2 Surveillance Reauirements 5i 9.4.2.1 Cumulative dose contributions from liquid efnuents for the current calendar quarter and the current g i

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') 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 requirements of ACTION a.(1) and (2) are applicable only if drinking water supply is taken from the receiving water body within 3 miles of the plant discharge. In the case of river-sited plants this is 3 miles downstream only.

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. APA-ZZ-01003 Rev. 6 9.5 LlOUID RADWASTE TREATMENT SYSTEM

$' 9.5.1 Controls (CTSN41834) 9.5.1.1 The Liquid Radwaste Treatment System shall be OPERABLE and appropriate portions of th'e system shall be used to reduce releases of radioactivity when the projected does due to the liquid effluent, from each unit, to UNRESTRICTED AREAS (see Technical Spcifications, Figure 5.1-4) would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ m a 31 day period.

APPLICABILITY: At alltimes, f ACTION: (CTSN 1161)

a. With radioactive liquid waste being discharged in excess of the above limits and the Liquid Radwaste Treatment Systems are not being fully utilized, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report g

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 i inoperability.

2) Action (s) taken to restore the inoperable equipment to OPERABLE status, and

3) Summary description of action (s) taken to prevent a recurrence.

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

9.5.2 Surveillance Reauirements 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,

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

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APA-ZZ-01003 g)

Rev. 6 5i 9.6 GASEOUS EFFLUENT!, 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 l 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 i 4

b. For lodine-131 and 133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ. g{l 3

APPLICABILITY: At alltimes. )

ACTION:

a. With the dose rate (s) exceeding the above hmits, immediately restore the release rate to i within the above limit (s).

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b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable. 9j 9.62 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.

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 Il.

and performing analyses in accordance with the sampling and analysis program specified in Table i 9.6 A.

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RADIOACTIVE GASEOUS EFFLUENTS SAMPLING AND ANALYSIS PROGRAM m, me ,mmaF > > xMMWaste Gas Decay Tanka +

2 e:m o e , *'s +

SAMPi ING FREQUENCY MINIMUM ANALYSIS TYPE OF ACTIVITY ANALYSIS LLD (1)

(9) FREQUENCY (pCi/mi)

--Pnor to each re6 ease- grab Pnor to each tank Pnncipal Gamma Emitters- particulate, iodine, noble gas (2) 1 E-4 sample l Continuous See footnote 8 r, ,> w 3 e g:; magm 5igrc:S "

!?2tContainment Purge or Vents '

se ,

^

4

, as SAMPLING FREQUENCY MINIMUM ANALYSIS TYPE OF ACTIVITY ANALYSIS LLD (1)

(9) FREQUENCY (pcilml)

Pnor to each release- grab Pnor to each release Pnncipal Gamma Emitters- particulate, iodine, noble gas (2) 1 E-4 sample H-3(oxide) 1E-6 Con'inuous See footnote 8

. sh - -

wem e < en ' ^

<in3.iUnit Vent-(3)n 65 <

+ s - ,

Nwina 4 SAMPLING FREQUENCY MINIMUM ANALYSIS TYPE OF ACTIVITY ANALYSIS LLD (1)

(9) FREQUENCY ( Ci/ml)

Monthly- grab sample (3)(4) Monthly (3) (4) Pnneipal Gamma Emitters- particulate, socine, noble gas (2) 1 E-4 H-3(oride) 1E-6 Continuous (6) Weekly (7) 1-131 1E-12 1-133 1E 10 Pnneipal Gamma Emitters- particulate nuclides only (2) 1E 11 Monthly Composite Gross Alpha 1E 11 Quarterly Composite Sr-89, Sr-90 1E 11 w.e.e + sm +:s sw swa > 3e WRadwaste Building Venth ese o 99 - -

SAMPLING FREQUENCY WilNIMUM ANALYSIS TYPE OF ACTIVITY ANALYSIS LLD (1)

(9) FREQUENCY (pCilml)

Monthly- grab sample Monthly Pnncipal Gamma Emitters- particulate, sodine, noble gas (2) 1E 4 Continuous (6) Weekly (7) 6-131 1E-12 1-133 1E-10 Principal Gamma Emitters- particulate nuclides only (2) 1E-11 Monthly Composite Gross Alpha 1E 11 l Quarterly Composite Sr -89, Sr-90 1E 11 I

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APA-ZZ-01003 Rev. 6  ;

TABLE 9.6-A (Cont'd) y TABLE NOTATIONS <

(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, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65,1-131, Cs-134, Cs-137, Cc-141, and Ce 144 in iodine and particulate releases. This list does not mean that only these nuclides are to be considered. Any nuclide which is identified in the sample and which is also listed in the ODCM gaseous effluents dose factor tables, shall be aalyzed and reported in the Annual Effluent Release Report.

(3) If the Unit Vent noble gas monitor (GT-RE-21B) shows that the effluent activity has increased (relative to the pre-transient activity) by more than a factor of 3 following a reactor shutdown, startup, er 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 lodines. 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 of 7 days or until the Unit Vent noble gas monitor no longer indicates a g factor of 3 increase in Unit Vent noble gas activity, whichever t.omes 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 refueling canal is flooded.

g (5) Deleted *

(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, 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 accordaace w:th footnote 3 (above), then the LLD may be increased by a factor of 10. -

(8) Continuous sampling of this batch release pathway is included in the continuous sampling g performed for the corresponding continuous release pathway. 'g (9) Samples shall be representative of the effluent release.

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APA-ZZ-01003 Rev. 6 '

9.7 DOSE - NOBLE GASES 9.7.1 Controis (CTSN 41834) 9.7.1.1 The air dose due to noble gases released in gaseous effluents, from each unit, to areas at and beyond the SITE BOUNDARY (see Technical Specifications Figure 5.1 3) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation, and

b. During any calendar year: Less than or equal to 10 mrads for gamma radiation and less l than or equal to 20 mrads for beta radiation.

APPLICABILITY: At all times.

ACTION: (CTSN 1161)

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

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

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

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APA-ZZ-01003 l

Rev. 6 l

s 9.8 DOSE -IODINE-131 AND 133. TRITIUM. AND RADIOACTIVE M ATERI AL IN  !

) PARTICULATE FORM l 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, to areas at and beyond the SITE BOUNDARY (see Technical Specifications, Figure 5.1-3) shall be lj limited to the following: 51 1

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

b. During any calendar year: Less than or equal to 15 mrems to any organ.

1 APPLICABILITY: At all tur es.  !

ACTION: (CTSN 1161)

a. With the calculated dose froa the release oflodine-131 and 133, tritium, and radionuclides in particulate form with half.$ives greater than 8 days, in gaseous effluents exceeding any 3:

of the above limits. ,trepare and submit to the Commission within 30 days, pursuant to gi Technical Specificaton 6.9.'., a Special Report that identifies the cause(s) for exceeding i the limits and defines ti e corrective actions that have been taken to reduce the releases and the proposed corrective avions to be taken to assure that subsequent releases will be in compliance with the above Smits.

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

9.8.2 Surveillance Reouirements )

9.8.2.1 Cumulative dose contributions for the current calendar quarter and current calendar year for Iodine-131 and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 days 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-01003 Rev. 6 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 each unit, to areas at and beyond the SITE BOUNDARY (see Figure Technical Specification's 5.1 3)would exceed:

a. 0.2 mrad to air from gamma radiation, or

b. 0.4 mrad to air from beta radiation, or

c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.

APPLICABILITV: At alltimes ACTION:

a. With radioactive gaseous waste being discharged in excess of the above limits, and the Gaseous Radwaste Treatment Systems are not being fully utilized, prepare and submit to the Commission within 30 days, pursuant to Technical Specifications 6.9.2, a Special Report that includes the following information:

1) Identification of any inoperable equipment or subsystems, and the reason for the inoperability,

2) Action (s) taken to restore the inoperable equipment to OPERABLE status, and

3) Summary description of action (s) taken to prevent a recurrence.

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

9.9.2 Surveillance Reouirements 9.9.2.1 Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM.

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 l 9.8.1.1.

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APA-ZZ-01003 .

Rev. 6 s 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 limited 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 gaseous effluents exceeding twice the limits of REC 9.4.1.1b,9.7.1.1a,9.7.1.1b,9.8.1.1a, or 9.8.1.1b, 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.I'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 g action to be taken to reduce subsequent release to prevent recurrence of exceeding the 3 above limits and includes the schedule for achieving conformance with the above limits.

This Special Report, as defined in 10 CFR 20.2203, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose (s) exceeds the above limits, and if the release l

e condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

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

9.10.2 Surveillance Reouirements 9.10.2.1 Cumulative dose contributions from gaseous effluents shall be determined in accordance with REC g' 9.7.2.1, and 9.8.2.1, and in accordance with the methodology and parameters in the ODCM. g 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 Rev. 6 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 9.11-A.

APPLICABILITY: At all times.

ACTION:

a. With the Radiological Environmental Monitoring Program not being conducted as specified in Table 9.ll A, prepare and submit to the Commission,in the Annual Radiological Environmental Operating Report required by Technical Specification 6.9.1.6, a description 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 effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 9.11-B when 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 identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive effluents 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 report shall be submitted if:

concentration (1) , concentration (2) + .. 21.0 reporting level (1) reporting (2)

T When radionuclides other than those in Table 9.11-B are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose

• to A MEMBER OF THE PUBLIC from all radionuclides is equal to or greater than the calendar year limits of REC 9.4.1.1,9.7.1.1 or 9.8.1.1. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Repon, required by Technical Specification 6.9.1.6.

c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 9.11-A, identify specific locations for obtaining replacement samples and add them within 30 days to the Radiological Environmental Monitoring Program **. The specific locations from which samples were unavailable may then be deleted from the monitoring program, in the next Annual Radiological Environmental Operating Report include the revised figure (s) and tables reflecting the new sample location (s) with supporting information identifying the cause of the unavailability of samples and justifying the selection of new location (s) for obtaining samples.

d. When LLDs specified in Table 9.ll-C are unachievat,te due to uncontrollable circumstances,(such as background fluctuations, unavailable small sample sizes, the presence ofinterfering nuclides, etc.) the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report.

e. The provisions of Technical Specifications 3.0.3 an.13.0.4 are not applicable.

• The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.

" Excluding short term or temporary unavailability.

~

APA-ZZ-01003 d

Rev. 6 9.11.2 Surveillance Reauirements

! 9.11.2.1 The radiological environmental monitoring samples shall be collected pursuant to Table 9.11 A and

' i shall be analyzed pursuant to the requirements of Table 9.ll-A and the detection capabilities I required by Table 9.11-C.

i i

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APA-ZZ-0100?

Rev.6 I TABLE 9.11-A RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NUMBER OF REPRESENTATIVE SAMPLES SAMPLING AND EXPOSURE PATIIWAY AND SAMPLE COLLECTION TYPE AND FREQUENCY ANDOR SAMPLE LOCATIONS (l) FREOUENCY OF ANALYSIS

1. Direct Radiationm Forty routine monitoring stations with two or more Quarterly Gamma dose quarterly dosimeters or with one instrument for measuring and recording dose rate continuously, placed as follows:

An inner ring cf sixteen stations, one in each meteorological sector in the general area of the SITE BOUNDARY; l

An outer ring of stations, 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 populstion centers, nearby residences, schools, md in one er two areas to serve as control stations.

2. Airborne Radioicdine and Sarnples from five locations;

~ Continuous sampler operation with Rsdioiodine Canister- l-131 anafysis Particulates sample conection weekly, or more weekly.

Three samples from close to the SITE frequeritty if required by dust loading.

BOUNDARY locations, in different sectors, with Particulate Sampler- Gross beta high calculated annual average ground level D/Qs.

radioactivity analysis foNowing filter change: Mand gamma isotopic One sample from the vicinity of a community analysis " of composite (by location) located near the plant with a high calculated quarterly, annual average ground level D/Q.

One sample from a locartion in the vicinity of Futton, MO. (Ref SOS 95-2280)

. _ _ _ _ =

APA-ZZ-01003 Rev.6 TABLE 9.11-A (Cont'd)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NUMBER OF REPRESENTATTVE SAMPLES SAMPLING AND EXPOSURE PATIIWAY AND SAMPLE COLLECTION TYPE AND FREQUENCY ANDW SAMPLE LOCATIONS (I) FREOUENCY OF ANALYSIS

3. Waterbome

a. Surface (6) One sampic upstream Composite sample over Gamma isotopic (5) and tritium One sampie downstream I. month period (7). analysis monthly

b. Drinking One sample of each of one to three of the nearest water I.13I analysis on each composite when the surplies within 11 miles dommstream that could be affected by Composite 2-week period (samp)le when 1131over analysis is dose calculated for the consumption of the its discharge. performed, monthly composite otherwise. water is greater than I mrem per year (8).

Composite for gross beta and gamma isotepic One sample from a cetrollocation. analyses (5) monthly. Composite for tritium analysis quarterly.

As there are no drinking water intakes within 10 miles downstream cf the discharge point, the drinking water pathway is currently not included as part of the Callaway Plant Radiological Environmental Monitoring Program. Should future water intakes be constructed within 10 river miles downstream of the discharge point,the program will be revised to include this pathway (Ref. I1.6.6).

c. Sediment from One sample from downstream area with existing or potential Semiannually Gamma isotopic analy's is(5) semiannually shoreline recreational value

4. Ingestion a Milk Samples Imm milking animals in three different Semimonthly when animals are on pasture, Gamma isotopic (5) and I.131 analysis meteorological sectors within 5 km (3 mile) distance having monthly at other times semimonthly when animals are on pasture the highest dose potential. If there are none, then one sample monthly at other times from milking anima!s in each of three different rneteorological sectors between 5 to 8 km (3 to 5 mile) distance where doses are calculated to be greater than I mrem per yr.

One sample from milking animals at a controllocation,15 to 30 km (10 to 20 mile) distance and in one of the least Frevaient wind directions. .

Due to the lack of milking animals which satisfy these requirements, the milk pathway is currently 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 wi!! be revised to include this pathway.

m m M M M

M M M M

g

,v _

APA-ZZ-01003 Rev.6  ;

i TABLE 9.11-A (Continued)

TABLE NOTATIONS i NUMDER OF REPRESENTATIVE SAMPLES SAMPLING AND EXPOSURE PATilWAY AND SAMPLE COLLECTION TYPE AND FREQUENCY AND/OR SAMPLE LOCATIONS (II FREOUENCY OF ANALYSIS

4. Ingestion (Cont'd)

b. Fish One sample of each commercially and recreationally important Sample in season, or semiannually if they are Gamma isotopic analysis (5) on edible species in vicinity of plant discharge area. not seasonal portions One sample of same species in areas not influenced by plant discharge,

c. Food Products One sample ofeach principal class of food products from any At time of harvest (9)(10) Gamma isotopic analysis (5) on edible portion area that is irrigated by water in which liquid plant wastes have been discharged.

l As there are no areas irrigated by water in which liquid plant wastes have been discharged within 50 miles downstream of the discharge point, this sampic type is not currently included as part.

of the Callaway Plant Radiological Environmental Monitoring Program. Should future irrigation water intakes be constructed within 10 river miles downstream of the discharge point,the l

g progra m will be revised to include this sample type (Ref. I1.7.4 and I1.7.5).

Samples of three different kinds of broad leaf vegetstion if Monthly when available Gamma isotopic (5) and I-13I analysis available grown nearest each of two different offsite locations i of highest predicted annual average ground Icvel D/Q if milk

. sampling is not pesformed One sample of each of the similar broad leaf vegetation grown . Monthly when available ,

Gamma isotopic (5) and I-131 analysis 15 to 30 km (10 to 20 mile) distant in one of the least i prevalent wind directions if milk sampling is not performed

_ _ _ _ _ _ - _ _ _ _ _ _ _ - _ . _ _ _ - _ _ _ _ - _ _ _ - - _ .--_____-________-_-__-___--__-____-____-____--______-_L

~ APA-ZZ-01003 g Rev. 6 m' 1

m. TABLE 9.11-A (Continued)

TABLE NOTATIONS (1) Specific parameters of distance and direction sector from the centerline of one unit, and additional description 0

where pertinent, shall be provided for each and every sample location in Table 9.ll A in a table and figure (s) g l in the appropriate plant procedure. Deviations are permined from the required sampling schedule if gI specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment, and other legitimate reasons. If specimens are unobtainable due to sampling equipment ,

malfunction, every effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be Leumented in the Annual Radiological i Environmental 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 continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable specific alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the Radiological Environmental Monitoring Program. Submit in the next Annual Radiological 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 andjustifying the selection of the new location (s) for obtaining samples.

The selection of sample locations should consider accessibility of the sample site, availability of power, "

wind direction frequency, sector population, equipment security, and the presence of potentially adverse environmental conditions (such as unusually dusty conditions, etc.).

(2) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet l B

are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct 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 g reduced accordingly. The frequency of analysis or readout for TLD systems will depend upon the 3 characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.

g (3) (Deleted) 3 (4) Airborne particulate sample filters shall be analyzed for gross beta radioactivi,y 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 thoron daughter decay. If gross beta activity in air particulate samples is l E

greater than an established baseline activity level, gamma isotopic analysis shall be performed on the individualsamples. (CTSN43303)

(5) Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.

(6) The " upstream sample" shall be taken at a distance beyond significant influence of the discharge. The a

" downstream" sample shall be taken in an area near the downstream edge of the mixing zone, g (7) In this program, composite sample aliquots shall be collected at time intervals that are very short (e.g.,

hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample.

(8) Groundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination.

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

=

the ODCM.

(10) If harvest occars rnac than once a year, sampling shall be performed during each discrete harvest. If harvest occurs continuously, sampling shall be monthly. Attention shall be paid to including samples of tuberous and root food products.

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APA-ZZ-01003 Rev 6 TABLE 9.1 I-B REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES l

REPORTING LEVELS ANALYSIS WATER - AIRDORNE PARTICULATE FISil MILK FOOD PRODUCTS (pCill)a ORGASES(pCi/m3 ) (pCi&g, wet)b g (pCi&g,wei)b H-3 20,000 y , - .,, y.

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,e-em '.,9 - ..,s ' 5 ,,~1,000 , 30,000 ,, . .,.. . .- _ Co-5 8. ,r ~ m- y,- .. m----,c.,,~.,--,<~..q,,m,ve,----s---.w <' ' 4. O' s'b > 0 i " A . - ss < ,' 0 ' , e .s s'  %' ,. > 0 s:. ', x' P se ' , ' ',.s. / > ' n 'N ' N s< 3 s 's' I ' o' * ,' V ~~,.<% <O ' !' 300 10,000 - s , .Co ,6,,0, p - .. -~;-,w my .ww gs , ~- ,u~~ <._ + wy,-' y-+s ~*~r s e,e 's" ~- ,~ ,,w , .S ~ r , .~, ,,.2,._,,,r,.~,, , , -, .~. v, - , .q , s us s+ i < n,' a- ,' ,. h < > , < .c %.m 'd ~ ~ , ' W. w . s th c ' e~ M r > .. s 8 . w '~.6 ' s s . s ' ' ,Zr.Nb,,-95 400 > , _ s . , - , , -,-.,,.---.,m,rs,, ) , , .. ,,,w_ss.,,.~,, y ,,.,- .~~, - - .,..-- ~~ -s.~~.,~,,.~~.s s s3' ,, .s, x, s s v *^- , sg' < s ^'s , 4..- *s , ' 'Y#^l4 * *P ', .- JP 4 4 s X ,\ /* ',# 2v, s s '. . O > ,. s '* << s >0 - ' ,s 'e v > - ^ ^ 'A I 0 9.-,,,,4 '3 ' ' 100 r k+

s. s _ ,.

os,_1_31p.--~,1-.~,,,.~.,,,2>.<-.-.s , ,} s ' > .s  %.. s <v v v2. . ~ ~ , , . . .. ' d i', u s ,. N 'n < 4 ' 'v.*1<<w,+ w$ y' ,' m-., <'s m yx,- , - .~ r - > 'si* , - - n s- ,.~ e. - ~~s , ,~, - v>- s,'4 ' J # . >/. >' s

• s' \^'$ 't+

- x s 1,000 . Cs-134 30 10 I,000 '~.~. -> -- s. m r.~ -s 60r s. e - y- -,_, - -r,, .-~.,. , - - m , - , ,-.,,,,s .-, - ~e:rs'* t ~ '1 s,- < .,__s , m ,.s, , , ~ , - . .,,,,,,, ,,~.,m s *3 'e , ' , sf>  ; s s, , y ,. f. < s;&> Q q :;;.o e ', u '  %.-' ,' : .' e ,, s " > w , ', s,,' ' s-,,, ., C.~s,-13 4 7 ys.>,-,,,.~4%, 50 .. ,.--, .. . 2.,,.20- , w m , y,y -<. ,7<w.~n 2,0 00~,3 e,- ~~<~ - ,c 70 y.,-- . ,4-v -. --.0p 2,00 ^ww q '! s.' > 's 's< s,s><., >' s* (' % s> ' 0 ^ J L.7 's 4 0 ,xs~ s <. s . 's x< v . i, '*\ ' a 200 300 Da-La-140 (a) Multiply the values in this table by IE-9 to convert to units of pCi/mL (b) Multiply the values in this table by IE-9 to convert to units of pCi/g. For drinking water samples. This is 40 CFR Part 141 value. For surface water samples, a value of 30,000 pCi/t may be used. Total activity, parent plus daughter activity. r i %# V _- APA-ZZ-01003 Rev.6 TABLE 9.11-C DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS LOWER LIMIT OF DETECTION (LLD)(I),(2),(3') ANALYSIS SURFACE DRINKING AIRDORNE FISII htILK TOOD I RODUCTS SEDthfENT WATER WATER PARTICULATE (pCi/kg, wet)b (pCi/kpt)b (pCWgJry)b (pCill)s (pCi/[) OR GASES (pCi/m3 ) Gross liets 4 4 0.01 - ,n. m .,y.,,;gewr _

::er.,,,.g y.n, m .,m31:., g g swu g ,v.n,,,.r.,myngen.c.m.

vmwn m;;i /. w re:ws m"'.m w,.nsgyny r m.wg.n my.rym.mnm nyra.ng:.mm,w,v g ,g m y.m.m " "m > , a .::+ 1 - II-3 3000,, 2000 . , :n ~ wr.,,y.y pgr

y. r.c

...:.. g g :wgygy->y.syp~e.ypsysnspre.g.mym3 em...,3.wn.cc e s.-wve <:mm p m e,me.-7 7m.m

.gu.us. g-: 4 4 e . . pp . ws. ~ - v<- >>9 y 4 ~ ..> s x m1a m mwyg.g  :- a hin-54 15 130 v .m v.~ e., ,.,,.,, m.;r. w- 15.g g,y

w%y ,,,rm ,wgm mgmmmm

e.,m .wmm.w ,nnr,y.: . - " - " m.w;wp.mamw7.+;g,rwv,.ym-n.,p.ww.gwmmmm:+y a g ev.mm

m s n .s . . g:s .;M.v; g.A;:: ~ m%m4h e.) - - :~ --1,... :su ,m - " a re-59 30 30 260

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Np.$ x4/S: M^ '> yy,mmm 592'b prespsg:pj9Ms.33

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4 ^ <v+s-- 'Y.bm N p:.- , ,e~ m .-> m th / sg - Zr-Nb ,95 I5 m g ,w 5 yu .,15,m.s, s- e, .s .pr . . wem .% ;v gv:m.g .es +ms y& ,.n y,g+ ,v.,m m ey g y,y/ .,.er.~w,~.m y.r y ,m~::y w y m. n mggy s.- y g y w.w.g.,.g m ,p p.,e .gnyg,n;qs < >.: < ::+ ., , . . . ~ . > I 131 1000 1 . 0.07, 1 60 y 3,.,- -w., . ,m+r,.sgyy., y.m.y y,m..7,.g s . . . . . . . .  : .<.e .> , . . M:i1 + g.gg i:i.% g/;;F:fM/$v . ,.,ww.p.y.,; ,xvs py.vyp  :.?# %g,.w7y,.:s >  :.v .r- e . ., ..,.y sw s3 w .- w .emm,pgr , q.myme;y.m.~.,..y.g 4 x s' '+O~ . A . . . z Cs-134 15 15 0.05 130 15 60 150 % :w w,m: r - W v.ppye y- w . . . . . .a. . .<m;y+~re:!Sp?. ' . . @X;r,f.? -y?-?wX*'4P^y%gwy?f;@wf-y .. p'?'M:4.7&e. ?5"?m: .* Nc:w:f ^ -@-^vMFr -:^77..:-^wff s .- s pTNA +.:.y * :r:5R .?w+:MN^M.w*ve T:wN;' w3. n+; ww ? _. . .. .... Cs.137 . . .18 18 0.06 150 18 80 .180 s , en ,, wr w., , ,.gm w;..,m.-wg p mg ms,.m mmxrgg.w,-- .. . , m; . , y.y . . . , , . ... . y~w ny;, r.- , eg g. .; ,;. ,g e v..r.m., me  ;,.n ;;v. m... , , ,.~ m s 15 15, , . , us.La-140" 15. . . ~ (a) hinttiply the values in this table by IE-9 to convert to tmits of pCi/ml. (b) hfuttiply the estocs in this table by lE-9_to convert to units oIpCi/g. .. Total activity, parent plus daughter activity. -80 E E E E E M E E E E E E E E E E E APA-ZZ-01003 Rev. 6 TABLE 9.11-C (Continued) 3 TABLE NOTATIONS (1) This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together 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 environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13, Revision 1, July 1977. (3) The LLD is described in Attachment 1. 81- APA-ZZ-01003 Rev. 6 9.12 RAD 10 LOG 1 CAL ENVIRONMENTAL MONITORING LAND USE CENSUS  ! 9.12.1 Controls (CTSN 41835) 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.

APPLICABILITY: At all times. ACTION:

a. With a Land Use Census identifying a location (s) that yields a calculated dose or dose commitment greater than the values currently being calculated in REC 9.8.2.1, identify the new location (s) in the next Annual Radioactive EfTluent Release Report, pursuant to Technical Specification 6.9.1.7.

b. With a Land Use Census identifying a location (s) that yields a calculated dose or dose l

commitment (via the same exposure pathway) 20% greater than at a location from which sampl:s are currently being obtained in accordance with REC 9.11.1.1, add the new I 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 I 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 i was conducted. In the next Annual Radiological Environmental Operating Repon include 3I the revised figure (s) and tables reflecting the new sample location (s) with information supporting the change in sample location. ll

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

9.12.2 Surveillance Reauirements i 9.12.2.1 The Land Use Census shall be conducted during the growing season at least once per 12 months gl 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 agriculturnuthorities and/or residents. The results of gi l the Land Use Census shall be included in the Annual Radiological Environmental Operating Repon i pursuant to Technical Specification 6.9.1.6. a g I I I

• Broad leaf vegetation sampling of at least three difTerent kinds of vegetation may be performed at the SITE BOUNDARY in each g to two different direction sectors with the highest predicted D/Q's in lieu of the garden census. Specifications for broad leaf vegetation sampling in Table 9.11-A, Part 4.c shall be followed, including analysis of control samples.

Q APA-ZZ-01003 Rev. 6 9.13 RADIOLOGICAL ENVIRONMENTAL MONITORING INTERLABORATORY COMPARISON 1 PROGRAM 9.13.1 Controls (CTSN 41835) 9.13.1.1 Analyses shall be performed on radioactive mateiials supplied as part of an Interlaboratory Comparison Program that has been approved by the USNRC. APPLICABILITY: At all times. ACTION:

a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in The Annual Radiological Environmental Operating Report pursuant to 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 9.13.2.1 The Interlaboratory Comparison Program shall be described in the plant procedures. A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 6.9.1.6. l ) O 9 1 APA-ZZ-01003 l Rev. 6 l ~. 10. ADMINISTRATIVE CONTROLS I 10.1 MAJOR CHANGES TO L10UID AND G ASEOUS RADWASTE TREATMENT SYSTEMS 10.1.1 Licensee-initiated major changes to the Radwaste Treatment Systems (liquid and gaseous):

a. Shall be reported to the Commission in the Annual Radioactive Efiluent Release Report for the period in which the evaluation was reviewed by the On-Site Review Committee (ORC). 3 The discussion of each change shall contain: g

1) A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59;

2) Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; i

3) A detailed description of the equipment, components and process involved and the interfaces with other plant systems; l

4) An evaluation of the change, which shows the predicted releases of radioactive g i materials in liquid and gaseous effluents that differ from those previously g predicted in the License application and amendments thereto;

5) An evaluation of the change, which shows the expected maximum exposures to a g' MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general 3; 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 gaseous effluents, to the actual releases for the period prior to when the changes

_s are to be made;

7) An estimate of the exposure to plant operating personnel as a result of the change; I and

8) Documentation of the fact that the change was reviewed and found acceptable by the ORC.

b. Shall become effective upon review and approval by the ORC and in accordance with Technical Specification 6.5.3.1.

l 3 10.2 CH ANGES TO THE OFFSITE DOSE CALCULATION MANU AL (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 revision of the ODCM must include, as a minimum, the Health 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 Efiluent Release Report pursuant to Technical Specificatiori 6.14. I I _) APA-ZZ-01003 Rev. 6 % 11. REFERENCES -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. I1.2 Title 10, " Energy" Chapter 1, Code of Federal Regulations, Part 50, Appendix I; U.S. Government Printing Office, Washington, D.C. 20402. I1.2.1 10 CFR 50.36 a (b) 11.3 Title 40, " Protection of Environment", Chapter 1, Code of Federal Regulations, Part 190; U.S. Govemment Print Office, Washington, D.C. 20402. I1.4 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 11.5.1 Letter NEO-54, D. W. Capone to S. E. Miltenberger, dated January 5,1983; Union Electric Company correspondence. -11.5.2 Letter BLUE 1285, "Callaway Annual Average X/Q and D/Q Values", J. H. Smith (Bechtel Power Corporation), to D. W. Capone (Union Electric Co.), dated February 27,1984. 11.53 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. 11.5.4 Reference Deleted 11.5.5 Private Communication, H. C. Lindeman & B.F. Holderness, August 6,1986 11.5.6- Calculation ZZ-67, " Annual Average Atmospheric Dispersion Parameters", April 1989. I1.6 Union Electric Company Callaway Plant, Unit 1, Final Safety Analysis Report. I1.6.1 Section 11.5.2.23.1 11.6.2 Section 11.5.2.2 3.4 - 11.63 Section 11.5.2.1.2 [ 11.6.4 Section I1.5.2.2J.2 11.6.5 Section 11.5.2.233 11.6.6 Section 11.23.3.4 {= 11.6.7 Section 11.2.3.4 3 11.6.8 Section 11.5.2J.3.1 11.6.9 Section 11.5.2.3J.2 11.6.10 Section 11.5.23.2.3 11.6.11 Seetton 11.5.2J.2.2 11.6.12 Section 23.5 3 11.6.13- Section 23.5.2.1.2 .. 11.6.14 Section 9.2.6 11.6.15 Section 9.2.7.2.1 85- APA-ZZ-01003 Rev. 6 11.6.16 Section 63.2.2 11.6.17 Table 11.1-6 11.6.18 Deleted 11.6.19 Deleted 11.6.20 Deleted 11.6.21 Deleted 11.6.22 Table 23-68 11.7 Union Electric Company Callaway Plant Environmental Report, Operating License Stage. I1.7.1 Table 2.1-19 11.7.2 Section 2.1.23 11.73 Section 2.133.4 11.7.4 Section 5.2.4.1 11.7.5' Table 2.1-19 11.8 U.S. Nuclear Regulatory Commission, Preparation of Radiological Effluent Technical Specificetion for Nuclear Power Plants", USNRC NUREG-0133, Washington, D. C. 20555, October 1978. I1.8.1 Pages AA-1 through AA-3 g 11.8.2 Section 5J.1J B 11.83 Section 43 11.8.4 Section 53.1.5 ~ l1.8.5 Section 5.1.1 11.8.6 Section 5.1.2 11.8.7 Section 5.2.1 11.8.8 Section 5.2.1.1 11.8.9 Section 5.3.1 11.8.10 Section 3.8 11.8.11 Section 33 11.9 U.S. Nuclear Regulatory Commission, "XOQDOQ, Program For the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations", USNRC NUREG-0324, Washington, D. C. 20555. I1.9.1 Pages 19-20 Subroutine PURGE 11.10 Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors", Revision 1, U. S. Nuclear Regulatory Commission, Washington, D. D. 20555, July,1977. 11.10.1 Section c.l.b 11.10.2 Figures 7 through 10 11.103 Section c.4 APA-ZZ-01003 s Rev. 6 11.11 Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor ) EfDuents for the Purposes of Evaluating Compliance with 10 CFR Pan 50, Appendix I", Revision 1. U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, October 1977. I1.11.1 Appendix C, Section 3.a 11.11.2 Appendix E, Table E 15 11.11.3 Appendix C, Section 1 11.11.4 Appendix E, Table E 11 11.11.5 Appendix E, Table E o 11.12 U. S. Nuclear Regs), tory Commission," Methods fcr Demonstrating LWR Compliance with the EPA Cranie. Fus! Cycle Standard (40 CFR Part 190)", USNRC NUREG-0543, Washington, D. C. 20555, Jancary 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 111, Page 8 11.13 Management Agreement for the Public Use ofLands, Union Electric Company and the State of Missouri Department of Conservation, December 21,1982. I1.13.1 Exhibit A i1.14 MISCELLANEOUS REFERENCES s 11.14.1 Drawing Number M-109-0007-06, Revision 5 11.14.2 Callaway Plant Annual Environmental Operating Report (updated annually) 11.14.3 UE Safety Analysis Calculation 87-001-00 11.14.4 Calculation ZZ-48, " Calculation ofinhalation and Ingestion Dose Commitment Factors for the Adult and Child", January,1988 11.14.5 HPCI 89-02," Calculation of ODCM Dose Commitment Factors", March,1989 11.14.6 HPCI 87-04," Calculation of the Limiting Setpoint for the Contaiwent 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 Effluent Streams", June,1988 11.14.8 Calculation ZZ-57, Dose Factors for Eu-154", January,1989 11.14.9 Calculation 22-78, Rev. 2, "ODCM Gaseous Pathway D-se Fa:tsrs for Adult Age Group", July,1992. 11.14.10 HPCI 88-08, " Performance Testing of the Enviror'raent TLD System at Callaway Plant", August, 1989. 11.14.11 Calculation ZZ-250, Rev. O, *ODCM Gaseous ?athway 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. 11.14.13 Calculation HPCI 95-004 (Rev. 0)," Calculation of Liquid Efnuent Dose Commitment factors (A,,) ) for the Adult Age Group", June,1996. i

-8 7-

. ~ APA-ZZ-01003 ) Rev. 6 l l . 11.15 U. S. Nuclear Regulatory Commission, "XOQDOQ: Computer Program for the Meteorological i Evaluation of Routine Effluent Releases at Nuclear Power Stations", USNRC NUREG/CR 2919 September 1982, Washington, D. C. 20555 1 11.15.1 Section 4, " Subroutine PURGE", pages 27 and 28 11.16 Regulatory Guide 4.13 " Performance, Testing, and procedural specifications for Thermoluminiscence Dosimetry: Environmental Applications "(Revision 1), July 1977; USNRC, l, Washington, D. C. 20555 W l1.17 TID-7004, " Reactor Shielding Design Manual", Rockwell, Theodore, ed; March 1956. 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,"1SOSHLD-II: Code Revision to include calculation of Dose Rate from Shielded Bremstrahlung Sources", Simmons, G. L., et al; March 1967 l 5 11.20 BNWL-236, Supplement 2, "A Revised Photon Probability Library for use with ISOSHLD- 111", Mansius, C. A.; April 1969. 11.21 ANSI N13.10-1974," Specification & Performance of On-Site Instrumentation for Continuously Monitoring Radioactivity in Effluents"; September,1974 11.22 Nuclear Regulatory Commission Generic Letter 89-01," Guidance for the Implementation of Programmatic Controls for RETS in the Administrative Controls Section of Technical Specifications and the Relocation of Procedural Details of Cu. vent RETT to the Offsite Dose g Calculation Manual or Process Control Program", January 1989 E I1.23 NRC Answers to 10 CFR 20 Implementation Questions 11.23.1 Letter, F. J. Congel to J. F. Schmidt, dated December 9,1991. / 11.23.2 Internal USNRC memo, F. J. Congel to V. L. Miller, et al, dated April 17,1992. I1 21.3 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. g 11.24 USNRC Inspection Report 50-483/92002(DRSS) Section 5, page 5. W l1.25 HPCI 96-005," Calculation of Maximum Background Value for HB RE-18". I1.26 EGG-PHY-9703," Technical Evaluation Report for the evaluation of ODCM Revison 0 (May, 1990) Callaway Plant, Unit 1", transmitted via letter, Samual J. Collins (USNRC) to D. F. Schnell (UE), dated July 12,1996. I I. I Ii ) Il I,I A PA-ZZ-01003 Rev.6 LOWER LIMIT OF DETECTION (LLD) A detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, H ASL-300 (revised annually), Curie, L.A. " Limits for Qualitative Detection and Qualitative Determination - Application to Radiochemistry", Anal Chem. 40. 586-93 (1986), and Hanwell, J.K.," Atlantic Richfield Hanford Company Repon 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. LLD = 4.56 S* E x V x 2.22E6 x Y x exp(-2at Where: LLD = the "a priori" lower limit of detection (microCuries per unit mass or volume), = gb the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute), _ E = the counting efficiency (counts per disintegration), V = the sample size (units of mass or volume), 2.22E6 = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, A " ~ the radioactive decay constant for the particular radionuclide (sec '), and At = the elapsed time between the midpoint of the sample collection period, and the time of counting (sec), for effluent samples, or A = the elapsed time between the end of the sample collection period, and the time of counting (sec), for environmental samples. Typical valees of E, V, Y, and At should be used in the calculation. It should be reco;,nized that the LLD is defined as a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori(after the fact) limit for a particular measurement. 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. Page1of1 ATTACHMENT 1 APA-ZZ 01003 Rev. 6 BASES FOR RADIOLOGICAL EFFLUENT CONTROLS i The BASES presented below summarize the reasons for the specified Radiological Effluent Control, but in accordance with 10 CFR 50.36 are not part of these controls. REC 9.1 R ADIOACTIVE L10UID EFFLUENT MONITORING INSTRUMENTATION Refer to FSAR CN #94-51 REC 9.2 RADIOACTIVE G ASEOUS EFFLUl?NT 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 effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in Appendix B, Table II, Column 2 to 10 CFR 20.1-20.601. 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 objectives of Appendix 1,10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR Part 20.1301 to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2. The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLD's). 1 Page 1 of 7 ATTACHMENT 2 APA ZZ-01003 Rev. 6 B ASES FOR R ADIOLOGICAL EFFLUENT CONTROLS REC 9.4 DOSE FROM LIOUlD EFFLUENTS This section is provided to implement the requirements of Sections ll.A and IV.A of Appendix 1,10 l CFR Part 50. The Limiting Condition for Operation implements the guides set fonh in Section ll.A gI of Appendix 1. The ACTION statements provide the required operating flexibility and at the same 3l ' time implement the guides set fonh in Section IV.A of Appendix 1 to assure that the releases of radioactive material in liquid emuents to UNRESTRICTED AREAS will be kept "as low as is a reasonably achievable" g, 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 concentrations in the finished drinking water that are in excess of the requirements of 40 ll = CFR Pan 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 g! Appendix ! be shown by calculational procedures based on models and data, such that the actual 3l exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be l substantially underestimated. The equations specified in the ODCM for calculating the doses due to g 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 B' Reactor Effluents with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic and Dispersion of Effluents from accidental and Routine React'or Releases i for the Purpose ofImplementing Appendix I", April 1977. REC 9.5 LIOUID RADWASTE TREATMENT SYSTEM I The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate ponions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This section implements the requirements of 10 CFR Pan 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Pan 50 and the design objective given in Section ll.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the g Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives g set forth in Section ll.A of Appendix I,10 CFR Part 50, for liquid effluents. I I I J Page 2 of 7 ATTACHMENT 2 APA-ZZ-01003 Rev. 6 , BASES FOR RADIOLOGICAL EFFLUENT CONTROLS REC 9.6 GASEOUS EFFLUENTS DOSE RATE This section is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY , from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED. AREAS. The dose rate limits a*e the doses associated with the concentrations of 10 CFR Part 20.1-20.601, Appendix B, Table 11, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the 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 THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODCM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrem / year to the whole body or to less than or equal to 3000 mrems/ year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/ year. The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLD's). 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 tne release of gaseous effluents 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* *' , 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 of iodine and particulate partitioning, plateout on plant and ventilation system surfaces, and the 99% efficiency 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 activity due to its relatively greater concentration in the Unit Vent. ' Cohen, Paul, Water Coolant Technoloev of Power Reactors, Table 5.19, page 198. American Nuclear Society. . 1980. 5 NUREG-0772, " Technical Bases for Estimating Fission Product Be'avior During LWR Accidents", Silberberg, M., editor, USNRC; Figure 4.3, page 4.22. June,1981. Page 3 of 7 A'ITACHMENT 2 . APA-ZZ-01003 ll Rev. 6 5l . g D ASES FOR RADIOLOGICAL EFFLUENT CONTROLS gI ) REC 9.7 DOSE - NOBLE G ASES This section is provided to implement the requirements of Sectior.s II.B. lil.A, and IV.A of Appendix 1,10 CFR Part 50. The Limiting Conditions for Operation implements the guides set forth in Section Il !!.B of Appendix 1. The ACTION statements provide the required operating flexibility and at the 3' same time implement the guides set fonh in Section IV.A of Appendix 1 to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is gI reasonably achievable". g The Surveillance Requirements implement the requirements in Section III.A of Appendix 1 that confonnance with the guides of Appendix 1 be shown by calculational procedures based on models l,i =l 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 methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of g! 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 El l Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, g October 1977 and Regulatory Guide 1.111," Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Efiluents in Routine Releases from Light-Water Cooled Reactors", Revision I, 5J) July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE l BOUNDARY are based upon the historical average atmospheric conditions. u REC 9.8 DOSE -LODINE-131. & 133. TRITIUM. AND RADIOACTIVE M ATERI AL IN PARTICULATE FORM Ii) l l This section is provided to implement the requirements of Sections ll.C, Ill.A, and IV.A of Appendix 1,10 CFR Pan 50. The Limiting Conditions for Operation are the guides set fonh in Section ll.C of ' Appendix 1. The ACTION sotements provide the required operating flexibility and at the same time implement the guides set forth ir. %ction IV.A of Appendix I to assure that the release of radioactive ll material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as reasonably W ' achievable". The ODCM calculational methods specified in the Surveillance Requirements implement the requirements in Section Ill.A of Appendix 1 that conformance with the guides of g Appendix I be shown by calculational procedures based on models and data such that the actual 3 exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methodology and parameters for calculating g; the doses due to the actual release rates of the subject materials are consistent with the methodology g provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1", Revision 1, October 1977 and Regulatory Guide 1.111," Methods for Estimating Atmospheric l Transport and Dispersion of Gaseous Effluent sin Routine Releases from Light-Water Cooled W ' 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 lodine-13), and 133, tritium, and radionuclides in paniculate form with half lives greater than 8 days are E1 g dependent upon the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were: (1) g individual inhalation of airbome radionuclides,(2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man,(3) deposition of radionuclides onto grassy areas E where milk animals and meat producing animals graze with consumption of the milk and meat by man, and (4) deposition on the ground with subsequent exposure of man. Page 4 of 7 ATTACHMENT 2 APA-ZZ-01003 Rev. 6 . BASES FOR RADIOLOGICAL EFFLUENT CONTROLS REC 9.9 G ASEOUS R ADWASTE TREATMENT SYSTEM The OPERABILITY of the WASTE GAS HOLDUP SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the system will be available for use whenever gaseous efnuents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable" This control implements the requirements of 10 CFR Pan 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section ll.D of Appendix ! to 10 CFR Part 50. The specified limits governing the use of appropriate ponions of the systems were specified as a suitable fraction of the dose design objectives set fonh in Sections ll.B and II.C of Appendix 1,10 CFR Part 50, for gaseous effluents. REC 9.10 TOTAL DOSE This REC is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Pan 20.1301. The control requires the preparation and submittal of a Special Report whenever the calculated doses due to releases of radioactivity and the radiation from uranium fuel cycle sources exceed 25 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 resul: ant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix 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 MEMBER OF THE PUBLIC to within the 40 CFR Pan 190 limits. For the purposes of the Special Repon, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose 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 Repon with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR Pan 190.11 and 10 CFR 20.2203, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in REC 9.3.1.1 and 9.6.1.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is pan of the nuclear fuel cycle. Page 5 of 7 ATTACHMENT 2 . APA-ZZ-01003 g Rev. 6 g BASES FOR R ADIOLOGICAL EFFLUENT CONTROLS There are three defined effluent release categories: 1.) Releases directly to the hydrosphere; 2.) noble j gas releases to the atmosphere; and,3.) radioiodine and particulate releases to the atmosphere. For nu each efnuent release category, it is assumed in the dose calculations that an individual with the highest dose potential is the receptor in general, the adult is considered to be the critical age group for liquid efnuents, and the child age group is the most limiting for radioiodines and particulates in gaseous efnuents. Thus, it is highly unlikely or impossible for the same individual to simultaneously receive the highest dose via all three effluent categories. For most reactor sites, it is also unlikely that g)) different potential dose pathways would contribute to the dose to a single Legl individual. Since it is difficult or impossible to continually determine actual food use patterns and critical age groups, for calculational purposes, assumptions are made which tend to maximize doses. Any refinements in the assumptions would have the effect of reducing the estimated dose. For radionuclides released to the hydrosphere, the degree of overestimation in most situations is such that no individual will receive a signincant dose. These conservative assumptions generally result in an overestimation of dose by one or two orders of magnitude. Since these assumptions are reflected in the Radiological Efnuent Controls limiting radionuclide releases to design objective individual doses, no offsite individual is likely to actually receive a significant dose. Since the doses from 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 airborne pathways and dmes resulting from direct radiation need to be considered in determining potential compliance to 40 CFR 190 REC. 9.11 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ' The Radiological Environmental Monitoring Program required by this REC provides representative g measurements of radiation and of radioactive materials in those exposure pathways and for those g " radionuclides that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the station operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the Radiological Ef0uent Monitoring Program by verifying that the measurable concentrations of radioactive materials and lev.els of radiation are not higher than expected on the basis of the efnuent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on 3 operational experience. g The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLD's). The LLT/s required by Table 9.ll-C are considered optimum for routine environmental measurements in industrial laboratories. I C I ' NUREG- 0543, " Methods for Demonstrating LWR Compliance with the EPA Uranium Fuel Cycle Standard (40 CFR 190)", Congel, F. J., Office of Nuclear Reactor Regulation, USNRC. January,1980, pp. 5- 8. Page 6 of 7 ATTACHMENT 2 . . ~ ~ . - . . .. . . - - - . . - . - .-. . .~ . . - - . ~ .. . - - . . -

APA ZZ-01003 Rev. 6 I B ASES FOR RADIOLOGICAL EFFLUENT CONTROLS i.

m j 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 l are identified and that modifications to the Radiological Environmental Monitoring Program given in l the ODCM are made ifrequired 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 agricuhural authorities, shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix 1 to 10 CFR Part

50. Restricting the census to gardens of greater than 50 m2 provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) 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 growl g broad leaf vegetation (i.e.,

similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m . j REC 9.13 RAD 10 LOGICAL ENVIRONMENTAL MONITORING INTERLABORATORY COMPARISON l 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 environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are valid for the purpose of Section IV.B.2 of Appendix 1 to 10 CFR Part 50. i. c e i 4 4 1 i ( Page 7 of 7 ATTACHMENT 2 APA-ZZ 01003 Rev. 6

SUMMARY

REVIEW OF RADIOLOGICAL EFFLUENT TECH SPECS POTENTI ALLY AFFECTED BY THE IMPLEMENTATION OF THE REVISED 10CFR20 The following is a summary review of the current Tech Specs that are potentially affemd by the implementation of the revised 10CFR20. In general, the potential impact is due to changes in the Emuent Concentration Values (ECVs)in 10CFR20, Appendix B, Table 2, Columns I and 2 (formerly MPC's), and 10CFR20.1601.

This summary is not intended to review those changes that may be necessary as a result of the eventual issuance of the Generic Letter.

The NRR staff has stated that the current level of effluent controls is sufficient to protect the health and safety 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, which basically requires that the more restrictive requirement (Tech Specs or 10CFR20) be implemented.

DEFINITIONS OF RESTRICTED AREA & MEMBER OF THE PUBLIC, AND TECH SPEC 5.1.2, SITE BOUNDARY FOR G ASEOUS 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 Specifications (TS 1.17). There is no corresponding definition of Controlled Area in the former rule.

s 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.

In addition, the NRC's backfit analysis t, performed pursuant to 10 CFR 50.109, concludes that the revisions 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 Area for currently licensed facilities.

There is also no requirement for the existence of a Controlled Area as defined in the revised rule2 , therefore it is not necessary that one be created at Callaway.

l 1 " Final Backfit Analysis for the Revision of 10CFR20, " Standards for Protection Against Radiation"", USNRC, office of Nuclear Regulatory Research, Division of Regulatory Applications. August,1990. (Available USNU' Public Documents Review.)

2 Refer to Question 26(a)(4th set).

Page 1 of 12 APPENDIX A

APA 22-01003 Rev. 6

) 'Ihe definition of the Member of the Public is significantly different in the revised rule relative to that provided 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 g 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 5

times, be within the Restricted Area. The major difference is that pursuant to the revised rule, the Member of the Public receives dose against the occupational dose limits of 10 CFR 20.1201 once inside the Restricted Area.

but the Tech Spec definition would limit the dose within the Restricted Area to the limits of 10 CFR 20.1301.

l

=

Since the limit provided in 20.1301 is much lower than that of 20.1201, the continued use of the more restrictive 40 CFR 190 and Tech Spec 1.17 definitions for the Member of the Public is appropriate and is reauired 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 array of definitions and dose limit applicability. For conservatism 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 activities are also limited to the more restrictive dose limits of 10 CFR 20.1301.

There are no changes recommended for those definitions and maps relative to the Restricted Area Site Boundary, and dose to the Member of the Public.

TECH SPEC 6.8.4.E.2, l LIQUID EFFLUENT RELEASE RATE l 5

LIMITS (REC 9.3)

On December 1,1992, Union Electric Co. provided notification3 ofintent to implement the revised 10 CFR 20. g Parts 20.1001- 20.2401 and associated appendices, pursuant to 10 CFR 20.1008(a). The revised rule was fully 3 implemented on January 1,1993. The following provides clarification with respect to compliance to l 10 CFR20.1001-20.2401 and Callaway Plant Technical Specifications 6.8.4.e (2) and 6.8.4.e (7).

Union Electric implemented the use of the revised Appendix B, Table 2 values concurrent with the l implementation of the revised rule. Technical Specification 6.8.4.e (2) requires that the concentration of radioactive material in liquid discha ;;ca 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 g shall implement all provisions of these sections,.. . and shall provide written notification.. . tid the licensee is g idopting early implementation (of the revised rule) and associated appendices." 10 CFR 20.10lM (b) requires tiw once implemented, "the applicable section of(the revised rule) shall be used in lieu of any section (of the o, rule) that is cited in license conditions or technical specifications." It further states, "if the requirements of (Q revised rule) at: more restrictive than the existing license condition, then the licensee shall comply with (the l

=

revind rum I

3 ULNRC 92-2729, D. F. Schnell to A. Bert Davis, dated December 1,1992.

Page 2 of 12 APPENDIX A

APA-ZZ-01003 Rev. 6 i

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) Additionally, the NRC had clarified the applicability of the revised Appendix B values to th: Technical Specification instantaneous release rate limits via their formal response to three separate lice nsee questions.

Question # 18 states4 that the Tech Spec instantaneous release rate limit is based on the old I art 20 concentrations, and asks if changes are required in the Tech Specs and ODCM as a result at the revised rule.

The NRC replies " the instantaneous release rates for liquid effluents, to the extent that they directly reference Appendix B concentration values, will need to be changed. The corresponding bases and certain alarm ser- j points will have to be changed by license amendment." l l

5 Question # 23 asks if computer data bases that use the old Appendix B values must be revised to the use the l new values. The NRC simply answers,"Yes". l 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 that the alarm set-points of liquid effluent monitors are likely to require change, since they are based on 10 CFR 20 Appendix B concentrations, as required by Tech Specs.

l Because Appendix B concentration values differ for many radio nuclides between the old and new versions of Part 20. these set points may have to be changed. This is analogous to a restriction in flow rate, and the NRC cites the reduction in Appendix B concentrations as the root cause of the change.

Based on the preceding information, Union Electric implemented the use of the revised Appendix B values concurrent with the implementation of the revised rule on January 1,1993. Because there were no values in the revised Appendix B for dissolved and entrained noble gases in liquid effluents, the old value of 2E-4 uCi/ml was used pending regulatory guidance.

l The Callaway Plant Technical Specifications contain, in Section 6.8.4.e, several specifications which provide appropriate limits on the maximum quarterly and annual whole body and organ dose to the Member of the Public from the discharge ofliquid and gaseous radioactive efnuents. Compliance with these specifications demonstrates compliance with the limits of 10 CFR 50, Appendix I, 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 20.1301.

l However,compliancs with the dose rate limits of Specifications 6.8.4.e items (2) and (7) with respect to the k implementation of the revised mle 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 provided annual average concentration limits on liquid and gaseous efnuents, and specifically referenced the limits of Appendix B, Table II, Columns 1 and 2.

4 Letter, F. J. Conjel (USNRC) to J. F. Schmitt (NUMARC), dated December 9,1991. page 16 of Enclosure 1.

5 bid, page 14 of Enclosure 1.

6 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.

Page 3 of 12 APPENDlX A

APA ZZ 01003 Rev. 6 Unlike the former rule, the values in the revised Appendix B, Table 2, Columns I and 2 do not of themselves constitute a limit on the release rate of radioactive effluents, but rather, as discussed in 10 CFR 20.1302 (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 restrictive than the i 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 corresponding provision in the new rule to 20.106.10 CFR 20.1008(e) requires that if a license condition cites a provision in the old rule for which there is no corresponding provision in the new rule, then the license condition remains in force until g, there is a technical specification change. g, The values of Appendix B, Table 2, Columns I and 2 of the revised rule did not change in a uniform fashion, j i.e., certain nuclides numerically decreased in value whereas others numerically increased in value. 1 I

Funbermore, 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 effluents. 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 10 CFR 50, Appendix 1 and 40 CFR 190, and the dose calculational methodologies of Regulatory G'uide 1.109, I which are largely based on the dosimetry system ofICRP 2 10, l

i Since the values of the revised Appendix B. Table 2, Columns 1 and 2 did not unifonnly increase or decrease in i value, it is not possible to determine whether Appendix B, Table 11 of the former rule or Appendix B, Table 2 of l the revised rule provides, in toto, the more conservative values for implementation of the subject license 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 1 and 40 CFR 190, and Regulatory Guide 1.109. Furthermore, the

~

operational history of the Callaway Plant demonstrates that the use of the 10 CFR 20.1- 20.601, Appendix B, B Table Il values is appropriate to maintain compliance with the requirements of 10 CFR 50, Appendix ! and 5l 40 CFR 190, which, in turn, demonstrates compliance with the 100 mrem /yr dose limit of 10 CFR 20.1301. The I concentration limits of the old Appendix B, Table 11 were based on a dose of 500 mrem /yr, which, when ,

expressed as a dose rate, is equal to .057 mrem /hr. Compliance with the requirements of Technical Specifications 6.8.4.e (2) and (7) using 10CFR 20.106, Appendix B, Table 11 values is conservative with respect  !

to the 2 mrem /hr limit of 10CFR20.1301(a)(2). Additionally, Technical Specifications 6.4.8.e(2) and (7) l specifically require the use of Appendix B, Table II to 10CFR20.1- 20.601, since there is no corresponding provision in the revised rule.

Thus,10 CFR 20.1008 (c) and (e) require the continued use of the values provided in Appendix B, Table 11 to g l 10 CFR 20.1 20.601 for the implementation of Technical Specifications 6.8.4.e, items (2) and (7). E Although the 2 mrem /hr limit of 10 CFR 20.1301(a)(2) was referenced in the preceding discussion, it is  ;

important to note that the regulation specifically states that this limit is applicable to external sources. Since, for the Callaway Plant, the only dose pathway to man from the discharge ofliquid radioactive effluent 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.

s International Commission on Radiation Protection, Publication 26, " Recommendations of the international E

Commission on Radiation Protection", Annals of the ICRP, Volume 1, No. 2,1977. 3 9 International Commission on Radiation Protection, Publication 30, " Limits for intakes of Radionuclides t'y E Workers", Annals of the ICRP, Volume 2, No. 3/4,1979. g to International Commission on Radiation Protection, Publication 2, " Report of Committee 11 on Permissible Dose for Internal Radiation",1960.

Page 4 of 12 APPENDIX A

APA-ZZ-01003 Rev. 6

} Union Electric re instituted the use of the values in Appendix B, Table 11, Columns 1 and 2, to 10 CFR l 20.1- 20.601 for Technical Specifications 6.8.4.e, items (2) and (7) pursuant to the requirements of 10 CFR 20.1008(c) and (e), on May 4,1993.

This position was affirmed by the USNRC on June 30,1%.'11 EFFLUFET CONCENTRATION VALUE FO!! GP.OSS ALPHA IN LIQUID EFFLUENTS There are two values in the revised Appendix B for unknown mixtures in liquid efnuents: 2E-9 and IE-6 uCi/ml. The less restrictive value is appropriate ifit is known that certain nuclides are "not present". The appropriate value for gross alpha in liquid effluents at the Callaway Plant from Appendix B, Table 2. Column 2 is 1E-6 uCi/ml.

He 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-l 13m.

Cd-113, In ll5,1-129, 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-23 8, 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 efnuents do not apply, since they specifically apply to gaseous effluents as indicated by the designation of applicable lung clearance classifications for ev.h o rth nuclides listed. The NRC's response to Question # 71 reiterates that ingestion ALI's do not have lung cie.rance classifications, which is also consistent with ICRP 30 and all other industry standards. Additionally, several of those listed in the list for liquid effluents also appear in the list of nuclides given for airbome activity, which indicates that only those specifically listed with the liquid effluent value apply.

Of those nuclides listed for unknown mixtures in liquid effluents, only Ra-224, Th-228, U-234, U-235, U-236, U-238, and Cm 242 are LWR produced alpha emitting nuclides. St-90, Cd-113m,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 l remainder of the nuclides in the list are not LWR produced. I 1

The phrase "not present" is not defined in the revised 10 CFR 20, however there is a large body of information '

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 activity, 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 cor.sistent with the basis of the revised rule, the NRC's response to questions l regarding the meaning of"not present", and the current ICRP guidance as shown below:

a. The revised rule is based on the dosimetry and methodology of!CRP 30 12, which in paragraph 3.1.3, describes the use of the current ten percent rule.

b. The NRC's response to Question # 14613 clearly indicates that the ten-percent rule is applicable to Appendix B.

Il Letter, Thomas E. Murley, Director, NRR, USNRC, to Thomas E. Tipton, NUMARC, dated June 30,1993.

12 j ICRP Publication 30, " Limits for intakes of Radionuclides by Workers", in Annals of the ICRP, Volume 2.

Number 3/4,1979.

13 Letter, Frank J. Congel, Director, DRPEP, USNRC, to John F. Schmidt, NUM ARC, dated September 14, 1992 (commonly referred to as the 4th set of Q&A)

Page 5 of 12 APPENDIX A

l APA-ZZ-01003 E Rev. 6

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, c. The current ICRP recommendations on the release of radioactive materials to the environment",

and the updated recommendations 15 to ICRP 30 continue to propagate the ten-percent rule, and I

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 effluents have been analyzed for transuranic 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 cencentration of the TRU nuclides can be inferred through the use of a tracer nuclide, such as Cc-144.

Ce-144 is particularly well suited for this purpose in that it is a fission product, can be measured by gamma ray I

spectroscopy, and is chemically similar to the TRU nuclides. Based on published ORIGEN code calculations 16 g of a representative LWR, and assuming a 90 day decay, the ratio of the nuclides of interest to Ce-144 is: E Ra-224/Ce 144 1.45 E-9 gi Th-228/Ce-144 1.45E-9 g' U-234/Ce-144 1.14E-6 U-235/Cc-144 1.75 E-8 U-236/Ce-144 2.58E-7 U-238/Ce-144 3.24 E-7 1 Cm 242/Ce-144 2.66E-2 Vollique, et al 37, found the Cm-242/Ce-144 ratio to be 6.5E-3, which is consistent with the above value. Based i;l on the above, it can be seen that Cm-242 is the only nuclide with a significant Ce-144 ratio. )

Based on the data contained in the Semiannual Effluent Release Reports for the period January,1989-July,1992, Ce 144 accounted for less than 0.3% of the total fission and activation product activity in liquid effluents, and less than SE-6% of the total activity discharged in liquid effluents during the same period. g Therefore, the maximum activity that could have been discharged of each of the above listed nuclides is much less than 10%. Accordingly, these nuclides are "not present".

g TECH SPEC 6.8.4.E.4, DOSE FROM LIQUID EFFLUENTS (REC 9.4), & TECH SPEC 6.8.4.E.5, LIQUID RADWASTE TREATMENT SYSTEM (REC 9.5)

I4 ,

These specifications are derived from 10CFR50, Appendix 1, and are not affected by the revised rule. Doses are calculated in accordance with Regulatory Guide 1.109 which has not been revised. No changes are anticipated for these specifications.

N ICRP Publication 56," Age-dependent doses to Members of tl e Public from the Intak'e of Radionuclides: Part I

1", Annals of the ICRP, Volume 20, Number 2,1989.

15 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.

i 16 Licht Weter Reactor Nuclear Fuel Cvele. Wymer, Raymond G. and Vondra, Benedict L., editors, Table 6, pages 70- 71 and Table 7, page 72. CRC Press,1981.

17 Vollique, P. G., et al, " Solubility of Transuranic Nuclides in Aerosols in Two Ginna Steam Generator Work

~.nvironments".

Proceedings of the Twenty-First Midyear Topical Meeting of the Health Physics Society, Pages

) 251-260. 1987.

Page 6 of 12 APPENDlX A g g

l t

1 APA ZZ 01003 Rev. 6

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l FSAR 16.11.1.1, CURIE CONTENT OF OUTDOOR LIQUID STORAGE TANKS Re purpose of this specification 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.

He effect of accidental contamination of the nearest ground water discharge locations due to accidental rupture of tanks containing radioactive liquids was performed as detailed in FSAR Section 2.4.13.3. It was assumed that the liquid contents 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 after a tank rupture will be below the original 10CFR20, Appendix B. Table 11, Column 2 values by the time the contaminated ground water reaches the nearest stream tributaries. The dilution capability of the streams is sufficient 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 1 and 11 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 l limit identified in Tech Spec 16.11.1.1 (Even though the spent resin storage tank is not an outdoor tank, the data j was used for this calculation since it is expected to have the highest curie contents for Sr-90, Cs-137 and Co-60 l' and the postulated accident assumes that all liquid released immediately merges with the ground water.) The resultant peak concentrations at the discharge point at Logan Creek were calculated using the normalized values 3_ then compared to the revised Appendix B effluent concentration values (ECV). All calculated concentrations at the discharge point were less than the applicable ECV. l

>f  !

Based on the above calculation, the existing Tech Spec limit of 150 Curies is conservative in comparison to the

.evised 10CFR20, Appendix B values and is therefore still applicable.

) Page 7 of 12 APPENDlX A

APA ZZ-01003 au Rev. 6 TABLEI A. Curie Content of Radionuclides in the Primary Spent Resin Storage Tank I

i NUCLIDE Cl* (in tank) Ci (normalized to 150 Ci, total)

Mn-54 2.91 E+01 8.17E-01 Co-58 6.1OE+02 1.7IE+01 Co-60 2.56E+02 7.19E+00 Sr-89 9.80E+00 2.75E-01 Sr-90 1.35E+00 3.79E-02 Nb-95 Zr-95 3.00E+00 8.42E-02 g 2.12E+00 5.95E-02 g I-131 1.17E+03 ' 3.28E+01 Cs 134 1.78E+03 5.00E+01 Cs-137 1.48E+03 4.15 E+01 Ba-140 1.63E+00 4.58E-02 TOTAL 5.343E+03 149.91

• Values are from FSAR Table 2.4-28.

B. Peak Concentrations of Radionuclides at the Logan Creek Discharge Point NUCLIDE pCl/ml* (original pCi/ml(based on ECV %ECV cale) 150 Ci total)

Mn-54 3.lE-22 8.7E-24 3E-05 3E-17%

Co-60 3.6E-23 1.0E-24 3E-06 3E-20%

Sr90 1.2E-05 3.4E-07 SE-07 67.4 %

Cs-137 5.5E-06 1.5E-07 IE-06 15.4 %

• Values are from FSAR Table 2.4-30.

E I

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APA-ZZ-01003 Rev. 6

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TABLE II A. Curie Content of Radionuclides in Refueling Water Storage Tank t

} NUCLIDE Ci* (in tank) Ci (normalized to 150 Ci, A

total)

Mn-54 6.99E-06 2.19E-02 Co-58 3.36E-04 1.05E+00 Co-60 4.58E-05 1.43E-01 Sr-89 5.92E-05 1.85E-01 Sr-90 1.92E-06 6.02E-03 Nb-95 1.31E-06 4.10E-03 Zr-95 1.25E-06 3.92E-03 l 1131 2.34E-02 7.33E+01 Cs-134 1.39E-02 4.35E+01 Cs-137 1.01E-02 3.16E+01 Ba-140 2.56E-05 8.02E-02 TOTAL 4.788E 149.9 1

• Values are from FSAR Table 2.4-28.

l B. Peak Concentrations of Radionuclides at Logan Creek Discharge Poir:t t

NUCLIDE pCL/ml* (original pCi/ml (based on ECV %ECV cale) 150 Ci total) l Co-60 1.lE-30 3.4E-27 3E 06 l E-19%

Sr-90 2.5E-13 7.8E-10 SE-07 0.16%

Cs-137 8.4E-13 2.6E-09 lE-06 0.26%

• Values are from FSAR Table 2.4 30.

I Page 9 of 12 APPENDIX A

g APA-ZZ-01003 E.

Rev.6 TECH SPEC 6.8.4.E.7. DOSE RATE LIMIT FOR GASEOUS EFFLUENTS (REC 9.6) ,

nis specification provides a gaseous emuent dose rate limit conforming to the ECV's in 10CFR20, Appendix B, Table 2, Column 1. For the nuclides ofinterest to Callaway, the revised ECV's are numerically greater, therefore the current REC is more restrictive than the dose rates confonning to the revised Appendix B gi values.10CFR20.1008 requires the implementation of the more restrictive of the requirements of 10CFR20. E' 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 l specified in Appendix B, Table 2, averaged over a period of one year. Although not specified as a limit, this 1 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.

1 He revised rule, in 20.1301, specifies two limits on radioactivity in emuents: An annual dose limit of gi 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 allow licensees to utilize the 3{

concentration values in Appendix B, Table 2 to demonstrate compliance with the limits of 20.130)

(20.1302(b)(2)). Note that 20.1302(b)(2)(i) describes these as " annual average concentrations" as opposed to instantaneous limits. Measurements and calculation means are also allowed (20.1302(b)(1)). l l Radiological Effluent Control (REC) 9.6 is required by Technical Specification 6.8.4.e.7 to contain:

" Limitations on the dose rate resulting from radioactive material released in gaseous emuents to areas 3 beyond the SITE BOUNDARY conforming to the doses associated with 10CFR20, Appendix B, gI J Table 2, Column 1." 3 The bases for this Control state that its purpose is to ensure that the dose at any time from gaseous emuents is within the annual dose limit of 10CFR20, which is the dose associated with the concentrations of 10CFR20, Appendix B, Table 2, Column 1. Additionally, this Control provides assurance that the release of gaseous effluents will not result in the exposure of a Member of the Public to annual average 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 g the dose rate limit of 2 mrem /h specified in 20.1301(a)(2), and is therefore more restrictive. 3 He preamble to the revised rule states that demonstration of compliance with the limits of 40CFR190 and with 10CFR50, Appendix I is sufficient to demonstrate compliance with the 100 mrem dose limit of 20.1301(a)(1).

l Other Controls are provided as required Technical Specification 6.8.4.e (items 8,9, and 10) which ensure that the limits of 40CFR190 and 10CFR50, Appendix ! are not exceeded.

The Bases for this Control reference the concentration values of 10CFR20, Appendix B, Table 2, Column I as a basis for the specified dose rate limits. These values were derived using ICRP 30 calculation methodology and the dose and dose rate values they represent are the Total Effective Dose Equivalent (TEDE) which is the g summation of the extemal and internal dose components. Compliance with the Control is demonstrated through 3 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.

Page 10 of 12 APPENDlX A

APA-ZZ-01003 Rev. 6 i

ne 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 effluents:

10CFR20 APPENDIX B CONCENTRATION VALUES Nuclide Former Rule Revised Rule New/Old Kr-85 3E-7 pCi/ml 7E 7 pCi/ml 2.3 Xe-133 3E-7 5E-7 1.7 Xe-135 1E-7 7E-8 0.7 I-131 lE-10 2E-10 2.0 1-133 4E-10 lE-9 2.5 Co-58 2E 9 1E-9 0.5 Co-60 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 effluents for the past three years (1989-1991). The concentration value for Xe-133 actually increased in the revise rule, as did that for Kr-85 and both iodine nuclides. Although the Co-58 and Co-60 values did decrease in the revised rule, they are relatively insignificant contributors to the whole body and organ dose from gaseous effluents 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 Xe 135 0.04 0.7 Xe-133m 0.01 2.0 Kr 85m 0.01 1.0 Kr-85 0.01 2.3 Particulates and lodines:

I-131 0.72 2.0 I-133 0.11 2.5 Co-58 0.03 0.5 Co-60 0.14 0.2 He NRC states, in their response to Question 19, that until 10CFR50, Appendix 1 is changed, licensees must ,

continue to show compliance with Tech Specs in terms of organ and whole body doses as per Regulatory l

Guide 1.109. He 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. F 1 the dose calculation methodology has not been revised, it would be more conservative to continue to utihzi , 1 current REC values vice dose rate limits calculated from the revised 10CFR20, Appendix B values. )

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1 Page 11 of 12 APPENDIX A

1 APA-ZZ.01003 Rev. 6 Refer to the discussion of T/S 6.8.4.e.2 (REC 9.3) for additional details.

TECli SPEC 6.8.4.E.6, GASEOUS RADWASTE TREATMENT SYSTEM OPERABILITY (REC 9.9)

TECII SPEC 6.8.4.E.8, DOSE FROM NOBLE GASES (REC 9.7)

TECII SPEC 6.8.4.E.9, DOSE FROM IODINES AND PARTICULATES IN GASEOUS EFFLUENTS (REC 9.8)

TECH SPEC 6.8.4.E.10, TOTAL DOSE FROM THE URANIUM FUEL CYCLE (REC E'

3 9.10)

These specifications are derived from 10CFR50, Appendix ! and 40CFR190 and are not affected by the revised rule. Doses continue to be calculated in accordance with Regulatory Guide 1.109 which has not been revised.

No chanF s are anticipated for these specifications. i I! 1 I

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