ML041270279

APA-ZZ-01003, Rev 014, Callaway Plant Off-Site Dose Calculation Manual.

ML041270279
Person / Time
Site:	Callaway
Issue date:	06/27/2003
From:	Abernathy J, Kerrigan J Union Electric Co
To:	Office of Nuclear Reactor Regulation
References
APA-ZZ-01003, Rev 014
Download: ML041270279 (156)

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APA-ZZ-01003 s- Rev. 014 NION ELCTRIC so CALLAWAY PLANT OFF-SITE DOSE CALCULATION MANUAL JUNE 17, 2003

APA-ZZ-01003 Revision 014 June 17, 2003 CALLAWAY PLANT ADMINISTRATIVE PROCEDURE APA-ZZ-01003 OFF-SITE DOSE CALCULATION MANUAL RESPONSIBLE DEPARTMENT HEALTH PHYSICS PROCEDURE OWNER HAROLD OSBORN WRITTEN BY JOHN KERRIGAN PREPARED BY JERRY ABERNATHY APPROVED BY JAA - b4

{JUN 2 6 2003 DATE ISSUED t p-l -C) ACCOUNTABLE This procedure contains the following:

Pages 1 through 63 Attachments through Tables through Figures through Appendices through Checkoff Lists through This procedure has checkoff list(s) maintained in the mainframe computer.

Conversion of commitments to TRS reference/hidden text completed by Revision Number:

Non-T/S Commitments

APA-ZZ-01003 Rev. 014 RECORD OF REVISIONS I PURPOSE AND SCOPE I 2 LIQUID EFFLUENTS 2 2.1 LIQUID EFFLUENT MONITORS 2 2.2 CALCULATION OF LIQUID EFFLUENT MONITOR SETPOINTS 4 2.3 LIQUID EFFLUENT CONCENTRATION MEASUREMENTS 5 2.4 DOSE DUE TO LIQUID EFFLUENTS 6 2.4.1 The Maximum Exposed Individual 6 2.4.2 Calculation Of Dose From Liquid Effluents 6 2.4.3 Summary, Calculation Of Dose Due To Liquid Effluents 7 2.5 LIQUID RADWASTE TREATMENT SYSTEM 7 2.6 DOSE FACTORS 7 3 GASEOUS EFFLUENTS 12 3.1 GASEOUS EFFLUENT MONITORS 12 3.2 GASEOUS EFFLUENT MONITOR SETPOINTS 13 3.2.1 Total Body Dose Rate Setpoint Calculations 14 3.2.2 Skin Dose Rate Setpoint Calculation 15 3.3 CALCULATION OF DOSE AND DOSE RATE FROM GASEOUS EFFLUENTS 15 3.3.1.1 Dose Rate from Noble Gases 15 3.3.1.2 Dose Rate from Radionuclides Other than Noble Gases 16 3.3.2 Dose Due to Gaseous Effluents 17 3.3.2.1 Air Dose Due to Noble Gases 17 3.3.2.2 Dose Due to Radionuclides Other than Noble Gases 18 3.4 GASEOUS RADWASTE TREATMENT SYSTEM 19 3.5 DOSE FACTORS 19 4 DOSE AND DOSE COMMITMENT FROM URANIUM FUEL CYCLE SOURCES 43 4.1 CALCULATION OF DOSE AND DOSE COMMITMENT FROM URANIUM FUEL CYCLE SOURCES 43 4.1.1 Identification of the MEMBER OF THE PUBLIC 43 4.1.2 Total Dose to the Nearest Resident 43 4.1.3 Total Dose to the Critical Receptor Within the SITE BOUNDARY 44 5 RADIOLOGICAL ENVIRONMENTAL MONITORING 47

5.1 DESCRIPTION

OF THE RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 47 5.2 PERFORMANCE TESTING OF ENVIRONMENTAL THERMOLUMINESCENCE DOSIMETERS 47

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APA-ZZ-01003 Rev. 014 RECORD OF REVISIONS 6 DETERMINATION OF ANNUAL AVERAGE AND SHORT TERM ATMOSPHERIC DISPERSION PARAMETERS 47 6.1 ATMOSPHERIC DISPERSION PARAMETERS 47 6.1.1 Long-Tenn Dispersion Estimates 47 6.1.2 Determination of Long-Term Dispersion Estimates for Special Receptor Locations 48 6.1.3 Short Term Dispersion Estimates 48 6.2 ANNUAL METEOROLOGICAL DATA PROCESSING 49 7 REPORTING REQUIREMENTS 56 7.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (COMN 2804) 56 7.2 RADIOACTIVE EFFLUENT RELEASE REPORT (COMN 2805) 56 8 IMPLEMENTATION OF ODCM METHODOLOGY (COMN 2791) 57 9 RADIOACTIVE EFFLUENT CONTROLS (REC) 58 10 ADMINISTRATIVE CONTROLS 59 10.1 MAJOR CHANGES TO LIQUID AND GASEOUS RADWASTE TREATMENT SYSTEMS 59 10.2 CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL (ODCM) (COMN 2815) 59 11 REFERENCES 60 FIGURES FIGURE 4.1 ....... 46 APA-ZZ-0 1003 Rev. 014 RECORD OF REVISIONS TABLES TABLE 2.1 .8 TABLE 2.2 .11 TABLE 3.1 .20 TABLE 3.2 .21 TABLE 3.3 - INHALATION PATHWAY..............................................................................23 TABLE 3.3 - MEAT PATHWAY .25 TABLE 3.3 - GRASS-COW-MILK PATHWAY...................................................................27 TABLE 3.3 - GRASS-GOAT-MILK PATHWAY .29 TABLE 3.3 - VEGETATION PATHWAY .31 TABLE 3.4 - INHALATION PATHWAY..............................................................................33 TABLE 3.4 - MEAT PATHWAY...........................................................................................35 TABLE 3.4 - GRASS-COW-MILK PATHWAY .37 TABLE 3.4 - GRASS-GOAT-MILK PATHWAY .39 TABLE 3.4 - VEGETATION PATHWAY.............................................................................41 TABLE 6.1 .50 TABLE 6.2 .51 TABLE 6.3 .52 TABLE 6.4 .53 TABLE 6.5 .54 TABLE 6.6 .55

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APA-ZZ-0 1003 Rev. 014 RECORD OF REVISIONS Rev. No. 0 Date: March 1983 Rev. No. 1 Date: November, 1983 Revised to support the current RETS submittal and to incorporate NRC Staff comments.

Rev. No. 2 Date: March, 1984 Revised to incorporate NRC Staff comments Rev. No. 3 Date: June,1985 Revised to incorporate errata identified by ULNRC00803 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 Minor clarifications, revised descriptions of liquid and gaseous rad monitors, revised liquid setpoint methodology to incorporate monitor background, revised dose calculations for 40CFRI 90 requirements, Revised Table 6 and Figures 5. IA and 5.1 B to refine descriptions of environmental TLD stations, incorporated description of environmental TLD testing required byReg. Guide 4.13, revised Tables 1, 2, 4 and 5 to add additional nuclides, deleted redundant material from Chapter 6.

Rev. No. 6 Date: May, 1989 Revised methodology for calculating maximum permissible liquid effluent discharge rates and liquid effluent discharge rates and liquid effluent monitor setpoints, provided methodology for calculating liquid effluent monitors response correction factors, provided an enhanced description of controls on liquid monitor background limits, provided additional liquid 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 8842.

APA-ZZ-01003 Rev. No. 0 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 operating license, issued 4/10/87.

Rev. No. 2 Date: May, 1991 Section 2A.2 - Changed gross alpha analysis frequency from "each batch" to a monthly composite as per Table 9.3-A, and the Callaway Plant NPDES permit (reissued March 15, 1991).

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APA-ZZ-0 1003 Rev. 014 RECORD OF REVISIONS Rev. No. 3 Date: June, 1993 Deleted HF-RE-45 and LE-RE-59 as effluent monitors. Revised table numbering for consistency with those in Section 9.0, deleted redundant material, incorporated 1992 Land Use Census results, moved LLD description to , moved REC Bases to Attachment 2. Deleted reporting requirements for solid radwaste, which are described in APA-ZZ-01011, PROCESS CONTROL PROGRAM. Addressed compliance with 10 CFR 20.1301.

Revised the dilution flow rate to allow values other than 5000 gpm, based on dilution flow monitor setpoint. Revised "MPC" terminology to "ECV". Added Action 46 to REC 9.2 to clarify actions for inoperable mid and high range WRGM Channels. Revised references to be consistent with the revised 10 CFR 20. Added Appendix A. Revised Action 41 of Rec 9.2 and the operability requirements of GT-RE-22/33. Incorporated the revised Ri values in Tables 3.2 and 3.3. Added Section 6.2 and Table 6.5.

Rev. No. 4 Date: September, 1994 Increased the minimum channels OPERABLE requirement of REC 9.2 for GT-RE-22 & 33 from I channel to 2 channels. Revised Action 41 and the Bases for REC 9.2 accordingly. Incorporated the operability requirements from Tcch Spec 3.9.9 into the Action statement for clarity. (Refer to CARS 199401176).

Rev. No. 5 Date: February, 1995 Removed the REMP station locations. Removed particulate nuclides with a half-life of less than 8 days from Tables 3.2-3.4 and removed C14, p 32 , Ni63, TeI25m, and from Tables 2.1, 2.2,3.2,3.3, and 3.4. Changed the reporting frequency of the Effluent Release Report from semiannual to annual. Removed the meat, milk and vegetable pathway dispersion parameters from Tables 6.1, 6.2, and 6.3, and clarified the applicability of the dispersion parameters and dose locations in Table 6.4. Relocated REC 9.1 and 9.2 to the FSAR. Revised footnotes 3 and 7 of Table 16.11-4 to require additional sampling of the Unit Vent in the event of a reactor power transient, only if the Unit Vent noble gas activity increases by a factor of 3 or greater. Added Section 4.1.3.1.3 for determination of dose due to the on-site storage of low level radioactive waste.

Rev. No. 6 Date September, 1996 Section 2: 7 Added dose factors (A11)for Ag""'n, Npe , Pu23, Pu '2 , PU241 Am241 , Cm242 and Cm234 244 to 239 40 Table 2.1, and Bioaccumulation Factors (Bf) for Ag, Pu, Am, and Cm to Table 2.2 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- 8.

Section 3: Eliminated y 9 1m 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 3.3.1.2. Augmented the definition of qi in Section 3.3.2.1.

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 Effluent 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.

APA-ZZ-01003 Rev. 014 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) Revised Table 9.3-A to incorporate sampling and analysis requirements for TRU nuclides in liquid effluents; (3) Eliminated sampling of Fuel Building Exhaust from Table 16.11-4 and the associated footnotes due to redundancy with Unit Vent sampling; revised the continuous sampling requirements for the gaseous batch release points consistent with plant design; revised the H3 analysis frequency for Purges from weekly to "prior to each purge"; 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 CARS 199502280).

Revised the location requirements for milk and vegetables. Revised description of use of baseline samples to trigger gamma isotopic analysis in footnote #4, revised requirement for location of downstream sample station in footnote 16.

Revised Surveillance Requirement 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 I 1: Added reference 11.14.13. : Revised the Bases for REC 9.10 to support the elimination of liquid effluents from Surveillance 9.10.2.1.

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

(This revision implements CARS 199502055, CARS 199600167, CARS 199600961, CARS 199502280, and CARS 199600986).

Rev. No. 7 Date February, 1997 Section 9: (1) REC 9.5, "Liquid Radwaste Treatment System", Action statement: Eliminated reference to COMN 1161, (2) Table 9.11-A, items 4a (milk) and 4c (vegetation): revised to required control stations in the least prevalentwinddirection. (SecCARS 199700166)

Appendix A: revised the discussion relative to the appropriate gross alpha Effluent Concentration Value.

Rev. No. 8 Date May, 1997 Section 1: The Purpose and Scope was revised to describe the split of the ODCM into two sections per FSAR Change Notice 95-058. Section 2: Sections 2.2 and 2.3 were revised to clarify the use of nuclide- specific alpha activity vice gross alpha activity for setpoint determination. Section 2.5 was revised to delete the description of the Liquid Radwaste Treatment System. Section 6: Tables 6.1, 6.2, and 6.3 were revised to reflect the results of the 1996 Annual Land Use Census. Section 7: The reporting requirements for the Annual Radiological Environmental Operating Report and the Effluent Release Report were relocated to the FSAR per FSAR Change Notice 95-058.

Section 9: RECs and the supporting Attachments 1 and 2 were relocated to the ESAR per FSAR Change Notice 95-058. Appendix A: Appendix A was deleted. Editorial changes were made throughout the ODCM reflecting the relocation of the REC's to the FSAR.

Rev. No. 9 Date March, 1998 Section 2.5: Revised projected liquid dose calculation to use previous 31 day cumulative doses.

Section 3.1.1: Added GL-RE-202, Laundry Decon Facility Dryer Exhaust Monitor. Added action to be taken when the particulate and/or iodine grab sampler is not operable. Section 3.2: Added setpoint calculation for GL-RE-202.

Section 3.2.1 and 3.3.2.2: Changes were made to correct typographical errors and have no technical impact.

Section 3.4: Revised projected gas dose calculation to use previous 31 day cumulative doses.

Section 3.5: Removed the word secular from "secular equilibrium" since the equilibrium mode could be secular or transient depending on the isotope. Table 6.2: Added Laundry Decon Facility Dryer Exhaust to title of table since these will be the dispersion factors used for this release point.

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APA-ZZ-01003 Rev. 014 RECORD OF REVISIONS Rev. No. 10 Date December 20, 1999 Section 3.1: Added explanation that GL-RE-202 only monitors particulate. Section 3.2: Changed LaundryDecon Facility Exhaust Monitor setpoint to less than or equal to 2000 cpm above equilibrium background with a maximum allowed background of 2000 cpm as calculated in HPCI 99-05. Tables 6.1. 6.2. 6.3: Updated values as calculated in HPCI 99-02. Section 5.1: Defined how REMP sample locations were determined. Removed reference to Plant Operating manual since it no longer exists.

Rev. No. 11 Date December 22,1999 Changes required to go from old Technical Specifications to Improved Technical Specifications. Technical Specification 4.9.4.2 changed to FSAR 16.11.2.4.1. Technical Specification 6.8.4.F changed to FSAR 16.11.4.

Technical Specification 6.8.1.F changed to Improved Technical Specification 5.4.1. Technical Specification 6.14 changed to Improved Technical Specification 5.5.1. Technical Specification 6.8.4.E changed to Improved Technical Specification 5.5.4. Technical Specification 6.9.1.6 changed to Improved Technical Specification 5.6.2.

Technical Specification 6.9.1.7 changed to Improved Technical Specification 5.6.3. Changed name of Annual Radiological Effluent Release Report to Effluent Release Report as stated in ITS. Added liquid releases are limited to 10 times the Appendix B, Table 2, Column 2 limits per FSAR CN 98-041 supporting implementation of ITS.

Rev. No. 12 Date December 01, 2000 Section 2.1 and 2.2.1: Updated 10CFR20, Appendix B, Table 11, Column 2 reference to the new IOCFR20 format.

Corrected typo for "f', flow setpoint should be undiluted waste low rate. Section 3.2.1: Corrected typo, default value for safety factor should be 0.1. Section 5.1: Updated crosscheck program used to EML since EPA program is no longer available. Section 6.2: Added vertical height of highest adjacent building used to perform concurrent year annual average atmospheric dispersion (X/Q) calculations and reference for this value. This information should be documented in the ODCM. Added responsibility for validation of meteorological data, since responsibility has changed from engineering to HPTS. Section 10.1.1: Revised to require a summary of Major Radwaste System changes to be included in the annual report. This was done to be consistent with FSAR 16.11.5.2. Several changes were made throughout the procedure to correct typographical errors and have no technical impact.

Rev. No. 13 Date September 19, 2002 Section 3.2: Revised to implement the approved OL 1218, Rev. 1; License Amendment no. 152 allowing equipment hatch and emergency air lock to remain open during refueling activities (FSAR CN-01-030 and CN-02-049). The amendment eliminated FSAR 16.11.2.4.1B and subsequently deleted the core alteration setpoint value 5.0 E-3 PCi/cc for Containment Purge Monitors GT-RE-22 and GT-RE-33. The alarm setpoints for the Containment Purge Monitors will be based on the methodology described in Section 3 of the ODCM.

Rev. No. 14 Date June 17,2003 Revised Table 2.1 (ingestion Dose Commitment Factor for Adult Age Group) to include dose factors for Pr-144.

(CARS 200303251). Revised Section 4.1.3.1 to adjust the Farmer's residence (critical receptor) from 3830 meters in the SE sector to 2897 meters in the NNW sector. The Farmer's residence (critical receptor) was changed in 2002 to a location directly across the street from the Nearest residence. For conservatism and ease in calculation, Table 6.1 and 6.2 were revised making the distances and dispersion parameters for the Farmer's residence (critical receptor) and the Nearest residence the same. Revised section 7.2 to reference Table 6.6. Revised section 10.2.2 to remove the requirement for QA department review of the ODCM for reach revision (CARS 200304509). Added a reference to 11.14.14, Calculation HPCI 03-004 (Rev. 0), "Calculation of Liquid Effluent Dose Commitment factor for Pr-144 (Ait) for the Adult Age Group", June, 2003.

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APA-ZZ-0 1003 Rev. 014 OFF-SITE DOSE CALCULATION MANUAL 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 Alarm/Trip 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 TIS AC 5.5.4 and FSAR 16.11.4, and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Effluent Release Reports required by T/S AC 5.6.2 and T/S AC 5.6.3.

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)

The ODCM consists of two parts: FSAR Section 16.11 which contains the Radiological Effluent Controls (REC's), and APA-ZZ-01003, which contains the methodology and parameters used to implement the REC's.

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APA-ZZ-0 1003 Rev. 014 2 LIOUID EFFLUENTS 2.1 LIQUID EFFLUENT MONITORS Gross radioactivity monitors which provide for automatic termination of liquid effluent releases are present on the liquid effluent lines. Flow rate measurement devices are present on the liquid effluent lines and the discharge line (cooling tower blowdown). Setpoints, precautions, and limitations applicable to the operation of the Callaway Plant liquid effluent monitors are provided in the appropriate Plant Procedures. Setpoint values are calculated to assure that alarm and trip actions occur prior to exceeding ten times the Effluent Concentration Values (ECV) limits in 10 CFR Part 20 at the release point to the UNRESTRICTED AREA. The calculated alarm and trip action setpoints for the liquid effluent line monitors and flow measuring devices must satisfy the following equation:

F + f Where:

C = The liquid effluent concentration value (ECV) implementing REC 16.11.1.1 for the site in (OCi/ml).

c = The setpoint, in (pCi/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, which, if exceeded, would result in concentrations exceeding ten times the values of 10 CFR Part 20 Appendix B, Table 2, Column 2, in the UNRESTRICTED AREA.

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

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

(Ref. 11.8.1)

If no dilution is provided, then c < C.

The radioactive liquid waste stream is diluted by the plant discharge line prior to entry into the Missouri River. Normally, the dilution flow is obtained from the cooling tower 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.

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

c < f (pCi / ml) (2.2) f The 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 analysis. Due to the possibility of a simultaneous release from more than one release pathway, a portion of the total site release limit is allocated to each pathway. The determination and usage of the allocation factor is discussed in Section 2.2. In the event the alarm/trip setpoint is reached, an evaluation will be performed using actual dilution and effluent flow values and actual isotopic analysis to ensure that REC 16.11.1.1 limits were not exceeded.

APA-ZZ-0 1003 Rev. 014 2.1.1 Continuous Liquid Effluent Monitors The radiation detection monitor associated with continuous liquid effluent releases is (Ref. 11.6.1, 11.6.2):

Monitor l.D. Description BM-RE-52 Steam Generator Blowdown Discharge Monitor The Steam Generator Blowdown discharge is not considered to be radioactive unless radioactivity has been detected by the associated effluent radiation monitor or by laboratory analysis. The sampling frequency, minimum analysis frequency, and type of analysis performed are as per FSAR TABLE 16.11-1.

2.1.2 Radioactive Liquid Batch Release Effluent Monitors The radiation monitor which is associated with the liquid effluent batch release system is (Ref. 11.6.4):

Monitor T.D. Description HB-RE-1 8 Liquid Radwaste Discharge Monitor This effluent stream is normally considered to be radioactive. The sampling frequency, minimum analysis frequency, and the type of analysis performed are as per FSAR TABLE 16.11-1.

APA-ZZ-01003 Rev. 014 2.2 CALCULATION OF LIQUID EFFLUENT MONITOR SETPOINTS The dependence of the setpoint (c), on the radionuclide distribution, yields, calibration, and monitor 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 of the release(s) being considered must be determined. This is obtained from the analyses required per Table 16.1 1-1, and is used to calculate an ECV sum (ECVSUM):

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

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

Where:

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

Ca = the concentration of Np 7, Pu23 8, Pu2392 4 0, pu242, Am24t , Cm242, & CM2431 2", in the quarterly composite sample.

Cs = the measured concentrations of Sr-89 and Sr-90 as determined by analysis of the quarterly composite sample.

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

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

ECVg, ECVS, ECVa, ECVf, ECVt = are ten times the limiting concentrations of the appropriate radionuclides from 10 CFR 20, Appendix B, Table 2, Column 2. For dissolved or entrained noble gases, the concentration shall be limited to 2x10 4 liCi/mI total activity.

For the case ECVSUM *1, the monitor tank effluent concentration meets the limits of REC 16.1 1.1.1 without dilution and the effluent may be released at any desired flow rate. If ECVSUM > 1 then dilution is required to ensure compliance with the concentration limits REC 16.1 1.1.1. If simultaneous releases are occurring or are anticipated, an allocation fraction, N, must be applied so that available dilution flow may be apportioned among simultaneous discharge pathways. The value of N may be any value between 0 and I for a particular discharge point, provided that the sum of the allocation fractions for all discharge points must be *1.

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

imax 5 (F + fp) (SF) (N) + (ECVSUM) (2.4)

Where:

fmax = maximum permissible liquid effluent discharge flowrate, (ingallons/minute);

p = 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 fluctuations and errors of measurements. This factor also provides amargin of safety in the calculation of the maximum liquid effluent discharge flowrate (fmax). The value of SF should be < 1.

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 Imax.

• Values for these concentrations are based on previous composite sample analyses as required by FSAR Table 16.1 1-1.

APA-ZZ-0 1003 Rev. 014 2.2.3 Calculation Of Liquid Effluent Monitor Setpoint The liquid effluent monitors are Nal(TI) based systems and respond primarily to gamma radiation. Accordingly, their setpoint is based on the total concentration of gamma emitting nuclides in the effluent:

c = BKG + ((Cg) + SF) iCiCml (2.5)

Where:

c = the monitor setpoint as previously defined, in (ICiVml);

BKG = the monitor background prior to discharge, in (pCi/ml);

Cg and SF are as previously defined.

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

Utilizing the methodology of ANSI NI3.10-1974 (Ref. 11.21), the background must be maintained at a value of less than or equal to 9E-6 ,Ci/ml (relative to Cs-137) in order to detect a change of 4E-7 p[Ci/ml of Cs-1 37. (Ref. 11.25).

In the event that there is no detectable gamma activity in the effluent or if the value of (j(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.

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 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 Technology (NIST) traceable calibration sources:

DBCFC = ( j(C

)) (CMR) x (ECF) (2.6)

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

CMR = the calculated response of the radiation monitor to the liquid effluent; ECF = the conversion factor for Cs-I 37, which converts count rate into concentration (pCi/ml).

Cg is as previously defined.

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

2.3 LIQUID EFFLUENT CONCENTRATION MEASUREMENTS Liquid batch releases are discharged as a discrete volume and each release is authorized based upon the sample analysis and the dilution flow rate existing in the discharge line at the time of release. To assure representative sampling, each liquid monitor tank is isolated and thoroughly mixed by recirculation of tank contents prior to sample collection. The methods for mixing, sampling, and analyzing each batch are outlined in applicable plant procedures. The allowable release rate limit is calculated for each batch based upon the pre-release analysis, dilution flow-rate, and other procedural conditions, prior to authorization for release. The liquid 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 gamma0 isotopic 24 and 241 H-3 analyses 242 of243 the t 4 liquid batch sample. For SrS9 , S90, Fe", Np237, Pu238, PuP , PU 1,Am , Cm , & Cm n , the measured concentrations from the previous quarterly composite analyses are used until laboratory results become available.

Composite samples are collected for each batch release and analyzed in accordance with FSAR TABLE 16.11-1. The dose from liquids discharged as continuous releases is calculated by utilizing the last measured values of samples in accordance with FSAR TABLE 16.11-1.

APA-ZZ-01003 Rev. 014 2.4 DOSE DUE TO LIQUID EFFLUENTS 2.4.1 The Maximum Exposed Individual

)

The cumulative dose determination considers the dose contributions from the maximum exposed individual's consumption of fish and potable water, as appropriate. Normally, the adult is considered to be the maximum exposed individual. (Ref. 11.8.3)

The Callaway Plant's liquid effluents are discharged to the Missouri River. As there are no potable water intakes within 50 miles of the discharge point (Ref. 11.7.1, 11.6.6), this pathway does not require routine evaluation. Therefore, the dose contribution from fish consumption is expected to account for more than 95% of the total man-rem dose from discharges to the Missouri River. Dose from recreational activities is expected to contribute the additional 5%,

which is considered to be negligible. (Ref. 11.6.7) 2.4.2 Calculation Of Dose From Liquid Effluents The dose contributions for the total time period.

m YAt, 1-1 are calculated at least once each 31 days and a cumulative summation of the total body and individual organ doses is maintained for each calendar quarter. Dose is calculated for all radionuclides identified in liquid effluents released to UNRESTRICTED AREAS using the following expression (Ref. 11.8.3):

D = Air At, Ci, F1 ] (2.12)

Where:

D = the cumulative dose commitment to the total body or any organ, -r, from the liquid effluents for the total period At in mrem.

Atl = the length of the I the time period over which C; , and F , are averaged for all liquid releases, in hours. At l corresponds to the actual duration of the release(s).

Cl = the average measured concentration of radionuclide, i, in undiluted liquid effluent during time period At, from any liquid release, in (pCi/mil).

Ai = the site related ingestion dose commitment factor to the total body or any organ X for each identified principal alpha, gamma and beta emitter listed in FSAR TABLE 16.11-1, (in mrem/hr) per (pCi/mil). The calculation of the Aix values given in Table 2.1 are detailed in Ref. 11.14.12, 11.14.13, 11.14.14, and 11.14.5.

Ft = the near field average dilution factor for C; , during any liquid effluent release:

F, fma -

(F+fm. )89.77 APA-ZZ-0 1003 Rev. 014 Where:

fmax = maximum undiluted effluent flow rate during the release F = average dilution flow 89.77 = site specific applicable factor for the mixing effect of the discharge structure.

(Ref. 11.5.1)

The term C; 1 is the undiluted concentration of radioactive material in liquid waste at the common release point determined in accordance with REC 16.11.1.1, Table 16.11-1, "Radioactive Liquid Waste Sampling and Analysis Program". All dilution factors beyond the sample point(s) are included in the F , term.

The nearest municipal potable water intake downstream from the liquid effluent discharge point into the Missouri River is located 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. Should 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. 11.6.6).

2.4.3 SummarM. Calculation Of Dose Due To Liquid Effluents The dose contribution for the total time period m

ZAt, 1-1 is determined by calculation at least once per 31 days and a cumulative summation of the total body and organ doses is maintained for each calendar quarter. The projected dose contribution from 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 in the dose calculations.

2.5 LIQUID RADWASTE TREATMENT SYSTEM The LIQUID RADWASTE TREATMENT SYSTEM is described in Chapter FSAR 11.2 of the Callaway Plant FSAR.

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 16.11.1.1. and 16.11.1.2.

Projected doses due to liquid releases to UNRESTRICTED AREAS are determined each 31 days. The prior 31 day period is used to calculate compliance. 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, Table 2.2 and other sources as necessary (Ref: 11.14.5, 11.14.13 and 11.14.12) Non-gamma emitting nuclides not listed in FSAR TABLE 16.11-1 are not considered.

APA-ZZ-0 1003 Rev. 014 TABLE 2.1 INGESTION DOSE COMMITMENT FACTOR (Ail) FOR ADULT AGE GROUP (mremlhr) per (giCi/ml)

Nmfidk Bone Live Total Thyroid Kidin 1 1 H-3 No Data 2.26E-01 2.26E-01 2.26E-01 2.26E-0I 2.26E-01 2.26E-01 Be-7 1.30E-02 2.98E-02 1.45E-02 No Data 3.15E-02 No Data 5.16E+00 Na-24 4.07E+02 4.07E+02 4.07E+02 4.07E+02 4.07E+02 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 6.81 E+02 8.13E+03 Co-57 No Data 2.09E+01 3.48E+01 No Data No Data No Data 5.3 1E+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 Zn-69 4.93E+01 9.44E+01 6.56E+00 No Data 6.13E+01 No Data 1.42E+O1 Br-82 No Data No Data 2.27E+03 No Data No Data No Data 2.60E+03 Br-83 No Data No Data 4.04E+01 No Data No Data No Data 5.8 1E+01 Br-84 No Data No Data 5.26E+01 No Data No Data No Data 4.13E-04 Br-85 No Data No Data 2.15E+00 No Data No Data No Data 0 Rb-86 No Data 1.01E+05 4.71E+04 No Data No Data No Data 1.99E+04 Rb-88 No Data 2.90E+02 1.54E+02 No Data No Data No Data 4.00E-09 Rb-89 No Data 1.92E+02 1.35E+02 No Data No Data No Data 0 Sr-89 2.21E+04 No Data 6.35E+02 No Data No Data No Data 3.55E+03 Sr-90 5.44E+05 No Data 1.34E+05 No Data No Data No Data 1.57E+04 Sr-91 4.07E+02 No Data 1.64E+01 No Data No Data No Data 1.94E+03 Sr-92 1.54E+02 No Data 6.68E+00 No Data No Data No Data 3.066E+03 Y-90 5.75E-01 No Data 1.54E-02. No Data No Data No Data 6.10E+03 Y-91M 5.44E-03 No Data 2.10E-04 No Data No Data No Data 1.60E-02 Y-91 8.43E+00 No Data 2.25E-01 No Data No Data No Data 4.64E+03 Y-92 5.05E-02 No Data 1.48E-03 No Data No Data No Data 8.85E+02 Y-93 1.60E-0 I No Data 4.42E-03 No Data No Data No Data 5.08E+03 Zr-95 2.40E-01 7.70E-02 5.21E-02 No Data 1.21 E-0 1 No Data 2.44E+02 Zr-97 1.33E-02 2.68E-03 1.22E-03 No Data 4.04E-03 No Data 8.30E+02 Nb-95 4.47E+02 2.48E+02 1.34E+02 No Data 2.46E+02 No Data 1.51 E+06 Mo-99 No Data 1.03E+02 1.96E+01 No Data 2.33E+02 No Data 2.39E+02 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 NoData 2.36E-01 6.70E-03 0 APA-ZZ-0 1003 Rev. 014 TABLE 2.1 (Cont'd)

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

Nuclide Bnau Liver Total Thyroid Kidn LAM aLLLI Ru-103 4.42E+00 No Data 1.90E+00 No Data 1.69E+01 No Data 5.17E+02 Ru-105 3.68E-01 No Data IA5E-01 No Data 4.76E+00 No Data 2.25E+02 Ru-106 6.57E+01 No Data 8.32E+00 No Data 1.27E+02 No Data 4.251E+03 Cd-109 No Data 5.54E+02 1.94E+01 No Data 5.31E-+02 No Data 5.59E+03 Ag-110m 8.83E-01 8.17E-01 4.85E-01 No Data 1.61E+00 No Data 3.33E+02 Sn-I 13 5.66E+04 1.61E+03 3.26E+03 9.18E+02 No Data No Data 1.69E+05 Sb-124 6.69E+00 1.26E-01 2.65E+00 1.62E-02 No Data 5.21E+00 1.90E+02 Sb-125 4.28E+00 4.78E-02 1.02E+00 4.35E-03 No Data 3.30E+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 1.1OE+04 4.1 IE+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 2.3 1E+OI 1.26E+02 No Data 2.27E+01 Te-131M 1.66E+03 8.09E+02 6.75E+02 1.28E+03 8.2 1E+03 No Data 8.03E+04 Te-1 31 1.89E+01 7.88E+00 5.96E+00 I.55E+01 8.25E+01 No Data 2.67E+00 Te-132 2.41E+03 1.56E+03 1.47E+03 1.72E+03 1.50E+04 No Data 7.38E+04 1-130 2.71E+O1 8.01E+01 3.16E+01 6.79E+03 1.25E+02 No Data 6.89E+01 1-131 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 6.82E+00 6.82E+02 3.1 1E+OI 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 1.79E+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 7.09E+05 5.80E+05 No Data 2.29E+05 7.62E+04 1.24E+04 Cs-136 3.12E+04 1.23E+05 8.86E+04 No Data 6.85E+04 9.39E+03 1.40E+04 Cs-137 3.82E+05 5.22E+05 3.42E+05 No Data 1.77E+05 5.89E+04 1.01E+04 Cs-138 2.64E+02 5.22E+02 2.59E+02 No Data 3.84E+02 3.79E+0 I 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 8.3 1E-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 Ce-141 2.24E-02 1.51E-02 1.72E-03 No Data 7.03E-03 No Data 5.78E+01 Ce-143 3.94E-03 2.92E+00 3.23E-04 No Data 1.28E-03 No Data 1.09E+02 Ce-144 1.17E+00 4.88E-01 6.26E-02 No Data 2.89E-01 No Data 3.94E+02 Pr-143 5.50E-01 2.2 1E-01 2.73E-02 No Data 1.27E-01 No Data 2.4 1E+03 Pr-144 1.80E-03 7.47E-04 9.14E-05 No Data 4.21E-04 No Data 2.59E-10 Nd-147 3.76E-01 4.35E-01 2.60E-02 No Data 2.54E-0I No Data 2.09E+03 Eu-154 3.67E+01 4.52E+00 .3.21E+00 No Data 2.16E+01 No Data 3.27E+03 Hf-181 3.99E-02 1.94E-01 1.80E-02 No Data 4.17E-02 No Data 2.21E+02 W-187 2.96E+02 2.47E+02 8.64E+01 No Data No Data No Data 8.09E+04 APA-ZZ-0 1003

'A Rev. 014 TABLE 2.1 (Cont'd)

INGESTION DOSE COMMITMENT FACTOR (AiT) FOR ADULT AGE GROUP (mrem/vhr) per (pCVml)

Nucdid~e Bo= Total Thyroid Ken 110f 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 2.801E-03 1.54E-03 No Data 8.72E-03 No Data 5.74E+02 Pu-238 5.69E+03 8.OI1E+02 1.43E+02 No Data 6.12E+02 No Data 6.1 IE+02 Pu-239* 6.58E+03 8.87E+02 1.60E+02 No Data 6.78E+02 No Data 5.67E+02 Pu-241 1.38E+02 7.061E+00 2.78E+00 No Data 1.28E+01 No Data 1.17E+ 01 Am-241 4.89E+04 1.72E+04 3.23E+03 No Data 2.43E+04 No Data 4.43E+03 Cm-242 1.23E+03 1.25E+03 8.19E+0I No Data 3.72E+02 No Data 4.73E+03 Cm-243* 3.82E+04 1.44E+04 2.24E+03 No Data 1.05E+04 No Data 4.67E+03

• Includes Pu-240 contribution "Includes Cm-244 contribution APA-ZZ-01003 Rev. 014 TABLE 2.2 BIOACCUMULATION FACTOR (Bfi)(a (pCi/kg per (pCi/liter)

Bf; Element Fish (Freshwater)

H 9.0E-01 Be 2.0E+00 Na 1.0E+02 Cr 2.0E+02 Mn 4.0E+02 Fe 1.0E+02 Co 5.0E+01 Ni 1.0E+02 Cu 5.0E+01 Zn 2.0E+03 Br 4.2E+02 Rb 2.0E+03 Sr 3.0E+01 Y 2.5E+01 Zr 3.3E+00 Nb 3.0E+04 Mo 1.OE+0I Tc 1.5E+01 Ru 1.0E+01 Rh 1.0E+01 Ag 2.3E+00 Cd 2.0E+02 Sn 3.0E+03 Sb 1.0E+00 Te 4.0E+02 1 1.5 E +01 Cs 2.0E+03 Ba 4.0E+00 La 2.5E+01 Ce 1.0E+00 Pr 2.5E+01 Nd 2.5 E +01 Eu 2.5E+01 Hf 3.3E+00 W 1.2E+03 Np 1.0E+01 Pu 3.5E+00 Am 2.5E+01 Cm 2.5E+01

° Values from Regulatory Guide 1.109, Rev. 1, Table A-1 and References 11.14.4, 11.14.8, and 11.14.13.

APA-ZZ-01003 Rev. 014 3 GASEOUS 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.

The alarm/trip (alarm & trip) setpoint for any gaseous effluent radiation monitor is determined based on the instantaneous noble gas total body and skin dose rate limits of REC 16.11.2.1, at the SITE BOUNDARY location with the highest annual average X/Q value.

Each gaseous monitor channel is provided with a two level system which provides sequential alarms on increasing radioactivity levels. These setpoints are designated as alert setpoints and alarm/trip setpoints. (Ref. 11.6.3)

The radiation monitor alarm/trip setpoints for each release point are based on the radioactive noble gases in gaseous effluents. It is not considered practicable to apply instantaneous alarm/trip setpoints to integrating radiation monitors sensitive to radioiodines, radioactive materials in particulate form and radionuclides other than noble gases. The exception is GL-RE-202. The only effluent released from the Laundry Decon Facility Dryer Exhaust is in the particulate form.

Conservative assumptions may be necessary in establishing setpoints to account for system variables, such as the measurement system efficiency and detection capabilities during normal, anticipated, and unusual operating conditions, the variability in release flow and principal radionuclides, and the time lag between alarm/trip action and the final isolation of the radioactive effluent. (Ref. 11.8.5) FSAR Table 16.11-6 provides the instrument surveillance requirements, such as calibration, source checking, functional testing, and channel checking.

3.1.1 Continuous Release Gaseous Effluent Monitors The radiation detection monitors associated with continuous gaseous effluent releases are (Ref. 11.6.8, 11.6.9):

Monitor I.D. Description GT-RE-2 1 Unit Vent GH-RE-10 Radwaste Building Vent GL-RE-202 Laundry Decon Facility Dryer Exhaust Monitor Each of the above continuously monitors gaseous radioactivity concentrations downstream of the last point of potential influent, and therefore measures effluents and not inplant concentrations.

The unit vent monitor continuously monitors the effluent from the unit vent for gaseous radioactivity. The unit vent, via ventilation exhaust systems, continuously purges various tanks and sumps normally containing low-level radioactive aerated liquids that can potentially generate airborne activity. 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 discharge.

The unit vent monitor provides alarm functions only, and does not terminate releases from the unit vent.

The Radwaste Building ventilation effluent monitor continuously monitors for gaseous radioactivity in the effluent duct downstream of the exhaust filter and fans. The flow path provides ventilation exhaust for all parts of the building structure and components within the building and provides a discharge path for the waste gas decay tank release line. These components represent potential sources for the release of gaseous and air particulate and iodine activities in addition to the drainage sumps, 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-0 1003 Rev. 014 The Laundry Decon Facility Dryer Exhaust Monitor continuously monitors the effluent of the dryer exhaust for particulate radioactivity during operation of the dryers. This effluent point is designed to release an insignificant quantity of radioactivity. The items to be placed in the dryers are typically washed before drying removing most of the radioactive material. The dryer effluent then passes through a HEPA filter before being sampled and released.

The Laundry Decon Facility Dryer Exhaust Monitor will secure the dryers and exhaust fans and isolate the dryer effluent upon a high radioactivity alarm or for a monitor failure.

The continuous Unit Vent and Radwaste Building Vent gaseous effluent monitor setpoints are established using the methodology described in Section 3.2. Since there are two continuous gaseous effluent 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. These 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.

For a loss of all isokinetic sampling and/or all heat tracing for the Unit Vent or Radwaste Building Vent grab samplers, one hour is allowed to restore a sampler to service. If sampling cannot be restored within one hour, all batch releases and ventilation not required for the operation of the plant should be secured. The best available sampling should be maintained during this period and normal sampling returned to service as soon as possible.

3.1.2 Batch Release Gaseous Monitors The radiation monitors associated with batch release gaseous effluents are (Ref. 11.6.9, 11.6.10, 11.6.11):

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

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 16.11.2.1 are not exceeded.

3.2 GASEOUS EFFLUENT MONITOR SETPOINTS The alarm/trip setpoint for Unit Vent and Radwaste Building Vent 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.

The alarm trip setpoint for the Laundry Decon Facility Exhaust Monitor is set to less then or equal to 2,000 cpm above equilibrium background. The maximum allowed background is 2,000 cpm as discussed in reference 11.27.

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APA-ZZ-O 1003 Rev. 014 3.2.1 Total Body Dose Rate Setpoint Calculations To ensure that the limits of REC 16.11.2.1 are met, ihe alarm/trip setpoint based on the total body dose rate is calculated according to:

Stb < DtbRtbFaF (3.1)

Where:

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

Dtb = REC 16.11.2.1 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 = 0. 1.

Fa the allocation factor which will modify the required dilution factor such that simultaneous gaseous releases may be made without exceeding the limits of REC 16.11.2.1.

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

R = C [(iQ) KI Qi] (3.2)

Where:

C = monitor reading of a noble gas monitor corresponding to the sample radionuclide concentrations for the release. Concentrations are determined in accordance with FSAR TABLE 16.114. The mixture of radionuclides determined via grab sampling of the effluent stream or source is correlated to a calibration factor to determine monitor response. The monitor response is based on concentrations, not release rate, and is in units of (pCi/cc).

X/Q = the highest calculated annual average relative concentration for any area at or beyond the SITE BOUNDARY in (sec/m 3 ). Refer to Tables 6.1, 6.2 and 6.4.

Ki = the total body dose factor due to gamma emissions for each identified noble gas radionuclide, in (mrem/yr) per (ptCi/m 3 ). (Table 3.1)

Qi = rate of release of noble gas radionuclide, i, in (pCi/sec).

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

APA-ZZ-0 1003 Rev. 014 3.2.2 Skin Dose Rate Setpoint Calculation To ensure that the limits of REC 16.11.2.1 are met, the alarm/trip setpoint based on the skin dose rate is calculated according to:

Ss < D.R,FF. (3-3)

Where:

Fs and Fa are as previously defined.

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

Ds = REC 16.11.2.1 limit of 3000 mrem/yr, conservatively interpreted as a continuous release over a one year period.

Rs = 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:

Rs =C-(RQ E: (L, +.1.1M; Q1 (3.4)

Where:

Li = the skin dose factor due to beta emissions for each identified noble gas radionuclide, in (mrem/yr) per (pCi/m 3).

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

M = the air dose factor due to gamma emissions for each identified noble gas radionuclide, in (mrad/yr) per (tCi/m 3 ).

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

3.3 CALCULATION OF DOSE AND DOSE RATE FROM GASEOUS EFFLUENTS 3.3.1 Dose Rate from Gaseous Effluents The following methodology is applicable to the location (SITE BOUNDARY or beyond) characterized by the values of the parameter (X/Q) which results in the maximum total body or skin dose rate. In the event that the analysis indicates a different location for the total body and skin dose limitations, the location selected for consideration is that which minimizes the allowable release values. (Ref. 11.8.6)

The factors Ki, Li, and M; relate the radionuclide airborne concentrations to various dose rates, assuming a semi-infinite cloud model.

3.3.1.1 Dose Rate from Noble Gases The release rate limit for noble gases is determined according to the following general relationships (Ref. 11.8.6):

D =Z[K,((XIQ)Q,)] S00mrem/yr (3.5)

DL=E(L, +1.1 M,)((XIQ)Q,)]

• 3000 mrem/yr (3.6)

Where:

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

1.1 = Units conversion factor; 1 mrad air dose = 1.1 mrem skin dose.

L;, M;, K;, (X / Q), DLb & D, are as previously identified.

- 15 -

APA-ZZ-01003 Rev. 014 3.3.1.2 Dose Rate from Radionuclides Other than Noble Gases The release rate limit for Iodine-131 and Iodine-133, for tritium, and for all radioactive materials in particulate form with half lives greater than 8 days is determined according to (Ref. 11.8.7):

D0 = Rj XQ Qi

• 1500 mrem/yr R (3.7)

Where:

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

R = 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/m 3 ).

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

(X I Q) is as previously defined.

The dose parameter (Ri) includes the internal dosimetry of radionuclide, i, and the receptor's breathing rate, which are functions of the receptor's age. The child age group has been selected as the limiting age group. All radiodines are assumed to be released in elemental form (ref. 11.8.7).

Ri values were calculated according to (Ref. 11.8.8):

R1 = K (BR) DFA1 (3.8)

Where:

K' = Units conversion factor: IE06 pCi/pCi BR = The breathing rate. (Regulatory Guide 1.109, Table E-5).

DFAi = The maximum organ inhalation dose factor for the ith radionuclide, in (mrem/pCi).

The total body is considered as an organ in the selection of DFA;. (Ref. 11.1 1.5 and 11.14.4)

The 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.

APA-ZZ-01003 Rev. 014 3.3.2 Dose Due to Gaseous Effluents 3.3.2.1 Air Dose Due to 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:

Dg = 3.17E -08 ((M, (xTQ)7 qj] *5 mrad (3.9)

During any calendar quarter, for beta radiation:

Db = 3.17E -08 [N, (XIQ) q,1J *l0mrad 1 (3.10)

During any calendar year, for gamma radiation:

Dg = 3.17E-08 Z Mi (X4-/Q) qj *lOmrad (3.11)

During any calendar year, for beta radiation:

Db = 3.17E -08  : [N, (X/ Q) q, JS20 mrad (3.12)

Where:

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.

N = The air dose factor due to beta emissions for each identified noble gas radionuclide, i, in (mrad/yr) per (,iCifm3 ).

q; = 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. qj is calculated as the product of the ventilation flow rate and the measured activity of the effluent stream as determined by sampling.

3.117E-08 = The inverse of the number of seconds per year.

X / Q & Mi are as previously defined.

APA-ZZ-0 1003 Rev. 014 3.3.2.2 Dose Due to Radionuclides Other than Noble Gases The dose to a MEMBER OF THE PUBLIC from Iodine-131 and 133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released to areas at and beyond the SITE BOUNDARY, is calculated according to the following expressions:

During any calendar quarter:

Zj Dij

• 7.5 mrem (3.13)

During any calendar year:

Ej Dij S 15 mrem (3.14)

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

Dij = 3.17E-8 E, R~jj [Wj qj] (3.15)

Where:

Dij = Dose in mrem, to a MEMBER OF THE PUBLIC from radionuclides other than noble gases, from pathway j, received by organ I (including total body).

R j.y = The dose factor for each identified radionuclide, i, in m2 (mrem/yr) per (pCi/sec) or (mrem/yr) per (iiCi/m3 ) as appropriate, for the pathway j, and exposed organ 1, appropriate to the age group of the critical MEMBER OF THE PUBLIC receptor.

(xIQ

= for the inhalation and tritium pathways, in (sec/mi3 ). Refer to Tables 6.1, 6.2, and 6.4 for applicability.

W = (DQ)for the food and ground plane pathways, in (meters72). 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.

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

qj 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 R Ijj values are used. If no real pathway exists within 5 miles from the center of the building complex, the cow-milk RjiJ 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 Iij for an existing pathway within 5 miles is less than a cow-milk R j at 4.5 to 5.0 miles, then the value of the cow-milk R 1,ij at 4.5 to 5.0 miles is used. (Ref. 11.8.9)

Although the annual average relative concentration(X / Q) and the average relative deposition rate (1) 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 ground plane pathways to coexist at the location of the nearest residence with the highest value of (X7/Q) . (Ref. 11.8.9) The Total Body dose from ground plane deposition is added to the dose for each individual organ. (Ref. 11.11.3)

APA-ZZ-0 1003 Rev. 014 3.4 GASEOUS RADWASTE TREATMENT SYSTEM Tlie 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 discharge. These systems will provide reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept ALARA.

The OPERABILITY of the gaseous radwaste treatment system ensures this system will be available for use when gases require treatment prior to their release to the environment.

OPERABILITY is demonstrated through compliance with REC 16.11.2.1, 16.11.2.2, and 16.11.2.3.

Projected doses (gamma air, beta air, and organ dose) due to gaseous effluents at or beyond the SITE BOUNDARY are determined each 31 days. The prior 31 day period is used to calculate compliance. 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 of less than 8 days are not considered (Ref: 11.24). Y-90, La-140, and Pr-144 are included because the parent half-life is greater than 8 days, and equilibrium is assumed. Non-gamma emitting nuclides not listed in FSAR TABLE 16.114 are also not considered. (COMN 43121)

APA-ZZ-G -3 Rev. 014 TABLE 3.1 DOSE FACTOR FOR EXPOSURE TO A SEMI-INFINITE CLOUD OF NOBLE GASES Total Body Gamma Air Beta Air Dose Factor Skin Dose Factor Dose Factor Dose Factor L i Mi Ni Radionuclide (mrem/yr) per (pCi/m 3 ) (mrad/yr) per (,uCi/m3) (mrad/yr) per (,uCi/m3) (mrad/yr per (pCi/m 3 )

Ei- *I 2;: S 12a.i:U5rz'7I ___ ___-

Kr-83m 7.56 E-02 1.93 E+01 2.88 E+02 Kr-85m 1.17E+03 1.46E+03 1.23 E+03 1.97 E+03 Kr-85 1.61 E+01 1.34 E+03 1.72 E+01 1.95 E+03 Kr-87 5.92 E+03 9.73 E+03 6.17 E+03 1.03 E+04 Kr-88 1.47 E+04 2.37 E+03 1.52 E+04 2.93 E+03 Kr-89 1.66 E+04 1.01 E+04 1.73 E+04 1.06 E+04 Kr-90 1.56 E+04 7.29 E+03 1.63 E+04 7.83 E+03 Xe-131m 9.15 E+01 4.76 E+02 1.56 E+02 1.11 E+03 Xe-133m 2.51 E+02 9.94 E+02 3.27 E+02 1.48 E+03 Xe-133 2.94 E+02 3.06 E+02 3.53 E+02 1.05 E+03 Xe-135m 3.12 E+03 7.11 E+02 3.36 E+03 7.39 E+02 Xe-135 1.81 E+03 1.86 E+03 1.92 E+03 2.46 E+03 Xe-137 1.42 E+03 1.22 E+04 1.51 E+03 1.27 E+04 Xe-138 8.83 E+03 4.13 E+03 9.21 E+03 4.75 E+03 Ar-41 8.84 E+03 2.69 E+03 9.30 E+03 3.28 E+03 20

APA-ZZ-0 1003

- Rev. 014 TABLE 3.2 PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Ground Plane Pathway (m 2 mrem/yr) per (tCi/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 1.63E+09 Fe-59 2.73E+08 3.21E+08 Co-57 2.98E+08 4.37E+08 Co-58 3.79E+08 4.44E+08 Co-60 2.15E+10 2.53E+10 Zn-65 7.47E+08 8.59E+08 Rb-86 8.99E+06 1.03E+07 Sr-89 2.16E+04 2.5 IE+04 Y-90 5.36E+06 6.32E+06 Y-91 1.07E+06 1.21E+06 Zr-95 2A5E+08 2.84E+08 Nb-95 2.50E+08 2.94E+08 Ru-103 1.08E+08 1.26E+08 Ru-106 4.22E+08 5.07E+08 Ag-110m 3.44E+09 4.01E+09 Cd-109 3.76E+07 1.54E+08 Sn-1 13 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 2.3 IE+07 21

APA-ZZ-01003 Rev. 014 TABLE 3.2 PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Ground Plane Pathway (m2 mrem/yr) per (pCi/sec)

NUCLIDE TOTAL BODY SKIN 1-130 5.5 1E+06 6.69E+06 1-131 1.72E+07 2.09E+07 1-132 1.25E+06 1.47E+06 1-133 2.45E+06 2.98E+06 1-134 4.47E+05 5.3 IE+05 I-135 2.53E+06 2.95E+06 Cs-134 6.85E+09 8.OOE+09 Cs-136 1.51E+08 1.7 1E+08 Cs-137 1.03E+10 1.20E+10 Ba-140 2.05E+07 2.35E+07 La-140 1.47E+08 1.66E+08 Ce-141 1.37E+07 1.54E+07 Ce-144 6.96E+07 8.04E+07 Pr-144 4.35E+07 5.00E+07 Nd-147 8.39E+06 1.OlE+07 Eu- 154 2.21E+10 3.15E+10

) Hf-181 1.97E+08 2.82E+08 22

APA-ZZ-0 1003 Rev. 014 TABLE 3.3 CHILD PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Inhalation Pathway (mrem/yr) per (pCi/m 3 )

Nuclide Liber Total hyraid iLung fiI-1A H-3 ND 1.12E+03 1.12E+03 1.12E+03 1.122+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 9.51 E+03 ND 1.00+04 1.58E+06 2.29E+04 Fe-55 4.74E+04 2.52E+04 7.77E+03 ND ND 1.11 E+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 3.162+03 ND ND 1.11 E+06 3.44E+04 Co-60 ND 1.31 E+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+08 ND 6.44E+06 ND ND 1.482+07 3.43E+05 Y-90 4.1 1E+03 ND 1.1 IE+02 ND ND 2.62E+05 2.68E+05 Y-91 9.14E+05 ND 2.44E+04 ND ND 2.631E+06 1.84E+05 Zr-95 1.90E+05 4.182+04 3.702+04 ND 5.96E+04 2.232+06 6.11 E+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-106 1.36E+05 ND 1.69E+04 ND 1.84E+05 1.432+07 4.29E+05 Ag-i lOim 1.69E+04 1.142+04 9.142+03 ND 2.122+04 5.48E+06 1.00E+05 Cd-109 ND 5.482+05 2.59E+04 ND 4.962+05 1.052+06 2.782+04 Sn-I 13 1.132+05 3.122+03 8.622+03 2.332+03 ND 1.46E+06 2.262+05 Sb-124 5.742+04 7.402+02 2.002+04 1.26E+02 ND 3.24E+06 1.642+05 Sb-125 9.84E+04 7.592+02 2.07E+04 9.1 02+0 I ND 2.32E+06 4.03E+04 Te-127m 2.49E+04 8.552+03 3.022+03 6.07E+03 6.362+04 1.482+06 7.14E+04 Te-129m 1.922+04 6.85E+03 3.042+03 6.33E+03 5.03E+04 1.762+06 1.82E+05 23

APA-ZZ-01003 Rev. 014 TABLE 3.3 CHILD PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Inhalation Pathway (mrem/yr) per (gCi/m 3 )

Nudie BD= MMe Total Kidny Lun (lLLI 1-130 8.18E+03 1.64E+04 8.44E+03 1.85E+06 2.45E+04 ND 5.11E+03 1-131 4.81E+04 4.81E+04 2.73E+04 1.62E+07 7.88E+04 ND 2.84E+03 1-132 2.12E+03 4.07E+03 1.88E+03 1.94E+05 6.25E+03 ND 3.202+03 1-133 1.66E+04 2.03E+04 7.70E+03 3.85E+06 3.38E+04 ND 5.48E+03 1-134 1.17E+03 2.162+03 9.95E+02 5.07E+04 3.30E+03 ND 9.55E+02 1-135 4.92E+03 8.73E+03 4.14E+03 7.92E+O5 1.34E+04 ND 4.44E+03 Cs-134 6.51E+05 1.0IE+06 2.252+05 ND 3.302+05 1.21E+05 3.85E+03 Cs-136 6.51E+04 1.71E+05 1.16E+05 ND 9.55E+04 1.45E+04 4.18E+03 Cs-137 9.07E+05 8.25E+05 1.28E+05 ND 2.82E+05 1.04E+05 3.62E+03 Ba-140 7.40E+04 6.48E+01 4.33E+03 ND 2.11E+01 1.74E+06 1.02E+05 La-140 6.44E+02 2.25E+02 7.55E+01 ND ND 1.832+05 2.26E+05 Ce-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 1.852+04 5.55E+03 9.14E+02 ND 3.OOE+03 4.332+05 9.732+04 Pr-144 5.96E-02 1.85E-02 3.00E-03 ND 9.77E-03 1.57E+03 1.972+02

) Nd-147 1.08E+04 8.73E+03 6.81E+02 ND 4.81E+03 3.28E+05 8.21E+04 Eu-154 1.01I+07 9.21E+05 8.402+05 ND 4.032+06 6.142+06 1.1OE+05 Hf-181 2.78E+04 1.0IE+05 1.25E+04 ND 2.052+04 1.062+06 6.62E+04 24

APA-ZZ-0 1003 Rev. 014 TABLE 3.3 CHILD PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Meat Pathway 2

(m mrem/yr) per (jtCi/sec) lucdile ___ Lixur Total ThyriQi if{ILLI 1g H-3 ND 2.34E+02 2.34E+02 2.342+02 2.34E+02 2.34E+02 2.34E+02 Be-7 7.38E+03 1.26E+04 8.07E+03 0.00+00 1.23E+04 0.002+00 7.002+05 Cr-51 0.002+00 0.00E+00 8.802+03 4.88E+03 1.33E+03 8.92E+03 4.67E+05 Mn-54 0.00E+00 8.022+06 2.14E+06 0.002+00 2.25E+06 O.OOE+00 6.73E+06 Fe-55 4.58E+08 2.43E+08 7.52E+07 0.00+00 0.00+00 1.37E+08 4.502+07 Fe-59 3.77E+08 6.102+08 3.04E+08 0.00E+00 0.OOE+00 1.77E+08 6.35E+08 Co-57 O.OOE+00 5.92E+06 1.20E+07 0.002+00 O.OOE+00 0.002+00 4.85E+07 Co-58 O.OOE+00 1.64E+07 5.03E+07 0.002+00 0.OOE+00 0.00E+00 9.59E+07 Co-60 0.00+00 6.94E+07 2.05E+08 0.002+00 O.OOE+00 0.00+00 3.84E+08 Zn-65 3.76E+08 1.002+09 6.23E+08 O.OOE+00 6.31 E+08 0.002+00 1.76E+08 Rb-86 0.00E+00 5.77E+08 3.55E+08 0.OOE+O0 0.00E+00 0.00+00 3.71 E+07 Sr-89 4.82E+08 O.OOE+00 1.38E+07 O.OOE+00 0.002+00 0.OOE+00 1.87E+07 Sr-90 .1.04E+10 O.OOE+00 2.64E+09 O.OOE+00 O.OOE+00 0.002+00 1.40E+08 Y-90 1.93E+05 0.002+00 5.162+03 0.002+00 0.002+00 0.002+00 5.49E+08 Y-91 1.802+06 0.002+00 4.82E+04 0.002+00 0.002+00 0.002+00 2.40E+08 Zr-95 2.67E+06 5.86E+05 5.22E+05 0.002+00 8.392+05 O.OOE+00 6.11 E+08 Nb-95 4.26E+06 1.66E+06 1.182+06 0.002+00 1.562+06 0.002+00 3.07E+09 Ru-103 1.55E+08 0.002+00 5.96E+07 0.002+00 3.90E+08 0.002+00 4.012E+09 Ru-106 4.44E+09 0.002+00 5.54E+08 0.002+00 6.00E+09 0.002+00 6.911E+10 Ag-l1lm 8.40E+06 5.67E+06 4.53E+06 O.OOE+00 1.06E+07 0.002+00 6.75E+08 Cd-109 O.OOE+00 1.912E+06 8.84E+04 0.002+00 1.70E+06 0.002+00 6.182+06 Sn-I 13 2.182+09 4.48E+07 1.24E+08 3.31 E+09 O.OOE+00 0.002+00 1.542+09 Sb-124 2.93E+07 3.80E+05 1.03E+07 6.46E+04 0.002+00 1.62E+07 1.832+08 Sb-125 2.852+07 2.20E+05 5.97E+06 2.64E+04 0.002+00 1.59E+07 6.812E+07 Te-127m 1.78E+09 4.782+08 2.11 E+08 4.25E+08 5.07E+09 0.002+00 1.44E+09 Te-129m 1.79E+09 5.002+08 2.782+08 5.78E+08 5.262+09 0.002+00 2.192+09 1-130 3.06E-06 6.18E-06 3.18E-06 6.802-04 9.232-06 0.002+00 2.89E-06 25

APA-ZZ-0 1003 Rev. 014 TABLE 3.3 CHILD PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Meat Pathway (m2 mrem/yr) per (pCi/sec)

Nuclldc Bkno Liur Total Thy rnid Ln ELLI 1-131 1.66E+07 1.67E+07 9.47E+06 5.511E+09 2.74E+07 O.OOE+OO 1.48E+06 1-132 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+IOO O.OOE+OO O.OOE+OO 1-133 5.70E-01 7.05E-01 2.67E-01 1.31 E+02 1.17E+OO O.OOE+OO 2.84E-O01 1-134 O.OOE+OO O.OOE+00 O.OOE+OO O.OOE+00 O.OOE+OO O.OOE+OOO O.OOE+OO 1-135 O.OOE+OO O.OOE+OO O.OOE+00 O.OOE+OO O.OOE+00 O.OOE+OO O.OOE+OO Cs-134 9.23E+08 1.51 E+09 3.20E+08 O.OOE+OO 4.69E+08 1.68E+08 8.17E+06 Cs-136 1.62E+07 4.46E+07 2.89E1+07 O.OOE+OO 2.38E+07 3.54E+06 1.57E+06 Cs-137 1.33E+09 1.28E+09 1.89E+08 O.OOE+OO 4.16E+08 1.50E+08 8.OOE+06 Ba-140 4.39E+07 3.85E+04 2.56E+06 O.OOE+00 1.2513+04 2.29E+04 2.2213+07 La-140 3.33E+02 1.171E+02 3.93E+01 O.OOE+00 0.0013+00 O.OOE+00 3.25E+06 Ce-141 2.22E+04 1.1IE+04 1.65E+03 O.OOE+00 4.86E+03 O.OOE+00 1.3813+07 Ce-144 2.32E+06 7.27E+05 1.24E+05 0.0013+00 4.02E+05 0.OOE+00 1.89E+08 Pr-143 3.34E+04 11.OOE+04 1.66E+03 0.0013+00 5.43E+03 0.OOE+00 3.6113+07 Pr-144 5.63E+02 1.74E+02 2.83E+01 0.OOE+00 9.21 E+O I 0.0013+00 3.7513+05 Nd-147 1.17E+04 9.48E+03 7.34E+02 0.OOE+00 5.2013+03 0.OOE+00 1.50E+07 Eu-154 1.12E+07 1.01IE+06 9.20E+05 0.OOE+00 4.4313+06 0.OOE+00 2.34E+08 Hf-181 4.77E+06 1.74E+07 2.15E+06 0.0013+00 3.53E+06 O.OOE+00 6.41 E+09 26

APA-ZZ-0 1003 Rev. 014 TABLE 3.3 CHILD PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Grass-Cow-Milk Pathway (m 2 mrem/yr) per ([iCi/sec)

LMr Total Thyroid Kidney GLl1.

H-3 O.OOE+OO 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 O.OOE+00 1.25E+04 0.00+00 7.12E+05 Cr-S I 0.0023+00 O.OOE+00 1.02E+05 5.66E+04 1.55E+04 1.03E+05 5.402+06 Mn-54 0.OOE+00 2.1 0E+07 5.59E+06 O.OOE+OO 5.89E+06 0.00+00 1.76E+07 Fe-55 1.12E+08 5.94E+07 1.84E+07 O.OOE+OO O.OOE+00 3.36E+07 1.IOE+07 Fe-59 1.20E+08 1.95E+08 9.70E+07 O.OOE+OO O.OOE+00 5.64E+07 2.03E+08 Co-57 O.OOE+00 3.84E+06 7.78E+06 O.OOE+00 O.OOE+00 0.OOE+00 3.15E+07 Co-58 O.OOE+OO 1.21E+07 3.72E+07 O.OOE+OO O.OOE+00 O.0OE+OO 7.082+07 Co-60 O.OOE+00 4.32E+07 1.27E+08 O.OOE+00 O.OOE+00 O.0OE+OO 2.39E+08 Zn-65 4.14E+09 1.10+10 6.86E+09 O.OOE+00 6.95E+09 O.OOE+00 1.942+09 Rb.86 O.OOE+00 8.78E+09 5.40E+09 O.OOE+OO O.OOE+OO O.OOE+00 5.65E+08 Sr-89 6.63E+09 O.OOE+OO 1.89E+08 O.OOE+00 O.OOE+00 O.OOE+OO 2.57E+08 Sr-90 1.12E+1 I O.OOE+OO 2.84E+10 O.OOE+OO O.OOE+OO O.OOE+00 1.51E+09 Y-90 3.38E+03 O.OOE+00 9.05E+OI O.OOE+00 O.OOE+00 O.OOE+0O 9.62E+06 Y-91 3.91E+04 O.OOE+OO 1.04E+03 O.OOE+OO 0.OOE+OO O.OOE+00 5.20E+06 Zr-95 3.84E+03 8.43E+02 7.51E+02 O.O0E+OO 1.21E+03 O.OOE+00 8.802+05 Nb-95 3.72E+05 1.45E+05 1.03E+05 O.OOE+OO 1.36E+05 O.0OE+OO 2.68E+08 Ru-103 4.29E+03 O.OOE+00 1.65E+03 0.00+00 1.08E+04 O.OOE+00 1.11 E+05 Ru-106 9.25E+04 O.OOE+OO 1.15E+04 O.OOE+00 1.25E+05 0.00+00 1.442+06 Ag-I lOnm 2.09E+08 1.41E+08 1.13E+08 O.OOE+00 2.63E+08 O.OOE+00 1.68E+10 Cd-109 0.OOE+00 3.86E+06 1.79E+05 O.OOE+OO 3A5E+06 O.OOE+0O 1.252+07 Sn-113 6.1 IE+08 1.26E+07 3.48E+07 9.29E+08 O.OOE+00 O.OOE+0O 4.322+08 Sb- 124 1.09E+08 1.411E+06 3.81E+07 2.40E+05 O.OOE+00 6.03E+07 6.80E+08 Sb-125 8.7 1E+07 6.72E+05 1.83E+07 8.07E+04 O.OOE+00 4.86E+07 2.08E+08 Te-127m 2.08E+08 5.61E+07 2.47E+07 4.98E+07 5.94E+08 O.OOE+OO 1.692+08 Te-129m 2.72E+08 7.59E+07 4.22E+07 8.76E+07 7.98E+08 0.OOE+00 3.3 1E+08 27

APA-ZZ-01003 Rev. 014 TABLE 3.3 CHILD PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Grass-Cow-Milk Pathway (m 2 mrem/yr) per (piCi/sec)

Nuide llonc LiMr Total Thyroid Kidney Lung G-Lldl 1-130 1.73E+06 3.50E+06 1.80E+06 3.85E+08 5.23E+06 O.OOE+00 1.64E+06 1-131 1.30E+09 1.31E+09 7.46E+08 4.34E+11 2.15E+09 O.OOE+O0 1.17E+08 1-132 6.92E-01 1.27E+00 5.85E-01 5.90E+01 1.95E+O0 O.OOE+OO 1.50E+OO 1-133 1.72E+07 2.13E+07 8.05E+06 3.95E+09 3.54E+07 O.OOE+OO 8.57E+06 1-134 O.OOE+OO O.OOE+00 O.OOE+0 O.OOE+OO O.OOE+00 O.OOE+OO 0.OOE+O0 1-135 5.4 1E+04 9.74E+04 4.61E+04 8.63E+06 1.49E+05 O.OOE+OO 7.42E+04 Cs-134 2.27E+10 3.72E+10 7.84E+09 O.OOE+OO 1.15E+10 4.14E+09 2.OOE+08 Cs-136 1.0 IE+09 2.78E+09 1.80E+09 O.OOE+O0 1.48E+09 2.21 E+08 9.78E+07 Cs-137 3.23E+10 3.09E+10 4.56E+09 O.OOE+OO 1.01E+10 3.62E+09 1.93E+08 Ba-140 1.17E+08 1.03E+05 6.84E+06 O.OOE+0O 3.34E+04 6.12E+04 5.94E+07 Ce-141 2.19E+04 1.09E+04 1.62E+03 O.OOE+O0 4.79E+03 O.OOE+O0 1.36E+07 La-140 1.78E+02 6.23E+01 2.1OE+O1 O.OOE+00 O.OOE+O0 O.OOE+O0 1.74E+06 Ce-144 1.62E+06 5.09E+05 8.67E+04 O.OOE+OO 2.82E+05 O.OOE+O0 1.33E+08 Pr-143 7.19E+02 2.16E+02 3.57E+O1 O.OOE+O0 1.177E+02 O.OOE+O0 7.76E+05

) Pr-144 5.04E+00 1.56E+O0 2.53E-01 O.OOE+OO 8.24E-O01 O.OOE+O0 3.35E+03 Nd-147 4.45E+02 3.611E+02 2.79E+O1 O.OOE+O0 1.98E+02 0.OOE+OO 5.711E+05 Eu-154 9.43E+04 8.48E+03 7.75E+03 O.OOE+OO 3.73E+04 O.OOE+O0 1.97E+06 Hf-181 6.44E+02 2.35E+03 2.9 1E+02 O.OOE+O0 4.76E+02 O.OOE+00 8.66E+05 28

APA-ZZ-0 1003 Rev. 014 TABLE 3.3 CHILD PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Grass-Goat-Milk Pathway (m 2 mrem/yr) per (jiCi/sec)

Niwlide BO=e Libr Total K GELU.

H-3 0.0012+00 3.20E+03 3.20E+03 3.20E+03 3.20E+03 3.202+03 3.202+03 Be-7 9.00E+02 1.532+03 9.842+02 0.002+00 1.50E+031 0.00E+00 8.55E+04 Cr-5 l 0.00+00 O.OOE+00 1.222+04 6.79E+03 1.85E+03 1.242+04 6.48E+05 Mn-54 0.00+00 2.52E+06 6.71E+05 0.002+00 7.06E+05 O.OOE+00 2.1 IE+06 Fe-55 1.45E+06 7.72E+05 2.39E+05 0.00+00 0.00+00 4.36E+05 1.43E+05 Fe-59 1.56E+06 2.53E+06 1.262+06 O.OOE+00 0.002+00 7.34E+05 2.64E+06 Co-57 O.OOE+O0 4.61E+05 9.33E3+05 O.OOE+00 0.002+00 O.OOE+00 3.78E+06 Co-58 0.002+00 1.46E+06 4.462+06 0.00+00 0.002+00 0.00E+00 8.50E+06 Co-60 O.OOE+00 5.19E+06 1.532+07 0.002+00 0.002+00 0.00+00 2.87E+07 Zn-65 4.97E+08 1.32E+09 8.23E+08 0.00+00 8.342+08 0.002+00 2.32E+08 Rb-86 0.00E+00 1.05E+09 6.48E+08 O.OOE+00 0.002+00 O.OOE+00 6.78E+07 Sr-89 1.39E+10 0.00+00 3.972+08 0.002+00 0.OOE+00 0.00E+00 5.39E+08 Sr-90 2.35E+1 I 0.OOE+00 5.95E+10 0.002+00 O.OOE+00 0 .OOE+00 3.16E+09 Y-90 4.06E+02 0 .0012+00 1.09E+01 0.002+00 0.00+00 0.002+00 1.152+06 Y-91 4.692+03 0.OOE+00 1.25E+02 O.OOE+00 0.00+00 0.00E+00 6.25E+05 Zr-95 4.60E+02 1.0 IE+02 9.012+01 0.00+00 1.45E+02 0.002+00 1.062+05 Nb-95 4.46E+04 1.74E+04 1.242+04 0.002+00 1.632+04 O.OOE+00 I 3.21E+07 Ru-103 5.14E+02 O.OOE+00 1.98E+02 0.002+00 1.29E+03 0.002+00 1.33E+04 Ru-] 06 1.1 IE+04 O.OOE+00 1.382+03 O.OOE+00 1.50E+04 O.OOE+00 1.732+05 Ag-l 1m 2.5 1E+07 1.69E+07 1.35E+07 0.002+00 3.152+07 0.002+00 2.012+09 Cd-109 0.00+00 4.64E+05 2.15E+04 0.002+00 4.142+05 0.002+00 1.502+06 Sn-I 13 7.33E2+07 1.511E+06 4.18E+06 1.1 12+08 0.002+00 O.OOE+00 5.18E+07 Sb-124 1.30E+07 1.69E+05 4.572+06 2.88E+04 O.OOE+00 7.24E+06 8.16E+07 Sb-125 1.05E+07 8.06E+04 2.19E+06 9.68E+03 0.00+00 5.832+06 2.50E+07 Te-127m 2.50E+07 6.73E+06 2.97E+06 5.982+06 7.13E+07 0.002+00 2.022+07 Te-129m 3.26E+07 9.10E+06 5.06E+06 1.052+07 9.57E+07 0.002+00 3.982+07 1-130 2.08E+06 4.20E+06 2.16E+06 4.62E+08 6.272+06 0.002+00 1.962+06 1-131 1.57E+09 1.57E+09 8.95E+08 5.21+11 I 2.582+09 0.00O+00 1.402+08 1-132 8.30E-0I 1.53E+00 7.02-01 I 7.08E+01 2.342+00 O.OOE+00 1.802+00 29

APA-ZZ-0 1003 Rev. 014 TABLE 3.3 CHILD PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Grass-Goat-Milk Pathway (m 2 mrem/yr) per (pCi/sec)

Nudift DO=c Lbhr Total ThLyrod Lung GLLLI k

1-133 2.06E+07 2.55E+07 9.66E+06 4.74E+09 4.25E+07 O.OOE+00 1.03E+07 1-134 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 1-135 6.49E+04 1.17E+05 5.53E+04 1.04E+07 1.79E+05 O.OOE+00 8.90E+04 Cs-134 6.80E+10 1.12E+11 2.35E+10 O.OOE+00 3.46E+10 1.24E+I 0 6.01E+08 Cs-136 3.04E+09 8.35E+09 5.40E+09 O.OOE+00 4.45E+09 6.63E+08 2.93E+08 Cs-137 9.68E+10 9.27E+10 1.37E+10 O.OOE+00 3.02E+10 1.09E+10 5.80E+08 Ba-140 1.41 E+07 1.23E+04 8.211E+05 O.OOE+00 4.011E+03 7.35E+03 7.13E+06 La-140 2.14E+01 7.47E+00 2.52E+00 O.OOE+00 O.OOE+00 O.OOE+00 2.08E+05 Ce-141 2.63E+03 1.31E+03 1.9SE+02 O.OOE+00 5.75E+02 O.OOE+00 1.63E+06 Ce-144 1.95E+05 6.11 E+04 1.04E+04 O.OOE+00 3.38E+04 O.OOE+00 1.59E+07 Pr- 143 8.63E+01 2.59E+01 4.28E+00 O.OOE+00 1.40E+01 O.OOE+00 9.31 E+04 Pr- 144 6.05E-01 1.87E-01 3.04E-02 O.OOE+00 9.89E-02 O.OOE+00 4.03E+02 Nd-147 5.34E+01 4.33E+01 3.35E+00 O.OOE+00 2.37E+01 O.OOE+00 6.85E+04 Eu-154 1.13E+04 1.02E+03 9.29E+02 O.OOE+00 4.47E+03 O.OOE+00 2.37E+05 Hf-181 7.73E+01 2.81 E+02 3.49E+01 O.OOE+00 5.72E+01 O.OOE+00 1.04E+05 30

APA-ZZ-0 1003

¢ Rev. 0 14 TABLE 3.3 CHILD PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES 1 OTHER THAN NOBLE GASES Vegetation Pathway (m 2mrem/yr) per ([tCi/sec)

Nuclide !Wne Lile Total Arhyroid K n Lung Gl-LL BDody H-3 ND 4.01 E+03 4.011E+03 4.0 1E+03 4.0 1E+03 4.0 1E+03 4.01 E+03 Be-7 3.38E3+o5 5.76E+05 3.70E+OS O.OOE3+OO 5.65E3+OS O.OOE+OO 3.21E+07 Cr-S I .OOE3+OO O.OOE+OO 1.17E3+OS 6.50E-+04 1.78E+04 1.1I9E+OS 6.21E3+06 Mn-54 O.OOE3+OO 6.65E3+08 1*77E3+08 O.OOE3+OO 1.86E+08 O.OOE+OO 5.58E+08 Fe-SS 8.011E+08 4.25E+08 1.32E3+08 O.OOE3+OO O.OOE+OO 2.40E3+08 7.87E3+07 Fe-S9 3.98E+08 6.43E+08 3.20E+08 O.OOE3+OO O.OOE+OO 1.87E+08 6.70E+08 Co-57 O.OOE3+OO 2.99E3+07 6.04E3+07 O.OOE3+OO O.OOE+OO O.OOE+OO 2.45E3+08 Co-58 O.OOE3+OO 6.44E3+07 1.97E2+08 O.OOE3+OO O.OOE3+OO O.OOE+OO 3.76E+08 Co-0 O.OOE3+OO 3.78E3+08 1.12E3+O9 O.OOE+OO O.OOE3+OO O.OOE+OO 2.1 OE3+O9 Zn-65 8.13E3+08 2.17E3+09 1.35E+09 O.OOE+OO 1.36E3+O9 O.OOE3+OO 3.80E3+08 Rb-86 O.OOE3+OO 4.52E+08 2.78E3+08 O.OOE3+OO O.OOE3+OO O.OOE+OO 2.911E+07 Sr-89 3.60E+ 10 O.OOE+OO 1.03E3+O9 O.OOE3+OO O.OOE+OO O.OOE+OO 1.39E+09 Sr-90 1.24E3+12 O.OOE3+OO 3.15SE+I I O.OOE3+OO O.OOE3+OO O.OOE+OO 1.67E+10 Y-90 3.011E+06 O.OOE3+OO 8.04E3+04 O.OOE3+OO O.OOE3+OO O.OOE+OO 8.56E+09 Y-91I 1.86E+07 O.OOE3+OO 4.99E+OS O.OOE3+OO O.OOE3+OO O.OOE+OO 2.48E3+09 Zr-9S 3.86E3+06 8.48E3+OS 7.55E+OS O.OOE3+OO 1.21 E+06 O.OOE3+OO 8.85E3+08 Nb-9S 7.48E3+OS 2.91 E+OS 2.08E3+OS O.OOE+OO 2.74E+OS O.OOE3+OO 5.39E3+08 Ru- 103 1.53E+07 O.OOE3+OO S.90E3+06 O.OOE3+OO 3.86E+07 O.OOE3+OO 3.97E+08 Ru-106 7.45E+08 O.OOE3+OO 9.30E3+07 O.OOE3+OO 1.0 IE+O9 O.OOE+OO 1.16E3+10 Ag-l l m 3.211E+07 2.17E+07 1.73E3+07 O.OOE+OO 4.04E3+07 O.OOE2+OO 2.58E3+09 Cd-109 O.OOE3+OO 2.45E+08 1.14E+07 O.OOE3+OO 2.18E3+08 O.OOE3+OO 7.94E+08 Sn-1 13 1.5813+09 3.25E+07 9.00E3+07 2.40E3+O9 O.OOE+OO O.OOE3+OO 1. 12E3+09 Sb-124 3.52E+08 4.57E+06 123E+08 7.77E+OS O.OOE3+OO 1.9SE3+08 2.20E3+09 Sb-125 4.99E3+08 3.85E+06 1.OSE3+08 4.63E3+OS O.OOE+OO 2.78E3+08 1.I919E+09 Te-127m 1.32E3+09 3.56E3+08 1.57E+08 3.16E3+08 3.77E3+O9 O.OOE3+OO 1.07E3+09 Te-129m 8.41IE+08 2.35E+08 1.31lE+08 2.711E+08 2.47E3+O9 O.OOE3+OO 1.03E3+09 1-130 6.16E+OS 1.24E+06 6.41E3+OS 1.37E3+08 1.86E3+06 O.OOE3+OO 5.82E+OS 1-131 1.43E+08 1.44E+08 8.17E+07 4.76E3+10 2.36E3+08 O.OOE3+OO 1.28E3+07 31

APA-ZZ-01003 Rev. 014 TABLE 3.3 CHILD PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Vegetation Pathway (m 2mrem/yr) per ([tCi/sec)

Nuclide Rnne lMur Total Thyroid Kiney Lung f@LLh 1-132 9.23E+01 1.70E+02 7.80E+01 7.87E+03 2.60E+02 O.OOE+OO 2.00E+02 1-133 3.53E+06 4.37E+06 1.65E+06 8.12E+08 7.28E+06 O.OOE+OO 1.76E+06 I-134 1.56E-04 2.89E-04 1.33E-04 6.65E-03 4.42E-04 O.OOE+OO 1.92E-04 1-135 6.26E+04 1.13E+05 5.33E+04 9.98E+06 1.73E+05 O.OOE+OO 8.59E+04 Cs-134 1.60E+10 2.63E+10 5.55E+09 O.OOE+OO 8.15E+09 2.93E+09 1.42E+08 Cs-136 8.24E+07 2.27E+08 1.47E+08 O.OOE+OO 1.211E+08 1.80E+07 7.96E+06 Cs-137 2.39E+10 2.29E+10 3.38E+09 O.OOE+OO 7.46E+09 2.68E+09 1.43E+08 Ba-140 2.77E+08 2.43E+05 1.62E+07 O.OOE+OO 7.90E+04 1.45E+05 1.40E+08 La-140 3.36E+04 1.18E+04 3.96E+03 O.OOE+OO O.OOE+OO O.OOE+OO 3.28E+08 Ce-141 6.56E+05 3.27E+05 4.86E+04 O.OOE+OO 1.43E+05 O.OOE+OO 4.08E+08 Ce-144 1.27E+08 3.98E+07 6.78E+06 O.OOE+OO 2.21E+07 O.OOE+OO 1.04E+10 Pr-143 1.46E+05 4.37E+04 7.23E+03 O.OOE+OO 2.37E+04 O.OOE+OO 1.57E+08 Pr-144 7.88E+03 2.44E+03 3.97E+02 O.OOE+OO 1.29E+03 O.OOE+OO 5.25E+06 Nd-147 7.15E+04 5.79E+04 4.48E+03 O.OOE+OO 3.18E+04 O.OOE+OO 9.17E+07 Eu-154 1.66E+08 1.50E+07 1.37E+07 O.OOE+OO 6.57E+07 O.OOE+OO 3.48E+09 Hf-181 4.90E+05 1.79E+06 2.21 E+05 O.OOE+OO 3.63E+05 O.OOE+OO 6.59E+08 32

APA-ZZ-0 1003 Rev. 014 TABLE 3.4 ADULT PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Inhalation Pathway (mrem/yr) per ([tCi/m 3 )

Nmclidc Bone Liver Total Kidna Lung (IlLLI 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 4.21 E+04 5.35E1+03 Cr-51 ND ND 1.OOE+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 7.21 E+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 Sr-89 3.04E+05 ND 8.72E+03 ND ND 1.40E+06 3.50E+05 Sr-90 9.92E+07 ND 6.1OE+06 ND ND 9.60E+06 7.22E+05 Y-90 2.09E+03 ND 5.61 E+O I ND ND 1.70E+05 5.06E+05 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 1.77E1+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 1.11OE+05 Ru-106 6.911E+04 ND 8.72E+03 ND 1.34E+05 9.36E+06 9.12E+05 Ag-i lOim 1.08E+04 11.OOE+04 5.94E+03 ND 1.97E+04 4.63E+06 3.02E+05 Cd-109 ND 3.67E+05 1.3 1E+04 ND 3.57E+05 6.83E+05 5.82E+04 Sn-I 13 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+O01 ND 2.48E+06 4.06E+05 Sb-125 5.34E+04 5.95E+02 1.26E+04 5.40E+O01 ND 1.74E+06 1.0 IE+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 1-130 4.58E+03 1.34E+04 5.28E+03 1.14E+06 2.09E+04 ND 7.69E+03 1-131 2.52E+04 3.58E+04 2.05E+04 1.19E+07 6.13E+04 ND 6.28E+03 33

APA-ZZ-0 1003 Rev. 014 TABLE 3.4 ADULT PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Inhalation Pathway (mrem/yr) per (IiCi/m 3 )

Nudid Bone MITE Total Thyroid KBey Lu" fI Bhds 1-132 1.16E+03 3.26E+03 1.16E+03 1.14E+05 5.18E+03 ND 4.06E+02 1-133 8.64E+03 1.48E+04 4.52E+03 2.15E+06 2.58E+04 ND 8.88E+03 1-134 6.44E+02 1.73E+03 6.15E+02 2.98E+04 2.75E+03 ND 1.OIE+OO 1-135 2.68E+03 6.98E+03 2.57E+03 4.48E+05 1.1 E+04 ND 5.25E1+03 Cs-134 3.73E+05 8.48E+05 7.28E+05 ND 2.87E+05 9.76E+04 1.04E+04 Cs-136 3.90E+04 1.46E+05 1.IOE+05 ND 8.56E+04 1.20E+04 1.17E+04 Cs-137 4.78E+05 6.21E+05 4.28E+05 ND 2.22E+05 7.52E+04 8.40E+03 T3a-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+O1 ND ND 1.36E+05 4.58E+05 Ce-141 1.99E+04 1.35E+04 1.53E+03 ND 6.26E+03 3.62E+05 1.20E-f 05 CU-i44 3.43E+06 1.43E+06 1.84E+05 ND 8.48E+05 7.78E+06 8.16E+05 Pt-J43 9.36E+03 3.75E+03 4.64E+02 ND 2.16E+03 2.81E+05 2.OOE+05 1'r-144 3.01E-02 1.25E-02 1.53E-03 ND 7.05E-03 1.02E+03 2.15E-08 Nd-147 5.27E+03 6.1OE+03 3.65E+02 ND 3.56E+03 2.21E+05 1.73E+05 EAu- I 54 5.92E+06 7.28E+05 5.18E+05 ND 3.49E+06 4.67E+06 2.72E+05 Hf-181 1.41E+04 6.82E+04 6.32E+03 ND 1.48E+04 6.85E+05 1.39E+05 34

APA-ZZ-0 1003 Rev. 014 TABLE 3.4 ADULT PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Meat Pathway 2

(m mrem/yr) per (iLCi/sec)

Nufid- BO= Libr Total Thyroid KidneM Lung EkLU H-3 ND 3.25E+02 3.25E+02 3.25E+02 3.25E+02 3.25E+02 3.25E+02 Be-7 4.57E+03 1.042+04 5.07E+03 ND 1.1 OE+04 ND 1.812E+06 Cr-51 ND ND 7.04E+03 4.21 E+03 1.552+03 9.34E+03 1.772+06 Mn-54 ND 9.17E+06 1.75E+06 ND 2.73E+06 ND 2.81 E+07 Fe-55 2.93E+08 2.022+08 4.72E+07 ND ND 1.13E+08 1.162+08 Fe-59 2.652+08 6.24E+08 2.39E+08 ND ND 1.742+08 2.08E+09 Co-57 ND 5.632+06 9.36E+06 ND ND ND 1.43E+08 Co-58 ND 1.82E+07 4.08E+07 ND ND ND 3.69E+08 Co-60 ND 7.51E+07 1.66E+08 ND ND ND 1.41E+09 Zn-65 3.56E+08 1.13E+09 5.1 lE+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 3.011E+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 1.21 E+05 ND 3.24E+03 ND ND ND 1.28E+09 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 9.432+05 ND 1.73E+06 ND 1.06E+10 Ru-103 1.05E+08 ND 4.53E+07 ND 4.01 E+08 ND 1.23E+10 Ru-106 2.80E+09 ND 3.54E+08 ND 5.40E+09 ND 1.8 1E+1 1 Ag-i lOim 6.68E+06 6.18E+06 3.67E+06 ND 1.21E+07 ND 2.52E+09 Cd-109 ND 1.59E+06 5.55E+04 ND 1.52E+06 ND 1.60E+07 Sn-I 13 1.37E+09 3.88E+07 7.86E+07 2.22E+07 ND ND 4.09E+09 Sb-124 1.982+07 3.74E+05 7.84E+06 4.79E+04 ND 1.542+07 5.612+08 Sb-125 1.91 E+07 2.13E+05 4.54E+06 1.94E+04 ND 1.47E+07 2.10E+08 Te-127m 1.111E+09 3.98E+08 1.36E+08 2.852+08 4.53E+09 ND 3.74E+09 Te-129m 1.13E+09 4.232+08 1.79E+08 3.89E+08 4.73E+09 ND 5.71E+09 1-130 2.122-06 6.27E-06 2.47E-06 5.31 E-04 9.782-06 ND 5.40E-06 1-131 1.08E+07 1.54E+07 8.822+06 5.042+09 2.64E+07 ND 4.06E+06 35

APA-ZZ-01003 Rev. 014 TABLE 3.4 ADULT PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Meat Pathway 2

(m mrem/yr) per (jiCi/sec)

Nuclide RD= LM Total Thyroid Kd Lung GELLI Body 1-132 O.OOE+OO O.OOE+OO O O.OOE+00 O.OOE+00 O.OOE+OO ND O.OOE+OO 1-133 3.67E-01 6.39E-01 1.95E-01 9.38E+O1 1.1 IE+OO ND 5.74E-Ol 1-134 O.OOE+OO O.OOE+OO O.OOE+00 O.OOE+OO O.OOE+OO ND O.OOE+OO 1-135 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO ND O.OOE+OO Cs-134 6.57E+08 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 7.81E+08 ND 4.04E+08 1.34E+08 2.31E+07 Ba-140 2.87E+07 3.61E+04 1.88E+06 ND 1.23E+04 2.07E+04 5.91E+07 La-140 2.21E+02 1.IlE+02 2.94E+OI ND ND ND 8.18E+06 Ce-141 1.40E+04 9.49E+03 1.08E+03 ND 4.41E+03 ND 3.63E+07 Ce-144 1.46E+06 6.09E+05 7.82E+04 ND 3.61E+05 ND 4.92E+08 Pr-143 2.1OE+04 8.40E+03 1.04E+03 ND 4.85E+03 ND 9.18E+07 Pr-144 3.52E+02 1.46E+02 1.79E+O1 ND 8.24E+O1 ND 5.06E-05 Nd-147 7.07E+03 8.17E1+03 4.89E+02 ND 4.77E+303 ND 3.92E+07 Eu-154 8.02E+06 9.86E+05 7.01E+05 ND 4.72E+06 ND 7.14E+08 Hf-181 3.01E+06 1.46E+07 1.35E+06 ND 3.14E+06 ND 1.66E+10 36

APA-ZZ-01003 Rev. 014 TABLE 3.4 ADULT PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Grass-Cow-Milk Pathway (m 2mrem/yr) per (ptCi/sec)

Nuclide Rnn Liver Total Lung ThLyrid G W 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 1.71 E+04 6.30E+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 2.51 E+07 1.74E+07 4.05E+06 ND ND 9.68E+06 9.96E+06 Fe-59 2.97E+07 6.98E+07 2.68E+07 ND ND 1.95E+07 2.33E+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 Zn-65 1.37E+09 4.37E+09 1.97E+09 ND 2.92E+09 ND 2.75E+09 Rb-86 ND 2.60E+09 1.21E+09 ND ND ND 5.12E+08 Sr-89 1.45E+09 ND 4.17E+07 ND ND ND 2.33E+08 Sr-90 4.68E+10 ND 1.15E+10 ND ND ND 1.35E+09 Y-90 7.43E+02 ND 1.99E+O 1 ND ND ND 7.87E+06 Y-91 8.59E+03 ND 2.30E+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 5.37E+04 2.89E+04 ND 5.31 E+04 ND 3.26E+08 Ru-103 1.02E+03 ND 4.39E+02 ND 3.89E+03 ND 1.19E+05 Ru-106 2.04E+04 ND 2.58E+03 ND 3.94E+04 ND 1.32E+06 Ag-l10m 5.82E+07 5.39E+07 3.20E+07 ND 1.06E+08 ND 2.20E+10 Cd-109 ND 1.133E+06 3.95E+04 ND 1.08E+06 ND 1.14E+07 Sn-113 1.34E+08 3.8 1E+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 7.31E+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 1.177E+07 1.86E+08 ND 1.54E+08 Te-129m 6.02E+07 2.25E+07 9.53E+06 2.07E+07 2.5 1E+08 ND 3.03E+08 1-130 4.21 E+05 1.24E+06 4.91 E+05 1.05E+08 1.94E+06 ND 1.07E+06 1-131 2.97E+08 4.25E+08 2.43E+08 1.39E+ I 7.28E+08 ND 1.12E+08 37

APA-ZZ-0 1003 Rev. 014 TABLE 3.4 ADULT PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Grass-Cow-Milk Pathway (m 2 mrem/yr) per (gCi/sec)

Nacut __ LI.ier Total Th yriLung LLLI 1-132 1.65E-01 4.42E-01 1.55E-01 1.55E+01 7.04E-O1 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 1-134 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO ND O.OOE+OO 1-135 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.IOE+1O 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.OIE+IO 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 La-140 4.14E+01 2.09E+01 5.51E+OO ND ND ND 1.53E+06 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 Pr-143 1.58E+02 6.34E+0I 7.83E+OO ND 3.66E+O1 ND 6.92E+05 Pr-144 1.IOE+OO 4.58E-01 5.61E-02 ND 2.58E-O1 ND 1.59E-07 Nd-147 9.42E+01 1.09E+02 6.51E+OO ND 6.36E+O1 ND 5.23E+05 Eu-154 2.37E+04 2.91E+03 2.07E+03 ND 1.39E+04 ND 2.1IE+06 Hf-181 IA2E+02 6.92E+02 6.41E+O1 ND 1.49E+02 ND 7.87E+05 38

APA-ZZ-0 1003 Rev. 014 TABLE 3.4 ADULT PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Grass-Goat-Milk Pathway (m2 mrem/yr) per (FiCi/sec)

Nucide Jkne Liver Total Aiyrod Midn la" fdALU 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 Cr-51 ND ND 3.43E+03 2.05E+03 7.56E+02 4.56E+03 8.63E+05 Mn-54 ND 1.011E+06 1.93E+05 ND 3.OIE+05 ND 3.1 OE+06 Fe-55 3.27E+05 2.26E+05 5.26E+04 ND ND 1.26E+05 1.30E+05 Fe-59 3.87E+05 9.08E+05 3.48E+05 ND ND 2.54E+05 3.03E+06 Co-57 ND 1.54E+05 2.56E+05 ND ND ND 3.90E+06 Co-58 ND 5.66E+05 1.27E+06 ND ND ND 1.15E+07 Co-60 ND 1.97E+06 4.35E+06 ND ND ND 3.70E+07 Zn-65 1.65E+08 5.24E+08 2.37E+08 ND 3.51 E+08 ND 3.30E1+08 Rb-86 ND 3.12E+08 1.45E+08 ND ND ND 6.15E+07 Sr-89 3.05E+09 ND 8.75E+07 ND ND ND 4.89E+08 Sr-90 9.84E+10 ND 2.41E+10 ND ND ND 2.84E+09 Y-90 8.92E+O1 ND 2.39E+OO ND ND ND 9.46E+05 Y-91 1.03E+03 ND 2.16E+O1 ND ND ND 5.68E+05 Zr-95 1.13E+02 3.63E+01 2.46E+O01 ND 5.70E+01 ND 1.15E+05 Nb-95 1.16E+04 6.45E+03 3.47E+03 ND 6.37E+03 ND 3.91 E+07 Ru-103 1.22E+02 ND 5.27E+O1 ND 4.67E+02 ND 1.43E+04 Ru-106 2.45E+03 ND 3.1OE+02 ND 4.73E+03 ND 1.59E+05 Ag-I lOOm 6.99E+06 6.47E+06 3.84E+06 ND 1.27E+07 ND 2.64E+09 Cd-109 ND 1.36E+05 4.74E+03 ND 1.30E+05 ND 1.37E+06 Sn-113 1.611E+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 7.50E+03 ND 2.41 E+06 8.78E+07 Sb-125 2.46E+06 2.74E+04 5.84E+05 2.50E1+03 ND 1.89E+06 2.70E+07 Te-127m 5.50E+06 1.97E+06 6.70E+05 1.41 E+06 2.23E+07 ND 1.84E+07 Te-129m 7.23E+06 2.70E+06 1.14E+06 2.48E+06 3.02E+07 ND 3.64E+07 1-130 5.05E+05 1.49E+06 5.88E+05 1.26E+08 2.32E+06 ND 1.28E+06 1-131 3.56E+08 5.09E+08 2.92E+08 1.67E+ I1 8.72E+08 ND 1.34E+08 1-132 1.98E-01 5.29E-01 1.85E-01 1.85E+01 8.43E-O01 ND 9.95E-02 39

APA-ZZ-0 1003 Rev. 014 TABLE 3.4 ADULT PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Grass-Goat-Milk Pathway (m2 mrem/yr) per (pCi/sec)

NudiLe Bo= Lixhr Total Tyroid Lu" CaLI 1-133 4.65E+06 8.09E+06 2.47E+06 1.19E+09 1.41E+07 ND 7.27E+06 1-134 O.OOE+OO O.OOE+OO O.OO1+00 O.OOE+OOO O.OOE+OO ND O.OOE+OO 1-135 1.54E+04 4.04E+04 1.49E+04 2.67E+06 6.48E+04 ND 4.57E+04 Cs-134 1.70E+10 4.04E+I0 3.30E+10 ND 1.31E+10 4.34E+09 7.07E+08 Cs-136 7.91E+08 3.12E+09 2.25E+09 ND 1.74E+09 2.38E+08 3.55E+08 Cs-137 2.22E+I0 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 1.38E+03 2.32E+03 6.65E+06 La-140 4.97E+00 2.51E+00 6.62E-OI ND ND ND 1.84E+05 Ce-141 5.82E+02 3.94E+02 4.46E+O1 ND 1.83E+02 ND 1.50E1+06 Ce-144 4.30E+04 1.80E+04 2.31E+03 ND 1.07E+04 ND 1.45E+07 Pr-143 1.90E+01 7.61E+00 9.40E-O1 ND 4.39E+OO ND 8.31E+04 Pr-144 1.33E-01 5.50E-02 6.74E-03 ND 3.1OE-02 ND 1.91E-08 Nd-147 1.13E+O1 1.31E+O1 7.82E-O1 ND 7.64E+OO ND 6.28E+04 Eu-154 2.84E+03 3.49E+02 2A9E+02 ND 1.67E+03 ND 2.53E+05 Hf-181 1.71E+O1 8.31E+01 7.70E+OO ND 1.79E+O1 ND 9.46E+04

)

40

APA-ZZ-0 1003 Rev. 014 TABLE 3.4 ADULT PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Vegetation Pathway (m2mrem/yr) per (tiCi/sec)

Nulide BOIe LLYrx Total d Kid=e (LLLi H-3 ND 2.26E+03 2.26E+03 2.26E+03 2.26E+03 2.26E+03 2.26E+03 Be-7 9.24E+04 2.1 IE+05 1.03E+05 ND 2.23E+05 ND 3.66E+07 Cr-SI 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 9.31 E+07 ND 9.59E+08 Fe-55 2.11OE+08 1.45E+08 3.38E+07 ND ND 8.08E+07 8.3 1E+07 Fe-59 1.26E+08 2.96E+08 1.14E+08 ND ND 8.28E+07 9.88E+08 Co-57 ND 1.1 7E+07 1.95E+07 ND ND ND 2.97E+08 Co-58 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 1.0 IE+09 4.56E+08 ND 6.75E+08 ND 6.36E+08 Rb-86 ND 2.19E+08 1.02E+08 ND ND ND 4.33E+07 Sr-89 9.97E+09 ND 2.86E+08 ND ND ND 1.60E+09 Sr-90 6.05E+I I ND 1.48E+1I ND ND ND 1.75E+10 Y-90 7.67E+os ND 2.06E+04 ND ND ND 8.14E+09 Y-91 5.1 lE+06 ND 1.37E+05 ND ND ND 2.81E+09 Zr-95 1.17E+06 3.77E+05 2.55E+05 ND 5.91 E+05 ND 1.1 9E+09 Nb-95 2.40E+05 1.34E+05 7.19E+04 ND 1.32E+05 ND 8.1 lE+08 Ru-103 4.77E+06 ND 2.06E+06 ND 1.82E+07 ND 5.57E+08 Ru-106 1.93E+08 ND 2.44E+07 ND 3.72E+08 ND 1.25E+10 Ag- lOnm 1.05E+07 9.75E+06 5.79E+06 ND 1.92E+07 ND 3.98E+09 Cd-109 O.OOE+OO 8.36E+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 4.11 E+07 2.51 E+05 ND 8.07E+07 2.94E+09 Sb-125 1.37E+08 1.53E+06 3.25E+07 1.39E+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 1.1 7E+09 Te-129m 2.51 E+08 9.382+07 3.98E+07 8.64E+07 1.05E+09 ND 1.27E+09 1-130 3.93E+05 1.16E+06 4.57E+05 9.8IE+07 1.8 IE+06 ND 9.97E+05 1-131 8.08E+07 1.16E+08 6.62E+07 3.79E+10 1.98E+08 ND 3.05E+07 41

APA-ZZ-0 1003 Rev. 014 TABLE 3.4 ADULT PATHWAY DOSE FACTORS (RI) FOR RADIONUCLIDES OTHER THAN NOBLE GASES Vegetation Pathway 2

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

Nudlif BDnn I vr Total Thyroid L1ung 1-132 5.77E+01 1.54E+02 5.40E+01 5.40E+03 2.46E+02 ND 2.90E+O01 1-133 2.09E+06 3.63E+06 1.1 IE+06 5.33E+08 6.33E+06 ND 3.26E+06 1-134 9.69E-05 2.63E-04 9.42E-05 4.56E-03 4.19E-04 ND 2.30E-07 1-135 3.90E+04 1.02E+05 3.77E+04 6.74E+06 1.64E+05 ND 1.15E+05 Cs-134 4.67E+09 1.1 IE+10 9.08E+09 ND 3.59E+09 1.19E+09 1.94E+08 Cs-136 4.27E+07 1.69E+08 1.21IE+08 ND 9.38E+07 1.29E+07 1.91E+07 Cs-137 6.36E+09 8.70E+09 5.70E+09 ND 2.95E+09 9.81 E+08 1.68E+08 Ba-140 1.29E+08 1.61 E+05 8.42E+06 ND 5.49E+04 9.24E+04 2.65E+08 La-140 1.58E+04 7.98E+03 2.1 IE+03 ND ND ND 5.86E+08 Ce-141 1.97E+05 1.33E+05 1.51E+04 ND 6.19E+04 ND 5.1OE+08 Ce-144 3.29E+07 1.38E+07 1.77E+06 ND 8.16E+06 ND 1.1 1E+10 Pr-143 6.26E+04 2.51 E+04 3.1OE+03 ND 1.45E+04 ND 2.74E+08 Pr-144 2.03E+03 8.43E+02 1.03E+02 ND 4.75E+02 ND 2.92E-04 Nd-147 3.33E+04 3.85E+04 2.31 E+03 ND 2.25E+04 ND 1.85E+08

) Eu-154 4.85E+07 5.97E+06 4.25E+06 ND 2.86E+07 ND 4.32E+09 Hf-181 1.40E+05 6.82E+05 6.32E+04 ND 1.47E+05 ND 7.76E+08 42

APA-ZZ-0 1003 Rev. 014 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, onsite storage of low-level radioactive waste, and outside storage tanks; b) Dose to skin due to beta radiation from submersion in a cloud 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. 11.12.5)

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

Since the doses via liquid releases are very 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 doses resulting from direct radiation are considered in determining compliance to 40 CFR 190 (Ref. 11.12.3).

There are no other Uranium Fuel Cycle Sources within 8km of the Callaway Plant.

4.1.1 Identification of the MEMBER OF THE PUBLIC The MEMBER OF THE PUBLIC is considered to be a real individual, including all persons not occupationally associated with the Callaway Plant, but who may use portions of the plant site for recreational or other purposes not associated with the plant (Ref. 11.4 and 11.8.10).

Accordingly, it is necessary to characterize this individual with respect to his utilization of areas both within and at or beyond the SITE BOUNDARY and identify, as far as possible, major assumptions which could be reevaluated if necessary to demonstrate continued compliance with 40 CFR 190 through the use of more realistic assumptions (Ref. 11.12.3 and 11.12.4).

The evaluation of Total Dose from the Uranium Fuel Cycle should consider the dose to two Critical Receptors: a) The Nearest Resident, and b) The Critical Receptor within the SITE BOUNDARY.

4.1.2 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 16.11.4.2.

APA-ZZ-0 1003 Rev. 014 4.1.3 Total Dose to the Critical Receptor Within the SITE BOUNDARY The Union Electric Company has entered into an agreement with the State of Missouri Department of Conservation for management of the residual lands surrounding the Callaway Plant, including some areas within the SITE BOUNDARY. Under the terms of this agreement, certain areas have been opened to the public for low intensity recreational uses (hunting, hiking, sightseeing, etc.) but recreational use is excluded in an area immediately surrounding the plant site (refer to Figure 4.1). Much of the residual lands within the SITE BOUNDARY are leased to area farmers by the Department of Conservation to provide income to support management and development costs. Activities conducted under these leases are primarily comprised of farming (animal feed), grazing, and forestry (Ref. 11.7.2, 11.7.3, 11.13, and 11.13.1).

Based on the utilization of areas within the SITE BOUNDARY, it is reasonable to assume that the critical receptor within the SITE BOUNDARY is a farmer, and that his dose from activities within 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. 11.5.5). Occupancy of areas within the SITE BOUNDARY is assumed to be averaged over a period of one year.

Any reevaluation of assumptions should 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 The Total Dose to the farmer from gaseous effluents is calculated for the adult age group using the methodology discussed in Section 3, utilizing concurrent meteorological data at the farmer's residence and historical meteorological data from Table 6.1 for activities within the SITE BOUNDARY. These dispersion parameters were calculated by assuming that the farmer's time is equally distributed over the areas farmed within the SITE BOUNDARY, and already have the total occupancy of 1100 hour0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br />s/year factored into their value (Ref. 11.5.6).

The residence of the current tenant is located at a distance of 2897 meters in the NNW sector.

The gaseous effluents 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 all of the ingestion pathways exist at this location.

It is assumed that food ingestion pathways do not exist within the SITE BOUNDARY, therefore the gaseous effluents dose within the SITE BOUNDARY is due to plume exposure from Noble Gases and the ground plane and inhalation pathways.

4.1.3.1.1 Direct Radiation Dose from Outside Storage Tanks The Refueling Water Storage Tank (RWST) has the highest potential for receiving significant amounts of radioactive materials, and constitutes the only potentially significant source of direct radiation dose from outside storage tanks to a MEMBER OF THE PUBLIC (Ref. 11.6.14, 11.6.15, 11.6.16 and 11.6.17).

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.

The RWST is a right circular cylinder approximately 12 meters in diameter, 14 meters in height with a capacity of approximately 1,514,000 liters (Ref. 11.6.17). The walls are of type 304 stainless steel and have an average thickness of .87 cm. (Ref. 11.14.1).

The direct radiation dose from the RWST is calculated based on the tank's average isotopic content and the parameters discussed above, considering buildup and attenuation within the volume source. Appropriate methodology for calculating the dose rate from a volume source is given in TID-7004, "Reactor Shielding Design Manual" (Ref. 11.17). The computer program ISOSHLD (Ref. 11.18, 11.19 and 11.20) will normally be utilized to perform this calculation.

APA-ZZ-0 1003 Rev. 014 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. The 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. 11.14.3).

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

4.1.3.1.3 Direct Radiation Dose From On-Site Storage Of Low Level Radioactive Waste The on-site storage area for radioactive wastes is located Plant Southwest of the radwaste building and consists of a concrete pad enclosed by a fence. The storage area is bounded on two sides by the radwaste building. The area is also partially bounded on a third side by the Discharge 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 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 HlICs 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.

The HICs typically are right circular cylinders approximately 1.7 meters in diameter and 1.8 meters in height. The HICs are stored inside OSCs which typically are constructed of concrete with additional shielding as necessary to minimize external doses. The individual parameters (e.g., 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-3 10. The computer program MICROSHIELD (Ref. 11.24) will normally be utilized to perform this calculation.

AP1s .iZ-01003 Rev. 014 46

APA-ZZ-0 1003 Rev. 014 5 RADIOLOGICAL ENVIRONMENTAL MONITORING

5.1 DESCRIPTION

OF THE RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM The Radiological Environmental Monitoring Program is intended to act as a background data base for preoperation and to supplement the radiological effluent release monitoring program during plant operation. Radiation exposure to the public from the various specific pathways and direct radiation can be adequately evaluated by this program.

Some deviations from the sampling frequency may be necessary due to seasonal unavailability, hazardous conditions, or other legitimate reasons. Efforts are made to obtain all required samples within the required time frame. Any deviation(s) in sampling frequency or location is documented in the Annual Radiological Environmental Operating Report.

REMP sampling locations that use meteorological sectors and or distance from the plant site were verified as described in reference 11.28.

Sampling, reporting, and analytical requirements are given in FSAR Tables 16.11-7, 16.11-8, and 16.11-9.

Airborne, 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 Environmental Measurements Laboratory Quality Assessment Program (EML) or an equivalent program. Participation includes all of the determinations (sample medium -

radionuclide combination) that are offered by EML 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. 11.16).

Energy dependence is tested at several energies between 30keV and 3MeV corresponding to the approximate energies of the predominant Noble Gases (80, 160, 200 keV), Cs-137 (662 keV),

Co-60 (1225 keV), and at least one energy less than 80 keV. Other testing is performed relative to either Cs-137 or Co-60. (Ref 11.14.10) 6 DETERMINATION OF ANNUAL AVERAGE AND SHORT TERM ATMOSPHERIC DISPERSION PARAMETERS 6.1 ATMOSPHERIC DISPERSION PARAMETERS The values presented in Table 6.1 and Table 6.2 were determined through the analysis of on-site meteorological data collected during the three year period of May 4, 1973 to May 5, 1975 and March 16, 1978 to March 16, 1979.

6.1.1 Longe-Trm Dispersion Estimates The variable trajectory plume segment atmospheric transport model MESODIF-II (NUREG/-CR-0523) and the straight-line Gaussian dispersion model XOQDOQ (NUREG/CR2919) were used for determination of the long-term atmospheric dispersion parameters. A more detailed discussion of the methodology and data utilized to calculate these parameters can be found elsewhere (Ref. 11.6.12).

The Unit Vent and Radwaste Building Vent releases are at elevations of 66.5 meters and 20 meters above grade, respectively. Both release points are within the building wake of the structures on which they are located, and the unit Vent is equipped with a rain cover which effectively eliminates the possibility of the exit velocity exceeding five times the horizontal wind speed. All gaseous releases are thus considered to be ground-level releases, and therefore no mixed mode or elevated release dispersion parameters were determined (Ref. 11.5.2).

APA-ZZ-0 1003 Rev. 014 6.1.2 Determination of Long-Term Dispersion Estimates for Special Receptor Locations Calculations utilizing the PUFF model were performed for 22 standard distances to obtain the desired dispersion parameters. Dispersion parameters at the SITE BOUNDARY and at special receptor locations were estimated by logarithmic interpolation according to (Ref. 11.6.13):

X = X, (d/d 1 )B (6.1)

Where:

B = In (X2 /IX) /In (d2 /d1 ).

Xl, X2 = Atmospheric dispersion parameters at distance dl and d2, respectively, from the source.

The distances dI and d2 were selected such that they satisfy the relationship.

dI < d < d2 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. 11.9.1 and 11.15.1):

F = (Ts /Ta)' (6.2)

Where:

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

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

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).

6.1.3.1 The Determination of the Slope Factor (S)

The general approach employed by subroutine PURGE of XOQDOQ (Ref. 11.15.1) was utilized to produce values of the slope of the (X/Q) curves for both the Radwaste Building Vent and the Unit Vent. However, instead of using approximation procedures to produce the 15 percentile (X/Q) values, the 15 percentile (X/Q) value for each release and at each location was determined by ranking all the I -hour((X/Q)I) values for that release and at that location in descending order.

The (X/Q)I value which corresponded to the 15 percentile of all the calculated (X/Q) values within a sector was extracted for use in the intermittent release (X/Q) calculation.

The intermittent release (X/Q) curve was constructed using the calculated 15 percentile (X/Q)I 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 I 1.15.1. The straight line connecting these points represents (X/Q)I 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)1 /(XIQ)a) (6.3) log (Ta /T1 )

or S (log (X/Q) 1 - log (X/Q)a) (6.4) log T - og T1 APA-ZZ-01003 Rev. 014 6.1.4 Atmospheric Dispersion Parameters for Farming Areas within the SITE BOUNDARY The dispersion parameters for farming areas within the SITE BOUNDARY are intended for a narrow scope application: That of calculating the dose to the current farmer from gaseous effluents while he conducts farming activities within the SITE BOUNDARY.

For the purpose of these calculations, it was assumed that all of the farmer's time, approximately 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> per year, is spent on croplands within the SITE BOUNDARY, and that his time is divided evenly over all of the croplands. Fractional acreage/time - weighted dispersion parameters were calculated for each plot as described in reference 11.5.6. The weighted dispersion parameters for each plot were then summed (according to type) in order to produce a composite value of the dispersion parameters which are presented in Tables 6.1 and 6.2. These dispersion parameters therefore represent the distributed activities of the farmer within the SITE BOUNDARY and his estimated occupancy period.

6.2 ANNUAL METEOROLOGICAL DATA PROCESSING The annual atmospheric dispersion parameters utilized in the calculation of doses for demonstration of compliance with the numerical dose objectives of 10 CFR 50, Appendix I, 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.

The vertical height of the highest adjacent building (V) used to perform concurrent year annual average atmospheric dispersion (X/Q) calculations is 169.16 meters (Ref. 11.29).

Meteorological Data is periodically verified to ensure valid data is being collected. Health Physics is responsible to ensure this review is performed.

APA--2-0 1003 Rev. 014 TABLE 6.1 HIGHEST ANNUAL AVERAGE ATMOSPHERIC DISPERSION PARAMETERS UNIT VENT DISTANCE X/Q X/Q LOCATION (b) SECTOR (METERS) X/Q DECAYED/ DECAYED/ D/Q UNDEPLETED DEPLETED (sec/m 3 ) (sec/m3 ) (sec/m 3 ) (m-2 )

SITE BOUNDARY(a) NNW 2200 L.OE-6 9.9E-7 8.5E-7 4.3E-9 Nearest Residence (c) (d) NNW 2897 6.7E-7 6.6E-7 5.6E-7 2.6E-9 Farmer's Residence(c) (f) NNW 2897 6.7E-7 6.6E-7 5.6E-7 2.6E-9 Farming Areas within the N/A N/A 2.6E-7 2.6E-7 2.4E-7 1.3E-9 Site Boundary (c) (e)

(a) Values given are from FSAR TABLE 2.3-82 (b) Data from 1998 Land Use Census (c) Values derived from FSAR TABLE 2.3-82, using the methodology presented in Equation (6.1) (Ref. 11.5.6)

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

(e) These values were derived for a narrow scope application. Extreme caution should be exercised when determining their suitability for use in other applications.

(f) In 2002, the Farmer moved to a new location directly across the street from the nearest resident. For conversatism, the same dispersion parameters are used.

Building Shape Parameter (C) = 0.5 (Ref. 11.5.3)

Vertical Height of Highest Adjacent Building (V) = 66.45 meters (Ref. 11.5.3) 50

APA-ZZ-G, 3 Rev. 014 TABLE 6.2 HIGHEST ANNUAL AVERAGE ATMOSPHERIC DISPERSION PARAMETERS RADWASTE VENT AND LAUNDRY DECON FACILITY DRYER EXHAUST DISTANCE XIQ X/Q LOCATION (b) SECTOR (METERS) X/Q DECAYED/ DECAYED/ D/Q UNDEPLETED DEPLETED (sec/m 3 ) (sec/r 3 ) (sec/ 3)-2 SITE BOUNDARY(a) NNW 2200 1.3E-6 1.3E-6 I.1E-6 4.3E-9 Nearest Residence (c) (d) NNW 2897 8.5E-7 8.5E-7 7.IE-7 2.6E-9 I Farmer's Residence(c)(f) NNW 2897 8.5E-7 8.5E-7 7.IE-7 2.6E-9 2.6E-9 Farming Areas Within N/A N/A 3.5E-7 3.5E-7 3.2E-7 1.3E-9 Site Boundary (c) (e)

(a) Values given are from FSAR TABLE 2.3-84, (b) Data from 1998 Land Use Census (c) Values derived from FSAR TABLE 2.3-81, using the methodology presented in Equation (6.1) (Ref. 11.5.6)

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

(e) These values were derived for a narrow scope application. Extreme caution should be exercised when determining their suitability for use in other applications.

I (f) In 2002, the farmer moved to a new location directly across the street from the nearest residence. For conservatism, the same dispersion parameters are used.

Building Shape Parameter (C) = 0.5 (Ref. 11.5.3)

Vertical Height of Highest Adjacent Building (V) = 19.96 meters (Ref. 11.5.3) 51

APA-ZZ-01003 Rev. 014 TABLE 6.3 SHORT DISPERSION PARAMETERS (a) (c)

Slope Factor(s)

Location (b) Sector Distance Unit Vent Radwaste Building Vent Site Boundary S 1300 -.328 -.320 Nearest Residence (d) NNW 2897 -.264 -.268 (a) Reference 11.5.3 (b) Data from 1998 Land Use Census (c) Recirculation Factor = 1.0 (d) All pathways are assumed to exist at the location of the nearest resident.

52

APt. '-0 1003 Rev. 014 TABLE 6.4 APPLICATION OF ATMOSPHERIC DISPERSION PARAMETERS Dose Pathway Dispersion Parameter Controlling Age Group Rec Controlling Location Noble Gas, x/Q, decayed/undepleted N/A 16.11.2.2 Site Boundary Beta Air & Gamma Air (2.26 day half-life)

Noble Gas, x/Q, decayed/undepleted N/A 16.11.2.1 Site Boundary Total Body & Skin (2.26 day half-life)

Inhalation x/Q, decayed/depleted Child 16.11.2.1 Nearest Resident (8 day half-life) 16.11.2.3 Site Boundary Ground Plane Deposition D/Q N/A 16.11.2.3 Nearest Resident Ingestion pathways D/Q* Child 16.11.2.3 Nearest Resident

• For H-3, x/Q, decayed/depleted is used instead of D/Q (Ref. 11.1 1.1).

53

APA-ZZ-0 1003 Rev. 014 TABLE 6.5 METEOROLOGICAL DATA SELECTION HIEARCHY Parameter Primary First Second Third Alternate Alternate Alternate Wind Speed lOm Pri lOm Sec 60m Pri 90m Pri Wind Direction I Om Pri I Om Sec 60m Pri 90m Pri Air Temperature I Om Pri I Om Sec Wind Variability l Om Pri l Om Sec 60m Pri 90m Pri Temp Different 60-lOm Pri 90-lOm Pri 90-60 Pri Dew Point I Om Pri Precipitation lm Pri (a) Pri indicates primary tower (b) Sec indicates secondary tower 54

- N APA-ZZ-0 100' ,;

Rev. 014 40

; ,

. .. Table 6.6 :

.ApDI io of Atios R t .1eM;- . . -:

.VAi:ata;'di of Atmos li Disv r V al t' :;;:'Co c eie t-Noble Gas, x/Q, decayed/undepleted N/A Concurrent Site Boundary Beta Air & Gamma Air Dose (2.26 day half-life) Nearest Resident Noble Gas, x/Q, decayed/undepleted N/A Concurrent Site Boundary Total Body & Skin Dose (2.26 day half-life) Nearest Resident Concurrent Farmer's Residence

.' Historical. Inside Site Boundary Ground Plane Deposition Dose D/Q N/A Concurrent Site Boundary Nearest Resident Concurrent Farmer's Residence Historical Inside Site Boundary Inhalation Dose - x/Q, decayed/depleted Child Concurrent Site Boundary (8 day half-life) Nearest Resident Adult Concurrent Farmer's Residence Historical Inside Site Boundary Ingestion Dose Pathways D/Q* Child Concurrent Site Boundary Nearest Resident Adult Concurrent Farmer's Residence

. Historical Inside Site Boundary

• For H-3, x/Q, decayed/depleted is used instead of D/Q (Ref. 11.11.1).

55

APA-ZZ-0 1003 Rev. 014 7 REPORTING REQUIREMENTS l7.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT The reporting requirements for the Annual Radiological Environmental Operating Report have been relocated to FSAR 16.11.5.1.

7.2 RADIOACTIVE EFFLUENT RELEASE REPORT The reporting requirements for the Radioactive Effluent Release Report have been relocated to FSAR 16.11.5.2. The application of atmospheric dispersion parameters in the report is presented in Table 6.6.

APA-ZZ-0 1003 Rev. 014 8 IMPLEMENTATION OF ODCM METHODOLOGY (COMN 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 implement the ODCM methodologies. These codes are evaluated in accordance with the requirements of plant operating procedures to ensure that they produce results consistent with the methodologies presented in the ODCM. Plant procedures implement the ODCM methodology.

APA-ZZ-0 1003 Rev. 014 9 RADIOACTIVE EFFLUENT CONTROLS (REC)

The Radioactive Effluent Controls have been relocated to FSAR 16.11, "Offsite Dose Calculation Manual Radioactive Effluent Controls". The former ODCM REC numbers appear on each of the REC's in the FSAR, and may be used as a cross-reference between the previous and the current numbering system if necessary.

APA-ZZ-01003 Rev. 014 10 ADMINISTRATIVE CONTROLS 10.1 MAJOR CHANGES TO LIOUID AND GASEOUS RADWASTE TREATMENT SYSTEMS 10.1.1 Licensee-initiated major changes to the Radwaste Treatment Systems (liquid and gaseous):

a. A summary of the change MUST be reported to the Commission in the Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the On-Site Review Committee (ORC). On site documentation MUST contain:

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

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

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

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

5) An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general 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 are to be made;

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

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

b. Changes to the Radwaste Treatment Systems Shall become effective upon review and approval by the ORC.

10.2 CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL (ODCM) (COMN 2815) 10.2.1 All changes to the ODCM shall be completed pursuant to T/S AC 5.5.1 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 Review for each revision of the ODCM must include the Health Physics Department.

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 Radioactive Effluent Release Report pursuant to T/S AC 5.5.1.

APA-ZZ-0 1003 4 Rev. 014 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, May21, 1991, Subpart D, page 23374.

11.2 Title 10, "Energy", Chapter 1, Code of Federal Regulations, Part 50, Appendix I; U.S.

Government Printing Office, Washington, D.C. 20402.

11.2.1 10 CFR 50.36 a (b) 11.3 Title 40, "Protection of Environment", Chapter 1, Code of Federal Regulations, Part 190; U.S.

Government Print Office, Washington, D.C. 20402.

11.4 U.S. Nuclear Regulatory Commission, "Technical Specifications Callaway Plant, Unit NO. I",

NUREG-1058 (Rev. 1), October 1984. Section 5.4.1 11.5 COMMUNICATIONS 11.5.1 Letter NEO-54, D. W. Capone to S. E. Miltenberger, dated January 5, 1983; Union Electric Company correspondence.

11.5.2 Letter BLUE 1285, "Callaway Annual Average X/Q and D/Q Values", J. H. Smith (Bechtel Power Corporation), to D. W. Capone (Union Electric Co.), dated February 27, 1984.

11.5.3 Letter BLUE 1232, "Callaway Annual Average X/Q Values and "S" Values", J. H. Smith (Bechtel Power Corporation) to D. W. Capone (Union 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.

11.6 Union Electric Company Callaway Plant, Unit 1, Final Safety Analysis Report.

11.6.1 Section 11.5.2.2.3.1 11.6.2 Section 11.5.2.2.3.4 11.6.3 Section 11.5.2.1.2 11.6.4 Section 11.5.2.2.3.2 11.6.5 Section 11.5.2.2.3.3 11.6.6 Section 11.2.3.3.4 11.6.7 Section 11.2.3.4.3 11.6.8 Section 11.5.2.3.3.1 11.6.9 Section 11.5.2.3.3.2 11.6.10 Section 11.5.2.3.2.3 11.6.11 Section 11.5.2.3.2.2 11.6.12 Section 2.3.5 11.6.13 Section 2.3.5.2.1.2 11.6.14 Section 9.2.6 11.6.15 Section 9.2.7.2.1 11.6.16 Section 6.3.2.2 11.6.17 Table 11.1-6 11.6.18 Deleted APA-ZZ-01003 Rev. 014 11.6.19 Deleted 11.6.20 Deleted 11.6.21 Deleted 11.6.22 Table 2.3-68 11.7 Union Electric Company Callaway Plant Environmental Report, Operating License Stage.

11.7.1 Table 2.1-19 11.7.2 Section 2.1.2.3 11.7.3 Section 2.1.3.3.4 11.7.4 Section 5.2.4.1 11.7.5 Table 2.1-19 11.8 U.S. Nuclear Regulatory Commission, Preparation of Radiological Effluent Technical Specification for Nuclear Power Plants", USNRC NUREG-0133, Washington, D. C. 20555, October 1978.

11.8.1 Pages AA-I through AA-3 11.8.2 Section 5.3.1.3 11.8.3 Section 4.3 11.8.4 Section 5.3.1.5 11.8.5 Section 5.1.1 11.8.6 Section 5.1.2 11.8.7 Section 5.2.1

) 11.8.8 Section 5.2.1.1 11.8.9 Section 5.3.1 11.8.10 Section 3.8 11.8.11 Section 3.3 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.

11.9.1 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 Commnission, Washington, D. D. 20555, July, 1977.

11.10.1 Section c.l.b 11.10.2 Figures 7 through 10 11.10.3 Section c.4 11.11 Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purposes of Evaluating Compliance with 10 CFR Part 50, Appendix I",

Revision 1, U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, October 1977.

11.11.1 Appendix C, Section 3.a 11.11.2 Appendix E, Table E-15 11.11.3 Appendix C, Section 1 11.11 A Appendix E, Table E-l l APA-ZZ-01003 Rev. 014 11.11.5 Appendix E, Table E-9 11.12 U. S. Nuclear Regulatory Commission, "Methods for Demonstrating LWR Compliance with the EPA Uranium Fuel Cycle Standard (40 CFR Part 190)", USNRC NUREG-0543, Washington, D.

C. 20555, January 1980.

11.12.1 Section 1,Page 2 11.12.2 Section IV, Page 8 11.12.3 Section IV, Page 9 11.12.4 Section 111, Page 6 11.12.5 Section 111, Page 8 11.13 Management Agreement for the Public Use of Lands, Union Electric Company and the State of Missouri Department of Conservation, December 21, 1982.

11.13.1 Exhibit A 11.14 MISCELLANEOUS REFERENCES 11.14.1 Drawing Number M-109-0007-06, Revision 5 11.14.2 Callaway Plant Annual Environmental Operating Report (updated annually) 11.14.3 UE Safety Analysis Calculation 87-001-00 11.14.4 Calculation ZZ-48, "Calculation of Inhalation and Ingestion Dose Commitment Factors for the Adult and Child", January, 1988 11.14.5 HPCI 89-02, "Calculation of ODCM Dose Commitment Factors", March, 1989 11.14.6 Deleted 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 ZZ-78, Rev. 2, "ODCM Gaseous Pathway Dose Factors for Adult Age Group",

July, 1992.

11.14.10 HPCI 88-08, "Performance Testing ofthe Environment TLD System at Callaway Plant", August, 1989.

11.14.11 Calculation ZZ-250, Rev. 0, "ODCM Gaseous Pathway Dose Factors for Child Age Group and Ground Plane Dose Factors", September, 1992.

11.14.12 UOTH 83-58, "Documentation of ODCM Dose Factors and Parameters", February, 1983.

11.14.13 Calculation HPCI 95-004 (Rev. 0), "Calculation of Liquid Effluent Dose Commitment factors (AQ) for the Adult Age Group", June, 1996.

11.14.14 Calculation HPCI 03-004 (Rev. 0), "Calculation of Liquid Effluent Dose Commitment factor for Pr-144 (Ai1) for the Adult Age Group", June, 2003.

11.15 U. S. Nuclear Regulatory Commission, "XOQDOQ: Computer Program for the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations", USNRC NUREG/CR-2919, September 1982, Washington, D. C. 20555 11.15.1 Section 4, "Subroutine PURGE", pages 27 and 28 11.16 Regulatory Guide 4.13, "Performance, Testing, and procedural specifications for Thermoluminiscence Dosimetry: Environmental Applications "(Revision 1), July 1977; USNRC, Washington, D. C. 20555 11.17 TID-7004, "Reactor Shielding Design Manual", Rockwell, Theodore, Ed; March 1956.

11.18 BNWL-236, "ISOSHLD - A computer code for General Purpose Isotope Shielding Analysis",

Engel, R. C., Greenberg, J., Hendrichson, M. M.; June 1966 APA-ZZ-01003 Rev. 014 11.19 BNWL-236, Supplement 1, "ISOSHLD- 11: Code Revision to include calculation of Dose Rate from Shielded Bremstrahlung Sources", Simmons, G. L., et al; March 1967

) 11.20 BNNWL-236, Supplement 2, "A Revised Photon Probability Library for use with ISOSHLD- III",

Mansius, C. A.; April 1969.

11.21 ANSI N13.10-1974 , "Specification & Performance of On-Site Instrumentation for Continuously Monitoring Radioactivity in Effluents"; September, 1974 11.22 Nuclear Regulatory Commission Generic Letter 89-01, "Guidance for the Implementation of Programmatic Controls for RETS in the Administrative Controls Section of Technical Specifications and the Relocation of Procedural Details of Current RETS to the Offsite Dose Calculation Manual or Process Control Program", January 1989 11.23 NRC Answers to 10 CFR 20 Implementation Questions 11.23.1 Letter, F. J. Congel to J. F. Schmidt, dated December 9, 1991.

11.23.2 Internal USNRC memo, F. J. Congel to V. L. Miller, et al, dated April 17, 1992.

11.23.3 Letter, F. J. Congel to J. F. Schmidt, dated April 23, 1992.

11.23.4 Letter, F. J. Congel to J. F. Schmidt, dated September 14, 1992.

11.23.5 Letter, F. J. Congel to J. F. Schmidt, dated June 8, 1993.

11.24 USNRC Inspection Report 50-483/92002(DRSS) Section 5, page 5.

11.25 HPCI 96-005, "Calculation of Maximum Background Value for HB-RE-18".

11.26 EGG-PHY-9703, "Technical Evaluation Report for the evaluation of ODCM Revision 0 (May, 1990) Callaway Plant, Unit 1", transmitted via letter, Samual J. Collins (USNRC) to D. F.

Schnell (UE), dated July 12, 1996.

11.27 HPCI 99-005, "Calculation of Setpoint for GL-RE-202".

11.28 HPCI 99-001, "Documentation of REMP Procedure Changes".

11.29 "Technical Specifications for Callaway Plant Meteorological Data Software" CALLAWAY - SP 16.11 OFFSITE DOSE CALCUALATION MANUAL (ODCM 9.0) RADIOACTIVE EFFLUENT CONTROLS 16.11.1 LIQUID EFFLUENT 16.11.1.1 LIQUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION (ODCM 9.3.1)

The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 16.11-1) shall be limited to 10 times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10 4 microCurie/ml total activity.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, immediately restore the concentration to within the above limits.

b. The provisions of Sections 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.1.1.1 SURVEILLANCE REQUIREMENTS (ODCM 9.3.2) 16.11.1.1 .1.a Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 16.11-1.

16.11.1.1 .1.b The results of the radioactivity analysis shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Section 16.11.1.1.

16.11.1.1.2 BASES This section is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than 10 times the concentration in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402. This limitation provides additional assurance that the levels of radioactive materials in bodies 16.11-1 Rev. OL-13f 1/04

CALLAWAY - SP of water in UNRESTRICTED AREAS will result in exposures within: (1)the Section IL.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-1 35 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).

16.11-2 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.1.2 DOSE FROM LIQUID EFFLUENTS LIMITING CONDITION FOR OPERATION (ODCM 9.4.1)

The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Figure 16.11-1) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrems to the whole body and to less than or equal to 5 mrems to any organ, and

b. During any calendar year to less than or equal to 3 mrems to the whole body and to less than or equal to 10 mrems to any organ.

APPLICABILITY: At all times.

ACTION:

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 a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. This Special Report shall also include: (1) the results of radiological analyses of the drinking water source, and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR Part 141, Clean Drinking Water Act.*

b. The provisions of Sections 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.1.2.1 SURVEILLANCE REQUIREMENTS (ODCM 9.4.2)

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

• 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.

16.11-3 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.1.2.2 BASES This section is provided to implement the requirements of Sections II.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section IL.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable".

Also, for fresh water sites with drinking water supplies that can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR Part 141. The dose calculation methodology and parameters in the ODCM implement the requirements in Section IIL.A of Appendix I which specify that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculations of Annual Doses to Man from Routine Releases of 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 Reactor Releases for the Purpose of Implementing Appendix I",April 1977.

The reporting requirements of Action(a) implement the requirements of 10CFR20.2203.

16.11-4 Rev. OL-13f 1/04

'1 CALLAWAY - SP 16.11.1.3 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION (ODCM 9.1.1)

The radioactive liquid effluent monitoring instrumentation channels shown in Table 16.11-2 shall be OPERABLE with their Alarm/Trip Setpoints set to ensure that the limits of Section 16.11.1.1 are not exceeded. The Alarm/Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.

APPLICABILITY: At all times.

ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel Alarm/Trip Setpoint less conservative than required by the above, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable.

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 16.11-2.

Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report, pursuant to Technical Specification 5.6.3, why this inoperability was not corrected within the time specified.

c. The provisions of Sections 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.1.3.1 SURVEILLANCE REQUIREMENTS (ODCM 9.1.2)

Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL OPERATIONAL TEST at the frequencies shown in Table 16.11-3.

16.11.1.3.2 BASES The radioactive liquid effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is 16.11-5 Rev. OL-13f 1/04

CALLAWAY - SP consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

16.11-6 Rev. OL-13f 1/04

.r CALLAWAY - SP 16.11.1.4 LIQUID RADWASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION (ODCM 9.5.1)

The Liquid Radwaste Treatment System shall be OPERABLE and appropriate portions of the system shall be used to reduce releases of radioactivity when the projected doses due to the liquid effluent, from each unit, to UNRESTRICTED AREAS (see Figure 16.11-1) would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31 day period.

APPLICABILITY: At all times.

ACTION:

I. 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 a Special Report that includes the following information:

1) Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability.

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

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

II. The provisions of Sections 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.1.4.1 SURVEILLANCE REQUIREMENTS (ODCM 9.5.2) 16.11.1 .4.1 .a 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.

16.11.1 .4.1 .b The installed Liquid Radwaste Treatment System shall be considered OPERABLE by meeting Sections 16.11.1.1 and 16.11.1.2.

16.11-7 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.1.4.2 BASES The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This section implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section lI.D of Appendix I to 10 CFR Part

50. The specified limits governing the use of appropriate portions of the Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives set forth in Section I.A of Appendix 1,10 CFR Part 50, for liquid effluents.

16.11-8 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.1.5 LIQUID HOLDUP TANKS (3/4.11.1.4) LIMITING CONDITION FOR OPERATION (3.11.1.4)

The quantity of radioactive material contained in each of the following unprotected outdoor tanks shall be limited to less than or equal to 150 Curies, excluding tritium and dissolved or entrained noble gases:

a. Reactor Makeup Water Storage Tank,

b. Refueling Water Storage Tank,

c. Condensate Storage Tank, and

d. Outside temporary tanks, excluding demineralizer vessels and the liner being used to solidify radioactive waste.

APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit, and describe the events leading to this condition in the next Radioactive Effluent Release Report, pursuant to Technical Specification 5.6.3.

b. The provisions of Sections 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.1.5.1 SURVEILLANCE REQUIREMENTS (4.11.1.4)

The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added and within 7 days following any addition of radioactive material to the tank. The provisions of Sections 16.0.2.2 and 16.0.2.3 are applicable, however the allowed surveillance interval extension beyond 25% shall not be exceeded. These tanks are also covered by Administrative Controls Section 5.5.12.

16.11.1.5.2 BASES The tanks listed above include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank 16.11-9 Rev. OL-13f 1/04

CALLAWAY - SP overflows and surrounding area drains connected to the Liquid Radwaste Treatment System.

Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20.1-20.602, Appendix B, Table II, Column 2, (redesignated at 56FR23391, May 21, 1991) at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.

16.11-10 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.2 GASEOUS EFFLUENTS 16.11.2.1 GASEOUS EFFLUENTS DOSE RATE LIMITING CONDITION OF OPERATION (ODCM 9.6.1)

The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 16.11-2) shall be limited to the following:

a. For noble gases: Less than or equal to 500 mrems/yr to the whole body and less than or equal to 3000 mrems/yr to the skin, and

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

APPLICABILITY: At all times.

ACTION:

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

b. The provisions of Sections 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.2.1.1 SURVEILLANCE REQUIREMENTS (ODCM 9.6.2) 16.11.2.1.1.a 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.

16.11.2.1.1.b The dose rate due to lodine-1 31 and 133, tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 16.11-4.

16.11-11 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.2.1.2 BASES 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 are the doses associated with the concentrations of 10 CFR Part 20.1-20.601, Appendix B, Table II, 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 20 10 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 the 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

• Cohen, Paul, Water Coolant Technology of Power Reactors, Table 5.19, page 198. American Nuclear Society. 1980.

• • NUREG-0772, Technical Bases for Estimating Fission Product Behavior During LWR Accidents",

Silberberg, M., editor, USNRC; Figure 4.3, page 4.22. June, 1981.

16.11-12 Rev. OL-13f 1/04

CALLAWAY - SP 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.

16.11-13 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.2.2 DOSE - NOBLE GASES LIMITING CONDITION OF OPERATION (ODCM 9.7.1)

The air dose due to noble gases released in gaseous effluents, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 16.11-2) shall be limited to the following:

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

APPLICABILITY: At all times.

ACTION:

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days 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 Sections 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.2.2.1 SURVEILLANCE REQUIREMENTS (ODCM 9.7.2)

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.

16.11.2.2.2 BASES This section is provided to implement the requirements of Sections II.B, III.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 I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable".

The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE 16.11-14 Rev. OL-13f 1/04

CALLAWAY - SP 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 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 Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors", Revision 1, July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

The reporting requirements of Action(a) implement the requirements of IOCFR20.2203.

16.11-15 Rev. OL-1 3f 1/04

CALLAWAY - SP 16.11.2.3 DOSE - IODINE-131 AND 133, TRITIUM, AND RADIOACTIVE MATERIAL IN PARTICULATE FORM LIMITING CONDITION OF OPERATION (ODCM 9.8.1)

The dose to a MEMBER OF THE PUBLIC from Iodine-131 and 133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 16.11-2) shall be limited to the following:

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

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

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of lodine-131 and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days a Special Report that identifies the cause(s) for exceeding the limits and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

b. The provisions of Sections 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.2.3.1 SURVEILLANCE REQUIREMENTS (ODCM 9.8.2)

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

16.11.2.3.2 BASES This section is provided to implement the requirements of Sections Il.C, Ill.A, and IV.A of Appendix 1,10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section Il.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the release of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as reasonably achievable". The ODCM calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix 16.11-16 Rev. OL-1 3f 1/04

CALLAWAY - SP I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.1 11, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors", Revision 1,July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate controls for Iodine-131, and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing 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)individual inhalation of airborne radionuclides, (2)deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition of radionuclides onto grassy areas where milk animals and meat-producing animals graze with consumption of the milk and meat by man, and (4)deposition on the ground with subsequent exposure of man.

The reporting requirements of Action(a) implement the requirements of 10CFR20.2203.

16.11-17 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.2.4 RADIOACTIVE GASEOUS EFFLUENT MONITORING (3/4.3.3.10) INSTRUMENTATION LIMITING CONDITION FOR OPERATION (ODCM 9.2.1)

The radioactive gaseous effluent monitoring instrumentation channels shown in Table 16.11-5 shall be OPERABLE with their Alarm/Trip Setpoints set to ensure that the limits of Section 16.11.2.1 are not exceeded. The Alarm/Trip Setpoints of these channels meeting Section 16.11.2.1 shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.

APPLICABILITY: As shown in Table 16.11-5.

ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel Alarm/Trip Setpoint less conservative than required by the above, immediately declare the channel inoperable.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 16.11-5.

Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the next Radioactive Effluent Release Report, pursuant to Technical.Specification 5.6.3, why this inoperability was not corrected within the time specified.

c. The provisions of Sections 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.2.4.1 SURVEILLANCE REQUIREMENTS (ODCM 9.2.2)

Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL OPERATIONAL TEST at the frequencies shown in Table 16.11-6.

16.11.2.4.2 BASES The radioactive gaseous effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The sensitivity of any noble gas activity monitor used to show 16.11-18 Rev. OL-1 3f 1/04

CALLAWAY - SP compliance with the gaseous effluent release requirements of Section 16.11.2.1 shall be such that concentrations as low as 1 x 10-6 pCi/cc are measurable.

The monitors GT-RE-22 and GT-RE-33 are only required fo automatic containment purge isolation in MODES 1 through 4. For plant conditions during CORE ALTERATIONS and during movement of irradiated fuel within containment, the function of the monitors is to alarm only and the trip signals for automatic actuation of CPIS may be bypassed. Based on the guidance provided in Regulatory Guide 1.97 concerning monitoring requirements for containment or purge effluent, the monitors GT-RE-22 and GT-RE-33 do not need to meet the single failutre criterion for an Alarm function only during CORE ALTERATIONS or during movement of irradiated fuel in containment. One instrumentation channel at a minimum is required for the alarm only function during refueling activities.

16.11-19 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.2.5 GASEOUS RADWASTE TREATMENT SYSTEM LIMITING CONDITION OF OPERATION (ODCM 9.9.1)

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 16.11-2) would exceed:

a. 0.2 mrad to air from gamma radiation, or

b. 0.4 mrad to air from beta radiation, or

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

APPLICABILITY: At all times ACTION:

I. With radioactive gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days 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.

II. The provision of Sections 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.2.5.1 SURVEILLANCE REQUIREMENTS (ODCM 9.9.2) 16.11.2.5.1.a 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.

16.11.2.5.1 .b 16.11-20 Rev. OL-13f 1/04

CALLAWAY - SP The installed VENTILATION EXHAUST TREATMENT SYSTEM and the WASTE GAS HOLDUP SYSTEMS shall be considered OPERABLE by meeting Sections 16.11.2.1 and 16.11.2.2 or 16.11.2.3.

16.11.2.5.2 BASES 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 effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable". This control implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix 1,10 CFR Part 50, for gaseous effluents.

16.11-21 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.2.6 EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION The concentration of oxygen in the WASTE GAS HOLDUP SYSTEM shall be limited to less than or equal to 3% by volume whenever the hydrogen concentration exceeds 4%

by volume.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM greater than 3% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

b. With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM greater than 4% by volume and the hydrogen concentration greater than 4% by volume, immediately suspend all additions of waste gases to the system and reduce the concentration on oxygen to less than or equal to 4% by volume, then take ACTION a. above.

c. The provisions of Sections 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.2.6.1 SURVEILLANCE REQUIREMENTS The concentrations of hydrogen and oxygen in the WASTE GAS HOLDUP SYSTEM shall be determined to be within the above limits by continuously monitoring the waste gases in the WASTE GAS HOLDUP SYSTEM with the hydrogen and oxygen monitors required OPERABLE by Section 16.11.2.7. The provisions of Sections 16.0.2.2 and 16.0.2.3 are applicable, however the allowed surveillance interval extension beyond 25%

shall not be exceeded. This system is also covered by Administrative Controls Section 5.5.12.

16.11.2.6.2 BASES This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM is maintained below the flammability limits of hydrogen and oxygen. Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

16.11-22 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.2.7 WASTE GAS HOLDUP SYSTEM RECOMBINER EXPLOSIVE GAS MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION At least one hydrogen and both the inlet and outlet oxygen explosive gas monitoring instrument channels for each WASTE GAS HOLDUP SYSTEM recombiner shall be OPERABLE with their Alarm/Trip Setpoints (with the exception of the "FEED H2 4%/

FEED 02 3%" and "FEED H2 4%/FEED 02 4%" alarms) set to ensure that the limits of Section 16.11.2.6 are not exceeded.

APPLICABILITY: During WASTE GAS HOLDUP SYSTEM operation.

ACTION:

a. With an outlet oxygen monitor channel inoperable, operation of the system may continue provided grab samples are 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 />.

b. With both oxygen or both hydrogen channels or both the inlet oxygen and inlet hydrogen monitor channels for one recombiner inoperable, suspend oxygen supply to the recombiner. Addition of waste gas to the system may continue provided grab samples are taken and analyzed at least; 1) once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during mechanical or chemical degassing in preparation for plant shutdown and 2) once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during other operations.

c. With the inlet Oxygen analyzer inoperable, operation of the system may continue provided the inlet Hydrogen is maintained less than 4%. If inlet Hydrogen is greater than 4%, suspend oxygen to the recombiner. Addition of waste gas to the system may continue provided grab samples are taken and analyzed at least; 1) once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during mechanical or chemical degassing operations in preparation for plant shutdown and 2) once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during other operations.

d. The provisions of Sections 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.2.7.1 SURVEILLANCE REQUIREMENTS The provisions of Sections 16.0.2.2 and 16.0.2.3 are applicable, however the allowed surveillance interval extension beyond 25% shall not be exceeded. This system is also covered by Administrative Controls Section 5.5.12.

Each waste gas holdup system recombiner explosive gas monitoring instrumentation channel shall be demonstrated OPERABLE by performance of:

a. A CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 16.11-23 Rev. OL-13f 1/04

CALLAWAY - SP

b. A CHANNEL OPERATIONAL TEST at least once per 92 days, and

c. A CHANNEL CALIBRATION at least once per 92 days with the use of standard gas samples containing a nominal:

1) One volume percent hydrogen, balance nitrogen and four volume percent hydrogen, balance nitrogen for the hydrogen monitor, and

2) One volume percent oxygen, balance nitrogen, and four volume percent oxygen, balance nitrogen for the inlet oxygen monitor, and

3) 10 ppm by volume oxygen, balance nitrogen and 80 ppm by volume oxygen, balance nitrogen for the outlet oxygen monitor.

16.11.2.7.2 BASES Mechanical 'degassing operation' is defined as the transfer of gas from the Volume Control Tank (VCT) to the Waste Gas Holdup System when establishing a nitrogen blanket on the VCT in preparation for a plant shutdown. Chemical 'degassing operation' is the process of adding hydrogen peroxide to the RCS after the VCT hydrogen blanket has been replaced with nitrogen per the Mechanical degassification process and the RCS has been reduced to less than 180'F. Both Mechanical and Chemical degassification may lead to an explosive gas mixture in the Waste Gas Holdup System, thus requiring the more restrictive 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> sampling. Other operations require 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> sampling.

The "FEED H2 4%/FEED 02 3%" AND "FEED H2 4%/FEED 02 4%" alarms are not required to be OPERABLE. These alarms result from the combination of inlet Hydrogen and inlet Oxygen analyzer outputs while the FSAR only addresses OPERABILITY of each separate analyzer. Only the individual alarms and control functions associated with each analyzer are to be used to determine its operability. These alarms and control functions are sufficient to ensure that the requirements of Section 16.11.2.6 are not exceeded.

16.11-24 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.2.8 GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to 2.5 x 105 Curies of noble gases (considered as Xe-133 equivalent).

APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, reduce the tank contents to within the limit, and describe the events leading to this condition in the next Radioactive Effluent Release Report, pursuant to Technical Specification 5.6.3.

b. The provisions of Sections 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.2.8.1 SURVEILLANCE REQUIREMENTS The provisions of Sections 16.0.2.2 and 16.0.2.3 are applicable, however the allowed surveillance interval extension beyond 25% shall not be exceeded. This system is also covered by Administrative Controls Section 5.5.12.

The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per 18 months.

16.11.2.8.2 BASES The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification. Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a MEMBER OF THE PUBLIC at the nearest SITE BOUNDARY will not exceed 0.5 rem. This is consistent with Standard Review Plan 11.3, Branch Technical Position ETSB 11-5, "Postulated Radioactive Releases Due to a Waste Gas System Leak or Failure," in NUREG-0800, July 1981.

16.11-25 Rev. OL-1 3f 1/04

CALLAWAY - SP 16.11.3 TOTAL DOSE 16.11.3.1 TOTAL DOSE LIMITING CONDITION FOR OPERATION (ODCM 9.10.1)

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 Section 16.11.2.2a, 16.11.2.2b, 16.11.2.3a, or 16.11 .2.3b, calculations should be made including direct radiation contributions from the units and from outside storage tanks to determine whether the above limits of Section 16.11.3.1 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days a Special Report that defines the corrective action to be taken to reduce subsequent release to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as 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 condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

b. The provisions of Technical Specifications 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.3.1.1 SURVEILLANCE REQUIREMENTS (ODCM 9.10.2) 16.11.3.1.1 .a 16.11-26 Rev. OL-1 3f 1/04

CALLAWAY - SP Cumulative dose contributions from gaseous effluents shall be determined in accordance with Sections 16.11.2.2.1, and 16.11.2.3.1, and in accordance with the methodology and parameters in the ODCM.

16.11.3.1.1 .b 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 Section 16.11.3.1.

16.11.3.1.2 BASES This specification is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20.1301. The control requires the preparation and submittal of a Special Report whenever the calculated doses due to releases of 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 resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the reactor units and from outside storage tanks are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits.

For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. Ifthe dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR Part 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 Sections 16.11.1.1 and 16.11.2.1.

An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

There are three defined effluent release categories: 1.) Releases directly to the hydrosphere; 2.) noble gas releases to the atmosphere; 3.) radioiodine and particulate releases to the atmosphere. For each effluent 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 effluents, and the child age group is the most limiting for radioiodine and particulates in gaseous effluents.

16.11-27 Rev. OL-13f 1/04

CALLAWAY - SP 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 all different potential dose pathways would contribute to the dose to a single real individual. Since it is difficult or impossible to continually determine actual food use patterns and critical age group, for calculational purposes, assumptions are made which tend to maximize doses. Any refinement 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 significant 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 Effluent 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 pathway. Therefore, only doses to individuals via airborne pathways and dose resulting from direct radiation need to be considered in determining potential compliance to 40 CFR 19*0 The reporting requirements of Action(a) implement the requirements of 10CFR20.2203.

• 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.

16.11-28 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.4 RADIOLOGICAL ENVIRONMENTAL MONITORING 16.11.4.1 MONITORING PROGRAM LIMITING CONDITION OF OPERATION (ODCM 9.11.1)

The Radiological Environmental Monitoring Program shall be conducted as specified in Table 16.11-7.

APPLICABILITY: At all times.

ACTION:

a. With the Radiological Environmental Monitoring Program not being conducted as specified in Table 16.11-7, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Technical Specification 5.6.2, 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 16.11-8 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days 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 Sections 16.11.1.2, 16.11.2.2, or 16.11.2.3. When more than one of the radionuclides in Table 16.11-8 are detected in the sampling medium, this report shall be submitted if:

concentration(1) + concentration (2) + >1.0 reporting level (1) reporting (2)

When radionuclides other than those in Table 16.11-8 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose*

to A MEMBER OF THE PUBLIC from all radionuclides is equal to or greater than the calendar year limits of Sections 16.11.1.2, 16.11.2.2 or 16.11.2.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report, required by Technical Specification 5.6.2.

c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 16.11-7, identify specific locations for

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

16.11-29 Rev. OL-1 3f 1/04

CALLAWAY - SP 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 16.11-9 are unachievable due to uncontrollable circumstances, (such as background fluctuations, unavailable small sample sizes, the presence of interfering nuclides, etc.) the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report.

e. The provisions of Sections 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.4.1.1 SURVEILLANCE REQUIREMENTS (ODCM 9.11.2)

The radiological environmental monitoring samples shall be collected pursuant to Table 16.11-7 and shall be analyzed pursuant to the requirements of Table 16.11-7 and the detection capabilities required by Table 16.11-9.

16.11.4.1.2 BASES The Radiological Environmental Monitoring Program required by this REC provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those 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 Effluent Monitoring Program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience.

The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLD's). The LLD's required by Table 16.11-9 are considered optimum for routine environmental measurements in industrial laboratories.

• Excluding short term or temporary unavailability.

16.11-30 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.4.2 LAND USE CENSUS LIMITING CONDITION OF OPERATION (ODCM 9.12.1)

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. The Land Use Census shall identify water intakes constructed within 10 river miles downstream of the plant discharge point.

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 by Section 16.11.2.3.1, identify the new location(s) in the next Radioactive Effluent Release Report, pursuant to Technical Specification 5.6.3.

b. With a Land Use Census identifying a location(s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with Section 16.11.4.1, add the new location(s) within 30 days to the Radiological Environmental Monitoring Program except for vegetation samples which shall be added to the program before the next growing season. The sampling location(s),

excluding the control station location, having the lowest calculated dose or dose commitment(s), via the same exposure pathway, may be deleted from this monitoring program after October 31 of the year in which this Land Use Census was conducted. In the next Annual Radiological Environmental Operating Report include the revised figure(s) and tables reflecting the new sample location(s) with information supporting the change in sample location.

c. With a Land Use Census identifying a water intake within 10 river miles downstream of the plant discharge point, implement the appropriate waterborne or ingestion sampling required by Table 16.11-7.

d. The provisions of Sections 16.0.1.3 and 16.0.1.4 are not applicable.

• Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each 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.

16.11-31 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.4.2.1 SURVEILLANCE REQUIREMENTS (ODCM 9.12.2)

The Land Use Census shall be conducted during the growing season at least once per 12 months using that information which will provide the best results, such as, but not limited to, door-to-door survey, aerial survey, or by consulting local agriculture authorities and/or residents. The results of the Land Use Census shall be included in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 5.6.2.

16.11.4.2.2 BASES This specification is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the Radiological Environmental Monitoring Program given in the ODCM are made if required by the results of this census. Information that will provide the best results, such as door-to-door survey, aerial survey, or consulting with local agricultural authorities, shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50.

Restricting the census to gardens of greater than 50 m2 provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: (1) 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m2 .

16.11-32 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.4.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION OF OPERATION (ODCM 9.13.1)

Analyses shall be performed on radioactive materials 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 5.6.2.

b. The provisions of Sections 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.4.3.1 SURVEILLANCE REQUIREMENTS (ODCM 9.13.2)

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 5.6.2.

16.11.4.3.2 BASES 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 I to 10 CFR Part 50.

16.11-33 Rev. OL-13f 1/04

CALLAWAY - SP 16.11.5 ADMINISTRATIVE CONTROLS 16.11.5.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (ODCM 7.1)

Routine Annual Radiological Environmental Operating Report covering the operation of the unit 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, 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 Section 16.11.4.2.

It shall also include a listing of new locations for environmental monitoring identified by the Land Use Census pursuant to Section 16.11.4.2.

The Annual Radiological Environmental Operating Report shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to Section 16.11.4.1, as well as summarized tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report. The reports shall also include the following: a summary description of the radiological 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 licensee participation in the Interlaboratory Comparison Program and the corrective action being taken if the specified program is not being performed as required by Section 16.11.4.3; reasons for not conducting the Radiological Environmental Monitoring Program as required by Section 16.11.4.1 and discussion of all deviations from the sampling schedule of Table 16.11-7, discussion of environmental sample measurements that exceed the reporting levels of Table 16.11-8, but are not the result of the plant effluents, pursuant to Section 16.11.4.1; and discussion of all analyses in which the LLD required by Table 16.11-9 was not achievable.

16.11.5.1.1 BASES The reporting requirement for the Annual Radiological Environmental Operating Report is provided to ensure compliance with Technical Specification 5.6.2. This requirement was relocated from the Offsite Dose Calculation Manual to FSAR Chapter 16.

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

16.11-34 Rev. OL-1 3f 1/04

CALLAWAY - SP 16.11.5.2 RADIOACTIVE EFFLUENT RELEASE REPORT (ODCM 7.2)

Routine 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 Radioactive Effluent Release Report shall include a summary of the quantities of radioactive liquid and gaseous effluents released from the unit as outlined in Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants, "Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

The Radioactive Effluent Release Report shall include an annual summary of hourly meteorological data collected over the previous calender year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distribution of wind speed, wind direction, and atmospheric stability* .

This report 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 effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figures 16.11-1 and 16.11-2) during the report period using historical average atmospheric conditions. All assumptions used in making these assessments, i.e., specific activity, exposure time and location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. Assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

The 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.

• 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.

16.11-35 Rev. OL-1 3f 1/04

CALLAWAY - SP The Radioactive Effluent Release Report shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Radioactive Effluent Release Report shall include a summary description of any major changes made during the year to any Liquid or Gaseous Treatment Systems, pursuant to Offsite Dose Calculation Manual. It shall also include a listing of new locations for dose calculations identified by the Land Use Census pursuant to Section 16.11.4.2.

Reporting requirements for changes to Solid Waste Treatment Systems are addressed in APA-ZZ-01011, PROCESS CONTROL PROGRAM (PCP).

The Radioactive Effluent Release Report shall also include the following information: An explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified, and a description of the events leading the liquid holdup tanks or gas storage tanks exceeding the limits of Section 16.11.1.5 or 16.11.2.8.

The Radioactive Effluent Release Report 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 5.5.1.

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

16.11.5.2.1 BASES The reporting requirement for the Radioactive Effluent Release Report is provided to ensure compliance with Technical Specification 5.6.3. This requirement was relocated from the Offsite Dose Calculation Manual to FSAR Chapter 16.

16.11-36 Rev. OL-13f 1/04

CALLAWAY - SP TABLE 16.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM

1. Discharge Monitor Tanks (Batch Release) (2)

SAMPLING FREQUENCY(7) TYPE OF LLD (1)

MINIMUM ANALYSIS ACTIVITY (jCimi)

FREQUENCY ANALYSIS Prior to Each Batch Prior to Each Batch Principal Gamma mitters (3)bE-7i 1-13`1 IE-6 Dissolved and Entrained IE-5 Gases (Gamma Emitters)

H-3 11E-5 Monthly Composite (4) Gross Alpha 1E-7 Quarterly Composite (4) 3r-89, Sr-90 5E-8 e-55 IE-6

~p-237 51E-9 3u-238 5E-9 3u-2391240 5E-9 Plu-241 5E-7 Am-241 5E-9

-242 51E-9 m-243/244 5E-9

2. Steam Generator Blowdown (Concinuous Release) (5)

SAMPLING FREQUENCY(7) TYPE OF LLD (1)

MINIMUM ANALYSIS ACTIVITY (AiCilml)

FREQUENCY ANALYSIS Daily trab Sample (6) Daily Principal Gamma Emitters 3) 5E

-131 1E-6 Dissolved and Entrained E-5 Gases (Gamma Emitters)

-3 E-5 MonthlyComposite (4) Gross Alpha 1E-7 Quarterly Composite (4) Sr-89, Sr-90 5E-8 Fe-55 1E-6 TABLE NOTATIONS (1) The LLD Is defined, for purposes of these specifications, as the smallest concentration of radioactive material In a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a 'real' signal.

For a particular measurement system, which may include radiochemical separation:

LLD - 4.66 Sb E x V x 2.22E6 x Y x exp(-%At)

Where:

LLD = the a priori lower limit of detection (microCuries per unit mass or volume),

Sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute),

E = the counting efficiency (counts per disintegration),

V = the sample size (units of mass or volume),

2.22E6= the number of disintegrations per minute per microCurie, Y = the fractional radiochemical yield, when applicable, X = the radioactive decay constant for the particular radionuclide (sec-1), and Rev. OL-13 5/03

CALLAWAY - SP TABLE 16.11-1 (Sheet 2)

At the elapsed time between the midpoint of the sample collection period, and the time of counting (sec). For batch releases, At=O.

Typical values of E, V. Y,and At should be used In the calculation.

It should be recognized that the LLD is defined as a '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.

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

(3) The principal gamma emitters for which the LLD control applies Include the following radionuclides: Mn-54, Fe-59, Co-58, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be considered.

Other gamma peaks that are Identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Radioactive Effluent Release Report pursuant to Technical Specification 5.6.3, in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974.

(4) A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results In a specimen that Is representative of the liquids released. Prior to analysis, all samples taken for the composite shall be thoroughly mixed In order for the composite samples to be representative of the effluent release.

(5) A continuous release is the discharge of liquid 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 effluent, the sample volume shall be proportioned to the effluent stream discharge volume.

The ratio of sample volume to effluent discharge volume shall be maintained constant for all samples taken for the composite sample.

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

Rev. OL-13 5/03

CALLAWAY - SP TABLE 16.11-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste Discharge Monitor 1 31 (HB-RE-18)

b. Steam Generator Blowdown 1 32 Discharge Monitor (BM-RE-52)

2. Flow Rate Measurement Devices

a. Liquid Radwaste Blowdown 1 34 Discharge Line (HB-FE-2017)

b. Steam Generator Blowdown 1 34 Discharge Line (BM-FE-0054)

c. Cooling Tower Blowdown and 1 34 Bypass FlowTotalizer(FYDB1017A)

3. Discharge Monitoring Tanks (DMTs) Level

a. DMT A(HB-LI-2004) 1 33

b. DMT B(HB-LI-2005) 1 33 ACTION STATEMENTS ACTION 31 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Section 16.11.1.1.1, and

b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving.

Otherwise, suspend release of radioactive effluents via this pathway.

Rev. OL-13f 1/04

CALLAWAY - SP TABLE 16.11-2 (Sheet 2)

ACTION STATEMENTS ACTION 32 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for principal I gamma emitters and 1-131 at a lower limit of detection as specified in Table 16.11-1:

a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activity of the secondary coolant is greater than 0.01 micro-Curie/gram DOSE EQUIVALENT 1-131, or

b. 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 specific activity of the secondary coolant is less than or equal to 0.01 micro-Curie/gram DOSE EQUIVALENT 1-131.

ACTION 33 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the volume discharged is determined by I alternate means.

ACTION 34 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per I 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump performance curves generated in place may be used to estimate flow.

Rev. OL-13f 1/04

CALLAWAY - SP TABLE 16.11-3 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL OPERATIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste D P R(2) Q(1)

Discharge Monitor (HB-RE-1 8)

b. Steam Generator D M R(2) Q(1)

Blowdown Discharge Monitor (BM-RE-52)

2. Flow Rate Measurement Devices

a. Liquid Radwaste D(3) N.A. R N.A.

Blowdown Discharge Line (HB-FE-2017)

b. Steam Generator D(3) N.A. R N.A.

Blowdown Discharge Line (BM-FE-0054)

c. Cooling Tower D(3) N.A. R N.A.

Blowdown and Bypass Flow Totalizer (FYDB1 017A)

3. Discharge Monitoring Tanks (DMT's) Level

a. DMT A(HB-LI-2004) Prior to N.A. R N.A.

release (4)

b. DMT B(HB-LI-2005) Prior to N.A. R N.A.

release (4)

Rev. OL-13f 1/04

CALLAWAY - SP TABLE 16.11-3 (Sheet 2)

TABLE NOTATIONS

1. The CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur as appropriate if any of the following conditions exists:

a. Instrument indicates measured levels above the Alarm/Trip Setpoint (isolation and alarm), or

b. Circuit failure (alarm only), or

c. Instrument indicates a downscale failure (alarm only), or

d. Instrument controls not set in operate mode (alarm only).

2. The initial CHANNEL CALIBRATION shall be performed using one or more of the reference (gas or liquid and solid) standards obtained from the National Institute of Standards and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system over its intended range of energy, measurement range, and establish monitor response to a solid calibration source.

For subsequent CHANNEL CALIBRATION, NIST traceable standard (gas, liquid, or solid) may be used; or a gas, liquid, or solid source that has been calibrated by relating it to equipment that was previously (within 30 days) calibrated by the same geometry and type of source standard traceable to NIST.

3. CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous, periodic, or batch releases are made.

4. CHANNEL CHECK shall consist of verifying indication of tank level during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous, periodic, or batch releases are made from the DMT.

Rev. OL-13f 1/04

CALLAWAY - SP TABLE 16.11-4 RADIOACTIVE GASEOUS EFFLUENTS SAMPLING AND ANALYSIS PROGRAM

2. Containment Purge or Vent SAMPLING FREQUENCY (9) MINIMUM ANALYSIS TYPE OF ACTIVITY ANALYSIS LLD (1)

FREQUENCY (jCimI)

Prior to each release- grab Prior to each release Pnncipal Gamma Emitters- 1E-4 sample particulate, iodine, noble gas (2)

H-3(oxide) 1 E-6 Continuous See footnote 8

3. Unit Vent (3)

SAMPLING FREQUENCY (9) MINIMUM ANALYSIS TYPE OF ACTIVITY ANALYSIS LLD (1)

FREQUENCY (ACi/ml)

Monthly- grab sample (3)(4) Monthly (3)(4) Principal Gamma Emitters-particulate, iodine, noble gas (2)

H-3(oxide) 1E-6 Continuous (6) Weekly (7) 1-131 1E-12 1-133 IE-10 Principal Gamma Emitters- 1E-11 particulate nuclides only (2)

MonthlyComposite Gross Alpha IE-11 Quarterly Composite Sr-89, Sr-90 IE-11

4. Radwaste Building Vent SAMPLING FREQUENCY (9) MINIMUM ANALYSIS TYPE OF ACTIVITY ANALYSIS LLD (1)

FREQUENCY (PCimI)

Monthly- grab sample Monthly Principal Gamma Emitters- 1E4 particulate, Iodine, noble gas (2)

Continuous (6) Weekly (7) 1-131 IE-12 1-133 IE-10 Principal Gamma Emitters- IE-11 particulate nuclides only (2)

Monthly Composite Gross Alpha IE-1I Quarterly Composite Sr-89, Sr-90 1E-11

5. Laundry Decontamination Facility Dryer Exhaust SAMPLING FREQUENCY (9) MINIMUM ANALYSIS TYPE OF ACTIVITY ANALYSIS LLD (1)

FREQUENCY . (AiCimI)

Continuous (6) Weekly (7) Principal Gamma Emitters- 1E-A1 particulate nuclides only (2)

Monthly (10) Composite Gross Alpha IE-11 Quarterly (10) Composite Sr-89, Sr-90 1E-11 Rev. OL-1 3 5/03

CALLAWAY - SP TABLE 16.11-4 (Sheet 2)

6. Containment ILRT Depressurization (Post-test Vent)

SAMPLING FREQUENCY (9) MINIMUM ANALYSIS TYPE OF ACTIVITY ANALYSIS LLD (1)

FREQUENCY (jiCUml)

Pnor to each release- grab Prior to each release Principal Gamma Emitters- 1E-4 sample particulate, iodine, noble gas (2)

H-3(oxide) I E-6 TABLE NOTATIONS (1) The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a 'real' signal.

For a particular measurement system, which may Include radiochemical separation:

LLD = 4.66 Sb E X V x 2.22E6 x Y x exp(-XAt)

Where:

LLD = the 'a priori" lower limit of detection (microCuries per unit mass or volume),

Sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute),

E = the counting efficiency (counts per disintegration),

V = the sample size (units of mass or volume),

2.22E6= the number of disintegrations per minute per microCurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (sect), and At = the elapsed time between the midpoint of the sample collection period, and the time of counting (sec).

Typical values of E, V, Y. and At should be used In the calculation.

It should be recognized that the LLD is defined as a '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.

(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, Ce-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 analyzed and reported in the Radioactive 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, or a thermal power change which exceeds 15% of the rated thermal power within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, samples shall be obtained and analyzed for noble gas, particulates and iodines.

This sampling shall continue to be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for a period of 7 days or until the Unit Vent noble gas monitor no longer indicates a factor of 3 Increase In Unit Vent noble gas activity, whichever comes first.

(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.

(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 Sections 16.11.2.1,16.11.2.2, and 16.11.2.3.

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

(8) Continuous sampling of this batch release pathway is included in the continuous sampling performed for the corresponding continuous release pathway.

(9) Samples shall be representative of the effluent release.

(10) Required only if Mn-54, Fe-59, Co-58, Co-60, Zn-65, Cs-134, Cs-137, Ce-141, or Ce-144 are detected in principle gamma emitter analyses.

Rev. OL-13 5/03

CALLAWAY - SP TABLE 16.11-5 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION

1. Unit Vent System

a. Noble Gas Activity 1 At all times 40,46 Monitor - Providing Alarm (GT-RE-21)

b. Iodine Sampler 1 At all times 43

c. Particulate Sampler I At all times 43

d. Unit Vent Flow Rate I At all times 45

e. Particulate and 1 At all times 43 Radioiodine Sampler Flow Rate Monitor

2. Containment Purge System

a. Noble Gas Activity Monitor Providing Alarm and 2 MODES 1,2,3, and 41 Automatic 4.

Termination of Release (GT-RE-22, GT-RE-33)

- Providing Alarm 1 During CORE 42 function only ALTERATIONS or movement of irradiated fuel within the containment

b. Iodine Sampler 1 MODES 1,2,3,4 and 43 during CORE ALTERATIONS or movement of irradiated fuel within the containment

c. Particulate Sampler 1 MODES 1,2,3,4 and 43 during CORE ALTERATIONS or movement of irradiated fuel within the containment

d. Containment Purge N/A N/A N/A Ventilation Flow Rate Rev. OL-13e 11/03

CALLAWAY - SP TABLE 16.11-5 (Sheet 2)

e. Particulate and 1 MODES 1,2,3,4 and 43 Radioiodine Sampler during CORE Flow Rate Monitor ALTERATIONS or movement of irradiated fuel within the containment

3. Radwaste Building Vent System

a. Noble Gas Activity 1 At all times 38,40 Monitor-Providing Alarm and Automatic Termination of Release (GH-RE-1 0)

b. Iodine Sampler I At all times 43

c. Particulate Sampler I At all times 43

d. Radwaste Building Vent N/A N/A N/A Flow Rate

e. Particulate and 1 At all times 43 Radioiodine Sampler Flow Rate Monitor

4. Laundry Decontamination Faciliy Dryer Exhaust

a. Particulate Monitor When the dryers 47 are operating

b. Particulate Monitor Flow When the dryers 47 Rate Meter are operating

c. Dryer Exhaust NA NA NA Ventilation Flow Rate ACTION STATEMENTS ACTION 38 - With the number of low range channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment for up to 14 days provided that prior to initiating the release:

a. At least two independent samples of the tank's contents are analyzed, and

b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup.

Rev. OL-13e 11/03

CALLAWAY - SP TABLE 16.11-5 (Sheet 3)

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 39- Deleted.

ACTION 40 - With the number of low range channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 41 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the affected channel to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the inoperable channel is not restored within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or with no channels OPERABLE, immediately suspend the release of radioactive effluents via this pathway (this Action must be completed whenever this default condition is entered).

ACTION 42- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, and if the containment equipment hatch is open, then immediately suspend CORE ALTERATIONS and movement of irradiated fuel assemblies within containment. If the containment equipment hatch is not open, then suspend the release of radioactive effluents via this pathway or immediately suspend CORE ALTERATIONS and movement of irradiated fuel assemblies within containment.

ACTION 43 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required in Table 16.11-4.

ACTION 44 - Deleted.

ACTION 45 - Flow rate for this system shall be based on fan status and operating curves or actual measurements.

ACTION 46 - For midrange and high range channels only - with the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, take the action specified in Section 16.3.3.4, ACTION C.

ACTION 47 - With the number of channels OPERABLE less than required by the Minumum Channels OPERABLE requirement, immediately suspend the release of radioactivite effluents via this pathway.

Rev. OL-13e 11/03

CALLAWAY - SP TABLE 16.11-6 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES FOR WHICH CHANNEL SOURCE CHANNEL CHANNEL OPERATIONAL SURVEILLANCE CHECK CHECK CALIBRATION TEST IS REQUIRED

1. Unit Vent System

a. Noble Gas Activity Monitor - D M R(3) 0(2) At all times Providing Alarm (GT-RE-21)

b. Iodine Sampler W N.A. N.A. N.A. At all times

c. Particulate Sampler W N.A. N.A. N.A. At all times

d. Unit Vent Flow Rate N.A. N.A. R(4) a At all times

e. Particulate and Radioiodine D N.A. R 0 At all times Sampler Flow Rate Monitor

2. Containment Purge System

a. Noble Gas Activity Monitor - N.A. P N.A. N.A. MODES 1,2,3,4 and during Providing Alarm and Automatic CORE ALTERATIONS or Termination of Release (GT-RE- movement of Irradiated fuel 22, GT-RE-33) within the containment

b. Iodine Sampler W N.A. N.A. N.A. MODES 1,2,3,4 and during CORE ALTERATIONS or movement of irradiated fuel within the containment

c. Particulate Sampler W N.A. N.A. N.A. MODES 1,2,3,4 and during CORE ALTERATIONS or movement of irradiated fuel within the containment

d. Containment Purge Ventilation N.A. N.A. R(4) N.A. MODES 1,2,3,4 and during Flow Rate CORE ALTERATIONS or movement of irradiated fuel within the containment Rev. OL-13 5/03

CALLAWAY - SP TABLE 16.11-6 (Sheet 2)

MODES FOR WHICH CHANNEL SOURCE CHANNEL CHANNEL OPERATIONAL SURVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION NEST IS REQUIRED

e. Particulate and Radiolodine D N.A. R N.A. MODES 1,2,3,4 and during Sampler Flow Rate Monitor CORE ALTERATIONS or movement of irradiated fuel within the containment

3. Radwaste Building Vent System

a. Noble Gas Activity Monitor - DP MP R(3) 0(1) At all times Providing Alarm and Automatic Termination of Release (GH-RE-1 0)

b. Iodine Sampler W N.A. N.A. N.A. At all times

c. Particulate Sampler W NA. NA. N.A. At all times

d. Radwaste Building Vent Flow NA. N.A. R(4) N.A. At all times Rate

e. Particulate and Radioiodine D N.A. R N.A. At all times Sampler Flow Rate Monitor

4. Laundry Decontamination Facility Dryer Exhaust

a. Particulate Monitor NA D A 0(5) When the dryers are operating

b. Particulate Monitor Flow Rate D NA A NA When the dryers are operating Meter

c. Dryer Exhaust Ventilation Flow NA NA R(4) NA When the dryers are operating Rate Rev. OL-13 5/03

CALLAWAY - SP TABLE 16.11-6 (Sheet 3)

1. The CHANNEL OPERATIONAL TEST shall also demonstrate that 3. The initial CHANNEL CALIBRATION shall be performed using one or more of the reference (gas automatic Isolation of this pathway and control room alarm or liquid and solid) standards certified by the National Institute of Standards & Technology (NIST) annunciation occur as appropriate if any of the following conditions or using standards that have been obtained from suppliers that participate in measurement exists: assurance activities with NIST. These standards shall pemnit calibrating the system over its Intended range of energy, measurement range, and establish monitor response to a solid

a. Instrument Indicates measured levels above the Alarm/rrip calibration source. For subsequent CHANNEL CALIBRATION, NIST traceable standard (gas, Setpoint (isolation and alarm), or liquid, or solid) may be used; or a gas, liquid, or solid source that has been calibrated by relating it to equipment that was previously (within 30 days) calibrated by the same geometry and type of

b. Circuit failure (alarm only), or source standard traceable to NIST.

c. Instrument indicates a downscale failure (alarm only), or 4. If flow rate is determined by exhaust fan status and fan performance curves, the following surveillance operations shall be performed at least once per 18 months:

d. Instrument controls not set In operate mode (alarm only).

a. The specific vent flows by direct measurement, or

2. The CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following b. The differential pressure across the exhaust fan and vent flow established by the fan's conditions exists: 'flow-AP" curve, or

a. Instrument indicates measured levels above the Alarm Setpoint, or c. The fan motor horsepower measured and vent flow established by the fans flow-horsepower" curve.

b. Circuit failure. or

5. The CHANNEL OPERATIONAL TEST shall also demonstrate that automatic Isolation of this

c. Instrument Indicates a downscale failure, or pathway and the shutdown of the dryers occur as appropriate if any of the following conditions exists:

d. Instrument controls not set in operate mode.

a. Instrument indicates measured levels above the AlarmlTrip Setpoint, or

b. Monitor failure.

Rev. OL-13 5/03

CALLAWAY - SP TABLE 16.11-7 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM' NUMBER OF REPRESENTATIVE EXPOSURE PATHWAY SAMPLES AND SAMPLE SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE LOCATIONS (1) COLLECTION FREQUENCY OF ANALYSIS

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

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

2. Airborne Radiolodine and Particulates Samples from five locations; Three samples from close to the SITE Continuous sampler operation with Radiolodine Canister: 1-131 analysis BOUNDARY locations, in different sample collection weekly, or more weekly.

sectors, with high calculated annual frequently if required by dust loading.

average ground level D/Qs. Particulate Samoler: Gross beta radioactivity analysis following filter One sample from the vicinity of a change( 4) and gamma isotopic community located near the plant with analysis( 5) of composite (by location) a high calculated annual average quarterly.

ground level D/Q.

One sample from a location in the vicinity of Fulton, MO.

3. Waterborne

a. Surface (6) One sample upstream Composite sample over 1-month Gamma Isotopic(5 ) and tritium One sample downstream period (7). analysis monthly Rev. OL-13 5/03

CALLAWAY - SP TABLE 16.11-7 (Sheet 2)

NUMBER OF REPRESENTATIVE EXPOSURE PATHWAY SAMPLES AND SAMPLE SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE LOCATIONS ") COLLECTION FREQUENCY OF ANALYSIS t

Qu a r e r l Ga ma I oto ic ~ an tr ti u

b. Ground Samples from one or two sources, Quarterly Gamma lsotopici5r and tritium only if likely to be affected(5) analysis quarterly As there are no ground water sources used for drinking or Irrigation purposes In areas where the hydraulic gradient or recharge properties are suitable for contamination, the ground waterbome pathway Iscurrently not included as part of the Callaway Plant Radiological Environmental Monitoring Program.

c. Drinking One sample of each of one to three of Composite sample over 2-week 1-131 analysis on each composite the nearest water supplies within 10 periodt 7)when 1-131 analysis is when the dose calculated for the miles downstream that could be performed, monthly composite consumption of the water is greater affected by its discharge. otherwise. than 1 mrem per year(9). Composite for gross beta and gamma isotopic analyses( 5) monthly. Composite for tritium analysis quarterly.

One sample from a control location.

As there are no drinking water intakes within 10 miles downstream of the discharge point, the drinking water pathway Iscurrently not included as part of the Callaway Plant Radiological Environmental Monitoring Program. Should the annual Land Use Census identify water intakes within 10 river miles downstream of the discharge point, the program will be revised to include this pathway.

d. Sediment from shoreline One sample from downstream area Semiannually Gamma isotopic analysis( 5) with existing or potential recreational semiannually value Rev. OL-13 5/03

CALLAWAY - SP TABLE 16.11-7 (Sheet 3)

NUMBER OF REPRESENTATIVE EXPOSURE PATHWAY SAMPLES AND SAMPLE SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE LOCATIONS ") COLLECTION FREQUENCY OF ANALYSIS

4. Ingestion

a. Milk Samples from milking animals in three Semimonthly when animals are on Gamma isotopic(5) and 1-131 analysis different meteorological sectors within pasture, monthly at other times semimonthly when animals are on 5 km (3 mile) distance having the pasture: monthly at other times highest dose potential. If there are none, then one sample from milking animals in each of three different meteorological sectors between 5 to 8 km (3 to 5 mile) distance where doses are calculated to be greater than 1 mrem per yr.(9)

One sample from milking animals at a control location, 15 to 30 km (10 to 20 mile) distance and in the least prevalent wind direction.

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 will be revised to include this pathway.

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

One sample of same species in areas not influenced by plant discharge.

c. Food Products One sample of each principal class of At time of harvest(t l0 Gamma Isotopic analysis(5) on edible food products from any area that is portion irrigated by water in which liquid plant wastes have been discharged.

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 sample type is not currently included as part of the Callaway Plant Radiological Environmental Monitoring Program. Should the annual Land Use Census Identify irrigation water intakes within 10 river miles downstream of the discharge point, the program will be revised to include this sample type.

Rev. OL-13 5/03

CALLAWAY - SP TABLE 16.11-7 (Sheet 4)

NUMBER OF REPRESENTATIVE EXPOSURE PATHWAY SAMPLES AND SAMPLE SAMPLING AND TYPE AND FREQUENCY ANDIOR SAMPLE LOCATIONS (') COLLECTION FREQUENCY OF ANALYSIS

c. Food Products Samples of three different kinds of Monthly when available Gamma isotopic (5) and 1-131 analysis broad leaf vegetation if available grown nearest each of two different offsite locations of highest predicted annual average ground level D/Q if milk sampling Is not performed One sample of each of the similar Monthly when available Gamma Isotopic (5) and 1-131 analysis broad leaf vegetation grown 15 to 30 km (10 to 20 mile) distant In the least prevalent wind direction if milk sampling is not performed TABLE NOTATIONS

1. Specific parameters of distance and direction sector from the centerline of one unit, and additional description where pertinent, shall be provided for each and every sample location in Table 16.11-7 In a table and figure(s) in the appropriate plant procedures. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment, and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 5.6.2.

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 and justifying the selection of the new location(s) for obtaining samples.

The selection of sample locations should consider accessibility of 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) Isconsidered to be one phosphor, two or more phosphors In a packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. The number of direct radiation monitoring stations may be reduced according to geographical limitations; e.g., at an ocean site, some sectors will be over water so that the number of dosimeters may be reduced accordingly. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.

3. Deleted.

Rev. OL-13 5/03

CALLAWAY - SP TABLE 16.11-7 (Sheet 5)

4. Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than an established baseline activity level, gamma isotopic analysis shall be performed on the individual samples.

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

6. The 'upstream sample' shall be taken at a distance beyond significant influence of the discharge. The 'downstream' sample shall be taken In an area near the downstream edge of the mixing zone.

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 maximum organ and age group, using the methodology and parameters In the ODCM.

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

Rev. OL-13 5/03

CALLAWAY - SP TABLE 16.11-8 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES REPORTING LEVELS FISH WATER AIRBORNE PARTICULA ATE (pCUkg, MILK FOOD PRODUCTS ANALYSIS (pcUta OR GASES (pCim 3) wet)b (pCV 0 a pCi/kg, wet)b H-3 20,000-Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zn-65 300 20,000 l Zr-Nb-95 400 1-131 2 0.9 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140" 200 300 (a) Multiply the values in this table by 1E-9 to convert to units of iCiIml.

(b) Multiply the values In this table by 1E-9 to convert to units of pCVg.

• For drinking water samples. This is 40 CFR Part 141 value. For surface water samples, a value of 30,000 pCI/e may be used.

- Total activity, parent plus daughter activity.

Rev. OL-13c 8/03

CALLAWAY - SP  :

.1 TABLE 16.11-9 DETECTION CAPABILITITES FOR ENVIRONMENTAL SAMPLE ANALYSIS LOWER LIMIT OF DETECTION (LLD) (1), (2), (3)

FOOD AIRBORNE FISH PRODUCTS SEDIMENT SURFACE DRINKING PARTICULATE (pCi/kg, MILK (pCikg, (pCi/kg, ANALYSIS WATER (pCVt)a WATER (pCVt)a OR GASES (pCVm 3) wet)b (pCie)a wet)b dry)b Gross Beta 4 4 0.01 H-3 3000 2000 Mn-54 15 15 130 Fe-59 30 30 260 Co-58,60 15 15 130 Zn-65 30 30 260 l Zr-Nb-95 15 15 1-131 1000 1 0.07 1 60 Cs-134 15 15 0.05 130 15 60 150 Cs-137 18 18 0.06 150 18 80 180 Ba-La-140* 15 15 15 (a) Multiply the values in this table by 1E-9 to convert to units of pCiml.

(b) Multiply the values in this table by 1E-9 to convert to units of tLCig.

• Total activity, parent plus daughter activity.

Rev. OL-13c 8/03

CALLAWAY - SP TABLE 16.11-9 (Sheet 2)

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, shalt 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 defined, for purposes of these specifications, 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:

4 LLD = .66Sb E x V x 2.22E6 x Y x exp(-XAt)

Where:

LLD = the "a priori lower limit of detection (microCuries per unit mass or volume),

Sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute),

E = the counting efficiency (counts per disintegration),

V = the sample size (units of mass or volume),

2.22E6 = the number of disintegrations per minute per microCurie, Y = the fractional radiochemical yield, when applicable, X = the radioactive decay constant for the particular radionuclide (seca1). and At = the elapsed time between the end of the sample collection period, and the time of counting (sec).

Typical values of E, V, Y, and At should be used In the calculation.

It should be recognized that the LLD is defined as a '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.

Rev. OL-13c 8/03

DATE: 01/14/04 UNION ELECTRIC COMPANY PAGE: 1 TIME: 12:29:39 CALLAWAY RECORDS INFORMATION SYSTEM REPORT: ARCP7301 4 -. CNOT COVER SHEET CN NUMBER: 03-050.

REVISION: 000 A210.0028/F126 FINAL SAFETY ANALYSIS REPORT CHANGE NOTICE - U.E.

VENDOR: U050-AMEREN/UE (FORMERLY UNION ELE CNOT TITLE AND DESCRIPTION:

REVISE SECTION 16.11.1.3 AND TABLE 16.11-2 OF FSAR TO ALLOW EITHER COOLING TOWER BLOWDOWN FLOW (FEDB1101) OR BYPASS FLOW (FEDB1006)

INSTRUMENTATION TO BE TAKEN CREDIT FOR INSTEAD OF REQUIRING BOTH INSTRUMENTS BE OPERABLE FOR A COMBINED FLOW MEASUREMENT (AS LONG AS THE DILUTION PATH USED IS FOR THE OPERABLE INSTRUMENT).

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APPROVED BY: DATE:

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F!ACIR C' oO'Vo CALLAWAY - SP 16.11.1.3 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION (ODCM 9.1.1)

The radioactive liquid effluent monitoring instrumentation channels shown in Table 16.11-2 shall be OPERABLE with their Alarm/Trip Setpoints set to ensure that the limits of Section 16.11.1.1 are not exceeded. The Alarm/Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.

APPLICABILITY: At all times.

ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel Alarm/Trip Setpoint less conservative than required by the above, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable.

b. With less than the minimum number of radioactive liquid effluent monitoring

-fnribmon-ta chiianne1ERABLE, take the ACTION shown in Table 16.11-2.

Restore the inoperable instrumentation to OPERABLE status within the time -30,AyS o.&Ji h ujI5I554~ithc ACT;, explain in the next Radioactive Effluent Release Report, pursuant to Technical Specification 5.6.3, why this inoperability was not c=t c iW te spe

c. The provisions of Sections 16.0.1.3 and 16.0.1.4 are not applicable.

16.11.1.3.1 SURVEILLANCE REQUIREMENTS (ODCM 9.1.2)

Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL OPERATIONAL TEST at the frequencies shown in Table 16.11-3.

16.11.1.3.2 BASES The radioactive liquid effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is 16.11-5 Rev. OL-1 3 5/03

CALLAWAY - SP TABLE 16.11-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste Discharge Monitor 1 31 (HB-RE-18)

b. Steam Generator Blowdown 1 32 Discharge Monitor (BM-RE-52)

2. Flow Rate Measurement Devices

a. Liquid Radwaste Blowdown 1 34 Discharge Line (HB-FE-2017)

b. Steam Generator Blowdown 1 34 Discharge Line (BM-FE-0054)

c. jeCmbined Cooling Tower Blowdown 1 34 EDD~~~~ff21210*lSta;~A-/7/03

3. Discharge onitoring n DMT's) Level

a. DMT A(HB-LI-2004) 1 33

b. DMT B(HB-LI-2005) 1 33 ACTION STATEMENTS ACTION 31 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue fo provided that prior to initiating a duayto14-days release:

a. At least two independent samples are analyzed in accordance with Section 16.11.1.1.1, and

b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving.

Otherwise, suspend release of radioactive effluents via this pathway.

Rev. OL-13 5/03

CALLAWAY - SP TABLE 16.11-2 (Sheet 2)

ACTION STATEMENTS ACTION 32 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pov dproied grab samples are bpy zed for principal gamma emitters and 1-131 at a lower limit o f /7 detectin as specified in Table 16.11-1:

a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activity of the secondary coolant is greater than 0.01 micro-Curie/gram DOSE EQUIVALENT 1-131, or

b. 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 specific activity of the secondary coolant is less than or equal to 0.01 micro-Curie/gram DOSE EQUIVALENT 1-131.

ACTION 33 - With the number of channels OPERABLE less than required by the Minimum Channels OPEBLE-meuirement, effluent releases via this pathway may continu fjo-y provided the volume discharged is determined by alter ate means.  ?

ACTION 34 - With the number of hannels OPERABL P less than required by the Minimum Channels ERABLE require ent, effluent releases via this pathway may continue days vided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> uring-actuaIleleases. Pump performance curves generated in place may be used to estimate flow.

Rev. OL-13 5/03

CALLAWAY - SP TABLE 16.11-3 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL OPERATIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste D P R(2) Q(1)

Discharge Monitor (HB-RE-1 8)

b. Steam Generator D M R(2) Q(1)

Blowdown Discharge Monitor (BM-RE-52)

2. Flow Rate Measurement Devices

a. Liquid Radwaste D(3) N.A. R N.A.

Blowdown Discharge Line (HB-FE-2017)

b. Steam Generator D(3) N.A. R N.A.

Blowdown Discharge Line (BM-FE-0054)

c. Coo ling D(3) N.A. R N.A.

Tower Blowdown and Bypass Flow aT ha I Fze r

\ FEDB10(6 (Ir'cOdlo17A0 '<

3. Dicagontrg Tanks (DMT's) Level

a. DMT A(HB-LI-2004) Prior to N.A. R N.A.

release (4)

b. DMT B(HB-LI-2005) Prior to N.A. R N.A.

release (4)

Rev. OL-13 5/03

DAT7.: 08/06/03 UNION ELECTRIC COMPANY PAGE: 1 TIME: b07:55.'03 CALLAWAY RECORDS INFORMATION SYSTEM REPORT: ARCP7301 CNOT COVER SHEET CN NUMBER: 03-020 REVISION: 000 A210.0028/F126 FINAL SAFETY ANALYSIS REPORT CHANGE NOTICE - U.E.

VENDOR: U050-AMEREN/UE (FORMERLY UNION ELE CNOT TITLE AND DESCRIPTION:

ZINC ADDITION TO RCS (CHAPTERS 5 AND 9)

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FSAR CN 03-020 ZINC ADDITION INTO RCS MARKED-UP PAGES

CALLAWAY - SP During reactor startup from the cold condition, and at other times as necessary, hydrazine is employed as an oxygen scavenging agent. The hydrazine solution may be introduced into the RCS in the same manner as described above for the pH control agent.

The reactor coolant is treated with dissolved hydrogen to control the net decomposition of water by radiolysis in the core region. The hydrogen also reacts with oxygen and nitrogen introduced into the RCS as impurities under the impetus of core radiation.

Sufficient partial pressure of hydrogen is maintained in the volume control tank so that the specified equilibrium concentration of hydrogen is maintained in the reactor coolant.

A self-contained pressure control valve maintains a minimum pressure in the vapor space of the volume control tank. This can be adjusted to provide the correct equilibrium hydrogen concentration.

Boron, in the chemical form of boric acid, is added to the RCS for long-term reactivity control of the core.

Suspended solids (corrosion product particulates) are minimized in the reactor coolant ID by controlling makeup water and by the use of small micron filters. Other impurity concentrations are maintained below specified limits by controlling the chemical quality of makeup water and chemical additives and by purification of the reactor coolant

.>u~gI the chemical and volume control system mixed bed demineralizer.

.72.3.2.2 Compatibility of Construction Materials with Reactor Coolant All of the ferritic low alloy and carbon steels which are used in principal pressure retaining applications have corrosion resistant cladding on all surfaces that are exposed to the reactor coolant. The corrosion resistance of the cladding material is at least equivalent to the corrosion resistance of Types 304 and 316 austenitic stainless steel alloys or nickel-chromium-iron alloy, martensitic stainless steel, and precipitation hardened stainless steel. The cladding of ferritic type base materials receives a post-weld heat treatment, as required by the ASME Code.

Ferritic low alloy and carbon steel nozzles have safe ends of either stainless steel wrought materials, stainless steel weld metal analysis A-7 (designated A-8 in the 1974 Edition of the ASME Code), or nickel-chromium-iron alloy weld metal F-Number 43. The latter buttering material requires further safe ending with austenitic stainless steel base material after completion of the post-weld heat treatment when the nozzle is larger than a 4-inch nominal inside diameter and/or the wall thickness is greater than 0.531 inches.

All of the austenitic stainless steel and nickel-chromium-iron alloy base materials with primary pressure retaining applications are used in the solution anneal heat treat condition. These heat treatments are as required by the material specifications.

5.2-14 Rev. OL-13 5/03

FSAR CN 03-020 INSERT A A soluble zinc compound may be added to the reactor coolant as a means to reduce radiation fields within the primary system. The zinc used may be either natural zinc or zinc depleted of 64Zn. When used, the target system zinc concentration is normally maintained to a concentration no greater than 40 ppb.

o AnN

• LM*5 --. )Lb CALLAWAY -SP 9.4.3. The isolation valve is the only moving part located inside the containment. The hydrogen purge subsystem is designed to vent containment atmosphere at a rate of 100 scfm.

The hydrogen purge subsystem has one penetration through which the containment air is purged and filtered. This purge line is located in a missile-protected area, and draws air from well-ventilated areas of the containment in a manner which prevents either spray or sump water from entering the pipe. As indicated in Section 6.2.5.3, purging would not be initiated before approximately 5.1 days after a LOCA, therefore, no separate air supply line will be needed. Makeup air will be available through the instrument air penetration; and, if this penetration is unavailable by the time purging would be necessary, an air bottle can be connected to a number of available penetrations. Should it be necessary to use this backup system, operational considerations and site meteorology would determine the timing and duration of the purges. In any case, sufficient purging would be performed to maintain the hydrogen concentration in the containment atmosphere below 4 volume percent.

6.2.5.2.3 Hydrogen Generation Hydrogen is generated within the containment by various mechanisms, as described below.

a. Radiolytic Hydrogen Generation Water is decomposed into hydrogen and oxygen by the absorption of energy emitted by nuclides contained in the fuel and those intimately mixed with the LOCA water. The quantity of hydrogen that is produced by radiolysis is a function of both the energy of ionizing radiation absorbed by the LOCA water and the net hydrogen radiolysis yield, G(H2), pertaining to the particular physical-chemical state of the irradiated water.

Evidence indicates that the net hydrogen yield from the radiolysis of pure water is 0.44-0.45 molecule per 100 eV of absorbed energy when the gaseous radiolysis products are continuously purged from the water. In the presence of reactive solutes and water in the absence of gas purging of the solution, significant recombination of the products of radiolysis can occur, thereby reducing the net hydrogen yield. However, in accordance with Regulatory Guide 1.7, a value of 0.5 molecule/ 00 eV has been assumed for the net yield of hydrogen from radiolysis of all LOCA water.

The assumptions given in Regulatory Guide 1.7 were used to determine the fission product distribution after the accident. This distribution is assumed to be instantaneous after the accident, and hydrogen production is assumed to begin immediately. Fifty percent of the halogens and I percent of the solids are assumed to be released from the fuel and intimately mixed with the water in the sump. All noble gas activity is 6.2.5-6 Rev. OL-13 5/03

• RR KCN 03-020 CALLAWAY -SP N eorn~crmco' O^'

released from the fuel and is present in the containment atmosphere. The decay energy of the solids was determined from Shure (Ref. 1),

conservatively assuming a 600-day reactor operating time for fission product buildup. Halogen and noble gas inventories were determined from Reference 2. Table 6.2.5-2 gives a summary of the remaining assumptions made in the analysis.

b. Zirconium-Water Reaction One of the major sources of hydrogen immediately following a LOCA is due to metal-water reaction. The extent of the metal-water reaction depends strongly on the course of events assumed for the LOCA and the effectiveness of the emergency core cooling systems. The extent of metal-water reaction is evaluated in accordance with the assumptions of Regulatory Guide 1.7.

Zirconium reacts with steam according to the reaction:

Zr + 2H 2 0 - ZrO2 + 2H 2 The hydrogen gas evolved from this reaction is calculated to be:

21b - mole H2 /lb - mole Zr 0.022 lb -moles H2 91.221b Zr/lb-mole Zr lb Zr The emergency core cooling system (ECCS) is designed to remove core heat at a rate that will prevent the fuel rods from heating to the point where significant Zr - H2 0 reactions will take place. The LOCA analysis shows that with the passive accumulators and the active elements of high-pressure and low-pressure safety injection, less than 0.1 percent of the zirconium cladding will react with water to generate hydrogen (Section 15.6.5). In analyzing postaccident hydrogen generation, it has been conservatively assumed that 5 percent (50 times the calculated amount) of the total mass of zirconium in Zircaloy-4 or Zirlo fuel cladding reacts. For additional conservatism, 5 percent of the estimated 54,000 pounds of zirconium metal in the reactor vessel, including burnable absorbers, is assumed to react. This amounts to 2,700 pounds of zirconium reacting.

The total hydrogen generation from this source is then estimated to be 59.2 lb-moles H2 -

c. Corrosion of Metals and Paints in the Containment Hydrogen is formed by corrosion of metals in the containment. The significant portion of this source of hydrogen is from the corrosion of zinc and aluminum. Figure 6.2.5-2 shows the maximum allowable quantities of 6.2.5-7 Rev. OL-1 3 5/03

CALLAWAY - SP zinc and aluminum inside the containment. Any combination of aluminum and zinc along the curve shown on Figure 6.2.5-2 will yield a maximum hydrogen concentration initiated at the time the concentration re.  % Awrst case combi itomely that combi ich causes the hydrogen concentration in coinment to ch 3 volume percent the fastest and which results in the highst peak hydrogen concentration when the recombiner is initiated at one d after a LOCA, is selected as a bounding case. This worst case combinati is used to generate Figures 6.2.5-3 through 6.2.5-7. Table 6.2.5-4 giv the temperature dependent corrosion rates used in the calculation for aluminuiand zinc.

Ziiirnnh containment is irktwo forms: zinc base paint and in galv zed steel. For the hydrogen generation calculation, the corrosion rate or zinc base paint are conservatively assumed to be the same as the co osion rates for galvanized steel. This is conservative because, as can e seen in Table 6.2.5-4, the corrosion rates for galvanized steel are greatr than those for zinc base paint. The containment, during the injecti phase, is sprayed with a borated solution having a pH as low as 4.0. uring the recirculation phase, the equilibrium pH of the spray is calc ated to be greater than or equal to 7.1. The corrosion rates for zin nd aluminum used in the hydrogen generation calculation are temp ature-dependent with the assumption of a constant recirculation su of 9.5. With the Hp-replacement of the spray additive system with t.odium phosphate baskets D in the containment recirculation sumps, a sessment was performed to evaluate the impact on the post-LOC drogen generation calculation.

The generation of hydrogen du the corrosion of aluminum for a minimum equilibrium sum of 7.1 would be less than that previously calculated for a sump of 9.5. The generation of hydrogen due to the corrosion of zinc f e lower sump pH would also be less than that previousl ated for a sump pH of 9.5 above a temperature of 170°F.

eneration of hydrogen due to the corrosion of zinc below a temperature of 1700 F would be greater at the lower sump pH; however, this effect would be compensated for by the decrease in the aluminum corrosion rates. Since aluminum is the dominant source of hydrogen from corrosion at Callaway, the use of trisodium phosphate baskets in replacing the spray additive system will not adversely impact the results of the post-LOCA hydrogen generation calculation.

The corrosion rates used in the hydrogen generation calculation, as given in Table 6.2.5-4, have been adjusted upward for higher temperatures which occur early in the accident, as requested in Regulatory Guide 1.7.

The surface areas for the corrosion of metals and paints are assumed constant throughout the analysis.

6.2.5-8 Rev. OL-1 3 5/03

CALLAWAY - SP C

TABLE 6.14 DESIGN CO.MPARISONTO REGULATORY POSITIONS OF REGULATORY GUIDE 144. REVISION O, DAMED MAY 1973, TITILED "CONTROL OF THE USE OF SENSITIZED STAINLESS STEEL" I-D Regulatory Guide PosTlon on Non-NSSS Posilion on 144 Position comonent NSSS Components

-CD Unstablized. auslenilic stainless steel of the AISI Type 3XX series used for components that are part of (1)the reactor coolanl pressute boundary, (2)systems required for reactol shutdown, (3)systems requirred for emergency core cooling.

and (4)reactor vessel internals that are nelied upon to permit adequate core cooling for any mode of normal operaaion or under credible postulated accident conditions should meet the, follovwinrr: .

1. Maerial should biuilabiy protd against contaminants 1. Complies 1. Complies, as discussed capable of ceusing stress corrosion cacking during fabrication. shipment, storage. constructon. testingo and operation of components and syslems.

2. Malerial from vAich components and systems ate to br 2. CompK.es. 2. Complies asdisrussedinSerlion5.2.3.42.

FSAP cA fabricatled 'shouldbe solution hea treated to produce a 03 -02(0.

nonsensilized condition in the materiel.

3. Nonsenstization of the Malteual should be verified using 3. All auslerilc stainless steels are furnished in the 3. Conmplies, as discussed in Secon 5.2.3.4.3.

ASTM A 262-70. 'Recommended Practices for Deteding solution annealed and water-quenched conditon.

Altstc in Stainle-s Sleel," Prmdices A or E or another Since susceptibility to stress corrosion cracking in this method that can be damonstrated to shovincnsensilization  : condition is minimal, testing Isnot performed.

In austenilic stainless steel. Test spedmens should be selected from mate.iai sub' ecl to eah diffarent hetal treatment practice and from each heat.

4. Material subjected to sensitizing temperature In the range 4. During fabrication and installation, austenilic stainless 4. Complies, as discussed In Secton 5.2.3.4.4.

ot 6i0 to 15D00F subsequent to solution heal treating In steels ore not permited to be expoed to tempesatures accordance with Subparagraph C.2. above and testing In In the range of 800.-1 50O'F excepl for welding.

arccordance wilh Subparagrsph C.3. above. should be L Wefrding practices are controlled to minimize Grads material Ihat Is,it should not have a carbon content sensitization, as discussed in Position 5 below.

greater than 0.03 peroenl. Exceptions are:

a. Matetial e)oosed to reactor coolant vhitch has a controlled concentration of less than 0.10 ppm dissolved oxygen at all temperatures above 200'F during nomial operatlon. or Rev. OL-13 5/03

J-- 1I- 'UJ__ u_; LLRWRY PLN OR

_-i _. P.3/4 JUN 1W We;U'_-JHM CHLLRWRY PLRNT OR P. 3/4 CALLAWAY- SP TABLE 6.1-5 DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.37, REVISION 0, DATED MARCH 1973, TITLED "QUALITY ASSURANCE REQUIREMENTS FOR CLEANING OF FLUID SYSTEMS AND ASSOCIATED COMPONENTS OF WATER-COOLED NUCLEAR POWER PLANTS" Regulatory Guide 1.37 Position Union Electric Position The requirements and recommendations for onsite cleaning of materials and components, cleanness control, and preoperational cleaning and layup of water-cooled nuclear power plant fluid systems that are included in ANSI N45.2.1-1973, "Cleaning of Fluid Systems and Associated Components During Construction Phase of Nuclear Power Plants," are generally acceptable and provide an adequate basis for complying with the pertinent quality assurance requirements of Appendix B to 10 CFR Part 50, subject to the following:

1. Subdivision 1.5 of ANSI N45.2.1, 1973 statesthat 1. Complies.

other documents required to be included as a part of the standard are either identified at the point of reference or described in Section 10 of the standard. The specific applicability or acceptability of these listed documents has been or will be covered separately in other regulatory guides or in Commission regulations,'where appropriate.

2. Although subdivision 1.2 of 'ANSI N4.5.2.1-1973 i . oNpI;e states that the requirements promulgated apply during the construction phase of a nuclear power plant, many of the requirements and recommendations contained in C.

0s2R the standard are also'apprf6riate to cleaning of fluid 03-20 systems and associated components durirhg'the operation phase of a nuclear power plant and they should be used when applicable. In this regard, however, it should be particularly noted that decontamination and cleanup of radioactively contaminated systems and components are not addressed by ANSI N45.2.1-1973. These operations will be considered separately in future regulatory guides.

Rev. OL-13 5/03

• 1 "I . i -f 4ano ^ a.

00',

P. 4/4 JUN JLTI 1010 '03

'03 02: 10PM CLLWY 02:1OPM PLANT QA CALLAWAY PL4T CA P. 4/4 CALLAWAY - SP TABLE 6.1-5 (Sheet 2)

Regulatory Guide 1.37 Position Union Electric Position

3. Subdivision 3.2 of ANSI N45.2.1-1973 tes that 3. Co lie the selection of the water quality for a speci C p QQ4! p application shall be made by the organiza on responsible for the cleaning operations u less otherwise r clki t icALb 5 .

specified in the purchase document. Tle water quality for final flushes of fluid systems and sociated components should be at least equi lent to the quality of the operating system water.

4. Section 5 of ANSI N45.2.1-1 73 states, in part, 4. Complie that low sulfur, low fluorine, and/or ow chlorine Ho l t 4 compounds may be used on auste itic stainless steels and that low sulfur and low lead c mpounds may be d L used on nickel-base alloys. Che ical compounds that could contribute to inter-granular cracking or stress-corrosion cracking shoul not be used with austenitic stainless steel and ni kel-base alloys.

Examples of such chemical co pounds are those containing chlorides, fluorides, lead, zinc, copper, sulfur, or mercury where such eleme ts are leachable or where they could be released by bre down of the compounds under expected environmental onditions (e.g., by radiation). This limitation is not intended to prohibit the use of trichlorotrifluoroethane wh h meets the requirements of Military Specificati Mil-C-81 302b for cleaning or degreasing of austenitic inless steel provided the precautions of subdivision .3(4) of ANSI N45.2.1-1973 are observed.'

5. Section 5 of ANSI N45.2.1-197 states, in part, 5. Compgle that operations such as' grinding an welding which ( OQA0 i A a generate particulate matter should e controlled. R 0t I) /

Adequate control of tools used in a rasive work .f3¢O. co.to/

operations such as grinding, sandin, chipping, or wire brushing should be provided. Speci lly, tools which contain materials that could contribute ranular cracking or stress-corrosion cracking or whic because U/

of previous usage may have becorme contarni ated with such materials, should not be used on surfac of corrosion-resistantalloys.

Examples of such materials are listed in Regulat ry Position 4.

Rev. OL-13 5/03 T.Ka, Mn cMn-7 - C - A '74 ACI)CA-In-l Col.,

CALLAWAY - SP 9.3.4.2.1.2 Reactor Coolant Purification and Chemistry Control System Reactor coolant water chemistry specifications are given in Table 5.2-5.

pH CONTROL - The pH control chemical employed is lithium hydroxide. This chemical is compatible with the materials and water chemistry of borated water/stainless steel/

zirconium/ inconel systems. In addition, lithium-7 is produced in the core region due to the irradiation of the dissolved boron in the coolant.

The concentration of lithium-7 in the RCS is maintained in the range specified for pH control (see Table 5.2-5). If the concentration exceeds this range the cation bed demineralizer is employed in the letdown line in series operation with a mixed bed demineralizer. Since the amount of lithium to be removed is small and its buildup can be readily calculated, the flow through the cation bed demineralizer is not required to be full letdown flow. The cation demineralizer is in use approximately one percent of the time (or as necessary). If the concentration of lithium-7 is below the specified limits, lithium hydroxide can be introduced into the RCS via the charging flow, or via alternate suitable flowpaths. The solution is prepared in the laboratory and is added via various installed plant systems or components. Reactor makeup water, or other suitable liquid, may then be used to flush the solution into the RCS.

In preparation for and during plant shutdown conditions, a mixed bed demineralizer loaded with the necessary quantity of lithiated or non-lithiated resin may be placed in service to extend the service life of the cation and remaining mixed bed demineralizer used during the cycle. Non-lithiated resin is also compatible with the materials of construction and water chemistry of the borated water/stainless steel/zirconium/inconel systems.

OXYGEN CONTROL - During reactor startup from the cold condition, and at other times as necessary, hydrazine is employed as an oxygen scavenging agent. The hydrazine solution may be introduced into the RCS in the same manner as described above for the pH control agent.

Dissolved hydrogen is employed to control and scavenge oxygen produced due to radiolysis of water in the core region. A sufficient partial pressure of hydrogen is maintained in the VCT so that the specified concentration of hydrogen is maintained in the reactor coolant. A pre re o y aint um pressure in the vapor space of the volu ontrol tank. This valve can be adjusted t rovide the correct equili gen concentration (25 to 50 cc hydrogen at STP pe ilogram of water).

Hy en is supplied from the hydrogen manifoli the -servicegas s tem.

ixed bed demineralizers are pr ided in the le down line to provide cle up for the letdown flow. The demineralize s remove ionic corrosion productsaeid-c rtain fission produc2 One demineralizer is in continuous service and can be supple nted ermittently by the cation bed demineralizer, if necessary, for additional p rification.

The cation resin removes principally cesium and lithium isotopes from the urification Rev OL-13 Rev9.3-32

.J 5/03 DPI

CALLAWAY - SP flow. The second mixed bed demineralizer serves as a standby unit for use if the operating demineralizer becomes exhausted during operation or for RCS purification purposes in support of plant shutdown conditions.

A further cleanup feature is available for use during cold shutdown and operation of the residual heat removal system (RHRS). A remote-operated valve admits a bypass flow from the RHRS into the letdown line upstream of the letdown heat exchanger. The flow passes through the heat exchanger, a mixed bed demineralizer, and the reactor coolant filter to the VCT. The fluid is then returned to the RCS via the normal charging route.

Filters are provided at various locations to ensure filtration of particulate and resin fines and to protect the seals on the reactor coolant pumps.

Fission gases are removed from the reactor coolant by purging of the VCT to the processing system.

4.2.1.3 Reactor Makeup Control System The soluble neutron absorber (boric acid) concentration is controlled by the BTRS and 0 -tra by the reactor makeup control system which controls the makeup water concentration at a pre-set value between 0 and 7700 ppm boric acid solution. The reactor makeup control system is also used to maintain proper reactor coolant inventory. In addition, for emergency boration and makeup, the redundant capability exists to supply borated water, at a minimum 2350 ppm boric acid concentration, directly from the refueling water storage tank to the suction of the charging pumps. When this source is used for boration, letdown from the RCS is required. Emergency boration utilizing only safety grade equipment is discussed in Appendix 5.4A.

The reactor makeup control system provides a manually preselected makeup concentration of boric acid to the charging pump suction header or to the volume control tank. The makeup control functions are those of maintaining desired operating level in the VCT and adjusting reactor coolant boron concentration for reactivity control. Reactor makeup water and boric acid solution (7000-7700 ppm) are blended together to achieve the desired boron concentration for use as makeup to maintain volume control tank level or to change the reactor coolant boron concentration.

A boron concentration measurement system (see Section 7.7) is provided to monitor the boron content of the reactor coolant in the letdown line. The boron concentration is indicated in the main control room.

Boric acid solution, at 7000 to 7700 ppm boron, is stored in two boric acid tanks. Two boric acid transfer pumps are provided which are capable of supplying the boric acid solution directly to the charging pumps' suction header upon remote manual demand from the main control room. The boric acid transfer pumps are normally aligned to recirculate the boric acid tank contents via the minimum flow lines and will supply boric acid to the boric acid blending tee upon demand of the reactor makeup control system.

9.3-33 Rev. OL-13 5/03

FSAR CN 03-020 INSERT B ZINC ADDITION - A soluble zinc compound may be added to the reactor coolant as a means to reduce radiation fields within the primary system. The zinc used may be either natural zinc or zinc depleted of 64Zn. When used, the target system zinc concentration is normally maintained to a concentration no greater than 40 ppb.

CAILLAWAY - SP TABLE 16.11-8 REPORTING LEVELS FOR RADIOACTIVITY CONCHNTRATIONS IN ENVIROIMENTAL SAMPLES REPORTING LEVELS WATER AIRBORNE PARTICULATE FISH MILK FOOD PRODUCTS n OR GASES (oCi/m)j weti t 3pCi ANALYSIS 20,000 <

H-3 1,000 30,000 r Mn-54 Pe-59 400 10,000 Co-So 1,000 30,000 Co-60 300 10,000 Zr-ITb.95** 400 I-131 2 0.9 3 100 CB-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140 200 300 (a) Multiply the values in this table by lS-9 to convert to units of pCi/ml.

(b) Multiply the values in this table by 15-9 to convert to units of pCi/g.

For drin1sing wzatcr samples. This is 40 CFR Part 141 value. For SUrfacewater samples, a value of 30,000 pCi/1 may be used.

Total activity, parent plus daughter adilviy.

Rev. OL.9 30D 200005197 FSAR c 03N

I a'

CAIjLAWAY - SP TABLE 16..11-9

DETECI'ION CAPABILITrES FOR ENVIRONMLNTrAr SAMPLE ANALYSIS LOWER LIMIT OF DETECTION (LLD) (1), (2)' (3)

SURFACE DRINXING AIRBORNE WATER WATER . MILK FOOD SEDIMENT PARTICULATE PISH.

ANAILYS IS (VCie/2 L Cxcie La OR CASES (Pci/mit (pC/ikq, fpci/e La PRODUCTS (pCi/ka .

wet) b (oCi/kq, drm) b feat) b Gross Beta 4.- 4 0.01 H-3 3000 2000 Mn- 54 15 15 130 Fe -59 30 30 260.

Co-58, 60 15 15 130 Zr*1rb-95 ** 15 15 1-131 1000 0,07 I 60 Co- 134 15 isI 0.05 130 15 60 150 Ca- 137 18 18 O.06 150 18 80 0SO Ba-La-140 15 15 15 (a) Multiply the values in this table by IR-9 to convert to units of pCi/ml.

(b) Multiply the values in this table by 1E-9 to convert to units of pCi/g.

Total activity, parent plus daughter aclivity.

tt2 °30

• 2(d0~ -U Rev. OL-9

• o3 - 09LO 5197

DATE: 10/01/03 UNION ELECTRIC COMPANY PAGE: 1 TIME- 1rs33:18 CALLAWAY RECORDS INFORMATION SYSTEM REPORT: ARCP7301

- CNOT COVER SHEET CN NUMBER: 03-045 REVISION: 000 A210.0028/F126 FINAL SAFETY ANALYSIS REPORT CHANGE NOTICE - U.E.

VENDOR: U050-AMEREN/UE (FORMERLY UNION ELE CNOT TITLE AND DESCRIPTION:

THIS CHANGE NOTICE PICKS UP SEVERAL EDITORIAL CHANGES THAT HAVE BEEN NOTED SINCE THE LAST REVISION TO THE FSAR.

THESE EDITORIAL CHANGES ARE MADE TO CLARIFY THE CONSISTENCY OF THE FSAR.

-w RESP ENGR: I DATE: of/_13

_____3 REVIEWED BY: DATE:

w APPROVED BY

~-Q A4-A.elxJ DATE:

=m TOTAL PAGES: 1 (COMPLETED BY:

ADMIN DEPT)

F11Y1 cA- oT-oq,5 CALLAWAY - SP TABLE 16.11-5 (Sheet 3)

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 39- Deleted.

ACTION 40- With the number of low range channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 41 - With the number of channels OPERABLE one less than required by the I Minimum Channels OPERABLE requirement, restore the affected channel to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the inoperable channel is not restored within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or with no channels OPERABLE, immediately suspend the release of radioactive effluents via this pathway (this Action

must be completed whenever this default condition is entered).

TION 42 - With the number of channels OPERABLE less than required by the Oki - Minimum ChannelsQPERABLE requirement, immediately initiate closure of the~equipm n suspend the release of radioactive effluents via is pathway, or suspend RE ALTERATIONS and movement of irradiated fuel assem blies witn containment.

ACTION 43 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required in Table 16.11-4.

ACTION 44 - Deleted.

ACTION 45 - Flow rate for this system shall be based on fan status and operating curves or actual measurements.

ACTION 46 - For midrange and high range channels only - with the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, take the action specified in Section 16.3.3.4, ACTION C.

ACTION 47 - With the number of channels OPERABLE less than required by the Minumum Channels OPERABLE requirement, immediately suspend the release of radioactivite effluents via this pathway.

Rev. OL-1 3 5/03

- s/us UNION ELECTRIC COMPANY PAGE: 1

13 CALLAWAY RECORDS INFORMATION SYSTEM REPORT: ARCP7301

- CNOT COVER SHEET CN NUMBER: 03-046 REVISION: 000 A210.0028/F126 FINAL SAFETY ANALYSIS REPORT CHANGE NOTICE - U.E.

VENDOR: U050-AMEREN/UE (FORMERLY UNION ELE CNOT TITLE AND DESCRIPTION:

REVISE ACTION 42 OF TABLE 16.11-5 TO CLARIFY REQUIRED ACTIONS

w

=m i_

i_

RESP ENGR: _ DATE: /1-13-2-v, 3

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-M REVIEWED BY: DATE:

im

. .ti APPROVED BY: DATE:

TOTAL PAGES:

(COMPLETED BY:

ADMIN DEPT)

• .. NO CtfRJ&L-FRo 1-JFNWM,4TZroP\I CALLAWAY - SP ON~ U TABLE 16.11-5 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION

1. Unit Vent System

a. Noble Gas -Activity 1 At all times 40,46 Monitor - Providing Alarm (GT-RE-21)

b. Iodine Sampler 1 At all times 43

c. Particulate Sampler At all times 43 1

d. Unit Vent Flow Rate I At all times 45 1

e. Particulate and At all times 43 Radioiodine Sampler Flow Rate Monitor

2. Containment Purge System

a. Noble Gas Activity Monitor

- Providing Alarm and 2 MODES 1,2,3, and 41 Automatic 4.

Termination of Release (GT-RE-22, GT-RE-33)

- Providing Alarm 1 During CORE 42 function only ALTERATIONS or movement of irradiated fuel within the containment

b. Iodine Sampler MODES 1,2,3,4and 43 during CORE ALTERATIONS or movement of irradiated fuel within the containment

c. Particulate Sampler MODES 1,2,3,4and 43 during CORE ALTERATIONS or movement of irradiated fuel within the containment

- d. Containment Purge N/A N/A N/A Ventilation Flow Rate Rev. OL-13 5/03

CALLAWAY - SP TABLE 16.11-5 (Sheet 3)

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 39- Deleted.

ACTION 40- With the number of low range channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 41 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the affected channel to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the inoperable channel is not restored within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or with no channels OPERABLE, immediately suspend the release of radioactive effluents via this pathway (this Action must be completed whenever this default condition is entered).

ACTION 42- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement,fimnediate initial cure othecntainent eipme hatc5and uspe the r ease f ra, ioactje effients ia thi pathwor uspdCO ALT RAT NS

$dmat eme ~tof ifadiat d fuel, sserlies ithin ntain ent ACTION 43- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required in Table 16.11-4.

ACTION 44 - Deleted.

ACTION 45 - Flow rate for this system shall be based on fan status and operating curves or actual measurements.

ACTION 46 - For midrange and high range channels only - with the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, take the action specified in Section 16.3.3.4, ACTION C.

ACTION 47 - With the number of channels OPERABLE less than required by the Minumum Channels OPERABLE requirement, immediately suspend the release of radioactivite effluents via this pathway.

Rev. OL-1 3d 9/03

FSAR-SP CN 03-046 INSERT and if the containment equipment hatch is open, then immediately suspend CORE ALTERATIONS and movement of irradiated fuel assemblies within containment. If the containment equipment hatch is not open, then suspend the release of radioactive effluents via this pathway or immediately suspend CORE ALTERATIONS and movement of irradiated fuel assemblies within containment.