ML20067D250

Semiannual Radioactive Effluent Release Rept Jul-Dec 1993

ML20067D250
Person / Time
Site:	Callaway
Issue date:	12/31/1993
From:	UNION ELECTRIC CO.
To:
Shared Package
ML20067D248	List: ML20067D245 ML20067D250
References
NUDOCS 9403080136
Download: ML20067D250 (250)

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d I I I SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORI l CALLAWAY NUCLEAR PLANT UNION ELECTRIC COMPANY LICENSE NPF - 30 JULY - DECEMBER 1993 I I I I t I J I. I E i l1

l TABLE OF CONTENTS

1.0 INTRODUCTION

I 2.0 SUPPLEMENTAL INFORMATION 2.1 Regulatory Limits I 2.2 Maximum Permissible Concentrations 2.3 Average Energy I 2.4 Mear:urements and Approximations of Total Radioactivity 2.5 Batch Releases 1 2.6 Abnormal Releases 3.0

SUMMARY

OF GASEOUS RADIOACTIVE EFFLUENTS I g 4.0

SUMMARY

OF LIQUID RADIOACTIVE EFFLUENTS 5.0 SOLID WASTES SHIFMENTS 6.0 RELATED INFORMATION 6.1 Unplanned Releases I 6.2 Changes to the Process Control Pro <3ra.m G.3 Changes to the Offsite Dose Calculation Manual 6.4 Major Changes to Radwaste Treatment Systems 6.5 Land Use Census Changes j 6.6 Inoperability of Effluent Monitoring Instrumentation 6.7 Instances of Liquid Holdup Tasks or Waste Decay Tanks Exceeding Technical Specification Limits 7.O METEOROLOGICAL DATJ6

8.0 ASSESSMENT

OF DOSES 8.1 Dose at the Site Boundary and Nearest Residence From Gaseous Effluents 8.2 Dose to the MEMBER OF THE PUBLIC from Activities Within the SITE BOUNDARY 8.3 Total Dose Due to the Uranium Fuel Cycle 8.4 Dose Due to Liquid Effluents 1 I I I

' TABLE OF CONTENTS Table 1A Semiannual Summation of Gaseous Releases Table IB Semiannual Airborne Continuous and Batch Releases Table 2A Semiannual Summation of Liquid Releases Table 2B Semiannual Liquid Continuous and Batch Releases Table 3 So'.id Waste and Irradiated Fuel Shipments Table 4 Cumulative Joint Frequency Distributions Table 5 Dose at the SITE BOUNDARY and Nearest Resident Table 6 Dose to the MEMBER OF THE PUBLIC from Activities within the SITE BOUNDARY Table 7 Total Dose Due to the Uranium Fuel Cycle Table 8 Dose Due tc; Liquid Effluents I I I t i I I I I I

i I 1 1

1.0 INTRODUCTION

This Semiannual Radioactive Effluent Release Report in submitted in accordance with Section 6.9.1.7 of the Callaway Plant Technical Specifications. ~ The report presents a summary of radioactivity released in liquid i and gaseous effluents, and solid waste shipped from the Callaway i ~ Plant during the period from July 1, 1993 to December 31, 1993. The l information is presented in the format outlined in Appendix B of Regulatory Guide 1.21, Revision 1, June 1974. All liquid and gaseous effluents discharged during this reporting period were in compliance with federal regulations and the limits of Union Electric Administrative Procedure APA-ZZ-01003, Offsite Dose d Calculation Manual (ODCM) 2.0 SUPPLEMENTAL INFOPF.ATION l 2.1 Peculatory Limits Specified as follows are the Radiological Effluent Control (REC) limits applicable to the release of radioactive material in liquid and gaseous effluents. 2.1.1 Fission and Activation Gases (Noble Gases) The dose rate due to radioactive noble gases released in gaseous effluents from the site to areas at and beyond the site boundary shall be limited to less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin. The air dose due to ncble gases released in gaseous effluents, from each unit, to areas at and beyond the site boundary shall be limited to the following: a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation and, l b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta j radiation. 2.1.2 Radiciodine, Tritium, and Particulates The dose rate due to Iodine 131 and 133, tritium and all radionuclides in particulate form with half lives greater than eight I (8) days released in gaseous effluents from the site to areas at and beyond the site boundary shall be limited to less than or equal to 1500 mrem /yr to any organ. I 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 eight (8) days in gaseous effluents released to areas at and 1 I beyond the site boundary shall be limited to the following: a. During any calendar quarter: Less than or equal to 7.5 mrem to I any organ and, I - 1 I

I b. During any calendar year: Less than or equal to 15 mrem to any organ. 2.1.3 Liquid Effluents The concentration of radioactive material released in liquid I effluents.to unrestricted areas shall be limited to the l' concentrations specified in Appendix B, Table II, Column 2 to I 10CFR20.001 to 20.601 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the l concentration shall be limited to 2.0E-04 microcuries/ml total i activity. The dose or dose commitment to an Individual from radioactive materials in liquid effluents released to unrestricted areas shall . I be limited: a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and less than or equal to 5 mrem to any organ, and b. During any calendar year to less than or equal to 3 mrem to the whole body and to less than or equal to 10 mrem to any organ. 2.1.4 Uranium Fuel Cycle Sources i The annual (calendar year) dose or d commitment to any member of I< the public due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrem. 2.2 Maximum Permissible Concentrations 2.2.1 The maximum permissible concentration values specified in I Appendix B, Table II, Column 2 to 10CFR20.001 to 20.601 are used to calculate release rates and permissible concentrations of liquid radioactive effluents at the unrestricted area boundary. A value of 2.0E-4 microcuries/ml is used as the limiting concentration for dissolved and entrained noble gases in liquid effluents. 2.2.2 For gaseous effluents, maximum permissible concentrations are not utilized in release rate calculations since the applicable limits are based on dose rate at the site baundary. The " Percent of Tech .I Spec Limit" for Table 1A is therefore not applicable to the Callaway Plant. 2.3 Averace Enercy This is not applicable to the Callaway Plant radiological effluent monitorin3 program. 2.4 Measurements and Acoroximations of Total Radioactivity The quantification of radioactivity in liquid and gaseous effluents I was accomplished by performing sampling and radiological analysis of effluents in accordance with the requirements of Table 9.3-A and Table 9.6-A of APA-ZZ-01003, Offsite Dose Calculation Manual. I Gamma spectroscopy was the primary analysis technique used to determine radionuclide composition and concentration of liquid and l -2 1 I i

I gaseous effluents. Composite samples were analyzed for Sr-89, i Sr-90, and Fe-55 by an independent laboratory. Tritium and alpha j were measured for both liquid and gaseous effluents using liquid

h scintillation counting and gas flow proportional counting

,3 techniques, respectively. 1l The total radioactivity in effluent releases was determined from the i measured concentrations of each radionuclide present and the total {g. volume of effluents d!.scharged. Gross beta or gamma radioactivity i measurement techniques were not utilized to approximate the total i radioactivity in effluents. 2.5 Batch Relaases i 2.5.1 Liquid f Number of batch releases: 128 Total time p7riod for batch releases: 54,094 minutes ) Maximum time period for a batch release: 936 minutes l Average time period f or batch releases: 423 minutes Minimum time period for a batch release: 14 minutes Average stream flow during periods of release of effluent into a flowing stream: 222,439 cfs* (

• Ref: Letter, United States Department of the Interior - Geological i

Survey - Missouri, dated January 10, 1994. 2.5.2 Gaseous Total for the Reoorting Period Number of batch releases: 41 Total time period for batch releases: 10,351 minutes Maximum time period for a batch release: 2,930 minutes Average time period for batch releases: 252 minutes Minimum time period for a batch release: 42 minutes 2.6 Un_ planned Rehases 2.6.1 Licuid Nutr.ber of releases: 0 'l Total Activity released: 0 2.6.2 gaseous Number of releases: 1 Total Activity released: 4.6E-03Ci I 4 I

1 I 1 .lE 3.0

SUMMARY

OF GASEOUS RADIOACTIVE EFFLUENTS l 3.1 The quantities of radioactive material released in gaseous effluents I are summarized in Tables 1A and IB. Note that for this reporting 1 period no gaseous effluents were considered as elevated releases. 4.0

SUMMARY

OF LIOUID PADI_QACTIVE EFFLUENTS 4.1 The quantities of radioactive material released in liquid effluents are summarized in Tables 2A and 2B. 5.0 SOLID WASTES 5.1 The quantities of radioactive material released in shipments of solid waste for burial and irradiated fuel transported from the site I during the reporting period are summarized in Table 3. The activity and fractional abundance of each nuclide was determined for each waste type by an independent laboratory based upon radiochemical analysis of samples of that waste type. The curie amount of each I nuclide listed in Table 3 was determined as the product of the fractional abundance and the total curies shipped. Those nuclides which comprise at least 1% of the total activity for a particular waste type are presented in Table 3. 6.0 RELATED INFORMATION 6.1 Unclanned Releases Unplanned releases are: 1) Inadvertent or accidental releases of i radioactive material; 2) Releases of radioactive material via nomal I pathways without a release permit, proper authorization, or proper sampling and analysis; and 3) Releases which are conducted in such a manner as to result in significant deviation from the requirements of the release permit. There was one unplannec release during the reporting period. On November 4, 1993, the auxiliary boiler was sampled and found to I be contaminated. Isotopic analysis indicated that I-131 was present at a concentration of 2.24E-8 pCi/ gram. The plant operations department was notified immediately to ensure I that area drains and boiler blowdown were routed to the liquid radwaste system. A 10CFR50.59 evaluation was initiated immediately along with an I investigation to determine the source of the -tivity which had concentrated in the auxiliary boiler. An evaluation was previously performed on auxiliary boiler I contamination assuming a maximum I-131 concentration of 5.0 E-08 pCi/ml. The resulting dose to a member of the public was calculated to be 2.8E-02 mrem to the child age group which is insignificant with respect to the quarterly and annual critical I organ dose limits of 7.5 and 15 mrem, respectively. The dose rate at the site boundary was calculated to be 1.35E-03 mrem /yr which is also well below the release rate limit of 1500 mrem /yr. This evaluation bounds the current operation of the auxiliary boiler with contamination identified. I ) i 1 l l

I 3 The post release evaluation showed that 3.9E-05 Ci of I-131 and 4.6E-03 Ci of Tritium was released to the environment. The dose to a member of the public from this release was calculated to be 3.3E-I 03 mrem which is negligible when compared to the quarterly and annual effluent control limits. As a result, the release of radioactive material from the operation of the auxiliary boiler did not endanger the health or safety of the public or the environment. The activity released is reported in Table 1A and 1B for this reporting period and will be included in the annual dose calculations (see Section 7.0). The safety evaluation completed for this event was documented in Request for Resolution (RFR) 14435 Rev. A. This event was also described by Suggestion Occurrence Solution (SOS) 93-1763. 6.2 Chances to the Process Control Program There were no changes made to Administrative Procedure APA-2Z-01011, " Process Control Program Manual", during the reporting period. 6.3 Changes to the Offsite Dose Calculation Manual I Revision 3 of Administrative Procedure APA-ZZ-01003, Offsite Dose Calculation Manual (ODCM) was approved September 23, 1993. The major changes incorporated in this revision were deletion of the turbine building drain and secondary liquid waste system liquid radiation monitors as effluent release monitors, the relocation of the reporting requirements for solid radwaste which are now described in APA-ZZ-01011, Process Control Program, and addressing compliance with 10CFR20.1301. A complete copy of Administrative Procedure APA-ZZ-01003, Revision 3 is included as Attachment 1. 6.4 Maior Chances to Radwaste Treatment Systemq During the reporting period, there were no changes to the plant l which would be considered a major change to a Liquid, Gaseous, or j Solid Radwaste Treatment System. 6.5 Land Use Census Chances There were no changes in critical receptor locations for dose i calculations during the reporting period. j 6.6 Inonerability of Effluent Monitorina Instrumentation All effluent monitoring instrumentation was OPERABLE within the limits specified by APA-ZZ-01003, Section 9.1.1 and 9.2.1 during the ] reporting period. 1 I E I-

I l 6.7 Instances of Liouid Holdup Tanks or Waste Gas Decav Tanks Excegdip_g Technical SDecification Limits i All liquid tanks and waste gas decay tanks were within the limits of Specifications 3.11.1.4 and 3.11.2.6 during the reporting period. 7.0 METEOROLOGICAL DATA The cogletion of the verification and validation of the new meteorological data processing software by a meteorologist (reference NRC inspection report unresolved item 50-483/93008-01) I was'not completed in time for incorporation of the 1993 meteorological data into this report. The data will be submitted as an addendum to this report when the verification and validation cf the new meteorological data processing software is completed. 8.0 ASSE_S_SEENT OF DDSES The assessment of doses to the maximum exposed individual from I gaseous effluents is not included in this report due to the verification and validation of the new meteorological data processing software which has not been completed as described in Section 7.0. The assessment of doses to the maximum exposed individual from liquid effluents was performed for locations representing the maximum dose. For liquid effluents, doses were well below Technical i Specification limits. The assessment of doses from gaseous effluents will be submitted as an addendum to this report with the meteorological data. 8.1 Dose at the SITE BOUNDARY From Gaseous Effluents An assessment of doses from gaseous effluents is performed in accordance with Administrative Procedure APA-ZZ-01003 for the maximum exposed individual at the SITE BOUNDARY location with the I highest ground level concentration of radioactive material, based upon actual meteorological conditions existing during the year. Doses are assessed at each location considering noble gas I submersion, inhalation and ground plane pathways. This assessment is performed for each age group, with the Child age group receiving the highest dose. The calculations for the SITE BOUNDARY location conservatively I assumed a hypothetical maximum exposed individual. The results of the assessment for the Child age group will be provided in an addendum to this report. 8.2 D_ggp at the Nearest Residence From Gaseous Effluents An assessment of doses from gaseous effluents is performed in accordance with Administrative Procedure APA-ZZ-01003 for the I maximum exposed i:Aividual at the Nearest Residence location with the highest ground level concentration of radioactive material, based upon actual meteorological conditions existing during the year. Doses are assessed at each location considering noble gas I submersion, inhalation, ground plane, and ingestion pathways. The ingestion pathways considered are produce, vegetable, goat's milk, cow's milk, and meat pathways. This assessment is performed for the Child age group. B . 6 - B

I The results of the assessment for the Child age group will be provided in an addendum to this report. The calculations for the Nearest Residence are for a "real" individual. It is conservatively I assumed that each ingestion pathway exists at the Nearest Residence location, and that the Child age group exists at each location. 8.3 Dose to the MEMBER OF THE PUBLIC from Activities Within the SITE I BOUNDARY The assessment of dose to the MEMBER OF THE PUBLIC from activities within the SITE BOUNDARY is performed in accordance with I Administrative Procedure APA-ZZ-01003 Section 4. The dose to the MEMBER OF THE PUBLIC from activities within the SITE BOUNDARY will be provided in an addendum to this report. I 8.4 Total Dose Due to the Uranium Fuel Cvele 4 Since there are no other Uranium Fuel Cycle facilities within 8 km of the Callaway plant, the total dose to the most likely exposed 8 MEMBER OF THE PUBLIC results from direct radiation and radioactive effluents from the Callaway Plant.. The methodology for assessing this dose is d. scribed in Administrative Procedure APA-ZZ-01003 Section 4. l The Total Dose from the Uranium Fuel Cycle is evaluated for the MEMBER OF THE PUBLIC who may use portions of the area within the SITE BOUNDARY for purposes not associated with plant operations. The Total Dose to the MEMBER OF THE PUBLIC (Table 7) is the sum of the dose due 1o activities within the SITE BOUNDARY (Table 6) and the dose due to gaseous effluents at his residence (Table 7). The Total Dose at the Nearest Residence is due to the dose from gaseous effluents, assuming that each food ingestion pathway exists at this location (Table 5) In each case, the whole body gamma dose from Noble Gases and ground plane exposure is added to the organ dose from the inhalation and ingestion pathways, The Total Dose from the Uranium Fuel Cycle will be presented in an a addendum to this report. B.S Dose Due to Licuid Effluents I The total dose to the maximum exposed individual from liquid 4 effluents released from the Callaway Plant during the year is presented in Table 8. I 1 I 1 ', g + r.4- - -... ~, w-_--,

LI L TABLE 1A 4 SEMIANNUAL SUMMATION OF GASEOUS RELEASES l ALL AIRBORNE EFFLUENTS L OUARTERS 3 AND 4.1993 I 1 f r THIRD FOURTH EST TOTAL' TYPE OF EFFLUENT UNITS OUARTER QUARTER ERROR % 4 A. FISSION AND ACTIVATION G ASES j

1. TOTAL RELEASE CURIES 1.42E+ 02 3.99E+ 02 20

2. AVERAGE RELEASE RATE FOR PERIOD vCi/SEC 1.78E+ 01 5.02E+01

3. PERCENT OF TECH SPEC LIMIT N/A N/A B. RADIOIODINES j g i 3

1. TOTAL IODINE-131 CURIES 2.10E-04 2.96E-04 23 g

i

2. AVERAGE RELEASE RATE FOR PERIOD pCi/SEC 2.64 E-05 3.73E-05

3. PERCENT OF TECH SPEC LIMIT N/A N/A 1'.

C. PARTICULATES

1. PARTICULATE (HALF-LIVES > 8 DAYS)

CURIES 1.27E-07 4.07E-06 30 {

2. AVERAGE RELEASE RATE FOR PERIOD pCi/SEC 1.60E-08 5.12E-07

3. PERCENT OF TECH SPEC LIMIT N/A N/A I

I

4. GROSS ALPHA RADIOACTIVITY CURIES 2.92E-07 4.52E-07

'I D. TRITIUM

1. TOTAL RELEASE CURIES 1.72E+ 01 3.74 E+ 01 14

2. AVERAGE RELEASE RATE FOR PERIOD efi/SEC 2.17E+ 00 4.70E + 00

3. PERCENT OF TECH SPEC LIMIT N/A N/A I

PAGE 1 OF 1 I I-

l TABLE 18 SEMIANNUAL AIRBORNE CONTINUOUS AND BATCH RELEASES GROUND LEVEL RELEASES I FISSION GASES, IODINES, AND PARTICULATES s' OUARTERS 3 AND 4,1993 7 CONTINUOUS RELEASES BATCH RELEASES I 4 THIRD FOURTH THIRD FOURTH NUCLlDE UNIT QUARTER OUARTER QUARTER QUARTER i g ~1. FISSION G ASES KR-85M CURIES 3.11 E-01 0.00E+00 1.80E-03 0.00E+00

g KR-85 CURIES 0.00E+00 0.00E+ 00 3.55E+ 00 0.00E+ 00 XE-131 M CURIES 0.00E+00 5.98E+ 00 5.91 E-01 7.32E-01 XE-133 CURIES 9.97E+ 01 3.54 E+ 02 3.02E+ 01 3.32E+ 01 I

XE-133M CURIES 0.00E+00 4.02E+ 00 1.80E-01 2.29E-01 i j XE-135 CURIES 6.56E+ 00 6.22E-02 1.42E-01 1.91 E-01 1 XE-135M CURIES 3.89E-01 0.00E4 00 0.00E+00 0.00E+ 00 AR-41 CURIES 0.00E+00 0.00E+ 00 1.58E-01 2.94 E-01 l l j TOTAL FOR PERIOD CURIES 1.07E+02 3.64E+02 l 3.48E+ 01 3.46E. 01 !l

2. IODINES i

!E l-131 CURIES 2.10E-04 2.57E-04 1.40E-07 3.91E-05 l i 1-132 CURIES 2.04E-05 2.06E-04 2.02E-08 0.00E+00 f I-133 CURIES 3.18E-04 1.06E-05 2.1 CE-07 0.00E+ 00 l-135 CURIES 5.44E-05 0.00E+ 00 0.00E+00 0.00E+00 TOTAL FOR PERIOD CURIES 6.03E-04 4.74 E-04 3.76E-07 3.91 E-05 I I ?. PARTICULATE. CR-51 CURIES 0.00E+00 6.23E-07 0.00E+ 00 0.00E+ 00 CO-58 CURIES 0.00E+00 3.15 E-06 0.00E+00 0.00E+00 I CO-60 CURfES 0.00E+00 1.59E-07 0.00E+ 00 0.00E+ 00 RB-88 CURIES 0.00E+00 0.00E+00 1.27E-07 0.00E+ 00 NS-95 CURIES 0.00E+00 1.36E-07 0.00E+ 00 0.00E+00 G ALPHA CURIES 2.91 E-07 4.52E-07 1.49E-09 5.01E-11 I TOTAL FOR PERIOD CURIES 2.91 E-07 4.52E-06 1.28E-07 5.01 E-11

4. TRITIUM H-3 CURIES 1.fAE+01 3.63E+ 01 8.30E-01 1.08E+ 00 I

PAGE 1 OF 1 I I

- ~. TABLE 2A SEMIANNUAL SUMMATION OF LIOUlD RELEASES ALL LIQUID EFFLUENTS OUARTERS 3 AND 4,1993

I 4

THIRD FOURTH EST TOT AL' l TYPE OF EFFLUENT UNITS OUARTER OUARTER ERROR % i A. FISSION AND ACTIVATION PRODUCTS j

1. TOTAL RELEASE (NOT INCLUDING l

TRITIUM, G ASES, ALPHA) CURIES 1.12E-02 4.89E-03 20 I }

2. AVERAGE D! LUTED CONCENTRATION DURING PERIOD vCi/ML 1.82E-08 1.09E-08 l

3. PERCENT OF APPLICABLE LIMIT N/A N/A

!l B. TRITIUM i i

1. TOTAL RELEASE CURIES 4.16E+02 1.12E+ 02 14 i

2. AVERAGE DILUTED CONCENTRATION i

DURING PERIOD vCi/ML 6.76E-04 2.4 BE-04 }

3. PERCENT OF APPLICABLE LIMIT N/A N/A E

1 C. DISSOLVED AND ENTRAINED G ASES ~

1. TOTAL RELEASE CURIES 7.87E-01 3.01 E-02 27 i

2. AVERAGE DILUTED CONCENTRATION DURING PERIOD vCi/ML 1.28E-06 6.67E-08 D. GROSS ALPHA RADIOACTIVITY

1. TOTAL RELEASE CUR!ES 0.00E400 1.77E-03 29 E. WASTE VOL RELEASED (PRE-DILUTION)

GAL 6.69E+ 06 5.12E+06 10 I F. VOLUME OF DILUTION WATER USED GAL 1.56E+ 08 1.14E+ 08 10 PAGE 1 OF 1 I 1

j 4 c TABLE 2B t!- SEM! ANNUAL LIQUID CONTINUOUS AND BATCH RELEASES ) TOTALS FOR EACH NUCLIDE RELEASED OUARTERS 3 AND 4,1933 f CONTINUOUS RELEASES BATCH RELEASES THIRD FOURTH THIRD FOURTH NUCLIDE UNITS OUARTER OUARTER OUARTER OUARTER E i

1. ALL NUCLIDES l

H-3 CUR!ES 0.00E+00 0.00E+00 4.16E+ 02 1.12E+ 02 3'l3 ( CR-51 CURIES 0.00E+ 00 0.00E+ 00 0.00E+ 00 1.59E-05 1 MN-54 CURIES 0.00E+00 0.00E+ 00 5.93E-05 9.06E-05 i FE-55 CURIES 0.00E+00 0.00E+ 00 516E-03 2.17E-03 FE-59 CURIES 0.00E+00 0.00E+ 00 0.00E+ 00 1.44 E-05 i CO-57 CURIES 0.00E+00 0.00E+ 00 0.00E+ 00 2.78E-06 CO-58 CURIES 0.00E+00 0.00E+ 00 1.09E-05 9.1 BE-04 5 CO-60 CUR!ES 0.00E + 00 0.00E+ 00 913E-04 9.25E-04 l l ZN-65 CURIES 0.00E+00 0.00E+ 00 0.00E+ 00 4.61 E-05 l 5 SR-89 CURIES 0.00E+00 0.00E+ 00 4.32E-03 0.00E+00 SR-90 CURIES 0.00E+ 00 0.00E+ 00 2.80E-04 0.00E+ 00 l ZR-95 CURIES 0.00E+ 00 0.00E+00 0.00E+00 1.35 E-04 { NB-95 CURIES 0.00E+00 0.00E+00 0.00E+00 1.9BE-04 TC-99m CURIES 0.00E+00 0.00E+00 2.13E-06 0.00E+00 4 l SB-124 CURIES 0.00E+00 0.00E+00 0.00E+ 00 3.36E-06 i lu CS-134 CURIES 0.00E+00 0.00E+00 1.75E-04 1.59E-04 s CS-137 CURIES 0.00E+00 0.00E+ 00 2.43E-04 1.26E-04 i CE-141 CURIES 0.00E+00 0.00E+ 00 0.00E+00 3.34 E-06 jg CE-144 CURIES 0.00E+00 0.00E+00 0.00E+00 5.03E-05 l g' I-131 CURIES 0.00E+00 0.00E+ 00 2.38E-05 3.49E-05 j XE-131M CURIES 0.00E+ 00 0.00E+ 00 4.92E-03 5.28E-04 j XE-133 CURIES 0.00E+ 00 0.00E+ 00 7.69E-01 2.94 E-02

• l XE-133M CURIES 0.00E+00 0.00E+00 9.59E-03 1.41 E-04 W

XE-135 CURIES 0.00E+00 0.00E+ 00 3.73E-03 0.00E+00 G ALPHA CURIES 0.00E+00 0.00E+ 00 0.00E+ 00 1.77E-03 I I I I TOTALS FOR PERIOD CUR!ES 0.00E+00 .00E+00 4.17E+ 02 1.12E+ 02 PAGE 1 OF 1 I l a

4 .I TABLE 3 SQLID WASTE & IRRADIATED FUEL SHIPMENTS i OUARTERS 3 & 4. 1993 A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR QISPOS% (DOES NOT INCLJDE j.. IRRADIATED FUEL) d TYPE OF WASTE 6-MONTH EST. TOTAL j PERIOD _ ERROR ( %)_ a. Spent resins, filter sludges 21.93 m* { evaporator bottoms, etc. 4.29E+02Ci 125% i Percent Abundance Curies i Fe-55 25.4834 1.09E+02 Co-60 25.083% 1.07E+02 Ni-63 11.617% 4.99E+01 i Cs-137 11.317% 4.86E+01 Co 58 10.095% 4.34E+01 Cs-134 8.218% 3.53E+01 l Mn-S4 2.743% 1.18E+01 Sb-125 1.127% 4.84E+00 b. Dry compressible waste, 17.96 m8 contaminated equipment, etc. 1.70E-01 Ci 125% Co-60 54.100% 9.37E-02 Ni-63 26.000% 4.50E-02 l Mn-54 12.200% 2.11E-02 j Fe-55 2.220% 3.84E-03 i Sb-125 2.040% 2.53E-03 f Co-58 1.450% 2.51E-03 c. Irradiated components, control 1.7 m8 j rods, etc. 2.8 Ci 1 d. Other: Solidified oil and oil sludges O m8 4 i 0 Ci f Solid Waste Discosition Number of Mode of Class of Solid Type of a i Shioments Transportation Destination Waste Shioned C_pntainer i 2 Cask Barnwell, SC A LSA 3 Cask Barnwell, SC B LSA } 3 Cask Barnwell, SC C LSA i 1 Truck Oak Ridge, TN A LSA (QUADREX) 9 Truck Oak Ridge, TN (SEG) A LSA I I I I Page 1 of 2 I

..c. ? !I lI TABLE 3 (cont'd i SOLID WASTE & IRPADIATED Fbst SHIPMENTS 7-l OUARTEP.S 3 & 4. 1993 lm i Solidification Agent i Cement (applicable to waste type a only) B. IEP1.DIATED FUEL S11IPMENTS (DISPOSITION) Ilu;rther of Shioments Mode of Transoortation Destination 0 N/A N/A I I i !I 4 1 i !I ii I I I I g e., _ I

TABLE 4 g CUMULATIVE JOlvr FREQUENCY DISTRIBUTIONS i i 5 1993 4 1 I The meteorological data will be submitted as an a Idendum to this report at a later date as ) discussed in Section 7.0. ) ,i t !I l l ! 81r i 4 2: I i ll I I I I I I I I I

TABLE 5 i

E DOSE AT Tile SITE BOUNDARY AND TO TIIE NEAREST RESIDENT g

FROM GASEOUS EFFLUENT 1993 i The dose at the site boundary and to the uearest resident from gaseous emuents will be submitted as an addendum to this report at a later date as discussed in Section 8.0. !I t I I J I I I I I I I I I

4 !I TABLE 6 DOSE TO TiiE MEMBER OF THE PUBLIC FROM ACTIVITIES l WITHIN THE SITE DOUNDARY 1993 !I i The dose to the MEMBER OF THE PUBLIC from activities within the site boundary wili j be submitted as an addendum to this report at a later date as discussed in Section 8.0. i !I i I I I I I I I I I

~, i-TABLE 7 1 g TOTAL-DOSE DUE TO THE URANIUM FUEL CYCLE l 3 (MEMBER OF THE PUBLIC) 1993 The total dose from the uranium fuel cycle will be submitted as an addendum to this report l at a later date as discussed in Section 8.0. iI I I I I I I a I I I I I I I I 1

I TABLE 8 DOSE DUE TO LIQUID EFFLUENTS (MEMBER OF THE PUBLIC) 1993 E ORGAN UNITS DOSE LIMIT * % LIMIT I

l. BONE MREM 9.llE-03 10.00

9. l l E-02 J

2. LIVER MREM 1.05E-02 10.00 1.05E-01 I

3. TOTAL BODY MREM 9.24 E-03 3.00 3.08E-Ol I

4. TllYR01D MREM 3.03E-03 10.00 3.03E-02

5. KIDNEY MREM 5.4SE-03 10.00 5.48E-02

6. LUNG MREM 3 84E-03 10.00 3.84E-02

7. GI-LLI MREM 6.57E-03 10.00 6.57E-02 I

• Annual dose limits of APA-ZZ-01003, Section 9.4.1.1.

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\\E ATTACllM ENT 1 APA-Z7 01003, OFFSITC DOSE CAL.CULATION MANUAL, REVISION 3 I I I I I I I I I I I p 3

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I APA-224)l003 Rev. 3 September 13,1993 NUCLEAR FUNCTION ADMINISTRATIVE PROCEDURE APA-ZZA)l003 OFFSITE DOSE CALCULATION MANUAL I RESPONSIBLE DEPARTMENT b EATW i 5 ' <- 5 I W,C, DATE GrPTmS era. 15, i M 3 PREPARED BY APPROVED EY O'72//d DATE 9 - 2 ? -f 3 t I / I DATE ISSUED Q-A7-43 I I This procedure centains the following: l Pages I through 127 Attachments 1 through 2 Figures I through 5 1 Appendices A through A CheckefrLists through I I /IgWco,,% n iss"" I ga.Ot;:D kb's ] kER _ i il3. l I I

I APA-ZZ-01003 . I Rev. 3 TABLE OF COtITENTS I Purpose and Scope. .. ] I Liquid EfDuents. .2 Liquid Emuent Monitors.. .2 Calculation of Liquid EiDuent Monitor Setpoints.. .3 Liquid Emuent Concentration Measurements.. .6 I Dose Due to Liquid EfDuents. .6 The Maximum Exposed Individual. .6 Calculation of Dos.e from Liquid Emuents. .7 I Summary, Calculation of Dose Due to Liquid Emuents.. .8 Liquid Radwaste Treatment System. .9 .. j a Gaseous Emuents.. Determination of Gaseous Emuent Monitor Setpoints... .16 I Total Body Dose Rate Setpoint Calculations. .16 Skin Dose Rate Setpoint Calculations... . 17 Noble Gases. . 18 l Radionuclides Other Than Noble Gases.. .18 E Dose Duc To Gaseous Emuents.. .19 Noble Gases. .19 Radionuclides Other Than Noble Gases.. .20 I Gaseous R.'dwaste Treatment System., .. 21 Dose and Dose Commitment from Uranium Fuel Cycle Sources.. . 56 Calculation of Dose and Dose Commitment from Uranium Fuel Cycle Sources., .56 I Identification of the MEMBER OF 7EE PUBLIC.. .56 Total Dose to the Nearest Resident.. ..56 Total Dose to the Critical Receptor Within the SITE BOUNDARY. . 57 Radiological Emironmenta! Monitoring. . 60 I Desenption of the Radiological Emironmental Monitoring Program.. .60 Performance Testing of Emironmental Thermoluminescence Dosimeters. .60 Determination of Annual Average and Short Term Atmospheric Dispersion Parameters.. . 73 Atmospherie Dispersion Parameters.. . 73 Ieng Term Dispersion Estimates. . 73 Determination of Long Term Dispersion Estimates for Special Receptor Locations. . 73 Short-Term Dispersion Estimates.. . 74 Reponing Requirernents.. . 81 Annual Radiological Emirocmental Operating Repon.. .'. 8 ! ) Semi-Annual Radioactive Emuent Release Report. . 81 Implementation of ODCM Methodology. .. 84 Radioactive Emuent Controls (REC).. . 85 Radioactive Liquid Emuent Monitoring Instrumentation.... ... 86 I Radioactive Gaseous Emuent Monitoring Instrumentation.. . 91 Liquid Emuents Concentradon. . 97 Dose from Liquid Emuents. .100 Liquid Pmdraste Treatment System.. .. 101 Gaseous Emuents Dose Rate. .102 Dose Noble Gases. .105 Dose - Iodine-131 and 133, Tritium, and Radioactive hbterial in Paruculate Form.. .106 I Gaseous Radwaste Treatment System.. . 107 Total Dose.. .108 Radiological Environmental Monitoring Program. .109 Radiological Ervironmental Monitoring Land Use Census.. ..I19 -j.

I APA-ZZ-01003 Rcv. 3 Tql.E OF CONTENTS I Radiological Environmental Monitoring Interlaturatory Comparison Program. .120 Administrative Controls.. .121 Major Changes to Liquid and Gascous Radwaste Treatment Systems. . 121 I Changes to the OfIsite Dose Calculation Manual (ODCM).. .121 Referenc es.. .123 Figure 4.1. .59 I Figure 5.1 A.. .61 Figure 5.1B.. . 62 Figurc 5.2A.. . 63 I Figure 5.2B.. . 64 Figure 3.3. .. 6 5 Table 2.1. .. 10 Table 2.2.. .13 Table 3.1. .22 Table 3.2.. .23 I Table 3.3.. .26 Table 3.4.. .. 41 Table 5.1. .66 Table 6.1. . 76 I Table 6.2. .77 Table 6.3.. . 78 Table 6.4.. .79 ll Table 6.5.. . 80 t us Table 9.1-A.. .87 f Table 9.1-B. .89 i Table 9.2-A.. . 92 !g l~3 Table 9.2-B. . 95 Table 9.3-A.. .98 Table 9.6 A.. .103 i Table 9.11-A.. . 111 != Table 9.11-B. . 116 l Table 9.ll-C.. .117 . - Lower Limit of Detection (LLD). .1Page

f. - Bases for Radiological E1Iluent Controls..

. 8 Pages Appendix A - Summary Review Of Radiological EfTluent Tech S;v:c Potentially Affected By The Implementation Of The Revised 10 CFR 20. .16 Pages (

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APA-ZZ ClOO3 Rev.3 RECORD OF REVISIONS Rnision Number Date Reason for Rnision Rev.O March 1983 R ev.1 November 1983 Revised to support the current RETS submittal nd to incorporate NRC Staff comments. .g 5 Rev.2 March 1984 Revised to incorporate NRC Staff comments. .. Rev. 3 June 1985 Revised to incorporate errata identified by ULhPC-803 and changes to the Emironmental Monitoring Program. Incorporate results of 1984 land Use Census. Rev. 4 February 1987 Minor clarifications, incorporated 31-day projected dose methodology. Change in the utilization of arcas within the Site Boundary. Rev.5 January 1988 Minor clarificadont, revised descriptions ofliquid and gaseous rad monitors, revised liquid setpoint I methodology to incorporate monitor background, rnised dose calculations for 40CFR190 requirements, Revised Table 6 and Figures 5.1 A and 5.lE to refine descriptions of emironmental TLD stations, I incorporated description of emironmental TLD testing required by Reg. Guide 4.13, rnised Tables 1,2,4 and 5 to add additional nuclides, deleted redundant material from Chapter 6. Rev.6 May 1989 Revised methodology for calculating maximum g permissible liquid efDuent discharge rates and liquid 3 -fIluent discharge rates and liquid eUluent mordtor sctpoints, provided methodology for calculating liquid efDuent monitors response correction factors, provided I an enhanced description of controls on liquid monitor background limits, provided additional liquid and gaseous dose conversion factors and bioaccumulation -l factors (Tables 1,2,4 & 5), prmided description of E the use of the setpoint required by Technical Specification 4,9.4.2 during Core Alterations, added discussion of gaseous and liquid monitor serpoint I selection in the event that the sample contains no detectable activity, added minimum holdup requirements for Waste Gas Decay tanks, rnised I dispersion parameters and accompanying description I per FSAR Change Notice 88-42. i -iii-

i I APA-ZZ-01003 Rev.3 RECORD OF REVISIONS 4 Revision Humber Date Reasc,n foyJevision APA-ZZ.010C0 August 1989 Radiological EfDuent Technical SpeciDeations u cre i Rev.O moved from the Callaway Plant Techidcal Specifications to Section 9.0, Radioactive EUluent Controls, of the ODCM as per NRC Generie Lctier 89-

01. At the same time, in order to formalize control of the entire ODCM, it was converted to APA-ZZ41003, OFFSITE DOSE CALCULATION MANUAL

{ Rev 1 October,1990 Revise Action 41 of Table 9.2-A to allow continued purging for 24 hours as per Amendment 20 to op: rating license, issued 4/10/S7, Rev.2 May,1991 Section 2.4.2 - Changed gross alpha analysis I frequency from "each batch" to a monthly compo:ite as per Table 9.3-A, and the Callawiy Plant NPDES permit (reissued March 15,1991). Rev.3 June,1993 Deleted HF-RE-45 and LE-RE-59 as efDuent morNrs. Revised table numbering for consistency j wit). xse in Section 9.0, deleted redundant material, I incorporated 1992 Land Use Census results, muved LLD description to Attachment 1, moved REC Bases to Attachment 2. Deleted reporting requirements for I 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 Dow monitor setpoint. Resised "MPC" terminology to "ECV". Added Action 46 to REC 9.2 to clarify actions for inoperable I mid and high range WRGM Channels. Resised references to be consistent with the revised 10 CFR 20. Added Appendix A. Resised Action 41 of Rec 9.2 and I the operability requirements of GT-RE-22/33. Incorporated the revised R; values in Tables 3.2 and 3.3. Added Section 6.2 and Table 6.5. I I I- -j v. I A

I APA-ZZ 01003 1 Rev.3 I OFFSITE DOSE CALCUL ATION M ANUAL 1.0 PURPOS.E AND SCOPE I The OFFSITE DOSE CALCULAT]ON MANUAL (ODCM) describes the methodology and parameters used in the calculation of offsite doses resulung from radioactive gaseous and liquid I emuents,in the calculadon of gaseous and liquid emuent monitoring Alarm / Trip Setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM also contains the Radioacdve Emuent Conuols and Radiological Environmental Monitoring Progran s I'~ required by Technical Specificadon 6.8.4,. and descriptions of the information that shou'd be included in the Annual Radiological Environmental Operating and Semi-annual Radioacdve Emuent Release Reports required by Technical Specifications 6.9.1.6 and 6.9.1,7. The ODCM also contains a list and descripdon of the specific sample locations for the radiologbil emironmental monitoring program. Compliance with Radiological Emuent Controls limits demonstrates compliance with the lindts of 10 CFR 20.1301. (Ref. I1 1.1,11.2.1,11.23.3) (CTS #4121) I I I I I I 1 I I 1-I

I APA-ZZW 1003 Rev.3 5 l 2.0 POUID EFFLUENTS 2.1 MOUTD EFFLUENT MONITORS Grost radioactivity monitors which provide for automatic termination ofliquid emuent releases are present on the liquid emuent lines. Flow rate measu:cment devices are present on the liquid efDuent lines and the discharge line (cooling tower blowdown). Setpoints, precautions, and I limitations applicable to the operation of the Callaway Plant liquid emuent monitors are provided in the appropriate Plant Procedures. Setpoint values are calculated to assure that alarm and trip actions occur prior to exceeding the Emuent Concentration Values (ECV) limits in 10 CFR Part 20 at the release point to the UNRESTR:CTED AREA. The calculated alarm and trip action setpoints for the liquid emuent line momtors and Dow measuring devices must satisfy the following equation: I Cf sC (2.1) F+f Where: C= The liquid emuent concentration value (ECV) implementing REC 9.3.1.1 for the site in (pCi/ml). The serpoint, in (pCi/ml), of the radioactisity monitor measuring the radioactivity c= concentration in the emuent line prior to dilution ar.d subsequent release; the setpoint, which is inversely related to the volumetric fic,. of the cmuent line and directly related to the volumetric Dow of the dilution sveun plus the emuent steam, represents a value, w hich, if exceeded, would result in concentrations exceeding the values of 10 CFR Pan 20 Appendix B, Table II, Column 2,in the UNRESTRICTED AREA. f= The flow setpoint as measured at the radiation monitor location, in volume per unit time, but in the same units as F, below. l F= The dilution water Dow rate setpoint as measured prior to the release point, in volume per unit time. (If(F)is large compared to (f), then F + f a F). (Ref. I1.8.1) If no dilution is provided, then e 5 C. The radioactive liquid waste strearn is diluted by the plant discharge line prior to entry into the Missouri River. Normally, the dilution Dow is obtained from the cooling tower blowdown, but I should this become unavailable, the plant water tr:atment facility supplies the necessary dilution Dow via a bypass line. The limiting concentration which corresponds to the liquid radwaste emuent monitor setpoint is to be calculated using methodology from the expression abos e. l Thus, the expression for determining the setpoint of the liquid radwaste emuent line monitor becomes: I C(F + f) (ACi/mi) (2.2) cs f I 2-B

I APA-ZZ-01003 Rev,3 I The alarm / trip setpoint calculations are based on the minimum dilution Dow rate (corresponding to the dilution Dow rate setpoint), the nuximum emuent stream Dow rate, and the actual isotopic analysis. Due to the possibility of a simultaneous release from more than one release pathway, a I portion of the total site release limit is allocated to each pathway. The determination and tragi cf the allocation factor is discussed in Section 2.2. In the event the alarm / trip serpoint is reached, an evaluation will be performed using actual dilution P.nd emuent Dow values and actual isotopic analysis to ensure that REC 9.3.1.1 limits were not execeded. 2.1.1 Continuous Liould Emuent Monitors I l The radiation detection monitor associated with continuous liquid emuent releases is (Ref. I161,11.6.2): Monitor i D Description BM-RE-52 Stam Generator Blowdown Discharge Monitor I i l The Steam Generator Blowdown disclurge is not considered to be radioactive unless radioactisity I has been detected by the associated emuent radiation monitor or by laboratory analysis. The sampling frequency, minimum analysis frequency, and type of analysis performed are as per l Table 9.3-A. 2.1.2 Radioactive Liouid Batch Release Emuent Monitors The radiation monitor which is associated with the liquid emuent batch release system is (Ref. I16.4): Monitor I.D. Description HB-RE-I S Liquid Radwaste Di: charge Monitor This emuent stream is normally considered to be radioactive. The sampling frequency, minimum analysis frequency, and the t3pe of analysis performed are as per Tabic 9.3-A. 2.2 CALCEATION OF LIOED EFFLUENT MONITOR SETPOINTS The dependence of the setpoint (c), on the radionuclide distribution, yields, calibration. and rnonitor parameters, requires that several variables be considered in setpoint calculations. (Ref. I1.8.1) I I 3 I

I APA-ZZ-01003 Rev.3 l 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 9.3-A, and is used to calculate an ECV sum I (ECVSUM): ECVSUM = ([(C,)/(ECV,)) i== g, a, s, t, f (2.3) I Where: C= the concentration of each measured gamma emitting nuclide observed by I } gamma-ray spectroscopy of the w _,ie sample. g C the measured concentration of alpha emitting r.aclides measured by gross alpha = a l analysis of the monthly composite sample. C*= the measured concentrations of Sr-89 and Sr-90 as determined by analysis of 3 the quarterly composite sample. l C= the measured concentration of H-3 in the waste sample. t I l C*= the measured concentration of Fe-55 as determined by analysis of the quarterly f composite sample. l ECV, ECV. ECV, ECVr, ECV = are the lintiting concentradons of the appropriate I g 3 a t radionuclides from 10 CFR 20, Appendix B, Table II, Column 2. For dissolved or entrained noble gases, the concentration shall be limited to 2x10-4 Ci/ml total actisity. For the case ECVSUM si, the monitor tank efnuent concentration meets the limits of REC 9.3.1.1 witheut dilution and the efIluent may be released at any desired flow rate. If ECVSUM > 1 then dil. don is required to ensure compliance with REC 9.3.1.1 concentration limits. If . I simultaneous releases are occurring or are anticipated, an allocation fraction, N, must be applied so that avanable 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, prosided that the sum total allocation fractions for all discharge points must be sl. 2.2.2 Calculation of the Maximum Permissible Licuid Effluent Discharre Flowri s t The maximum permissible liquid efDuent discharge flowrate is calculated by: l f s (F + f ) (SF)(N) + (ECVSUM) (2.4) max p I

• Values for these concentrations are based on previous composite sample analyses as required by Table 9.3-A.

I I

APA-22-01003 Rev.3 1 I i i i i Where: f = maximum permissible liquid emuent discharge Dowrate, in (gallons / minute); nm f= the expected undiluted liquid efiluent flowrate, in gpm. p N= the allccation fraction which apportions dilution flow among simultaneous discharge pathways (see discussion above) SF = the safety factor; an administrative factor used to compensate for statistical Ductuations and errors er measurements. This factor also provides a margin of safety in the I calculation of the maximum liquid efDuent discharge flowrate (fnm). The value of SF should be 51. I l F & ECVSUM, are previously defined. I The dilution water supply is furnished with a flow monitor which isolates the liquid cinuent discharge if the dilution flow rate falls below its sctpoint value. In the event that f is less than f. then the value of f is substituted into the equation for max max f, and a new value of f is calcubated. This substitution is performed for three iterations in p max order to calculate the correct value of fmax-2.2.3 Calculation of Liauid Efiluent. Monitor Serpoint The liquid efDuent monitors are Nal(TI) based systems and respond primarily to gamnu ra:iiation. Accordingly, their setpoint is based on the total concentration of gamma emitting I nuclides in the efDuent: c = BKG + ( E (C ) + SF ) = Cihnl(2.5) s Where: ) I e= the monitor se! point as previously defined, in (pCi/ml); I BKG = the monitor background prior to discharge, in (pCi/ml); C and SF are as previously defined. g The monitor's background is controlled at an appropriate limit to ensure adequate sensithity. Utilizing the methodology of ANE! N13.10-1974 (Ref. I1.21), the background must be I j maintained at a value ofless than or equal to 2.23E-6 pCi/ml (relative to Cs 137)in order to detect a change of IE-7 Ci/ml of1-134 (the most restrictive nuclide in Table 1 of reference I1.21). In the event that there is no detectable gamma activity in the efDuent or if the value af C(C ) + SF)is less than the background of the monitor, then the monitor setpoint will be set at g twice t' e current background of the monitor. h I i ) I 5- ) e

1 APA-ZZ-01003 Rev.3 I As previously stated, the monitor's response is dependent on the gamma emitting radionuclide distribution of the emocnt. Accordingly, a new database conversion factor is calculated for each release based upon the results of the gamma spectrometric analysis of the educ'd sample and the measured response of the monitor te National Institute of Standards and Technology (NIST) traceable calibration sources. DBCF = I(C,)) + (CMR) x (ECF) (2.6) c I DBCF = the monitor data base conversion factor which converts count rate into e concentration (pCi/ml); CMR= the calculated response of the radiation monitor to the liquid emuent; g ECF = the conversion factor for Cs-137, which converts count rate into concentration g (t Ci/ml). C is as presiously defined. g The new value of the DBCF is calculated and entered into the monitor data base prior to each e discharge. A more complete discussion of the derivation and calculation of the CMR is given in reference 11.14.7. 2.3 MOUID EFFLUENT CONCENTRATION ME ASUREMENTS Liquid batch releases are discharged as a discrete volume and each release is authorized based i upon the sample analysis and the dilution flow rate existing in the discharge line at the time of release. To assure representative rampling, each liquid monitor tank is isolated and thoroughly I 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 prceedural conditioru, prior to authorization for release. The liquid emuent I discharge is monitored prior to entering the dilution discharge line and will automatically be terminated if the pre-selected alarm / trip setpoint is exceeded. Concentrations are determined primarily from the gamma isotopic and H-3 analyses of the liquid batch sample. For gross alpha, I Sr 89, Sr-90, & Fe 55, the measured concentration from the previous composite analysis is used. Composite samples are collected for each batch release. Monthly analysis for gross alpha and quarterly analyses for Sr-89, Sr-90, and Fe-55 are performed in accordance with Tab!c 9.3-A. I Doses from liquids discharged as continuous releases are calculated by utilizing the last measured values of samples required in accordance with Table 9.3-A. 2.4 DOSE DUE TO UDUID EFFLILENTJ 2.4.1 The Maximum _E xposed Individual I The cumulative oose determination considers the dose contributions from the maximum exposed individual's consamption of fish and potable water, as appropriate. Normally, the adult is considered to be the maximum exposed individual. (Ref.11.8.3) I I I

I APA-ZZ 01003 Rev.3 I The Callaway Plant's liquid emuents are discharged to the Missouri River. As there are no potable water intakes within 50 miles of the discharge point (Ref. I1.7.1,11.6.6), this pathway does not require routine evaluation. Therefore, the dose contribution from fish consmupdon is I expected to account for more than 95% of the total man-rem dose from discLrges to the Missouri River. Dose from recreational activities is expected to contribute the additional 5%, which is considered to be negligible. (Ref. I1.6.7) I 2.4.2 Qticulation of Dow From Liould Emuents The dose contributions for the total time period. m I [dt, tal are calculated at least once cach 31 days and a cumulative sununation of the total body and individual organ doses is maintained for ca:h calendar quarter. Dose is calculata! foi d! ~ radionuclides identified in liquid emuents released to UNRESTRICTED AREAS using the following expression (Ref. I1.E.3): D,= A;, T At, C, F, (2.12) i i ti Where: D= the cumulative dose commitment to the total body or any organ, t, from the liquid emuents for the total period I T At, m ta in mrem. At, = the length of the / th time period over which Cj, and F, are averaged for allliquid releases,in hours. At, corresponds to the actual duration of the release (s). Cj,= the average measured concentration of radionuclide, i, in undiluted liquid emuent during time period 41, from any liquid release, in ( Ci/ml). Aj,; = the site related ingestion dose commitment factor to the total tody or any organ t for each identified principal gamma and teta emitter listed in Table 9.3-A, (in mrem /hr) per ( Ci/ml). The calculation of the Ajg values is detailed in Ref. I1.14.5 and are given I in Table 2.1. I I

I APA-ZZ-01003 Rev.3 F, = the near 6 eld average d.ilution factor for C, during any liquid emuent release: i f~ F, = (F + f_) 89.77 Where: fmax maximum undiluted emuent flow rate during the release = F= average diludon flow 89.77 = site specific applicable factor fut the mixing effect of the discharge structure. (Ref. I1.5.1) .g .R The term Cj, is the undiluted concentration of radioactive material in liquid waste at the common release poim deterndned in secordance with REC 9.3.1.1, Table 9.3-A, " Radioactive I Liquid Waste Sampling and Analysis Program". All dilution f'clors beyond the sample point (s) are included in the F, tenn. The nearest municipal potable water intake downstream imm the liquid efIluent discharge point I l 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 I exposed individual, or in dose estimates to the population. Should future potable water intakes be l constracted within 10 siver miles downstream of the discharge point, then this manual will be revised 'o include this pathway in dose estimates. (Ref. I1.6.6).. 2.4.3 Summarv. Calculation of Dose Due to Licuid Emuents The dose contribudon for the total time period m [ At, g is determined by calculation at least once per 31 days and a cumulative summadon of the total body and organ doses is maintained for each calendar quarter. The projected dose contribudon I from liquid emuents 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 rad >onuclides identified in liquid emuents released to UNRESTRICTED AREAS. Nuclides which are not detected in the analyses are reported as "less I than" the nuclide's Minimum Detectable Aedvity (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. I I .g. I

I APA-7Z-01003 Rev.3 l 2.5 L10UID RADWASTE TREATMENT SYSTEM The LlQUID RADWASTE TREATMENT SYSTEM is capable of varying treatment, depending on waste type and product desired. It is capable of concentrating, gas stripping, and distillation of I liquid wastes through the use of the evaporator system. The demineralization system is capable of removing radioactive ions from solutions to be reused as makeup water. Filtration is performed on certain liquid wastes and it may, in some cases, be the only requued treatment prior to release. I The system has the ability to absorb hahdes through the use of charcoal filters prior to their release. The design and operation requirements of the LIQUID RADWASTE TREATMENT SYSTEM I provide assurance that releases of radioactive materials in liquid efGuents will be kept "As Low As Reasonably Achievable"(ALARA). l The OPERABILITY of the LIQUID RADWASTE TREATMENT SYSTEM ensures this system

E will be available for use when liquids require treatment prior to their release to the emironment.

OPERABILITY is demonstrated through compliance with REC 9.3.1.1. and 9.4.1.1. Projected doses due to liquid releases to UNRESTRICTED AREAS are determined each 31 days ) by dividing the cumulative annual total by the number of elapsed months. I I I I I I I I I 9 I

I APA-ZZ-01003 Rev.3 TABLE 2.1 4 E INGESTION DOSE COMMITMENT FACTOR (A;2) FOR ADULT AGE GROUP I (rnrenvhr) per ( CUml) I Nuclides Bone Liver Total Thyroid Kidney Lung GI-LLI Bodv I ~ y.,,, .m. y ~ H-3 No Data 2.26E-01 ' '2.26E-01 2.26E-01 2.26E-01 2.26E 51 2.26E'-01 Be-7 1.30E-02 2.98E-02 1.45 E-02 No Dam 3.15E-02 No Data 5,16E+00 C-14 3.13E@4 6.26Ee3 6.26E403 6.26E+03 6.26E+03 6.26E403 6.26E+03 Na-24 4.07E+02 4.07E+02 4.07E+02 4.07E402 4.07E+02 4.07E+02 4.07E+02 P-32 4.62E+07 2.87E+06 1.78E+06 No Data No Data No Data 5.19E+06 I Cr-51 No Data No Data 1.27E+00 7.62E-01 2.81E-01 1.69E+00 3.20E402 Mn-54 No Data 4.38E43 8.35E+02 No Data 1.30E+03 No Data 1.34 E+04 Mn-56 No Data 1.10E402 1.95E+01 No Data 1.40E+02 No Data 3.52E403 Fe-55 6.57E+02 4.54E402 1.06E+02 No Data No Data 2.53E42 2.61E+02 I Fe 59 1.04E403 2.44E+03 9.34E402 No Data No Data 6.81E+02 8.13E+03 Co-57 No Data 2.09E+01 3.48E+01 No Data No Data No Data 5.31E+02 I Co-58 No Data 8.94E+01 2.00E42 No Data No Data No Data 1.81E403 Co-60 No Data 2.57E402 5.66E+02 No Data No Data No Data 4.82E+03 Ni-63 3.1IE404 2.15E+03 1.04 E+03 No Data No Data No Data 4.49E402 I Ni-65 1.26E+02 1.64 E+01 7.4 SE+00 No Data No Data No Data 4.16E+02 3 Cu-64 No Data 1.00E401 4.69E40 No Data 2.52E+01 No Data 8.52E+02 - Zn(35 2.32E44 7.3 8E +0 ' 3.33E+04 No Data 4.93 E+0 i No Data 4.65E+02 I Zn-69 4.93E401 9.44E401 6.56E+00 No Data 6.13 E+01 No Data 1.42E+01 Br-82 No Data No Data 2.27E+03 No Data No Data No Data 2.60E403 Br-83 No Data No Data 4.04E+01 No Data No Data No Data - 5.81E+01 i Br-S4 No Data No Data 5.26E+01 No Data No Data No Data 4.13E-04 Br-85 No Data No Data 2 15E+00 No Data No Data No Data 0 l Rb-86 No Data 1.01E+05 4.7 IE+04 No Data No Data No Data 1.99E+04 I Rb-88 No Da"t 2.90E-02 1.54E+02 No Data No Data No Data 4.00E-09 Rb-59 No Data 1.92E+02 1.35E+02 No Data No Data No Data 0 I St-89 2.21 E+04 No Data 6.35E402 No Data No Data No Data 3.55E403 { Sr 90 5.44E+05 No Data 1.34E 405 No Data No Data No Data 1.57E+04 l St-91 4 07E+02 No Data 1.64E+01 No Data No Data No Data 1.94 E+03 Sr-92 154E402 No Data 6.68E400 No Data No Data No Data 3.06E+03 . l I Y-90 5.75E 01 No Data 1.54E-02 No Data No Data No Data 6.10E+03 1 I I \\ I

APA-ZZ-01003 Rev.3 I TABLE 2.1 (Cont'd) INGESTION DOSE COMMITMENT FACTOR (A ) FOR ADULT AGE GROUP iz I (mremAtr) per (pCi/ml) Nuclide Bone Liver Total Thyroid Kindey Lung GI-LL1 .. Body ,.,r,_ i

u...

I

1. g.

^ Y-91 S.4 3E+00 No Data 2.25E-01 No Data No Data No Data 4.64 E+03 Y-92 5.05E-02 No Data 1.48E-03 No Data No Data No Data E.85E+02 Y-93 1.60E-01 No Data 4.42E-03 No Data No Data No Data 5.08E+03 I Zr-95 2.40E-01 7.70E-02 5.21E-02 No Data 1.21E-01 No Data

2. '4 E+02 Zr-97 1.33E-02 2.6SE-03 1.22E-03 No Data 4.04E43 No Data 8.30E+02 I

Nb-95 4.47E42 2.48E402 1.34E+02 Ne Data 2.46E+02 No Data 1.51E+06 Mo-99 No Data 1.03E42 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 1.4 EE+01 Tc-101 9.l lE-03 1.31E-02 1.29E41 No Data 2.36E-01 6.70E-03 0 Ru-103 4.42E+00 No Data 1.90E400 No Data 1.69E+01 No Data 5.17E+02 Ru 105 3.68E-01 No Data 1.45E-01 No Data 4.76E+00 No Data 2.25E+02 I Ru-106 6.5 +01 No Data 8.32E40 No Data 1.27E+02 No Data 4.25E+03 Cd-109 '4o Data 5.54E+02 1.94E+01 No Data 5.31E+02 No Data 5.59E+03 Sn lo 5.66E+04 1.6]E+03 3.26E+03

9. lSE+02 No Data No Data 1.69E+05 Sb 124 6.69E+00 1.26E-01 2.65E+00 L62E-02 No Data 5.21E*00 1.90E+02 Sb-125 4.28E+00 4.78E-02 102E+00 4.35E-03 No Data 3.30Ev00 4.71E401 -

Te-125M 2.57E+03 9.30E+02 3.44E+02 7.72E+02 1.04E+04 No Data 1.02E+04 - I Te-127M 6.47E+03 2.32E+03 7.90E+02 1.66E'03 2.63E+04 Nu Data 2.17E+04 Te-127 1.05E+02 3.78E+01 2.28E401 7.80E+01 4.29E+02 No Data 8.30E+03 Te-129M 1.10E+04 4.11E+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 ] E+01 1.26E+02 No Data 2.27E+01 Te-131M 1.66E M3 8.09E+02 6.75E+02 1.28E+03 8.21E403 No Data 8.03E+04 Te-131 189E+01 7.88E400 5.96E+00 1.55E401 8.25E401 No Data 2.67E+00 I Te-132 2.41 E+03 1.56E+03 1.47E+03 1.72E+03 1.50E+04 No Data 7.38E404 I-130 2.71 E+01 8.01E41 3.16E+01 6.79E+03 1.25E+02 No Data 6.89E+01 I I-131 1.49E+02 2.14E+02 1.22E+02 7.00E+04 3.66E+02 No Data 5.64E+01 1-132 7.29E+00 1.95E401 6.82E+00 6.82E+02 3.11E+01 No Data 3.66E+00 1-133 5.10E+01 8.67E+01 2.70E+01 1.30E+04 1.55E+02 No Data 7.97E+01 - I-134 3.81E+00 1.03E+01 3.70E+00 1.79E+02 1.64E+01 No Data 9.01E43 I 1-135 1.59E+01 4.16E+01 1.54E+01 2.75E+03 6.68E+01 No Data 4.70E40.1 I I I

4 APA-ZZ-O li.03 Rev. 3 TABLE 2.1 (Cont'd) INGESTION DOSE COMMITMENT FACTOR (Ajy FOR ADULT AGE GROUP I i (mrem /hr) per (pCi/ml) i-Nuclide Bone Liver Total Thyroid Kindey Lung GI-LLI } Bodv l y.m s,

.a

... ~.. ..,...s.. j Cs-134 2.95E45 7.09E+05 5.80E+05 No Data 2.29E+05 7.62E+04 1.24E+04 i Cs-136 3.12 E+04 1.23 E+05 8.86E+04 No Data 6.85E+04 9.39E43 1.40E+04 l Cs-137 3.82E+0.P 5.22E+05 3.42E +05 No Data 1.77E+05 5.89E+04 1.01E44 - Cs 138 2.64 E+02 5.22E+02 2.59E+02 No Data 3.84E+02 3.79 E+01 2.23 E-03 4 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 2.44E4)1 1.27E+01 No Data 8 3IE-02 1.40E-01 4.00E42

• E Ba1/1 4.50E-01 3.40E-04 1.52E-02 No Data 3.16E 04 1.93E-04 2.12E-10

5 Ba-l.a 2.04E-01 2.09E-04 1.28E-02 No Data 1.77E-04 1.19E-04 0

La-140 1.50E-01 7.53E-02 1.99E-02 No Data No Data No Data 5.53E+03 La-142 7.65E-03 3.48E-03 8.66E-04 No Data No Data No Data 2.54E+01 I i Cc-141 2.24E-02 1.51E-02 1.72E-03 No Data 7.03E-03 No Data 5.78E+01 ?. Cc-143 3.94E-03 2.92E+00 3.23E-04 No Data 1.28E-03 Nc Data 1.09E+02 - l Ce-144 1.17E+00 4 88E-01 6 26E-02 No Data 2.89E-01 Nc Data 3.94E+02 i Pr-143 5.50E-01 2.21E-01 2.~/ 3E-02 No Data 1.27E-01 No Data 2.41E+03 } Nd-147 3.76E-01 4.35E-01 2.60E-02 No Data 2.54E-01 No Data 2.09E+03 Eu-154 3.67E+01 4.52E+00 3.21 E+00 No Data 2.16E+01 No Data 3.27E43 j Hf-181 3.99E-02 1.94E-01 1.80E-02 No Data 4.17E-02 No Data 2.2 ] E+02 J W-187 2.96E+02 2.47E+02 8.64E+01 No Data No Data No Data 8.09E44 Np-239 2.84E-02 2.80E-03 1.54E-03 No Data 8.72E-03 No Data 5.74E+02 4 i ! I r I I I I I

I j

APA-ZZ411003 Rev.3 i-TABIJdJ BIOACCUMULATION FACTOR (EFj)_USED IN TIIE ABSENCE OF SITE-SPECIFIC D ATA(aj (pfiirl per (nCi/liicd 4i E i 3 { Bi E.lement Fish (Freshwat.n) l H 9.0 E - 01 [E Be 2.0 E + 00 C 4.6 E + 03 ! g Na 1.0 E + O2 g P 1.0 E + 05 j Cr 2.0 E + 02 j Mn 4.0 E + O2 l t Fe 1.0 E + O2 i 5 Co 5.0 E + 01 i Ni 1.0 E + O2 }g Cu 5.0 E + 01 1 g Zn 2.0 E + 03 t Br 4.2 E + 02 Rb 2.0 E + 03 l i Sr 3.0 E + 01 i i = Y 2.5 E + 01 Zr 3.3 E + 00 lg Nb 3.0 E + 04 g Mo 1.0 E + 01 Te 1.5 E + 01 Ru 1.0 E + 01 Rh 1.0 E + 01 Cd 2.0 E + O2 Sn 3.0 E + 03 I Sb 1.0 E + 00 Te 4.0 E + 02 I 1.5 E + 01 Cs 2.0 E + 03 I Ba 4.0 E + 00 La 2.5 E + 01 Cc 1.0 E + 00 I Pr 2.5 E + 01 Nd 2.5 E + 01 Eu 2.5 E + 01 Hf 3.3 E + 00 I W l.2 E + 03 Np 1.0 E + 01 I (a) Values from Regu?atorv Guide 1.109. Rev.1, Table A-1 and References 11.14.4 and 11.14.8. I I I

l I APA-ZZ-01003 Rev.3 I 3.0 GASEOUS EFFLUENTS 3.1 G ASEOt)S EFFIJJENT MONITORS Nobic gas acuvity monitors are present on the containment building ventilation system, plant unit ventilation system, and radwaste building ventiladon system. The alarm / trip (alarm & trip) setpoint for any gaseous emuent radiation monitor is determined based on the instantaneous noble gas total body and skin dose rate limits of REC 9.6.1.1, at the SITE BOUNDARY location with the highest annual average X/Q value. Each monitor channel is provided with a two level system w hich provides sequential alarms on increasing radioactivity levels. These setpoints are desipated as sdert setpoints and alarm / trip setpoints (Ref. I1.6.3) The radiadon monitor alarm / trip setpoints for ca:h release point are aased on the radioactive I noble gases in gaseous emuents. It is not considered practicable to apply instantaneous alarm / trip setpoints to integrating radiation monitors sensitive to radiciodines, radioactive materials in particular: form and radionuclides other than noble gases. Conservadve assumptions may be necessan in establishing setpoints to account for system variables, such as the measurement I system emciency and detection capabilities during normal, anticipated, and unusual operating condidons, the variability in release Dow and principal radionuclides, and the time lag between alarm / trip action and the final isolation of the radioactive efIluent. (Ref. I1.8.5) Table 9.2-B I prosides the instrument surveillance requirements, such as calibration, source checking, functional tesdng, and channel checking. 3.1.1 Conlipuous Release Gaseous Emuent Menitors The radiaticn detection monitors associated with continuous gaseous cinuent releases are (Ref. 11.6.8,11.6.9); Monitor I D. Descriotion GT-RE-21 Unit Vent GH-RE 10 Radwaste Building Vent Each of the above continuously monitors gaseous radioacUvity concentrations downstream of the. last point of potendal influent, and therefore measures emuents and not inplant concentrations.

g The unit vent monitor continuously monitors the emuent from the unit vent for gaseous

$5 radioacdvity, The unit vent, sia ventilation exhaust systems, continuously purges various tanks l and sumps normally containing low +1cvel radioactive aerated liquids that can potentially generate g airborne acuvity. The exhaust systems which supply air to the unit vent are from the fuel g 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. I I .I APA-71A)l001 Rev.3 The Radwaste Building ventilation emuent monitor continuously monitors for gaseous radioacuvity in the effluent duct downstream of the exhaust filter and fans. The flow path I provides ventilation es haust for all pans of the building structure and components within the building and prosides 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 addidon to the drainage susups, tanks, and equipment purged by the waste processing I system. This monitor will isolate the waste gas decay tank discharge line upon a high gaseous radioactivity alarm. The continuous gaseous emuent monitor setpoints are established using the methodology described in Section 3.2. Since there are two continuous gascous emuent release points, a I fractiot, of the total dose rate limit (DRL) will be allccated to cach 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 cl.anged as I required to support plant operational needs, but shall not be allowed to execed 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 excccded; the alarm only indicates that the specific release point is contributing a greater fraction of the dose rate limit than was allocated to the I associated monitor, and will necessitate an evaluation of both systems. 3.1.2 IhtchJLeicase Gaseous Monitors The radiation monitors associated with batch release gaseous emuents are (Ref. I1.6.9,11.6.10, 11.6.11): Monitor I D. DeSaiption 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. I 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 presiously described. A pre-release isotopic analysis is performed for each batch release to determine the identity and I quantity of the principal radionuclides. The alarnVtrip setpoint(s)is adjusted accordingly to ensure that the limits of REC 9.6.1.1 are not exceeded. 3.2 G ASEOUS EFFLUENT MONITOR ELTgNTS The alarm / trip setpoint for gaseous emuent monitors is determined based on the lesser of the total body dose rate (equation 3.1) and siin dos: rate (equation 3.3), as calculated for the SITE I BOUNDARY. I I

I APA-Z.Z-01.33 Rev. 3 l l During core alteradons, th: setpoint for the Containment Purge Monitors, GT-RE-22 and GT RE-33 is set at a value ofless than or equal to SE-3 Ci/cc, as required by Technical Spccification 4.9.4.2. The actual sctpomt value will be reduced according to the Instrument Loop Uncertainty Estimate (ILUE). This value will also be udlized in the event that there is no detectable noble gas activity in the containment atmosphere sample analyzed in accordance with REC 9.6.1.1. The full derivadon of this value is discussed in reference 11.14.6. 3.2.1 Total Body Dose Rate Se2 point Calculations To ensure that the limits of REC 9.6.1.1 are met, the alartn/ rip setpoint based on the total body t dose rate is cajculated according to: S s D,R,F,F, (3.1) ( Where: 9th = the alann' trip setpoint based on the total body dose rate (pCi/cc). Dtb = REC 9.6.1.1 limit of 500 mrem'yr. censcrvatively interpreted as a condnoous release over a one year period. F= the safety factor a conservadve factor used to compensate for stadstical fluctuadons and 3 g enors of measurement. (For example, F = 0.5 corresponds to a 100% variation.) 3 g Default value is F = 1.0. s F= the allocation factor which will modify the required diludon factor such that a simu!taneous gaseous releases may be made without exceeding the limits of REC 9.6.1.1 i The default value is 1/n, where n is the number of pathways planned for re! case. II Rib = factor used to convert dose rate to the effluent ccncentradon as measured by the emuent menitor, in (pCi/cc) per (mrenVp) to the total body, determined according to: R, = C + X/Q) K, Q, (3.21 l Where: ( C= monitor reading of a noble gas monitor corresponding to the sample radionuclide concentratwns far the batch to be released. Concentradons are deter. ned in accordance B with Table 9.6-A. The mixture of radionuclides determined via grab wpling of the i emuent stream or source is correlated to a calibration factor to determine monitor 1 response. The monitor response is based on concer.trations, net release rate, and is in units of( Ci/cc). X / Q the highest calculated annual average reladve concentration for any area at or beyond 3 the SITE BOUNDARY in (ccc/m ). Refer to Tables 6.1,6.2 and 6. l }:j= the total bady dose factor due to gamma emissions for cach identiDed noble gas 3 radionuclide. in (mrern'yr) per (pCi/m ). (Table 3.1) I Q= rate of release of noble gas radionuchde. i, in (uCi/sec). 4 - I APA-ZZ-OlOO3 { Rev.3 l Qi s calculated as the product of the vendlation path Dow rate and the measured actisity of the i effluent stream as determined by grab sampling. 3.2.2 Skin Dose Rate Sc! point Calculation I To ensure that the lindts of REC 9.6.1.1 are met, the alsnn/ trip setpoint based on the skin dose j rate is calculated according to: i l S, s D,R,F,F, (3.3) Where: F and F are as presiously defined. j 3 a f" S= the alann/ trip setpoint based on the skin dose rate. 3

!E o=

REC 9.6.1.1 limit of 3000 mrcrn'yr, conservatively interpreted as a continuous release s jg over a one year period. i l R= factor used to convert dose rate to the efDuent concentration as measured by the c0luen' 3 j monitor, in (pCi/cc) per (mrem /yr) to the skin, determined according to: j I R, = C + (X/Q) [ (L + 1.!M ) Q, (3.4) 3 i t s l 4 l Where: l Lj = the skin dose factor due to betr emissions for each identified noble gas radionuclide, in l (mrem /yr) per (pCi/m'). 1 j L1 = conversion factor; I mrad air dose = 1.1 miem skin dose. M= the air dose factor due to gamma emissions for each identified noble gas radionuclide, in I i 3 (mrad'p) per (pCi/m ). C, (X / Q) and Qj are previously defined. l 3.3 CALCULATION OF DOSE AND DOSE RATE FROM GASEOUS EFFLUENTS 3.3.1 Calculation of Dose Rate The following methodology is applicable to the location (SITE BOUNDARY or tx: yond) I characterized by the values of the parameter (X/Q) which results in the maximum total body or skin due rate. In the event that the analysis indicates a different location for the total body and skin dose limitations, the locadon selected for consideradon is that w hich minimizes the allowable release values. (Ref. I1.8.6) The factors Kj. L. and M relate the radionuclide airborne concentrations to various dose rates, i i assumine a semi-infinite cloud model. -17 I

I r j APA-ZZ-01003 Rev.3 -I l-3.3.1.1 b'oble Ga.s.g The rele:rse rate limit for noble gases is determined according to the following general reladonships (Ref.11.8.6): D, = [ K,((X/ Q)Q,) s 500 mrem / yr (3.5) ( i D,=Tj (L, + 1.1 M,)((X / Q)Q,) s 3000 mrem /yr (3.6) i Where: l Dtb = Total body dose rate, consen advely averaged as cr a period of onc > car. Kj = Total body dose factor due to gamma emissions for each identified noble gas 3 radionuclide, in (mrem /yr) per ( Cum A X / Q = The highest calculated annual average relative concentradon for any area at or beyond the SITE BOUNDARY. Refer to Tables 6.1,6.2, and 6.4 for applicability. Qi = The releasc rate of noble gas radionuclides, i, in gaseous effluents, from all vent releases in ( CUsec). Qj is calculated as the product of the ventilation path flow rate and the measured activity of the efIluent stream as determined by grab sampling. D= Skin dose rate, consen advely averaged over a period of one year. 3 I L; = Skin dose factor due to beta emissions for each identified noble gas radionuclide. in I 3 (mrem /yr) per (pCi/m ), 1.1 = Units conversion factor; I mrad air dose = 1.1 mrem skin dose. Mj= Air dose factor due to, gamma emissions for each identified noble gas radionuclide,in (mradlyr) per (pCUm ). 3 l 3.3.1.2 Radionu:lides Other Than Noble Gases The release raie limit for Iodine-131 and lodine-133, for tritium, and for all radioacdve materials in pardculate form with halflives greater than 8 days is determined according to (Ref. I1.S.7): D, = R, X/Q Q, s 1500 mrem /yr (3.7) Where: D= Dose rate to any cridcal organ, in (mrem /yr). o R; = Dose parameter for radionuclides other than noble gases for the inhalation pathway for 3 the child based on the crideal organ. in (nuem/yr) per (pCi/rn ). l { i

I 4 3 APA-ZZ-01003 j Rev.3 t-i 4 Q, and (X / Q) are as previously defined. The dose parameter (F ) includes the internal dosimetry of radionuclide,i, and the receptor's i l breathing rate, which are functions of the receptor's age. The child age group has been selected W. as the limiting age group. All radiodines are assumed to be released in elemental form (ref. 11.8.7). Rj values were calculated according to (Ref. I118): l R, = K' (BR) DFA, (3.8) j Where: ~ K' = Units conversion factor: lE06 pCi/pCi j BR = The breathing rate. (Regulatory Guide 1.109, Tab':E-5). l DFAj = Th: maximum organ inhalation dose factor for the ith radionuclide, in (mrern/pCi). The total body is considered as an organ in the selection of DFA;. (Ref. I1.11.5 and 11.14.4) f l3.3.2 Dose Due to Gaseous Emuents j 3.3.2.1 Noble Gas The air dose at the SITE BOUNDARY due to noble gases is calculated according to the i following methodology (Ref. I1.8.9): i During any calendar quarter, for gamma radiation: D, = 3.17E-08 [ M, (X/ Q) qi s 5 mrad (3.9) I During any calendar quarter, for beta radiation: D, = 3.17E-08 [ N (X/Q) qi s lo mrad (3.10) i i During any calendar year, for gamma radiation: D, = 3.17E -08 T' M (X/Q) q, s 10 mrad (3.11) 3 l During any calendar year, for beta radiation: f D, = 3.17E-08 [ N, (X / Q) q, s 20 mrad (3.12) } 6 s l. f

1 APA-ZZ-01003 Rev.3 Where: l l D= Air dose from gamma radiation due to noble gases released in gaseous emuent. g i D= Air dose from beta radiadon due to noble pses released in gaseous emuents. b Nj = The air dose factor due to beta emissions for each identified noble gas radionuclide, i, in 3 (mrad /yT) per (pCi/m ). l q; = The releases of radiciodines, radioanive materials in particulate form, and radionuclides other than noble gases i, in gaseous emuents, for all gaseous releases in (pCi). Releases are cumulative over the calendar quarter or year as appropriate. I l 3.17E-08 = The inverse of ae number of seconds p:r year. l X / G & Mj are as previcusly defined l 3.3.2.2 Radionuclides Other Than Noble Gases The dose to a MEMBER OF THE PUBLIC from Iodine-131 and 133, tndmn, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous emuents released, to areas at and beyond the SITE BOUNDARY, is calculated according to the following expressions: During any calendar quarter: D, = 3.17E- 08 Ri[W q;) s; 7.5 mrem (3.13) During any calendar year: D, = 3.17E-08 T Ri [W q;] s 15 mrem (3.14) Where: Dj = Dose to a hEMBER OF TIE PUBLIC from radionuclides other than noble gases. l 2 Rj = The dose factor for each identified radionuclide, i, in m (mrerrvyr) per (pCi/sec) or 3 (mrem /yr) per ( Ci/m ). (X / G) for the inhalation and tritium pathways, in (sec/m ). Refer to Tables 6.1,6.2, 3 W= and 6.4 for applicability. (D / G) for the food and ground plane pathways, in (meters-2). Refer to Tables 6.1, W= j 6.2 and 6.4 for applicabihty i 1 I

I APA-ZZ-01003 Rev.3 (D / Q) = the average relative deposidon of the emuent at or beyond the SITE BOUNDARY, considering depledon of the plume during transport. 3.17 E-08 = The inverse of the number of seconds per year. l gj is as previously defined. For the dircedon sectors with exisdng pathways within 5 niiles from the site, the appropriate R; I values are used. If no real p'thway exists within 5 miles from the center of the building complex, the cow-milk Rj value is used, and it is assumed that this pathway exists at the 4.5 to 5.0 mile distance in the limidng-case sector. If the Rj for an existing pathway within 5 miles is less than a cow-milk R at 4.5 to 5.0 miles, then the value of the cow-milk Rj at 1.5 to 5.0 miles is used. (Ref. 9.8.10) Although the annual average relative concentradon (X/Q) and the average reladve deposition rate (D/Q) are generally considered to be at the approxintate receptor locadon in lieu of the SITE BOUNDARY for these calcu!ations, it is acceptable to consider the ingestion, inhalation, and ground plane pathways to coexist at the locadon of the nearest residence with the highest value of (X/Q). (Ref. I1.8.9) The Total Body dose from ground plane deposidon is added to the dose for eacl. individual organ. (Ref. I1.11.3) l 3.4 G ASEOUS RADWASTE TREATMENT SYSTEM The gaseous radwaste treatment system and the ventilation exhaust system are available foi use whenever gaseous emuents require treatment prior to being released to the emironment. The I pascous radwaste treatment system is designed to allow for the retendon of all gaseous fission products to be discharged from the reactor coolant system. The retendon system consists of eight l (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 I materials in gaseous emuents will te kept ALARA. The OPERABILITY of the gaseous radwaste treatment system ensures this system will be I available for use when gases require treatment prior to their release to the emironment. OPERABILITY is demonstrated through compliance with REC 9.6.1.1,9.7.1.1, and 9.8.1.1. Projected doses (gamma air, beta air, and organ dose) due to gaseous emutnts at or beyond the SITE BOUNDARY are determined each 31 days by dividing the cumuladve annual total by the number of elapsed months. I I I I

m M M M M M M M M M APA-7.Z-01003 Rev.3 TABLE 3.1 DOSE FACTOR FOR EXPOSURE TO A SEMI-INFINITE CLOUD OF NOULE GASES' [ Total Body Gamma Air Beta Air Dose Factor Skin Dose Factor Dose Factor Dose Factor Ki L; M Ni 3 3 3 3 Radionuclide (mrem /yr) per (pCi/m ) (mrad /yr) per (pCi/m ) (mradlyr) per (pCi/m ) (mradlyr per ( Ci/m ) t ~ m m m m, - y.n3 w.a 'aw - _ _ p v~nym nn m gmmm, ~ .n a . om xc.. :. >; wh s:;a;p yx .-m ~s as., 3 sg z Kr-83 m 7.56 E-02 1.93 E401 283 E+02 Kr-85m 1.17E+ 03 1.46E403 1.23 E+03 1.97 E+03 K r-85 1.61 E+ 01 1.34 E+03 1.72 E+01 1.95 E403 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 E4 04 1.73 E404 1.06 E+04 Kr 90 1.56 E4 04 7.29 E403 1.63 EiO4 7.83 E+03 Xc-131m 9.15 E+01 4.76 E+02 1.56 E+02 1.11 E+03 Xc-133m 2.51 E402 9.94 E+02 3.27 E+02 1.4R E+03 Xc-133 2.94 E+02 3.06 E+02 3 53 E+02 1.05 E+03 Xc-135m 3.12 E+03 7.11 E+02 3.36 E4 03 7.39 E402 Xc-135 1.81 E+03 1.86 E403 1.92 E+03 2.46 E+03 Xc-137 1.42 E+ 03 1.22 E+04 1.51 E+03 1.27 E+04 Xc-138 8.33 E+03 4.13 E+03 9.21 E+ 03 4.75 E+03 Ar-41 S.34 E+03 2.69 E+03 9.30 E+03 3.28 E+03 (a) The listed dose factors are derivc<1 from Reg. Guide 1.109, Ta')!c B-1 (Rev.1,1977).

APA-ZZ-01003 TABLE 3.2 i PATHWAY DOSE FACTORS (Rj) FOR RADIONUCLIDES I OTilER THAN NOBLE GASESa Ground Plane Pathway 2 (m mremlyr) per (pCi/sec) I l NUCLTDE TOTAL BODY SKIN Bc-7 2.2 4 E+07 3.21E407 I Na-24 1.19E+07 1.39E+07 Cr-51 4.66E+06 5.51E-06 hin-54 1.39E+09 1.63E+09 hin-56 9.04 E+05 1.07E+06 Fe-59 2.73E+08 3.21E+08 Co-57 2.98E+08 4.37E+08 I Co-58 3.79E+08 4.44 E+08 Co40 2.15E+10 2.53E+10 Ni45 2.97E+05 3.45E+05 Cu44 6.07E+05 6.88E+05 Zn45 7.47E+08 8.59E+08 g Br 82 3.14 E+07 4.4 9E+07 3 Br-S3 4.87E+03 7.0SE+03 Br-84 2.03E+05 2.36E+05 I Rb-86 8.99E46 1.03 E+07 Rb-88 3.3 ]E+04 3.78E+04 Rb-89 1.23E+05 1.4 8E+05 I Sr-89 2.16E+04 2.51E+04 Sr-91 2.15E+06 2.51E+06 S r-92 7.77E+05 8.63E405 I Y-90 3.35E+06 6.32E+06 Y-91m 1.00E*05 1.16E+05 Y-91 1.07E+06 1.21E+06 Y-92 1.89E+05 2.14E405 Y-93 1.83E405 2.51E+05 Zr-95 2.45E+08 2.84E+08 I Zr-97 2.96E46 3.45E+06 Nb-95 2.50E+C8 2.94E+08 i I bio-99 4.00E+06 4.63E+06 I I I.

B APA-ZZ-01003 i TABLE 3.2 (Cont'd) PATHWAY DOSE FACTORS (R;) FOR RADIONUCLIDES =~ OTHER TIIAN NOBLE GASES ~ Ground Plane Pathway 2 l (m mrem /yr) per (pCi/sec) 1 l NUCLIDE TOTAL BODY SKIN Tc-99m 2.02E+06 2.31E+06 I Tc-101 2.04E+04 2.26E+04 Ru-103 1.0SE+08 1.26E+0S Ru-105 6.36E+05 7.21E+05 I Ru-106 4.22F-: Us 5.07E +0S Ag-11Om 3.44E+09 4.01E+09 Cd-109 3.76E+07 1.54Et08 I Sn-113 1.4 3E+07 4.09E+07 Sb-124 8.74E+08 1.23E+09 Sb-125 3.57E+09 5.19E+09 Tc 125m 1.55E+06 2.13E +06 Te-127m 9.17E+04 1.08E+05 i g Te-127 2.98E403 3.28E+03 E Te-129m 1.98E+07 2.31E+07 Te-129 2.62E+04 3.10Et04 I Te-131m 8.03E+06 9.46E406 Te-131 2.92E+04 3.45E+04 Te-132 4.23E+06 4.98E+06 I 1130 5.51E+06 6.69E+06 1131 1.72E+07 2.09E+07 I-132 1.25E+06 1.47E+06 I. 1 133 2.45E+06 2.98E406 l-134 4.47E+05 5.31E+05 1-135 2.53E+06 2.95E+06 Cs-134 6.85E+09 8.00E+09 Cs-136 1.51E+08 1.71E+08 i Cs-137 1.03E+10 1.20E+ 10 Cs-138 3.59E+05 4.10E+05 Ba-139 1.06E+05 1.I9E+05 Ba-140 2.05E+07 2.35E+07 4 1 :

I APA-ZZ-01003 TABLE 3.2 (Contd) PATHWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTHER THAN NOBLE GASES I Ground Plane Pathway 2 (m mrem /yr) per (pCi/sec) I NUCLIDE TOTAL BODY SKIN Ba-141 4.17E+04 4.75E 504 Ba-14 2 4.4 9E+04

5. l l E+04 I

La-140 1.4 7E-08 1.66E408 La-142 7.60E+ 05 9.12E+05 Cc-141 1.37E+07 1.54E407 Ce-143 2.3 IE+06 2.63E+06 Cc-144 6.96E+07 8.04E+ 07 Pr-144 4.35E+07 5.00E+07 I Nd-147 8.39E+06 1.01 E+C7 Eu-154 2.21E+10 3.15E+10 Hf-181 1.97E+08 .2.82 E+0S W-187 2.35E+06 2.73E+06 Np-239 1.71E+06 1.98E+06 I I I I I I I I,

I AP A-ZZ-01003 Rev. 3 TAILLE 3.3 Cl!!LD PATilWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTilER TilAN NOBLE GASES 2 Inhalation Pathw ay (mrem /) r) per (pCi/m') i TOTAL NUCLIDE HONE LIVE R BODY TIIYROID KTDNEY LUNG G T-LLI ~ j H-3 ND l 1.12E+03 l 1.12E+03 1 1.12E+03 l 1.12E+03 i L12E+03l 1.12E+03 l l Be-7 l 8.47E+02 l 1.44E+03 l 9.25E+02 l ND l ND l 6.47E+04 l 2.55E+03 l l C-14 l 3.59E+04 ' 6.73E+03 l 6.73E403 l 6.73E+03 1 6.73E+03 l 6.73E+03 l 6.73E+03 l l Na-24 j 1.61E+04 l 1.61E+04 l 1.6]E+04 l 1.61E+04 l 1.61E+04 i L61E+04 [ 1.61E+04 l 1 P-32 l 2.60E+06 l 1.14E405 l 9.88E+04 l ND l ND l ND l 4.22E+04 l I l Cr-51 l ND l ND l 154E+02 l 8.55E+01 l 2.43E+01 l 1.70E+04 l LOSE +03 l l Mn-54 l ND l 4.29E+04 l 9.51E+03 l ND l 1.00E+04 l 1.5SE+06 l 2.29E+04 l l Mn-56 l ND l 166E+00 l 3.12E-01 i ND l 1.67E+00 l 1.31E+04 l 123E+05 l l Tc-55 1 4.74E+04 l 2.52Er04 l 7.77E+03 l ND l ND l 1.1lE+05 l 2.87E+03 l I l Fe-59 l 2.07E+04 l 3.34E+04 l 1.67E+04 l ND l ND l 1.27E+06 l 7.07E+04 l l Co-57 l ND l 9.03E+02 l 1.07E+03 l ND l ND l 5.07E+05 l 1.32E+04 l I l Co-58 l ND l L77E+03 1 3.16E+03 l ND l ND l 1.llE+06 l 3.44E+04 l l Co-60 l ND 1 1.3]E+04 l 2.26E+04 l ND l ND l 7.07E+06 l 9.62E+04 l l Ni-63 l 8.21E+05 l 4.63E+04 l 2.80E+04 l ND l ND l 2.75E+05 l 6.33E+03 l l Ni-65 l 2.99E+00 l 2.96E-01 l 1.64E-01 l ND l ND l

8. LEE +03 l 8.40E+04 l l Cu-64 l

ND l 1.99E @ 0 l 1.07E+00 l ND l 6 03E+00 l 9.58E+03 l 3.67E+04 l l Zn 65 l 4.26E+04 l 1.13E+05 l 7.03E+04 l ND l 7.14E+04 l 9.95E+05 l _ L63E+04 l 5 l Zn-69 l 6.70E-02 l 9.66E-02 [ $ 92E-03 l ND l 5.SSE-02 l 1.42E+03 [ 1.02E+04 l l Br-82 l ND l ND l 2.09E+04 l ND l ND l ND l 0.^9E+00 l l Br-83 i ND l ND l 4.74E*02 l ND l ND l ND l 3.70E-15 l l Br-84 l ND l ND l 5.4SE+02 l ND l ND l ND l 3.70E-15 l l Br-85 l ND l ND l 2.53E+01 l ND [ ND l ND l 3.70E-15 l l Rb-86 l ND l 198E+05 l 1.14E+05 l ND l ND l ND l 7.99E+03. l l Rb-88 l ND l 5.62E+02l 3.66E+02 l ND l ND l ND 1 1.72E+01 l l Rb-89 l ND l 3.45E+02 1 2.90E+02 l ND l ND [ ND l 1.89E+00 l I l Sr-89 l 5.99E+05 l ND l 1.72E+04 l ND l ND l 2.l(E46 l 1.67E+05 l l Sr-90 l 1.01E+08 l ND l 6.44E+06 l ND l ND l L4SE+07l 3.43E405 l lSr-91 l 1.21E+02 l ND l 4.59E+00 l ND l ND l 5.33E+04 l 1.74E+05 l lSt-92 l 1.31E+01 l ND l 5.25E-01 l ND l ND l 2 40E+04 l 2.42E+05 l 5 l Y-90 l 4.llE203 l ND l 1.llE+02l ND l ND l 2.62E+05 l 2.68E+05 l I I I I

APA ZZ-01003 Rev. 3 TABLE 3.3 (Con't) Cll!LD PATI!WAY DOSE FACTORS (R;) FOR RADIONUCLlDES OTilER TIIAN NOIILE CASES" Inhalation Pathway (m rem /) r) per (p Ci/r.0) TOTAL j NUCLIDE BONE LIVE R BODY THYROID NIDNEY LUNG GT-LLI I l Y-91m l 5.07E-01 l ND l L842-02 l ND l ND l 2.81E+03 i L72E+03 l 1 l Y-91 l 9.14E+05 l ht i 2.44E+04 l ND l ND l 2.63E+06 l 1.84E+05 l l Y 92 1 2.03E+01 l ND l 5.81E-01 l ND l ND l 2.39E+04 l 2.39E+05 l l Y-93 l 1.86E+02 l ND l 5.1lE+00 l ND l ND l 7.44E+04 l 3.8SE+05 l t l Zr-95 l 1.90E+05 l 4.18E+04 l 3.70E+04 l ND l 5.96E+04 l 2.23E+06 l 6.llE4 04 l j l Zr 97 l 1.SSEw.: l 2.72EM)] l 1.60E+01 l ND l 3.89E+01 l 1.13E+05 l 3.51E+05 l I l Nb-95 l 2.3 5E+04 l 9.18E*03 l 6.55E+03 l ND l 8.62E+03 l 6.14E+05 l 3.70E+04 l l Mo-99 l ND l 1.72E+02 l 4.26E+01 l ND l 3.92E+02 1 1.35E+05 l 127E+05 l I Tc 99m l ! 78E-03 l 3.4SE-03 l 5.77E-02 l ND l 5.07E-02 l 9.51E+02 1 4.81E+03 l l Tc-101 l S.10E-05 l 8.5]E-05 l 1.08E-03 l ND l 1.45E-03 l 5.85E+02 l 1.63E+01 l l Ru-103 l 2.79E+03 l ND l 1.07E+03 l ND l 7.03E+03 l 6.62E+05 l 4.48E+04 l l Ru 105 l 1.53E+00 l ND l 5.55E-01 l ND l L34E+00 l 1.59E+04 l 9.95E+04 l I l Ru-106 l 1.36E+05 l ND l L69E+04 l ND l 1.84EM)5 l 1.43E+07 l 4.29E+05 l I Ag-110m l 1.69E+04 l 1.14E+04 1 9.14E+03 l ND l 2.12E+04 l 5.48E+06 l 1.00E+05 l l Cd 109 l ND l 5.4SE+05 l 2.59E+04 l ND l 4.96E+05 l 1.05E+06 l 2.78E+04 l l Sn 113 l 1.13E405 l 3.12E+03 l 8.62E+03 l 2.33E+03 } ND l 1.46E+06 l 2.26E+05 l l Sb-124 l 5.74E+04 l 7.40E4 02 l 2.00E+04 l 1.26E+02 ! ND l 3 24E+06 l 1.64E+05 l l 5b-125 l 9.84E+04 l 7.59E+02 l 2.07E+04 l 9.10E+01 l ND l 2.32E+06 l 4.03E+04 l I l Tc-125m l 6.73E+03 l 2 33E+03 l 9.14E+02 l L92E+03 l 0.00E+00l 4.77E+05 l 3.38E+04 l l Te-127m l 2.49E+04 l 8.55E+03 l 3.02E+03 l 6.07E+03 l 6.36E+04 l 1.4SE+06l 7.14E+04 l lTe-127 l 2.77E+00 l 9.5]E-01 l 6.11E-01 l 1.96E+00 l 7.07 +00 l 1.00E+04 l 5.62E+04 l l Tc-129m l 1.92E+04 l 6.84E+03 l 3.04E+03 l 6.33E+03 l 5.03E+04 l 1.76E+06 l 1.82E+05 l l Te-129 l 9.77E-02 l 3.50E-02 l 2.3SE-02 l 7.14E-02 l 2.57E-01 l 2.93E+03 l 2.55E+04 l l Tc-131m l 1.34E+02 l 5.92E+01l 5.07E+01 l 9.77E+01 l 4.00E402 l 2.06E+05 l 3.0SE+05 l l Te-131 l 2.17E-0: { 8.44E-03 l 6.59E-03 l 1.70E-02 l 5.8SE-02 l 2.05E+03 l 1.33E+03 l l Tc-132 l 4.81E+02 l 2.72E+02 1 2 63E402 l 3.17E+02 l 1.77E+03 l 3.77E+05 l 1.38E+05 l I l l-130 1 8.18E+03 l 1.64E+04 l 8.44E+03 l 1.85E+06 l 2.45E+04 l ND l 5.llE+03 l l 1 131 1 4.8]E+04 l 4.81E+04 l 2.73E+04 l 162E+07 l 7.8SE+04 l ND l 2.84E+03 l l l-132 l 2.12E+03 l 4.07E403 i 1.8SE+03 l 1.94E+05 l 6.25E+03 l ND 1 3.20E+03 l l l-133 l 1.66E+04 l 2.03E+04 l 7.70E+03 l 3.85E+06 l 3.3SE+04 l ND l 5.48E+03 l I l l-134 l 1.17E+03 l 2.16E+03 l 9.95E+02 l 107E+04l 3.30E+03 l ND l 9.55E+02 l I I I -2 7-I i

. ~. - I APA-ZZ-01003 Ret 3 j TA Bl.E 3.3 (Con't) ClllLD PATllWAY DOSE FACTORS (Rj) FOlt RADIONUCLIDES OTilER Til AN NOBLE GASESa Inhalation Patlova3 (m remlyr) per (p Ci/m') TOTAL NUCLIDE BONE LWER BODY THYROID NIDNEY LUNG GI-LLI I-l 1 135 l 4.92E+03 l 8.73E+03 l 4.14E+03 l 7.92E+05 l 1.34E+04 l ND j _ 4.44E+03 l l Cs-134 1 6.51E+05 l

1. ole +06 l 2.25E+05 l ND

-l 3.30E+05 l 1.21E+05l 3.85E+03 l l Cs-136 l 6.5]E+04 l 1.71E-05 l 1.16E+05l ND l 9.55E+04 l 1.45E+04 l 4.18E+03 l I l Cs-137 i 9.06E+05l 8.25E+05 l 1.2SE+05 l ND l 2.82E+05 l 1.04E+05 l 3.6?E+03 l } Cs-lu l 6.33E+02 l 8.40E+02 l 5.55E+02 l ND 1 6.22E+02 l 6.8 t h+01 l 2.70E+62 l j l Ba-139 l 1.84E+00l 9.84 E-04 1 5.37E-02 l ND l 8.62E-04 l 5.77F+03 l 5.77E+04 l l Ba-140 l ~.40E+04 1 6.47E401 l 4.33E+03 l ND l 2.llE+0! l 1.74E+06 l 1.02E+05 l l Ba-141 l 1.96E-01 l 1.09E-04 l 6.36E-03 l ND l 9.47E-05 l 2.92E-03 1 2.75E+02 I l Ba-142 l 5.00E-02 l 3.60E-05 l 2.79E-03 I ND l 2.91E-05 l 1.64E+03 ! 2.74E+00 l l La 140 l 6.44E+02 l 2.25E+02 1 7.55E+01 1 ND l ND l 1.83E+05 2.26E+05 1 i La-142 l 1.29E+001 4.1lE-01 l 1.29E-01 l ND l ND l 8.69E+03 l 7.59E+04 l l Cc-141 l 3.92E+04 l 1.95E+04l 2.90E+03 l ND l 8.55E403 l 5.44E+05 l 5.66E+04 l 1 l Cc-143 l 3.66E+02 l 1.99E+02 l 2.87E+01 l ND l 8.36E+01 l 1.15Evo5 l 1.27E+05 l l Cc-144 l

6. tie +06 l 2.12E+06 l 3.61E+05 l ND l 1.17E+06 l 1.20E+07l 3.88E+05 l l Pr-143 l 1.85E+04 l 5.55E+03 l 9.14E+02 l ND

} 3.00E+03 l 4.33E+05 l 9.73E+04 l I 1 l Pr-144 l 5.96E-02 l 1.55E-02 [ 3.00E-03 l ND l 9.77E-03 l 1.57E+03 l 1.97E+02l l Nd-147 l 1.0SE+04 l 8.73E+03 l 6.81E+02 } NU l 4.SIE+03 l 3.28E+05 l 8.21E+04 l l Eu-154 l 1.01E407 l 9.21E+05 l 8.40E+05 l ND l 4.03E+06 l 6.14E+06 l 1.10E+05 l 1 l Hf 181 l 2.78E+04 l 1.01E+05 l 1.25E+04 l ND l 2.05E+04 j l.06E+06 l 6.62E+04 l l W-187 l 1.63E+01 l 9.66E+00 l 4.33E+00 l ND l ND l 4.llE+04 l 9.10E+04 l l Np-239 l 4.66E+02 l 3.34E+01 l 2.35E+01 l ND l 9.73E+01 l 5.81E+04 l 6.40E+04 l I g l 1 I I I

APA ZZ-01003 Rev. 3 T A Hl.E 3.3 (ContNI) ClllLD PATilWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTilER Til AN NOBLE CASES 8 Meat Pathway (m mrern/ r) per (pCihec) 2 3 I. TOTAL NUCI.IDE HONE LIVER HODY TilVRO1D KIDNEY !UNG GI-I.L1 I l H-3 l ND l 2.34E+02 l 2.34E+02 l 2.34E+02 l 2.34E+02 J 2.34EM)2 l 2.34E+02 l l Be-7 l 7.3SE+03 l 1.26E+04 ! 8.07E403 l 000E+00l 1.23E+04 l 0.00E+00 l 7.00E+05 l l C 14 l 3.84E+08 j 7.68E+07 l 7.68E+07 l 7.6SE+07 1 's.68E+07 l 7.6hE+07 l 7.68E+07 l l Na-24 l 1.73E-03 l 1.73E-03 l 1.73E-03 l 1.73E-03 l 1.73E-03 l 1.7.'E-03 l 1.73E-03 l l P-32 l 7.43E+09 l 3.4SE+08 l 2.86E4 0s l 0.00E+00 l 0.00E-00l 0 00E+06 l 2.05E-08 l l Cr-51 l 0.00E400l 0.00E+00 l 8.80E+03 l 4.88E*03 l 1.33E403 ! S.92E+03 l 4.67E+05 l I l Mn 54 l 0 00E+00 l 8.02E+06l 2.14E+06 l 0.00E+00l 2.25E+06 l 0.00E+00l 6.73E+06 l l Mn-56 l 0.00E+00l 1.82E-53l 4.1lE54l 0.00E+00 l 2.20E-51 l 0.00E-00 l 2.64E-SI l l Fe-55 l 4.58E+08 l 2.4.1E+0S l 7.52E+07 l 0.00E+00l 0.00E 'oO I 1.37E+0S l 4.50E+07 l l Fe 59 l 3.77E+0S l 6.10E40S l 3.04E+0S l 0.00E-00 l 0.00E-00l 1.77E+08 l 635E+0S l I l Co 57 l 0.00E+00l 5.92E+06 l 1.20Et07 l 0.00E+00 1 0.00E+00 l 0.00E+00 1 4.S$E+07 l l Co-SS l 0.00E+00 l L64E+07 l 5.03E+07 l 0.00E+00 l 0.00E+00 l 0.00E+00 l 9.59E+07 l I l Co-60 l 0.00E+00l 6.94E+07 l 2.05E+0S l 0.00E+00l 0.00E+00l 0.00E+001 3.84E+0S l l Ni-63 l 2.92E+10 1 1.56E+09l 9.92E+08 l 0.00E*00 1 0.00E+00 l 0.00E+00 l 1.05E+0S l l Ni-65 l 3.62E-52 l 3.41E-53 l 1.99E-53 l 0.00E+00 l 0.00E+00l 0.00E+00l 4.lSE-51 l l Cu-64 l 0.00Et00 l 2.99E-07 l 1.80E-07 l 0.00E+00 l 7.22E-07 l 0.00E+00l 140E-05 l l Zn-65 l 3.76E+0S l 1.00E+09 l 6.23E+0S l 0.00E+00 l 6.31E+0S l 0.00E+00 l 1.76E+0S [ l Zn-69 l 0.00E+00l 0.00E+00l 0.00E+00 l 0.00E+00l 0.00E+00 j 0.00E+00 l 0.00E+00 l I l Br-S2 l 0.00E+00 l 0.00E+00 l 1.53E+03 l _0.00E+00l 0.00E+00l 0.00E+001 0.00E+00 l l Br-83 l 0.00E-00 l 0.00E+00l 9.82E-57 l 0.00E+00 l 0.00E40l 0.00E+00 l 5.74E-74 ) l Br-84 1 0.00E+00 l 0.00E+00l 0.00E+00l 0.00E+00l 0.00E+00 l 0.00E+00 l 0.00E+00 l I-l Br-85 l 0.00E+00[ 0.00E+001 0.00E+00 l 0.00E+00 ! 0.00E+00 l 0.00E+00 l 0,00E+00 l [Rb86 l 0.00E-{J0 l 5.77E+0S l 3.55E+0S l 0.00E+00l 0.00E+00l 0.00E+00l 3.7]E'07 l l Rb-0S l 0.00E+00l 0.00E+001 0.00E+00 l 0.00E+00 l 0.00E+00 l 0.00E+00j 0.00E+00 l-l Eb.89 [ 0.00E+00l 0.00E+00l 0.00E+00l 0.00E+00l 0.00E+00l 0.00E+00l 0.00E+00 l l j Sr 89 l 4.82E+0S l 0.00E+00 l 1.3SE+07l 0.00E+00 l 0.00E+00 1 0.00E+00 ! 1.87E+07 l l g l Sr-90 l 1.04E+10 l 0.00E+00 l 2.64E+09 l 0.00E+00l 0.00Ed-00 l 0.00E+00 l 1.40E+0S 1 ' g l Sr-91 l 2.42E-10 l 0.00E+00 l A13E-12 l 0.00E+00 l 0.00E+00 l 0.00E+00l 5.34E-10 l l Sr-92 l 1.90E-49 l 0.00E+00 l 7.60E-51 l 0.00E+00 l 0.00E+00 l 0.00E+00 l 3.59E-48 l l Y-90 l 1.93E+05 l 0.00E+00 l 5.16E403 l 0.00E+00 1 0.00E+00 l 0.00E+00 l 5.49E+0Sl ) I I I I

i

{

APA-ZZ-01003 j Rev. 3 TARI.E 3.3 (Cont'd) ClllLD PATIIWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTllER TilAN NOBLE GASESa Meat Pathway { (m mrcm/,s r) per (p Cihec) 8 TOTAL NU CL1D E BONE LIVER HODY TIf YROID KIDNEY LUNG GI-LLI l Y-91m l 0.00E+00 1 0.00E+00 l 0.00E*00 l 0.00E+00 l 0.00E+00 1 0.00E+00 l 0.00E+00 l l Y-91 l 1.80E+06 l 0 00E400 l 4.82E+04 l 0.00E+00 l 000E+00l 0 00E+00 l 2.40E+0S l l Y-92 l 2.46E-39 l 0.00E+00l 7.04 E-41 1 0.00E+00 l 0.00E+00 l 0.00E+00 l 7.1IE-35 l l Y-93 l 7.50E-12 l 0.00E+00 l 2.06E-13 l 0.00E+00l 0.00E+00l 0 00E+00 l 1.12E-07 l l Zr 95 l

2. 67E4 00 l 5.86E405 l 5.22E+05 j 0.00E+00 l 8 39E+05 l 0 00E+00 l 6.llE+0S l l Zr-97 l

3.22E-05 l 4.66E-06 l 2.75E-06 l 0.00E+00 l 6.6SE-06 l 0.00E+00l 7.05E-Ol l l Nb-95 l 4.26E+06 l 1.66E+06 l

1. lSE+06 l 0.00E+00 l 1.56E+06 l 0.00E+00 l 3 07E+091 j

l Mo-99 l 0.00E+0a l 1.15E+05 l 2.85E+04 l 0.00E400 l 2.46E+05 l 0.00E+ 00 l 9.53E+04 l l l Tc-99m l 4.00E +02 1 7.84 E*02 l 1.30E-04 l 0.00E+00 l 1.14E+04 l 3.98E+02 l 4.46E+05 l l Tc-101 l 0.00E+00l 0.00E+00 l 0.00E+00 l 0.00E+00l 0.00E+00l 0.00E+00 1 0.00E40 l r" l Ru-103 l 1.55E+0S l 0.00E+00 l 5.96E+07 l 0.00E+00l 3.90E+0S l 0.00E+00 l 4.01E+09 l l l Ru-105 l 9.17E-28 l u.00E+00 l 3.33E-28 l 0.00E+00 l S.06E-27 l 0.00E+00 l 5.99E-25 ' l l l Ru-106 l 4 44E+09 l 0.00E+00l 5.54E+0S l 0.00E+00 l 6.00E409 l 0.00E+00 l 6.91E+10 l l l Ag-l10m l 8 40E+061 5.67E+06 I 4.53E+06 l 0.00E+00 l 1.06E+07 l 0.00E+00l 6.75E+0S l l Cd-109 l 0.00E+00 l 1.91E+06 l S.84E+04 l 0.00E+00 l 1.70E+06 l 0.00E+00 l

6. LEE +06 l l'

l Sn-!!3 l 2.lSE+09 l 4.4SE+07 l L24E+0S l 3.31E+09 l 0.00E+00l 0.00E+00 l 1.54E+09 l l Sb-124 l 2.93E+07 l 3.80E+05 l 1.03E+07 l 6 46E+04 l 0.00E+00 l 1.62E+07 l 1.83E+0S l l Sb-125 l 2.85E+0? l 2.20E+05 l 5.97E+06 l 2 64E+04 l 0.00E+00 l 1.59E+07 l 6.81E+07 l l Te-125m l 5.70E+0Sl L54E+0Sl 7.59E+07 l 1.60E+0S l 0.00E+00 l 0.00E+00 l 5.50E+08 l r j Tc-127m l 1.78E+09l 4.78E+0S l 2.1 J E+0S l 4.25E+03 l 5.07E+09 l 0.00E+00l 1.44E+09 l [ l Te-127 l 3.42E-10 l 9.21E-Il l 7.33E-f l l 2.36E-10 l 9.72E-10 l 0.00E+00 l 133E-OS l l Te-129m l L79E+09 l 5.00F+0S l 2.78E+0S l 5.78E+03 l 5.26E+09 l 0.00E+00 l 2.19E+09 l l l Te-129 ) 0.00E+00 l 0.00E+00 l 0.00E+00l 0.00E400 l 0.00E+00l 0.00E+00l 0.00E+00l l Te-131m l 7 02E+02 l 2.43E+02 l 2.58E+02 l 4.99E+02 l 2.35E+03 J 0.00E+00l 9.85E+03 l l Te-131 l 0.00E+001 0.00E+00 l 0.00E+00 l 0.00E+00 l 0.00E+00l 0.00E+00 l 0.00E+00l f l Te-132 l 2.12E406 l 9.40E+05 l Ll4E+06 l L37E+06 l S.73E+06 l 0.00E+00 l 9.47E+06 l f l l-130 l 3.06E-06 l 6.18E-06 l 3.lSE-06 l 6.80E-04 l 9.23E-06 l 0.00E+00 l 2.89E-06 l l l-131 l L66E+07 l 1.67E+07 l 9 47E+06 l 5.51E+09 l 2.74E+07 l 0.00E+00l 1.4SE+06 l j l l-132 l LooE-SS l 1.94E-58 l S.93E-59 l 9.01E-57 l 2.97E-58 l 0.00E+00l 2.2SE-58 l lI-133 l 5.70E-01 l 7.05E-01 l 2.67E-01 l 1.3]E+02l 1.17E+00l 0.00E+00 l 2.84E-01 l lI134 l 0.00E+00 l 0.00E+00l 0.00E+00l 0.00E+00j 0.00E+00 l 0.00E+00 l 0.00E+00l [

'l APA ZZ-01003 3 Rev. 3 i i TA BI E 3.3 (Cont $,1) ClllLD PATllWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTiiER TII AN NOULE GASESn Meat Pathway (m' m rem /,s r) per (pCihec) TOTAL NUCLIDE BONE LIVER BODY TilYROID KIDNEY 1.UNG G I.L1.1 l l-135 l 6.59E-17 l 1.19E 16 l 5.62E-17 l 1.05E-14 l 1.82E-16 1 0.00E+00 1 9.04E-17 l l Cs-134 l 9.23E+08 l 1.51E+09l 3.20E+08 l 0.00E+00 l 4.69E+0S l 1.68E+0S 1 8.17E+06 l = l Cs-136 l 1.62E+07 l 4 46E+07 l 2.89E+07 l 0.00E+00 l 2.38E+07 l 3.54E+06 l 1.57E+06 l I l Cs-137 l 1.33E+C9 l l 2SE+09 l 1.89E+08 l 0.00E400 l 4.16E+08 l 1.50E+08 ; 8 00E+06 l l Cs 138 l 0.00E+00 l 0.00E+00l 000E+00l 0.00E+ 00 l 0.00E+00 l 0.00E+00l 0.00E+00l i l Ba-139 l 0 00E+00 l 0.00E+001 0.00E+00 l 0.00E+00l 0.00E+00l 0.00E+00 l 1.20E-99 l l Ba 140 l 4.39E+07 l 3.85E+04 l 2.56E+06 l 0.00E400 l 1.25E+04 l 2.29E+04 l 2.22E+07 l l Ba 141 l 000E-00l 0.00E+00 1 0.00E+00 l 0.008 00 1 0.00E+00 l 0.00E+00 l 0.00E+00 l l Ba.142 l 0.00E+00l 0.00E+00l 0.00E+001 0.00E+00 l 0 00E+00 l 0 00E+00 l 0 00E+00 l l La-140 e 3.33E+02 l 1.17E+02 l 3.93E+01 l 0.00E+00l 0.00E+00 l 0.00E+00l 3.25E-06 l l La 142 l 5.50E 92 l 1.75E-92 l 5.49E-93 [ 0.00E+00l 0.00E+00l 0.00E+00l 3.4SE-87 l l Cc-141 l 2.22E+04 l 1.] IE+04 l 1.65E+03 l 0.00E+00 l 4.86E+03 l 0.00E+00 l 1.3SE+07 l I l Ce-143 l 3.18E-02 l 1.72E+0! l 2.50E-03 l 0.00E+00l 7.23E-03 l 0.00E+00l 2.52E+02 l l Cc-144 l 2.32E+06 l 7.27E+05 l 1.24E+05 l 0.00E+00 1 4.02E+05 l 0.00E+00l 1 SPE+0S l l Pr 143 l 3.34E+04 l 1.00E+04 l 1.66E+03 1 0.00E+00 l 5.43E+03 1 0.00E+00 l 3.61E+07 l l Pr 144 l 5.63E+02 l 1.74E+02 l 2.83E+01 l 0.00E+00l 9.21E+01 l 0.00E+00 l 3.75E+05 l l Nd-147 l 1.17E+04 l 9.4SE+03 l 7.34E+02 l 0.00E+001 5.20E+03 l 0.00E+00 l 1.50E+07l l Eu 154 l 1.12Et07 l 1.01E+06 l 9.20E+05 l 0.00E+00 l 4.43E+06 l 0.00E+00l 2.34E+08 l I l Hf-lS1 l 4.77E+06 l 1.74E+07 l 2.15E+06 l 0.00E+00 l 3.53E+06 l 0.00E+00 l 6.4 ] E+09.l l W-187 l 3.22E-02 l 1.91E-02 l 8.56E-03 l 0.00E+001 C.00E+00l 0.00E+00l 2.68E+00 l l Np-239 l 4.33E-01 l 3.1lE-02 l 2.19E-02 l 0.00E+00 l 9.00E-02 l 0.00E+00 l 2.30E+03 ) l 1 I I I I -31

I APA ZZ-01003 Rev. 3 TA HLE 3.3 (Cont'd) CHILD PATHWAY DOSE FACTORS (R;)FOR RADIONUCLIDES OTIIER TIIAN NOBLE GASES 3 Grass-Cow-Milk Pathway (m mrcm/) r) per (pCihec) 8 TOTA L NUCl.TDE BONE LIVER IlO D T TIIYROID KIDNEY LUNG GI-LLI l H-3 l 0.00E+00 l 1.57E+03 l 1.57E+03 l 1.57E+03 l1.57E+03 l 1.57E+03 l 1.57E+03 l I l Bc.7 l 7.50E+03 l 1.28E+04 l 8.20E+03 l0.00E+00 l1.25E+04 1 0.00E+00 l 7.12E+05 l l C-14 l L20E+09 l 2.39E+08 l 2.39E+0S l 2.39E+0S l 2.39E+0F l 2.39E+08 l 2.395+08 j l Na-24 l 8.87E+06 l L.87E+06 lS.87E+06 l8.87E+06 l 8.87E+06 l887E+06 [ 8.8'E+06 l l P-32 l 7.79E+10 l 3.64E+09 l 3.00E+09 l0.00E+00 l0.00E+00 l0.00E+00 l 2.15E+09 l I l Cr-51 1 0.00E+00 l0.00E+00 l LO2E+05 l 5.66E+04 j L55E+04 l 1.03E+05 l 5.40E+06 l l Mn-54 l 000E+00 l 2.10E+07 l 5.59E+06 l0.00E400 l5.89E+06 l0.00E+00 l 1.76E+07 l I l Mn-56 l 0.00E+00 i 1.30E-02 l 2.93E-03 l0.00E+00 i L57E-02 1 0.00E+00 l1.S8E+00 l l Fe 55 l 1.12E+0S l5.94E+07 l 1.84E+07 l0.00E+00 l0.00E+00 l 3.36E+07 l 1.10E+07 l l Fe-59 l 1.20E+08 l1.95E+0S l 9.70E+07 l 0 00E+00 l0.00E+00 l 5 04E+07 l 2.03E+08 [ l Co-57 l0.00E+00 l 3.84E+06 l 7.78E+06 l 0.00E+00 l0.00E+00 l0.00E400 l 3.15E+07 l l Co-58 l0.00E+00 l1.2!E+07 l 3.72E+07 l0.00E+00 1 0.00E+00 l0.00E*00 l 7.08E+07 ( l Co-60 l0.00E+00 l 4.32E+07 l 1.27E+0S l0.00E+00 l0.00E+00 1 0.00E+00 l 2.39E+08 l I 1 Ni 63 l 2.97E+10 l 1.59E+09 l1O!E-09 l 0.00E'00 l0.00E+00 l0.00E+00 l LO7E+0S I l Ni-65 l L66E+00 l 1.56E-01 l 9.13E-02 l0.00E+00 j0.00E+00 l0.00E+00 l 1.92E+01 l f l Cu-64 l 0.00E+00 l 7.47E+04 l 4.52E+04 ] 0.00E+00 l 1.81E+05 l0.00E+00 l 3.51E+06 l l Zn-65 l 4.14E+09 l 1.10E+10 l 6.86E+09 l0.00E+00 l 6.95E+09 l0.00E+00 l 1.94E+09 l l Zn-69 l 9.55E-12 l1.38E-11 l 1.2SE 12 l0.00E+00 l 8.37E.12 1 0.00E+00 l 8.70E-10 l [ Br-82 l 0.00E+00 l0.00E+00 l1.15E+0S l0.00E+00 l0.00E+00 l0.00E-00 l0.00E+00 l l Br-83 l0.00E400 l0.00E+00 l 4.42E-01 l0.00E+00 j0.00E+00 l 0.00E+00 l 2.58E-18 l l l Br.84 [ 0.00E+00 l 0.00E+00 l 6.60E-23 1 0.00E+00 l0.00E+00 l0.00E+00 l 3.34E-40 l l l Br-85 l 0.00E+00 l 0.00E+00 l0.00E+00 l 0.00E+00 l 0.00E-00 1 0.00E+00 1 0.00E+00 l l Rb-86 l 0.00E+00 l 8.73E+09 l 5.40E409 l0.00E+00 l0.00E+00 l0.00E+00 l5.65E+0S l l Rb-88 l 0.00E-00 l 7.33E-45 l 5.09E-45 l0.00E+00 l0.00E+00 l0.00E-00 [ 3.60E-46 [ l Rb 89 l 0.00E+00 l L38E-52 I L23E 52 l0.00E+00 l0.00E+00 1 0.00E+00 l 1.21E-54 l l Sr-89 l 6.63E+09 l0.00E+00 l1.89E+0S l 0.00E+00 l0.00E+00 j0.00E+00 l 2.57E+08 l l Sr 90 l 1.12E+11 l0.00E+00 l 2.84E+10 l 0.00>00 l0.00E+00 l 0.00E+00 l 1.51E+09 l I l Sr-91 l 1.31E+05 l 0.00E+00 l 4.93E403 l 0.00E+00 l 0.00E+00 l 0.00E+00 l 2.88E+05 l l Sr-92 l 2.19E+00 l 0.00E+00 l 8 79E-02 l0.00E+00 l 0.00E+00 l0.00E+00 l 4.15E401 l l Y-90 l 3.38E+03 l0.00E+00 l 9 05E+01 l0.00E+00 l0.00E+00 l0.00E+00 l 9.62E+06 l t t !I !

APA-ZZ-01003 B Ret 't TA HLE 3.3 (Cont'd) CIIILD PATiiWAY DOSE FACTORS (R;)FOR RADIONUCLIDES OTIIER i HAN NOBLE G ASESa Grass-Cow-Milk Pathway (m mremly r) per (pCihec) 2 TOTAL I N11CL1DE BONE L1VER HODY TilVRO1D KIDNEY IUNG GTLLI l Y-91m l 2.70E-19 l0.00E+00 l 9.83E-21 1 0 00E+00 1 0.00E+00 1 0.00E+00 l 5.29E 16 l I l Y.91 1 3.91 E+04 l 0.00E+00 l1.04E+03 l0.00E+00 1 0.00E+00 l 0 00E+00 l5.20E+06 l l Y-92 l 2..'.4 E-04 l0.00E+00 l 7.26E-06 l0.00E+00 l0.00E+00 1 0.00E +00 l 7.33E+00 l l Y-93 l1.01E400 l0.00E+00 l 2.78E-02 1 0.00E+ 00 l0.00E+00 l0.00E+00 l 1.51E4 04 l l Zr-95 l 3.84E+03 l 8.43E+02 l 7.51E+02 l0.00E+00 l1.21E+03 l0.00E+00 l 8.80E+05 l .I l Zr-97 l 1.92E4 00 l 2.78E-Ol l 1.64E-01 l0.00E+00 l 3.99E-01 l0.00E+00 l 4.21E+04 l l Nb-95 l 3.72E4 05 l 1.45E+05 l 1.03E+05 l000E+00 l 1.36E405 1 0.00E+00 l 2.6SE+08 l I l Mo 99 l0.00E+00 l 815E+07 l 2.02E+07 l0.00E+00 l 1.74E+0S l0.00E+00 l 6.74E+07 l l Tc-99m l 1.88E+04 l 3.70E+04 l 6.13E+05 l0.00E+00 l 5.37E+05 l1.8SE+04 l 2.10E+07 l l Tc-101 l 1.20E-59 l 1.26E-59 l1.59E-58 l0.00E+00 l 2.14E-58 l 6.64E-60 l 3.99E-59 l l Ru-103 l 4.29E+03 l0.00E+00 l1.65E+03 l0.00E+00 l1.08E+04 l0.00E+00 l 1.1]E+05 l l Ru-105 l 3.83E-03 1 0.00E+00 l 1.39E-03 1 0.00E+00 l 3.37E-02 l0.00E+00 l 2.50E+00 l 1 l Ru-106 l 9.25E+04 l 0.00E+00 l 1.15E+04 l0.00E+00 l1.25E+05 l0.00E+00 l 1.44E+06 l ) !l l Ag-110m l 2.09E+0S l L41E+08 l 1.13E+0S l0.00E+00 l 2.63E+0S l0.00E+00 l1.68E+10 l l Cd-109 l 0.00E+00 l 3.86E+06 l1.79E+05 l 0.00E+00 l 3.45E+06 l 0.00E+00 l1.25E+07 l 1 l Sn-ll3 l 6.llE+0S l1.26E+07 l 3.4SE+07 l 9.29E+0S l0.00E+00 l 0.00E+00 l 4.32E+0S l l Sb-124 l 1.09E+0S l 1.41E+06 l 3.81E+07 l 2.40E+05 l0.00E+00 l 6.03E+07 l 6,80E+02 l t i Sb-125 l 8.71E+07 l 6.72E+05 l 1.83E+07 l 8.07E+04 l 0.00E+00 l 4.86E+07 l 2.0SE+0S l: jg l Te 125m l 7.38E+07 l 2.00E+07 l 9.85E+06 l 2.07E+07 l0.00E+00 l 0.00E+00 l 7.13E+07 l g l Te-127m l 2.0SE+08 l 5.6]E+07 l 2.47E+07 l 4.98E+07 l 5.94E+08 l0.00E+00 l 1.69E+08 l j l Te-127 l 2.98E+03 l 8.04E+02 l 6.39E+02 l 2.06E+03 l 8.48E+03 l0.00E+00 l 1.16E+05 l j s 3l l Tc.129m l 2.72E+08 l 7.59E+07 1 4.22 E+07 l8.76E+07 l 7.98E+08 l0.00E+00 l 3.31E+0S l ! N l Te-129 l 1.29E-09 j 3.61E-10 l 3.07E-10 l 9.22E-10 l 3.78E-09 l0.00E+00 l 8 04E-OS l l Tc-131m l 160E+06 1 5.54 E+ 05 l 5.90E+05 l114E+06 l 5.36E+06 l0.00E+00 l 2.25E407 l 2 lTe-131 l 1.65E-32 l 5.02E-33 l 4.90E-33 l 1.26E-32 l 4.98E-32 1 0.00E+00 l 8.64E 32 l l Te-132 l1.03E+07 l 4.54E+06 l 5.49E+06 l 6.61E+06 l 4.22E+07 l 0,00E+00 l 4.57E+07 l ll-130 l 1.73E+06 l 3.50E+06 l 1.80E+06 l 3.85E+08 l 5.23E+06 l0.00E+00 l 1.64E+06 l ) I l l-131 l1.30E+09 l1.31E+09 l 7.46E+08 l 4.34E+1I l 2.15E+09 l0.00E+00 l 1.17E+0S l ll132 l 6 92E-01 l 1.27E+00 l 5.85E-01 l5.90E+01 l1.95E+00 l 0.00E+00 l 1.50E+00 l ll133 l 1.72E+07 l 2.13E407 l 8.03E+06 l 3.95E+09 l 3.54E+07 l0.00E400 l 8.57E+06 l lI-134 l 8.56E-12 l1.59E-11 l 7.32E-12 l 3.66E-10 l 2.43E-Il l0.00E+00 l 1.05E-Il l I i i I I i iI

I_ APA-ZZ-01003 j Rev. 3 l TAHLE 3.3 (Cmil'd) . I CIHLD PATHWAY DOSE FACTORS (R;)FOR RADIONUCL,lDFS OTHER THAN NOBLE GASES Grass-Cow-Milk Pathway (m' mrem /yr) per (p Ci/sec) TOTAL _ I NUCLIDE BONE LIVER BODY Tl!YRO1D KIDNEY LUNG GI-LLI l l-135 l5.4]E+04 l 9.74E+0 l l 4.61E+04 l 8 63E+06 l1.49E+05 l0.00E+00 l 7 42EH)4 l I l Cs-134 l 2.27E+10 l 3.72E+10 l 7.84E+09 l0.00E+00 l 1.15E+10 l 4.14F-09 l 2.00E+08 l l Cs-136 l 1.01E+09 l 2.7EE+09 l1.80E+09 l0.00E+00 l 1.4SE+09 l 2.21E,08 l 9.78E+07 l l Cs-137 l 3.23E+10 l 3.09E+10 1 4.56E+09 l0.00E+00 l1.01E+10 l 3.62E+09 l 1.93E+0S l l Cs-138 l 4.03E-23 l 5.61E-23 l 3.56E-23 l0.00E+00 l 3.95E-23 l 4.25E-24 l 2.58E-23 l l Ba-139 l 2 022-07 l 1 OSE-10 l ~.86E-09 l0.00E+00 l 9.42E-11 l 6.34E-11 l 1.17E-05 l l Ba-140 l 1.17E+0S l LO3E+05 l 6.84E+06 l0.00E+00 l 3.34E+04 l 6.12E+04.1 5.94 E+07 l I l Ba-141 l 1.90E-45 l 1.06E-48 l 6.17E-4 7 l0.00E+00 l 9.19E 49 l 6.24E-48 j 1.0SE-45 l l Ba-142 l 1.21E-79 l 8.70E-83 l 6.75E-81 l0.00E+00 l 7.04E-S3 l 5.12E-83 l1.5SE-Si l l La-140 l1.78E+02 l 6.23E+01 l 2.10E+01 l 0.00E+00 l0.00E+00 1 0.0^E+00 l1.74E,06 l l La-142 lS.14E-l1 l 2.60E-11 l 8.13E-12 l0.00E+00 l000E+00 l0.00E+00 1 514E-06 l l Cc-141 l 2.19E+04 l 1.09E+04 l 1.62E+03 l0.00E+00 l 4.79E+03 l0.00E+00 l1.36E+07 l l Cc-143 l1.8SE+02 l 1.02E605 l 1.47E+01 l0.00E+00 l 4.27E+01 l0.00E+00 l1.49E+06 l l Ce-144 l 1.62E+06 l 5.09E+05 lS.67E+04 l0.00E+00 l 2.82E+05 1 0.00E+00 l1.33E+0S l l Pr-143 l 7.19Ee02 l 2.16E+02 l 3.57E+01 1 0.00E+00 l 1.17E+02 [0.00E+00 l 7.76E+05 l I l Pr-144 l 5.04E+00 l 1.56E+00 l 2.53E-01 l 0.00E+00 l S.24E-01 l0.00E+00 i 3.35E+03 l l Nd-147 l 4.45E F02 l 3.61E+02 l 2.79E+01 l 0.00E+00 l 1.98E+02 l0.00E+00 l 5.71E+05 l l Eu-154 l 9.43E+04 l S.4SE+03 l 7.75E+03 l0.00E+00 l 3.73E+04 l0.00E+00 l1.97E+06 l. .g l Hf-181 l 6 44E+02 l 2.35E+03 l 2.91E+02 l0.00E+00 l 4.76E+02 l0.00E+00 l8.66E+05 l E l W-lS7 l 2.89E+04 l 1.71E+04 l 7.69E+03 l 0.00E+00 l 0.00E+00 l 0.00E+00 1 2.41E+06 l l Np 239 l L73E+01 l1.24E+00 l 8 74E-01 l0.00E+00 l 3.60E+00 1 0.00E+00 l 9.2]E+04 l I I I I I I I

I APA-ZZ-01003 Rev. 3 TA HLE 3.3 (Cont'd) CIIILD PATIIWAY DOSE FACTORS (R;) FOR RADIONUCLlDES OTIIER Til AN NOBLE GASESa Grass-Gual-Milk Pathway (m' mrem /yr) per (pCi/sec) TOTAL NUCLIDE BONE I TVER BODY THYROID KIDNEY LUhG GTLLI IH3 l 0.00E+00 l 3.20E+03 l 3.20E+03 l 3.20E+03 i 3.20E+03 [ 3.20E+03 1 3.20E+03 l l Be-7 l 9.00E+02 l 1.53E+03l 9.84E+02 l 0.00E+00 l 1.50E+03 1 0.00E+00 l 8.55E+04 l I l C-14 l 1.20E+09 l 2.39E+0S l 2.39E+08 l 239E+08 l 2.39E+08 l 2.39E+0S l 239E+0Sl l Na 24 l 1.06E+06 l 1.06E+06 l 1.06E+06 l 1.06E+06 l 106E+06l 1.06E+06 l 1.06E+06l I P-32 l 9.34E+10 l 4.37E+09 l 3.60E+09 l 0.00E+00l 0.00E+00l 0.00E+00l 2.58E+09 l [ Cr 51 l 000E+00l 0.00E+00 l 1.22E+04 l 6 7'.~.4 03 l 1.85E+03 l L24E+04 l 6 4SE+05 l l Mn-54 l 0.00E+00 1 2.52E+06 l 6 7]E+05 l 0.00E+00 l 7.06E-05 ; 0.00E+00 l 2.llE+06 l I l Mu-56 l 0.00E+00 l 1.56E-03 l 3.52E-04 l 0.00E+00 l 1.89E 03 l 0.00E+00 l 2.26E-01 l l Fe 55 l 1.45E406 l 7.72E+05 l 2.39E+05 l 0.00E+00l 0.00E+00 l 4.36E+05 l 1.43E+05 l lFc-59 l 156E+06 l 2.53E+06 l 1.26E+06 l 0.00E+00l 0.00E+00 l 7.34E+05 l 2.64E+06 l l Co 57 l 0.00E+00 l 4.6iE+05 l 9.33E+05 l 0.00E400l 0.00E+00l 0.00E+00 1 3.78E+06 l l Co-58 l 0.00E+00l 1.46E+06l 4.46E+06 l 0.00E+00 l 0.00E+00l 0.00E+00 l 8.50E+06 l l Co-60 l 0.00E+00 l 5.19E+06 l 1.53E+07 l 0.00E+001 0.00E+00 l 0.00E+00l 2.87E+07 l I l Ni43 l 3.56E+09 l 1.91E+0S l 1.21E+08 l 0.00E+00 l 0.00E+00 l 0.00E+00.1 1.28E+07 l ! Ni-65 l 1.99E-01 l L8SE-02 l 1.10E-02 l 0.00E+00 l 0.00E+00 l 0.00E+00l 2.30E+00 l 1 Cu-64 l 0.00E+00 l S.33E+03 [ 5.03Et031 0.00E+00 l 2 ole +04l 0.00E+00 l 3.91E+05 l I l Zn-65 l 4.97E+0S l 1.32E+09l 8.23E+0S l 0.00E+00l 8 34E+0S l 0.00E+00l 2.32E+08 l l Zn-69 l 1.15E-12 l 1.66E-12 l 1.53E-13 1 0.00E+00 1 1.00E-12 l 0.00E+00 l 1.04E-10 l l Br-8f l 0.00E+00 l 0.00E+00 l 1.3SE+0? l 0.00E+00 1 0.00E+00 l 0.00E+00 1 0.00E+00 l l Br-83 l 0.00E+00 l 0.00E+00 l 5.30E-02 1 0.00E+00 l 0.00E+00[ 0.00E+00 l 3.10E 19 l 1 l Br-84 l 0.00E+00 l 0.00E+00 l 7.92E-24 l 0.00E+001 0.00E400 l 0.00E+001 4.00E-41 l j ig l Br-85 l 0.00E+00 l 0.00E+00l 0.00E+00l 0.00E+00 l 0.00E+00l 0.00E+00l 0.00E+00 l -l l'g l Rb-86 l 0.00E+00 l 1.05E+09 l 6 48E+0S j 0.00E+00l 0.00E+00 l 0.00E+00 l 6.78E+07 l { l Rb-88 l 0 00E-no l S.80E-46 ! 611E-46 l 0.00E+00 1 0.00E+00 l 0.00E+001 4.32E-47 l l l Rb 89 l 0.00E+00 l 1.66E-53 l 14SE 53 i 0.00E+00 l 0.00Ew30l 0.00E+00 l 1.45E 55 l 1

W l Sr-89 l 1JuE+10 l 0.00E+00 l 3.97E+08 l 0.00E+00{ 0.00E+00[ 0.00E+00 l 5.39E+0S [

j l Sr 90 l 2.35E+11 } 0.00E+00l 5.95E+10 l 0.00E+00 l 0.00E+00 l 0.00E+00 l 3,16E+09 l i l Sr-91 l 2.74E+05 l 0.00E+00 l 1.04E+04 l 0.00E+00l 0.00E+00l 0.00E+00 l 6.06E+05 l lg3 i St-92 l 4.60E+001 0.03E+00 l 1.84E-01 l 0.00E+00l 0.00E+00 l 0.00E+00 l 8.72E+01 l l l Y-90 l 4.06E+02 l 0.00E+00l 1.09E+01 1 0.00E+001 0.00E+00 l 0.00E+00 l 1.15E+06 l 35- . I

l APA-ZZ-01003 IW Res,3 3-E T^nLE 3 3 (cand)

E-l CHILD PATHWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTIIER TilAN l

NOIlLE GASESa Grass-Goat Milk Pathway (m mrem /yr) per (pCihec) 2 TOTAL l NT'CLIDE IlONE LIVER llOD Y TIIYHOID h'ID NE Y LUNC GILLI l Y-91m l 3.24E-20 1 0.00E+00 l 1.18E-21 l 0.00E+00l 0.00E+00l 0.00E+00 l 6.35E-17 l ll l Y-91 l 4.69E+03 l 0.00E+00l 1.25E+02 l 0.00E400l 0.00E+00 l 0.00E+00 l 6.25E+05 l j W l Y-92 l 3.0$E-05 l 0.00E+00 l 8.72E-07 l 0.00E+00l 0.00E+00l 0.00E+00 l 8.80E-01 1 l Y-93 l 1.22E-01 l 0.00E+00 l 3.34E-03 l 0.00E+00 l 0.00E+00l 0.00E+00l 1.8]E+03l l Zr 95 l 4.60E+02 l 1.0 l E+02 l

9. ole +01 l 0.00E400l 1.45E+02 l 0.00E+00l 1.06E+05l

!I j l Zr-97 l 2.3 ] E-01 1 3.33E-02 l 1.97E 02 l 0.00E+00 l 4.78E-02 l 0.00E+00 l 5.03E+03 l ll l Nb-95 l 4.46E404 l 1.74E+04 l 1.24 E + 04 1 0.00E+00 l 1.63E+04 l 0.00E+0C l 3.21E+07 l 3 l Mo-99 l 0.00E+00 l 9.78E+06 l 2.42E+06 l 0.00E+00l 2 09E+071 0.00E+00 l S.09E+06 l j W [ Tc 99m l 2 26E+03 l 4.44E+03 l 7.35E+04 l 0.00E+00 l 6.44E+04 l 2.25E+03 l 2.52E+06 l j l Tc-101 l 1.44E-60 l 1.51E-60 l 1.91E-59 l 0.00E+00l 2.57E 591 7.96E-61 l 4.79E-60 l l Ru-103 l 5.14E+02 l 0.00E+00 l 1.98E+02 1 0.00E+00 l 1.29E+03 l 0.00E+00l L33E+04 l l Ru 105 l 4.60E-04 } 0.00E+00 l 1.67E-04 l 0.00E+00 l 4.04E-03 l 0_00E+00 l 3.00E-01 l 4 i l Ru-106 l 1.llE+04 l 0.00E+00 l 1.3SE+03 l 0.00E+00l 1.50E+04 l 0.00E+00 l 1.73E405 l ll l Ag-110m l 2.51E+07 l 1.69E+07 l 1.35E*07l 0.00E+00 l 3.15E+07 l. 0.00E+00 l 2 ole +09 l l um l Cd-109 l 0.00Et00 l 4.64E+05 l 2.15E+04 l 0.00E+00l 4.14E+05 l 0.00E+00 l 1.50E+06 l l Sn-113 l 7.33E+07 l 1.51E+06 l 4.18E+06 l L1lE+08l 0.00E+00l 0.00E+00 l 5.18E+07 l l Sb 124 l 1.30E+07 l 1.69E+05 l 4.57E+06 l 2.SSE+04 l 0.00E+00 l 7.24E+06 l 8.16E+07 l l l Sb-125 l 1.05E+07 l S.06E+04 l 2.19E+06 l 9.6SE+03 l 0.00E+00 l 5.83E+06 l 2.50E+07 1 f i Te-125m ! 8.86E+06 l 2.40E+06 l 1.lSE+06 l 2.49E+06 l 0.00E+00 l 0.00E+00 l 8.55E+06 l I l Te-127m 1 2.50E+07 l 6.73E+06 l 2.97E+06 l 5.98E+06l 7.13E+07 l 0.00E+00l 2.02E+07 l l Te 127 l 3.5SE+02 l 9.65E+01 1 7.67E+01 l 2.48E+02 l 1.02E+03 l 0.00E+00 l 1.40E404 l I I Tc-129m l 3.26E+07 l 9.10E+06 l 5.06E+06 l 1.05E+07 l 9.57E+07 l 0.00E+00l 3.9SE+07 l l Te-129 l 1.55E 10 l 4.33E Il ) 3.6SE-11 l 1.!lE10l 4.54E 10 l 0.00E+00 l 9.65E-09 l l Te-131m l L92E+05 l 6.65E+04 l 7.07E-04 l 1.37E+05 l 6.43E+05 l 0.00E+00 l 2.70E+06 l I Te-131 l 1.97E-33 l 6.02E-34 l 5.8SE-34 l 1.51E 33 l 5.97E-33 l 0.00E+00 l 1.04E-32 l lTe-132 1 1.23E+06 l 5.45E+05 l 6.58E+05 l 7.94E+05 l 5.06E+06 l 0.00E+00 l 5.49E+06 l l l-130 l 2.08E+06 l 4.20E+06 l 2.16E+06 1 4.62E+08 l 6.27E+06 l 0.00E+00 l 1.96E+06 l I ll-131 l L57E+09 l 1.57E+09 l 8.95E408 l 5.21E+11 l 2.58E+09 l 0.00E+00( l40E+0Sj l I-132 l 8.30E-01 l 1.53E+00l 7.02E-01 l 7.08E+01 l 2.34E+00 l 0.00E400 l 1.80E+00 l l l-133 l 2.06E+07 l 2.55E+07 l 9.66E+06 l 4.74E+09 l 4.25E+07 l 0.00E+00 l 103E+07 l i 1-134 l 1.03E-11 l 1,91E-11 l 8.7SE-12 l 4.39E.10 l 2.92E-11 l 0.00E+00 l 1.27E-11 l I I . I -3 6-I

• +

I APA-ZZ-01003 Rev. 3 TAllt.E 3.3 (Cont'd) CillLD PATHWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTilER TII AN NOBLE GASESa Grass-Goat-Milk Pathway (m' mrem /yr) per (pCi/sec) TOTAL NUCLTDE BONE LTVER BODY TIIYROID KIDNEY LUNG CI-LLI l l-135 l 6.49E+04 l 1.17E+05 l 5.53E+04 l 1.04E+07 l 1.79E+05 1 0.00E+00 l 8.90E+04 l I l Cs-134 l 6 80E+10 l 1.12E+11 l 2.35E+10 l 0.00E+001 3.46E+10 l 1.24E+10l 6.01E+0S l l Cs-136 l 3 04E+09 l S.35E+09 l 5.40E-09 l 0.00E+00l 4.4 5E+09 I 6.63E+08 l 2.93E+0S i l Cs-137 l 9 6SEv10 l 9.27E+10 l 1.37E+10 1 0.00E+ 00 1 3.02E410 l 1.09E+10 l 5.80E+0S l l Cs-138 l 1.21E 22 l L68E-22 l LO7E 221 0.00E+00 l 1.18E-22 l 1.27E-23 l 7.75E-23 l I J Ba-139 l 2.42E-OS l 1.29E-1! l 7.03E-10 ! 0.00E+00l L13E-11l 7.61E-12 l 1.40E-06 l l Ba-140 l L41E+07 l 123E+04 l 8.21E+05 l 0.00E+00 l 4.0!E403 l 7.35E+03 l 7.13E406 j I l Ba-141 l 2.2SE-46 l 1.27E-49 l 7.4 IE-48 l 0.00E+00 l 1.10E-49 l 7.4 9E-4 9 l 1.30E-46 l l Ba-142 ] 1.45E-80 l 1.04E-S3 l 8.10E-S2 1 0.00E+00 l 8.45E-84 [ 6.14E-84 l 1.89E-82 l l La !40 l 2.14E+01 1 7.47E+00 j 2.52E+00 l 0.00E+001 0.00E+00 l 0.00E+00 l 2.0SE+05 l l La-142 l 9.77E-12 l 3.1lE-12l 9.75E-13 1 0.00E+00 l 0.00E+00l 0.00E+00 l 6.17E-07. l l Ce 141 l 2.63E+03 l 1.31E+03 l 1.95E+02 l 0.00E+00 l 5.75E+02 l 0.00E+001 1.63E+06 l l Cc-143 l 2.25E+01 l 1.22E+04l 1.77E+00l 0.00E+00l 5.12E+00 l 0.00E+00 l 1.79E+05 l I-l Cc-144 l 1.95E+05 l 6.llE+04 l 1.04E+04 l 0.00E+00l 3.3SE+04 [ 0,00E+00[ 1.59E+07 l l Pr-143 1 8 63E+01 l 2.59E+01 1 4.2SE+00 l 0.00E+00 l 1.40E+01l 0.00E+00 l 9.3]E+04 l j g l Pr-144 } 6.05E-01 l 1.87E-01 l 3.04E-02 I 0.00E+00 l 9.89E-02 1 0.00E+00 l 4.03E+02 l g i Nd-147 l 5.34E+01 l 4.33E+01 l 3.35E+00 l 0.00E+00l 2.37E+01 l 0.00E+00 l 6.85E+04 l l Eu 154 [ 1.13E+04 l 1.02E+03 l 9.29E+02 l 0.00E+001 4.47E+03 1 0.00E+00 } 2.37E+05 { l Hf-181 1 7.73E+01 l 2.SIE+02 l 3.49E+01 l 0.00E+00 l 5.72E+01 1 0.00E+00 l 1.04E+C5 l },l l W-]S7 I 3.47E+03 l 2.06E+03 l 9.22E+02 l 0.00E+001 0.00E+00 l 0.00E+00l 2.89E+05 l m l Np-239 l 2.08E+00 l 1.49E-01 l 1.05E-01 l 0.00E+00 l 4.32E-01 l 0.00E+00 l 1.10E+04 l 1 i l3 i 4 i g 37

' g' APA ZZ-01003 g Rev. 3 TAllII 3.3 (Cont'rl) j g i 3-CHILD PATHWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTilER THAN J g NOBLE GASES 3 - E Vegetation Pathway i (m' mrem /yr) per (pCihec) TOTAL 2 NUCLIDE BONE LIVER BODY TIIYROID KIDNEY ~ LUNG GILLI e j j143 l ND l 4.01E+03 l 4 ole +03 l 4.01E+03 l 4 ole +03 l 4.01E+03 l 4.01E+03 l f l Be-7 l 3.38E-05 l 5.76E+05 l 3.70E+05 l 0.00E+00 l 5.65E+0) l 0.00E400 l 3.21E+07 l !gE l C-14 l 8.89E+0S l 1.78E+0S l 1.78E*0S l 1.78EtOS l 1.78E408 l 1.78E+0S l 1.78E+0S l l l Na-24 l 3.73E+05 l 3.73E+05 l 3.73E+05 l 3.73E405 l 3.73E+05 l 3.73E+05 l 3.73E+05 l I j l P-32 l 3.37E+09 l 1.5SE+0Sl 1.30E+0Sj 0.00E+00l 0.00E+00 1 0.00E+00 l 9.32E+07 l us l Cr-51 l 0.00E+001 0.00E+00 l 1.17E+05 l 6.50E+04 l 1.78E+04 l 1.19E+05 { 6.21E+06 l l Mn-54 1 0.00E+00 l 6 65E+0S l 1.77E+0S l 0.00E+00l 186E+0S l 0.00E+00 l 5.58E+0S l l Mn-56 l 0.00E+00 l 1.89E+01l 4.26E+00 l 0.00E+00 l 2.28E+0 ) l 0.00E+00l 2.74E+03 l I l Fe 55 l 8.01E+08 l 4.25E+0S l 1.32E+0S l 0.00E+00l 0.00E+00l 2.40E+0S l 7.87E+07 l J l Fe-59 l 3.98E+08 l 6.43E+08 l 3.20E+0S l 0.00E+001 0.00E+00 l 1.87E+08 l 6.70E+08 l I l Co-57 l 0.00E+00 l 2.99E+07 l 6.04E+07 l 0.00E+00l 0.00E+00l 0.00E+00l 2.45E+08 l l Co-58 l 0.00E+00l 6.44E+07 l 1.97E+0Sl 0.00E+001 0.00E*00 l 0.00E+00 l 3.76E+0S l l Co-60 1 0.00E+00 l 3.78E+0S l 1.12E+09j 0.00E400l 0.00E+00l 0.00E+001 2.10E+09 l -l _I l Ni-63 l 3.95E+10 l 2.1lE+09 l 1.34E+09 1 0.00E+00 l 0.00E+00l 0.00E+00 l - 142E+0S j l Ni-65 l 105E+02l 9.90E+00 j 5.7EE+00 1 0.00E+00 l 0.00E+00 l 0.00E+00 l 1.21E+03 l l Cu-64 [ 0.00E+00 l 1.10E+04 l 6.64E+03 1 0.00E+00 l 2.66E+04 l 0.00E+00 l 5.16E+05 l I l Zn-65 l S.13E+0S l 2.17E+09 l 1.35E+09 l 0.00E+00l 1.36E409l 0.00E+00 l 3.80E+08 l l Zn-69 l 9.53E-06 l 1.38E-05 l 1.27E-06 l 0.00E+00 l 8.35E-06 l 0.00E+00 l 8.68E-04 l l Br-82 l 0.00E+00 l 0.00E+00l 2.04E+06 l 0.00E+00l 0.00E+001 0.00E+00 l 0.00E+00 l l Br-83 l 0.00E+001 0.00E+00 l 5.38E+00 l 0.00E+00 l 0.00E+00l 0.00E+00 l 3.14E-17 l l Br44 l 0.00E+00l 0.00E+00l 3.85E-111 0.00E+00 l 0.00E+00 l 0.00E+00l L94E-28 l l Br-85 l 0.00E+00l 0.00E+00l 000E+00l 0.00E+00l 0.00E+00 l 0.00E+00 l 0.00E+00 l I l Rb S6 j 0.00E+00l 4.52E+0S l 2.78E+0S l 0 00E+00 l 0.00E+00 1 0.00E+00 l 2 9]E+07 l l Rb-88 l 0.00E+00 l 4.43E-22 l 3.0SE-22l 0.00E+001 0.00E+00 l 0.00E+00 l 2.17E-23 l l Rb-89 l 0.00E+00l 4.67E 26 l 4.15E-26 l 0.00E+00 l 0.00E+00l 0.00E+00 l 4.07E-28 l l Sr-S9 l 3.60E+10 l 0.00E+00l 1.03E+09 l 0.00E+00 l 0.00E+00 1 0.00E+00 l 1.39E+09 l l Sr-90 l 1.24E+12 l 0.00E+00 l 3.15E41 ) [ 0.00E+00 l 0.00E+00l 0.00E+00 l 1.67E+10 l l Sr-91 l 5.24E+05 l 0.00E+00 l 1.98E+04 l 0.00E+00l 0.00E+00 J 0.00E+00 l 1.16E+06 l I l Sr-92 l 7.29E+02 l 0.00E+00 l 2.92E+01 ! 0.00E+00 l 0.00E+00l 0.00E+00 l 1.3SE+04 l i j Y-90 l 3.01E+06 l 0.00E+00 l 8.04 E+04 1 0.00E+00 l 0.00E+00 l 0.00E+00 l 8.56E+09 l l I I I I

1 I' APA-ZZ-01003 Rev. 3 TAI!LE 3.3 (Cont'd) I CHILD PATIIWAY DOSE FACTORS (Rj) FOR RADIONUCLIDES OTIIER TilAN I. NOllLE GASESa l Vegelation Pathway (m' mrem /yr) per (pCihec) TOTAL NUCLTT)E BONE LTVE R BODY Til V R O 1 D KIDNEY LUNG GI-LLI l Y-91 m l 8.95E-09 l 0.00E40l 3.26E-10 l 0.00E+00l 0.00E+00 l 0.00E+00l 1.75E-05 l I l Y-91 l 1.86E+07 1 0.00E+00 l 4.99E+05 l 0.00E+00 1 0.00E+00 1 0.00E+00 l 2.48E+09 l l Y-92 l 1.59E+00 l 0.00E+00l 4.54E-02 l 0.00E+00 1 0.00E+001 0.00E+00 l 4.5SE+04 i l Y-93 l 2.93E+02 l 0.00E+00 l 8.05E+00 l 0.00E+00l 0.00E+00l 0 00E+00 l 4.37E+06 l l Zr-95 l 3 86E+06 l 8 4SE+05 l 7.55E+05 l 0.00E+00 l 1.21E+06 l 0.00E+00 l 8.85E+0S l I l Zr-97 l 5.70E+02 l 8.24E+0!l 4.86E4 01 l 0.00E+00 i L18E+02l 0 00E+00 l 1.25E+07 l l Nb-95 1 7.4 S E+05 1 2.9)E+05 l 2.0SE+05 l 0.00E+00l 2.74E+05 1 0.0(.E+00 [ 5.39E+08 I I l Mo.99 l 0.00E+00 l 7.7]E+06 l 1.9]E+06 l 0.00EM)0 l 1.65E+07l 0.00E,00[ 6.38E+06 1 l Tc-99m l 5.35E+02 l 1.05E+03 l 1.74E+04 l 0.00E+00l 1.53E+04 l 5.33E+021 5.97E+05 l l Tc-101 l 1.43E-30 l 1.49E-30 l L89E-5 l 0.00E+00 l 2.55E-29 l 7.90E-31 l 4.75E-30 l l Ru-103 l 1.53E+07 l 0.00E+00 l 5.90E+06 l 0.00E+00 1 3.86E+07 l 0.00E+00 l 3.97E+0S l l Ru-105 l 9.17E+0i l 0.00E+00 l 3.33E+01 l 0.00E+00 l 8.06E+02 l 0.00E+00i.5.99E+04 l l Ru-106 l 7.45E+08 l 0.00E+00 l 9.30E+07 l 0.00E+00 l 1.01E+09 l 0.00E+00 l 1.16E+10 l ) I ] Ag-l10m l 3.21E+07 l 2.17E+07 l 1.73E+07 1 0.00E+001 4.04E+07 l 0.00E+00 l 2.58E+09 l l Cd-109 l 0.00E+00 l 2.45E+0S l 1.14E+07 l 0.00E+00l 2.18E+0S l 0.00E+00l 7.94E+08 l l Sn-ll3 l 1.58E+09 l 3.25E+07 l 9.00E+07 l 2.40E+09 l 0.00E+00 l 0.00C+00 l 1.12E+09 l l Sb-124 l 3.52E+0S l 4.57E+06 l 1.23E+0S l 7.77E+05 l 0.00E+00 l 1.95E+08 l 2.20E+09 l l Sb-125 l 4.99E+0S l 3.S5E+06 l 1.05E+0Sl 4.63E+05 l 0.00E+00 l 2.78E+08 l 1.19E+09 l l Te-125m j 3.5]E+0S l 9.50E+07 l 4.67Ev07 l 9.84E+07 l 0.00E+001 0.00E+00 l 3.38E+08 l I l Te-127m l L32E+09 l 3.56E+08 l 1.57E+0S l 3.16E+0S l 3.77E+09 l 0.00E+00 l 1.07E+09 l lTc-127 l 9.85E+03 l 2.66EH)3 l 2.llE+03 l 6.82E+03 l 2.80E+04 l 0.00E+00 l 3.85E+05 l l Te-129m l 8.4]E+08l 2.35E+0S l 1.3]E+0Sl 2.71E+0S l 2.47E+09 l 0.00E+00 l 1.03E+09 l I l Te-129 l 1.33E-03 l 3.70E-04 l 3.15E-04 [ 9.46E-04 l 3.88E-03 l 0.00E+00l 8.25E-02 l l Te-131m i 1.54E+06 l 5.33E+05 l 5.68E+05 l 1.10E+06 l 5.16E+06 l 0.00E+00 l 2.16E+07 l ] Te-131 l 2.59E-15 l 7.90E-16 l 7.7]E-16l 1.98E-15 l 7.84E-15 l 0.00E+00 l 1.36E-14 l l Te-132 l 7.00E+06 l 3.10E+06 l 3.74E+06 l 4.5]E+06 l 2.8SE+07 l 0.00E+00 l 3.12E+07 l l1130 l 6.16E+05 l L24E+06 l 6.41E+05 l 1.37E+08 l 1.86E+06 l 0.00E+00 l 5.82EM)5 l g l l-131 l L43E+0S! 1.44E+0S l 8.17E+07 l 4.76E+10 l 2.36E+0S l 0.00E+00 l L28E+07l 3 ll-132 l 9.23E+01 l 1.70E+02 l 7.80E+01 [ 7.87E+03 l 2.60E+02 1 0.00E+00 l 2.00E+02 l l1133 l 3.53E+06 l 4.37E+06 l 1.65E+06 [ 8.12E+08 l 7.28E+06 l 0.00EH)0l 1.76E@6 l ll-134 l 1.56E-04 l 2.89E-04 l 1.33E-04 l 6.65E-03 1 4.42E-04 l 0.00E+00 l 1.92E 04 l I I -39 I

!l APA-ZZ-01003 4 LE llev. 3 g TA!!LE 3.3 (Cont'd) 5 CHILD PATilWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTilER THAN NOBLE GASESa [ Vegetation Pathway (m mremlyr) per (pCihec) 2 l' TOTAL l NUCLlDE HONE LnTR HODY TIIYROID KIDNEY LUNG El-Lil I l l-135 { 6.26E+04 l 1.13E+05 l 5.33E404 l 9.98E+06 l 1.73E+05 1 0.00E+00 l 8.59E+04 l jg l Cs-134 l 1.60E+10 ! 2.63E+10 l 5.55E+09 l 0 00E+00 l S.15E+09l 2.93E+09 l 1.42E+0S l l Cs-136 l 8.24E+07 l 2.27E+0S l 1.47E+0S l 0,00E+00 l 1.21E+0S l 1.80E+07 l 7.96E+06 l }3 l Cs-137 l 2.39E+10 l 2.29E+ 10 1 3.3SE409 l 0.00E400 l 7.46E+09 l 2.6SE+09 l 1.43E+0Sl i l Cs-138 l 6 61E-Il l 9.20E-Il l 5.83E-11 l 0.00E400 l 6.47E-Il I 6.96E-12 l 4.24E-Il l l En-139 l 4.97E-02 l 2.65E-05 l 1.44E-03 l 0.00E+00 l 2.32E-05 1 1.5CE-05l 2.87E+00 l i Ba-140 1 2.77E+0S l 2.43E+05 l 1.62E+07 l 0.0.' E+00 l 7.90E+04 l 1.45E+05 l 1.40E+0S [ i ig l Ba.141 l 2.01E-21 l 1.13E-24 l 6.54E-23 l 0.00E400 l 9.74E-25 l 6.61E-24 l 1.15E-21 j jg l Ba-142 l 1.02E-38 l 7.31E-42 l 5.67E-40 l 0.00E+00 l 5.91E-42 l 4.30E-42 l 1.32E-40 l t l La-140 l 3.36E+04 l 1.lSE+04 l 3.96E+03 l 0.00E+00 ! 0.00E+00 l 0.00E+00 l 3.2SE+0S l l La-142 l 3.37E-04 l 1.07E-04 l 3.36E-05 l 000E+00l 0.00E+00 l 0.00E+00 l 2.13E+01 l l Cc 141 l 6.56E+05 l 3.27E+05 l 4.86E+04 1 0.00E+00 l 1.43E+05 l 0.00E+00 l 4.0SE+0S l l Ce-143 l 1.72E+03 l 9.31E+05 l 1.35E+02 l 0.00E+00 l 3.91E+02 l 0.00E+00l 1.36E407 l lI l Cc-144 l 1.27E+0S l 3.98E+07 l 6.78E+06 ! 0.00E+00 l 2.2]E+07 l 0.00E+00J 1.04E+10 l l Pr-143 l 1.46E-05 1 4.37E+04 l 7.23Et03 l 0.00E-00 l 2.37E+04 l 0.00E+00 l 1.57E+0S l l Pr-144 l 7.8SE+03 1 2.44E'03 l 3.97E+02 l 0.00E+00l 1.29E+03 1 0.00E+00 l 5.25E+06 l j I l Nd-147 l 7.15E+04 l 5.79E+04 l 4.4SE+03 l 0.00E+00 l 3.lSE+04 l 0.00E+00 l 9.17E+07 l l Eu-154 l 1.66E+0S l 1.50E+07 l 1.37E+071 0.00E+00 l 6.57E+07 l 0.00E+00{ 3.48E+09 l l Hf-181 l 4.90E+05 l 1.79E+06 l 2.21E+05 l 0.00E+00l 3.63E+05 l 0.00E+00 l 6.59E+08 l I l W-IS7 l 6.44E+04 l 3.SIE+04 l 1.71E+04 l 0.00E+00l 0.00E+001 0.00E+00 l 5.36E+06 l [ Np-239 l 2.57E+03 J 1.84E+02 l 1.30E+02 l 0.00E+00 l 5.33E+02 l 0.00E+00 l 1.36E+07l I I I . I I I -4 0-I

I APA-ZZ-01003 ' Rev. 3 I -TAllLE 3.4 ADULT PATHWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTilER TIIAN NOBLE GASES" Inhalation Pathway (mrem /) r) per (pCi/rv) I TOTAL N t1CI IDE BONE LTVE R BODY Tl!YROlD KIDNEi' l.llNG GI-LLI l H-3 i ND l 1.26E+03 1 1.26E+03 l 1.26E+03 l1.26E+03 l 1.26E+03 l 1.26E+03 l I l Be-7 l 4.27E+02 l 9.6EE+02 l 4.70E-02 l ND l ND l 4.21E+04 1 5.35E+03 l l C-14 l 1.S*'E+04 l 3.41E+03 l 3.41E+03 l 3.41E+03 l 3.41E+03 l 3.41E+03 l 3.41E+03 l [ Na-24 l1.02E+04 l1.02E+04 l 1.02E+04 l 1.02E+04 l1.02E+04 l1.02E+04 l 1.02E+04 l lP-32 l L32E+06 1 7.71 E+04 l 5.01E+04 l ND l ND l ND l 8.64E+04 l l Cr-51 l ND l ND l 1.00E402 l 5.95E+01 l 2.2SE+01 l 1.44E+04 1 3.32E+03 l l MnA4 l ND l 3.96E+04 l 6.30E+03 i ND l 9.84E403 l 1.40E+06 l 7.74E+04 l I l Mn-56 l ND l1.24E+00 l 1.83E-01 l ND l1.30E+00 l 9.44E+03 l 2.02E+04 j l Fe 55 l 2.46E+04 l 1.70E+04 l 3.94E403 l ND l ND l 7.21E+04 l 6.03E+03 l l Fe-59 l 1.lSE+04 [ 2.78E+04 l 1.06E+04 l ND l ND l 1.02E+06 l 1.8SE+05 l l Co-57 l ND l 6.92E+02 l 6.7]E+02 j ND l ND l 3.70E+05 l 3.14E+04 l l Co-58 l ND l1.5SE+03 l 2.07Ev03 l ND l ND l 9.28E+05 l 1.06E+05 l l Co-60 l ND l 1.15E+04 l1.4SE+04 l ND l ND l 5.97E+06 l 2.85E+05 l I l Ni-63 l 4.32E+05 l 3.14E+04 l1.45E+04 l ND l ND l 1.78E+05 l1.34E+04 l l Ni-65 l1.54E+00 l 2.10E-01 l 9.12E-02 l ND l ND l 5.60E+03 l1.23E+04 l -l l Cu-64 l ND l 1.46E+00 l 6.15E-01 l ND l 4.62E+00 l 6.78E+03 l 4.90E+04 l M l Zn-65 l 3.24E+04 l 1.03E+05 1 4.66E+04 l ND l 6.90E+04 l 8.64E+05 l5.34E+04 l l Zn-69 l 3.38E-02 j 6.51E-02 l 4.52E-03 l ND l 4.22E-02 ! 9.20E+02 l 1.63E+0! l I l Bi-82 l ND l ND l 1.35E+04 l ND l ND l ND l 1.04E+04 l l Br-83 J ND l ND l 2.41E+02 l ND l ND l ND l 2.32E+02 l l Br-84 l ND l ND l 3.13E+02 l ND l ND l ND l 1.64E-03 l l Br 85 l ND J ND j1.28E+01 l ND l ND l ND l S.00E-15 l l Rb-86 l ND l 1.35E+05 l 5.90E+04 l ND l ND l ND l1.66E+04 l l Rb-88 l ND l 3.87E+02 l 1.93E402 l ND l ND l ND l 3.34E-09 l l Rb-89 l ND j 2.56E402 l1.70E+02 l ND l ND l ND l 9,28E-12 l l St-89 l 3.04E+05 l ND l 8.72E+03 l ND l ND l 1.40E+06 1 3.50E+05 l l Sr-90 l 9.92E+07 l ND l 6.10E+06 l ND l ND l 9.60E+06 l 7.22E+05 l I l Sr-91 l 6.19E+01 l ND l 2.50E+00 l ND l ND l 3.65E+04 l1.91E+05 l l Sr-92 l 6.74E+00 l ND l 2.91E-01_ l ND l ND 1 1.65E+04 l 4.30E+04 l l Y 90 l 2.09E+03 l ND l5.61E+01 l ND 1 ND l 1.70E+05 1 5.06E+05 l I I -41 I 3 .c e

I APA ZZ-01003 Rev. 3 TA HLE 3.4 (Cont'd) I. AL) ULT PATilWAY DOSE FACTORS (R;) FOR RADIONUCLLDES OTilER TilAN NOHLE GASESn Inhalation Pathway (m rem /,s r) per (p cum') TOTAL NUCLIDE BONE LIVER BODY TIIYROID NIDNEY LliNG GI-LLI l Y-91m l 2.6 l E-01 l ND l 1.02E-02 l ND l ND l1.92E+03 l 1.33E+00 l l Y-91 l 4.62E405 [ ND l 1.24E+04 l ND l ND lJ70E+06 l 3.85E+05 l I J Y-92 l1.03E+01 l ND l 3 02E-01 l ND l ND l 1.57E+04 l 7.35E+04 l l Y 93 l 9.44E+01 l ND l 2 61Et00 1 ND l ND l 4.85E44 l 4.22E+05 l j Zr-95 l 1.0'Ed o$ l 3.44E4 04 l 2.33E+04 l ND l 5.42E+04 l 1.77E+06 - l 1.50E+05 l l Zr-97 l 9.68E+0! l 1.96E+01 l 9.04E+00 l ND l 2.97E+01 l 7.87E+04 l 5.23E+05 l l Nb-95 l1.4IE+04 l 7.82E+03 l 4.21E+03 l ND l 7.74E+03 l 5.05E+05 l 1.04E+05 l l Mo-99 l ND l 1.21E+02 l 2.30E+0! l ND l 2.9]E+02 l 9.12E+04 l 2.48E+05 l I l Tc-99m l 1.03E-03 l 2.91E-03 1 3.70E-02 l ND l 4.42E-02 l 7.64E+02 l 4.16E+03 l l Tc-101 l 4.lSE-05 l 6.02E-05 l 5.90E-04 l ND 1 1.08E-03 l 3.99E+02 l 1.09E-Il l I l Ru-103 l 1.53E+03 l ND l 6.58E+02 l ND l 5.83E+03 l 5.05E+05 l1.10E+05 l l Ru 105 l 7.90E-01 l ND l 3.1lE 01 l ND l 1.02E+00 l 1.10E+04 l 4.82E4)4 l l Ru 106 l 6.91E+04 [ ND l8.72E+03 i ND l 1.34E+05 l 9.36E+06 l 9.12E+05 l' I l Ag-110m l 1.0SE+04 l 1.00E404 l 5.94E+03 l ND l 1.97E+04 l 4,63E+06 l 3.02E+05 l l Cd-109 l ND l 3.67E+05 l1.31E+04 l ND l 3.57E+05 l 6.S3E+05 1 5.82 E+04 l l Sn-113 l 5.72E+04 l 2.18E+03 l 4.39E+03 l 1.24E+03 i ND l 9.44E+05 l 1.18E+05 l l Sb-124 l 3.12E+04 l5.89E+02 l 1.24E+04 l 7.55E+01 l ND l 2.48E+06 l 4.06E+05 l l Sb-125 l 5.34E+04 l 5.95E+02 l 1.26E+04 l 5.40E+01 l ND 1 1.74E+06 l 1.01E+05 l l Te-125m l 3.42E+03 l 1.58E+03 l 4.67E+02 l1.05E+03 l 1.24E+04 l 3.14E+05 l 7.06E+04 l I l Te-127m l 1.26E+04 l5.77E+03 l 1.57E+03 l 3.29E+03 l 4.58E+04 l 9.60E+05 l 1.50E+05 l l Te-127 i L40E+00 l 6.42E-01 l 3.10E-01 l1.06E+00 l 5.10E+00 l 6.5]E+03 - l 5.74E+04 l l Te-129m l 9.76E+03 l 4.67E+03 ) 1.58E403 l 3.44E+03 l 3.66E+04 l 1.16E+06 l 3.83E+05 l I l Tc-129 l 4.98E-02 l 2.39E-02 l 1.24E-02 l 3.90E-02 l 1.87E-Ol l 1.94E+03 l 1.57E+02 l l Te-131m l 6.99E+01 l 4.36E+0) l 2.90E+01 15.50E+01 l 3.09E+02 l1.46E+05 l 5.56E+05 l l Te-131 l1.11E-02 l 5.95E-03 l 3.59E-03 l 9.36E-03 l 4.37E-02 l 1.39E+03 l 1.84E+01 l l Te-132 l 2.60E+02 l 2.15E+02 l1.62E+02 l1.90E+02 l1.46E+03 l 2.88E+05 l 5.10E+05 l l l-130 l 4.58E+03 l 1.34E+64 l5.28E+03 l 1.14E+06 l 2.09E+04 l ND l 7.69E+03 l ll-131 l 2.52E+04 l 3.58E+04 l 2.0$E+04 l1.19E+07 l 6.13E+04 l ND l 6.28E+03 l I l l-132 l1.16E+03 1 3.26E+03 l 1.16E+03 l 1.14E+05 l 5.18E+03 l ND l 4.06E+02 l Il133 l 8.64E+03 l1.48E+04 j 4.52E+03 l 2.15E+06 l 2.58E+04 l ND I 8.88E+03 l Il-134 l 6.44E+02 l 1.73E+03 l 6.15E+02 l 2.98E404 l 2.75E+01 l ND l1.01E+00 l I -4 2 - 1 m t --- 4

~ APA-ZZ-01003 Rev. 3 TAllLE 3.4 (Coni'd) . I ADULT PATIIWAY DOSE FACTORS (R;) FOR llAD10NUCLIDES OTilEll TIIAN NOBLE GASESa Inlialation Pathway (mremlyr) per (p Ci/m') TOTAL NilCLIDE BONE ITVER HODY TilVROID KIDNEY LUNG GILLI l1135 l 2.68E+03 l 6.98E+03 l 2.57E403 l 4.48E+05 l1.1lE+04 l ND l5.25E+03 l l Cs 134 l 3.73E+05 ] 8.4SE+05 l 7.2SE+05 l ND l 2.87E+05 l 9.76E+04 l LO4E+04 l I i Cs-136 l 3.90E+04 l 1.46E+05 l1.10E+05 l ND lS.56E+04 l1.20E+04 [1.17E+04 l l Cs-137 l 4.78E+05 l 6.21E+05 l 4.28Et05 l ND l 2.22E+05 l 7.52E+04 l 8.40E+03 l l Cs 138 l 3.31E+02 l 6.21E+02 l 3.24E+02 l ND l 4.80E+02 l 4.86E+0l l 1.86E-03 l l Ba-139 ! 9.36E-01 l 6.66E-04 l 2.74E-02 l ND l 6.22E-04 l 3.76E403 l 8 96E+02 l l Ba-140 l 3.90E+04 l 4.90E+01 l 2.57E-03 l ND l1.67E+01 l 1.27E+06 l 2.18E+05 ! l Ba ;41 l 1.00E-01 l 7.53E-05 l 3.36E-03 l ND l 7.00E-05 l1.94E+03 l 1.16E-07 l I l Ba 142 l 2.63E-02 l 2.70E-05 l 1.66E-03 i ND l 2.29E-05 l119E+03 l 1.57E-16 l l La-140 l 3.44E+02 l1.74E+02 l 4.58E+01 i ND l ND l1.36E405 1 4.58E+05 l I l La 142 l 6.83E-01 l 3.10E-01 l 7.72E-02 l ND l ND l 6.33E+03 l 2.1lE+03 l l Ce-141 l L99E+04 l135E+04 l 1.53E+03 l ND l 6.26E+03 l 3.62E+05 l1.20E+05 l lCe-143 l 1.86E+02 l 1.3SE+02 l 1.53E+01 l ND l 6.0SE+01 l 7.98E+04 l 2.26E+05 l I l Cc-144 l 3.43E+06 l1.43E+06 l1.S4E+05 l ND l 8.48E+05 1 7.78E+06 l8.16E+05 l l Pr 143 l 9.36E+03 l 3.75E+03 l 4.64E+02 l ND l 2.16E+03 l 2.SlE+05 l 2.00E+05 l l Pr-144 l 3.01E-02 l L25E-02 l 1.53E-03 l ND l 7.05E-03 l1.02E+03 l 2.15E-08 l I l Nd-147 l5.27E+03 l 6.10E+03 l 3.65E+02 l ND l 3.56E+03 l 2.21E+05 l 1.73E+05 l l Eu 154 l 5.92E+06 l 7.2SE+05 l 5.lSE+05 l ND l 3.49E+06 l 4.67E+06 l 2.72E405-l l Hf-181 l1.41E+04 l 6.82E+04 l 6.32E+03 l ND l 1.4SE+04 l 6.85E+05 l 1.39E+05 l I l W 187 l 8.48E+00 l 7.0SE+00 l 2.48E+00 l ND l ND l 2.90E+04 l 1.55E+05 l l Np 239 l 2.30E+02 l 2.26E+01 l 1.24E+01 l ND l 7.00E+01 l 3.76E+04 l Ll9E+05 l I I I I

I APA-ZZ-01003 Rev. 3 TAllII 3.4 (Cont'd) ADULT PATHWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTIIER THAN NOllLE GASESa I Meat Patliway 2 (m mrern/,i r) per (p Ci/sec) TOTAL NUCLIDE IlONE LIVE R EODY TIIYROID NIDNEY LUNG GI-LLI l 113 l ND l 3.25E+02 l 3.25E+02 l 3.25E+02 l 3.25E+02 l 3.25E+02 l 3.25E+02 l l Ec-7 l 4.57E+03 l 1.04E404 l 5.07E*03 l ND l 1.10E+04 l ND l 1 SIE406 l l C-14 l 2.41E+0S l 4.82E+07 l 4.82E+07 l 4.82E+07 l 4.S2E+07 l 4.82E+07 l 4.82E+07 l l Na-24 l 1.36E-03 l 1.36E-03 l 1.36E-03 l 1.36E-03 l 1.36E-03 l 1.36E-03 l 1.36E-03 l l P-32 l 4.65E+09 l 2.89E+0S l 1.SOE+0S l ND l ND } ND l 5.23Et08 l l Cr-51 l ND l ND l 7.04E+03 l 4.21E+03 l 1.55E+03 l 9.34E+03 l 1.77E+06 l l Mn-54 l ND l 9.17Et06 l 1.75E+06 l ND l 2.73E+06 l ND l 2.SIE+07 l l Mn-56 l ND l 1.6SE-53 l 2.98E-54 l ND i 2.13E-53 l ND l 5.36E-52 l l Fe-55 l 2.93E+08 l 2.02E*08 l 4.72E+07 l ND l ND l 1.13E+0S l 1.16E+0Sl l Fe-59 l 2.65Et08 l 6.24E+0S l 2.39E+0S l ND l ND l 1.74E+0S l 2.0SE+09 l l Co-57 l ND l 5.63E+06 l 9.36E+06 l ND l ND l ND l 1.43E+0S l l Co-58 l ND l 1.S2E+07l 4.0SE+07 i ND l ND l ND l 3,69E+08 l l Co-60 l ND l 7.5]E+07 l 1.66E+0Sl ND J ND l ND l 1.4]E+09l l Ni-63 l 1.89E+10 l 1.3]E+09 l 6.33E+0S l ND l hT) l M l 2.73E+0S l l Ni-65 l 2.31E-52 l 3.00E 53 l 1.37E-53 l ND l ND l hT l 7.61E-52 l I l Cu-64 l ND l 2.72E-07 l 1.2SE-07 l ND l 6.86E-07 l N1 l 2.32E-05 l l Zn-65 l 3.56E+0S l 1.13E+09 l 5.1lE+0S l ND l 7.57E+0S l ND l 7.13E+0S l 1 l Zn-69 l 0.00E+00 l 0.00E+00 l 0.00E+00l ND l 0.00E+00 l ND l 0.00Et00 l l Br-S2 l ND l ND l 1.22E+03 l ND l ND l ND l 1.40E+03 l l Br-83 l ND l ND l 6.lSE-57 l ND l ND l ND l S.90E-57 l l Br 84 l ND l ND l 0.00E+00 l ND l ND l ND l 0.00E+00 l l Br-85 l ND l ND l 0.00E+00l ND l ND l ND l 0.00E+00 l i j kb-86 l ND l 4.87E*08 l 2.27E+0S l ND l ND l ND l 9.60E+07 l l l Rb-S8 l ND l 0.00E+00 l 0.00E+00 l ND l ND l N9 l 0.00E+00 l l Rb-89 l ND l 0.00E+00l 0.00E+00 l ND l ND l ND l 0.00E+00 l l l Sr-89 l 3.01E+0S l ND l 8.65E+06 l ND l ND l ND l 4.83E+07 l l Sr-90 l 1.24E+10 i ND l 3.05E+09 l ND l ND l ND l 3.59E+08 l l l Sr-91 l 1.53E-10 l ND l 6.lSE-12 l ND' l ND l ND l 7.2SE-10 l 1 Sr-92 l 1.21E-49 l ND l 5.23E 51 l .ND l ND l ND l 2.39E 48 l l lY-90 l 1.2]E+05 l ND l 3.24E+03 l ND l ND l ND l 1.2SE+09 l 4 I lI t 44-

1I APA-ZZ-01003 Ret 3 4 TAllLE 3.4 (ContNI) i:I ADULT PATIIWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTHER TilAN NOllLE GASESa LI Meat Palliway 1 I I (m mrem / r) per (pCihec) 2 3 TOTAL NUCLiDE BONE IIVER UODY TIIYROID K1DNEY LUNG GI LLI l l Y-91m l 0.00E+00 l ND l 0 00E+00 l ND l ND l ND l 0.00E+00l l Y-91 l 1.13E+06 l ND l 3.02E+04 l ND l ND l ND l 6.23E+08 l I l Y 92 l 1.55E-39 l ND i 4.52E-41 l ND j ND l ND l 2.7]E-35 l l Y-93 l 4.72E-12 j ND l 1.30E-13 [ ND l ND l ND l 1.50E-07 [ l Zr-95 l 1.87E+06l 6.00E+05 l 4.06E+05 l ND l 9.42E+05 l ND l 1.90E+09 l ') I 1 l Zr-97 l 2.07E-05 l 4.18E-06 l 1.91E-06 l ND l 6.32E-06 l ND l 1.30E+00 l l Nb-95 1 3.15E+06 l 1.75E+06 l 9.43E+05 l ND l 1.73E+06 l ND l 1.06E+10 l l Mo-99 l ND l 1.00E+05 l 1.90E+04 l ND l 2.26E+05 l ND l 2.32E+05 l l Tc-99m 1 2.87E+02 l 8.10E+02 l LO3E+04 l ND l 123E+04 l 3.97E+02 l 4.79E+05 l ' Tc-101 l 0.00E+00l 0.00E+00 l 0.00E+001 ND l 0.00E+00 l 0.00E+00l 0.00E+00 l I l Ru 103 l 1.03E+0S[ ND l 4.53E+07 l ND [ 4.0lE+0S j ND l 1.23E+10 l l Ru 105 l 5.86E-28 l ND l 2.32E-28 l ND l 7.58E-27 l ND l 3.59E-25 l ) l Ru-106 l 2.80E+09 l ND l 3.54E+0S l ND l 5.40E+09 l ND l 1.81E+11 l 1 I l Ag-110m l 6.68E+06 l 6.18E+06 l 3.67E+06 l ND l 1.2 IE+07 l ND [ 2.52E+09 l l Cd-109 l ND l 1.59E+06 l 5.55E+04 l ND 1 1.52E+06 l ND l 1.60E+07 l l Sn-ll3 l 1.37E+09] 3.8SE+07 l 7.86E+07 l 2.22E+07 l ND l ND l 4.09E+09 l l Sb 124 l 1.98E+07l 3.74E+05 l 7.84 E+06 1 4.79E+04 l ND j 1.54E+07l 5.61E+03l l Sb-12.. l 1.91E+07l 2.13E+05 l 4.54E+06 l 1.94E+04 l ND l 1.47E+07 l 2.10E+08 l l Te-125m l 3.59E+0S [ 1.30E+08[ 4.80E+07 l 1.0SE+08 l L46E+09l ND l 1.43E+09 i I l Tc 127m l 1.1lE+09 l 3.98E+0S l 1.36E+08 l 2.85E+0S l 4.53E+09 l ND l 3.74E+09 l l Te-127 l 2.14E 10 l 7.67E-11 l 4.62E-11 l 1.5SE-10l 8.70E-10 [ ND l 1.69E-08 l l Te-129m l 1.13E+09 l 4.23E+0S l L79E+0Sl 3.89E+08 l 4.73E+09 l ND l 5.71E409 l l Te-129 l 0.00E+00l 0.00E+00 l 0.00E+00l 0.00E+00l 0.00E+00 l ND l 0.00E400l l Te-131m l 4.52E+02 l 2.21E+02 l 1.84E+02 l 3.50E+02 l 2.24E+03 l ND l 2.19E+04 l ] l Te 131 l 0.00E+00j 0.00E+00 l 0.00E*00l 0.00E4001 0.00E400 l ND l 0.00E+00 l l Te-132 l 1.42E+06 l 9.18E+05 l 8.62E+05 l 1.01E+06 l 8.84E+06 l ND l.4 34E+07 l l l-130 [ 2.12E-06 l 6.27E-06 l 2.47E-06 l 5.31E-04 l 9.78E-06 l ND l 5.40E-06 l l l-131 l 1.0SE+07 l 1.54E+07 l 8 82E+06 l 5.04E+09 l 2.64E+07 l ND l 4.06E+06 l - I l l-132 l 7.20E-59 l 1.93E-58 l 6.74E-59 l 6.74E-57 l 3.07E 58 l ND l 3.62E 59 l lI133 l 3.67E-01 l 6.39E-01 l 1.95E-01 l 9.38E+01 l

1. ) I E+00 l ND

[ 5.74E-01 l ll-134 l 0.00E+00 l 0.00E+001 0.00E+00 l 000E+00l 0.00E+00 l ND l 0.00E+00 l I -4 5 -

APA-ZZ-01003 Ret 3 a TAllLE 3.4 (Cont'd) ADULT PATHWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTIIER THAN NOIlLE GASES" I Meat Patliway (m mremly r) per (pCihec) 2 TOTAL NUCLIDE BONE LI\\TR HODY T11YROID K1DNEY LUNG GI-LL1 l l-135 l 4.4SE-17 l 1.17E-16 l 4.33E-17 l 7.74E-15 l 1.SSE-16 l ND l 1.33E-16 l l Cs-134 l 6.57E+08 l 1.56E+09 l 1.2SE+09 l ND l 5.06E+0S l 1.6SE F08 l 2.74E+07 l I l Cs-136 l 1.20E+07 l 4.76E+07 l 3.42E407 i ND l 2.65E4 07 [ 3.63E+06 j 5.40E+06 l l Cs-137 l 8.71E+0S l 1.19E+09 l 7.K l E + 08 l ND l 4.04E+t,0 l 1.34E+0Sl 2.3 ] E+07 l l Cs-138 l 0.ME400l 0.00E+00 1 0.00E+00 l ND l 0.00E+00 l 0.00E+00 l 0.00E+00 l l Ba-130 1 0.00E+00 l 0.00E+00l 0.00E+00 i ND l 0.00E+00 l 000E+00l 0.00E+00l l Ba-140 l 2 87E+07 l 3.61E+04 l 1.88E+06 l ND l 1.23E+04 l 2.07E+04 l 5.91E+07l l Ba-141 l 0.00E+0i 1 0.00E+00 l 0.00E+00 l ND i 0.00E+03l 0.00E+00 l 0.00E+00 l l Ba 142 l 000E+00l 0.00E+00l 0.00E+00 l ND l 0.00E+00l 0.00E+00l 0.00E+00 l l La-140 l 2.2 IE+02 l 1.1lE+02l 2.94E+01 l ND l ND l ND l 8,18E+06 l l La-142 l 3.60E-92 l 1.64E-92 l 4.0SE-93 l ND l ND l ND l 1.20E-88 l l Ce-141 l 1.40E+04 l 9.49E+03 1 1.08E+03l ND l 4.41E+03 i ND l 3.63E+07 l l Cc-143 l 2.01E-02 l 1.49E+01l 1.64E-03 l ND l 6.54E-03 l ND l 5.55E+021 1 I l Cc-144 l 1.46E+06 l 6.09E+05 l 7.82E+04 l ND l 3.61E+05 l ND l 4.02E+08 l l Fr-143 l 2.10E+04 l S.40E+03 l 1.04E+03 l ND l 4.S5E+03 l ND l 9.18E+07 l l l Pr-144 l 3.52E+02 l 1 A6E+02 l 1.79E+01 l ND l S.24E+0! l ND l 5.06E-05 1 I' i l Nd-147 l 7.07E-03 l S.17E+03l 4.89E+02 I ND l 4.77E+03 l ND l 3.92E+07 l l Eu-154 l 8.02E406 l 9.86E+05 l 7.01E+05 l ND. I 4.72E+06 l ND l 7.14E+08 l 1 l Hf-181 l 3.01E+06 l 1.46E+07 l 1.35Et06 l ND l 3.14E+06 l 'ND l 1.66E+10 l J I I W-187 l 2.07E-02 l 1.73E-02 l 6.03E-03 l ND l ND l ND l 5.67E+00 l l Np-239 l 2.63E-01 l 2.59E-02 l 1.43E-02 [ ND 1 8.07E-02 l ND l 5.30E+03 l 1 1 I -4 6-I

APA-ZZ-01003 4 net 3 TA RI.E 3.4 (ContNI) ADULT PATilWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER TilAN l 4 NOBLE GASESa Grass-Cow-Milk Pathway (m mrem /)r) per(pCihec) 2 TOTAL r NUCLIDE BONE LIVER HODY TIIYMOID KIDNEY L11NG GI.L l l H-3 l ND l 7.63E+02 l 7.63E+02 l 7.63E+02 l 7.63E+02 l 7.63E+02 l 7.63E i g l Be-7 l 1.63E+03 l 3.72E+03 l 1.81E+03l ND l 3.93E+03 l ND l 6.45E jE l C-14 l 2.63E+08 [ 5.27E+07 l 5.27E-07 1 5.27E+07l 5.27E+07 ) 5.27E+07l 5.27F r l l Na-24 l 2.44E+06 l 2.44E+06 l 2.44E+06 l 2.44E406 l 2.44E+06 l 2.44E+06 l 2.44E l P-32 l 1.71E+10 l 1.06E+09 l 6.6]E+0S l ND l ND l ND l 1.92E f EN l Cr-51 l ND l ND l 2.86E+04 l 1.71E+04 l 6.30E+03 l 3.79E+04 1 7.19E j l Mn-54 l ND l 8.42E+06 l 1.61E+06 l ND l 2.50E406 l ND l 2.5SE ' l g l Mn-56 l ND l 4.20E-03 l 7A5E-04 l ND l 5.33E-03 l ND l 1.34E j g l Fe 55 l 2.51E+07 l 1.74E+07 l 4.05E+06 l ND l ND l 9.68E+06 [ 9.96E ' I [ Fe-59 l 2.97E+07 l 6.98E+07 l 2.68E+07 l ND l ND l 1.95E+07 l 2.33E l Co-57 l ND l 1.28E+06 l 2.13E+06 l ND l ND l ND . I 3.25E l Co-58 l ND l 4.72E+06 l 1.06E+07 l ND l ND l ND l 9.56E I Co-60 i ND l 1.64E+07 l 3.62E+07 i ND l ND l ND l 3.0SE I L l Ni-63 l 6.73E+09 l 4.67E+08 l 2.26E+08 l ND l ND l ND l 9.73E l Ni-65 l 3.71E-01 l 4.82E-02 l 2.20E-02 l ND (_ ND l-ND l 1.22E r l Cu-64 l ND l 2.39E+04 l 1.12E+04 l ND l 6.01E+04 l ND l 2.03E I-l Zn-65 l 1.37E+09 l 4.37E+09 l 1.97E+09l ND l 2.92E+09 l ND l 2.75E l Zn-69 l 2.llE-12 1 4.03E-12 l 2.80E 13 l ND l 2.62E-12 I ND l 6.06F l Br-82 1 ND l ND l 3.23E+07 l ND l ND l ND l 3.71E l Br-83 l ND l ND l 9.75E-02 l ND l ND l ND l 1.40E l Br-84 l ND l ND l 1.63E 23 l ND l ND l ND l 1.28f l Br 85 l ND l ND l 0.00E+00 l ND l ND l ND l 0.00E I l Rb-86 l ND l 2.60E+09 l 1.2]E+09 l ND l ND l ND l 5.12E l Rb-88 l ND J 2.19E-45 l 1.16E-45 l ND l ND l ND l 3 03I l Rb-89 l ND l 4.4 5E-53 1 3.13E-53 l ND l ND l ND l 2.59E l Sr-89 l 145E409 l ND l 4.17E+07 l ND l ND l ND l 2.33E l Sr 90 l 4.68E+10 l ND l 1.15E+10 l ND l ND l ND l 1.35E l Sr-91 l 2.90E+04 l ND l 1.17E+03 l ND l ND l ND l 1.38E I l Sr-92 1 4.90E-Ol l ND l 2.12E-02 l ND l ND l ND l 9.71E - l Y-90 l 7.43E+02 l ND l 1.99E+0! l ND l _ ND l ND l 7.87E-I i -4 7-I

APA-ZZ-010(G Rev. 3 TA RLE 3.4 (Cont'<!) ADULT PATIIWAY DOSE FACTORS (Rj) FOR RADIONUCLIDES OTIIER TIIAN NOBLE GASESa Grass-Goat-Milk Pathway (m' m rem /yr) per (p Cihec) TOTAL I, NUCLIDE BONE LWER BODY THYROID KIDNEY LUNG GI-LLI ! H-3 l ND l 1.56E+03l L56E+03 l L56E+03 l 1.56E+03 l 1.56E+03 l 1.56E+03 l I l Be-7 l 1.96E+0 l 4.47E*02 l 2.17E+02 l ND l 4.72E+02 l ND l 7.74E+04 l l C-s4 l 2.64E+0S l 5.27E+07 l 5.27E-07 l 5.27E407 l 5.27E+07l 5.27E+07 l 5.27E+07 l l Na-24 l 2.93E+05 1 2.93E-05 l 2.93E405 l 2.93E+05 l 2.93E405 l 2.93E+05 l 2.93E+05 l l P-32 l 2.05E+10 [ 1.28E+09 l 7.94E+0S l ND l ND l ND l 2.3]E+09 l l Cr-51 l ND l ND l 3 A3E+03 l 2 05E+03 l 7 56E+02 l 4.56E-03 l 8.63E+05l l Mn-54 l ND l 1.0lE+06 i 1.93E+05l ND l 3.01E+05 l ND l 3.10E+06 l I l Mn-56 l ND l 5.04E-04 l 8.94E-05 l ND l 6.40E-04 l ND l 1.61E-02 l lFe-55 l 3.27E+05 l 2.26E+05 l 5.26Etu4 l ND l ND l 1.26E+05 l 1.30E+05 l l Fe-59 l 3 87E+05 l 9.0SE+05 l 3ASE405 l ND l ND l 2.54E+05 l 3.03E+06 l l Co-57 l ND l 1.54E+05 l 2.56E+05 l ND l ND l ND l 3.90E+06 l l Co-58 l ND l 5.66E+05 l 1.27E+06 l ND l ND l ND l 1.15E+07 l l Co40 l ND l 1.97E+06 l 4.35E+06 l ND l ND l ND l 3.70E+07 l l Ni43 l S.0SE+0S l 5.60E+07 l 2.7]E'07 l ND l ND l ND l 1.17E+07 j l Ni-65 l 4.46E-02 l 5.79E-03 l 2.64E-03 l ND l ND l ND l 1.47E-01 l I-l Cu-64 l ND l 2.66E+03 l 1.25E+03 l ND l 6.71E+03 l ND l 2.27E+05 l l Zn-65 l L65Ev08 1 5.24E+08 l 2.37E+0S l ND l 3.51E+0S l ND l 3.30E +0 S l l Zn-69 l 2.53E-13 l 4.84E-13 l 3.37E-14 l ND l 3.15E-13 l ND l 7.28E-14 l l Br-82 l ND l ND l 3.88E+06 l ND l ND l ND l 4.45E+06 l I l Br-83 l ND l ND j Ll7E-02 l ND l ND l ND l 1.69E-02 l l Br-84 l ND l ND l 1.96E-24 l ND [ ND ND l L54E-29 l ln l Br-85 l ND l ND l 0.00E+00i ND l ND l ND l 0.00E+00 l l Rb-86 ND l 3.12E+0S l 1.45E+0S l ND l ND l ND l 6.15E+07 l l Rb SS l ND l 2.63E-46 l 1.40E-46 l ND l ND j ND l 3.64E-57 l l Rb-89 l ND l 5.35E-54 1 3.76E-54 l ND l ND l ND j 3.1lE-67 l l Sr-89 l 3.05E+09 l ND l 8.75E+07 l ND j ND l ND l. 4.89E+08 l l Sr-90 l 9.84E+10 l ND l 2.41E+10 l ND l ND l ND l 2.84E+09 l 'l lSt-91 l 6.09E+04 l ND l 2.46E+03 l ND l ND l ND l 2.90E+05 l i 5 lSr-92 l 1.03E+00 l ND l 4.45E-02 l ND l ND l ND l 2.04E+01 l j Y-90 l 8.92E+01 l ND l 2.39E+00 l ND l ND l ND l 9.46E+05 l I I I I

APA-ZZ-01003 Rev. 3 TABLE 3.4 (Conl51) ADULT PATIIWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN I 2 l NOBLE CASES 3 Grass-Guat-Milk Patinvay = (m mn y r) per (p Cihec) 2 TOTAL NUCLIDE BONE LIVER BODY TIIYRO1D N1DNEY LUNG GI LLI l Y-91 m l 7.25E-21 l ND l 2.SIE-22 i ND l ND l ND l 2.13E-20 l l l Y-91 l LO3E+03 i ND l 2.76E+01 l ND l ND l ND l 5.683405 l } l Y-92 l 6.72E-06 l ND l 1.96E-07 l ND l ND l ND 1 1.]SE-01 l l Y-93 l 2 69E-02 l ND l 7.42E-04 l ND l ND l ND l 8.52E+02 l l Zr-95 l 1.13E+02 l 3.63E+01 l 2.46E+0! l ND l 5.70E+01 1 ND j 1.15E+05 l - l Zr-97 l 5 21E-02 l 1.05E-02 l 4.80E-03 l ND l 159E-02 l ND l 3.25E+03 l l Nb-95 l

1. l 'iE+04 1 6.45E+03 l 3.47E+03 l ND l 6.37E+03 i ND l 3.91E+07 l I

} Mo-99 l ND l 2.98E+06 l 5.66E+05 l ND l 6.74E+06 l ND l 6.90E+06 [ l Tc-99m l 5.69E+02 l 1.61E+03 l 2.05E+04 l ND l 2.44E+04 l 7.87E+02 l 9.51E+05 l l Tc-101 1 3 21E-61 l 4.62E-61 1 4.54E-60 l ND 1 S.33E-60 l 2.36E-61 l L39E-72 l l Ru-103 l 1.22E+02 l ND l 5.27E+01 I ND l 4.67E+02 l ND } 1.43E+04 l l Ru-105 l LO3E-04 l ND l 4.07E-05 l ND l 1.33E-03 l ND l 6.31E-02 l l Ru-106 l 2.45E+03 i ND l 3.10E+02 l ND l 4.73E+03 l ND l 1.59E+05 l I l A;;-)10m l 6.99E*06 l 6 47E+06 l 3.84E+06 l ND l 1.27E+07 l ND l 2.64E+09 l l Cd-109 l ND l 1.36E+05 l 4.74E+03 l ND l 1.30E+05 l ND l 1.37E+06 l lSn-113 l 1.61E*07 l 4.58E-05 l 9.2SE+05 l 2.62E+05 l ND l ND l 4.83E-07 l 1 Sb-124 l 3.09E+06 l 5.84E+04 l 1.23E+06 l 7.50E+03 l ND l 2.41E-06 l 8.78E+07 l l Sb-125 l 2.46E+06 l 2.74E+04 l 5.84E+05 l 2.50E+03 l ND l 1.89E+06 l 2.70E+07 l l Te-125m l 1.96E+06l 7.09E+05 l 2.62E+05 l 5.89E+05 l 7.96E+06 l ND l 7.81E+06 l I l Te-127m l 5.50E+06 l 1.97E+06 l 6.70E+05 l 1.4]E+06 l 2.23E+07 l ND l 1.84E+07 l j Te-127 l 7.85E+01 l 2.82E+01 l 1.70E+01 l 5.82E+01l 3.20E+02 l ND l 6.19E+03 l I l Te-129m l 7.23E+06 l 2.70E+06 l 1.14E+06 l 2.4SE+06 l 3.02E+07 l ND l 3.64E+07 l l Tc 129 l 3.41E-11 l 1.2SE-11 l 8.32E-12 l 2.62E-11 l 1.43E-10 l ND l 2.5SE-11 l 1 Te-131m l 4.34E+04 l 2.12E+04 l 1.77E404 l 3.36E+04 l 2.15E+05 i ND l 2.llE+06 l l l Te-131 l 4.40E-34 1 4.84E-34 l 1.39E-34 l 3,62E-34 l 1.93E-33 l ND l 6.24E-35 l l Te-132 l 2.89E+05 l 1.87E+05 l 1.75E+05 l 2.06E+05 l 1.80E+06[ ND l 8.83E+06 l l1130 l 505E+05 l 1.49E+06 l 5.88E+05l 1.26E+0S l 2.32E+06 l ND 1 1.2SE+06 l I l l-131 l 3.56E+0S l 5.09E+08 l 2.92E+08 l 1.67E+11 l 8.72E+0S l ND l 1.34E+0S l l l-132 l 1.98E-01 l 5.29E-01 l 1.85E-01 l 1.85E+0 ) l E.43E-01 l ND l 9.95E-02_ l lI-133 1 4.65E+06 l 8.09E+06 l 2.47E+06 l 1.19E+09 l 1.41E+07 l ND l 7.27E+06 l ll-134 l 2.14E-12 l 6.64E-12 l 2.37E-12 l 1.15E-10 l 1.06E-11 ] ND l 5.78E-15 l 1 5,.

'I-- APA ZZ-01003 ner. 3 j TAHI E 3.4 (Cont'il) l ADULT PATIIWAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTiiER TilAN NOllLE GASESa Grass-Goat-Milk Pathway (m mrem / r) per (pCihec) 2 3 i' TOTAL NUCLIDE HONE LIVER BODY T!!YROID KIDNEY LUNG GI-LLI { l l-135 l 1.54E+04 l 4.04E404 l L49E+04 l 2.67E+06 l 6.4SE+04 l ND l 4.57E+04 l 1 Cs-134 l 1.70E+ 10 l 4.04E+10 l 3.30E+ 10 1 ND l 1.31E+10 l 4.34E+09 [ 7.07E+0S l i l Cs-136 l 7.91E4 0S l 3.12E+09 l 2.25E+09 l ND 1 1.74E+09 l 2.3SE+08 l 3.55E408 j 1.- l Cs-137 l 2.22E+10 l 3.03E+ 10 l 1.99E+10 l ND l 1.03E+10 l 3.42E+09 l 5.87E+08 l l Cs-138 l 2.75E-23 l 5.43E-23 l 2.69E-23 l ND l 3.99E-23 l 3.94E-24 l 2.32E-28 l 1 J Ba-139 l 5.34E-09 l 3.80E-12 l 1.56E-10 l ND l 3.55E.12 l 2.16E-12 l 9.46E-09 l l Ba-140 l 3.23E+06 l 4.06E+03 [ 2.12E+05 i ND l L3SE+03 l 2.32E403 l 6.65E+06 i lI l Ba-141 l 5.03E-47 l 3.80E-50 l 1.70E-48 l ND l 3.54E-50 l 2.16E-50 l 2.37E-5C l l Ba-142 l 3.32E-S! l 3.42E-84 l 2.09E-82 l ND l 2.89E-84 l 1.93E-84 l 4.68E-99 l l La-140 l 4.97E+00 l 2.51E+00 l 6.62E-01 l ND l ND l ND l 1.84E+05 l l La 142 l 2.24E-12 l 1.02E 12 l 2.54E-13 l ND l ND l ND l 7.45E-09 l l l Cc-141 l 5.82E+02l 3.94E+02 l 4.46E+01 l ND l 1.83E+02 l ND l 1.50E+06 l l Cc-143 J 4.99E+00 l 3.69E+03 l 4.09E-01 l ND l 1.63E+00 l ND l 1.38E+05 l l Ce-144 l 4.30E+04 l 1.80E+04 l 2.31E+03 l ND 1 1.07E @4 l ND l 1.45E+07l l Pr-143 l 1.90E+01l 7.6]E+00l 9.40E-01 l ND l 4.39E+00 l ND 1 8.31E+04 l g l Pr-144 j 1.33E-Ol l 5.50E-02 l 0.74E-03 l ND l 3.10E-02 l ND l 1.91E-OS ; W l Nd-147 l 1.13E+01 l 1.3]E+Cl l 7.82E-01 l ND l 7.64E+00 l ND l 6.28E+04 l l l Eu 154 l 2.84E+03 l 3.49E+02 l 2.49E+02 l ND l 1.67E+03 l ND l 2.53E+05 l l Rf-181 l 1.71E+01 l 8.3]E+01 l 7.70E+00 l ND l 1.79E+01 i ND l 9.46E+04 l I l W-]S7 l 7.83E+02 l 6.54E+02 l 2.29E+02 1 ND l ND l ND l 2.14E+05 l l Np 239 l 4.43E-01 l 4.35E-02 l 2.40E-02 l ND l 1.36E-01 l ND l 8.93E+03 l I i I I I ,I I - -

APA-ZZ-01003 Rev. 3 TAltl,E 3.4 (Cont'<l) ADULT PATI 1WAY DOSE FACTORS (R;) FOR RADIONUCLIDES OTIIER TilAN NOBLE GASESa Vegetation Palliway (m mrem /)r) per (pCihec) 2 TOTAL NUCLIDE !!ONE LIVER !!O D Y THYROID KIDNTY LUNG G1LLI l R3 l ND l 2.26E403 l 2.26E403 l 2.26E+03 l 2.26E+03 l 2.26E+03 l 2.26E+03 l j l Be 7 l 9.24E+04 l 21lE+05 l 1.03E+05l ND l 2.23E+05 l ND l 3.66E+07 l l l C-14 l 2.2SE40S l 4.55E+07 l 4.55E+07 l 4.55E+07 l 4.55E+07 l. 55E+07 l 4.55E407 l i l Na-24 l 2.69E+05 l 2.69E+05 l 2.69E+05 l 2.69E+05 l 2.69E+05 l 2.69E+05 l 2 69E+05 l l P-32 1 1.40E+09 l 8.74E+07 l 5.43E+07 i ND l ND l ND l 1.5SE+0S l 4i l 4 um l Cr-51 l ND l ND l 4.64E+04 j 2.78E+04 l 1.02E+04 l 6.16E+04 l 1.17E+07 l l Mn-54 l ND I 3.13E+0S l 5.97E+07 i ND l 9.3]E+07 l ND l 9.59E+08 l l l Mn-56 l ND l 1.60E+01 l 2.84E+00 l ND l 2.03E+0! l ND l 5.10E+02 l i Fe-55 l 2.10E+0S l 1.45E+0S I 3.3SE+07 l ND l ND l S.0SE+07 l S.31E+07 l l l Fe-59 l 1.26E+08 l 2.96E+0S l 1.14E+0S l ND l ND l 8.2SE+07l 9.8SE+0S l

l

[ Co-57 l ND l 1.17E+07 l 1.95E+07 l ND l ND l ND l 2.97E+08 l l= l Co-58 1 h"D l 3.07E+07 l 6.89E407 l ND l ND l ND l 6.23E+0S l i l Co-60 l ND l 1.67E+08 l 3.69E+0S l ND l ND l ND l 3.14E+09 l iE l Ni-63 l 1.04E+10 l 7.2]E+0S l 3.49E+0S l ND l ND l ND l 1.50E+0S l 1

g i Ni-65 l 6.16E+01 l 8.00E+00 l 3.65E+00 l ND l

ND l ND l 2.03E+02 l } } l Co-64 l ND l 9.20E+03 l 4.32E+03 l ND l 2.32E+04 l ND l 7.84E+05 l l Zn-65 l 3.17E408 l 1.01E+09 l 4.56E+0S l ND l 6.75E+0S l ND l 6.36E+0S l l Zn-69 l 5.52E.06 l 1.05E-05 l 7.34E-07 i ND l 6.S5E-06 l ND l 1.59E-06 l l Br-82 i ND l ND 1 1.50E+06 l ND l ND l ND l 1.72E+06l l Br-83 l ND l ND { 3.1lE+00 l ND l ND l ND l 4.4SE+00 l i ! Br-84 l ND l ND l 2.49E 11 l ND l ND l ND l 1.96E-16 l +g l Br-85 l ND l ND l 0.00E+00l ND l ND l ND l 0.00E+00 l lg 4 l Rb 86 i ND l 2.19E+0S l 1.02E+0S I ND l ND l ND l 4.33E+07 l l Rb-88 i ND l 3.47E-22 l 1.S4E-22 [ ND l ND l ND l 4.79E 33 l j-l Rb-89 l ND l 3.94E 26 l 2.77E-26 l ND l ND l ND. I 2.29E-39 l o 1 Sr-89 { 9.97E+09 l ND l 2.86E+0S l ND l ND l ND l 1.60E409 l l Sr-90 l 6.0$E+1! l ND l 1.4SE+11 l ND l ND l ND l 1.75E+10 l jg l5r91 l 3.05E+05 l ND l 1.23E+04l ND l ND l ND l 1.45E+06 l lE i sr 92 i 4.27E+02i ND i i.s5E+0i i ND i ND i ND i ~ 8.46E+03i l Y-90 l 7.67E+05 l ND l 2.06E+04 l ND l ND l ND l 8.14E+09 l 't !I i lI 4 }i .= n - s

{ )'lN i APA-ZZ-01003 Ret 3 1 TAllLE 3.4 (Cont'd) ADULT PATIIWAY DOSE FACTORS (Rj) FOR RADIONUCLIDES OTIIER TIIAN 4 NOIILE GASESn i' Vegetation Pathway L (m' mremlyr) per (pCi/sec) TOTAL NUCLIDE DONE LIVER ILODY TilVROID NIDNEY LUNG GI.ll! l Y-91m j 5.24E-09 l ND l 2.03E-10 l ND l ND l ND l 1.54E-08 l l l Y 91 l 51IE+06 l ND l 1.37E+05l ND l ND l ND l 2.81E+09 l

. 3 l Y-92 l

9.16E-01 i ND l 2.6SE-02 l ND l ND l ND l 1.60E+04 i J Y-93 l L70E+02 } ND l 4.GSE-00 J ND l ND l ND } 5.38E+06l j l Zr-95 l 1.17E+06l 3.77E+05 l 2.55E405 l ND l 5.91E+05 l ND l 1.19E+09 l ~ l Zr-97 l 3.37E402 l 6SIE+0!! 3.1lE401 l ND l 1.03E+02l ND j 2.1lE+07 l I l Nb-95 l 2.40E+05 l L34E+05 l 7.19E+04 l ND l 1.32E+05 I ND l 8.1]E+0SI ll l Mo-99 l 0.00E+00l 6.15E+06 l 1.17E+06 l ND l 1.39E+07 l ND l 1.43E4 07 l l 5 l Tc-99m l 3.53E+02 l 9.96E+02 l L27E+04 l ND l 1.51E+04 [ 4.83E+02 l 5.90E+05 l - i l Tc.101 l 8.34E-31 j L20E-30 l L18E-29 l ND l 2J6E-29 l 6.14E-31 1 3.61E-42 l l _ l Ru-103 l 4.77E406 l ND l 2.06E+06 l ND l 1.82E+07 l ND l 5.57E+08l l Ru-105 l 5.39E+01l ND l 2.13E+01 l ND l 6.97E+02 l ND j 3.30E+04 l l l Ru-106 l 1.93E+0S l ND l 2.44E+07 l ND l 3.72E+0S l ND l 1.25E+10 l l l Ag 110m l 1.05E+07l 9.75E46 l 5.79Evo6l ND l L92E+07 l ND l 3.98Ev09 l 1 l 15 l Cd-109 l 0.00E+00 l 8.36E+07 l 2.92E+06 l ND l 8,00E+07 l ND l 8.43E+0S j j l Sn-113 l 4.16E408 l 1.lSE+07 l 2.40E+07 l 6.75E+06 l ND l ND l 1.25E+09 l [ Sb-l'4 l 1.04E+08 l 1.96E+06 l 4.1lE+07 l 2.51E+05 l ND l 8.07E+07 l 2.94E+09 l l l Sb-125 1 1.37E+0S l 1.53E+06 1 3.25E-07 l 1.39E+05 l ND l 1.05E+08 l 1.50E+09 l l l Te-125m l 9.66E+07 j 3.50E+07 l 1.29E+07 l 2.90E+07 l 3.93E+08 l ND [ 3.86E+08 l i l Te-127m l 3.49E+08 l 1.25E+0S l 4.26E407 l 8.92E+07 l 1.42E+09 l ND l 1.17E+09 1 i lTe-127 l 5.66E+03 l 2.03E+03 l 1.23E+03 l 4.20E+03 l 2.31E+04 l ND l 4.47E+05 l !!g l Te-129m l 2.51E+08 l 9.38E+07 l 3.98E+07 l 8.64E+07 l LO5E+09 l ND l 1.27E+09 l l3 l Te-129 l 7.65E-04 l 2.87E-04 l 1.86E-04 l 5.87E-04 l 3.22E-03 l ND l 5.77E-04 l f l Te-131m l 9.12E+05 l 4.46E+05 l 3.72E+05 l 7.07E+05 l 4.52E+06 l ND l 4.43E+07 l - l l Te-131 l 1.51E-15 l 6.32E-16 l 4.78E-16 l L24E-15 l 6.63E-15 l ND l 2.14E-16 l l Te-132 l 4.30E+06 l 2.78E+06 l 2.61E+06 1 3.07E+06 l 2.68E4 07 l ND l L32E+08 l l i-130 l 3.93E+05 l 1.16E+06l 4.57E+05 l 9.81E+07 l L81E+06l ND l 9.97E+05 l I l l-131. l 8.08E+07 l

1. '6E+08 l 6.62E+07 l 3.79E+10 l 1.98E+08l ND l 3.05E+07 l.

l1-132 l 5.77E+01 l 1.54E+021 5.40E+01 l 5.40E+03 l 2.46E+02 l ND l 2.90E+01 l l l-133 l 2.09E+06 l 3.63E+06 l 1.1lE+06 l 5.33E+0Sl 6.33E+06 [ ND l 3.26E+06 l t l l-134 1 9.69E-05 1 2.63E-04 l 9.42E-05 l 4.56E-03 l 4.19E-04 l ND l 2.30E-07 l 1 I -54 8

l APA-ZZ-01003 m Rev. 3 l TA BLE 3.4 (Cont'd) ADULT PATilWAY DOSE FACTORS (R ) FOR RADIONUCLIDES OTHER THAN l NOBLE CASESa Vegetation Patliway (m' m remly r) per (p Ci/sec) TOTAL i NUCLIDE llONE LIVER IlODY T il V R O ID KIDNEY 1.UNG GI-LLI l l-135 l 3.90E+04 l 1.02E+05 l 3.77E+04 l 6.74E406 l L64E+05 l ND l 1.15E+05 l I l Cs 134 l 4.67E+09 l 1.1 l E+ 10 l 9.0SE+09 l ND l 3.59E+09 l 1.19E+09 1 1.94E+08 l l Cs 136 l 4.27E+07 l 1.69E+0Sl 1.21E+0Sj ND l 9.3SE*07 l 1.29E+07 l 1.9]E+07 l l Cs 137 l 6.36E+09 l 8.70E+o9 l 5.70E+09 l ND l 2.95E+09 l 9.81E*08 l 1.6SE+08 l l Cs-138 1 3.94E-Il l 7.78E-Il l 3 86E-11 l ND l 5.72E-Il l 5.65E-12 l 3.32E 16 l l Ba-139 l 2.S6E-02 l 2.04E-05 i S 39E-04 l ND l 1.9]E-05 l. 1.16E-05 l 5.0SE-02 l l Ba-140 l 1.29E+08 l 1/.lE+05 l S.42E+06l ND l 5.49E+04 l 9.24E+04 l 2.65E+08 l I l Ba-141 l 1.16E-21 l 8 81E-25 l 3.93E 23 ) ND l 8.19E 25 l 5.00E-25 l 5.49E-31 l l Ba-142 l 6.09E-39 l 6.26E-42 l 3.83E-40 l ND l 5.29E-42 l 3.55E-42 l 8.SSE-57 l l La-140 l 1.58E+04 l 7.98E+03 l 2.1 J E+03 l ND l ND l ND l 5.86E+0S l l La-142 l 2.03E-04 ] 9.2]E-05 l 2.29E-05 l ND l ND I ND l 6.72E-01 l l Ce-141 l 1.97E+05 l 1.33E+05 l 1.51E+04 l ND l 6.19E+04 l ND l 5.10Eo0S l l Cc 143 l 9.98E+02 l 7.3SE+05 l 8.17E+01J ND l 3.25E+02 l ND l 2.76E+07 l I l Cc-144 l 3.29E+07 l 1.3SE407 l 1.77E+06l ND l 8,16E+06 l ND l 1.1lE+10 l l Pr-143 l 6.26E+04 l 2.51E+04 l 3.10E+03 i ND l 1.45E+04 l ND 1 2.74E+08 l l Pr-144 l 2.03E+03 l 8.43E+02 l 103E+02l ND l 4.75E+02 l ND l 2.92E-04 l I l Nd 147 [ 3.33E+04 l 3.85E+04 l 2.31E+03 l ND l 2.25E+04 j ND l 1.85E+0S l l Eu 154 l 4.85E+07 l 5.97E+06 l 4.25E+06 l ND l 2.86E+07 [ ND j 4.32E+09 l l Hf 181 l 1.40E+05 l 6 82E+05 l 6.32E+04 l ND l 1.47E+05 l ND l 7.76E+0S l I l W-187 l 3.80E+04 l 3.18E+04 l 1.11E+04 l ND l ND l ND l 1.04E+07 l l Np-239 l 1.43E+03 l 1.41E+02 l 7.76E+01 l ND l 4.39E+02 l ND l 2.89E+07 l I I-I E I-I 55-I

I APA 7101003 Rev. 3 I 4.0 DOSE AND DOSE COMMITMENT FROM URANIUM FUEL CYCLE SOURCES 4.1 CALCULATION OF DOFE AND DOSE COMhji FMFNT FROM URANIUM FUEL CYCLE m EQFRCES The annual dose or dose commitment to a MEMBER OF TiiE PUBLIC for Uranium Fuel Cycle Sources is deternuned as: l I Dose to the total body and internal organs due to gamma ray exposure from submersion a. in a cloud of radioactive noble gases, ground planc exposure, and direct radiation from the Unit and outside storage tanks; I b. Dose to skin due to beta radiadon from submersion in - cloud of radioactise noble, ses, and ground planc er,esure; Thyroid dose due to inha!adon and ingestion of radiciodines; and c. d. Organ dose due to inhalation and ingestion of radioactive material. It is assumed that total body dose from sources of gamma radiation irradiates internal body organs at the same numerical rate. (Ref. I1.12.5) I The cose from gaseous efnuents is considered to be the summation of the dose at the individual's ressdence and the dose to the individual from actisities within the SITE BOUbDARY. Since the doses via liquid releases are very conservatively evaluated, there is reasonable I assurance that no real individual wi)) 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. I1.12.3). l There are no other Uranium Fuct Cycle Sources within 8km of the Callaway Plant. 1.1.1 Identification of the MEMBER OF THE PUBilC 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 I recreational or other purposes not associated with the plant (Ref.11.4 and 11.8.10). Accordingly, it is necessary to characterize this indisidual with respect to his utilization of areas both within and at or bevond 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. I1.12.3 and 11.12.4). The evaluation of Total Dose from the Uranium Fuel Cycle should consider the dose to two

I i

Critical Receptors: a) The Nearest Resident, and b) The Critical Receptor within the SITE BOUNDARY. i 4.1.2 Total Dose to the Nearest Residan.1 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) exis:s at the location of the Nearest Resident. 1

I

I

APA-Z2-01003 Rev. 3 lt is assumed that direct radiation dose from operation of the Unit and outside storage tanks, and .g dose from gaseous emuents due to sethitics within the SITE BOUNDARY, is negligible for the 3 Nearest Resident. The total Dose from the Uranium Fuel Cycle to the Nearest Resident is j 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. i I 1 The location of the Ncarest Resident in ea:h meteorological sector is determined from the Annual Land Use Census conducted in accordance with the Requirements of REC 9.12.1.1. I 1 4.1.3 Total Dose to the Critical Regator Within the SITE BOUND ARY The Union Electric Company has entered into an agreement with the State of Missouri I Department of Conservadon for managemerd of the residuallands surrounding the Callaway Plant, includmg some areas within the SITE BOUNDARY. Under the terms of this agreement, i certain areas have been opened to the public for low intensity recreational uses (hunting, idking, I 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 Depanment of Conservation to provide income to support management and I development costs. Activities conducted under these leases are primarily comprised of farming (animal fecd), grazing, and forestry (Ref. I1.7.2,11.7.3,11.13, and 11.13.1). Based on the utilization of areas within the SITE BOUNDARY, it is reasonab!c to assume that I the critical receptor within the SITE BOUNDARY is a famier, and that his dose from activides within the SrE BOUNDARY is due to exposure incurred while conducting his farming activities. The current tenant has estimated that he spends approximately 1100 hours per year I working in Ods area (Ref. I1.5.5). Occupancy of areas within the SITE BOUNDARY is assumed to be averaged over a yriod of one year. Any reevaluation of assumpdons 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 efDuents is calculated for the adult age poup using the methodology discussed in Section 3, utilizing concurrent meteorological data at the farmer's I residence and historical meteorological data from Table 6.1 for activities within the SITE BOUND ARY. 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 lave the total occupancy of 1100 hours / year factored into their value (Ref. I1.5.6). The residence of the current tenant is located at a distance of 3830 meters in the SE sector. No meat or milk animals or vegetable gardens were identified by the latest Land Use Census for this I location, therefore, the gaseous efDuents dose at the farmer's residence is due to plume exposure from Noble Gases and the ground plane and inhalation pathways. For conservatism, it is acceptable to assume that the ingestion pathways exist at this locadon. It is assumed tLat food ingestion pathways do not exist within the SITE BOUNDARY, therefere the gaseous efDuents dose within the SITE BOUNDARY is due to plume exposure from Noble Gases and the ground plane and inhalation pathways I I

i I APA-ZZ-01003 Rev. 3 4.1.3.1.1 Direct Radiation Dose frp_m._ Qutside Storane Tank.s The Refueling Water Storage Tank (RWST) has the highest potential for receiving significant amounts of radioactive materials, and constitutes the only potentially significam source of direct radiation dose from outside storage tanks to a hEMBER OF THE PUBLIC (R.:f. I1.6.14, 11.6.15,11.6.16 and 11.6.17). Direct radiation dose from the RWST to a hEMBER OF TIE PUBLIC is determined at the I nearest point of the Owner Controlled Area fence which is not obscured by significant plant l structures, w hich is 450 meters from the RWST. The RWST is a right circular cylinder approximately 12 meters in diameter,14 meters in height I with a capacity of approximately 1,514,000 liters (Ref. I1.6.17). The walls are of type 304 stainless steel and have an average thickness of.87 cm. (Ref. I1.14.1). I The direct radiation dose from the RWST is calculated based on the tank's average isotopic content and the parameters discussed above, considering buildup and attenuation within the volume source. Appropriate methodology for calculating the dose rate from a volume source is given in TID-7004, " Reactor Shielding Design Manual"(Ref. I1.17). The computer program I' ISOSHLD (Ref. I1.18,11.19 and 11.20) will normally be utilized to perform this calculation. 4.1.3.1.2 Direct Radiation Dose from the Reactor The maximum direct radiation dose from the Unit tc a MEhiBER 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 I containment dome, taking credit for shielding provided by the containment dome and for j 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 I converted to a dose rate at the given distance by use of ANSUANS-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. I1.14.3). I l 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 hours per year. l I E I E I

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I APA-ZZ-01003 Rev. 3 l5.0 BADIOLOGICAL ENVIRONMENTAL MONITORING il DESCRTW12N OF '1HE RADIOLOGLQ_AL ENVIRQFMENTAL MONITORING PROGRAM The Radiological Emironmental Monitoring Program is intended to act s a background data base for preoperation and to supplement the radiological effluent releas.e monitoring program during plant operation. Radiation exposure to the public from the various specific pathways and direct radiation can tc adequately evaluated by this program, Some desiations frorn the sampling frequency may be necessary due to seasonal unavailability, t hazardous conditions, or other legitimate reasons. Effons are made to obtain all required samples within the required time frame. Any deviation (s) in sampling frequency or locatior, is I documented in the Annual Radiological Environmental Operating Report. l The Emironmental samples are collected and analyzed at the frequency shown in Table 5.1. g Sampling, reporting, and analytical requirements are given in Tables 9.11-A,9.11-B, and 3

9. I 1 -C.

Airborne, waterborne, and ingesdon samples collected under the monitoring prograrn are I analyzed by an independent, third-party laboratory. This laboratory is required to participate in the Environmental Protecdon Agency's (EPA) Emironmental Radioactivity Laboratory Intercomparison Studies (Crosscheck) Program or an equivalent program. Panicipation includes I all of the determinadons (sample medium - radionuclide combinadon) that are offered by the EPA and that are also included in the monitoring program. .g 5.2 PERFORMANCE TESTING OF ENVIRONMENTAL THERMOLUMTNESC_ENCE g DOS &iFR$ Thermoluminescence Detectors (TLD's) used in the Emironmental Monitoring Program are I tested for accuracy and precision to demonstrate compliance with Regulatory Guide 4.13 (Ref. I 1. I6). Energy dependence is tested at several energies between 30 kev and 3McV 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 l to either Cs-137 or Co-60. (Ref. I1.14.10) I I lI LI iI

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.__--n..~.. - - - - - - ~ ~ - - - - - " - ~ ' g g g 'm g g g-M M M-M M M M M E E E APA-ZZ-01003 Rev.3 TABLE 5.1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 1. Direct Rad:ation 40 twtine monitoring stat.nns either with two or mnre dWnwten or with one instrument for meaturing and recordicg dose rn'e cominuoudy. etmsiting of an inner rmg of sixteen station *, one in each meteexological sector in the genc1 at area of the SITE PDUNDARY. l Station i Cate Sedor She Descrjption Ircation 04 A 0.3 miles east ofIlwy O arst CC Amdion. Callaway Dectric Corp. Uti!.ty role 1.91 mi @ 349' N No.18392 47 D County Road 441,0.9 miles south ofIIwy 0, Callaway Eledric Cone-. thinty Pole 0 9 mL @ 17' NNE No. 23151 4R C County road 4 it.1.3 miles south ofIIwy O Plant Security Are Sign Pact 0.4 mi @ 45' NE 03 D Primary hieteorologient Tower i i.3 mi. 6? 75* ENE 49 E County road 442. Callaway Einiric Canp Utility Pole No. 06939. Refenn 1.7 mi. @ 93* E Wildlife Management Parking Area 32 F Ught role near East Plant Security Fence 0 4 mi @ l14* ESE 5I O Located in the "Y" ofibe sharukmed rnitroad spur northwest of sludge lagann 0.7 mi. @ 137' SE l j 30 11 County road 439,3.3 miles north ol IIwy. 94. Cattaway E!cctric Cone. thility Pete 0.9 mi.@ 163* SSE No. 33086. c7 J Coimty road 439,2.6 miles north ofI!=y. 94, Callaway Electric Coop. Utttity Pn;c I.3 mL Q l E1* 5 No UO97 37 K County road 459. 0.9 rmlet sou:h ofIiwy CC, Callaway Electric Coop. Utility 0.7 mi 2 201* SSW Pole No. 35077 43 L County road 439,0 7 miles smeh of!!wy CC, Callaway Electric Cuy. Utility 0.3 rrei. @ 230* SW Pole No. 33073 4 M liighway CC,1.0 miles =auth ofcminty road 439. Calinway Electric Co.y. Utility 1.7 mi. @ 257' TSW Pole No. I8769 -

~ ~.. - -..- -. - -. - ~ -.. ~. m m' m e a m m m m m m e e M 'M e M APA-72-01003 Rev.3 TABLE 5,1 (Omt'd) RADIOLOGICAL ENVIRONMENTAL MONTTORING PROGRAM Sution Cate Seda Site Decrigjo_n [pjeon 06 N Coun*y road 422,1.2 miles we4 of!!wy CC,Callsway Dectric Coop Dihty role 2 0 mi. @ 277* W No.19609 s 43 P County rond 42R,0.1 nules west of!!wy CC, Callsway Eledric Co.y t>tility Pete 10 mi, g 29y wNw No.1331t0 03 Q 0.1 miles west ofIIwy CC nn gravet rond 0.8 miles south ofliny O. Callaway 1.3 mi. @ 30R' NW r.ledric Conp3hility Pole Na. I R339 46 R Nuth.eut side ofItwy CC an.1 Cmmty Rnad 416 Intenedien Calf awmy Dedric t.3 mi @ 333' NSW Coop. Utility role No. 23242 An outer ring of sixteen statiom, one in each metemological sedor in the 6 to R k n range frmn the site. 36 A Cmmtypond 133, O R mite

• nouth of Cmmty Resd 132 Catlew my Dedric Coop.

3.2 mi. @ 7* N Utihty Pole No.19137 21 B Comty Rnad 133,1.9 miles north of fluy O. Callaway Dectric Crwy Unhty Pale 4.0 mi. @ 23' NNE No.19100 20 C Ilighway D,0.4 miles rwwth ofIlwy K. Callaway Eledric Conp. Utihty Pole No. 4

• mi. @ 47' NE 12833 1S D

liighway D,0 4 miles south ofIlwy O. Callaway Dedric Conp. UtJity Pete No. 3.g mi. @ 63* ENE I2932 i 17 F. Cmmty Road 4033,0.3 miles east of fiwy D, Kingdmn Telephone Comp 3"y Pole 4.0 mi. @ 87 E No.3 x 12 14 F South-cut shte ofIlwy 94 ml!!wy D Intenedinn Collawmy Dedric Canp Utdity 10 mi @ l2:* ESE l l Pole No. I1940 I1 G City of Portiuxt, Catiau ay Mcdric Coop. Utility Pole No.12112 4.R mi. @ l39* SE 10 J liighway 94,1.3 miles east of Cmmty Road 439 Callawmy Dectric Coop. tMity 4.0 mi. @ 137' SSE Pole No.12122 { -

.. ~.. - e e e a e e m W W M M M M M - M M W APA-ZZ-01003 Rev.3 TA13LE 5.1 (Cont'd) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM St.itstm CNIf Sn10r h Niptinft MaliNt 09 J North weq sMe tiny 94 amt County Road 459 Junctimt Callswsy Llnisic Cmp. 3.7 mi qt 183* 5 thility Pete No 06754 30 K Weet side of County Road 447 at the Junction with County Road 463, King &wn Teicphone Contany Pole No. 2KI 4.6 mi @ 20** SSW 42 L County Rond 447. 2 6 mdes north erCounty Road 461. Celinmy Electsic Conp 4.4 mi g 233* sw tkhty Pole No. C6326 32 hf liighwmy VY,0.6 miles west of Cosmty Road 447, Callaway Electric Cocy Utility 3.4 mi. g 2$l' WSW Pole No. 27031 i 4i N Ilighway AD,2 R miles cast ofIlwy C, Callaway Dcctric Cay. thihiy role No. 4

• mi @ 279* W l8239 40 P

North. cast side of County Rond i12 arul Ihvy 0 Junction Callaway riectric Coop. 4.2 mi@ 294* WNW Utihty role No. 06326 l 39 Q County Road i12,0.7 miles cast nrCounty Read II f Callaway rintric Coop 5 4 mi. @ 31$* NW Utility role No.17316 t i 3R R Cmmty ResJ 133,1.5 miles south ofilwy UU Colisway Electric Conp. Utility 4 R mi @ 337' NNW Pole No. 34708 Eight Statiom to be placed in special intcrest areas such as population cemen, nearby residences, schools, and in 1 or 2 arcat to sen e as contml s'atinns. 33 N City crilams Prairie, south-east of the Ihty C and !!ny AD Junction 7.4 mi. @ 273* W 3i L City of hiokane, Callaway Electric Co v. (Niity Pole No. 06037 7.4 mi @ 2It* SW 26 E Town er Americus, Cattaway Elcoric Cene. Utility Po!< No. I I I 37 12.1 mi. @ E2* E I -6 8 -

g g M M M M M M M E E A PA-72-01003 Rev, 3 TARLE 5.1 (Cont'd) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM .$l Alion 05e Sedor 3 Sile Dex_Tjtson Loc atin_n 0 27 T Town of Hiufnon, Callmay Elettric Conp. tFtility Pole No. II496 9.6 mi. O I10' ESE 35 R City of Toledo, Callaway Electric Corp.1:tility role Na.17684 5.5 ni ft 342' NNW 23 D City of Yuentan, Callaway Electric Coop. Utility role No, t 2670 6.8 mi. @ 16' NNE j 11 O City of Portland, Callaway Eledric Coop. Utility Pole No. I2112 4.8 mi @ 139' SE 20 C City of Re.ulsville, Call.may Eledric Coop. Laility Pole No.12R30 4 g mi. ft 47' NE i 34 P North <aq pide ofIIwy C and Cmmty Road.103 Jundian IP Contrnl) 9.7 mi. @ 293* WNW 0I Q liighway 7. 0.2 miles end of thitincu 54, Caliaway I.ledric Coop. Ii ility Pole 11.0 mi. @ 312* NW t (Q Control) No. 21344 1 i 2 Airborne Radioimline and Particulates + 'three anmpfes from close to the three SITE dol >NDARY locations, in different sedors, of the average highest calet lated annuai avnage ground level D'Q. 51trion Q3e Sedre f Site Description location t A1 D Primary hieteorological Tower 1.3 mi.@ ?R* ENE AR B Cotmty Road 44R,1.0 miles snuth ofIhvy 0 0.8 mi @ 24' NNE 113 A. 0.3 miles east ofItwy O and iIwy CC Junction 1.9 ni @ 34Y N -69

.. ~. ~ - m m m m mm m m m M m m m M M M M M APA-7.Z-01003 Rev.3 TABLE 5.1 (Cont'tu RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM One sample from the cmenunity with the highest D.'Q. Sta: inn Code Seem Site Descrirtion Locati_n n A9 R Comnnmity of Ref.mn 1.7 mi @ 338' NNW One sample from a control locatim, as for example 13-30 km distant and in the least preva!cnt wind direction. Station i Code Seder Site Descriggim_ location A7 Q Dartley farm 9.5 mi. @ 312* NW 3. Watertw.me

n. Surf $ce One sample upecam Station Coje Sedq Site DescripMan leccion SGI II 84 feet uMre:un of discharge, north bank 4 R mi. fa 144' SE One sample uretam Station yte Sector Site Descrirfinn

_lecatim l 502 0' l.1 miles dowtream of discharge, north bank 5.2 mi. (<? I33' SE L r

g g g g g M M r M M M-E E E E APA-ZZ41003 Rev. 3 TABLE 5.1 (Cont'd) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM

b. Drinking One nample each of the nea est water supplies within 10 miles dnumtream that could be affected by its dinbarge, to a rnaximum of thrn sampfes, and one sample frcen a centrol location.

As there are no drinking water intakes within 10 miles onwmtream of'be discharge point,the & inking water pathway is currently nM included as part of the Callow my I1 ant Radiological Environmental Monitoring Progrant Should futurewater intakes be constructed within 10 river miles dowmtream of the daharge poir( then the program will be revised to imiude this pathway (Ref. I t.6.6).

c. Seement One sample fram dowmtrearn area with existing er potential recreational value.

Station Code Setw Site Descripig!) Laca'i<m l C O 1.3 river mile downtream of dixharge, north bank 3.2 rni. @ !35' SE I 4 Ingestion

a. Milk Sample, trem mil king animals in three difTerent meteorological sedMs within a 3 km distance having the hig)ct dme pr<ential. If there are nme, then one sample Em rni! king animals in each of three di!Terent meteorological sectors between 3 to 8 km distance where drmes are calculated to be greater than 1 mrrm per year.

Due to a lack of milk animals which satisfy these requirements, the milk pathway is currently nr4 included an a part of the Cartsway !1 ant Radiciegical I mirmmental Monitoring Program. Should the Annual Land Une Census idenlify the existence t>f milking animals in locations which antisfy these rerpiremmt=, then the proeram will be revised to include this pathway.

b. Fi=h One sample of each emmercially and recreationa!!y important species in vicinity erplant discharge area.

Statinn C_p e Sector Si.te Dncrir4 ion Leentiott l C O 1.3 river mile dowmtream of dinharge, north bank 3.2 mi. @ t33' SE - -

m,.... ..m_ _ _ _. _ ~. _ _.. _. g m e m_ m M M M M M M M J APA-ZZ-01003 Rev.3 TAf1LE 5.1 (Cont'd) R ADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM One narnple of same species in arcas not influenced by plant discharge Ststion C"de Seder SS3 JRsEdgls5, { Espi;3 i l A 0.6 river miles upstream of discharge, noeth bank 4 9 mi. @ 154* SSE l e food One sample of ea<.h pincipal class of prmfucts from any area that is irrigated by water in which liquid pf ant wastes have been discharged. Products As there are no areas irrigated by water in which liquid plant wastes have been dischsrged within 30 river miles dowwtremn of the discharge pint, this sarr ple fm is net currently incl dcJ ss part of the Callsway Plant Radiological Emironmental Monitoring Propant Shnuld future irrigation water intakes be concruded within 10 river miles u downstrearn of the discharge point. then the pogram will be revised ta include this sample type (Ref.11.7.4 arvi 11.7.5). Sample / of thece ddferent kinds of broad leaf vegetation grown nearest est.h of two ddTere,d offsite locations of highe<t gedided animal aserage gemerwi 1cvel D Q (if mi;k sanyling in not performed). Starum Cmle Sedor Site f%viptien . LMation V6 R Decker's farm 1.8 mi. @ 3 44' NNW V7 A Mechan's farm 1.it mi. @ 376* N One sample of each of similar teond leaf vegetation grown ! 3 to 30 km distant in the lean nevalent wind direction (if milk sarryUi g is not perf.wmed). Siniinn Omic Sedce Site !]em_jplion f.ncation V3 L llearicy's fami !5 0 mi. @ 227' SW L - -

I APA-ZZ-OlG03 Rev.3 I 6.0 DETERMTNAT10N OF ANNUAL AVERAGE AND SHORTTERM ATMOSPHERIC DISPERSION PARAMETERS 6.1 ATMOSPHERIC DISPEP,SION PARANETERS 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 hLuch 16,1978 to March 16,1979. 6.1.1 Lont-Term Dimersion Estimates The variable trajectory plume segment atmospheric transpon model MESODIF Il (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 fotmd elsewhere (Ref. I1.6.12), The Unit Vent and Radwaste Building Vent releases are at elevatierts of 66.5 meters and 20 meters above grade, respectively. Both release points are within the building wake of the structures on which they are located, and the unit Vent is equipped with a rain cover which effectively climinates the possibility of the exit velocity exceeding five times the horizontal wind speed. All gaseous releases are thus considered to be ground-level releases, and therefore no mixed mode or elevated release dispersion parameters were determined (Ref.11.5.2). 6.1.2 Determination of Lone Term Dimersion 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 rpecial receptor locations were estimated by logarithnue interpolation according to (Ref. I1.6.13): X = X, (d / d,)" (6.1) Where: B = In (X, /X,)/in (d /d )- 2 i X,X= Atmospheric dispersion parameters at distance d and d, respectively, from I I 2 3 2 the mrce. i The distances d and d were selected such that they satisfy the relationship. 3 2 d<d<d i 2 73-I

-g E APA-ZZ-01003 Rev.3 6.1.3 Sho1 Term Dignion Estimates Airborne releases are classified as short tenn if they are less than or equal to 500 hours during a calendar year and not more than 150 hours in any quaner. Short term dispersion estimates are determined by multiplying the appropriate long term dispersion estimate by a ccrrecdon factor (Ref. I1.9.1 and 11.15.1): I F = (T, / T,)* (6.2) Where: Tg = The total number of hours of the short tenn release. T= The total number of hours in the data collection period from w hich the long term a diffusion estimate was determined (Refer to Section 6.1). Values of the slope factor (S), are presented in Table 5.3. l Short term dispersion esdmates are not applicable to short term releases which are sufficiently random in both time of day and duration (e.g., the shon term release periods are not dependent j sok!) on atmospheric condidons or time of day) to be represented by the annual average j dispersion canditions (Ref. I1.E.1). f 6.1.3.1 The Determination of the Slox Factor (S) I j The general approach employed by subroudne PURGE of XOQDOQ (Ref. I1.15.1) was utilized l to produce values of the slope of the (X/Q) curves for both the Radwaste Building Vent and the j Unit Vent. However, instead of using approximation procedures to produce the 15 percentile j (X/Q) values, the 15 percentile (X/Q) value for each release and at each locadon was determined t l by ranking all the 1-bour((X/Q)3) valu'es for that release and at that location in descending order. 4 The (X/Q); value which corresponded to the 15 percentile of all the calculated (X/Q) values within a ecctor was extracted for use in the intermittent release (X/Q) calculation. t i lI 4 4

I I

-74 ~

I APA-ZZ-01003 Rev.3 l, l \\ I l The intermittent release (X/Q) curve was constructed using the calculated 15 percentile (X/Q)j and its corresponding annual average (X/Q)a. A graphic representadon of how the ~ computadonal procedure works is ilhistrated by Figure 4.8 of reference 11.15.1. The straight line connecting these points represents (X/Q)j values for intermittent releases, ranging in duration from one hour to 8760 hours. He slope (S) of the curve is expressed as: I -log ((X / Q), /(X / Q),) S= (6.3) log (T, / T,) or -(log (X/Q), - log (X/Q),) S= (6.4) log T, - log T, 6.1.4 Atmost,herie Dispersion Parameters for Farmine Areas within the SITE BOUND ARY The dispersion parameters for farming areas within the SITE BOUNDARY are intended for a l narrow scope application: nat of calculating the dose to the current farmer from gaseous cHluents while he conducts farming activities within the SITE BOUNDARY. For the purpose of these calculadons, it was assumed that all of the farmer's time, approximately 1100 hours per year, is spent on croplands within the SITE BOUNDARY, and that his time is divided evenly over all of the croplands. Fractional acreage /dme - 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 I dispersion parameters therefore represent the distributed actisities of the farmer within the SITE BOUNDARY and his estimated occupancy period. ! 6.2 ANNUAL METEOROLOGICAL DATA PROCESSING The annual atmospheric dispersion parameters udlized in the calculadon of doses for demonstration of compliance with the numerical dose objecdves of 10 CFR 50, Appendix I, are determined using computer codes and models consistent with XOQDOQ (Ref. I1.15). These codes have been validated and verified by a qualified meteorologist prior to implementadon. Muldple sensors are udlized to ensure 90% valid data recovery for the wind speed, wind direction, and ambient air temperature parameters as required by Regulatory Guide 1.23. The selection hierarchy is presented in Tabic 6.5. I I -75 I

m m M M M M M M M M M M M M M M E E' M APA-7E01003 Rev.3 TABLE 6.1 IIIGIIEST ANNUAL AVERAGE ATMOSPIIERIC DISPERSION PARAMETERS (a} UNIT VENT DISTANCE X/Q X/Q LOCATION (b) SECTOR (METERS) XT) DECAYED / DECAYED / D'Q UN Dr.PI.ETED DEPLETED e--...+,-,

zw.m

.r,. .,,g p.m.~,:. _. wmmm:cy... u :M > L.& &:Li %,,% m ,.ymp nw iAKO4). < :n: .z 2 ':q-u c.m n ....: wx: p- ) iu. ..,%.pn &. n .m (tec/m-(sec/m- ) (=cc!m (m) SITE DOUNDARY NNW 2200 1.0 E-6 9 9E-7 8 SE 7 4.3 E-9 Nearest Cow (c) WSW 2172 4 4E-7 4.4E 7 3.RE-7 f.6 E-9 Nearest Goat (c) WSW 2172 4.4 E-7 4.4 E-7 3.RE-7 1.6 E-9 Neare*t M-at Animal (d) NNW 2864 6.R E-7 6.RE-7 3.7 E-7 2 6E.9 Nearest Vegetab!c (c) NNW 2R64 6 SE-7 6 RE-7 3.7E-7 2.6E-9 Garden Nearest Residence (c) NNW 28(;4 6.NE-7 6.RC-7 3.7E-7 2.6E-9 Tarming Arce within the N/A. N/A 2.1E-7 2.1E.7 1.9E-7 1.1 E-9 Site Boundary (c)(e) (m) Values given are from TSAR Table 2.3-82 (b) Data from 1992 Imd Use Cemus (t) values d,ived from TSAR Table 23-83, using the metixwlofegy presented in Equation (6.1)(Ref. I1.5.6) (<!) De nearest meat animal is assurned to exist at the Incation of the nearest resident. (c) new values were derived for a narrow scope application. Extreme caution should be exercised when determining their suitability for me in other applications. Huilding Shape Parameter (C)- 0.3 (Ref. I1.3.3) ( Veitical licight ofliighest Adjacent Building (V) = 66.45 meters (Ref. I1.53) i -

~ .~ m m mm m m m m m W m m M M M m M M M APA-ZZ-01003 Rev.3 T_ABLE 6.2 HIGHEST ANNUAL AVERAGE ATMOSPHERIC DISPERSION PARAMETERS (a) RADWASTE BU!LDING VENT DISTANCE X/Q N/Q II) CATION (b) SECTOR (METERS) X/Q DECAYEDI 1)ECAYED/ D/Q UNDEP!J.TED l'EPIITED x--mm ..w,. , a.x m. i ;. -,r - - ~~ y y y n n y w n e.A:y m m.%.. . p.,-~.m y r p g y< m p m..,.s.m.q) r n.& r e ... x (sec/nr ) c.a.,. (sedm ) (secim-(m') l S!TE llOUNDARY NNW 2200 13E4 13 E-6 1.t E-6 43 E-9 Nearest Cow (c) WSW 2172 5.7 E-7

5. 7 E-7 4 9E-7 1.6D9 Netrett Goat (c)

WSW 2172 5.7E-7 5.7E-7 4.9E-7 1.6 E-9 Nearest Meant Animmt(d) htNW 2K61 R.7D7 8.7E-7 7.2D7 2.6 E-9 Nearest Vegetable Garden (c) NNW 2F64 R.7E.7 8.7E 7 7.2 E-7 2 6E-9 Nearest Residence (c) NNW 2R64 It.7E-7 87D7 7.2 E-7 2.6 E-9 ranning Areas Within N/A N/A 2.9 E-7 2.9E-7 2.6 E-7 1.1 E-9 Site 11oundarv (c)(c) (a) Values given are from FSAR Tahic 23-84 (h) Data from 1992 (>md Uw Cemus (c) Values derived frern ESAR Table 23-81, u. ting the methodology presented in Equation (6 l} (Ref. I 1.5.6) (d) 1hc nearest mest animal is as*umed to exist at the location of the nearest resident. (c) These values were derived for a narrow scope applicatinrt Extreme caution Omid be esercised whm determining their suitahiiity for use in Wher applicatie. Ituildmg Nec Parameter (C) = 0.3 (Ref.11.53) Vertical licight of IIighest Aljacent Ehdiding (V)

• 19.96 meters (Ref. I 1.53) -

APA-ZZ4)l003 TABLE 6.3 SJJORT TERM DISPER$10N PARAMETERS (a) (c) Slope t actor (s) I Location (b) Sector Distance Unit Vent ladwaste Building Vent I l Site Boundary S 1300 .32S .320 Nearest Cow NW 5053 .263 266 Nearest Goat NW 5053 263 266 Nearest Meat Animal NNW 2736 .262 .268 Nearest Vegetabl; Garden NNW 2865 .264 .268 I Nearest Retidence NNW 2865 264 268 I (a) Refertnce 11.5.3 l (b) Dat/4 from 1992 Land Use Ceru;us (c) Recircidation Factor = 1.0 LI ll lE

lI

'I ,I I 1

'M m m m m W M M m W W W M M M M M 'M.W APA-ZZ-01003 Rev.3 TABLE 6_4 APPLICATION OF ATMOSPIIERIC_ DISPERSION PARAMETERS DOSE PAT 11WAY D SPERSION PARAMETER COffrROLLING AGE QROW [ONTROLLING LOCATION l Nobic Gat, Ucta Air x/Q, decay ed/undepleted Site Beundary (2.26 day hainife) Nobic Gas. Garnma Air x/O, decayed /undepicted Site Boundary (2.26 day halflife) Noble Gas, Total Body x/Q. decayed /undepleted Site Boundary (2.26 day halflife) Noble Gas. Skin x/Q, decayed /und picted Site Doundary (2.26 day half ~ ire) Ground Planc Deposition D/Q Nearest Resident t Inhalation x/Q, decay ed/ depicted Child Nearest Resident [ 1 (8 day halflife) Vegetation D/Q' Child Nearest Resident Milk D/Q' Child Nearest Resident i Meat D/Q* Child Nearest Resident

• For 11-3 and C-14, x/Q, decayed' depleted is used instcati of DiQ (Ref. I 1.11.1).

1 l. 79-w w e s--

APA-ZZ-01003 Rev.3 I TABLE 6.5 METEOROLOGICAL D ATA SELECTION lirEARCHY I I I Parameter Primary Firs Second Third Alictnate Alternate Alternate Wind S;x:cd 10m Pri 10m Sec 60m Pri 90m Pri Wind Direction 10m Pii 10m Sec 60m Pri 90m Pri Air Tc:aperature 10m Pri 10m Sec Wind Variabili:y 10m Pri 10m Sc 60m Pri 90m Pri Temp Diflerence 60-10m Pri 90-10m Pri 90-60m Pri Dew Point 10m Pri Precipitaiton Im Pri I l (a) Priindicates pJman tower l (b) Sec indicates secondary tu cmr 15 lI I I B .I l I I

I APA-ZZ-01003 Rev. 3 7.0 EsPORTING REOUIREMENTS l7.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (CTS #2804) Routine Annual Radiological Emironmental Operating Report covering the operation of the unit I l during the previous calendar year shall be submitted prior to May 1 of each year. The Annual Radiological Emironmental Operating Report shall include summaries, I interpretatior.s, and an analysis of trends of the results of the radiological emironmental surveillance activities for the report period, including a comparison with preoperational studies, with operational controls and with previous emironmental surveillance reports, and an cg assesstnent of the observed impacts of the plant operadon on the emironment. The reports shall B i iso ine'"de the rcSu2's of tand use Censuses required by P2C 9.12 l The Annual Radiological Emiron.nental Operating Report shall include the results of analysis of I all radiological emironmental samples and of all emironmental radiadon measurements taken l during the period pursuant to the locations specified in the ODCM, as well as summarized and tabulated results of these arnlyses and measurements in the format of the table in th: Radiological Assessment Branch Technical Posidon, Revision 1, November 1979. In the event I that scme indhidual results are not available for inclusion with the report, the report shall be i submitted noting and explaining the reasons for the missing results. The missing data shall bc l submitted as soon as possible in a supplementary report. l The reports shall also include the following: a summary descripdon of the radiological emironmental monitoring program; at least two legible maps

• covering all sampling locadons

!I keved to a table giving distances and directions from the centerline of one reactor; the results of 1;censee pardcipation in the Interlaboratory Comparison Program and the correcdve action being taken if the specified program is not being performed as required by REC 9.13.1; reasons for not conducting the Radiological Emironmental Monitoring Program as required by REC 9.11.1 and ,I discussion of all deviations from the sampling schedule of Table 9.11-A. discussion of emironmental sample measurements that exceed the reprting levels of Table 9.]l-B, but are not l the result of the plant effluents, pursuant to REC 9.1 L1; and discession of all analyses in which i the LLD required by Table 9.ll-C was not achievable. l7.2 EEiMIANNUAL RADIOACTIVE EFFLUENT PILEASE REPORT (PTS # 2805) Roudne Sendannual Radioactive Emuent Release Reports covering the operatica of the unit l during the previous 6 months of operation shall be submined within 60 days after knuary I and } July 1 of each year. t The Semiannual Radioactive Emuent Release Report shall include a summary of the quantities of radioactive liquic and gaseous emuents released from the unit as outlined in Regulatory Guide 1.21,

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

I 4

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

I . ' I

I APA-ZZ-01003 Rev.3 The Semiannual Radioactive Emucnt Release Report to be submitted within 60 days after i January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the fonn of an hour by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form ofjoint frequency distribudon of wind speed, wind direction, and I atmospheric stability This same report shall also include an assessment of the radiation doses due to the radioactive I-liquid and gaseous emuents released from the unit during the previous calendar year. This same repon shall also include an assessment of the radiation doses from radioactive liquid and gaseous emuents to MEMBERS OF THE PUBLIC due to their activides inside the SfrE BOUNDARY (Technical Specificadons, Figures 3.1-3 and 5.1-4) during the report period using I historical average atmospheric conditions. All assumptions used in making these assessments, i.e., specific activity, exposure time and location, shall be included in these repons. The rnetcorological conditions concurrent with the time of release of radioactive materials in gaseous - I emuents, as detennined by sampling frequency and measurement, shall be used for deterndning the gaseous patinvay doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM). The Semiannual Radioactive Emuent Release Repon to be submitted within 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most I exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary efDuent pathways and direct radiation. for the presious calendar year to show conformance with 40 CFR Pan 190, "Emironmental Radiation Protection I Standards for Nuclear Power Operation." Doses to the MEMBER OF THE PUBLIC shall be calculated using the methodology and parameters of the ODCM. The Semiannual Radioactive Emuent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in ganous and liquid emuents made during the reporting period. g The Sendannual Radioactive Emuent Release Repons shall include a summary description of 3 any major chang s made during the reporting period to any Liquid or Gaseous Treatment Systems, pursuant to Section 10.1. It shall also include a listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to I PIC 9.12.1. Reporting requirements for changes to Solid Waste Treatrnent Systems is addressed in APA-Z.Z-01011, PROCESS CONTROL PROGRAM (PCP). The Semiannual Radioactis e Emuent Release Reports shall also include the followi ig information: An explanation as to why the inoperabihty ofliquid or gaseous efflant monitoring I instrumentadon was not corrected within the time specified in REC 9.1.1 or REC 9.2.1, and a descripdon of the events leading the liquid holdup tanks or gas storage tanks exceeding the limits of Technical Specification 3.11.1.4 or 3.11.2.6.

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

I APA-ZZ41003

- E-Rev.3 _ g The Semiannual Radioactive Ef!1uent Release Reports shall also include as a part of or submitted I concurrent with, a complete and legible copy of all revis6ns of the oDCM that occurred during the reponing period pursuant of Technical Specificadon 6.14 E soiid wasic reponing is adaresscd in Apr.zz.oi0ii. pnocess cosTaot raocaru i i l g (PCP). I !I R. I ) I I I I I I I .e3

I APA-ZZ4)l003 Rev.3 l8.0 IMPLEMENTATION OF ODCM METHODOLOGY (CTS #2791) . I The ODCM provides the mathematical relationships used to implement the Radioactive EfIluent Controls. For routine effluent release and dose assessment, computer codes are utilized to I implement the ODCM methodologies. These codes are evaluated in a::cordance with the requirements of HDP ZZ-04500, ' Rad / Chem Computer Systems Conduct of Operations", to ensure that they produce results consistent with the methodoisgies presented in the ODCM. Procedures which implement the ODCM methodology are contained in the Plant Operatini, I Manual. I I I I I I I L I I I l l I .s,.

1 i I APA-Z.Z-01003 Rev.3 I 9.0 RADIO ACTIVE EFFLUENT CONTROLS (REQ NOTE: 1. The terms in this section that appear in capitalized type are defined in Tecimical Specificador.s. 2. All frequency notations are per Table 1.1 of Technical Specifications. 9.0.1 Compliance with the Controis contained in the su.cceding Controls is required during the I OPERATIONAL MODES or other conditions specified therein; except that upon failure to meer the Control, the associated ACTION requ;tements shall be met. I 9.0.2 Noncompliance with a Control shall exist when the requirements of the Control and associated ACTION requirements are not met within the specified time intervals if the Controlis restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required I I I I I I I I I I e

I APA ZZ-01003 Rev.3 .I 9.1 RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION 9.1.1 . Controls (CTS # 41834) 9.1.1.1 The radioactive liquid emuent monitoring instrumentation channels shown in Table 9.1-A shall be OPERABLE with their Alarnt Trip Setpoints set to ensure diat the limits of REC 9.3.1.1 are f not exceeded. The AlarmA^ rip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM). APPLICABILITY: At all times. 6CT10N a. With a radioactive liquid emuent monitoring instrumentation channe! AlarntrTrip Se: point less conservative than requited by the above Control, immediately suspend the release of radioactive liqu:d emuents monitcred by the affected channel, or declare die channel inoperable. b. With less than the minimum number of radioactive liquid emuent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 9.1-A. Restore the inoperable instrumentation to OPERABLE status within the time specified in the I ACTION, or explain in the next Semiannual Radioactive Emuent Release Report, pursuant to Section 7.2, why this inoperability was not conected within the time specified. l c. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable. 9.1.2 Surveillance Reauirements 9.1.2.1 Each radioactive liquid emuent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 9.1-B. I I I I 1 I I .u.

-. - ~... - - -.... -. ~. - ~. . -. -. - -. ~. . -.. ~. m m m m m 'm m m m 'M M m m M. M M m M M ,l i APA-Z7e01003 l Rev.3 TABLE 9.1-A RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM INSTRUMENT C!!ANNELS OPERABLE ACTION I Radioactivity Monitors Providing Alann and Automatic Termination of Rc! case

a. Liquid Radwaste Discharge Monitor (llB-RE-IR) 1 31

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

l

2. Flow Rate Measurement Devices

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

b. Steam Generator B!owdown Dix harge Line (UM-FE-0054) 1 34

c. Combined Cooling Blowdown and Dypass Flow I

34 1 -

I APA-ZZ4:1003 Rev.3 I TABLE 9.1-A (Cont'd) I ACTION STATEMENTS ACTION 31 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, cmuent releases sia this pathway may continue for up to 14 days provided that prior to initiating a release: l At least two independent samples are analyzed in accordance with REC 9.3.2.1, and a. I b. At least two technically quzliDed members of the facility staffindependently verify the release rate calculations and discharge line valving. Otherwise, suspend eclease of radioactive emuents via this pathway. ACTION 32 - With the number of channels OPERABLE less than required by the Miniraum Channels OPERABLE requirements, emuent releases via this pathway may continue for up to 30 I days provided grab samples are analyzed for principal gamma emitters and 1-131 at a low r limit of detection as specified in Table 9.3-A: a. At le'st once per 12 hours when the spccific activity of the secondary coolant is greater than 0.01 micro Curie / gram DOS:; IQUIVALENT l-131, or b. At least once per 24 hours u hen the specific activity of the secondary coolant is less than or equal to 0.01 microcurie / gram DOSE EQUlVALENT I-131. l ACTION 33 - (Deleted) l ACTION 34 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, emuent releases sia this pathway may continue for up to 30 days provided the Dow rate is estimated at least once per 4 hours during actual releases. Pump performance curves generated in place may be used to estimate flow. I I I I I .ss. g

. - -.. - - - - - - - - -... -. ~ ~....... ~. -... - - ~.... -.. -.. l APA-ZZ-01003 Rev.3 l TABLE 9.1-11 l RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIR_EhiENTS I ANALOG CilANNEL CIIANNEL SOURCE CIIANNEL OPERATIONAL INSTRUMENT CilECK CilECK CALIBRATION TEST

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste Discharge Monitor (HD-RE-IR)

D P R(2) Q(1)

b. Steam Generator Blowdown Discharge Monitor D

M R(2) Q(1) (DM-RE-52) l l

2. Flow Rate Measurement Devices

a. Liquid Radwaste Blowdown Discharge Line (IIB-FE-2017)

D(3) N.A R N.A.

b. Steam Generator Blowdown Discharge Line (DM-FE-0054)

D(3) N.A. R N.A.

c. Combined Cooling Tower Blowdown and Bypass Flow D(3)

N A. R N.A. l -

APA-ZZ-01003 Rev.3 TABLE 9.1 B (Cont'd) TABLE NOTATIONS (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic I isolation of this pathway and control room alarm annunciation occur as appropriate if any of the following conditions exists: I Instrument indicates newred levels above the Alarm / Trip Sctpoint (isolation and alarm), a. or b. Circuit failure (alarm only), or Instrument indicates a downscale failure (alarm only), or c. d. Instrument controls not set in operate mode (alarm only). (2) The initial C}WINEL CALIBRATION shall be performed using one or more of the reference I (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 mor.itor response to I 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 cortsist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days on which continuous, I periodic, or batch releases are made. I I I I I I I ' I

I I APA-ZZ-01003 Rev.3 l 9.2 ~ BADlOACTIVE GASEOUS EFFLILENT MONITORING INSTRUMENTATION 9.2.1 Controls (CTS # 41834) 9.2.1.1 The radioactive gaseous emuent monitoring instmmentation channels showm in Table 9.2-A shall bc OPERABLE with their AlarmfTrip Sctpoints set to ensure that the limits of REC 9.6.1 1 and Technical Specification 3.11.2.5 are not exceeded. The AlarmfTrip Serpoints of these I channels meeting REC 9.6.1.1 shall be detennined and adjusted in accordance with the methodology and parameters in the ODCM. AIPPLICABILITY: As shour in Table 9.2-A. I ACTION: I With a radioactive gaseous emuent monitoring instrumentation channel Alarmffrip Sctpoint a. less conservative than required by the above specification, immediately suspend the release of radic. active gasecus emucnts monitored by the afTected channel, or declare the channel inoperable. b. With less than the minimum number of radioactive gaseous emuent monitoring instrumentation charmels OPERABLE, take the ACTION shown in Table 9.2-A. Restore I the inoperable instrumentation to OPERABLE status within the time specified i ACTION, or explain in the next Semiannual Radioactive Emuent Release Report, pursuant to Technical Specification 6.9.1.7, w by this inoperability was not corrected within the time specified. l The prosisjons of Technical Specifications 3.0.3 and 3.0.4 are not applicable c. l9.22 Surveillance Requirements 9.2.2.1 Each radioactive gaseous emuent monitoring instrumentation channel shall be demonstrated OPEPABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CILANNEL I CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 9.2-B. I I I I I L-. a ~

M M M M M M M M E E E E E E E' APA-ZZ-01003 Rev.3 TABLE 9,2-A RADIOACTWE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINI AfUNI CII ANNELS INSTP.UMENT OPEE6DLE A.PIPMCADl!JTY ACTION

1. Unit Vent System l

NW'le Gxt Ac1ivity Moniter - Providing Mann 1 At all tunes 40,46 a { (GT-RE-21) b Intsne Sampin 1 At n!! times 43

c. Putwulate Sanper i

At all times 43 \\ d Unit Vent Flow Rate i At att times 43 l l

e. Particulate and Indine Sampler Flow Rate himitor i

At all times 39 l

2. Containment Purge Sydem l

a Noble Gss Activity Monitor - Providing Alvm and 1 M odes 1, 2, 3, 4 41 Antmuntic Termination ol'Relcue (GT.RE-22, nr.1 chiring GT-R E-3.1) Core Alterations l

b. Imtme Sanper i

1 MMes 1, 2, 3, 4 43 and doring Core Alicrations ]' l

c. Particulate Sampler 1

MMes 1,2,3,4 43 ami during Core Alterat.nns l J. I' low Raie N/A N/A N/A l Particulate atxt lodine Sunpler T!ow Rare Monitor I MMes I,2,3,4 39 e and durin; Core Altaatiom 1,

m W M M M M m M M M M ' mM APA-71A1003 Rev.3 TABLE 9.2-A_(Gppfd) RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTAT]_ON MINIMlIM CIIANNELS INSTRtIMpg OPERARL1 APPLICABILITY ACTION f 3 Radwsde Building Vent Systern l

a. Noble Gas Ac1hity Monitew Preiding Alarm and I

31,40 Autmutim Tem 6narkm of He: case (GII-RE-10) At a!I times i k fa!rne Samp;cr 1 At alt t;nws 43

r. Particulate Sampler I

At n!! times 41

d. Flow Rate N/A N/A N/A l

Particulate am! Iodine Sampier Flow Rate Moniter 1 At all times 39 e i k b - - -

APA-ZZ-01003 TABLE 9.2-A (Cont'd) TABLE NOTATIONS ACTION STATEMEES ACTION 38 - With the number oflow range channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank (s) may be I released to the emironment for up to 14 days provided that prior to initiadng the release: At least two independent samples of the tank's contents are analyzed, and a. b. At least two technically qualified members of the facility staffindependently verify the relcase rate calculadons and discharge valve lineup. I Oth ruise, suspend release of radioactive efIluents via this pathway. I ACTION 39 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, emuent releases sia this pathway may continue prosided the flow rate is estimated based on fan status and operadng curves or actual measurements at least once per 4 hours. ACTION 40 - With the number oflow range channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, emuent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 12 hours and these samples are analyzed for radioactivity within 24 hours. ACTION 41 - With the number of channels OPERABLE one less than required by the Minimum I l Channels OPERABLE requirement, immediately suspend PURGING of radioactive emuents via this pathway. i I ACTION 43 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE regturement, emuent releases via the affected pathway may continue for up to 30 days provided samples are continuously colletted with auxiliary sampling equipment as required in Table 9.6-A. i 1 ACTION 45 - Flow rate for this system shall be based on fan status and operating currer or actual ) measurements. l l 1 I ACTION 46 - For midrange and high range channels only - with the number of OPERABLE channels i less than the minimum channels OPERABLE requirements of Table 9.2-A, take the action specified in Technical Specification 3.3.3.6, ACTION C. i

. -. _ -. ~,. - ~. - - - - - - - - - - - - - - - - - - - ~ ~ m m m m m M m M M M M M M M M APA-7E01003 Rev.3 TABLE 9.2-0 RADIOACTIVE G ASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQI)JRIiMENTS ANAL 00 Cll ANNEL MCDES FOR W1(ICll CIIANNEL SOURCE CilANNEL OPERATIONAL SLTR VEll1ANCE INJTRlLjD4T C11ECE CilLCK CA!.!!1R ATION TEST IS REQUIRED

1. linit Vent S>Mem Noble Gas Aoivity Moniinr a

hwi.fing Af arm (GT-R F.-2 t) D M R(3) Q(2) At all times

b. Init.n Swrin W

NA N A. NA At all tirnes Partkula'e Sar rler W NA NA NA At sti times c l d tinit Vr.: Mow Rate NA N A. P(4) Q At all times [ e Particulate an 1 Ra,tioiMine Sampler Mme P,re Maniter D N A. R Q At sit times

2. Cmtamment 1%rge Sptem t'et.le Gas Activity Monitor -

t Prosting Ala m arkt Autrinstic '!ceminatim of Release Mc=les 1. 2,3,4 and during (tt F RT. 22, GI-R EJ3) D P H(3) Q(1) Core Altantims

b. jutmc Sampler W

NA NA NA MMes I,2,3,4 and duing Core Aherations e Particulate Sampler W NA NA NA Modes I, 2. 3, 4 and duricg Core Ahcratims l

d. Cmtainment INrge Ventdatim now Rate NA NA R(4)

NA Modes 1,2,3,4 and during Core Aheratims l Particulate and Radiciodine Sampler Dow Rate Monste-D NA R NA Modes t,2,3,4 e and arrir g Core Ahcratmns

3. R aihs noe Thildrig Vent Spten

a. Noti?erias Actidry Mmiteir.

Irnvid rig Alarm and Automatic Tern 6natinn o( Retense (Gif.Ria10) D,P M,P RO) O(1) At n!! times b Inline Sampler W NA NA NA At att times

c. Paiticulate Swpler W

NA NA NA At alltimes a RaJete Dnilding Vent now Rate NA NA R(4) NA At all times

c. Paniculate snel Radioiodine Sampler Mow Rate Monitor D

N A. R NA At all times.

I APA-2.Z4)l003 I Rev.3 IABLE 9 2-B (Cont'd) TABLE NOTATIONS (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isoladon of this pathway and control room alarm annunciation occur as appropriate if any of the l following conditions exists: I Instrument indicates measured levels above the Alarm / Trip Setpoint (isolation and alarm), a. or b. Circuit failure (alarm only), or Instrument indicates a downscale failure (alarm only), or c. d. Instrument controls not set in operate mode (alarm only). (2) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciadon occurs if any of the follewing condidons exists: a. Instrument indicates measured levels above the Alarm Setpoint, or b. Ciremt failure, or c. Instrument indicates a downscale failure, or d. Instrument controls not set in operate mode. (3) The inidal CHANNEL CALIBRATION shall be perfonned using one or more of the reference (gas or liquid and solid) standards certified by the Nadonal Institute of Standards & Technology { (NIST) or using standards that have been obtained from suppliers that participate in ] I measurement assurance activities with NIST. These standaids shall permit calibrating th. system over its intended range of energy, measu ement rangt, and establish monitor response to a solid calibration source, For subsequent CHANNEL CALIBRATION, NIST traceable standard I (gas, liquid, or solid) may be used; or a gas. liquid, or solid source that has been calibrated by relating it to equipment that was previously (within 30 days) by the same geometry and type of source traceable to NIST. (4) If flow rate is determined by exhaust fan status and fan performance cmves, the following surveillance operations shall b: performed at least once p:r 18 months. a. The specific vent flows by direct measurement, or b. The differential pressure across the exhaust fan and vent flow established by the fan's " flow-AP" Curve, or c. The fan motor horsepower measured and vent flow established by the fan's

• flow-horsepower" curs c.

1 I . I 1 \\

) . E APA-Z.Z-0100'3 Rev.3 l9.3 LIOU1D EFFLUENTS CONCENTRATION I l l9.3.1 fontrol (CTS # 41834) J f \\ The concentration of radioact ve material released in liquid emuents to UNRESTRICTED i 9.3.1.1 AREAS (see Technical Specifications, Figure 5.1-4) shall be limited to the concentration l specified in 10 CFR Part 20.1-20.601, Appendix B, Table II, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10-4 microCuric/ml total activity. (CTS

• 4160) i APPLICABILITY: At all times.

ACTION: a With the concentration of radioactive material rcleased in liquid emuents to UNRESTRICTED AREAS exceeding the abo,e limits, immediately restore the concentration to within the above limits. l b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable. l9.3.2 Surveillance Reauirements 9.3.2.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 9.3-A. I 9.3.2.2 The results of the radioactivity analysis shall be used in acccrdance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of REC 9.3.1.1. I I I I I I I ' I

) E APA-ZZ-01003 Rev.3 TABLE 9 3-A RADIOACTIVE L10UID WASTE SAMPLING AND ANALYSIS PROGRAM UQUID FLELEASE SAMPUNG MINIMUM -TYPEGr ~1LWER UMIT ' TYPE . FREQUENCY (7) ANALYSIS ACTBTTY OF DETICTION FREQUENCY AN-ALYSIS (LLD)(1) I (ucilm!) J. Batch Waste P Principal Gamma I Release Tanks G) Each Ttatch Eminen (3) 3 x10'7 I 4 l lal 1 x 10 I d I

a. Discharge Dissoh-ed, J lal0 1

Monitor Entrained Gucs Tank (Gamma Emium) 0 Each Bauh 11-3 1x10 i g M Gross Alpha lx10-7 Comrosite (#) l Q St-89. Sr-90 5x10'8 Composite (4) 4 Fe45 lx10

2. Condnuous Daily

}Tincipal Gamma S x10-7 ReleasesO) ~ EmittenO) 4 1-131 lx10 0 Steam Dissolved and lx10 I Generator En*. rained Gases Blowdown (Gamma Enuncrs) l DaihlO Grab S' amp!c !! 3 1xlod M Gnss Alpha i n 10'7 Con.c..siteIN I St-89, Sr-90 3x10'8 -Q, 4 Fe45 lx10 I I

I APA-ZZ-01003 - TABLE 9.3-A (Cont'd) 4 TABLE NOTATIONS l (1) The LLD is described in Attachment 1. (2) A batch release is the discharge ofliquid wastes of a disa ete volume. Prior to sampling for ~ analyses, each batch shall be isolated, and then thoroughly mixed a method described in the ODCM to assure representativt sampling. (3) The principal gamma emi:ters for which the LLD control applies include the following I radionuclides: Mn-54, Fe-59, Co-58, Zn45, Mo-99, Cs-134, Cs-137, Cc-141, and Cc-144, This list does not mca-that only these nuclides are tv be considered. Other gamma peaks that are identifiable, together with those of the above in the Semianmut! Radioactive Emnent Release Report pursuant to Technical Specification 6.9.1.7, in the format outlined in Regulatory Gtdde 1.21, Appendix B, Revision 1, June 1974. (4) A composite sample is one in which the quantity ofliquid sampled is proportional to the quantity g ofliquid waste discharged and in which the method of sr.mpling employed results in a specium i 1 3 that is representative of the liquids released. Prior to analysis, all samples taken for the composite shall be thoroughly mixed in order for the composite samples to be representative of the emuent release. (5) A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release. I (6) Samples shall be taken at the initiation of emuent flow and at least once per 24 hours thereafter while the release is occurring. To be representative of the liquid emuent, the sample volume shall be proportioned to the emuent stream discharge volume. The ratio of sample volume to I emuent discharge volume shall be maintained constant for all samples taken for the composite sample. 4 l (7) Samples shall be representative of the emuent release. 4 I l I I i I 99

3 4 APA-ZZ-01003 i Rev.3 e 9.4 QQ1E_ FROM LIOUID EFFLUENTS 1 l 9.4.1 .C_ontrols (CTS # 41834) i g 9.4.1.1 The dox or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in 3 liquid effluents released, from cach unit, to UNRESTRICTED AREAS (see Technical j - l SpeciDeations, Figure 5.1-4) shall be lirruled: During any calendar quaner to less than or equal to 1.5 mrems to the whole body and to a. j less than or equal ta 5 mrems to any organ, and i !g b. During any calendar year to less than or equal to 3 mrems to the whole body and to less g than or equal to 10 mrems to any organ. l 6PPLICABILTD': At all times. I j l ACTION: (CTS A i161) i }E with the calculated dose rrom tac reicas= or radioa=tive materi is in iia"id ernuents a-W excceding any of the above limits, prepare and submit to the Commission within 30 days, i pursuant to Techr.ical Speci6 cation 6.9.2, Special Report that identines the cause(s) for execeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the prop 3 sed corrective actions to be taken to assure that subsequent releases 4 will be in compliance with the above limits. This Special Report shall also include: (2) the t results of radiological impact on fmished drinking water supplies with regard to the requirements of 40 CFR Part 141, C!can Drinking Water Act. l b. The provisions of Technical Specincations 3.0.3 and 3.0.4 are not applicable. 9.4.2 Surveillance Reauirements 9.4.2.1 Cumulath e dosc contributions from liquid efDuents for the current calendar quarter and the j I current calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days. ) I I I I I

• The requirements of ACTION a (1) and (2) are applicable only if drinking water supply is taken from ihe rece:ving water body within 3 miles of the plant discharge. in the case of river-sited plants this is 3 miles dowanretirn only.

I 100-

- _ ~ - - - =... - LI APA-ZZ-01003 1, Rev,3 i 9.5 L10UID RADWASTE TREATMENT SYSTEM }'=l i l9.51 Controls (CTS # 41834) !g 9.5.1.1 The Liquid Radwaste Treatment System shall be OPERABLE and appropriate portions of the

3 system shall be used to reduce releases of radioacdvity when the projected doses due to the liquid

{ emuent, from each unit, to UNRESTRICTED AREAS (sec Technical Specificadons, Figure 5.1-l

4) would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31 day period.

i j APPLICABILITY: At all dmes. a l ACTION: (CTS #11til) { a. With radioactive liquid waste being discharged without treaunent and in excess of the +g above limits and any portion of the Liquid Radwaste Treatment System not in operation, 'g prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that includes the following informadon: I 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.) Sursmary descripdon of action (s) tien to prevent a recurrence. l b. The provisions of Tecimical Specifications 3.0.3 and 3.0.4 are not applicable. l9.5.2 Surveillance Reauirements 9.5.2.1 Doses due to liquid releases from each unit to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodokgy and parameters in the ODCM when Liquid Radwaste Treatment Systems are not being fully utilized. 9.5.2.2 The installed Liquid Radwaste Treatment System shall be considered OPERABLE by meeting I REC 9.3.1.1 aad 9 4.1.1. I I I I >I -101-

f APA-ZZ41003 Rev.3 9.6 9ASEOUS EFFLUENTS DOSE RATE l9.6.1 Controls (CTS # 41834) 9.6.1.1 The dose rate due to radioactive materials released in gaseous efIluents from the site to areas at and beyond the SITE BOUNDARY (see Technical Specifications, Figurc 5.1-3) shall te limited { l to the following: a. For noble gases: Less than or equal to 500 mremvyr to the whole body and less than or equal to 3000 mremvyt 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 nuems/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 witidn the above limit (s). l b. The provisions of Technical Specifications 3.0.3 and 3.0,4 are not applicable. E 9.6.2 Surveillance Recuirements 9.6.2.1 'Ihe dose rate due to noble gases in gaseous e!Iluents shall be determined to be within the above limits in accordance with the methodology and parameters in the CDCM. L 9.6.2.2 The dose rate due to Iodine-131 and 133, tritium and all radionuclides in paniculate form with half-lives greater than 8 days in gaseous efIluents shall be determined to h. within the above f limits in accordance with the methodology and parameters in the ODCM by obtaining L representative samples and perfonning analyses in accordance with the sampling and analysis program specified in Table 9.6-A. I L:. f: I -102-I g L

~ ~. - - - - - - - - ~ ~ - - - - M M M M M M M M-M M M M N M M M M' APA-ZZ-01003 Rev.3 TABl.E 9.6-A RADIOACTIVITY GASEOUS WAST3 iPLING AND ANAU 'OGIMM MINIMUM AtiAIA SG LOWER LIMIT L'F 4 O ASEOUS REIIA3E TYPE 1 SAMPt.!NG FREQUENCY (9) . FREQUENCY TYPE OF ACTlV!T Y ANALA Yl33 DETICTION (Lt D)(1) i' (pCi/mt) I Waec Gu IVemy P P thncipal Gamma EmittersUI l ank ls10 F=h Tar k Grob $=mple l'arh Tank 2 ( u= nnent hup m VentU) P P Ihipal Onmma Emitten(2) 4 Each PURGEU) g,g g Each PURGEU) Ginh Sample M It-3(emide) Is10 # 31: nit Vent MUMd) A[3) }W;pst Gamma EmetenU) J is10 G ah 5emple kdU 11.3 (osik) 1=10 4 4 Spent Fuct Dmidq fishaust A[3) M hmcip=1 Gamme Emitten#2) 1510 Grab snmpic W \\d3) Il-3 (mi&) 1510# Gesh hmple I i fe%I4s*te Dmidmg Vent M M Ptinegwl Gamma EmoterO) Is10 Ge=S Sample 6 Ai! Reic==c T)T** *, hsied in I.,2.,3.,4, Continumg6xt) gg U I-131 1x10 ast 5. above Chartoel Sample 40 l-133 1x10 i C,wimnm(6)(1) idD Principal Gamma Tmitte's(2) g, g n-1 I I' articulate Sample CentmunusNI M Oms. A!pha 1 = 10'I I Compns e e P=riits!=te Sampic Continunus (6X8) t Q Sr.29, St-90 1x10 Camprete P=tmelete Sante l -103- /

I APA-ZZ-01003 Rev.3 i i TABLE 9.6-A (Cont'd) TABLE NOTATIONS (1) The LLD is described in Attachment 1. (2) The principal gamma emitters for which the LLD specification applies include the fohowing I 8 radionuelides: Kr-87, Kr-88, Xc 133, Xe-133m, Xe-135, and Xc-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co40,2n45, Mc-99, I-131, Cs-134, Cs-137, Cc-141, and Cc-144 in iodine and particulate releases. This list does not mean that only these nuclides are to be I considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reponed in the Semiannual Radioactive EfTluent Rc! case Report pursuant to Technical Specification 5.9.1.7, in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974. (3) Sampling and analysis sha'l also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within I hour period. (4) Tritium grab sampics shall be taken and analyzed at least once per 24 hours when the refueling canal is flooded. (5) Grab samples need to be taken only when spent fuel is in the spent fuel pool. (6) The ratio of the sample flow rate for the sampled stream flow rate shall be knowTi for the time I period covered by each dose or dose rate calculation made in accordance with PIC 9.6.1.1, 9.7.1.1, and 9.8.1.1. (7) Samples shall be changed at least once per 7 days and analyses shall be completed within 48 I hours after changing, or after removal from sampler. For unit vent, sampling shall also be performed at least once per 24 hours for at least 7 days following each shutdown, STARTUP or THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour I period and analyses shall be completed within 48 hours of changing. When samples collected for 24 hours are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement does not apply if: '(1) analysis shows that the DOSE EQUIVALENT I-131 I concentration in the reactor coolant has not increased more than a factor of 3, and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3. (8) Continuous sampling of the spent fuel building eOtaust needs to be perfonned only when spent fuel is the spent fuel pool. l (9) Samples shall be representative of the efDuent release. I I I I -104-

I S

APA ZZ4)l003 Rev.3 'I i 9.7 I>OSE - NOBLE G ASES l9.7.1 Contrpls (CTS # 41834) l 9.7,1.1 The air dose due to noble gases released in gascous emuents, from each unit, to areas at and j beyond the SITE BOUNDARY (see Technical Specificadons Figure 5.1-3) shall be limited to the I following: i a. During any calendar quarter: Less tha i or equal to 5 mrads for gamma radiadon and lets than or equal to 10 mrads for beta radiadon, and b. During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mras for beta radiadon. 6PPLICABILITY: At all times. l ACTION: (CTS # 1161) With the calculated air dose from radioacdve noble gases in gaseous emuents exceeding a. any of the above limits, prepare and submit to the Commission within 30 days, pursuant to i Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the correcdve a: dons that have been taken to reduce the releases and the proposed corrective aedons to be taken to assure that subsequent releases will be in compliance with the above limits. i l b. The prousions of Techrical Specificadons 3.0.3 and 3.0.4 are not applicable. l9.7.2 Surveillance Reauirements I 9.7.2.1 Cumulative dose contribudons 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. 4 I I I I I -105-

I APA-22-01003 Rev.3 9.8 DOSE - LODINE-131 ANDl3.1 TRITIUM. AND R ADIO&CTIVE M ATEP1 AL IN I PARTICULATE FORM l9.8,1 Controls (CTS # 41834) l 9.8.L1 The dose to a MEMBER OF TIIE PUBLIC from lodinc-131 and 133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous efIluents released, from each unit, to areas at and beyond the SI m BOUNDARY (see Technical Specifications, I Figure 5.1-3) shall be lintiled to the following: a Dunng any calendar quancr: Less than or equal to 7 5 mrems to any organ, and b. During any calendar ye.,r: Lers than or equal to 15 mrems to any organ. APPLICABILrn': At all times. I l ACTION: (CTS # 1161) I a. With the calculated dose from the reicase ofIodine-131 and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous efIluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to I Technical Specification 6.9.2, a Special Repon that identifics the cause(s) for exceeding the limits and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. I l b. The provisions of Technical Speci.fications 3.0.3 and 3.0.4 are not applicable. l9.8.2 Surveillance Requiremenn 9.8.2.1 Cumulative dose contributions for the current calendar quaner and current calendar year for lodine-131 and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 I days shall be determined in accordance with the methodology and parameters in the ODCM at ) least once per 31 days. I I I I-I I 4 -106-l

I APA ZZ-01003 Rev. 3 9.9 G ASEOUS RADWASTE TREATMENT SYSTEM l9.9.1 Contrels (CTS # 41834) l 9.9.1.1 The VElfffLAT10N EXHAUST TREATMENT SYSTEM and the WASTE GAS HOLDUP SYSTEM shall bc OPERABLE and appropriate pordons of these systems shall be used to reduce i releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases, ,g from each unit, to areas at and beyond the SITE BOUNDARY (see Figure Technical g Specification's 5.1-3) would exceed: i a. 0.2 mrad to air from gamma radiadon, or b. 0.4 mrad ta air from t<ta radiation, or i c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC. i j APPLICABILITY: At all times A_CTION: l a. With radioacuve gaseous waste being discharged without treatment and in excess of the j above limits, prepar e and submit to the Commission within 30 days, pursuant to Technical j Specificadons 6 9.2, a Special Report that includes the following information-l 1 L) Identification of any inoperable equipment or subsystems, and the reason for the I inoperability, 2.) Action (s) taken to restore the inoperable equipment to OPERABLE status, and 3.) Summary description cf action (s) taken to prevent a recurrence. l b. The provision of Technical Specifications 3.0.3 and 3.0.4 are not applicable. 9.9.2 Surveillance Reauirements I 9.9.2.1 Doses due to gaseous releases from each unit to areas at and beyond the SITE POUNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in the l ODCM wben Gaseous Radwaste Treatment Systems are not being fully udlized. 9.9.2.2 The installed VENTILATION EXHAUST TREATMElfr SYSTEM and the WASTE GAS liOLDUP SYSTEMS shall be considered OPERABLE by mecung REC 9.6.1.1 and 9.7.1.1 or 9.8.1.1. I I I -107-

1 I l t APA.ZZ41003 Rev.3 9.10 TOTAL DOSE I l 9.10.1 Comrols (CTS # 41834) I 9.10.1.1 The annual (:alendar 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. APPL 1CB31LITY: At all times. ACTION: With the calculated doses from the release of radioactive materials in liquid or gaseous a. ellluents exceeding twice the limits of REC 9.4.1.la, 9.4.1.lb, 9.7.1.la, 9.7.1,1b, 9.8.1.1a, I or 9.8.1.lb, calculadons should be made including direct radiadon contnbudons from the units and from outside storage tanks to determine w hether the above limits of REC 9.10.1.1 have been exceeded. If such is the case, prepare and submit to the Comndssion within 30 I days, pursuant to Technical 5;aiEcadon 6.9.2, a Special Report that dc0nes the correedve 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 j Special Report, as defined in 10 CFR 20.2203, shall include an analysis t'.at estimates the 'I radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all efIluent pathways and direct radiation, for the calendar year that i includes the release (s) cos cred by this report. It shall also describe levels of radiation and I concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the esumated 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 i I Special Report shall include a request for a variance in accordance with the prosisions 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 SpeciEcations 3.0.3 and 3.0.4 are not applicable. 9.10.2 Syrveilla_nce Requirements 9.10 2.1 Cumulative dose contribudons from liquid and gaseous efnuents shall be deterndeed in accordance with REC 9.4.2.1, 9.7.2.1, and 9.8.2.1, and in accordance with tl.e mea.odology and parameters in the ODCM. I 9.10.2.2 Cumuladve dose contributions from direct radiadon from the units and from radwaste storage tanks shall be determir ed in accordance with the methodology and parameters in the ODCM. This requirements is applicable only under condidons set forth in ACTION a. of REC 9.10.1.1. I I I

T APA-ZZ-01003 I Rcv. 3 9.11 RADIOLOGICAL ENVTRQNMENTAL MONTTORING PROGRAM l 9.11.1 Controls (CTSn 41835) I 9.11.1.1 The Radiological Environment Monitoring Program shall be conducted as specified in Table 9.11-A. APPLICABILITY: At all tirnes. I ACTION: I a. With the Radiological Emironmental Monitoring Program not being conducted as speciDed in Table 9.4-A, prepare and submit to the Commission, in the Annual Radiological l Emironmental Operating Report required by Technical Specification 6.9.1.6, a description I of the reasons for not conducting the program as required and the plans for preventing a recurrence. b. With the level of radioactivity as the result of plant emuents in an emironmental sampling medium at a specified location exceeding the reponing levels of Table 9.11-B when averaged over any calendar quancr, prepare and submit to the Commission within 30 days, pursuant to Technical Speci5 cation 6.9.2, a Special Report that identifies the cause(s) for I emuents so that the potential annual dose

• to a hEMBER OF TIE PUBLIC is less than exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive the calendar year limits of REC 9.4.1.1,9.7.1.1, or 9.8.1.1, When more than one of the radionuclides in Table 9.11-B are detected in the sampling medium, this report shall be I

submitted if: I concentration (1) concentration (2) + + 2 1.0 reponing level (1) reporting (2) I When radionuclides other than those in Table 9.11-B are detected and are the result of plant emuents, this report shall be submitted if the potential annual dose

• to A LEhGER OF I

l REC 9.4.1.1. 9.7.1.1 or 9.8.1.1. This report is not required if the measured level of TIE PUBLIC from all radionuclides is equal to or greater than the calendar year limits of radioactivity was not the result of plant emuents; however,in such an event, the condition shall be reported and described in the Annual Radiological Emironmental Operating - I l Report, required by Technical Specification 6.9.1.6. I I l The methodology and parameters used to estimate the potential annual dose to a hEhBER OF THE PUBLIC sl.all be indicated in this report. -109- . I W--- r --- a a

I APA 7241003 Rev. 3 With milk or fresh leafy vegetable samples unavailable from one or more of the sampic c. locadens required by Table 9.11-A, identify specine locations for obtaining replacement samples and add them within 30 days to the Radiological Environmental Monitoring Program given in the ODCM." The specific locations from which samples were unavailable may then be deleted from the monitoring program. Pursuant to Technical Specification 6.14, submit as part of, or concurrent with, the next Semiannual Radioactive Efiluent Release Report a complete and legible copy of the entire ODCM, including the revised figure (s) and table reflecdng the new location (s) with supporting infonnation identifying the cause of the unavailability of samples andjustifying the selection of new locadon(s) for obtaining samples. lI d. When LLDs specined in Table 9.11-C are unachievable due to uncontrollable circumstances, (such as background fluctuadons, unavailable small sample sizes, the presence of interfering nuclides, etc.) the contributing factors shall be identined and described in the Annual Radiological Emirnnmental Operaung Report. 1 l The provisions of Technical Specincadons 3.0.3 and 3.0.4 are not appheable. c. l 9.11.2 Surveillance Requirements 9.11.2.1 The radiological emironmental monitoring samples shall be collected pursuant to Table 9.11 A I from the specific locadons given in the table and figure (s) in the ODCM, and shall be analyzed pursuam to the requirements of Table 9.ll-A and the detecdon capabilities required by Table 9.11 -C. I I . I I I I l " Excludmg short term or temporary unavailabihty. I 1 I -110-

.... - ~. - - - - - _ - -. m m m m. m m m m M M M E - l l APA-ZZ-01003 Rev.3 TABLE 9.11-A RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NUMBER OF REPRESENTATIVE SAMPLES SAMPLING AND EXPOSURE FATilWAY AND SAMPLE COLLECTION TYPE AND TREQUT.NCY M y-UR S Alf_P.LE !_dCATICNS(I) Tf M UENCY Of ANAL)1ij

1. Direct Rad,atims(2)

Forty routine monitoring stations either with two or mne Quec ty Garnma dm sparerly desimeters or with one instnmwnt for measuring and remding dose rate enedieuously, placed as followr An ince rh of sixteen stations, one in each meicerofagical scoor in the g ncrsi area cf the SITE DOUNDARY; An cuter ring of stations, one in each meteorological sector in the 6-to R-km (3 to 3 mile) range from the site; and E ght statens to be pinced in special interest areas such as population centers. nearby residences, sdets, and in one er two areas te serve as control stantes.

2. Nrborne Ha,1.ohtme and Samples frorn five locatinns; Continuous sampler eperation wM senpie Padic.indine Canister: 1-13I ans1pis weekly ruiadates co!!cction weekly, or nwe frequently if required lhree sacipics from close to the dvec SITE DOUNDARY by Just leading.

Particutare Sampig Grees beta radioactinty Iccations,in di!Terent sectors, of tbc highest calcula:ed annual analpis fellowing fi'ter change:(4) aM garnma average ground icvet D1Q. ive ie anatpis (3) of ecmpnite (by Ixs1 ion) r i qusticely. One sample from the vicinity of a comrnunity having the Lighest calculated armel average ground Icvet DQ. Om sample frem a cordrollocatiert, as fee exanple 15 to 30 bn (10 to 20 mile) distant and in the levt prevalent wind direction (3). I i -1Il-

M M M M M M M E 'E E E E E t APA-ZZ-01003 Rev.3 TABLE 9.11-A (Cont'd) i RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NUMDER OF R EPRESENTATIVE SAMTI.ES SAMPLfNG AND E.NPOSURE PATitWAY AND SAMPLE COLIIC1 TON TYPE AND FREQUENCY AND/Olt S AMPLE LOCATIONSW [R_r,r2UJ.NCY OF ANALYSIS - 1 3 Waletl=wne

a. Surface [')

One umple uMream Composite sample over Osmma iactT cd) and tritmen i One sample dowmtscam ianonth pakwl(71 ansipin wth?y t b thinking One naryte of each of one to three of the nearest water supplies Composite ssqle over 1-13I analysis m each evenpewite w+ct tw dme within lC rnes doveticarn that could be sffeded by its 2-week period (') when 1.131 ana!pis is calculated for the cmumption of the wa'er is discharge. pafe med, nmnthly cornprile otheruise. greater than I mrem per year (*). Comtwsite for gress twta and ganwne inntarie anahme,(I) One unyle from a control location nmthly. Compwite for tritham ansipis 1 m uter ly. t

c. Sed ment finen One nample froen dowmtream area with esisting or pntentini Semiannually Ganwns isotopic antipis(?) semiannuait<

{ sheveline recreatinent value I e

4. }DRC$tktM1

n. Mdk Sampics from mifking animals in three difTerent metentogical Seminumthly when animsts are nn pasture.

Gamma iactopic[) and !-131 analpis i sedors witEn 3 km(3 mile)dierance having the highe=t dose monthly at other times seminymtJJy uhen animals are nn pasture: potential If there are nnne,then ene sainple frnen milking nmth!y at other times ] animals in each of three different metentlogical sedar-between 3 to 5 h m (3 to 5 mite) datance where dmes are calculated to be i greater than I mrem per yr. O'x nampte from mi! king animals at a contrcl hwation,13 to 30 km (in so 20 milc) distance saf in ti e Icast prevntent wind directim -112-1 m .. 1.. A-

e e m m m m m m m' m m M m APA-7E01003 Rev.3 TADl.E 9.ll-A (Cont'd) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NUMf3ER OF REPRt7ENTAllVE 5 AMr!.ES SANtri.lNO AND EXPOSUR C PAT!!WAY ANDSAMPII CO!.IJCTION DTE AND FREOtINCY f ANirOR SARIJ 1.OCATirjNs(I) FR EOUENCY OF ANAL.YSIS -1. Ingestion (Conrd) i

h. I ish One sample of each enmmercistly and recreationatly impariant Sample in senvs or smdannustly if they are Omem iMaric ana!pi% ) en e4ble pmtions 3

species in vicinity of plant discharge area. n<< seasonal One sample of smw species in areas not influenced by plant dAharge. t

c. Food prmfuds One sample of each principal clus ofibed products frnen any aren At time of harvest (9x)

Gamme iwtmic analpis (3)on cable pc tion 10 that is irrigated by water in which lapdd plant wasses have been discharged. Samples orth:ec 6!T-rent kim!= cf broad leaf vegetation if Gamma juryic (3) and I 1.11 analysis l evmilable grown nenrest carh of two &fTerent ofNte locations of Monthly uhen available I highest predided annual average ground level D'Q if mi:k sampling is not performed. 3 One sample of each of the similar broad testvcFetation ymm 15 to 30 km (10 to 20 mile) distant in the teso preva! cut wind Monthly when avastable Gamma isotopic (3) and I-1}! anafpis direction if milk sampling is not performed. i t 113-j

ApA-ZZ-01003 TABlJJl1 A (Continued) TABLE NOTATIONS (1) Specific parameters of distance and direction sector from the centerline of one unit, and additional description where pertinent, shall be prosided for each and every sample location in Table 9.ll A in a table and figure (s) in the ODCM. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automade sampling eqtupment, and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every effort shall be made to complete correcnvc action prior to the end of the next scmpting period. All deviations from the sampling schedule shall be documented I in the Annual Radiological Environmental Operating Repon pursuant to Technical Specification 6.9.1.6. (CTS # 2804) It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of I the media of choice at the most desired location or time. In these instances suitable specific alternative media and locations may be chosen for tL particular pathway in quesdon and appropriate substitutions made within 30 days in the Radiolo6ical Environniental Monitoring Program gis en in the ODCM. Purst. ant to Specification 6.14, submit in the next Semiannual I Radioactive EfIluent Release Repon documentation for a change in the ODCM including the revised figure (s) and table reflecting the new locadon(s) with supponing information identifying the cause of the unavailability of samples for that pathway andjustifying the selection of the new location (s) for obtaining samples. (2) One or rnore instruments, such as a pressurized ion chamber, for measuring and recording dose rate i I continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of i this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are cortsidered as two or more dosimeters. Film badges shall not be used as ] dosimeters for measuring direct radiation. The number of direct radiadon 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 optimem dose information with minimal fading (3) The purpose of this sample is to obtain backgiound informadon. Ifit is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites that provide I valid background data may be substituted. (4) Airborne paniculate sample filters shall be analyzed for gross beta radioactisity 24 hours or more I. after sampling to allow for radon and thoron daughter decay. If gross beta aedsity in air paniculate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the indisidual samples. (5) Garruna isotopic analysis means the identification and quantification of gamma-endtting radionuclides that may be attributable to the efIluents from the facility. (6) The

• upstream sample" shall be taken at a distance beyond significant inCuence of the discharge.

The

• downstream" sample shall be taken in an area beyond but near the mixing zone.

(7) In this program, composite sample aliquots shall be collected at time intervals that are very shon (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample. I -114-I

APA-ZZ-01003 Rev.3 TABLE 9.11-A (Continued) TABLE NOTATIONS l (8) Groundwa:er samples shall be taken when this source is tapped for drinking or irrigation purposes in W areas where the hydraulic gradient or recharge properties are suitable for contandnation. (9) The dose shall be calculated for the minimum organ and age group, using the methodology and parameters in the ODCM. (10) If harvest occurs more than once a year, sampling shall be performed during each discrete han est. If harvest occus continuously, sampling shall te monthly Attention shall be paid to including samples of tuberous and root food products. I ) I I I I I -115-

m m m, m m-M M M M-M M M M M M M~ APA-ZZ-01003 Rev.3 TABLE 9.11-B REPORTING LEVELS FOR RADLOACTIVITY CONCEBTTPATIONS IN ENVIRONMFETAL S AMPLES REPORTING LEVELS ANALYSIS WAT ER AIRliORNEl'ARTICU{ ATE F151f MILK gp[ ya OR GMES(pCi/m ) (pCiA g. wet)h TOODPRODUpS t 3 .If 4 3

x. ~,..

v, 20.000 y.m._. 7 -n,.7 -3., , m m s.yn. , ~--y.. a. yg..y q.7.g..xy. q., m_.1,.. .;,.s.,,._ g-n.ny 1.,000. Mn.34. _c d

,.'b

.c .400 v., 'll:a n g., g.,30. 000. w ~;, ,n, ~ y a. .:.y +, ,c. .,g .~m.n~~.e_ .,g.,, x 1L...- ~. Te.59 I m,.:sn g e:ew ~v 10.tM0 ~ ~~m~.,.: - - ~, ~, vw - w-m :.-:. ..5Mr

.; Ji'.

). i:L Sv ': ; e s .. m g,. n r ~ m s, .....y.3 ,,yy...~m.m y m xCo-60 .. '. k 'h c. 2.y, sp r e, y y,.7mm s- .,c e . wg.y~ ..,~.f .,r-N 4., y A . ;.d y 'b d',, .v / a.. . A10.000 .c ./-.. 300 ~- 3 m~n c~ymyg-p;s - ~ amy::+;x:pm

y~mw-; m 3 rm ~~;.

m ~<~~~mw

. -... ni w. wy v. ~ <.e mm-~m-- ~cmec., .7.r-Nb-?$ , ~

a

....s ~. ~..

w< %;2.

s u% N ~d' , ;d.% r:si t:;, ~ ^h. 400,;.u ~.e.- .y -. m;my gr y.n ,y.y,.g :p grn .v.,n;-,y

y n.g.

v - n::

m. n~w.y y, -

g., ~ 0 9 m. g p g., n-~my :n.:..,

..a. -...

p.m n7, y --r. ..D + 1-I 31. ~;p.s, m,,. -,mry,

v. m J.+w

.:a ^ --r.M w.u m. .w,.; - : x-2 1 m..,.,..~.ue..,.

c..,r.;. D.4.,-; y; ^

.w.;. ; ,p.- . w.~.y a-.3 100 I v 'M .L 4% -' r-,.mg

yrv.

J-.'s, %,2 ^:w, v a. - --.yg.~.w , ~, ,v v, Ct-134 a g ~,.,. g,,,10g.n. :.m. 7,,._m.n: gy,.; 1,000. .. 60.... 3 .nym,; e,~ r -.0 , v; 1,.M3 < r~e kan..u ^ , r o _ -e.w m, m m Cs 137~...v s <......,. 30 ...c, c P;u 5O.0.:... ., b,C. x .~70- .,2,000 - + c-20 2.000 - ~mm, v :gy m y-.nw., - m.~.m ~, zry nw. - -~ rm e:g ~~n . ;y';

,;.2w

. ~...i ;-., ,.<1",.

w.'e a+

~....~r.v wy -~.- v s1 !!>La-140 200 300 t {a) Multiply the values in thh 12 hie by 1E-9 to convert to units orpCUml. (h) Mohiply the values in this table by 1E-9 to comert to umts of pCUg. I . For drinking water samples. This is 40 CFR Part 141 value. For surface water sampfes, a value of 30,000 pCi/C may be used. Total activity, parent plus daughter activity. -116-

- M M M M. M M Q M M-M Mi & M M M M O Y APA-ZZ-01003 Rcn 3 i TABLE 9.11-C DETECTION CAPADILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS LOWER LIMIT OF DETECTION (LLD) [l),(2), (3) i ANALYSIS SURFACE DRINKING AIRDORNE FISH MILK FOOD PRODL{)(pcia g, dry)b S SEDIMENT WATER WATER PARTICUIATE (pCi h. wet)b (pCu[ )* (PCM g. wrt 3 @Ci'[ )* (pCU( )s OR GASI S (tGm) Gro- n, .y ,es, y.,. 3 o.a - ^ .q.,.. x_.. m,,.,, .._.c . ~.,,.

xa.:e

-m.15 n.. .n. s i u.,' s s ? Mn-54 15

m.,130 y,,wy m7,, _

c 2. wgsg.y., r.3m,- .~7.<., . m., s s a, ..m.:.. > s ,...y.y.,-,,.-~ %:n n.

ni...., s es.

w s.

a Fe-59,..,g:e., ...,30,,. .n n.< 1 30 ,26,0- .~,,n-, ? y n., n. ~ w L '.w:n,.L L< ^^ e.,,,,.nm..n.,,.., w.,, , n : ' ' s:.i . :.3 i ^' a+ ^~ .,..n.. .,. ~ - Co-3 R.60 15 .a.n c ' s ; i-15 130.... vr. y~avere g:.sw:w:v;::m.,m:~e w y ecmmyg.m-rn u.~;. ~ ~.m.ng ;g:.g :;mge.;.

v..my~ v.y; m,y v.n-

-~.r.- ,.~ ..a-.~ . I5 ~ m ...c n w1:

,

2.:

a u: nw s r ~ ~.a vrr. ~.c, e 7.r,N,..b.9 5 c. I5 4 .v n n n um.~.ny. e r.wyn ;. rv,gmn.--n ~~.a. ^ , w.-.. g ~ ~ p

e,

m,,~y., ~ m.-.w ~.g-em, w - ve-. e ,:-m - w, n.. . ~, < m ):%. .1.* t.G d ~. :w.1. c. n a.3 x L, '.E..s ? n l ,,e I-til 1000 t <: ~.. ~.,1 0.07 t 60 ..r.r- - ~.. g.nnyn:.vy - .s w, Cs-134 15 ...15.. ~.

• rgr ~v;m we;;m g~.m. w,.:, my-m..

w.., y., 7;-n, -- e.w m,~ ~ - ~ ~. - w.a.....,., - +

aa: n.. ~..

c.v: - s .a ., ygg, 3. s~..z.gy.c, g m.m.~, 0.05,, g- ~..... .. w. .,...60 u - - -. ~y e ~ 130 15 ....130 l ~y g.g:.eg7.yy. cgm. gy _ se

n...

.y::y.:; Cs-I37 - I8 ~ .m .g v.; n. I8 0.06 150 18 ......p .....120 7 20 .gm mc y.ym:grn:p.~wmm~.n pr:7;m~mp,.e s, Ila-I+140 15 n... s -K.s-W a m:,7.~w y my: + mm- ~ vwr n~mm m n: ~. ~ rxew m m- . nd.,,,,. e .:+ V 5 ~ .n. 15 13 (a) Mu!6p!y the values in this table try 1 E-9 to convert to units of pCUmL (b) Multiply the values m this table by IE-9 to convert to smits crpti/g. t. l .. Tctal activity, parent plus daughter activity. L t i r b -117-t 1

. _ = J APA-ZZ-01003

g Rev. 3 ig TABLE 9.11-C (Continued) 1 IABLE NOTATIONS (1) This list does not mean that only these nuclides are to be considered. Other peaks that are lI identifiable, together with those of the listed nuclides, shall also be analyzed imd reported in the Annual Radiological Ensitorunental Operating Report.

4 (2) Required detection capabilities for thermoluminescent dosimeters used for emironmental measurements shall tx in accordance with the recommendations of Regulatory Guide 4.13, Revision 1, July 1977 a l l (3) The LLD is descrikd in Attachment 1.

I J

'!g a e !g I I . I I i 1 I .lle.

I APA 22-01003 Rev. 3 9.12 RADIOkQQlCAL ENVIRONMENTAL MONITORING LAND USE CENSUS l 9.12.1 Controls (CTS *41S3') I 9.12.1.1 A Land Use Census shall be conducted and shall identify within a distance of 8 km (5 unles) die location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence and the nearest garden

• of greater than 50m (300 rt ) producing broad leaf,egetation.

2 2 APPL.lCABILITY: At all times. AQ.QB a. With a Land Use Census identifying a locadon(s) that y:cids a calculated dose or dose commitment greater than the values currendy being calculated in REr' 9.8.2.1, identify the new locadon(s) in the next Semiannual Radioactive EfDuent Release Repon, pursuant to Technical Specincadon 6.9.1.7 b. Widi a Land Use Census identifying a locatior s) that yields a calculated dose or dose s l commitment (via the same exposure pathw :y) 20% greater than at a location from which m' samples are currently being obtained in accordance with REC 9.11.1.1, add the new location (s) withia 30 days to the Radiological Emironmental Monitoring Program given in the ODCM. The sampling location (s), excluding the control station location, having the I lowest calculated dose or de.w commitment (s), via the same exposure pathw2y, may be deleted from this monitoring program after October 31 of the year in which this Land Use Census was conducted. Pursuant to Technical Specification 6.14, submit as part of, or ' l concurrent with, the next Semiannual Radioactive EfDuent Release Repon a complete and legible copy of the entire ODCM, including the revised figure (s) and table (s) reflecting the = new location (s) with information supporting the change in sampling locations. c. The provisions of Technical Specificadons 3.0.3 and 3.0.4 are not applicable. l 9.12.2 Surveillance Reauiremen_t1 5 9.12.2.1 The l2nd Use Census shan be conducted during the growing season at least once per 12 months using that information which will provide the best rest'lts, such as, but not limited te, I door-to-door survey, aerial survey, or by consulting local agriculture authorides and/or residents. The results of the Land Use Census shall be included in the Annual Radiological Emironmental l Operating Report pursuant to Technical SpeciDeation 6.9.1.6. I i

• Broad leaf vegetadon sampling of at least three difTerent kinds of vegetation may b perform:d at the

[lW SITE BOUNDARY in each to two different direction sectors with the highest predicted D/Q's in lieu of the ga* den census. Specifications for broad leaf vegetation sampling in Table 9.ll-A, Part 4.c shall be - followed, including analysis of control samples. -119 I 4

-= --.-.-w.., ..w._ ..a- +. -- u. a I APA 72411003 Rev.3 9.13 BAplOLQQLCAJJNVIRONMFNTAL MONITORING INTERLABOR ATORY A I COMPARISON PROGRAhj l 9.13.1 Controh (CTS # 41S35) 9.13.1.1 Analyses shall be performed o's radioactive materials supplied as part of an Interlaboratory l Comparison Program that hr, been approved by the USNRC. APPLICABILIT* At all dmes. ACTION: l Wi0i analyses net being performed as required above, repon the corrective actions taken to a. prevent a recurrence to the Commission ir. The Annual Radiological Environmenta! l Operating Report purruan: to Technical Specification 6.9.1.6. l b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable. l 9.13.2 Eurveillance Reguirements l 9.13.2.1 The Interlaboratory Comparison Program shall be described in the ODCM. A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall bc a included in the Annual Radiological Emironmental Operat.ing Report pursuant to Technical Specification 6.9.1.6. I I E I I I E I I .m.

==.+

I APA ZZ-01003 Rev.3 10.0 ADMINISTRATIVE CONTROLS I 10.1 MAJOR CHANGES TO LIQUID AND GASEOUS RADWASTE TREATMENT SYSTEMS.* 10.1.1 Licensee-initiated major changes to the Radwaste Treatment Systems (liquid and gascoue.): a. Shall be reported to the Comndssion in the Semiannual Radioactive Efiluent Release Report for the period in which the evaluation was reviewed by the On-Site Resicw i Committee (ORC). The discussion of each change shall contain: 1.) A summary of the evaluation that led io 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 efIluents that differ from those previously predicted in the Licer se application and amendments thereto; 5.) An evaluadon of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the LTriRESTRICTED AREA and to the general I population that differ from thase y reviously estimated in the License application and amendments thereto; I 6.) A comparison of the predicted releases of radioactive materials, in liquid and gaseous efDuents, to the actual releases for the p:riod prior to when the changes are to be nude; 7.) An estimate of the exposure to plant operating personnel as a result of the change; and 8.) Documentation of the fact that the change was reviewed and found acceptable by the ORC. b. Shall become effective upon review and approval by the ORC and in accordance with Technical Specification 6.5.3.1. 10.2 QiANGES TO THE OFFSITE DOSE CA1.,C. ULATION MANUAllODCM)(CTS # 2815) I 10.2.1 All changes to the ODCM shall be completed pursuant to Technical Specification 6.14 and approved as per APA-ZZ@l01, ' Preparation, Review, Approval and Control of Procedures". I 10.2.1.1 All changes shall tx: approved by the ORC prior to implementation.

• Union Electric Co. may choose to submit the information called for in this control as part of the annual FS AR update.

I .n,.

I APA.22-01003 Rev.3 10.2.2 Cross Disciplinary Resiew for each resision of the ODCM must include, as a minimum, the I l Health Physics, Quality Assurance, and Licensing and Fuels 16diological Engir.ccring Departments. I 10.2.3 A complete and legible copy of each revision of the ODCM that became effective during the last semiannual period shall be submitted as a part of, or concurrent with that periods Semiannual Radioactive EfIluent Release Report pursuant to Technical Specification 6.14 I I I -1 I I I E 4 g t i I I I I .m.

!'I. l-APA ZZ-01003 Rev.3 l f 11.0 [GFERENCF,3 il 11.1 Title 10. " Energy", Chapter 1, Code of Federal Regulations, Part 20; U.S. Government Printing l Omce, Washington, D.C. 20402. I1.1.1 Statements of Consideration, Federal Register, Vol. 56, No. 98, Tuesday, May 21,1991, Subpart D, page 23374. I l 11.2 Title 10. " Energy *, Chapter 1. Code of Federal Regulations, Part 50, Appendix 1; U.S. j Government Printing OfIice, Washington, D.C. 20402. 1 l11.2.1 10 CFR 50.36 a (b) 'I 11.3 Title 40, " Protection of Environment", Chapter 1, Code of Federal Regulations, Part 190, U.S. i Government Print Omcc, Washington, D.C. 20402. i I j 11.4 U.S. Nuclear Regulatory Commission, " Technical Specifications Callaway Plant, Unit NO.1", } NUREG 1038 (Rev.1), October 1984 I1.4.1 Section 6.8.1 (CTS # 2791) i !!.4.2 Section 6.8.4f (CTS # 41834) 'l 11.5 Cornmunications 4 i i 11.5.1 Letter NEO-54, D. W. Capone to S E. Mittenberger, dated January 5,1983; Union Electric j Company correspondence. 11.5.2 Letter BLUE 1285, "Callaway Annual Average X/Q and D/Q Values", J. H. Smith (Bechtel 8 Power Corporation), to D. W. Capone (Union Electric Co.), dated February 27,1984. ,i l } l1.5.3 Letter BLUE 1232, "Callaway Annual Average X/Q Values and "S" Values", J. Il Smith (Bechtel Power Corporation) to D. W. Capone (Union Electric Co.), dated February 9,1984 l l 11.5.4 Reference Deleted i 11.5.5 Private Communication, H. C. Lindeman &. B F. Holderness, August 6,1986 l 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 I l 11 6.5 Section 11.5 2.2 3.3 l i1.6.6 Sc~ tion 112.314 I e.

-~ APA-ZZ-01003 Rev.3 ! EB i l1.6.7 Section i1.2.3 4.3 i 11.6.8 Section i1.5.2.3.3,i 11,6.9 Section 11.5.2.3.3.2 i I1.6.10 Section 11.5.2.3.2.3 11.6.I1 Section 11.5.2.3.2.2 I 11.6.12 Section 2.3.5 11.6.13 Section 2.3.3.2.1.2 l 11.6.14 Section 9.2.6 1 j 11,6.15 Section 9.2.7.2.1 11.6.16 Section 6.3.2.2 i 11.6.17 Table 11.1-6 i I l 11.6.18 Deleted l 11.6.19 Deleted l 11.6.20 Deleted l 11.6.21 Deleted I l 11.6.22 Table 2.3-68 11.7 Urdon Electric Cornpany Callaway Plant Erwiwnmental Report, Operating License Stage. I1.7.1 Table 2.1 19 l ~ 11.7.2 Section 2.1.2.3 11.7.3 Section 2.1.3.1.4 11.7.4 Sectien 5.2.4.1 j I1.7.5 Table 2.1-l9 i 1.8 U.S Nuclear Regulatory Commission, Preparation of Radiological Efiluent Technical I~ Specification for Nuclear Power Plants", USNRC NUREG-0133, Washtngton, D. C. 20555, October 1978. I1.8.1 Pages AA-! throu;h AA 3 11.8.2 Section 5.3. l.3 i I 124

I APA-ZZ-01003 Rev.3 .I 11.8.3 Section 4.3 11.8.4 Section 5.3.1.5 11.8.5 Section 5.1.1 I 11.8.6 Section 5.1.2 11.8.7 Section 5.2.1 11.8.8 Secdon 5.2.1.] 11.8.9 Section 5.3.1 11.8.10 Secdon 3.8 11.8.11 Section 3.3 I l Roudne Emuent Releases at Nuclear Power Stations", USNRC NUREG-0324, 11.9 U.S. Nuclear Regulatory Commission, "XOQDOQ, Program For the Meterological Evaluation of Washington, D. C. 20555. I1.9.1 Pages 19 20 Subroutine PURGE I1.10 Regulatory Guide 1.111, " Methods for Estimating Atinospheric Transport and Dispersion of I Gaseous Emuents in Routine Releases from Light-Water-Cooled Reactors", Revision 1, U. S. Nuclear Regulatory Commission, Washington. D. D. 20555, July,1977. 11.10.1 Section c.).b I 11.10.2 Figures 7 threugh 10 11.10.3 Section c.4 11.11 Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Emuents for the Purposes of Evaluating Compliance with 10 CFR Part 50, Appendix 1", ReTision i 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 l 11.11.3 Appendix C, Section 1 I. I1.11.4 Appendix E, Table E-11 11.11.5 Appendix E, Table E-9 11.12 U. S. Nuclear Regulatory Comndssion,

• Methods for Demonstrating LWR Compliance with the I

EPA Uranium Fuel Cycle Standard (40 CFR Part 190)", USNRC NUREG43543, Washington, D. C. 20555, Janu.ny 1960 11.12.1 Section I, Page 2 -125-I

I APA-Z.Z-01003 I Rev.3 11.12.2 Section IV, Page 8 11.12.3 Section IV, Page 9 11.12.4 Section Ili, Page 6 11.12.5 Section III, Page 8 I 11.13 Management Agreement for the Public Use of Lands, Union Electric Company and the State of Misorri Depanment of Conservadon, December 21,1982. I1.13 Exhibit A 11.14 Miscellaneous Refercnces 11.14.1 Drawing Number M-109-0007-06, Revision 5 11.14.2 Callaway Plant Annual Environmental Operating Report (updated arumally) l 11.14.3 UE Safety Analysis Calculadon 87-001-00 11.14.4 Calculation ZZ-48, " Calculation of inhalation and Ingestion Dose Commitment Factors for the Adult and Child", January,1988 l 11.14.5 HPCI 89-02, " Calculation of ODCM Dose Commitment Factors", March,1989 11.14.6 HPCI 87-04, " Calculation of the Limiting Setpoint for the Containment Purge Exhaust Monitors, GT-RE-22 and GT RE-33", March,1987 l. f 11.14.7 HPCI 88-10, " Methodology for Calculating the Response of Gross NaI(TI) Monitors to Liquid Etlluent Streams", June,1988 f 11.14.8 Calculation ZZ-57, " Dose Factors for Eu-154", January,1989 i ! E ' ^ ' * " ' " ' ' " z " ' ' '"""'^d""^8 ' " " ~ ' " ' ' ' " - !W 11.14.10 HPCI 88-OS. " Performance Testing cf the Environment TLD System at Callaway PlaM", August,1989, ij W l1.14.11 Catculation ZZ-250, Rev. I "ODCM Gaseous Pathway Dose Factors for Child Age Group and Ground Plane Dosc Factors". l 11.15 U. S. Nuclear Regulatory Commission, "XOQDOQ: Computer Program for the Meterological Evaluation of Routine Effluent Reicases at Nuclear Power Stations", USNRC NUREG/CR-ll 2919. September 1982, Washington, D. C. 20555 t ,. m 11.15.1 Section 4. " Subroutine PURGE *, pages 27 and 28 [g 11.16 Regulatory Guide 4.13, "Perfonnance. Testing, and procedunt specifications for 3 Thermoluminiscence Dosimetry: Erwironmental Applications "(Revision 1), July 1977; USNRC, Washington, D. C. 20555 !I i t -126-

I s.

i

~ 1 APA-ZZ-01003 g Rev,3 2 kE i j 11.17 TID-7004, " Reactor Shielding Design Manual", Rockwell, Theodore, ed; March 1956. l 11 18 BNWL-2%, *lSOSilLD - A computer code for General Purpose isotope Shiciding Analysis", Engel, R. C., Greenberg, J., liendrichson, M. M.; June 1966 11.19 BNWL-236, Supplement I, "lSOSHLD. II: Code Revision to include calculation of Dose Rate from Shielded Bremstrahlung Sourecs", Simmons, G. L., et al; March 1967 l 11.20 BNWL-236, Supplement 2. "A Revised Photon Probability Library for use with ISOSHLD-i= 111", Mansius, C. A.; April 1969. ] l-g 11.21 ANSI N13.10-1974, " Specification & Performan.c of On-Site Instrumentation for jg Continuousiv Monitoring Radioactnity in E0luents"; September,1974 11.22 Nuclear Regulatory Commission Generic Letter 89-01, " Guidance for the 1mplementation of Programmatic Controls for RETS in the Administrative Controls Section of Technical } Specifications and the Rclocation of Prc,:edural Details of Current RETS to the Offsite Dose Calculation Manual or Process Control Program", January 1989 l l 11.23 NRC Answers to 10 CFR 20 Implementation Questions l 11.23.1 Lettcr, F. J. Congel to J. F. Schmidt, dated December 9,1991. t t j l 1123.2 Internal USNRC memo, F. J. Congel to V. L. Miller, et al, dated April 17,1992. 1 l 11.23.3 Letter, F. J. Congel to J. F. Schmidt, dated April 23,1992. l l 11.23.4 Letter, F. J. Congel to J. F. Schmidt, dated September 14,1992. l 1123.5 Letter, F. J. Congel to J. F. Schmidt, dated June 8,1993. 4 f i B 1 IlI f I

I

,I 4 E

a t

t ~

I APA-ZZ-01003 Rev.3 . I A'ITACHMENT 1 LOWER LIMIT OF DE'ITICT10N (LLDJ A detailed discussion of the LLD, and other detection limits, can be found in ll ASL Procedures Manual, I HASL-300 (rnised annually), Curie, L. A. " Limits for Qualitadve Detecdon and Qualitative l Determination - Application to Radiochemistry", Anal Chem 40. 585-93 (1986), and Hartwell, J. K., " Atlantic Richfield Hanford Company Report ARH S A-215 (June 1975). The LLD is defined, for purposes of these controls, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability a with on!) 5% probability of falsely concludir.g that a blank observadon represents a ".real" signal. For a particular measurement system, which may include radiochendcal separation: I LLD = E x V x 2.22E6 x Y x exp(-26t) ~ Wherc: LLD = the "a priori" lower lii, tit of detection (microCuries per unit mass or volume), l S= the standard deviadon of the background counting rate or of the counting rate of a blank b

-i W sample as appropriate (counts per minute),

E= the counting efficiency (counts per disintegration), a V= the sample si7.c (units of mass or volume), 2.22E6 = the number of disintegradons per minute per ndcroCurie, Y= the fractional radiochemical yield, w hen applicable, ). = the radioactive decay constant for the particular radionuclide (sec-l), and 61 = the elapsed time between the midpoint of the sample collection period, and the time of counting (sec), for cIIluent samples, or At = the clasped time between the end of the sample collection period, and the time of coundng (sec), for emironmental samples. 5 Typical values of E, V, Y, and At should be used in the calculadon. It should be recognized that the LLD is defined as a apriori(before the fact) limit representing the I I 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 routinc conditions. I The derm; don of at applies only to the calculation of the LLD. A more rigorous treatment of the buildup and decay during the sample collecdon and/or counting period (s) may be applied to actual sampic anajvris - I of desired. I Page1of1 ATTACHMENT 1 I i

4 I APA Z2.-01003 Rev.3 ATTAClIMENT 2 DASES FOR RADIOLOGICAL EFFLUENT CONTROLS. FO_.E: The BASES presented below summarize the reasons for the specified Radiological Emuent Control, but in accordance with 10 CFR 5036 are not pan of these controls. REC 9.1 ]pdioactive Liauid Emuent Monitorine Instrumentation j The radioactive liquid emuent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid emuents during actual or potential releases ofliquid emuents 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 values of 10 CFR l Pan 20.1-20,60, Ap;rndix B. 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. REC 9.2 Radioactive Gaseous Emuent Monitorine Instrumentation The radioactive gaseous emuent monitoring instrumentation la provided to morutor and control, as applicable, the releases of radioactive materials in gaseous emuents during actual or potential releases of gaseous emuents. The Alarm / Trip Setpoints for these I 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 1 l consistent with the requirements of General Design Criteria 60,63, and 64 and Appendix ,j E A to 10 CFR Part 50. The sensitivity of any noble gas activity monitor used to show 4 compliance with the gaseous emuent release requirements of REC 9.7.1.1 shall be such that concentrations as low as 1 x 104 pCi/cc are measurable. Y1 The Containment Purge Monitor (noble gas channel) closes the Containment Isolation Valves in the Mini Purge System and the Shatdown Purge System. This action isolates I the Containment atmosphere from the emironment to minimize releases of radioactivity in the event of an accident. The Mini Purge System is normally used during Reactor operation and the Shutdown Purge System will be in use with the Reactor shutdown. Each I of the puige systems has inner and outer Containment Isolation Valves in its supply and exhaust ducts..A high radiation signal from either monitor initiates Containment Purge Isolation, which closes both inner and outer Containrnent Isolation Valves in the Mini Purge System and the Shutdown Purge System. Th,: Containment Purge Isolation I Radiation Monitoring hutrumentation isolates the Containment atmosphere from the emironment to minimize releases of radioactivity in the event of an accident. I I I I I Page 1 of 8 ATTAC}BtENT 2

i I APA-ZZ-01003 Rev.3 ATTACHMENT 2 j D ASES FOR RADIOLOGICAL EFFLUENT CONTROLS The safety analyscs assume that the Containment remains intact with penetrations I unnecessary for core cx>oling isolated early in the event, and that the purge valves isolate rapidly. The Containment purge isolation radiation monitors act as backup to the SI signal to ensure closing of the purge valves. They are alsn the primary means for automadcally isolating Containment in the event of a fuel handling axident during shutdowrt Containment isolation in turn ensures meeting the Containment leakage rate assumptions of the safety analyses, and ensures that the calculated accidental offsite radiological doses are below 10 CFR 100.11 limits. The Containment Purge Isolation I Functions are required OPERABLE in MODES 1,2,3, and 4, and during CORE ALTERATIONS or movement ofirradiated fuel assemblics within Contairunent. Under these conditions, the potential exists for an accident that could release fission product radioacthity into Containment. Therefore, the Contain.nent Purge Isoladon I instrumentation must be OPERABLE in these MODES. While in MODES 5 and 6 without fuel handling in progress, the Containment Purge I Isoladon Instrumentation need not be OPERABLE since the potential for adioactive releases is minimized and operator action is sufficient to ensure post accident offsite doses are maintained within the lindts of 10 CFR 100.11. l I NUREG-1431 requires one Containment Purge Isolation noble gas monitor OPERABLE t in MODES 1,2,3, and 4, and during CORE ALTERATIONS. The associated ACTION allows purging operations. :.mtinue for four hours, provided that the particulate, iodine, II and area radiation mordtors remain OPERABLE. These monitors are assumed to proside a CPIS on high radiation. Since the four Containment radiation monitors measure difIerent parameters, and are not recundant, the failure of a single channel may resuh in loss of the radiation monitoring Function for certain events. Consequendy, the failed I channel must be restored to OPERABLE status. The 4 hours allowed to restore die affected channel is justified by the low likelihood of events occurring during this interval, and recognition that one or more of the remaining channels will respond to most events. I However, the Ca!!away particulate, iodine, and area radiation monitors do not provide ESFAS actuation, therefore a four hour restoration period is inconsistent with the assumptions stated in NUREG-1431. The Callaway design provides totally rcJundant Containment Purge Isolation monitors, both providing an independent ESFAS actuation signal. Since both monitor die same ventilation duct, only one unit is required OPERABLE in order to ensure ESFAS acu.ation upon high Containment purge acuvny. The LCO and associated actions are consistent with the redundancy requiren.ents cf I 2 Regulatory Guide 1.97. Table 3 of this Regulatory Guide identifies the Contairunent Purge monitors as Category 2 equipment, and Table I shows that there are no specific l requirements for redunancy. Therefore, only one monitor is required OPERABLE for purging operations. I i 1 " Standard Technical Specifications Westinghouse Plants *, NUREG-1431, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation. September,1992. 2 Instrumentation for Light-Water - Cooled Nuclear Power Plants to Assess Plant and Envitcas Conditions During and following an Accident *, Regulatw Guide 1.97, Revision 3, USNRC, Office of Nuclear Regula:ory Res.carch. May,1983. Page 2 of 8 ATTACHMENT 2 I I l I

I APA-ZZ-01003 Rev.3 ATTA CHMENT 2 B ASES FOR RADIOLOGICAL EFFLUENT CONTRQ1.4 The LCO and associated ACTIONS are consistent with NUREG 1431 and 3 NUREG-1301. REC 9.3 J,10UID EFFLUENTS CONCENTRATION I This secdon is provided to ensure that the concentradon of radioactive materials released in liquid waste effluents to UNRESTR]CTED AREAS will be less than the concentradon .I levels specified in Appendix B, Table II, Column 2 to 10 CFR 20.1-20.601. This limitation provides additional assurance that the levels of radioactive materith 4- " 2.ies of water in UNRESTRICTED AREAS will result in exposures within: (1) die Section II.A design objectives of Appendix 1,10 CFR Part 50, to a MEhGER OF THE PUBLIC, I and (2) the limits of 10 CFR Part 20.1301 to the population. The concentration limit for dissolved or entrained nob!c gases is based upon die assumpdon that Xc-135 is the controlling radioisotope and its MFC ir. air (submersion) was converted to an equivalent I concentration in water using the methods descnbed in Internadonal Commission on Radiological Protection (ICRP) Publication 2. The required detection capabiliues for radioactive materials in liquid waste samples are I tabulated in terms of the lower limits of detection (LLD's). I lI i l t l i lI lI !j 3'Offsite Dose Calculation Manual Guidance: Standard Radiologic.J Effluent Contro!s For Pressurized

5 Water Reactors, Generi

Letter 89-01, Supplenient No.1 *, NUREG-1301, USNRC, office of Nuclear Reactor Regulation. April,1991.

lI Page 3 of 8 ATTACHMENT 2 e n

I APA-ZZ-01003 ATTACHMENT 2 B ASES FOR RADIOLOGICAL EFFLUENT CONTROLS REC 9.4 Dose From Liould Emuents This section is prosided to implement the requirements of Sections ll. A and IV. A ef Appendix 1,10 CFR Part 50. The Linuting Condition for OperaUon implements the I guides set fonh in Section ILA of Appendix 1. Th: ACTION statements proside the required operating flexibility and at the same nme implement the guides set forth in Section IV. A of Appenex 1 to assure that the releases of radioactive material in liquid emuents 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 operat ons, there is reasonable assurance that the operation of the facility will not i I result in radionuclide concentratiers in the finished drinking water that are in excess of the requirements of 40 CFR Part 141. The dose calculadon methodology and parameters in the ODCM iniplement the requirements in Section llLA of Appendix 1 which specify that conformance with the guides of Appendix 1 be shown by calculational procedures based on models and data. such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unhkely to be substantially underestimated. The equations specified it. the ODCM for calculaung the doses due to the actual release rates I of radioactive materials in liquid emuents are consistent with the methodology prosided in Regulatory Guide 1.109, " Calculations of Annual Doses to Man from Routine Releases of Reactor Emuents with 10 CFR Pan 50, Appendix I", Resision 1. October 1977 and I Regulatory Guide 1.113, " Estimating Aquade and Dispersion of Eduents from accidental and Routine Reactor Releases for the Purpox ofimplemenung Appendix I", April 1977. REC 9.5 LIOUID RADWASTE TREATMENT SYSTEM The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid emuents require treatment prior to release to the I environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid emuents will be kept "as low as is reasonably achievable". This section implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 I CFR Pan 50 ara; the design objective given in Section II.D of Appendix I to 10 CFR Part

50. The speciDed limits governing the use of appropriate portions of the Liquid Radwaste Treatment System were speciDed as a suitable fraction of the dose design objectives set forth in Section 11.A of Appendix 1,10 CFR Part 50, for liquid emuents.

I E I I Page 4 of 8 ATTACHMENT 2

I APA-ZZ-01003 Rev.3 ATTACHMENT 2 B ASES FOR RADIOLOGICAL EFFLUENT CONTROLS REC 9.6 G ASEOUS EFFLUENTS DOSE RATE This section is provided to ensure that the dose at any time at and beyond the SfTE BOUNDARY from gaseous emuents from all units on the site will be within die annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS. The dose rate limits are the 2 doses associated with the concentrations of 10 CFR Part 20.1-20.601, Appendix B. Table II, Column 1. These limits prmide reasanable assurance that radioactive material discharged in gaseous efnuents will no'. result in the exposure of a MEMBER OF THE 'E PUBLIC in an UNRESTRICTED APJiA, either within or outside the SITE BOUNDARY, E to annual average concentrations execeding the dose limits speci.fied in 10 CFR Part 20 10 CFR 20.1301. For MEMBERS OF TIE PUBLIC who may at times be within du: g SITE BOUNDARY, the cccupanci of that hEMBER OF THE PUBLIC will usually be -g sumciendy low to compensate for any increase in the atmospheric diffusion factor above i that for the SITE BOUNDARY. Examples of calculadons for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODCM. The I specified release rate limits restrict, at all times, the corresponding gamma and beta dose i rates above background to a MEhGER OF THE PUBLIC at or bcyond the SITE BOUNDARY to less than or equal to 500 mrem / year to the whole body or to less Otan or equa: to 3000 mrems/ year to the skin. These release rate limits also restrict, at all times, I the corresponding th>Toid dose rate above background to a child sia the inhalation pathway to less than er equal to 1500 mrems/ year. l The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLD's). i REC 9.7 POSE - NOE LE GASES This section is provided to implement the requirements of Sections II,B, IU.A, and IV. A of i Appendix I,10 CFR Part 50. The Limiting Conditions for Operation implements the j guides set forth in Section ll.B of Appendix I. The ACTION statements proside 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 emuents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable" The Surveillance Requirernents implement the requirements in Section Ul.A of Appendix 1 that conformance with the guides of Appendix 1 be shown by calculational procedures based on models and data suci, that the actual exposure of a MEhBER OF THE PUBLIC through appropriate pathways in urdikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the I doses due to the actual release rates of radioactive noble gases in gaseous efDuents are consistent with the methodoloD' Provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases on Reactor Efiluents for the Purpese of Evaluating Compliance with 10 CFR Part 50, Appendix 1", Revision 1, October 1977 and lI' Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Ef! bents in Routine Releases from Light-Water Cooled Reactors", Resision 1, Ady 1977. The ODCM equations provided for determining d.e air doses at and beyond the SITE BOUNDARY are bas.ed upon the historical average atmospheric conditions. I Page 5 of 8 ATTACHMENT 2

APA-ZZ-01003 Rev.3 ATTACHMENT 2 DLAJES FOR RADIOLOGICAL EFFLUENT CONTROLS REC 9.8 DOSE - 10 DINE-13 L & 133. TRITIUM. AND RADIOACTIVE M ATERI AL IN PARTICULATE FORM iI This section is provided to implement the requirements of Sections II.C, IILA. and IV.A of Appendix I,10 CFR Pan 50. The Limiting Conditions for Operation are the guides set forth in Secuon II.C of Appendix 1. The ACTION statements provide the required operating Dexibility and et the same time imp!cment the guides set forth in Section IV.A of Appendix 1 to assure that the release of radioactive traterial in gaseous emuents to UNRESTRICTED AREAS will be kept "as low as reasonably achievable" The ODCM 1 calculational methods sp;cified in the Surveiilance Requirements implenient the requirements in Section Ill.A of Appendix 1 that conformance with the guides of Appendix I be shown by calculatior.al proc,dures based on models and data such that the 'l actual exposure of a hEMBER OF THE PUBLIC through appropriate pathways is W unlikely to be substantially underestimated. The ODCM calculational methodology and parameters for caletdating the doses due to the actual release rates of the subject materials

• g are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of g

Annual Doses to Man from Routine Releases of Reactor Emuents for the Purpose of i Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.111, "Metheds for Estimating Atmospheric Transport and Dispersion 1 of Gaseous Emuent sin Routine Releases from Light-Water Cooled Reactors", Resision 1, j July 1977. These equations also provide for determining the actual doses based upon the historical average atraospheric conditions. The release rate controls for Iodine-131, and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are I dependent upon the existing radionuclide pathways to man, in the areas at and beyond the SrTE BOUND ARY. The pathways that were examined in the development of these calculations were: (1) individual inhalation of airborne radionuclides, (2) deposition of I 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. REC 9.9 GASEOUS RADWASTE TRE ATMENT SYSTEM The OPERABILITY of the WASTE GAS HOLDUP SYSTEM and the VEffrILATION EXHAUST TREAThENT SYSTEM casures that the system will be available for use wh:never gaseous emuents require treatment prior to release to the etnironment. The I requirement that the appropriate ponions of these systems be used, when specified, prosides reasonsole assurance that the releases of radioactive materials in gaseous emuents will be kept *as low as is reasonably achies able" This control irupleme.ts the requirements of 10 CFR Pan 50.36a, General Design Criterion 60 of Appendix A to 10 I CFR Part 50, and the design objectives given in Section ILD of Appendix 1 to 10 CFR Part

50. The specified I;mits 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 ofAppendix I,10 CFR Pan 50, for gaseous emuents.

I I Page 6 of 8 ATTACHMENT 2 I

I APA-ZZ-01003 ATTACHMENT 2 B ASES FOR R ADIOLOGICAL EFFLUENT CObrrROLS REC 9.10 TOTAL DOSE This REC is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20.1301. The control requires the preparation and submittal of a Special Report whenever the calculated doses due to releases of radioaedsity I and the radiation from uraruum fuel cyc!c sources exceed 25 mrems to the whole body or any urgan 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 - hEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the I individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the reactor urdts and from outside storage tanks are kept small. The Special Report will describe a course of acdon that should result in the limitadon of the annual dose to a MEMBER OF T1E 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 hEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the I exception that dose contributions from other nuclear fuel cycle facilides at the same site or within a radius of 8 km must be considered. If the dosc to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report I with a request for a variance (provided the release conditions resulting in siolation of 40 CFR Part 190 have not already been corrected),in accordance with the prosisions of 40 CFR Part 190.11 and 10 CFR 20.2203, is considered to be a timely request and fulfills the I requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to 40 CFR Part 190, and docs not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in REC 9.3.1.1 and 9.6.1.1. An indisidual is not conddercd a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying oct any operadon that is part of the nuclear fuel cycle. REC. 9.11 RADIOLOGICAL ENVIFONMENTAL MONITORING PROGP A.ld l The Radiological Emironmental Monitoring Program required by this REC prosides representative measurements of radiation and of radioactive materials in those ex7esure I pathways and for those radionuclides that lead to the highest potential radiadon exposures of hEMBERS OF THE PUBLIC resulting from the station operation. This monitoring program implements Section IV.B.2 of Appendix I to 10 CFR Part 50 and thereby I supplements the Radiological EfIluent Monitoring Program by verifying that the measurable concentradons of radioactive matenals and levelr of radiation are not higher ) than expected on the basis of the efDuent measurements and the modeling of the emironmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Emironmental 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 operadonal experience. The required detecdon capabilides for emironmental sample analyses are tabulated in I terms of the lower hmits of detection (LLD's). The LLD's required by Table 9.11-C are corsidered optimum for routine environmental measurements m industrial laboratories. Page 7 of 8 ATTACHhENT 2 )

I APA-ZZ41003 I Rev. 3 l ATTACHMENT 2 1 {MSES FOR RADIOLQGICAL EITLUENT CONTRO!J REC 9.12 RADIOLOGICAL ENVIRONMENTAL MONITORIRQ L AND USFdENSUS ) l This REC is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the Radiological Emironmental Mordtoring 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 consuldng with local agricuhural authorides, shall be uvd. This census satisfies the l requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to 2 I gardem of greater than 50 m provides assurance that signincant exposure pathways via leafy vegetables will be identi5ed and monitored since a garden of this size is the { minimum required to prcxiuce the quantity (26 kg/ year) ofleafy vegetables assumed in j Regulatory Guide 1,109 for consumption by a child. To determine this minimum garden i I 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 2 3 kg/m REC 9.13 EA_D_1G. LOGICAL ENVIRONMEtITAL MONTTORING INTERLABORATORY COMPARISON PROGRAM i 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 i part of the quality assurance program for emironmental monitoring in order to demonsfate that the results are valid for the purpose of Section IV.B.2 of Appendix I to 10 CFR Part 50 I I ' I Page 8 of 8 ATTACllMENT 2

APA-ZZ-01003 Rev. 3 1. 1 i 1 APPENDIX A 1 l' . I

SUMMARY

REVIEW OF RADIOLOGICAL EFFLUENT TECH SPECS POTENTIALLY AFFECTED BY THE i IMPLEMENTATION OF THE REVISED 10 CFR 20 i ti 1 I s i I i f i i 4 1 t 4 !I Page 1 of 16 APPENDIX A

I 4

APA-ZZ-01003 Rev. 3

SUMMARY

REVIEW OF RADIOLOGICAL EFFLUENT TECH SPECS I POTENTIALLY AFFECTED BY THE IMPLEMENTATION OF THE REVISED 10CFR20 The following is a summary review of the current Tech Specs that are potentially affected by the implementation of the revised 10CFR20. In general, the potential impact is due to changes in the Efnuent Concentration Values (ECV's)in 10CFR20, Appendix B, Table 2, I Columns 1 and 2 (formerly MPC's), and 10CFR20.1601. This summary is not intended to review those changes that may be necessary as a result of the eventual issuance of the Generic Letter. 1 The NRR staff has stated that the current level of effluent controls is sufDcier.t to protect the health and safety of the public, and further restrictions resul.ing from the revision to Appendix B, T able 2, were unintentional. They are currently preparing a Gener:c Letter I that will provide guidance for submitting Te:h Spec changes that will return to the current level of control. This is currently anticipated during late 1993. Those who implement the revised rule prior to January 1,1994, will have to do so under the requirements of I 10CFR20.1008, which basically requires that the more restrictive requirement (Tech Specs or 10CFR20) be implemented. I DEFINITIONS OF RESTRICTED AREA & MEMBER OF TIIE PUBLIC, AND TECH SPEC 5.1.2, SITE BOUNDARY FOR GASEOUS EFFLUENTS The definition of Restricted Area has not changed significantly from that in the former rule. The definition of the Member of the Public in the revised rule is significantly j different from that in the Callaway Plant Technical Specifications (TS 1.17). There is no corresponding definition of Controlled Area in the former rule. The Callaway Plant was licensed to operate wi:h a Restricted Area as defined in the FSAR and shown on the figures in TS 5.1.4 and in the ODCM. Since the requirements have not been revised, there is no compelling reason to change the Callaway Plant Restricted Area from its current boundaries. In addition, the NRC's backfit analysis 2, performed pursuant to 10 CFR 50.109, concludes that the revisions to 10CFR20 apply primarily to operational procedures and should cause no modifications in facility design. Since the plant siting and the location and size of the i Restricted Area are considered to be a part of the facility design, it is clearly not the intent I of the NRC that revisions to 10CFR20 would require changes to the Restricted Area for currently licensed facilities. I " Final BackEt Analysis for the Revision of 10CFR20, " Standards for Protecdon Against Radiadon'"', USNRC. 05cc of Nucle.ar Regulatory Research. Division of Regulatory Applications August,1990. (Available USNRC Pubbe Do:oments Review.) I Page 2 of 16 APPENDIX A

APA-ZZ-01003 Rev.3 There is also no requirement for the existence of a Controlled Area as defined in the revised rule 2, therefore it is not necessary that one be created at Callaway. The definition of the Member of the Public is significantly different in the revised rule relative to that provided in TS 1.17 and in 40 CFR 190. The revised rule defines the Member of the Public as anyone who is not in the Restricted Area. The Tech Specs and 40 CFR 190 generally define the Member of the Public as anyone who is not occupationally associated with plant operations, and also recognizes that the Member of the Public may, at times, be within the Restricted Area. The major difference is that pursuant to the revised rule, the Member of the Public receives dose against the occupational dose limits of 10 CFR 20.1201 once inside the Restricted Area, but the Tech Spec dermition would limit the dose within the Restricted Area to the limits of I 10 CFR 20.130: Since the limit provided in 20.1301 is much lower than that of 20.1201, the continued use of the more restrictive 40 CFR 190 and Iech Spec 1.17 definitions for the Member of the Public is appropriate and is required pursuant to 10 CFR 20.1008(c). A more thorough and detailed analysis of the definitions of the Member of the Public found in 10 CFR 20,40 CFR 190, and Tech Spec 1.17, focusing on the applicability of Occupational Vs. Non-occupational dose limits, indicates a confusing and inconsistent array of definitions and dose limit applicability. For conservatism and simplicity, Union Electric has defined occupational dose as dose received while working with or around I radioactive materials. This definition is more restrictive than the definition in 10 CFR 20 in that the more restrictive dose limits of 10 CFR 20.1301 are applied to Members of the Public within the Restricted Area, instead of the less restrictive limits of 10 CFR 20,1201. It is more restrictive than the Tech Spec definition in that delivery persons, service technicians, and others who may enter the site to perform non-radiologica! work activities are also limited to the more restrictive dose limits of 10 CFR 20.1301. 1 There are no changes recommended for those definitions and maps relative to the Restricted Area, Site Boundary, and dose to the Member of the Public. i TECH SPEC 6.8.4.F.2, LIQUID EFFLUENT RELEASE RATE LIMITS (REC 9.3) i On December 1,1992, Union Electric Co. provided notification) ofintent to implement the revised 10 CFR 20, Parts 20.1001-20.2401 and associated appendices, pursuant to 10 CFR 20.1008(a). The revised rule was fully implemented on January 1,1993. The l following provides clarification with respect to compliance to 10 CFR 20.1001-20.2401 and Callaway Plant Technical Specifications 6.8.4.f(2) and 6.8.4.f(7). l 2 Refer to Question 26(a)(4th set). 3 ULNRC 92-2729, D. F. Schndl to A. Ber: Davis, daa:d December 1.1992. Page 3 of 16 APPENDIX A

I AP A-ZZ-01003 Rev. 3 Union Electric implemented the use of the revised Appendix B, Table 2 values concurrent with the implementation of the revised rule. Technical Specification 6.8.4.f(2) requires that the concentration of radioactive material in liquid discharges not exceed the values of 10 CFR 20, Appendix B, Table II, Column 2. The NRC had indicated via the revision to 10 CFR 50.72 that the concentration values have nominally decreased by a factor of 10, and the NRC staff had stated on numerous occasions that they considered the values in the revised rule to be more restrictive than the those in the old rute. This was frequently referred to as an ' implicit" change to the Technical Specifications. 10 CFR 20.1008 (a) requires that if the revised rule is implemented prior to January 1, I 1994, then "the licensee shall implement all provisions of these sections,... and shall provide written notification... that the licensee is adopting early implementation (of the revised rule) and associated appendices." 10 CFR 20.1008 (b) requires that once I implemented, "the applicable section of(the revised rule) shall be used in lieu of any section (of the old rule) that is cited in license conditions or technical specifications." It further states, "if the requirements of(the revised rule) are more restrictive than the existing license condition, then the licensee shall comply with (the revised rule). I Additionally, the NRC had clarified the applicability of the revised Appendix B values to the Technical Specification instantaneous release rate limits via their formal response to three separate licensee questions. Question # 18 states that the Tech Spec instantaneous d I release rate limit is based on the old Pan 20 concentrations, and asks if changes are required in the Tech Specs and ODCM as a result of the revised rule. The NRC replies " the instantaneous release rates for liquid effluents, to the extent that they directly reference I Appendix B concentration values, will need to be changed. The corresponding bases and cenain alarm set-points will have to be changed by license amendment." Question # 23 asks if computer data bases that use the old Appendix B values must be 5 revised to the use the new values. The NRC simply answers, "Yes" I I I d Letter, F. J. Conjel (USNRC) to J. F. Schmitt (NUhiARC), dated December 9,1991. page 16 of. 5 ibid, page 14 of Enclosure 1. Page 4 of16 APPENDIX A

i. APA-ZZ-01003 Rev. 3 Question # 22 states that many alarm set points are based on 10 CFR 20 Appendix B 6 concentrations, and asks if they will have to be changed. The NRC answers that the alarm set-points ofliquid emuent monitors are likely to require change, since they are based on 10 CFR 20 Appendix B concentrations, as required by Tech Specs. Because Appendix B concentration values differ for many radio nuclides between the old and new versions of. Part 20, these set points may have to be changed. This is analogous to a restriction in flow rate, and the NRC cites the reduction in Appendix B concentrations as the root cause of the change. Based on the preceding information, Union Electric implemented the use of the revised Appendix B values concurrent with the implementation of the resised rule on January 1, I 1993. Because there were no values in the revised Appendix B for dissolved and entrained noble gases in liquid emuents, the old value of 2E-4 uCi/ml was used pending regulatory guidance. The Callaway Plant Technical SpeciEcations contain, in Section 6.8.4.f, several specifications which provide appropriate limits on the maximum quarterly and annual whcle body and organ dose to the Member of the Public from the discharge ofliquid and gaseous radioactive eGuents. Compliance with these specifications demonstrates compliance with the limits of 10 CFR 50, Appendix I, and 40 CFR 190 and, as stated in the supplemental information to the revised rule, demonstrates compliance with the 7 100 mrem /yr dose limit of 10 CFR 20.1301. j I However, compliance with the dose rate limits of Specifications 6.8.4.f items (2) and (7) with respect to the implementation of the revised rule is less clear, as there is no longer a regulatory basis for these Specifications. These Specifications formerly implemented the j requirements of10 CFR 20.106, which provided annual average concentration limits on i liquid and gaseous emuents, and specifically referenced the limits of Appendix B, Table II, Columns 1 and 2. I I l I I c USNRC Mernorandum. F. J. Conjel to V. L Miller, et al, dated April 17,1992. page 13 of Enclosure 1. 7 Federal Regis1er. Vol 58. No. 98, Tuesday, Ma) 21,1991. pages 23360-23474 I Page 5 of 16 APPENDD: A I

i APA-ZZ-01003 Rev. 3 1 Unlike the fonner mie, the values in the revised Appendix B, Table ?, Columns 1 and 2 do not of themselves constitute a limit on the release rate of radioactive emuents, but rather, as discussed in 10 CFR 20.1302 (b)(2)(i), merely provide one means of demonstrating compliance with the annual dose limit of 10 CFR 20.1301. Since there is no release rate limit provided in the revised rule, the subject Specifications are therefore license I. conditions.10 CFR 20.1008 (c) requires that any existing license condition that is more restrictive than the revised rule remain in force until there is a technical specifcation change. Additionally, since the values in the revised Appendix B, Table 2 are not limits as I. was the case with 20.106, there is no corresponding provision in the new rule to 20.106. 10 CFR 20.1008(e) requires that is a license condition cites a provision in the old rule for which there is no corresponding provision in the new rule, then the license condition I remains in fotee until there is a technical specification change. I The values of Appendix B, Table 2, Columns 1 and 2 of the revised mle did not change in a uniform fashion, i e., certain nuclides numerically decreased in value whereas others numerically increased in value. Funhermore, the values did not change by a consistent amount, varying by as much as a factor of 20 with respect to the corresponding nuclide in the former rule. This inconsistency is clearly evident for those nuclides which are j commonly associated with nuclear power plant effluents. In addition, the bases for the I revised values is the dosimetry system ofICRP 26 8 and ICRP 30 9 This is inconsistent with the bases for the dose limits of 10 CFR 50, Appendix 1 and 40 CFR 190, and the dose calculational methodologies of Regulatory Guide 1.109, which are largely based on the dosimetry system ofICRP 2 R l I I I I I 8 International Commission on Radiation Protection, Publication 26, " Recommendations of the International Conunission on Radiation Protection", Annals of the ICRP, Volume 1, No. 2,1977. ' International Comnussion on Radiauor. Protection, Publicadon 30, " Limits for Intakes of Radionuelides by Workers", Annals of the ICRP, Volume 2. No. 3/4,1979 10 International Commission on Radiation Protection. Publication 2, " Report of Committee 11 on Permissible Dose for Jnternal Radiation",1960. Page 6 of 16 APPENDIX A l

APA-ZZ-01003 Rev.3 Since the values of the revised Appendix B, Table 2, Columns 1 and 2 did not uniformly increase or decrease in value, it is not possible to determine whether Appendix B, Table 11 of the former mle or Appendix B, Table 2 of the revised rule provides,in toto, the more conservative values for implementation of the subject license conditions.- It is clear, however, that the bases for the revised Appendix B, Table 2 values are inconsistent with the bases of 10 CFR 50, Appendix I end 40 CFR 190, and Regulatory Guide 1.109. Furthermore, the operational history of the Callaway Plant demonstrates that the use of the 10 CFR 20.1-20.601, Appendix B, Table II values is appropriate to maintain compliance with the requirements of 10 CFR 50, Appendix I and 40 CFR 190, which, in turn, demonstrates compliance with the 100 mremlyr dose limit of 10 CFR 20.1301. The concentration limits of the old Appendix B, Table II were based on a dose of .I 500 mrem /yr, which, when expressed as a dose rate,is equal to 057 mrem /hr. Compliance with the requirements of Technical Specifications 6.4.8.f(2) and (7) using 10CFR 20.106, Appendix B, Table II values is conservative with respect to the 2 mrem /hr I limit of 10CFR20.1301(a)(2). Additionally, Technical Specifications 6.4.8.f(2) and (7) specifically require the use of Appendix B, Table II to 10CFR20.1-20.601, since there is no corresponding provision in the revised rule. Thus,10 CFR 20.1008 (c) and (e) require the continued use of the values provided in Appendix B, Table II to 10 CFR 20.1-20.601 for the implementation of Technical I Specifications 6.8.4.f, items (2) and (7). g Although the 2 mrem /hr limit of 10 CFR 20.1301(a)(2) was referenced in the oreceding 5 discussion, it is important to note that the regulation specifically states that this limit is applicable to external sources. Since, for the Callaway Plant, the only dose pathway to I man from the discharge ofliquid radioactive emuent is through the consumption of fish, there are no external dose pathways, and therefore the requirements of 10 CFR 20.1301(a)(2) are satisfied apriori. Union Electric re instituted the use of the values in Appendix B, Table II, Columns 1 and 2, to 10 CFR 20.1-20.601 for Technical Specifications 6.8.4.f, items (2) and (7) pursuant to the requirements of 10 CFR 20.1008(c) and (e), on May 4,1993. This position was affirmed by the USNRC on June 30,1993". ErrLIJENT CONCENTRATION VALUE FOR GROSS ALPHA IN LIQUID EFFLUENTS I There are two values in the revised Appendix B for unknown mixtures in liquid efIluents: 2E-9 and IE-6 uCi/ml. The less restrictive value is appropriate ifit is known that certain nuclides are "not present. The appropriate value for gross alpha in liquid effluents at the Callaway Plant from Appendix B, Table 2, Column 2 is IE-6 uCUml. I 11 Lctter, Thomas E. Murley, Director, NRR, USNRC, to Thomas E. Tipton, NUMARC. dated June 30,1993. Page 7 of 16 APPENDIX A I

I APA-ZZ-01003 Rev. 3 The value of IE 6 uci/ml in Appendix B, Table 2, Column 2 only applies to an unknown mixture of nuclides where those listed opposite the value are known to be "not present" These nuclides are Fe-60, St-90, Cd-113m, Cd-l 13, in-115,1-129, Cs-134, Sm-147, Gd-148, Gd-152, Hg-194 (organic), Bi-210m, Ra-223, Ra-224, Ra-225, Ac-225, Th-2?8, Th-230, U-233, U-234, U-235, U-236, U-23 8, U-nat., Cm-242, Cf-248, Es-254, Fm-257, I and Md-258. The other nuclides listed in the immediately preceding values for ur.known. mixtures in gaseous emuents do not apply, since they specifically apply to gaseous emuents as indicated by the designation of applicable lung clearance classifications for I each of the nuclides listed. The NRC's response to Question # 71 reiterates that ingestion ALI's do not have lung clearance classifications, which is also consistent with ICRP 30 and all other industry standards. Additionally, several of those listed in the list for liquid ' I emuents also appear in the list of nuclides given for airbome activity, which indicates that only those specifically listed with the liquid emuent value apply. Of those nuclides listed for unknown mixtures in liquid emuer,ts, only Ra-224, Th-228, U-234, U-235, U-236, U-238, and Cm-242 are LWR produced alpha emitting nuclides. I Sr-90, Cd-l13m, I-129, and Cs-134 are also LWR produced, but are beta or beta / gamma emitters, and are not determined via a gross alpha analysis. The remainder of the nuclides in the list are not LWR produced. The phrase "not present" is not defined in the revised 10 CFR 20, however there is a large body ofinformation which can be applied to determine the meaning of"not present". The I former mie, in footnote 5 to Appendix B, stated that a nuclide may be considered to be "not present" ifit constitutes less than 10% of the total activity, provided that the aggregate of all such not present" nuclides does not exceed 25% of the total activity. I The use of the " ten percent rule"is consistent with the basis of the revised rule, the NRC's response to questions regarding the meaning of"not present", and the current ICRP guidance as shown below: The revised rule is based on the dosimetry and methodology ofICRP 30 12, a. which in paragraph 3.1.3, describes the use of the current ten percent rule.

b. The NRC's response to Question # 146 13 clearly indicates that the ten-percent rule is applicable to Appendix B.

I I 12 ICRP Publication 30, " Limits for intakes of Radionuclides by Workers", in Annals of the ICRP, VoSme 2, Number 3/4,1979. IN Lette, Frank J. Congel, Director, DRPEP. USNRC, to John F. Schmidt, NUMARC, dated September I 14,1992 ' commonly referred to as the 4th set of Q&A) l I Page 8 of 16 APPENDIX A I 1

r APA-ZZ-01003 Rev. 3

c. The current ICIU' recommendations on the release of radioactive materials to the environmentl4, and the updated recommendations 5 to ICRP 30 continue l

to propagate the ten-percent rule, and apply it to ofTsite dose as well as dose to radiation workers. It is therefore clear that the ten-percent rule continues to apply to the values in Appendix B of the revised rule. Callaway Plant liquid efIluents hase been analyzed for transuranic nuclides (TRU) on two separate occasions, during the second and third quarters of1987. In each instance, TRU nuclides were not detectable, with an MDA of IE-8, uCi/ml, which is a factor of 10 belcw the gross alpha LLD of IE-7 uCJml. I The concentration of the TRU nv!; des can be inferred through the use of a tracer nuclide, such as Ce-144. Cc-144 is particularly well suited for this purpose in that it is a fission product, can be measured by gamma ray spectroscopy, and is cheretically sinJlar to the I TRU nuclides. Based on published ORIGEN code calculationsl6 of a representative LWR, and assuming a 90 day decay, the ratio of the nuclides ofinterest to Ce-144 is: I Ra-224/Ce-144 1.45E-9 Th-228/Ce-144 1.45E-9 U-234/Cc-144 1.14E-6 I U 235/Ce-144 1.75E-8 U-236/Ce-144 2.58E-7 U-238/Ce-144 3.24E-7 Cm-242/Ce-144 2.66E-2 Vollique, et al 17, found the Cm-242/Ce-144 ratio to be 6.5E-3, which is consistent with the above value. Based on the above, it can be seen that Cm-242 is the only nuclide with a significant Ce-144 ratio. I I 14 ICRP Publication 56. " Age-dependent doses to Members of the Publi: from the intake of Radionuclides: Part 1*, Annals of the ICRP, Volume 20, Number 2,1989. 35 ICRP Publicadon 61, " Annual Limits on Intake of Radionuclides by Workers Based on the 1990 Recommendations", Annals of the ICRP, volume 21, Number 4,1991. I i 161.icht Water Reartor Nuclear Fuel Ovele, Wymer, Rayuend G. and Vondra, Benedict L, editors, Tabic 6, pages 70 71 and Table 7, page 72. CRC Press,1981. 37 Vollique, P. G., et al,

• Solubility of Transurani: Nuclid:s in Aerosols in Two Ginna Steam Generator Work Environments" Proceedings of the Twenty-First Midyear Topical Meeting of the Health Physics Society, Pages 251-260,1987.

Page 9 of 16 APPENDIX A

APA-22-01003 Rev. 3 Based on the data contained in the Semiannual Efiluent Release Reports for the period January,1989-July,1992, Ce 144 accounted for less than 0.3% of the total fission and activation product activity in liquid efIluents, and less than SE-6% of the total activity discharged in liquid effluents during the same period. Therefore, the rnaximum activity that could have been discharged of each of the above listed nuclides is much less than 10%. Accordingly, these nuclides are "not present" TECH SPEC 6.8.4.F.4, DOSE FROM LIQUID EFFLUENTS (REC 9.4), & TECH SPEC 6.8.4.F.5, LIQUID RADWASTE TREATMENT SYSTEM (REC 9.5) I These specifications are derived from 10CFR50, Appendix 1, and are not affected by the revised rule. Doses are calculated in accordance with Regulatory Guide 1.109 which has not been revised. No changes are anticipated for these specifications. I ~ TECH SPEC 3.11.1.4, CURLE CONTENT OF OUTDOOR LIQUID STORAGE TANKS The purpose of this specificatio, is to limit the activity in the nearest receiving waters, excluding tritium and entrained noble gases, to the concentrations in 10CFR20, I Appendix B, Table 2, Column 2. The effect of accidental contamination of the nearest ground water discharge locations due I to accidental rupture of tanks containing radioactive liquids was performed as detailed in FSAR Section 2.4.13.3. It was assumed that the liquid contents of a ruptured tank would I immediately merge with the ground water 5 feet below plant grade and travel directly from the tank to the nearest down-gradient well (Well 23). The results of the calculation show that, with the exception of H-3 and Sr-90, the radio nuclide concentrations found in I ground water after a tank rupture will be below the original 10CFR20, Appendix B, Table II, Column 2 values by the time the contaminated ground water reaches the nearest stream tributaries. The dilution capability of the streams is sufficient to reduce the I concer.tration of H-3 and Sr-90 below the original Appendix B values All computed concentrations at Well 23 were below the Appendix B limits for unrestricted areas. I l I I Page 10 of 16 APPENDIX A I

r APA-ZZ-01003 Rev. 3 Tables I and II list the curie contents of the primary spent resin storage tank and refueling water storage tank used in the FSAR calculations. These values were adjusted to reflect a total tank curie content of 150 Curies, the limit identified in Tech Spec 3.11.1.4. (Even though the spent resin storage tank is not an outdoor tank, the data was used for this calculation since it is expected to have the highest curie contents for Sr-90, Cs.137 and I Co-60 and the postulated accident assumes that all liquid released immediately merges with the ground water.) The resultant peak concentrations at the discharge point at Logan Creek were calculated using the normalized values then compared to the revised I Appendix B efIluent concentration values (ECV). All calculated concentrations at the discharge point were less than the applicable ECV. Based on the above calculation, the existing Tech Spec limit of 150 Curies is conservative in comparison to the revised 10CFR20, Appendix B values and is therefore still applicable. I I I I I I I I I I I Page11 of16 APPENDIX A

r APA-ZZ-01003 TABLEI A. Curie Content of Radionuclides in the Primary Spent Resin Storage Tank NUCLIDE Ci' (in tank) Ci (normalized to 150 Ci, total) I I Mn-54 2.91 E+01 8.17E-O l Co-58 6.10E+02 1.71 E+01 Co-60 2.56E+02 7.19E+00 I Sr-89 9.80E+00 2.75 E-01 Sr-90 1.35E+ 00 3.79E-02 hb-95 3.00E+00 8.42E-02 Zr-95 2.12E+00 5.95E-02 1-131 1.17E+03 3.28E+01 Cs-134 1.78E+03 5.00E+0) Cs-137 1.48E+03 4.15E+01 Ba-140 1.63E400 4.58E-02 TOTAL 5.343E+03 149.91

• Values are from FSAR Table 2.4-28.

I B. Peak Concentrations ofRadionuclides at the. Logan Creek Discharge Point j NUCLIDE p Ci/ml* pCi/ml(based ECV %ECV (original cale) on 150 Ci total) Mn 54 3.1 E-22 8.7E-24 3E-05 3E-17% I Co-60 3.6E-23 1.0E-24 3 E-06 3E-20% i Sr-90 1.2E-05 3.4E-07 SE-07 67.4 % Cs-137 5.5E-06 1.5E-07 IE-06 15.4 % I

• Values are from FSAR Table 2.4-30.

I I Page 12 of 16 APPENDIX A I

fI j APA-ZZ-01003 TABLEU j A. Curie Content of Radionuclides in Refueling Water Storage Tank I NUCLIDE Cia (in tank) Ci (normalized to 150 Ci, total) Mn-54 6.99E-06 2.19E-02 Co-58 3.36E-04 1.05E+00 I Co-60 4.58E-05 1.43E-01 Sr-89 5.92E-05 1.85E-01 Sr-90 1.92E-06 6.02E-03 I Nb-95 1.3 I E-06 4.10E-03 Zr-95 1.25E-06 3.92E-03 1-131 2.34E-02 7.33 E+01 I Cs-134 1.39E-02 4.35E+01 Cs-137 1.01E-02 3.16E+01 Ba-140 2.56E-05 8.02E-02 TOTAL 4.78SE-02 149.9 ' Values are from FSAR Table 2.4-28. I B. Peak Concentrations ofRadionuclides at Logan Crer* iischarge Point NUCL1DE pCihr!* pCi/ml (based ECV %ECV (original calc) on 150 Ci total) I Co-60

1. l E-30 3.4E-27 3E-06 lE-19%

Sr-90 2.5E-13 7.SE-10 SE-07 0.16% Cs-137 8.4E-13 2.6E-09 1E-06 0.26 % I

• Values are from FSAR Table 2.4-30.

I .I I Page 13 of16 APPENDIX A I

I APA-ZZ-01003 Rev. 3 TECH SPEC 6.S.4.F.7. DOSE RATE LIMIT FOR GASEOUS EFFLUENTS (REC 9.6) Tids specification provides a gaseous emuent dose rate limit conforming in the ECV's in 10CFR20, Appendix B, Table 2, Column 1. For the nuclides ofinterest to Callaway, the ,I revised ECV's are numerically greater, therefore the current REC is more restrictive than the dose rates conforming to the revised Appendix B values.10CFR20.1008 requires the implementation of the more restrictive of the requirements of 10CFR20, technical specifications, or any speciallicense conditions. The current REC represents the more restrictive requirement and will be implemented without revision. The former rule, in 20106(a), limited the amount of radioactivity released in emuents to the concentrations specified in Appendix B, Table 2, averaged over a period of one year. I Ahhough not specified as a limit, this corresponded to an annual whole body dose lindt of 500 mrem to the Member of the Public. The former rule did not specify a dose rate limit. The revised rule, in 20.1301, specifies two limits on radioactivity in emuents: An annual dose limit of 100 mrem, TEDE (20.1301(a)(1)) and a dose rate limit of 2 mrem /h i TEDE (20.1301(a)(2)). Note that the revised rule does not specify limits on concentration as did the former rule but does allow licensees to utilize the concentration values in Appendix B, Table 2 to demonstrate compliance with the limits of 20.1301 4 (20.1302(b)(2)). Note that 20.1302(b)(2)(i) describes these as " annual average concentrations" as opposed to instantaneous limits. Measurements and calculation means are also allowed (20.1302(b)(1)). Radiological Emuent Control (IGC) 9.6 is required by Technical Specification 6.8.4 f.7 to contain: " Limitations on the dose rate resulting from radioactive material released in gaseous emuents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10CFR20, Appendix B, Table 2, Column 1." The bases for this Control state that its purpose is to ensure that the dose at any time from gaseous emuents is within the annual dose limit of 10CFR20, which is the dose associated I with the concentrations of 10CFR20, Appendix B, Table 2, Column 1. Additionally, tids Control provides assurance that the release of gaseous emuents will not result in the .'g exposure of a Member of the Public to annual average concentrations in excess of the g values of 10CFR20, Appendix B, Table 2, Column 1. Note that in each case, the bases references an annual dose limit but makes no reference to a dose rate limit. JC establishes a release rate limit of 500 mrem /y that is equal to approximately a rnRem/h, well below the dose rate limit of 2 mrem /h specified in 20.1301(a)(2), and is therefore more restrictive, I Page 14 of16 APPENDIX A

L APA-ZZ-01003 Rev. 3 The preamble to the revised rule states that demonstration of compliance with the limits of 40CFR190 and with 10CFR50, Appendix 1 is sumcient to demonstrate compliance with I the 100 mrem dose limit of 20.1301(a)(1). Other Controls are provided as required Technical Specification 6.8.4.f(items 8,9, and 10) which ensure that the limits of 40CFR190 and 10CFR50, Appendix I are not exceeded. The Bases for this Control reference the concentration values of 10CFR20, Appendix B, Table 2, Column 1 as a basis for the specified dose rate limits. These values were derived I using ICRP 30 calculation methodology and the dose and dose rate values they represent are the Total Effective Dose Equivalent (TEDE) which is the summ:. tion of the external and intemal dose components. Compliance with the Control is demonstrated through I calculation methodologies and parameters as established in Regulatory Guide 1.109 and NUREG 0133, which are based on the ICRP 2 maximum organ methodology, and thus cannot be used to calculate emuent cases and dose rates that correspond to the I concentration values specified in the revised 10CFR20, Appendix B, Table 2, Column 1. The table below compares the numerical value of the former and revised Appendix B I values for those nuclides most commonly reported in the Callaway Plant's gaseous emuents: 10CFR20 APPENDIX B CONCENTRATION VA. LUES Nuclide Former Rule Revised Rule New/Old Kr-85 3E-7 pCi/ml 7E-7 pCi/ml 2.3 I Xe-133 3 E-7 5E-7 1.7 Xe-135 lE-7 7E-8 0.7 I-l 31 1E-10 2E-10 2.0 I-133 4E-10 1E-9 2.5 Co-58 2E-9 1E-9 0.5 Co-60 3E-10 SE-11 0.2 I Of these, Xe-133 accounts for greater than 90% of the total activity released from the Callaway Plant in gaseous emuents for the past three years (1989-1991). The concentration value for Xe-133 actually areased in the revise rule, as did that for Kr-85 and both iodine nuclides. Although the Co-58 and Co-60 values did decrease in the revised rule, they are relatively insignificant contributors to the whole body and organ dose from gaseous emuents discharged from the Callaway Plant as summarized below. I L

I Page 15 of 16 APPENDIX A

APA-ZZ-01003 Rev.3 GASEOUS EFFLUENT ACTIVITY PROFILE 1989 - 1991 I Fraction of Total Ratio of Appendix B Nuclide Activity Released Concentration Values Noble Gases: Xe-133 0.92 1.7 I Xe-13 5 0.04 0.7 Xe-133m 0.01 2.0 Kr-85m 0.01 1.0 I Kr-85 0.01 2.3 Particulates and lodines. I I-131 0.72 2.0 I-133 0.11 2.5 Co-58 0.03 0.5 I Co-60 0.14 0.2 I The NRC states, in their response to Question 19, that until 10CFR50, Appendix Iis changed, licensees must continue to show compliance with Tech Specs in terms of organ and whole body doses as per Regulatory Guide 1.109. The response to Question 21 states I that Regulatory Guide 1.109 will not be resised at this time, thus Regulatory Guide 1.109 methodology continues to be utilized to show compliance with Tech Specs. Since the dose calculation methodology has not been revised, it would be more conservative to I continue to utilize the current REC values vice dose rate limits calculated from the revised 10CFR20, Appendix B values. Refer to the discussion of T/S 68.4.f.2 (REC 9.3) for additional details. TECH SPEC 6.8.4.F.6, GASEOUS RADWASTE 'lCATMENT SYSTEM OPERABILITY (REC 9.9) TECH SPEC 6.8.4.F.8, DOSE FROM NOBLE GASES (REC 9.7) TECH SPEC 6.8.4.F.9, DOSE FROM IODINES AND PARTICULATES IN GASEOUS EFFLUENTS (REC 9.8) TECH SPEC 6.8.4.F.10, TOTAL DOSE FROM THE URANIUM FUEL CYCLE (REC 9.10) These specifications are derived from 10CFR50, Appendix I and 40CFR190 and are not I affected by the revised rule. Doses continue to be calculated in accordance with Regulatory Guide 1.109 which has not been revised. No changes are anticipated for these specifications. I Page 16 of 16 APPENDlX A}}