ATTACHMENT 1

OCONEE NUCLEAR STATION PROPOSED TECHNICAL SPECIFICATION REVISION

RADIOLOGICAL EFFLUENT CONTROL

Pages

1-5 4,.1-1 4.21-1 1-6 4.1-2 6.1-2 3.5-31 4.1-5 6.1-3 3.5-32 4.1-10 6.1-3a 3.5-33 4.1-11 6.1-4 3.5-34 4.1-12 6.1-5 3.5-35 4.1-13 6.1-5a 3.5-36 4.1-14 6.4-1 3.9-1 4.1-15 6.4-2 3.9-2 4.1-16 6.5-2 3.9-3 4.1-17 6.6-1 3.9-4 4.1-18 6.6-2 3.9-5 4.1-19 6.6-3

3.10-1 4.11-1 6.6-4 3.10-2 4.11-2 6.6-5 3.10-3 4.11-3 6.7-1 3.10-4 4.11-4 6.7-2 3.10-5 4.11-5 6.8-1 3.10-6 4.11-6 6.9-1 3.11-1 4.11-7 3.11-2 4.11-8

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1.8 RADIOLOGICAL EFFLUENT CONTROL

1.8.1 Source Check

A source check is the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

1.8.2 Channel Calibration

The initial channel calibration for radioactivity measurement instrumentation shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards should permit calibrating the system over its intended range of energy and rate capabilities. For subsequent channel of calibration, sources that have been related to the initial calibration should be used, at intervals of at least once per eighteen months. This can normally be accomplished during refueling outages.

1.8.3. Offsite Dose Calculation Manual

The Offsite Dose Calculation Manual (ODCM) is a manual containing the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm/trip setpoints.

1.8.4 Process Control Program (PCP)

The Process Control Program shall contain the sampling, analysis, and formulation determination by which solidification of radioactive wastes from liquid systems is assured.

1.8.5 Solidification

Solidification shall be the conversion of radioactive wastes to an immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).

1.8.6. Gaseous Radwaste Treatment System

A gaseous radwaste treatment system is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

1.8.7 Ventilation Exhaust Treatment System

A Ventilitation Exhaust Treatment System is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be Ventilation Exhaust Treatment System components.

1-5

1.8.8 Purge-Purging

Purge or Purging is the controlled process of discharing air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

1.8.9 Venting

Venting is the controlled process of-discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during Venting. Vent, used in system names, does not imply a venting process.

1-6

3.5.5 Radioactive Effluent Monitoring Instrumentation

Applicability

Applies to radioactive liquid effluent, gaseous effluent, and gaseous process monitoring instrumentation.

Specifications

3.5.5.1 Liquid Effluents

a. The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.5.5-1 shall be operable with their alarm/trip setpoints set to ensure that the limits of Specification 3.9.1 are not exceeded.

b. If a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint is less conservative than required, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable.

c. In the event that the number of operable radioactive liquid effluent monitoring instrumentation channels falls below the limit given under Table 3.5.5-1, Column A, action shall be as shown in Column B.

3.5.5.2 Gaseous Process and Effluents

a. The radioactive gaseous process and effluent monitoring instrumentation channels shown in Table 3.5.5-2 shall be operable with their alarm/trip setpoints set to ensure that the limits of Specification 3.10.1 are not exceeded.

b. If a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint is less conservative than required, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel or declare the channel inoperable.

c. In the event that the number of radioactive gaseous process or effluent monitoring instrumentation channels falls below the limit given under Table 3.5.5-2, Column A, action shall be taken as shown in Column B.

3.5.5.3 Setpoints

The setpoints shall be determined in accordance with the methodology described in the ODCM and shall be recorded.

Bases

The radioactive liquid effluent instrumentation is. provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents

3.5-31

during actual or potential releases. The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to assure that the alarm/trip will occur priorto exceeding the limits of 10 CFR Part 20. The operability and use of this instrumentation is .consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases. The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to assure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring (and controlling) the concentration of potentially explosive gas mixtures in the waste gas holdup system. 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.

.3.5-32

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TABLE 3.5.5-1 NOTES

(a) Effluent releases may continue for up to 14 days, provided that prior to initiating a release:

1. Two independent samples are analyzed in accordance with Specification 3.9 and;

2. Two independent data entry checks for release rate calculations and valve lineup verifications of the effluent pathway are conducted.

Otherwise, suspend release of radioactive effluents by this pathway.

(b) Effluent releases may continue for up to 14 days, provided that prior to batch release of the sump, grab samples are collected and analyzed for gross radioactivity (beta and/or gamma) at a lower limit of detection of at least 10 ' pCi/ml.

(c) Effluent releases may continue for up to 14 days provided flow rate is estimated at least once per four hours during actual releases.

(d) Effluent releases may continue for up to 30 days provided that daily grab samples are collected and analyzed for gross radioactivity (beta and/or gamma) at a lower limit of detection of *at least 10 7 pCi/ml.

*At all times.

**Flow determined from number of hydro units operating; if hydro is not operating, leakage flow, which is measured periodically, is used.

3.5-34

TABLE 3.5.5-2 GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION

OPERATING CONDITIONS

A B INSTRUMENT MINIMUM OPERATOR ACTION IF

OPERABLE MINIMUM NUMBER OF CHANNELS APPLICABILITY OPERABLE CHANNELS IS

NOT MET

1. Waste Gas Holdup Tanks

Hydrogen Monitor 1 (2) (c)

2. Unit Vent Monitoring System

a. Noble Gas Activity Monitor 1 (3) * (a) (RIA - 45)***

b. Iodine Sampler 1 (3) * (e) c. Particulate Sampler 1 (3) * (e) d. Effluent Flow Rate 1 (3) * (b)

Monitor e. Sampler Flow Rate 1 (3) * (f)

Monitor

3. Interim Radwaste Building Ventilation Monitoring

System

a. Noble Gas Activity 1 * (d) Monitor (RIA - 53)

b. Iodine Sampler 1 (e) c. Particulate Sampler 1 * (e) d. Effluent Flow Rate

Monitor 1 (b) e. Sampler Flow Rate

Monitor 1 * (f)

( ) Per station.. At all times.

** During waste gas holdup system operation. 0 Providing alarm and automatic termination of release.

3.5-35

TABLE 3.5.5-2 NOTES

(a) Effluent releases from waste gas tanks or containment purges may continue for up to 14 days provided that prior to initiating a release:

1. Two independent samples are analyzed and;

2. Two independent data entry checks for release rate calculations and valve lineup verifications of the effluent pathway are conducted and;

Effluent release from ventilation system or condenser air ejectors may continue for up to 30 days provided that grab samples are taken once per 8 hours and these samples are analyzed for gross activity (beta and/or gamma) within 24 hours. Otherwise, suspend release of radioactive effluents via this pathway.

(b) Effluent releases may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours.

(c) If the hydrogen monitoring instrumentation is inoperable, it shall be returned to an operable status within 7 days. Otherwise, the waste gas tank on service shall be sampled every 3 days and all other waste gas tanks not in service shall be sampled every 7 days.

(d) Effluent releases may continue for up to 30 days provided grab samples are taken once per 8 hours and these samples are analyzed for gross activity (beta and/or gamma) within 24 hours.

(e) Effluent releases may continue for up to 30 days, provided samples are continuously collected with auxiliary sampling equipment for periods of 7 days and analyzed within 48 hours of the end of sample collection.

(f) Alarms indicating low flow may be substituted for flow measuring devices.

3.5-36

3.9 RADIOACTIVE LIQUID EFFLUENTS

Applicability

Applies at all times to the controlled release of all liquid waste.discharged from the site which may contain radioactive materials.

Objective

To establish conditions for the controlled release of radioactive liquid effluents. To implement the requirements of 10 CFR 20, 10 CFR 50.36a, Appendix A to 10 CFR 50, Appendix I to 10 CFR 50, 40 CFR 141 and 40 CFR 190.

Specification

3.9.1 Concentration

a. The concentration of radioactive material released at anytime from the site to unrestricted areas (denoted in Figure 3.9-1) shall be limited to the concentration specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases the concentration shall be limited to 2 x 10 4 pCi/ml total activity.

b. If the concentration of radioactive material released from the site to unrestricted areas exceed the above limits, promptly restore concentration to within the above limits and notify the regional NRC Office of Inspection and Enforcement of the occurrence.

3.9.2 Dose

a. The dose or dose commitment to an individual from radioactive materials in liquid effluents to unrestricted areas shall be limited to:

1) during any calendar quarter:

< 4.5 mrem to the total body

< 15 mrem to any organ and;

2) during any calendar year:

< 9 mrem to the total body

< 30 mrem to any organ.

b. If the calculated dose from the release of radioactive materials in liquid effluents exceeds any of the above limits, in lieu of any other report required by Section 6.6.2, a report shall be

3.9-1

submitted within 30 days to the regional NRC Office of Inspection and Enforcement which includes the following:

1. Cause(s) for exceeding the limit(s)

2. Corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and during the remaining quarters of the calendar year, so that the cumulative dose or dose commitment to an individual from these releases is within 9 mrem to the total body and 30 mrem to any organ.

3. Results of radiological analyses of the drinking water source and the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR 141.

3.9.3 Liquid Waste Treatment

a. The appropriate subsystems of the liquid radwaste treatment system shall be used to reduce the radioactive materials in liquid waste prior to their discharge, if the projected dose due to liquid effluent releases to unrestricted areas, when averaged over 31 days would exceed 0.18 mrem to the total body or 0.6 mrem to any organ.

b. If radioactive liquid waste is discharged without treatment and in excess of the above limit, a report shall be submitted within 30 days to the regional NRC Office of Inspection and Enforcement which includes the following:

1. Cause of equipment or subsystem inoperability.

2. Corrective action to restore equipment and prevent recurrence.

3.9.4 Chemical Treatment Ponds

a. The quantity of radioactive material in the Chemical Treatment Ponds (CTP) shall be limited so that, for all radionuclides identified, excluding noble gases, the sum of the ratios of activity (in curies) to the limits in 10 CFR 20, Appendix B, Table II, Column 2 shall not exceed 1.7 x 10s.

7I Ai < 1.7 x 105

j Cj

where Aj = pond inventory limit for single radionuclide 'j' (curies)

Cj = 10 CFR 20, Appendix B, Table II, Column 2, concentration for single radionuclide 'j' (curies)

3.9-2

b. After a primary to .secondary leak is detected, the initial batch of used Powdex resin shall not be transferred to the CTP. No batch of used powdex resin shall be transferred to the CTP unless the sum of the ratios of the activity of the radionuclides identified in the preceeding batch from any powdex cell in the same unit is less than 0.1% of the limit identified in 3.9.6.a.

2 -< 10 j Aj

where Qj = radionuclide activity in the batch

Aj = pond inventory limit for radionuclide 'j'

C. The radionuclide inventory per batch of used powdex resin transferred, averaged over the transfers of the previous 13 weeks, shall not exceed 0.01% of the pond radionuclide inventory limit. If this average.exceeds 0.01% of the pond radionuclide inventory limit, then a report will be submitted within 30 days to the Regional NRC Office of Inspection and Enforcement describing the reason or reasons for exceeding the objective and plans for future operation. Decay of radionuclides may be taken into account in determining inventory levels.

Qj + Qj + Qj + Qjn < .01% x Aj 1 2 (n-1)

n

where Qj = activity of radionuclide 'j' in the batch

n = number of batches transferred to the chemical treatment ponds during.the previous 13-week period.

Bases

The concentration specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

The dose specification is provided to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I

3.9-3

that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.

For units with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned among the units sharing that system.

The requirements that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and design objective Section 11.D of Appendix A to 10 CFR Part 50.

The inventory limits of the chemical treatment ponds are based on limiting the consequences of an uncontrolled release of the pond inventory. The short term rate limit (2 mrem/hr) of 10CFR20.105 is applied to 10CFR20.106 in the following expression:

AJ x 106 pCi x gal < 2 mrem/hr x 8760 hr 1.3 x 10 gal curie 3786 ml 500 mrem/yr yr

Cj Aj < 1.7 x 10s Cj

where Aj = pond inventory limit for radionuclide 'j' (curies)

Cj = 10CFR20 Appendix B, Table II, Column 2 concentration for radionuclide 'j'

1.3 x 106 gal = estimated volume of smaller chemical treatment pond

The batch limits provide assurance that activity input to the CTP will be minimized.

3.9-4

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OCONEE NUCLEAR STATION Figure 3.9-1

3.10 RADIOACTIVE GASEOUS EFFLUENTS

Applicability

Applies at all times to the controlled release of all gaseous waste discharged from the site which may contain radioactive materials.

Objective

To establish conditions for the controlled release of radioactive gaseous effluents. To implement the requirements of 10CFR20, 10CFR50.36a, Appendix A to 10CFR50, Appendix I to 10CFR50, and 40CFR190.

Specifications

3.10.1 Dose Rate

a. The instantaneous dose rate at the site (exclusion area) boundary (Figure 3.10-1) due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:

1. The dose rate limit for noble gases shall be:

< 500 mrem/yr to the total body

< 3000 mrem/yr to the skin and;

2. The dose rate limit for all radioiodines and for all radioactive materials in particulate form and radionuclides other than noble gases with half-lives greater than 8 days shall be < 1500 mrem/yr to any organ.

b. If the dose rate exceeds the above limits, immediately decrease the release rate to within the above limits and notify the regional NRC Office of Inspection and Enforcement of the occurrence.

3.10.2 Dose

a. The air dose due to noble gases released in gaseous effluent from the site shall be limited to the following:

1. During any calendar quarter:

< 15 mrad for gamma radiation < 30 mrad for beta radiation

2. During any calendar year:

< 30 mrad for gamma radiation < 60 mrad for beta radiation

3.10-1

b. The dose to an individual from radioiodines, radioactive materials in particulate form, and radionuclides other than noble gases with half-lives greater than 8 days in gaseous effluents released from the site, shall be limited to the following:

1. During any calendar quarter:

< 22.5 mrem to any organ

2. During any calendar year:

< 45 mrem to any organ.

c. If the calculated dose from these gaseous effluents exceeds any of above limits in lieu of any other report required by Specification 6.6.2, a report shall be submitted within 30 days to the regional NRC Office of Inspection and Enforcement which includes the following:

1. Cause(s) for exceeding the limit(s);

2. Corrective action to be taken to reduce the releases of. these radioactive materials in gaseous effluents during the remainder of the current calendar quarter and during the remaining quarters of the calendar year so that the cumulative dose or dose commitment to an individual from these releases is within 30 mrad for gamma radiation, 60 mrad for beta radiation, and 45 mrem to any organ.

3.10.3 Gaseous Radwaste Treatment

a. The gaseous radwaste treatment system shall be used to reduce the noble gases in gaseous wastes prior to their discharge, if the projected gaseous effluent air dose due to gaseous effluent releases from the site, when averaged over 31 days exceeds 0.6 mrad for gamma radiation and 1.2 mrad for beta radiation.

b. High efficiency particulate filters and charcoal filters where available shall be used to reduce radioactive materials other than noble gases in gaseous waste prior to their discharge when the projected doses due to effluent releases to unrestricted areas when averaged over 31 days would exceed 0.9 mrem to any organ.

c. If radioactive gaseous waste is discharged without treatment for more than 31 days and in excess of the above limits in lieu of any other report required by specification 6.6.2, a report shall be submitted within 30 days to the regional NRC Office of Inspection and Enforcement which includes the following:

1. Cause of equipment or subsystems inoperability

2. Corrective action to restore equipment and prevent recurrence

3.10-2

3.10.7 Explosive Gas Mixture'

a. The concentration limit of hydrogen in the waste gas holdup tanks is 2% by volume.

b. If the concentration of hydrogen in the waste gas hold up tanks exceeds 2% by volume but is less than or equal to 4% by volume, then within 48 hours, reduce the concentration of hydrogen to within the limit in Specification 3.10.4.a.

c. If the concentration of hydrogen in the waste gas holdup tanks exceeds 4% by volume, then immediately suspend all additions of waste gases to the tank and, within 48 hours, reduce the concentration of hydrogen to less than 4% by volume within one hour, and within 48 hours reduce the concentration of hydrogen to within the limit in Specification 3.10.4.a.

3.10.8 Waste Gas Holdup Tanks

a. The quantity of radioactivity contained in each waste gas holdup tank shall be limited to < 3.8 E + 05 curies noble gases (considered as Xe-133).

b. Daily, when radioactive materials are being added to a waste gas holdup tank, the quantity of radioactive material contained in the tank being.filled shall be determined.

c. If the quantity of radioactive material in any waste gas holdup tank exceeds the above limit, promptly suspend all additions of

radioactive material to the tank and within 48 hours, reduce the tank contents to within the above limit.

Bases

Specification 3.10.1 is provided to assure that the dose rate at anytime at the exclusion area boundary from gaseous effluents from all units on the site

will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not

result in the exposure of an individual in an unrestricted area, either within or outside the exclusion area boundary, to annual average concentrations exceeding the limits specified in Appendix B, Table 11 of 10 CFR Part 20 (10 CFR Part 20.106(b)). For individuals who may at times be within the exclusion area boundary, the occupancy of the individual will be sufficiently low to

compensate for any increase in the atmospheric diffusion factor above that for

the exclusion area boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an

individual at or beyond the exclusion area boundary to < 500 mrem/year to

the total body or to < 3000 mrem/year to the skin. These release rate

3.10-3

limits also restrict, at all times, the corresponding thyroid rate above background to an infant via the cow-milk-infant pathway to < 1500 mrem/year for the nearest cow to the plant.

For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.

Specification 3.10.3 is provided to implement the requirements of Appendix I, 10 CFR Part 50. The specification provides the required operating flexibility and at the same time implement the guides set forth in Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable". Surveillance requirements are implemented to meet the requirements of Appendix I. Calculational procedures based on models and data show that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated.

The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials will be consistent with the methodology provided in Regulatory Guide 1.109, "Calculating of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, "Revision I, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors."

Equations in the ODCM are provided for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for radioiodines, radioactive material in particulate form.and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the unrestricted area. The pathways which are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and design objective Section IID of Appendix I to 10 CFR Part 50.

Specification 3.10.7 is provided to assure that the concentration of potentially explosive gas mixtures contained in the waste gas treatment system is maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen below the flammability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3.10-4

Restricting the quantity of radioactivity contained in each waste gas holdup tank provides assurance that in the event of an uncontrolled release of the tanks contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem.

0 3.10-5

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3.11 SOLID RADIOACTIVE WASTE

Applicability

Applies to the processing of radioactive solid waste prior to shipment from the site.

Specification

3.11.1 Solid Radioactive Waste

a. The solid radwaste system shall be operable and used as applicable in accordance with a PROCESS CONTROL PROGRAM for the solidification and packaging of other solid radioactive wastes to ensure meeting the requirements of 10CFR Part 20 and of 10CFR Part 71 prior to shipment of containers of radioactive wastes from the site.

b. If the requirements of 10CFR Part 20 and/or 10CFR Part 71 are not satisfied, suspend shipments of defectively packaged solid radioactive wastes from the site.

c. If the solid radwaste system is inoperable for more than 31 days, in lieu of any other report required by Specification , prepare and submit to the Commission within 30 days pursuant to Specification

a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystem and the reason for inoperability,

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

3. A description of the alternative used for SOLIDIFICATION and packaging of radioactive wastes, and

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

d. The solid radwaste systems shall be demonstrated operable at least once per 92 days by:

1. Operating the solid radwaste system at least once in the previous 92 days in accordance with the PROCESS CONTROL PROGRAM, or

2. Verification of the existence of a valid contract for SOLIDIFICATION to be performed by a contractor in accordance with a PROCESS CONTROL PROGRAM.

3.11.2 Process Control Program

The Process Control Program shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive waste (e.g., filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions.

3.11-1

1. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM.

2. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least 3 consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.8, to assure SOLIDIFICATION of subsequent batches of waste.

Bases

The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require processing and packaging prior to being shipped offsite. This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL .PROGRAM may include, but are not limited to waste type, waste pH, waste/liquid/ solidification agent/catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times.

3.11-2

4.1 OPERATIONAL SAFETY REVIEW

Applicability

Applies to items directly related to safety limits and limiting conditions for operation.

Objective

To specify the frequency and type of surveillance to be applied to unit equipment and conditions.

Specification

4.1.1 The frequency and type of surveillance required for Reactor Protective System and Engineered Safety Feature Protective System instrumentation shall be as stated in Table 4.1-1.

4.1.2 The frequency and type of surveillance required for selected equipment shall be as stated in Table 4.1-2.

4.1.3 Required sampling should be performed as detailed in Table 4.1-3.

4.1.4 The frequency and type of surveillance required for radioactive effluent monitoring instrumentation shall be as stated in Table 4.1-4.

4.1.5 Using the Incore Instrumentation System, a power map shall be made to verify expected power distribution at periodic intervals not to exceed ten effective full.power days.

Bases

Failures such as blown instrument fuses, defective indicators, and faulted amplifiers which result in "upscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system. Furthermore, such failures are, in many cases, revealed by alarm or annunciator action. Comparison of output and/or state of independent channels measuring the same variable supplements this type of built-in surveillance. Based on experience in operation of both conventional and nuclear systems, when the unit is in operation, the minimum checking frequency stated is deemed adequate for reactor system instrumentation.

Calibration is performed to assure the presentation and acquisition of accurate information. The nuclear flux (power range) channels amplifiers are calibrated (during steady-state operating conditions) when core thermal power exceeds indicated neutron power by more than two percent. During non-steady-state operation, the nuclear flux channels amplifiers are calibrated daily to compensate for instrumentation drift and changing rod patterns and core physics parameters.

I4.1-1

Channels subject only to "drift" errors induced within the instrumentation itself can tolerate longer intervals between calibrations. Process system instrumentationerrors induced by drift can be expected to remain within acceptable tolerances if recalibration is performed at the intervals specified.

Substantial calibration shifts within a channel (essentially a channel failure) are revealed during routine checking and testing procedures. Thus, the minimum.calibration frequencies set forth are considered acceptables.

Periodic use of the Incore Instrumentation System for power mapping is sufficient to assure that axial and radial power peaks .and the peak locations are controlled in accordance with the provisions of the Technical Specifications.

REFERENCE

(1) FSAR, Section 7.1.2.3.4

4.1-2

TABLE 4.1-1 (Continued)

Channel Description Check Test Calibrate Remarks

20. Reactor Building Spray NA MO NA System Logic

21. Reactor Building Spray NA MO AN System Analog Channel Reactor Building High Pressure

22. Pressurizer Temperature ES NA AN

23. Control Rod Absolute ES(l) NA AN(2) (1) Check with Relative Position Indicator. Position (2) Calibrate rod misalignment channel.

24. Control Rod Relative ES(l) NA AN(2) (1) Check with Absolute Position Indicator. Position (2) Calibrate rod misalignment channel.

25. Core Flood Tanks Un

a. Pressure ES NA AN b. Level ES NA AN

26. Pressuizer Level ES NA AN

27. Letdown Storage Tank DA NA AN

28. Delete

29. High and Low Pressure NA NA AN Injection Systems Flow Channels

TABLE 4.1-3

Minimum Sampling Frequency

Item Check Frequency

1. Reactor Coolant a. Gamma Isotopic Analysis a. 3 times/week* j b. Radiochemical Analysis for Sr 89, 90 b. Monthly* c. Tritium c. Monthly* d. Gross Beta Activity (1) d. 5 times/week* e. Chemistry (Cl, F and 02) e. 5 times/week* f. Boron Concentration f.. 2 times/week**

g. Gross Alpha Activity g. Monthly* h. E Determination (2) h. Semi-annually

2. Borated Water Storage Boron Concentration Weekly* and after each Tank Water Sample makeup

3. Core Flooding Tank Boron Concentration Monthly* and after each makeup

4. Spent Fuel Pool Water - Boron Concentration Monthly*** and after Sample each makeup

5. Secondary Coolant a. Gross Beta Activity a. Weekly* b. Gamma Isotopic Analysis (3)

6. Concentrated Boric Acid Boron Concentration Twice weekly* Tank

*Not applicable if reactor is in a cold shutdown condition for a period exceeding the sampling frequency. **Applicable only when fuel is in the reactor. ***Applicable only when fuel is in wet storage in the spent fuel pool.

TABLIE 4.1-3 Continuted

li ill Ilm Sanll I io F ie r!!q!IIIkcY

I t -II Check k'reqjoency -1lbAilsi o at

7. CA'Illsae T:SLTaji, a Prncipl Gimia Eni~trs a. Crab Sample prior t~o re- a. Gammsa acie<5l7[i/il Cm d floigj~ Mci Ltriiig - iniltidigig D~issolve~d Noble lease of each batch (11). Diissolve~d Gases <10 5pci-/mi1 Tatik , I.Lstimiry-llot Shower Gases 1-131 (106 mCi/il

b. lladiocheiinica I Analysis b. Mlonthly fronk all compost ted 1). <5x108

1Ci/mnl for Sr's Sr H9, 90O, Fe-55, P-32 batches (10-6 pCi/mi for Fe, P

C. Tfri L i In C. mloith Iy C. <10-5 iCi/mIl

di. Gross A lpha Activity d. Monhily d . (10 7 pCi/mIl

6. t1nkii Vvill. Samp Iisig a. I oi je Sj)CctraIIm 4) a. Coot mluoils weekly samlple (8) a. <10-1 lj Ci/cc (1-133)

I if, I odels Waste Gasi lDcay(4 o12pic(-3)

Taliks, leacw.1 Bilioig b. larLicailates4

<02pic(-1) Ios~cs, Aiiihiary llil~dlig (7) (8)~lI ic

Vvuii. IA gul, Speoti Fill 1 (1) Principl1 Gammja Emi tiers (1) Weekly Compos i te~ (1) <10 0ic/, l I V..-ni hat ion, Air

F . .s)(2) Gross Al pha Activy (2) Monthly oil a sample (2) <10 '' i/cc of onle week diirat ion

(3) HaicenclAnalysis .(3) Quarterly Composite (3) <10-1 piCi/cc

SH 89, 90O

U:. Gases l'y prinicipal Gaiiuna C. Weekly Grab Sample C. <10~ p Ci/cc Emi t Lers

d. Tlri I ion ItilI. Weekly Grab Sample d. <106 piCi/cc

%ivi icwg uiisis arce liciuig mottc to tank.

TABILE 4.1-:3 Cont jiiie

fti liilillil ai lll i iaJ r ll! ,

Lower Ljiit of' Detecti'on 5

I 1cm Check. FreijuenLcyofabAays rWse

tbi. Waste ("at Decay a. Pinjcipal Gaimma cittLers () a. Grab Sample prior to re- a. <:04 pCi/cc (gasUS)

Ta iik lease of each batch <I01 pjCi/cc (particulates and iodiiies)

b,. TriLiim 1). Grab, sample prior to 11. <10 p lci/cc release of each batchi

81 -- ',. 1~c HIS li 1.11 t ig a. Pr incipal I Cummia Emi Iteis ( a. Grab Sampl ec achi purge as. <10- [, Ci/cc (gasesi <to 10 pCi/cc (particulates. aiid i odijlies)

II. Tiit i um b. Grab Samsple e achi purge b . (10-6 pi:i/cc

9. -i Hydro Diam Mleasiire Leakage Flow~ RaLe Aiiiial I iy I)i~ I o L i oi FI ow~

I.Ciii'Itollser Air Ejecti ir Ileasiire jodinei Parit ion One Lille if and whenl

Pa tLi Liol fi acto, lac tor ini Codietise r pr imiary to secondary leaks develop

II - i..rsi Recci vioig Tanks Painicijile Ganutia Ei~lItrs Grab Samiple prior to includiiig DissolvedI Noble release of each batch (11)

wa: I, ( lii I C I let i o Piicil Ganuila Emit ters a. - lliiliy from Compositec sample a. G aiiia Nisciides

ha;i o l II hwut <5 X 10 7 pici/mi Dissolved Cases '10 ici /Il 1 13J <10 6 piCi/mtl

l'ABLIE 4. 1-3 Colt~a id

Hiliimu Sampling Frii leEre

Lower LiiiA of IeetAion

1 tcal Che.ck Fr~onyof 1, , *Anitye foj Va ite

12 cooit. d1 bi. liddiocliemical Aiialysis 1j. Monthly frout composite sample Lb. <5 x 10 8 OCi/lil

Sr-H9, Sr-90, lVe-55, P-32 for sr's <06 piCi/Iai

for Fe, P

Tr i t i ma C . Mlonithly from. colupos i~e sample C . <10 J pCi html.

di. Gross Ailus Activity dl. tuLlily froml compoitec Sammple d . <10 7 piCi/mi

TABLE 4.1-3 NOTES

(1) When radioactivity level is greater than 10 percent of the limits of

Specification 3.1.4, the sampling frequency shall be increased to a mini

mum of once each day.

(2) E determination will be started when gross beta activity analysis indicates

greater than 10 pCi/ml and will be redetermined for each 10 pCi/ml increase

in gross beta activity analysis thereafter. A radiochemical analysis for

this purpose shall consist of a quantitiative measurement of 95 percent of

the radionuclides in the reactor coolant with half lives greater than 30

minutes. This is expected to consist of gamma isotopic analysis of the

primary coolant, including dissolved gaseous activities, radiochemical

analysis for Sr-89 and Sr-90, and tritium analysis.

(3) When gross beta activity increases by a factor of two above background,

an iodine analysis will be made and performed thereafter when the gross

beta activity increases by 10 percent.

(4) Samples shall be changed at least once per 7 days and analyses shall be

completed within 48 hours after changing (or after removal from sampler).

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 percent of RATED THERMAL POWER in one hour and analyses shall be

completed within 48 hours of changing. When samples collected for 24

hours are analyzed, the corresponding LLD's may be increased by a factor

of 10.

(5) The LLD is the smallest concentration of radioactive material in a

sample that will be detected with 95% probability with 5% probability of

falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical

separation):

LLD _4.66 sb

E * V * 2.22 x 106 Y exp (-MAt)

Where:

LLD is the "a priori" lower limit of detection as defined above

(as microcurie per unit mass or volume),

s is the standard deviation of the background counting rate or

0 the counting rate of a blank-sample as appropriate (as counts

per minute),

E is the counting efficiency (as counts per transformation),

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

4.1-14

TABLE 4.1-3 NOTES

2.22 x 106 is the number of transformations per minute per microcurie,

Y is the fractional radiochemical yield (when applicable),

X is the radioactive decay constant for the particular radionuclide, and

At is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).

The value of s b used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the backround counting rate or of the counting rate of the blank samples (as appropriate) rather than go an unverified theoretically predicted variance. Typical values of E, V, Y and At shall be used in the calculation.

(6) The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

(7) The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, CO-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

(8) The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specification 3.10.1, 3.10.2.a and 3.10.2.b.

(9) A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released.

(10) To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.

4.1-15

TABLE 4.1-3 NOTES

(11) A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then

thoroughly mixed, by a method described in the ODCM, to assure representative sampling.

4.1-16

TABLE 4.1-4

RADIOACTIVE EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL

INSTRUMENT CHECK CHECK CALIBRATION TEST

1. Liquid Radwaste Effluent Line a. Effluent Line Monitor DA* PR AN(3) QU(1) b. Flow Rate Measurement Device *(5) NA AN NA

2. Turbine Building Sump a. Sump Monitors DA* MO AN(3) QU(2) b. Flow Rate Measurement Device (6) NA NA NA NA

3. Waste Oil Collection Basin Outflow *(5) NA AN NA Sampler and Flow Rate Monitor

4. Waste Gas Holdup Tanks a. Hydrogen Monitor DA** NA QU(4) NA

5. Unit Vent Monitoring a. Noble Gas Activity Monitor DA* MO AN(3) QU(2)

b. Iodine Sampler DA* NA NA NA

c. Particulate Sampler DA* NA NA NA

d. Effluent Flow Rate Monitor DA* NA AN NA

e. Minimum Flow Device DA* NA AN NA

TABLE 4.1-4 CONTINUED

RADIOACTIVE EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL

INSTRUMENT CHECK CHECK CALIBRATION TEST

6. Interim Radwaste Building Ventilation Monitoring a. Noble Gas Activity Monitor DA* MO AN(3) QU(2)

b. Iodine Sampler DA* NA NA NA

c. Particulate Sampler DA* NA NA NA

d. Effluent Flow Rate Monitor DA* NA AN NA

e. Minimum Flow Device DA* NA AN NA

---- *During releases via this pathway. **During waste gas holdup system operation (treatment for primary system of gases)

Frequency Notation

DA - Daily MO - Monthly PR - Completed prior to each release

QU - Quarterly AN - Annually NA - Not Applicable

TABLE 4.1-4 (Continued)

TABLE NOTATION

(1) The channel functional test shall demonstrate automatic isolation of this pathway if the instrument indicates measured levels above the alarm/trip setpoint and;

(2) The channel functional test shall demonstrate that control room alarm annuciation occurs if the instrument indicates measured levels above the alarm/trip setpoint.

(3) The initial channel calibration for radioactivity measurement instrumentation shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) using standards that have been obtained from suppliers that participate in measurements assurance activities with NBS. These standards should permit calibrating the system over its intended range of energy and rate capabilities. For subsequent channel calibration, sources that have been related to the initial calibration shall be used, at intevals given. This can normally be accomplished during refueling outages.

(4) The channel calibration shall include the use of standard gas samples containing a nominal:

1. One volume percent hydrogen, balance nitrogen; and

2. Four volume percent hydrogen, balance nitrogen.

(5) The channel check shall consist of verifying indication of flow during periods of release. Channel check shall be made at least once per 24 hours on days on which continuous, periodic, or batch releases are made.

(6) No surveillance is required because pump curves and pump run times are used to compute total flow.

(III

4.1-19

4.11 RADIOLOGICAL ENVIRONMENTAL MONITORING

Applicability

Applies to the surveillance of the station environ for radiation and radioactive materials attributable to station operation and effluent releases.

Specification

4.11.1 Radiological Environmental Monitoring Program

a. The radiological environmental monitoring samples shall be collected in accordance with Table 4.11-1 and shall be analyzed pursuant to the requirements of Tables 4.11-1, 4.11-2, and 4.11-3.

b. If the radiological environmental monitoring program is not conducted as required., a description of the reason for not conducting the program as required and plans to prevent a recurrence shall be included in the Annual Radiological Environmental Operating Report. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, or to malfunction of automatic sampling equipment. If the latter, every effort shall be made to complete corrective action prior to the end of the next sampling period.

C. If milk samples become permanently unavailable from any of the required sample locations, a report shall be submitted within 30 days to the regional NRC Office of Inspection and Enforcement which identifies the cause of the unavailability of samples and identifies locations for obtaining replacement samples. The locations from which samples were unavailable may then be deleted from Table 4.11.1 provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations, if available.

4.11.2 Land Use Census

a. A land use census shall be conducted and shall identify the location of the nearest milk animal and the nearest residence in each of the 16 meterological sectors within a distance of five miles. Broad leaf vegetation sampling shall be performed at the site boundary in the direction sector with the highest D/Q in lieu of the garden census.

b. If a land use census identifies a location which yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.21, then a report shall be submitted within 30 days to the regional NRC Office of

Inspection and Enforcement identifying the new location.

4.11-1

C. If a land use census identifies a location which yields a calculated dose or dose commitment (via the same exposure pathway) greater than at a location from which samples are currently being obtained pursuant to Specification 4.11.1, then a report shall be submitted within 30 days to the NRC Office of Inspection and Enforcement identifying the new location. The new location. shall be added to the radiological environmental monitoring program within 30 days. The sampling location having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted.

d. The land use census shall be conducted annually between the dates of April 1 and October 1, by door to door survey, aerial survey, or by consulting-local authorities..

4.11.3 Interlaboratory Comparison Program

a. Analyses .shall be performed on radioactive materials supplied

as part of an Interlaboratory Comparison Program which has been

approved by the NRC.

b. If these analyses are not performed as required, report corrective actions in the Annual Radiological Environmental Ope

rating Report.

Bases

The environmental monitoring program required by this specification provides

measurements of radiation and of radioactive materials in those exposure path

ways and for those radionuclides which lead to the highest potential radiation

exposures of individuals resulting from the station operation. This monitoring

program thereby supplements the radiological effluent monitoring program by

verifying that the measurable concentrations of radioactive materials and

levels of radiation are not higher than expected on the basis of the effluent

measurements and modeling of program will be effective for at least the first

three years of commercial operation. Following this period, program changes

may be initiated based on operational experience.

The detection capabilities required by Table 4.11-1 are state-of-the-art for

routine environmental measurements in industrial laboratories. The specified

lower limits of detection correspond to less than the 10CFR50, Appendix I,

design objective dose-equivalent of 15 mrem/year for atmospheric releases to

the most sensitive organ and individual.

The land use census specification is provided to assure that changes in the

use of unrestricted areas are identified and that modifications to the monitor

ing program are made if required by the results of this census.

The requirement for participation in an Interlaboratory Comparison Program is

provided to assure that independent checks on the precision and accuracy of the

measurements of radioactive material in environmental sample matrices are performed as part of a quality assurance program for environmental monitoring in

order to demonstrate that the results are reasonably valid.

4.11-2

TABLE 4.11-1

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM

Exposure Pathway Number of Sampling and Type and Frequency

and/or Sample Sample Locations** Collection Frequency of Analysis**

1. AIRBORNE

a. Radioiodine 5 Locations Continuous operation of Radioiodine canister.

and Particu- sampler with sample col- Gamma isotopic analy

lates lection as required by sis for 1-131 or each dust loading but at sample. Particulate least once per 7 days. sampler. Gamma isotopic

analysis on each sample.

2. DIRECTION RADIATION

40 Locations Continuous integration Gamma dose on each with collection at dosimeter. least once per 92 days.

3. WATERBORNE

a. Surface 2 Locations Composite* sample col- Gamma isotopic analysis lected over a period of each composite sample of < 31 days. by location. Tritium

analyses of composite sample at least once per 92 days.

*Sample locations are identified in the ODCM. **Frequency of analysis stated only if different from collection frequency.

TABLE 4.11-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM

Exposure Pathway Number of Sampling and Type and Frequency

and/or Sample Sample Locations** Collection Frequency of Analysis***

b. Drinking 3 Locations Composite* sample col- Gross beta and gama lected over a period isotopic analysis of of < 31 days. each composite sample.

Tritium analysis of composite sample at least once per 92.days.

c. Sediment from 2 Locations - At least once per 184 Gamma isotopic analysis

Shoreline days. of each sample.

4. INGESTION

a. Milk 4 Locations At least once per 15 Gamma isotopic and 1-131 days when animals are analysis of each sample. on pasture; at least once per 31 days at other times.

b. Fish 2 Locations At least once per 184 days. One sample of each of Gamma isotopic analysis the following species: on edible portion. 1. Bass 2. Catfish

c. Broad-leaf 2 Locations At least once per 31 Gamma isotopic analysis.

Vegetation days.

*Composite samples shall be collected by collecting an aliquot at intervals not exceeding 2 hours.

**Sample locations are identified in the ODCM. **Frequency of analysis stated only if different from collection frequency.

TABLE 4.11-2

MAXIMIM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)ac

Airborne Particulate Broadleaf

Water or Gas Fish Mile Vegetation Sediment

Analysis (pCi/1) (pCi/ma) (pCi/kg,wet) (pCi/1) (pCi/kg,wet) (pCi/kg,dry)

gross beta 4

3 H 2000

5 4Mn 15 130

59Fe 30 260

5 8 ,6 0 Co 15 130

65 3. 260 Zn 30

95Zr- 95Nb 30

b131 15 7 x 10- 2 1 60

I

134,137 15,18 5,6 x 10-2 130,150 15,18 60, 80 150, 180 Cs

140 60 60

140 15 15

TABLE 4.11-2 (Continued)

TABLE NOTATION

a - The LLD is the smallest concentration of radioactive material in a sample with a 95% probability of detection and with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation):

4.66 s LLD = b

E * V * 2.22 * Y * exp(-Adt)

where

LLD is the lower limit of detection as defined above (as pCi per unit mass or volume)

s is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute)

E is the counting efficiency (as counts per transformation)

V is the sample size (in units of mass or volume)

2.22 is the number of transformation per minute per picocurie

Y is the fractional radiochemical yield (when applicable)

X is the radioactive decay constant for the particular radionuclide.

At is the elapsed time between sample collection (or end of the sample collection period) and time of counting

The value of s used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide deter

mined by gamma-ray spectrometry, the background shall include the

typical contributions of other radionuclides normally present

in the sampels (e.g., potassium-40 in milk samples). Typical

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

4.11-6

TABLE 4.11-2 (Continued)

TABLE NOTATION

The LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement*.

Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interferring nuclides, or other uncontrollable circumstances, may render these LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

b - LLD for gamma isotopic analysis for 1-131 in drinking water samples. Low level 1-131 analysis on drinking water will not be routinely performed because the calculated dose from 1-131 in drinking water at all locations is less than 1 mrem per year. Low level 1-131 analyses will be performed if abnormal releases occur which could reasonably result in > 1 pCi/liter of 1-131 in drinking water. For low level analyses of 1-131 an LLD of 1 pCi/liter will be achieved.

c - Other peaks which are measurable and identifiable, together with the radionuclides in Table 4.11-2, shall be identified and reported.

*For a more complete discussion of the LLD, and other detection limits, see the following:

(1) HASL Procedures Manual, RASL-300 (revised annually). (2) Currie, L.A., "Limits for. Qualitative Detection and Quantitative

Determination - Application to Radiochemistry" Anal. Chem. 40, 586-93 (1968).

(3) Hartwell, J. K., "Detection Limits for Radioisotopic Counting Techniques," Atlantic Richfield Hanford Company Report (ARH-2537 June 22, 1972).

4.11-7

TABLE 4.11-3

REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES

Reporting Levels

Broadleaf

Water Airborne Particulate Fish Milk Vegetatation

Analysis (pCi/) or Gases (pCi/m) (pCi/Kg,wet) (pCi/1) (pCi/Kg,wet)

H-3 2 x 104 *

Mn-54 1 x 103 3 x 104

Fe-59 4 x 102 1 x 10

Co-58 1 x 103 3 x 104

00 Co-60 3 x 102 1 x 10

Zn-65 3 x 102 2 x 10

Zr-Nb-95 4 x 102

1-131 2** 1.0 3 1 x 102

Cs-134 30 10 1 x 103 60 lx 103

Cs-137 50 20 2 x 103 70 2 x 103

Ba-La-140 2 x 102- 3 x 102

*For drinking water samples. This is 40 CFR Part 141 value.

*If low level 1-131 analyses are performed.

4.21 DOSE CALCULATIONS

Applicability

Applies to the projected and cummulative dose contributions from all radioactive liquid and gaseous effluents.

Specification

4.21.1 Dose From All Sources

a. The dose or dose commitment to any member of the public due to releases of radioactivity and radiation from uranium fuel cycle sources shall not exceed 25 mrem to the whole body or any organ (except the thyroid, which shall not exceed 75 mrem) over a period of 12 consecutive months and shall be calculated using the methodology contained in the Offsite Dose Calculation Manual.

b. If the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeds twice the limits of Specifications 3.9.2, 3.10.2.a, or 3.10.2.b, then limit subsequent releases such that Specification 4.21.1.a is met and within 30 days submit a report and in lieu of any other report required by Specification 6.6.2 to the regional NRC Office of Inspection and Enforcement which includes an analysis which demonstrates that radiation exposure to the critical individual from uranium fuel cycle sources are within 40 CFR Part 190 requirements.

4.21.2 Dose Due to Liquid Effluents

a. Monthly, cummulative dose contributions from liquid effluents shall be determined in accordance with the Offsite Dose Calculation Manual.

4.21.3 Dose Due to Gaseous Effluents

a. Monthly, cummulative dose contributions from gaseous effluents shall be determined in accordance with the Offsite Dose Calculation Manual.

4.21-1

6.1.2 Technical Review and Control

6.1.2.1 Activities

a. Procedures required by Technical Specification 6.4 and other procedures which affect station nuclear safety, and changes (other than editorial or typographical changes) thereto, shall be preapred by a qualified individual/organization. Each such procedure, or procedure change, shall be reviewed by an individual/group other than the individual/group which prepared the procedure, or procedure change, but who may be from the same organization as the individual/group which prepared the procedure, or procedure change. Such procedures and procedure changes may be approved for temporary use by two members of the station staff, at least one of whom holds a Senior Reactor Operator's License on the unit(s) affected. Procedures and procedure changes shall be approved prior to use or within seven days of receiving temporary approval for use by the station Manager; or by the Operating Superintendent, the Technical Services Superintendent or the Maintenance Superintendent, as previously designated by the station Manager.

b. Proposed changes to the Technical Specifications shall be prepared by a qualified individual/organization. The preparation of each proposed Technical Specificationschange shall be reviewed by an individual/group other. than the individual/group which prepared the proposed change, but who may be from the same organization as the individual/group which prepared the proposed change. Proposed changes to the Technical Specifications shall be approved by the station Manager.

c. Proposed modifications to station nuclear safety-related structures, systems and components shall be designated by a qualified individual/ organization. Each such modification shall be reviewed by an individual/ group other than the individual/group which designed the modification, but who may be from the same organization as the individual/group which designed the modification. Proposed modifications .to station nuclear to safety-related structures, systems and components shall be approved prior to implementation by the station Manager; or by the Operating Superintendent, the Technical Services Superintendent, or the Maintenance Superintendent, as previously designated by the station Manager.

d. Individuals responsible for reviews perofmed in accordance with 6.1.2.1.a, 6.1.2.1.b, and 6.1.2.1.c shall be members of the station supervisory staff, previously designated by the station Manager to perform such reviews. Each such review shall include a determination of whether or not additional, cross-disciplinary, review is necessary. If deemed necessary, such review shall be performed by the appropriate designated station review personnel.

e. Proposed tests and experiments *which affect station nuclear safety and are not addressed in the FSAR or Technical Specifications shall be reviewed by the station Manager; or by the Operating Superintendent, the Technical Services Superintendent or the Maintenance Superintendent, as previously

designated by the Station Manager.

6.1-2 A 77/77/74 7/20/79

f. Incidents reportable pursuant to Technical Specification 6.6.2.1 and violations of Technical Specifications shall be investigated and a report prepared which evaluates the occurrence and which provides recommendations to prevent recurrence.. Such reports shall be approved by the station Manager and transmitted to the Vice President, Steam Production, or his designee; and to the Director of the Nuclear Safety Review Board.

g. The station Manager shall assure the performance of special reviews and investigations, and the preparations and submittal of reports thereon, as requested by the Vice President, Steam Production.

h. The station security program, and implementing procedures, shall be reviewed at least annually. Changes determined to be necessary as a result of such review shall be approved by the station Manager and transmitted to the Vice President, Steam Production, or his designee; and to the Directed of the Nuclear Safety Review Board.

i. The station emergency plan, and implementing procedures, shall be reviewed at least annually. Changes determined to be necessary as a result of such review shall be approved by the station Manager and transmitted to the Vice President, Steam Production, or his designee; and to the Director of the Nuclear Safety Review Board.

j. The station manager shall assure that an independent fire protection and loss prevention inspection and audit shall be performed annually utilizing qualified off-site personnel and that an inspection and audit by a qualified fire consultant shall be performed at intervals no greater than three years.

k. Unplanned onsite releases of radioactive material to the environs shall be investigated and a report prepared which evaluates the occurrence and which provides recommendations to prevent recurrence. Such reports shall be approved by the station Manager and transmitted to the Vice President, Steam Production, or his designee; and to the Director of the Nuclear Safety Review Board.

1. Proposed changes to the Process Control Program (PCP), Offsite Dose Calculation Manual (ODCM), and radwaste treatment systems shall be prepared by a qualified individual/organization. Each proposed change shall be reviewed by an individual/group other than the individual/group which

prepared the proposed change, but who may be from the same organization as the individual/group which prepared the proposed change. Proposed changes to the PCP,.ODCM, and radwaste treatment systems shall be approved by the station Manager and the System Health Physicist prior to implementation.

6.1.2.2 Records

Records of the above activities shall be maintained.

6.1-3 A 77/77/74 7/20/79

, 6.1.3 Nuclear Safety Review Board

6.1.3.1 Function

The NSRB shall function to provide independent review and audit of designated activities in the areas of:

a. Nuclear power plant operations b. Nuclear Engineering c. Chemistry and radiochemistry d. Metallurgy e. Instrumentation and control f. Radiological safety g. Mechanical and electrical engineering h. Administrative control and quality assurance.practices

6.1.3.2 Organization

a. The Director, members and alternate members of the NSRB shall.be formally appointed by the Vice President, Steam Production, and shall have an

academic degree in an engineering or physical science field; and in addition, shall have a minimum of five years technical experience, of which a minimum of three years shall be in one or more areas given in 6.1.3.1.

b. The NSRB shall be composed of at least five members, including the Director, Members of the NSRB may be from the Steam Production Department, from other departments within the Company or from external to the Company. A maximum of one member of the NSRB may be from the Oconee Nuclear Station staff.

c. . Consultants may be utilized by the NSRB to provide expert advice to the. NSRB, as determined necessary by the Director of the NSRB.

d. Staff assistance may be provided in order to promote the proper, timely

and expeditious performance of its functions.

e. The NSRB shall meet at least once per six months. The period between such meetings shall not exceed eight months.

f. A quorum of the NSRB shall consist of the Director, or his designated alternate, and at leadt two other NSRB members or alternate members. No more than a minority of the quorum shall have line responsibility for operation of Oconee Nuclear Station.

6.1-3a

6.1.3.3 Subjects Requiring Review

The following subjects shall be reported to and reviewed by the NSRB:

a. The safety evaluations for (1) changes to procedures, equipment or systems, and (2) tests or experiments completed under the provisions of 10 CFR 50.59(a)(1) to verify that such actions did not constitute an unreviewed safety question.

b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in 10 CFR 50.59.

c. Proposed tests or experiments which involve an unreviewed safety question as defined in 10 CFR 50.59.

d. Proposed changes in Technical Specifications or the Facility Operating Licenses.

e. Violations of applicable statutes, codes, regulations, orders Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.

f. Significant operating abnormalities or deviations from normal and expected performance of station equipment that affect nuclear safety.

g. Incidents that are the subject of non-routine reports submitted to the Commission.

h. Quality Assurance Department audits relating to station operations and actions taken in response to these audits.

6.1-4

6.1.3.4 Audits

Audits of station activities shall be performed under the cognizance of the NSRB. These audits shall encompass:

a. The conformance of station operation to provisions contained within the Technical Specifications and applicable facility operating license conditions at least once per year.

b. The performance, training and qualifications of the station staff at least once per year.

c. The results of actions taken to correct deficiencies occuring in equipment, structures, systems or methods of operation that affect nuclear safety at least once per six months.

d. The performance of activities required by the quality assurance program to meet the criteria of Appendix B to 10 CFR 50 at least once per two years.

e. The station emergency plan and implementing procedures at least once per two years.

f. The station security plan and implementing procedures at least once per two years.

g. Any other area of station operation considered appropriate by the NSRB or the Vice President, Steam Production.

h. The station fire protection program and implementing procedures at least once per 24 months.

i. The Offsite Dose Calculation Manual and implementing procedures at least once per 24 months.

j. The radiological environmental monitoring program and the results thereof at least once per 24 months.

k. The PROCESS CONTROL PROGRAM and implementing procedures for solidification of radioactive wastes at least once per 24 months.

1. The performance of activities required by the Quality Assurance Program to meet the criteria of Regulatory Guide 4.15, December 1977 at least once per 12 months.

6.1-5

6.1.3.5 Responsibilities and Authorities

a. The NSRB shall report to and advise the Vice President, Steam Production on those areas of responsibility specified in Specifications 6.1.3.3 and 6.1.3.4.

b. Minutes shall be prepared and forwarded to the Vice President, Steam Production, and to the Senior Vice President, Production and Transmission, within 14 days following each formal meeting of the NSRB.

c. Records of activities performed in accordance with Specifications 6.1.3.3. and 6.1.3.4 shall be maintained.

d. Audit reports encompassed by Section 6.1.3.4 shall be forwarded to the Vice President, Steam Production, and to the Senior Vice President, Production and Transmission and to the management positions responsible for the areas audited within 30 days of completion of each audit.

6.1-5a

6.4 STATION OPERATING PROCEDURES

Specification

6.4.1

The station shall be operated and maintained in accordance with approved pro

cedures. Written procedures with appropriate check-off lists and instructions shall be provided for the following conditions:

a. Normal startup, operation, and shutdown of the complete facility and of all systems and components involving nuclear safety of

the facility.

b. Refueling operations.

c. Actions taken to correct specific and foreseen potential malfunc

tions of systems or components involving nuclear safety and radia

tion levels, including responses to alarms, suspected primary system leaks and abnormal reactivity changes.

d. Emergency procedures involving potential or actual release of

radioactivity.

e. Preventive or corrective maintenance which could affect nuclear safety or radiation exposure to personnel.

f. Station survey following an earthquake.

g. Personnel radiation protection procedures.

h. Operation of radioactive waste management systems.

i. Control of pH in recirculated coolant after loss-of-coolant accident. Procedure shall.state that pH will be measured and the addition of appropriate caustic to coolant will commence within 30 minutes after switchover to recirculation mode of core cooling to adjust the pH to a range of 7.0 to 8.0 within

24 hours.

j. Nuclear safety-related periodic test procedures.

k. Long-term emergency core cooling systems. Procedures shall include provision for remote or local operation of system components necessary to establish high and low pressure in

jection within 15 minutes after a line break.

1. Fire Protection Program implementation.

m. Offsite Dose Calculation Manual implementation.

n. Process Control Program implementation.

6.4-1

@ 6.4.2

Quarterly selected drills shall be conducted on site emergency procedures including assembly preparatory to evacuation off site and a check of the adequacy of communications with off-site support groups.

6.4.3

A respiratory protective program approved by the Commission shall be in force.

6.4-2

b. By-product material inventory records.

i. Minutes of Nuclear Safety Review Board Meetings.

j. Training records.

k. Test results, in units of microcuries, for leak tests performed pursuant to Specification 4.16.

1. Radioactive liquid effluent,.gaseous effluent, and gaseous process monitoring instrumentation alarm/trip setpoints.

6.5-2

6.6 STATION REPORTING REQUIREMENTS

6.6.1 Routine Reports

In addition to the applicable reporting requirements of Title. 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Office of Inspection and Enforcement, Region II unless otherwise noted.

6.6.1.1 Startup Report

A summary report of unit startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the facility license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit. Startup reports shall be submitted (1) within 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) nine months following initial criticality, whichever occurs first. If a startup report does not cover all three events, i.e., initial criticality, completion of the startup test program and resumption or commencement of commercial power operation supplementary reports shall be submitted at least every three months until all three events are completed.

6.6.1.2 Monthly Operating Report

Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Management Information and Program Control, U.S. Nuclear Regulatory Commission, Washington, D.C., 20555, with a copy to the appropriate Regional Office, to be submitted by the fifteenth of each month following the calendar month covered by the report.

6.6.1.3 Personnel Exposure and Monitoring Report

Prior to March 1 of each year, a tabulation shall be submitted to the NRC of the

number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man-rem exposure according to work and job functions, e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total body dose received

from external sources shall be assigned to specific major work functions.

6.6.1.4 Radioactive Effluent Release Report

Routine radioactive effluent release reports covering the operating of the unit during the previous 6 months of operation shall be submitted within 60

days after January I and July 1 of each year.

6.6-1

The radioactive effluent release reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the station.

The radioactive effluent release reports shall include a summary of the meteorological conditions concurrent with the release of gaseous effluents during each quarter.

The radioactive effluent release reports shall include an assessment of the radiation doses from radioactive effluents to individuals due to their activities inside the unrestricted area boundary during the report period. All assumptions used in making these assessments (e.g., specific activity, exposure time and location) shall be included in these reports.

The radioactive effluent release reports shall include the following information for all unplanned releases to unrestricted areas of radioactive materials in gaseous and liquid effluents:

a. A description of the event and equipment involved.

b. Cause(s) for the unplanned release.

c. Actions taken to prevent recurrence.

d. Consequences of the unplanned release.

The radioactive effluent release reports shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the station during each calendar quarter. In addition, the unrestructed area boundary maximum noble gas gamma air and beta air doses shall be evaluated. The meteorological conditions concurrent with the releases of effluents shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the Offsite Dose Calculation.Manual.

The radioactive effluent release reports shall include the following information for each type of solid waste shipped offsite during the report period:

a. container volume,

b. total curie quantity (determined by measurement or estimate),

c. principal radionuclides (determined by measurement or estimate),

d. type of waste (e.g., spent resin, compacted dry waste evaporator bottoms),

e. type of container (e.g., LSA, Type A, Type B, Large Quantity), and

f. solidification agent (e.g., cement, urea formaldehyde).

The radioactive effluent release reports shall include any changes to the Offsite Dose Calculation Manual, the Process Control Program, and the radwaste treatment systems made during the reporting period.

6.6-2

6.6.1.5 Radiological Environmental Monitoring

Routine radiological environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.

The annual radiological environmental operating reports shall include summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an.assess-. ment of the observed impacts of the plant operation on the environment. The reports shall also include the results of the land use censuses required by Specification 4.11. If harmful effects are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.

The annual radiological environmental operating reports shall include summarized and tabulated results of the radiological environmental samples required by Specification 4.11 taken during the report period. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

* The initial report shall also include the following: a summary description of the radiological environmental monitoring program including sampling methods for each sample type, size and physical characteristics of each sample type, sample preparation methods, analytical methods, and measuring equipment used; a map of all sampling locations keyed to a table giving distances and directions from one reactor; and the result of land use censuses required by Specification 4.11. Subsequent reports shall describe all substantial changes in these aspects.

6.6-3

b. Thirty-Day Written Reports

The types of events listed below shall be the subject of written reports to the Director, Office of Inspection and Enforcement, Region II, within 30 days of discovery of the event. (Copy to the Director, Office of Management Information and Program Control.)

(1) Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems.

(2) Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or shutdown required by a limiting condition for operation.

(3) Observed inadequacies in the implementation of administrative or procedural controls during operation of a unit which could cause reduction of degree of redundancy provided in the Reactor Protective System or Engineered Safety Feature Systems.

(4) Occurrence of radioactive material contained in liquid or gaseous holdup tanks in excess of that permitted by the limiting condition for operation established in the technical specifications.

(5) An unplanned offsite release of 1) more than 1 curie of radioactive material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radioiodine in gaseous effluents. The report of an unplanned offsite release of radioactive material shall include the following information:

1. A description of the event and equipment involved.

2. Cause(s) for the unplanned release.

3. Actions taken to prevent recurrence.

4. Consequences of the unplanned release.

(6) Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table 4.11-3 when averaged over any calendar quarter sampling period. When more than one of the radionuclides in Table 4.11-3 are detected in the sampling medium, this report shall submitted if:

concentration (1) concentration (2) limit level (1) + limit level (2) + ... > 1.0

When radionuclides other than those in Table 4.11-3 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year objectives of Specifications 3.9 and 3.10. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

6.6-4

* 6.6.3 Special Reports

Special reports shall be submitted to the Director, Office of Inspection and Enforcement, Region II, within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

a. Single Loop Operation Specification 3.1.8

b. Auxiliary Electrical Systems 3.7

c. Radioactive Liquid Effluents, Dose, Specification 3.9.2 Liquid Waste Treatment, Specification 3.9.3 Chemical Treatment Ponds, Specification 3.9.4

d. Radioactive Gaseous Effluents, Dose, Specification 3.10.2 Gaseous Radwaste Treatment, Specification 3.10.3

e. Fire Protection and Detection Systems, Specification 3.17

f. Reactor Coolant System Surveillance, Inservice Inspection, Specification 4.2.4 Reactor Vessel Speciman, Specification 4.2.8

g. Reactor Building Surveillance, Containment Integrated Leak Rate, Specification 4.4.1.1.7 Annual Inspection Specification 4.4.1.4

h. Structural Integrity Surveillance, Tendon Stress, Specification 4.4.2.2 End Anchorage Concrete, Specification 4.4.2.3 Liner Plate, Specification 4.4.2.4.

i. Radiological Effluent and Environmental Monitoring Program, Specification 4.11.1 Land Use Census, Specification 4.11.2

j. Fuel Surveillance Program, Specification 4.13

k. Dose Calculation (40 CFR 190), Specification 4.21

6.6-5

6.7 OFFSITE DOSE CALCULATION MANUAL (ODCM)

6.7.1

The ODCM shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid instrumentation alarm/trip setpoints consistent with the applicable LCO's contained in these Technical Specifications. Methodologies and calculational procedures acceptable to the Commission are contained in NUREG-0133.

The ODCM shall be submitted to the Commission at the time of proposed Radiological Effluent Technical Specifications and shall be subject to review and

approval by the Commission prior to implementation.

6.7.2

Any changes to the ODCM shall be made by either of the following methods:

A. Licensee initiated changes:

.1. Shall be submitted to the Commission by inclusion in the semiannual Effluent Release Report for the period in which the change(s) was made and shall contain:

a. sufficiently detailed information to totally support the

rationale for the change without.benefit of additional or supplemental information. Information submitted should consist.of a package of those pages of the ODCM to be

changed with each page numbered and provided with an

approval and date box, together with appropriate analyses or evaluations justifying the change(s);

b. a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and

c. documentation of the fact that the change has been reviewed in accordance with Technical Specification 6.1.2.1.2 and found acceptable by the station manager and the System Health Physicist.

2. Shall become effective upon review and acceptance by both the station

manager and the System Health Physicist after confirmation of receipt unless otherwise acted upon by the Commission through written notifi

cation to the licensee.

B. Commission initiated changes:

1. Shall be determined by the station manager to be applicable to

the facility after consideration of facility design.

6.7-1

2. The licensee shall provide the Commission with written notification of their determination of applicability including any necessary revisions to reflect facility design.

3. Shall be reviewed by the NSRB at its next regularly scheduled meeting.

6.7-2

6.8 PROCESS CONTROL PROGRAM (PCP)

6.8.1

The PCP shall describe the program of sampling, analysis, and evaluation within

which solidification is assured, consistent with Specification 3.11. The PCP

shall be submitted to the Commission at the time of proposed Radiological

Effluent Technical Specifications and shall be subject to review and approval by

the Commission prior to implementation.

6.8.2

A. Licensee initiated changes to the PCP:

1. Shall be submitted to the Commission by inclusion in the semi

annual Radioactive Effluent Release Report for the period in

which the change(s) was made and shall contain:

a. sufficiently detailed information to totally support the

rationale for the change without benefit of additional or

supplemental information;

b. a determination that the change did not reduce the overall

conformance of the solidified waste product to existing criteria for solid wastes; and

c. documentation of the fact that the change has been reviewed

in accordance with Technical Specification 6.2.2.1.9 and

found acceptable by the Station manager and the System Health Physicist.

2. Shall become effective upon review and acceptance by both the

station manager and the System Health Physicist, unless otherwise

acted upon by the Commission through written notification to the

licensee.

6.8-1

6.9 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous and Solid)

6.9.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid):

1. Shall be reported to the Commission in the Radioactive Effluent Release Report for the period in which the evaluation was reviewed in accordance with Technical Specification 6.1.2.1.2. The discussion of each change shall contain:

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

b. Sufficient detailed information to totally support the reason for the change withodt benefit of additional or supplemental information;

c. A detailed description of the equipment, components and processes involved and the interfaces with other plant systems;

d. An evaluation of the hange which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously

predicted in the license application and amendments thereto;

e. An evaluation of the change which shows the expected maximum exposures to individual in the unrestricted area and to the

general population that differ from those previously estimated in the license application and amendments thereto;

f. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to

be made;

g. An estimate of the exposure to plant operating personnel as a result of the change; and

h. Documentation of the fact that: the change was reviewed in accordance with Technical Specification 6.1.2.1 and found acceptable by the

station manager and the System Health Physicist.

2. Shall become effective upon review and acceptance by the station manager and the System Health Physicist.

6.9-1

ATTACHMENT 2

OCONEE NUCLEAR STATION OFFSITE DOSE CALCULATION MANUAL

OFFSITE DOSE CALCULATION MANUAL

FOR

DUKE POWER NUCLEAR STATIONS

TABLE OF CONTENTS

SECTION Page

INTRODUCTION 111

1.0 RELEASE RATE CALCULATIONS 1-1

1.1 LIQUID EFFLUENTS 1-1 1.2 GASEOUS EFFLUENTS 1-1 1.2.1 Noble Gases 1-1 1.2.2 Radioiodines, Particulates, and Others 1-2

2.0 RADIATION MONITORING SETPOINTS 2-1

2.1 LIQUID MONITORS 2-1 2.2 GAS MONITORS 2-1

3.0 DOSE CALCULATIONS 3-1

3.1 DOSE MODELS FOR MAXIMUM EXPOSED INDIVIDUAL 3-1 3.1.1 Liquid Effluent 3-1 3.1.2 Gaseous Effluents 3-2 3.1.2.1 Noble Gases 3-2 3.1.2.2 Radioiodines, Particulates, and Others 3-3 3.1.3 Direct Radiation 3-7 3.2 SIMPLIFIED DOSE PROJECTIONS 3-7 3.3 FUEL CYCLE CALCULATIONS 3-8 3.3.1 Milling 3-8 3.3.2 Conversion 3-8 3.3.3 Enrichment 3-8 3.3.4 Fuel Fabrication 3-8 3.3.5 Nuclear Power Production 3-9 3.3.6 Fuel Reprocessing 3-9

Appendix A - Oconee Nuclear Station Site Specific Information

Appendix B - McGuire Nuclear Station Site Specific Information

Appendix C - Catawba Nuclear Station (Later) Site Specific Information

. Revision 1

LIST OF TABLES

Table 1.2-1 Dose Factors for Noble Gases and Daughters Table 1.2-2 Dose Parameters for Radioiodines and

Radioactive Particulate, Gaseous Effluents

Table 3.1-1 Bioaccumulation Factors to be Used in the Absence of Site Specific Data

Table 3.1-2 Ingestion Dose Factors for Adults Table 3.1-3 Ingestion Dose Factors for Teenager Table 3.1-4 Ingestion Dose Factors for Child Table 3.1-5 Ingestion Dose Factors for Infant Table 3.1-6 Inhalation Dose Factors for Adults Table 3.1-7 Inhalation Dose Factors for Teenager Table 3.1-8 Inhalation Dose Factors for Child Table 3.1-9 Inhalation Dose Factors for Infant Table 3.1-10 External Dose Factors for Standing on

Contaminated Ground Table 3.1-11 Stable Element Transfer Data

(II

INTRODUCTION

The Offsite Dose Calculation Manual provides the methodology and parameters to be used in the calculation of off-site doses due to radioactive liquid and gaseous effluents to assure compliance with the dose limitations of the Technical Specifications. These dose limitations assure that:

1) the concentration of radioactive liquid effluents from the site to the unrestricted area will be.limited to the concentration levels of 10CFR20, Appendix B, Table II;

2) the exposures to any individual from radioactive liquid effluents will not result in doses greater than the design objectives of 10CFR50, Appendix I;

3) the dose rate at any time at the site boundary from radioactive gaseous effluents will be limited to the annual dose limits of 1OCFR20 for unrestricted areas; and

4) the exposure to any individual from radioactive gaseous effluents will not result in doses greater than the design objectives of 10CFR50, Appendix I.

The methodology used to assure compliance with the dose limitations described above shall also be used to prepare the radioactive liquid and gaseous effluent reports required by the Technical Specifications. To assure compliance with 40CFR190 when twice the design objectives of 10CFR50, Appendix I are exceeded, the methodology and parameters to be used in calculating the off-site dose to any individual resulting from the entire fuel cycle except mining and waste management facilities are provided in this Manual.

The Manual also provides the methodology and parameters to be used in the calculation of radioactive liquid and gaseous effluent monitoring instrumentation alarm/trip setpoints to assure compliance with the concentration and dose rate limitations of the Technical Specifications. Changes to the methodology and parameters used in.this Manual shall be reviewed by a qualified reviewer(s) and approved by the Station Manager and the System Health Physicist prior to implementation and shall be audited by the Nuclear Safety Review Board. Changes to this Manual shall be submitted to the Nuclear Regulatory Commission within 90 days of the date the change was made effective by inclusion in the Monthly Operating Report to comply with the Technical Specifica.tions.

This Manual does not replace any station implementing procedures.

iii . Revision 1

1.0 RELEASE RATE CALCULATIONS

The release rate calculations presented in the following sections are site release limits. Sites containing two or more units shall administratively control releases to assure that the release rate calculations limit releases as stated in the Technical Specifications. Administrative controls could limit the number of releases occurring at one time or apportion the release rate between the units.

1.1 LIQUID EFFLUENTS

To comply with Technical Specifications and to assure that the concentration of radioactive liquid effluents from the site to the unrestricted area is limited to the concentrations of 10CFR20, Appendix B, Table II, Column 2, the following release rate calculation shall be performed:

a f < F +(a Ci )

i=1 MPC.

where:

C. = The concentration of radionuclide, 'i', in undiluted liquid effluent, 2. in pCi/ml.

* PC. = the concentration of radionuclide, 'i', from 10CFR20, Appendix B, Table 2 II, Column 2, in pCi/ml.

f = the undiluted effluent flow from the tank, in gpm.

F = the dilution flow from the site discharge structure to unrestricted area receiving waters, in gpm.

a = recirculation factor at equilibrium; this factor accounts for the fraction of discharged water reused by the station; this factor is one for stations on rivers or lakes where discharged water cannot be reused, and varies for sites where water is recirculated and is specified in the appropriate Appendix.

1.2 GASEOUS EFFLUENTS

In order to comply with the Technical Specifications and to assure that the dose rate, at any time, in the unrestricted area due to radioactive materials released in gaseous effluents from the site is limited to <500 mrem/yr to the total body and <3000 mrem/yr to the skin for the noble gases and is limited to <1500 mrem/yr to any organ for all radioiodine and for all radioactive materials in particulates form and radionuclides other than noble gases with half lives greater than 8 days, the following release rate calculations shall be performed. These calculations, when solved for 'f', i.e. flowrate, are the release rates for noble gases and for radioiodines, particulates and other radionuclides with half-lives greater than 8 days. The most conservative of release rates calculated shall control the release rate.

1-1 Revision 1

1.2.1 Noble Gases

1 K. x [(X/Q)Q.] <500 mrem/yr, and 11

I (L. + 1.1 M.) [(x/Q)Q.] <3000 mrem/yr . 1. 1 1 1

where:

K. = The total body dose factor due to gamma emissions for each identified noble gas radionuclide, in mrem/yr per pCi/m 3 from Table 1.2-1.

L. = The skin dose factor due to beta emissions for each identified noble gas radionuclide, in mrem/yr per pCi/m 3 from Table 1.2-1.

M. = The air dose factor due to gama emissions for each identified noble gas radionuclide, in mrad/yr per pCi/m 3 from Table 1.2-1 (unit conversion constant of 1.1 arem/mrad converts air dose to skin dose).

P. = The dose parameter for radionuclides other than noble gases for the 1 inhalation pathway, in arem/yr per pCi/m 3 and for the food and ground

plane pathways in m2 (mrem/yr) per pCi/sec from Table 1.2-2. The dose factors are .based on the critical individual organ and most restrictive age group (child or infant).

Qi The release rate of radionuclides, 'i', in gaseous effluent from all 1 release points at the site, in pCi/sec.

(x/Q)= The highest calculated annual average dispersion parameter for any area at or beyond the unrestricted area boundary.

W = The highest calculated annual average dispersion parameter.for estimating the dose to an individual at the controlling location.

Q. = kjC.f + k2 = 4.72E+2C.f i1 1

where:

C. = the concentration of radionuclide, 'i', in undiluted gaseous effluent, 1 in pCi/ml.

f = the undiluted effluent flow, in cfm.

k, = conversion factor, 2.83E4 ml/ft3 .

k2 = conversion factor, 6E1 sec/min.

1.2.2 Radioiodines, Particulates, and Others

I P. [W Q.] < 1500 arem/yr V i. 1 1

where the terms are as defined above.

1-2

TABLE 1.2-1 (0 of 1)

D)OSE FACTORS FOR.NOBLE GASES AND DAUGHTERS'.

Total Body Gamma Air Beta Air

Dose Factor Skin Dose Factor Dose Factor Dose Factor

K. L. . N. 1. 1) 11

Radionuclide Onirem/yp, r ici/oi3 (mrem/yr per pCi/rn3) (mrad/yr per pjCi/rn 3). (mrad/yr~per VCi/rn3)

Kr-83n 7.56E-02** --- 1 .93E+01 2.88E+02

Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+t03

Kr-85 1.61E+01 1.34E+03 1.72E+01 1.95Et03

Kr-87 5.92E+03 9.73Et03 6.17E-t03 1.03Et04

Kr-88 1.47E+04 2.37E+'03 1.52E+04 2-93E+03

Kr-89 1..66E+04 1.01E+04 1.73E+04 1.06E+04

Kr-90 .1.56E+-04 7.29Et03 1.63E1-04 7.83Et03

Xe-131mi 9.15E+O1 4.76E+02 1.56E+02 1.11Et03

Xe-133mi 2.51E+02 9.94E+02 3.27E+02 1.48E+03

Xe-133 2.94E+02 3-06E+02 3.53E+02 1.05E+03

Xe- 135mn 3.12E+03 7.11E+02 3.36E+03 7.39E+02

Xe-135 1.81E+03 1.86E+03. 1.92E+03 2.46E+03

Xe-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04

Xe-138 8.83E+03 4.13E+03 9.21E+03 4.75E+03

H.

U1Ar-41 8.84E+03 2.69E+03 9.30E+03 3.28E+03

0

*'The listed dose factors are for radionuclides that may be detected in gaseous effluents.

*"*7.56E-02 =7.56 x 10 2.

TABLE.2-2 (1 of 1)

DOSE PARAMETERS FOR RADIOIODINES AND RADIOACTIVE

PARTICULATE, GASEOUS EFFLUENTS*

P(I), DOSE PARAMETERS FOR RADIOIODINES AND RADIOACTIVE PARTICULATES IN GASEOUS EFFLUENTS

Pathways Pathways.

Inhalation Milk and Ground Inhalation Milk and Ground

(mrem/yr per m2 . mrem/yr per (mrem/yr per (m2 .mrem/yr per

Radionuclide pCi/m 3) pCi/sec) Radionuclide pCi/m 3 ) pCi/sec)

H 3** 6.5E+02 2.4E+03 RU 103 1.6E+04 1.6E+08

RU 106 1.6E+05 3.OE+08

P 32 2.OE+06 1.6E+11 AG 110M 3.3E+04 .1.5E+10

CR 51 3.6E+02 1.1E+07 CD 115M 7.OE+04 5.2E+07

MN 54 2.5E+04 1.IE+09 SN 123 2.9E+05 3.7E+09

FE 55 2.OE+04 1.1E+08 SN 126 1.2E+06 1.1E+10

FE 59 2.4E+04 7.2E+08 SB 124 5.9E+04 1.4E+09

CO 58 1.1E+04 5.8E+08 SB 125 1.5E+04 9.1E+08

CO 60 3.2E+04 4.6E+09 TE 127M 3.8E+04 1.OE+09

NI 63 3.4E+05 3.OE+10 TE 129M 3.2E+04 1.3E+09

ZN 65 6.3E+04 1.8E+10 CS 134 7.OE+05 5.6E+10

RB 86 1.9E+05 2.1E+10 CS 136 1.3E+05 5.7E+09

CS 137 6.1E+05 5.OE+10

SR 90 4.1E+07 1.OE+11 BA 140 5.6E+04 2.6E+08

Y 91 7.OE+04 5.9E+06 CE 141 2.2E+04 3.2E+07

ZR 95 2.2E+04 3.5E+08 CE 144 1.5E+05 . 1.6E+08

NB 95 1.3E+04 3.8E+08 I 131 1.5E+07 1.OE+12

MO 99 2.6E+02 3.2E+08. I 133 3.6E+06 9.6E+09

*If SR-90 analysis is performed, use P(I) given in.I-131 for unidentified components. If SR-90 and 1-131

analyses are performed, use P(I) given in CS-137 for unidentified components. If SR-90, 1-131, and CS-137

analyses are performed, use P(I) given in ZN-65 for unidentified components.

*'For tritium, the units of dose parameters are mrem/yr per pCi/m3 for all pathways, and they must be

multiplied by X/Q in Specification 3.11.2.1.B.

2.0 RADIATION MONITORING SETPOINTS

Effluent radiation monitor alarm/trip setpoints shall be determined using the calculations presented in the following sections. The calculations define the

relationships between the measured effluent activity, the maximum allowable

effluent activity, the effluent flowrate, and the dilution available in the

restricted area (as defined for effluent releases in the Technical Specifica

tions) which must be controlled to assure that the instantaneous release rate

is not exceeded.

The setpoints shall be determined for those monitors listed in the appropriate

tables of the Technical Specifications.

2.1 LIQUID MONITORS

The following equation shall be used to calculate liquid radiation monitor

setpoints;

Cf < MPC F+f

where:

MPC = the effluent concentration limit implementing 10CFR20 for the site, in PCi/ml.

C the radioactivity concentration in pCi/ml, in the effluent line prior to

dilution and subsequent release, which may be the setpoint and, if so,

represents a value which, if exceeded, would result in concentrations

exceeding the limits of 10CFR20 in the unrestricted area.

f = the flow measured at the radiation monitor location in gpm.

F = the dilution water flow as measured prior to the release point in gpm.

(Note that if no dilution is provided, C < MPC. Also, note that when (F) is

large compared to (f), then F + f F.)

2.2 GAS MONITORS

The following equation shall be used to calculate noble gas radiation monitor

setpoints based on Xe-133:

K.(X/Q)Q. < 500 1 1

Q. = 4.72E+2 C f (See Section 1.2.1)

where:

C = the gross activity in undiluted effluent, in pCi/ml.

f = the flow from the tank or building and varies for various release

sources, in cfm.

2-1

K = from Table 1.2-1 for Xe-133, 2.94E-2 mrem/yr per pCi/m 3 .

(x/Q) = the highest calculated annual average dispersion parameter for any area at or beyond the unrestricted area boundary for long term releases.

2

2-2

3.0 DOSE CALCULATIONS

3.1 DOSE MODELS FOR MAXIMUM EXPOSED INDIVIDUAL

3.1.1 Liquid Effluents

Of the possible exposure pathways in the aquatic environment, only two

contribute significantly to the total dose; these pathways are ingestion of

potable water and aquatic foods. The dose contributions, from these pathways, for measured quantities of radioactive materials identified in liquid effluents released to unrestricted areas shall be calculated for the maximum exposed individual in each.age group using:

m D =I (A At C. at aiT = 2 2

where:

D = the cumulative dose commitment to the total body or any organ, at t, for an individual of age group, a, from the liquid effluent

m for the total time period I At, in mrem.

at,= the length of the 2th time period over which C and F are averaged for all liquid releases, in hours.

C. = the average concentration of radionuclide, 'i', in undiluted liquid effluent during time period At from any liquid release, in pCi/ml.

F = the near field average dilution factor for Ci2 during any liquid effluent release where:

F Ff 9 F Fx f

where:

a recirculation factor at equilibrium; this factor accounts for the frac

tion of discharged water reused by the station. This factor is one

for stations on rivers or lakes where discharged water cannot be reused, and varies for sites where water is recirculated and is specified in the

appropriate Appendix.

f liquid.radwaste flow, in gpm.

F dilution flow, in gpm.

A. = the site related ingestion dose commitment factor for an individual alt of age group, a, to the total body or any organ, 'T', for each identi

fied principal gamma and beta emitter, mrem/hr per pCi/ml.

Aai = 1.14E5 (U /D + U BF.)DF aiT aw w aF 1 ailt

3-1 Revision 1

where:

1.14E5 = 106pCi/pCi x 103ml/kg 8760 hr/yr.

U = Water consumption by age group, 2/yr. aw infant 330

child 510 teen 510 adult 730

D= Dilution factor from the near field area to the potable water intake.

U = fish consumption by age group, kg/yr. aF infant -

child 6.9 teen 16 adult 21

BF. = Bioaccumulation factor for radionuclide, 'i', in fish, pCi/kg per pCi/l, from Table 3.1-1.

DF aiT = Dose conversion factor for radionuclide, 'i', by age group in preselected organ, T, in mrem/pCi, from Tables 3.1-2, 3.1-3, 3.1-4, and

3.1-5, respectively.

3.1.2 Gaseous Effluents

The dose contributions from measured quantities of radioactive materials

identified in gaseous effluent released to unrestricted areas shall be cal

culated for the maximum exposed individual using the following equations:

3.1.2.1 Noble Gases

For gamma .radiation:

D = 3.17 E-8 I M. [(X/Q) Qi] y i=1

For beta radiation:

D = 3.17 E-8 I N. [(XQ) QI i=1

where:

3.17 x 10 8 = The inverse of the number of seconds in a year.

1. = The air dose factor due to gamma emissions for each identified noble

gas radionuclide, in mrad/yr per pCi/m3 from Table 1.2-1.

N. = The air dose factor due to beta emissions for each identified noble

gas radionuclide, in mrad/yr per pCi/m3 from Table 1.2-1.

(XIQ) = The highest calculated annual average relative concentration for any area at or beyond the unrestricted area boundary.

3-2

Q.= The release of noble gas radionuclides, 'i', in gaseous effluents, Iin pCi.

3.1.2.2 Radioiodines, Particulates, and Others

These calculations apply to all radioiodines, radioactive materials in partic

ulate form and radionuclides other than noble gases with half-lives greater

than 8 days:

D= 3.17 E-8 1 R. [WQ.] .1 1

where:

3.17 x 10-8 = The inverse of the number of seconds in a year.

Q. = The release of radioiodines, radioactive materials in particulate form 1 and radionuclides other than noble gases in gaseous effluents, 'i',

in pCi. Releases shall be cumulative over the calendar quarter or year as appropriate.

W= The annual average dispersion or deposition parameter for estimating the

dose to an individual at the controlling location.

W = (x/Q) for the inhalation pathway, in sec/m.

W = (D/Q) for the food and ground plane pathways, in meters2 .

R. = The dose factor for each identified radionuclide, 'i', in m2 (mrem/yr) per 1 pCi/sec or mrem/yr per pCi/m.

R. = RI (x/QI + R G(D/Q + R CD/Q,x/Q] + R [D/Q,x/Q] + R [D/Q,x/Q

where:

Inhalation Pathway Factor, R. [x/Q]

R [x/Q] = K'(BR) (DFA.) (mrem/yr per pCi/m ) 1 a la

where:

K' = a constant of unit conversion, 106 pCi/pci.

(BR) = the breathing rate of the receptor of age group (a), in m3 /yr. a

The breathing rates (BR) a for the various age groups are tabulated below, as

given in Regulatory Guide 1.109.

Age Group (a) Breathing Rate (m3/yr)

Infant 1400 Child 3700 Teen 8000 Adult 8000

3-3

(DFA.) = the maximum organ inhalation dose factor the receptor of age group 1 a (a) for the ith radionuclide, in mrem/pCi. The total body is con

sidered as an organ in the selection of (DFA )a. See Tables 3.1-6,

3.1-7, 3.1-8, and 3.1-9.

Inhalation dose factors (DFA.) for the various age groups are given in

Tables 3.1-6, 3.1-7, 3.1-8, ana 3.1-9 (taken from Regulatory Guide 1.109

(Rev.1)).

Ground Plane Pathway Factor, R. [D/Q] 1

G -A t R. [D/Q] = K'K"(SF)DFG.[(1-e i )/X.] (m 2 mrem/yr per pCi/sec)

where:

K' = a constant of unit conversion, 106 pCi/pCi.

K" = a constant of unit conversion, 8760 hr/year.

X. = the decay constant for the ith radionuclide, sec 1 .

t = the exposure time, 4.73 x 108 sec (15 years).

DFG. = the ground plane dose conversion factor for the ith radionuclide (mrem/hr per pCi/m2).

SF = the shielding factor (dimensionless), 0.7 (Regulatory Guide 1.109 (Rev. 1)).

Ground plane dose conversion factors, DFG, are found in Table 3.1-10.

Grass-Cow-Milk Pathway Factor, RC [D/Q]

QF(U ) ff_(1- _f_)e _ ifA~ R. [D/Q] = K' E F ap (r)(DFL i a Y e 1 Xi w a p Ys

(m *mrem/yr per pCi/sec)

where:

K' = a constant of unit conversion, 106 pCi/pCi.

Q = the cow's consumption rate, in kg/day (wet weight),.50 (Regulatory

Guide 1.109 (Rev. 1)).

U = the receptor's milk consumption rate for age (a), in liters/yr. ap

U (liters/yr) - Infant 330 ap - Child 330

- Teen 400

- Adult 310 (Regulatory Guide 1.109 (Rev. 1))

3-4

Y = the agricultural productivity by unit area of pasture feed grass, in kg/m 2 , 0.7.

Y = the agricultural productivity by unit area of stored feed, in kg/m2 , 2.0.

F = the stable element transfer coefficients, in days/liter, Table 3.1-11. m

r = fraction of deposited activity retained on cow's feed grass, r = 1 for radioiodine and r = 0.2 for particulates (Regulatory Guide 1.109).

(DFL.) = the maximum organ ingestion dose factor for the ith radionuclide ia for the receptor in age group 'a', in mrem/pCi. See Tables 3.1-2,

3.1-3, 3.1-4, and 3.1-5.

X. = the decay constant for the ith radionuclide, in sec1.

A = the decay constant for removal of activity on leaf and plant surfaces w by weathering, 5.73 x 10 7 sect1 (corresponding to a 14 day half-life).

t = the transport time from pasture to cow, to milk, to receptor, in sec, 1.73 x 10s (2 days).

t h = the transport time from pasture, to harvest, to cow, to milk, to reh ceptor, in sec, 7.78 x 106 (90 days).

f = fraction of the year that the cow is on pasture (dimensionless), 1.0.

f = fraction of the cow feed that is pasture grass while the cow is on pasture (dimensionless), 1.0.

E = an adjustment fraction which accounts for the fraction of radionuclides in elemental form which contribute dose for this pathway,

E = 0.5 for radioiodine, E = 1.0 for all others.

The concentration of tritium in milk is based Con the airborne concentration

rather than the deposition. Therefore, the R. is based on [x/QI: 1

R C[X/Q] = K'K'"F QU (DFL.) [0.75(0.5/H)] (mrem/yr per pCi/m 3) 1 mFap 1a

where:

K'" a constant of unit conversion, 103 gm/kg.

H = absolute humidity of the atmosphere, 8 gm/m 3 , (Regulatory Guide 1.109)

0.75 = the fraction of total feed that is water.

0.5 = the ratio of the specific activity of the feed grass water to the atmos

pheric water.

0 3-5

Grass-Cow-Meat Pathway Factor, R.[D/Ql 1

The integrated concentration in meat follows in a similar manner to the development for the milk pathway, therefore:

QF(U ) ff (1-f f )e-ith -At f R [D/Q] = K' ap F (r)(DFL...) a - + e

1 X. + f Y I w p s

(in2 mrem/yr per pCi/sec)

where:

F = the stable element transfer coefficients, in days/kg, Table 3.1-11. f

U = the receptor's meat consumption rate for age (a), in kg/yr. ap.

U (kg/yr) - Infant 0 ap - Child 41

- Teen 65 - Adult 110 Taken from Regulatory Guide 1.109 (Rev. 1).

t = the transport time from pasture to receptor, in sec.

th = the transport time from crop field to receptor, in sec.

The concentration of tritium in meat is based on its airborne concentration. rather than the deposition. Therefore, the R. is based on [X/Q]:

1

R0[x/Q] = K'K'"F Q U (DFL.) [0.75(0.5/H)] (mrem/yr per pCi/m 3) 1 fF ap 1 a

where all terms are defined above.

Vegetation Pathway Factor, RY[D/Q]

The integrated concentration in vegetation consumed by man follows the expression developed in the derivation of the milk factor. Man is considered to

consume two types of vegetation (fresh and stored) that differs only in the

time period between harvest and consumption, therefore:

R [D/Q] = K' (r) (DFL ULf e LL + USf a th

[Y v(X.+ X ) DLi aL a g

(m2 * mrem/yr per pCi/sec)

where:

K' = a constant of unit conversion, 106 pCi/pCi.

UL = the consumption rate of fresh leafy vegetation by the receptor in age a group (a), in kg/yr.

3-6

UL = (kg/hr) - Infant 0 a - Child 26

- Teen 42 - Adult 64

S U = the consumption rate of stored vegetation by the receptor in age group

a (a), in kg/yr.

US (kg/yr) - Infant 0 a - Child 520

- Teen 630 - Adult 520

f= the fraction of the annual intake of fresh leafy vegetation grown locally. L

f = the fraction of the annual intake of stored vegetation grown locally. g

t = the average time between harvest of leafy vegetation and its consumption, in seconds, 8.6 x 10 (1 day).

t = the average time between harvest of stored vegetation and its consumption, h in seconds, 5.18 x 106 (60 days).

Y= the vegetation area density, 2.0 kg/m2 . v

and all other factors are previously defined.

The concentration of tritium in vegetation is base4 on the airborne concentra

tion rather than the deposition. Therefore, the Ri is based on [x/Q]:

R [x/Q] = K'K'" Lf + Usff (DFL) [0.75(0.5/H) (mrem/yr per pCi/m3). i[ a L ag7 ) ia

All terms defined previously.

3.1.3 -Direct Radiation

Direct radiation is that radiation from confined sources and does not include

any external component from radioactive effluents. The point kernel method

has been used to calculate offsite dose rates from radioactive materials stored

in the refueling water storage tanks and reactor makeup water storage tanks.

Dose calculations using this method performed for Duke Nuclear Stations indicate

direct radiation doses are much less than 0.01 mrem/yr. and, therefore, makes a

negligible contribution to individual dose. Direct radiation doses will not be

calculated routinely.

3.2 SIMPLIFIED DOSE PROJECTIONS

To estimate the cumulative dose contributions to the maximum exposed individ

ual for 31 day dose projection calculations, the calculations presented in

Section 3.1 can be simplified. The simplified calculations would be for an

individual in the critical population using only data for the critical pathway and critical radionuclide(s). Critical populations, critical pathways, and

critical radionuclides have been determined for each Duke Nuclear Station from

the dose calculations performed to evaluate compliance with Appendix I to 10CFR50.

3-7 Revision 1

Simplified 31-day .dose projection calculations are presented in the section on

site specific information.

3.3 FUEL CYCLE CALCULATIONS

In accordance with the requirements of 40CFR190, the annual dose commitment to any member of the general public shall be calculated to assure that doses

are limited to 25 millirems to the total body or any organ with the exception of the thyroid which is limited to 75 millirems. In accordance with the

requirements of the Technical Specifications, the annual dose commitment shall

also be calculated any time that one of the quarterly dose limits of the Technical Specifications is exceeded; these annual dose commitments may not just be calculated for the calendar year.

The "Uranium fuel cycle" is defined in 40CFR Part 190.02(b) as:

"Uranium fuel cycle means the operations of milling or uranium ore, chemical conversion of uranium, isotopic enrichment of uranium, fabrication of uranium fuel, generation of electricity by a light-water

cooled nuclear power plant using uranium fuel, and reprocessing of spent uranium fuel, to the extent that these directly support the production of electrical power for public use utilizing nuclear energy, but excludes mining operations, operations at waste disposal sites, transportation of any radioactive material in support of these operations, and the reuse of recovered non-uranium special nuclear and by-product materials from the cycle."

Based on this definition of the fuel cycle and the information in 10CFR51

Table S-3 and WASH-1248, the radiological impact of the following operations

has been assessed for Duke Nuclear Stations:

3.3.1 Milling

No milling operations occur within fifty miles of any Duke Nuclear Station.

The increment of dose from milling operations to any individual within fifty miles of any Duke Nuclear Station is negligible.

3.3.2 Conversion

No uranium hexafluoride production occurs within fifty miles of any Duke Nuclear Station.. The increment of dose from UF6 production to any individual

within fifty miles of any Duke Nuclear Station is negligible.

3.3.3 Enrichment

No uranium enrichment operations occur within fifty miles of any Duke Nuclear

Station. The increment of dose from enrichment operations to any individual

within fifty miles of any Duke Nuclear Station is negligible.

3.3.4 Fuel Fabrication

No fuel fabrication operations occur within fifty miles of any Duke Nuclear

Station. The increment of dose from fabrication operations to any individual

within fifty miles of any Duke Nuclear Station is negligible..

3-8 Revision 1

3.3.5 Nuclear Power Production

The production of electricity for public use using light-water-cooled nuclear power stations results in increments of dose to individuals within fifty miles of any station due to liquid and gaseous effluent releases and direct radiation or skyshine. The increments of dose resulting from liquid and gaseous effluent releases will be calculated using the methodology presented in Sections 3.1.1 and 3.1.2. The dose from direct radiation, skyshine, and radiation from the station storage facilities has been estimated using conservative assumptions (see Section 3.1.3), the estimates of this dose will be presented in the section on site specific information.

In certain situations more than one nuclear power station site may contribute to the doses to be considered in making fuel cycle dose assessments in accordance with 40CFRI90. Situations involving more than one station will be presented in the section on site specific information.

3.3.6 Fuel Reprocessing

No fuel reprocessing operations occur within fifty miles of any Duke Nuclear Station. The increment of dose from reprocessing operations to any individual within fifty miles of any Duke Nuclear Station is negligible.

To summarize, only dose increments from nuclear power production operations (Section 3.3.5) need be considered in calculations to demonstrate compliance with the requirements of 40CFR190.

Revision 1

TABLE 3.1-1* (1 of 1)

BIOACCUMULATION FACTORS TO BE USED IN THE ABSENCE OF SITE-SPECIFIC DATA (pCi/kg per pCi/liter)

FRESHWATER ELEMENT FISH INVERTEBRATE

H 9.0E-01 9.OE-01

Na 1.0E-02 .2.OE 02

P 1.0E 05 2.0E 04

Cr 2.OE 02 2.OE 03

Mn 4.OE 02 9.0E 04

Fe 1.OE 02 3.2E 03

Co 5.0E 01 2.0E 02

Ni 1.OE 02 1.0E 02

Cu 5.OE 01 4.OE 02

Zn 2.OE 03 1.OE 04

Br 4.2E 02 3.3E 02

Rb 2.OE 03 1.OE 03

Sr 3.OE 01 1.0E 02

Y 2.5E 01 1.OE 03

Zr 3.3E 00 6.7E 00

Nb 3.OE 04 1.OE 02

Mo 1.OE 01 1.OE 01

Tc 1.5E 01 5.OE 00

Ru 1.OE 01 3.OE 02

Rh 1.OE 01 3.OE 02

Te 4.OE 02 6.1E 03

I 1.5E 01 5.OE 00

Cs 2.OE 03 1.OE 03

Ba 4.OE 00 2.OE 02

La 2.5E 01 1.OE 03

Ce 1.OE 00 1.OE 03

Pr 2.5E 01 1.OE 03

Nd 2.5E 01 1.OE 03

W 1.2E 03 1.OE 01

Np 1.OE 01 4.OE 02

Table taken from Regulatory Guide 1.109 (Rev.1)

TABLE 3.1-1 (1 of 1) Revision 1

TABLE 3.1-2 (1 of 3)

INGESTION DOSE FACTORS FOR ADULTS (MREM PER PCI INGESTED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

H 3 NO DATA 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07

NA 24 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06

P 32 1.93E-04 1.20E-05 7.46E-06 NO DATA NO DATA NO DATA 2.17E-05 CR 51 NODATA NO DATA 2.66E-09 1.59E-09 5.86E-10 3.53E-09 6.69E-07 MN 54 NO DATA 4.57E-06 8.72E-07 NO DATA 1.36E-06 NO DATA 1.40E-05

MN 56 NO DATA 1.15E-07 2.04E-08 NO DATA 1.46E-07 NO DATA 3.67E-06 FE 55 2.75E-06 1.90E-06 4.43E-07 NO DATA NO DATA 1.06E-06 1.09E-06 FE 59 4.34E-06 1.02E-05 3.91E-06 NO DATA NO DATA 2.85E-06 3.40E-05

CO 58 NO DATA 7.45E-07 1.67E-06 NO DATA NO DATA NO DATA 1.51E-05 CO 60 NO DATA 2.14E-06 4.72E-06 NO DATA NO DATA NO DATA 4.02E-05 NI 63 1.30E-04 9.01E-06 4.36E-06 NO DATA NO DATA NO DATA 1.88E-06

NI 65 5.28E-07 6.86E-08 3.13E-08 NO DATA NO DATA NO DATA 1.74E-06 CU 64 NO DATA 8.33E-08 3.91E-08 NO DATA 2.10E-07 NO DATA 7.10E-06 ZN 65 4.84E-06 1.54E-05 6.96E-06 NO DATA 1.03E-05 NO DATA 9.70E-06

ZN 69 1.03E-08 1.97E-08 1.37E-09 NO DATA 1.28E-08 NO DATA 2.96E-09 BR 83 NO DATA NO DATA 4.02E-08 NO DATA NO DATA NO DATA 5.79E-08 BR 84 NO DATA NO DATA 5.21E-08 NO DATA NO DATA NO DATA 4.09E-13

BR 85 NO DATA NO DATA 2.14E-09 NO DATA NO DATA NO DATA LT E-24 RB 86 NO DATA 2.11E-05 9.83E-06 NO DATA NO DATA NO DATA 4.16E-06 RB 88 NO DATA 6.05E-08 3.21E-08 NO DATA NO DATA NO DATA 8.36E-19

RB 89 NO DATA 4.01E-08 2.82E-08 NO DATA NO DATA NO DATA 2.33E-21 SR 89 3.08E-04 NO DATA 8.84E-06 NO DATA NO DATA NO DATA 4.94E-05 SR 90 7.58E-03 NO DATA 1.86E-03 NO DATA NO DATA NO DATA 2.19E-04

SR 91 5.67E-06 NO DATA 2.29E-07 NO DATA NO DATA NO DATA 2.70E-05 SR 92 2.15E-06 NO DATA 9.30E-08 NO DATA NO DATA NO DATA 4.26E-05 Y, 90 9.62E-09 NO DATA 2.58E-10 NO DATA NO DATA NO DATA 1.02E-04

Y 91M 9.09E-11 NO DATA 3.52E-12 NO DATA NO DATA NO DATA 2.67E-10 Y 91 1.41E-07 NO DATA 3.77E-09 NO DATA NO DATA NO DATA 7.76E-05 Y 92 8.45E-10 NO DATA 2.47E-11 NO DATA NO DATA NO DATA 1.48E-05

Y 93 2.68E-09 NO DATA 7.40E-11 NO DATA NO DATA NO DATA 8.50E-05

ZR 95 3.04E-08 9.75E-09 6.60E-09 NO DATA 1.53E-08 NO DATA 3.09E-05

ZR 97 1.68E-09 3.39E-10 1.55E-10 NO DATA 5.12E-10 NO DATA 1.05E-04

TABLE 3.1-2 (1 of 3)

Revision 1

TABLE 3.1-2 (2 of 3)

INGESTION DOSE FACTORS FOR ADULTS (MREM PER PCI INGESTED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

NB 95 6.22E-09 3.46E-09 1.86E-09 NO DATA 3.42E-09 NO DATA 2.10E-05 MO 99 NO DATA 4.31E-06 8.20E-07 NO DATA 9.76E-06 NO DATA 9.99E-06 TC 99M 2.47E-10 6.98E-10 8.89E-09 NO DATA 1.06E-08 3.42E-10 4.13E-07

TC 101 2.54E-10 3.66E-10 3.59E-09 NO DATA 6.59E-09 1.87E-10 1.10E-21 RU 103 1.85E-07 NO DATA 7.97E-08 NO DATA 7.06E-07 NO DATA 2.16E-05 RU 105 1.54E-08 NO DATA 6.08E-09 NO DATA 1.99E-07 NO DATA 9.42E-06

RU 106 2.75E-06 NO DATA 3.48E-07 NO DATA 5.31E-06 NO DATA 1.78E-04 AG 110M 1.60E-07 1.48E-07 8.79E-08 NO DATA 2.91E-07 NO DATA 6.04E-05 TE 125M 2.68E-06 9.71E-07 3.59E-07 8.06E-07 1.09E-05 NO DATA 1.07E-05

TE 127M 6.77E-06 2.42E-06 8.25E-07 1.73E-06 2.75E-06 NO DATA 2.27E-05 TE 127 1.10E-07 3.95E-08 2.38E-08 8.15E-08 4.48E-07 NO DATA 8.68E-06 TE 129M 1.15E-05 4.29E-06 1.82E-06 3.95E-06 4.80E-05 NO DATA 5.79E-05

TABLE 3.1-2 (2 of 3)

TABLE 3.1-2 (3 of 3)

INGESTION DOSE FACTORS FOR ADULTS (MREM PER PCI INGESTED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

TE 129 3.14E-08 1.18E-08 7.65E-09 2.41E-08 1.32E-07 NO DATA 2.37E-08 TE 131M 1.73E-06 8.46E-07 7.05E-07 1.34E-06 8.57E-06 NO DATA 8.40E-05 TE .131 1.97E-08 8.23E-09 6.22E-09 1.62E-08 8.63E-08 NO DATA 2.79E-09

TE 132 2.52E-06 1.63E-06 1.53E-06 1.80E-06 1.57E-05 NO DATA 7.71E-05 I 130 7.56E-07 2.23E-06 8.80E-07 1.89E-04 3.48E-06 NO DATA 1.92E-06 I 131 4.16E-06 5.95E-06 3.41E-06 1.95E-03 1.02E-05 NO DATA 1.57E-06

I 132 2.03E-07 5.43E-07 1.90E-07 1.90E-05 8.65E-07 NO DATA 1.02E-07 I 133 1.42E-06 2.47E-06 .7.53E-07 3.63E-04 4.31E-06 NO DATA 2.22E-06 I 134 1.06E-07 2.88E-07 1.03E-07 4.99E-06 4.58E-07 NO DATA 2.51E-10

I 135 4.43E-07 1.16E-06 4.28E-07 7.65E-05 1.86E-06 NO DATA 1.31E-06 CS 134 6.22E-05 1.48E-04 1.21E-04 NO DATA 4.79E-05 1.59E-05 2.59E-06 CS 136 6.51E-06 2.57E-05 1.85E-05 NO DATA 1.43E-05 1.96E-06 2.92E-06

CS 137 7.97E-05 1.09E-04 7.14E-05 NO DATA 3.70E-05 1.23E-05 2.11E-06 CS 138 5.52E-08 1.09E-07 5.40E-08 NO DATA 8.01E-08 7.91E-09 4.65E-13 BA 139 9.70E-08 6.91E-11 2.84E-09 NO DATA 6.46E-11 3.92E-11 1.72E-07

BA 140 2.03E-05 2.55E-08 1.33E-06 NO DATA 8.67E-09 1.46E-08 4.18E-05 BA 141 4.71E-08 3.56E-11 1.59E-09 NO DATA 3.31E-11 2.02E-11 2.22E-17 BA 142 2..13E-08 2.19E-11 1.34E-09 NO DATA 1.85E-11 1.24E-11 3.OOE-26

LA 140 2.50E-09 1.26E-09 3.33E-10 NO DATA NO DATA NO DATA 9.25E-05 LA 142 1.28E-10 5.82E-11 1.45E-11 NO DATA NO DATA NO DATA 4.25E-07 CE 141 9.36E-09 6.33E-09 7.18E-10 NO DATA 2.94E-09 NO DATA 2.42E-05

CE 143 1.65E-09 1.22E-06 1.35E-10 NO DATA 5.37E-10 NO DATA 4.56E-05 CE 144 4.88E-07 2.04E-07 2.62E-08 NO DATA 1.21E-07 NO DATA 1.65E-04 PR 143 9.20E-09 3.69E-09 4.56E-10 NO DATA 2.13E-09 NO DATA .4.03E-05

PR 144 3.01E-11 1.25E-11 1.53E-12 NO DATA 7.05E-12 NO DATA 4.33E-18 ND 147 6.29E-09 7.27E-09 4.35E-10 NO DATA 4.25E-09 NO DATA 3.49E-05 W 187 1.03E-07 8.61E-08 3.01E-08 NO DATA NO DATA NO DATA 2.82E-05

NP 239 1.19E-09 1.17E-10 6.45E-11 NO DATA 3.65E-10 NO DATA 2.40E-05

TABLE 3.1-2 (3 of 3)

TABLE 3.1-3* (1 of 3)

INGESTION DOSE FACTORS FOR TEENAGER (MREM PER PCI INGESTED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

H 3 NO DATA 1.06E-07 1.06E-07 1.06E-07 1.06E-07 1.06E-07 1.06E-07 NA 24 2.30E-06 2.30E-06 2.30E-06 2.30E-06 2.30E-06 2.30E-06 2.30E-06

P 32 2.76E-04 1.71E-05 1.07E-05 NO DATA NO DATA NO DATA 2.32E-05 CR 51 NO DATA NO DATA 3.60E-09 2.00E-09 7.89E-10 5.14E-09 6.05E-07 MN 54 NO DATA 5.90E-06 1.17E-06 NO DATA 1.76E-06 NO DATA 1.21E-05

MN 56 NO DATA 1.58E-07 2.81E-08 NO DATA 2.OOE-07 NO DATA 1.04E-05 FE 55 3.78E-06 2.68E-06 6.25E-07 NO DATA NO DATA 1.70E-06 1.16E-06 FE 59 5.87E-06 1.37E-05 5.29E-06 NO DATA NO DATA 4.32E-06 3.24E-05

CO 58 NO DATA 9.72E-07 2.24E-06 NO DATA NO DATA NO DATA 1.34E-05 CO 60 NO DATA 2.81E-06 6.33E-06 NO DATA NO DATA NO DATA 3.66E-05 NI 63 1.77E-04 1.25E-05. 6.OOE-06 NO DATA NO DATA NO DATA 1.99E-06

NI 65 7.49E-07 9.57E-08 4.36E-08 NO DATA NO DATA NO DATA 5.19E-07 CU 64 NO DATA 1.15E-07 5.41E-08 NO DATA 2.91E-07 NO DATA 8.92E-06 ZN 65 5.76E-06 2.OOE-05 9.33E-06 NO DATA 1.28E-05 NO DATA 8.47E-06

ZN 69 1.47E-08 2.80E-08 1.96E-09 NO DATA 1.83E-08 NO DATA 5.16E-08 BR 83 NO DATA NO DATA 5.74E-08 NO DATA NO DATA NO DATA LT E-24 BR 84 NO DATA NO DATA 7.22E-08 NO DATA NO DATA NO DATA LT E-24

BR 85 NO DATA NO DATA 3.05E-09 NO DATA NO DATA NO DATA LT E-24 RB 86 NO DATA 2.98E-05 1.40E-05 NO DATA NO DATA NO DATA 4.41E-06 RB 88 NO DATA 8.52E-08 4.54E-08 NO DATA NO DATA NO DATA 7.30E-15

RB 89 NO DATA 5.50E-08 3.89E-08 NO DATA NO DATA NO DATA 8.43E-17 SR 89 4.40E-04 NO DATA 1.26E-05 NO DATA NO DATA NO DATA 5.24E-05 SR 90 8.30E-03 NO DATA 2.05E-03 NO DATA NO DATA NO DATA 2.33E-04

SR 91 8.07E-06 NO DATA 3.21E-07 NO DATA NO DATA NO DATA 3.66E-05 SR 92 3.05E-06 NO DATA 1.30E-07 NO DATA NO DATA NO DATA 7.77E-05 Y 90 1.37E-08 NO DATA 3.69E-10 NO DATA NO DATA NO DATA 1.13E-04

Y 91M 1.29E-10 NO DATA 4.93E-12 NO DATA NO DATA NO DATA 6.09E-09 Y 91 2.01E-07 NO DATA 5.39E-09 NO DATA NO DATA NO DATA 8.24E-05 Y 92 1.21E-09 NO DATA 3.50E-11 NO DATA NO DATA NO DATA 3.32E-05

*Taken from Regulatory Guide 1.109 (Rev. 1)

TABLE 3.1-3 (1 of 3)

Revision 1

TABLE 3.1-3 (cont'd) (2 of 3)

INGESTION DOSE FACTORS FOR TEENAGER (MREM PER PCI INGESTED)

VUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

Y 93 3.83E-09 NO DATA 1.05E-10 NO DATA NO DATA NO DATA 1.17E-04 ZR 95 4.12E-08 1.30E-08 8.94E-09 NO DATA 1.91E-08 NO DATA 3.OOE-05 ZR 97 2.37E-09 4.69E-10 2.16E-10 NO DATA 7.11E-10 NO DATA 1.27E-04

NB 95 8.22E-09 4.56E-09 2.51E-09 NO DATA. 4.42E-09 NO DATA 1.95E-05 MO 99 NO DATA 6.03E-06 1.15E-06 NO DATA 1.38E-05 NO DATA 1.08E-05 TC 99M 3.32E-10 9.26E-10 1.20E-08 NO DATA 1.38E-08 5.14E-10 6.08E-07

TC101 3.6QE-10 5.12E-10 5.03E-09 NO DATA 9.26E-09 3.12E-10 8.75E-18 RU103 2.55E-07 NO DATA 1.09E-07 NO DATA 8.99E-07 NO DATA 2.13E-05 RU105 2.18E-08 NO DATA 8.46E-09 NO DATA 2.75E-07 NO DATA 1.76E-05

RU106 3.92E-06 NO DATA 4.94E-07 NO DATA 7.56E-06 NO DATA 1.88E-04 AG110M 2.05E-07 1.94E-07 1.18E-07 NO DATA 3.70E-07 NO DATA 5.45E-05 TE125M 3.83E-06 1.38E-06 5.12E-07 1.07E-06 NO DATA NO DATA 1.13E-05

TE127M 9.67E-06 3.43E-06 1.15E-06 2.30E-06 3.92E-05 NO DATA 2.41E-05 TE127 1.58E-07 5.60E-08 3.40E-08 1.09E-07 6.40E-07 NO DATA 1.22E-05 TE129M 1.63E-05 6.05E-06 2.58E-06 5.26E-06 6.82E-05 NO DATA 6.12E-05

TE129 4.48E-08 1.67E-08 1.09E-08 3.20E-08 1.88E-07 NO DATA 2.45E-07 TE131M 2.44E-06 1.17E-06 9.76E-07 1.76E-06 1.22E-05 NO DATA 9.39E-05 TE131 2.79E-08 1.15E-08 8.72E-09 2.15E-08 1.22E-07 NO DATA 2.29E-09

TE132 3.49E-06 2.21E-06 2.08E-06 2.33E-06 2.12E-05 NO DATA 7.OOE-05 I 130 1.03E-06 2.98E-06 1.19E-06 2.43E-04 4.59E-06 NO DATA 2.29E-06 I 131 5.85E-06 8.19E-06 4.40E-06 2.39E-03 1.41E-05 NO DATA 1.62E-06

I 132 2.79E-07 7.30E-07 2.62E-07 2.46E-05 1.15E-06 NO DATA 3.18E-07

I 133 2.01E-06 3.41E-06 1.04E-06 4.76E-04 5.98E-06 NO DATA 2.58E-06

I 134 1.46E-07 3.87E-07 1.39E-07 6.45E-06 6.10E-07 NO DATA 5.10E-09

I 135 6.10E-07 1.57E-06 5.82E-07 1.01E-04 2.48E-06 NO DATA 1.74E-06

CS134 8.37E-05 1.97E-04 9.14E-05 NO DATA 6.26E-05 2.39E-05 2.45E-06

CS136 8.59E-06 3.38E-05 2.27E-05 NO DATA 1.84E-05 2.90E-06. 2.72E-06

CS137 1.12E-04 1.49E-04 5.19E-05 NO DATA 5.07E-05 1.97E-05 2.12E-06

CS138 7.76E-08 1.49E-07 7.45E-08 NO DATA 1.10E-07 1.28E-08 6.76E-11

BA139 1.39E-07 9.78E-11 4.05E-09 NO DATA 9.22E-11 6.74E-11 1.24E-06

TABLE 3.1-3 (2 of 3)

TABLE 3.1-3 (cont'd) (3 of 3)

INGESTION DOSE FACTORS FOR TEENAGER (MREM PER PCI INGESTED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

BA140 2.84E-05 3.48E-08 1.83E-06 NO DATA 1.18E-08 2.34E-08 4.38E-05 BA141 6.71E-08 5.01E-11 2.24E-08 NO DATA 4.65E-11 3.43E-11 1.43E-13 BA142 2.99E-08 2.99E-11 1.84E-09 NO DATA 2.53E-11 1.99E-11 9.18E-20

LA140 3.48E-09 1.71E-09 4.55E-10 NO DATA NO DATA NO DATA 9.82E-05 LA142 1.79E-10 7.95E-11 1.98E-11 NO DATA NO DATA NO DATA 2.42E-06 CE141 1.33E-08 8.88E-09 1.02E-09 NO DATA 4.18E-09 NO DATA 2.54E-05

CE143 2.35E-09 1.71E-06 1.91E-10 NO DATA 7.67E-10 NO DATA 5.14E-05 CD144 6.96E-07 2.88E-07 3.74E-08 NO DATA 1.72E-07 NO DATA 1.75E-04 PR143 1.31E-08 5.23E-09 6.52E-10 NO DATA 3.04E-09 NO DATA 4.31E-05

PR144 4.30E-11 1.76E-11 2.18E-12 NO DATA 1.01E-11 NO DATA 4.74E-14 ND147 9.38E-09 1.02E-08 6.11E-10 NO DATA 5.99E-09 NO DATA 3.68E-05 W 187 1.46E-07 1.19E-07 4.17E-08 NO DATA NO DATA NO DATA 3.22E-05

NP239 1.76E-09 1.66E-10 9.22E-11 NO DATA 5.21E-10 NO DATA 2.67E-05

TABLE 3.1-3 (3 of 3)

ITABLE

3.1-4* (1 of 3)

INGESTON DOSE FACTORS FOR CHILD (MREM PER PCI INGESTED)

.NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

H 3 NO DATA 2.03E-07 2.03E-07 2.03E-07 2.03E-07 2.03E-07 2.03E-07 NA 24 5.80E-06 5.80E-06 5.80E-06 5.80E-06 5.80E-06 5.80E-06 5.80E-06

P 32 8.25E-04 3.86E-05 3.18E-05 NO DATA NO DATA NO DATA 2.28E-05 CR 51 NO DATA NO DATA 8.90E-09 4.94E-09 1.35E-09 9.02E-09 4.72E-07 MN 54 NO DATA 1.07E-05 2.85E-06 NO DATA 3.OOE-06 NO DATA 8.98E-06

MN 56 NO DATA 3.34E-07 7.54E-08 NO DATA 4.04E-07 NO DATA 4.84E-05 FE 55 1.15E-05 6.10E-06 1.89E-06 NO DATA NO DATA 3.45E-06 1.13E-06 FE 59 1.65E-05 2.67E-05 1.33E-05 NO DATA NO DATA 7.74E-06 2.78E-05

CO 58 NO DATA 1.80E-06 5.51E-06 NO DATA NO DATA NO DATA 1.05E-05 CP 60 NO DATA 5.29E-06 1.56E-05 NO DATA NO DATA NO DATA 2.93E-05 NI 63 5.38E-04 2.88E-05 1.83E-05 NO DATA NO DATA NO DATA 1.94E-06

NI 65 2.22E-06 2.09E-07 1.22E-07 NO DATA NO DATA NO DATA 2.56E-05 CU 64 NO DATA 2.45E-07 1.48E-07 NO DATA 5.92E-07 NO DATA 1.15E-05 ZN 65 1.37E-05 3.65E-05 2.27E-05 NO DATA 2.30E-05 NO DATA 6.41E-06

ZN 69 4.38E-08 6.33E-08 5.85E-09 NO DATA 3.84E-08 NO DATA 3.99E-06 BR 83 NO DATA NO DATA 1.71E-07 NO DATA NO DATA NO DATA LT E-24 BR 84 NO DATA NO DATA 1.98E-07 NO DATA NO DATA NO DATA. LT E-24

BR 85 NO DATA NO DATA 9.12E-09 NO DATA NO DATA NO DATA LT E-24 BR 86 NO DATA 6.70E-05 4.12E-05 NO DATA NO DATA NO DATA 4.31E-06 RB 88 NO DATA 1.90E-07 1.32E-07 NO DATA NO DATA NO DATA 9.32E-09

RB 89 NO DATA 1.17E-07 1.04E-07 NO DATA NO DATA NO DATA 1.02E-09

SR 89 1.32E-03 NO DATA 3.77E-05 NO DATA NO DATA NO DATA 5.11E-05

SR 90 1.70E-02 NO DATA 4.31E-03 NO DATA NO DATA NO DATA 2.29E-04

SR 91 2.40E-05 NO DATA 9.06E-07 NO DATA NO DATA NO DATA 5.30E-05 SR 92 9.03E-06 NO DATA 3.62E-07 NO DATA NO DATA NO DATA 1.71E-04 Y 90 4.11E-08 NO DATA 1.10E-09 NO DATA NO DATA NO DATA 1.17E-04

Y 91M 3.82E-10 NO DATA 1.39E-11 NO DATA NO DATA NO DATA 7.48E-07 Y 91 6.02E-07 NO DATA 1.61E-08 NO DATA NO DATA NO DATA 8.02E-05

Y 92 3.60E-09 NO DATA 1.03E-10 NO DATA NO DATA NO DATA 1.04E-04

*Taken from Regulatory Guide 1.109 (Rev. 1).

TABLE 3.1-4 (1 of 3)

Revision 1

TABLE 3.1-4 (cont'd) (2 of 3)

INGESTION DOSE FACTORS FOR CHILD (MREM PER PCI INGESTED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

Y 93 1.14E-08 NO DATA 3.13E-10 NO DATA NO DATA NO DATA 1.70E-04 ZR 95 1.16E-07 2.55E-08 2.27E-08 NO DATA 3.65E-08 NO DATA 2.66E-05 ZR 97 6.99E-09 1.01E-09 5.96E-10 NO DATA 1.45E-09 NO DATA 1.53E-04

NB 95 2.25E-08 8.76E-09 6.26E-09 NO DATA 8.23E-09 NO DATA 1.62E-05 MO 99 NO DATA 1.33E-05 3.29E-06 NO DATA 2.84E-05 NO DATA 1.10E-05 TC 99M 9.23E-10 1.81E-09 3.OOE-08 NO DATA 2.63E-08 9.19E-10 1.03E-06

TC101 1.07E-09 1.12E-09 1.42E-08 NO DATA 1.91E-08 5.92E-10 3.56E-09 RU103 7.31E-07 NO DATA 2.81E-07 NO DATA 1.84E-06 NO DATA 1.89E-05 RU105 6.45E-08 NO DATA 2.34E-08 NO DATA 5.67E-07 NO DATA 4.21E-05

RU106 1.17E-05 NO DATA 1.46E-06 NO DATA 1.58E-05 NO DATA 1.82E-04 AG110M 5.39E-07 3.84E-07 2.91E-07 NO DATA 6.78E-07 NO DATA 4.33E-05 TE125M 1.14E-05 3.09E-06 1.52E-06 3.20E-06 NO DATA NO DATA 1.10E-05

TE127M 2.89E-05 7.78E-06 3.43E-06 6.91E-06 8.24E-05 NO DATA 2.34E-05 TE127 4.71E-07 1.27E-07 1.01E-07 3.26E-07 1.34E-06 NO DATA 1.84E-05

TE129M 4.87E-05 1.36E-05 7.56E-06 1.57E-05 1.43E-04 NO DATA 5.94E-05

TE129 1.34E-07 3.74E-08 3.18E-08 9.56E-08 3.92E-07 NO DATA 8.34E-06 TE131M 7.20E-06 2.49E-06 2.65E-06 5.12E-06 2.41E-05 NO DATA 1.01E-04 TE131 8.30E-08 2.53E-08 2.47E-08 6.35E-08 2.51E-07 NO DATA 4.36E-07

TE132 1.01E-05 4.47E-06 5.40E-06 6.51E-06 4.15E-05 NO DATA 4.50E-05

I 130 2.92E-06 5.90E-06 3.04E-06 6.50E-04 8.82E-06 NO DATA 2.76E-06

I 131 1.72E-05 1.73E-05 9.83E-06 5.72E-03 2.84E-05 NO DATA 1.54E-06

I 132 8.OOE-07 1.47E-06 6.76E-07 6.82E-05 2.25E-06 NO DATA 1.73E-06

I 133 5.92E-06 7.32E-06 2.77E-06 1.36E-03 1.22E-05 NO DATA 2.95E-06 I 134 4.19E-07 7.78E-07 3.58E-07 1.79E-05 1.19E-06 NO DATA 5.16E-07

I 135 1.75E-06 3.15E-06 1.49E-06 2.79E-04 4.83E-06 NO DATA 2.40E-06 CS134 2.34E-04 3.84E-04 8.10E-05 NO DATA 1.19E-04 4.27E-05 2.07E-06

CS136 2.35E-05 6.46E-05 4.18E-05 NO DATA 3.44E-05 5.13E-06 2.27E-06

CS137 3.27E-04 3.13E-04 4.62E-05 NO DATA 1.02E-04 3.67E-05 1.96E-06 CS138 2.28E-07 3.17E-07 2.01E-07 NO DATA 2.23E-07 2.40E-08 1.46E-07

BA139 4.14E-07 2.21E-10 1.20E-08 NO DATA 1.93E-10 1.30E-10 2.39E-05

TABLE 3.1-4 (2 of 3,)

TABLE 3.1-4 (3 of 3)

INGESTION DOSE FACTORS FOR CHILD (MREM PER PCI INGESTED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

BA140 8.31E-05 7.28E-08 4.85E-06 NO DATA 2.37E-08 4.34E-08 4.21E-05 BA141 2.OOE-07 1.12E-10 6.51E-09 NO DATA 9.69E-11 6.58E-10 1.14E-07 BA142 8.74E-08 6.29E-11 4.88E-09 NO DATA 5.09E-11 3.70E-11 1.14E-09

LA140 1.01E-08 3.53E-09 1.19E-09 NO DATA NO DATA NO DATA 9.84E-05 LA142 5.24E-10 1.67E-10 5.23E-11 NO DATA NO DATA NO DATA 3.31E-05 CE141 3.97E-08 1.98E-08 2.94E-09 NO DATA 8.68E-09 NO DATA 2.47E-05

CE143 6.99E-09 3.79E-06 5.49E-10 NO DATA 1.59E-09 NO DATA 5.55E-05 CE144 2.08E-06 6.52E-07 1.11E-07 NO DATA 3.61E-07 NO DATA 1.70E-04 PR143 3.93E-08 1.18E-08 1.95E-09 NO DATA 6.39E-09 NO DATA 4.24E-05

PR144 1.29E-10 3.99E-11 6.49E-12 NO DATA 2.11E-11 NO DATA 8.59E-08 ND147 2.79E-08 2.26E-08 1.75E-09 NO DATA 1.24E-08 NO DATA 3.58E-05 W 187 4.29E-07 2.54E-07 1.14E-07 NO DATA NO DATA NO DATA 3.57E-05

NP239 5.25E-09 3.77E-10 2.65E-10 NO DATA 1.09E-09 NO DATA 2.79E-05

TABLE 3.1-4 (3 of 3,)

TABLE 3.1-5* (1 of 3)

INGESTION DOSE FACTORS FOR INFANT (MREM PER PCI INGESTED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

H 3 NO DATA 3.08E-07 3.08E-07 3.08E-07 3.08E-07 3.08E-07 3.08E-07 NA 24 1.01E-05 1.01E-05 1.01E-05 1.01E-05 1.01E-05 1.01E-05 1.01E-05

P 32 1.70E-03 1.OOE-04 6.59E-05 NO DATA NO DATA NO DATA 2.30E-05 CR 51 NO DATA NO DATA 1.41E-08 9.20E-09 2.01E-09 1.79E-08 4.11E-07 MN 54 NO DATA 1.99E-05 4.51E-06 NO DATA 4.41E-06 NO DATA 7.31E-06

MN 56 NO DATA 8.18E-07 1.41E-07 NO DATA 7.03E-07 NO DATA 7.43E-05 FE 55 1.39E-05 8.98E-06 2.40E-06 NO DATA NO DATA 4.39E-06 1.14E-06 FE 59 3.08E-05 5.38E-05 2.12E-05 NO DATA NO DATA 1.59E-05 2.57E-05

CO 58 NO DATA. 3.60E-06 8.98E-06 NO DATA NO DATA NO DATA 8.97E-06 CO 60 NO DATA 1.08E-05 2.55E-05 NO DATA NO DATA NO DATA 2.57E-05 NI 63 6.34E-04 3.92E-05 2.20E-05 NO DATA NO DATA NO DATA 1.95E-06

NI 65 4.70E-06 5.32E-07 2.42E-07 NO DATA NO DATA NO DATA 4.05E-05 CU 64 NO DATA 6.09E-07 2.82E-07 NO DATA 1.03E-06 NO DATA 1.25E-05 ZN 65 1.84E-05 6.31E-05 2.91E-05 NO DATA 3.06E-05 NO DATA 5.33E-05

ZN 69 9.33E-08 1.68E-07 1.25E-08 NO DATA 6.98E-08 NO DATA 1.37E-05 BR 83 NO DATA NO DATA 3.63E-07 NO DATA NO DATA NO DATA LT E-24 BR 84 NO DATA NO DATA 3.82E-07 NO DATA NO DATA NO DATA LT E-24

BR 85 NO DATA NO DATA 1.94E-08 NO DATA NO DATA NO DATA LT E-24 RB 86 NO DATA 1.70E-04 8.40E-05 NO DATA NO DATA NO DATA 4.35E-06 RB 88 NO DATA 4.98E-07 2.73E-07 NO DATA NO DATA NO DATA 4.85E-07

RB 89 NO DATA 2.86E-07 1.97E-07 NO DATA NO DATA NO DATA 9.74E-08 SR 89 2.51E-03 NO DATA 7.20E-05 NO DATA NO DATA NO DATA 5.16E-05 SR 90 1.85E-02 NO DATA 4.71E-03 NO DATA NO DATA NO DATA 2.31E-04

SR 91 5.OOE-05 NO DATA 1.81E-06 NO DATA NO DATA NO DATA 5.92E-05 SR 92 1.92E-05 NO DATA 7.13E-07 NO DATA NO DATA NO DATA 2.07E-04 Y 90 8.69E-08 NO DATA 2.33E-09 NO DATA NO DATA NO DATA 1.20E-04

Y 91M 8.10E-10 NO DATA 2.76E-11 NO DATA NO DATA NO DATA 2.70E-06 Y 91 1.13E-06 NO DATA 3.01E-08 NO DATA NO DATA NO DATA 8.10E-05 Y 92 7.65E-09 NO DATA 2.15E-10 NO DATA NO DATA NO DATA 1.46E-04

*Taken from Regulatory Guide 1.109 (Rev. 1)

TABLE 3.1-5 (1 of 3)

Revision 1

TABLE 3.1-5 (2 of 3)

INGESTION DOSE FACTORS FOR INFANT (MREM PER PCI INGESTED)

NUCLIDE BONE LIVER T.'BODY THYROID KIDNEY LUNG GI-LLI

Y 93 2.43E-08 NO DATA 6.62E-10 NO DATA NO DATA NO DATA 1.92E-04 ZR 95 2.06E-07 5.02E-08 3.56E-08 NO DATA 5.'41E-08 NO DATA 2.50E-05 ZR 97 1.48E-08 2.54E-09 1.16E-09 NO DATA 2.56E-09 NO DATA 1.62E-04

NB 95 4.20E-08 1.73E-08 1.OOE-08 NO DATA 1.24E-08 NO DATA 1.46E-05 MO 99 NO DATA 3.40E-05 6.63E-06 NO DATA 5.08E-05 NO DATA 1.12E-05 TC 99M 1.92E-09 3.96E-09 5.10E-08 NO DATA 4.26E-08 2.07E-09 1.15E-06

TC101 2.27E-09 2.86E-09 2.83E-08 NO DATA 3.40E-08 1.56E-09 4.86E-07 RU103 1.48E-06 NO DATA 4.95E-07 NO DATA 3.08E-06 NO DATA 1.80E-05 RU105 1.36E-07 NO DATA 4.58E-08 NO DATA 1.00E-06 NO DATA 5.41E-05

RU106 2.41E-05 NO DATA 3.01E-06 NO'DATA 2.85E-05 NO DATA 1.83E-04 AG110M 9.96E-07 7.27E-07 4.81E-07 NO DATA 1.04E-06 NO DATA 3.77E-05 TE125M 2.33E-05 7.79E-06 3.15E-06 7.84E-06 NO DATA NO DATA 1.11E-05

TE127M 5.85E-05 1.94E-05 7.08E-06 1.69E-05 1.44E-04 NO DATA 2.36E-05 TE127 1.00E-06 3.35E-07 2.15E-07 8.14E-07 2.44E-06 NO DATA 2.10E-05 TE129M 1.00E-04 3.43E-05 1.54E-05 3.84E-05 2.50E-04 NO DATA 5.97E-05

TE129 2.84E-07 9.79E-08 6.63E-08 2.38E-07 7.07E-07 NO DATA 2.27E-05 TE131M 1.52E-05 6.12E-06 5.05E-06 1.24E-05 4.21E-05 NO DATA 1.03E-04 TE131 1.76E-07 6.50E-08 4.94E-08 1.57E-07 4.50E-07 NO DATA 7.11E-06

TE132 2.08E-05 1.03E-05 9.61E-06 1.52E-05 6.44E-05 NO DATA 3.81E-05 I 130 6.OOE-06 1.32E-05 5.30E-06 1.48E-03 1.45E-05 NO DATA 2.83E-06 I 131 3.59E-05 4.23E-05 1.86E-05 1.39E-02 4.94E-05 NO DATA 1.51E-06

I 132 1.66E-06 3.37E-06 1.20E-06 1.58E-04 3.76E-07 NO DATA 2.73E-06 I 133 1.25E-05 1.82E-05 5.33E-06 3.31E-03 2.14E-05 NO DATA 3.08E-06 I 134 8.69E-07 1.78E-06 6.33E-07 4.15E-05 1.99E-06 NO DATA 1.84E-06

I 135 3.64E-06 7.24E-06 2.64E-06 6.49E-04 8.07E-06 NO DATA 2.62E-06 CS134 3.77E-04 7.03E-04 7.10E-05 NO DATA 1.81E-04 7.42E-05 1.91E-06 CS136 4.59E-05 1.35E-04 5.04E-05 NO DATA 5.38E-05 1.10E-05 2.05E-06

CS137 5.22E-04 6.11E-04 4.33E-05 NO DATA 1.64E-04 6.64E-05 1.91E-06 CS138 4.81E-07 7.82E-07 3.79E-07 NO DATA 3.90E-07 6.09E-08 1.25E-06 BA139 8.81E-07 5.84E-10 2.55E-08 NO DATA 3.51E-10 3.54E-10 5.58E-05

TABLE 3.1-5 (2 of 3)

TABLE 3.1-5 (3 of 3)

INGESTION DOSE FACTORS FOR INFANT (MREM PER PCI INGESTED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

BA140 1.71E-04 1.71E-07 8.81E-06 NO DATA 4.06E-08 1.05E-07 4.20E-05 BA141 4.25E-07 2.91E-10 1.34E-08 NO DATA 1.75E-10 1.77E-10' 5.19E-06 BA142 1.84E-07 1.53E-10 9.06E-09 NO DATA 8.81E-11 9.26E-11 7.59E-07

LA140 2.11E-08 8.32E-09 2.14E-09 NO DATA NO DATA NO DATA 9.77E-05 LA142 1.10E-09 4.04E-10 9.67E-11 NO DATA NO DATA NO DATA 6.86E-05 CE141 7.87E-08 4.80E-08 5.65E-09 NO DATA 1.48E-08 NO DATA 2.48E-05

CE143 1.48E-08 9.82E-06 1.12E-09 NO DATA 2.86E-09 NO DATA 5.73E-05 CE144 2.98E-06 1.22E-06 1.67E-07 NO DATA 4.93E-07 NO DATA 1.71E-04 PR143 8.13E-08 3.04E-08 4.03E-09 NO DATA 1.13E-08 NO DATA 4.29E-05

PR144 2.74E-10 1.06E-10 1.38E-11 NO DATA 3.84E-11 NO DATA 4.93E-06 ND147 5.53E-08 5.68E-08 3.48E-09 NO DATA 2.19E-08 NO DATA 3.60E-05 W 187 9.03E-07 6.28E-07 2.17E-07 NO DATA NO DATA NO DATA 3.69E-05

NP239 1.11E-08 9.93E-10 5.61E-10 NO DATA 1.98E-09 NO DATA 2.87E-05

TABLE 3.1-5 (3 of 3)

TABLE 3.1-6* (1 of 3)

INHALATION DOSEFACTORS FOR ADULTS (MREM PER PCI INHALED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

H 3 NO DATA 1.58E-07 1.58E-07 1.58E-07 1.58E-07 1.58E-07 1.58E-07 NA 24 1.28E-06 1.28E-06 1.28E-06 1.28E-06 1.28E-06 1.26E-06 1.28E-06

P 32 1.65E-04 9.64E-06 6.26E-06 NO DATA NO DATA NO DATA 1.08E-05 CR 51 NO DATA NO DATA 1.25E-08 7.44E-09 2.85E-09 1.80E-06 4.15E-07 MN 54 NO DATA 4.95E-06 7.87E-07 NO DATA 1.23E-06 1.75E-04 9.67E-06

MN 56 NO DATA 1.55E-10 2.29E-11 NO DATA 1.63E-10 1.18E-06 2.53E-06 FE 55 3.07E-06 2.12E-06 4.93E-07 NO DATA NO DATA 9.01E-06 7.54E-07 FE 59 1.47E-06 3.47E-06 1.32E-06 NO DATA NO DATA 1.27E-04 2.35E-05

CO 58 NO DATA 1.98E-07 2.59E-07 NO DATA NO DATA 1.16E-04 1.33E-05 CO 60 NO DATA 1.44E-06 1.85E-06 NO DATA NO DATA 7.46E-04 3.56E-05 NI 63 5.40E-05 3.93E-06 1.81E-06 NO DATA NO DATA 2.23E-05 1.67E-06

NI 65 1.92E-10 2.62E-11 1.14E-11 NO DATA NO DATA 7.OOE-07 1.54E-06 CU 64 NO DATA 1.83E-10 7.69E-11 NO DATA 5.78E-10 8.48E-07 6.12E-06 ZN 65 4.05E-06 1.29E-05 5.82E-06 NO DATA 8.62E-06 1.08E-04 6.68E-06

ZN 69 4.23E-12 8.14E-12 5.65E-13 NO DATA 5.27E-12 1.15E-07 2.04E-09 BR 83 NO DATA NO DATA 3.01E-08 NO DATA NO DATA NO DATA 2.90E-08 BR 84 NO DATA NO DATA 3.91E-08 NO DATA NO DATA NO DATA 2.05E-13

BR 85 NO DATA NO DATA 1.60E-09 NO DATA NO DATA NO DATA LT E-24 RB 86 NO DATA 1.69E-05 7.37E-06 NO DATA NO DATA NO DATA 2.08E-06 RB 88 NO DATA 4.84E-08 2.41E-08 NO DATA NO DATA NO DATA 4.18E-19

RB 89 NO DATA 3.20E-08 2.12E-08 NO DATA NO DATA NO DATA 1.16E-21 SR 89 3.80E-05 NO DATA 1.09E-06 NO DATA NO DATA 1.75E-04 4.37E-05 SR 90 1.24E-02 NO DATA 7.62E-04 NO DATA NO DATA 1.20E-03 9.02E-05

SR 91 7.74E-09 NO DATA 3.13E-10 NO DATA NO DATA 4.56E-06 2.39E-05 SR 92 8.43E-10 NO DATA 3.64E-11 NO DATA NO DATA 2.06E-06 5.38E-06 Y 90 2.61E-07 NO DATA 7.01E-09 NO DATA NO DATA 2.12E-05 6.32E-05

Y 91M 3.26E-11 NO DATA 1.27E-12 NO DATA NO DATA 2.40E-07 1.66E-10 Y 91 5.78E-05 NO DATA 1.55E-06 NO DATA NO DATA 2.13E-04 4.81E-05 Y 92 1.29E-09 NO DATA 3.77E-11 NO DATA NO DATA 1.96E-06 9.19E-06

*Taken from Regulatory Guide 1.109 (Rev. 1)

TABLE 3.1-6 (1 of 3)

Revision 1

TABLE 3.1-6 (2 of 3)

INHALATION DOSE FACTORS FOR ADULTS (MREM PER PCI INHALED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI.

Y 93 1.18E-08 NO DATA 3.26E-10 NO DATA NO DATA 6.06E-06 5.27E-05 ZR 95 1.34E-05 4.30E-06 2.91E-06 NO DATA 6.77E-06 2.21E-04 1.88E-05 ZR 97 1.21E-08 2.45E-09 1.13E-09 NO DATA 3.71E-09 9.84E-06 6.54E-05

NB 95 1.76E-06 9.77E-07 5.26E-07 NO DATA 9.67E-07 6.31E-05 1.30E-05 MO 99 NO DATA 1.51E-08 2.87E-09 NO DATA 3.64E-08 1.14E-05 3.10E-05 *TC 99M 1.29E-13 3.64E-13 4.63E-12 NO DATA 5.52E-12 9.55E-08 5.20E-07

TC 101 5.22E-15 7.52E-15 7.38E-14 NO DATA 1.35E-13 4.99E-08 1.36E-21 RU 103 1.91E-07 NO DATA 8.23E-08 NO DATA 7.29E-07 6.31E-05 1.38E-05 RU 105 9.88E-11 NO DATA 3.89E-11 NO DATA 1.27E-10 1.37E-06 6.02E-06

RU 106 8.64E-06 NO DATA 1.09E-06 NO DATA 1.67E-05 1.17E-03 1.14E-04 AG 110M 1.35E-06 1.25E-06 7.43E-07 NO DATA 2.46E-06 5.79E-04 3.78E-05 TE 125M 4.27E-07 1.98E-07 5.84E-08 1.31E-07 1.55E-06 3.92E-05 8.83E-06

TE 127M 1.58E-06 7.21E-07 1.96E-07 4.11E-07 5.72E-06 1.20E-04 1.87E-05 TE 127 1.75E-10 8.03E-11 3.87E-11 1.32E-10 6.37E-10 8.14E-07 7.17E-06 TE 129M 1.22E-06 5.84E-07 1.98E-07 4.30E-07 4.57E-06 1.45E-04 4.79E-05

TE 129 6.22E-12 2.99E-12 1.55E-1 2 4.87E-12 2.34E-11 2.42E-07 1.96E-08 TE 131M 8.74E-09 5.45E-09 3.63E-09 6.88E-09 .3.86E-08 1.82E-05 6.95E-05 TE 131 1.39E-12 7.44E-13 4.49E-13 1.17E-12 5.46E-12 1.74E-07 2.30E-09

TE 132 3.25E-08 2.69E-08 2.02E-08 2.37E-08 1.82E-07 3.60E-05 6.37E-05 I 130 5.72E-07 1.68E-06 6.60E-07 1.42E-04 2.61E-06 NO DATA 9.61E-07 I 131 3.15E-06 4.47E-06 2.56E-06 1.49E-03 7.66E-06 NO DATA 7.85E-07

I 132 1.45E-07 4.07E-07 1.45E-07 1.43E-05 6.48E-07 NO DATA 5.08E-08 I 133 1.08E-06 1.85E-06 5.65E-07 2.69E-04 3.23E-06 NO DATA 1.11E-06 I 134 8.05E-08 2.16E-07 7.69E-08 3.73E-06 3.44E-07 NO DATA 1.26E-10

I 135 3.35E-07 8.73E-07 3.21E-07 5.60E-05 1.39E-06 NO DATA 6.56E-07 CS 134 4.66E-05 1.06E-04 9.10E-05 NO DATA 3.59E-05 1.22E-05 1.30E-06 CS 136 4.88E-06 1.83E-05 1.38E-05 NO DATA 1.07E-05 1.50E-06 1.46E-06

CS 137 5.98E-05 7.76E-05 5.35E-05 NO DATA 2.78E-05 9.40E-06 1.05E-06 CS 138 4.14E-08 7.76E-08 4.05E-08 NO DATA 6.OOE-08 6.07E-09 2.33E-13 BA 139 1.17E-10 8.32E-14 3.42E-12 NO DATA 7.78E-14 4.70E-07 1.12E-07

TABLE 3.1-6 (2 of 3)

TABLE 3.1-6 (3 of 3)

INHALATION DOSE FACTORS FOR ADULTS (MREM PER PCI INHALED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

BA 140 4.88E-06 6.13E-09 3.21E-07 NO DATA 2.09E-09 1.59E-04 2.73E-05 BA 141 1.25E-11 9.14E-15 4.20E-13 NO DATA 8.75E-15 2.42E-07 1.45E-17 BA 142 3.29E-12 3.38E-15 2.07E-13 NO DATA 2.86E-15 1.49E-07 1.96E-26

LA 140 4.30E-08 2.17E-08 5.73E-09 NO DATA NO DATA 1.70E-05 5.73E-05 LA 142 8.54E-11 3.88E-11 9.65E-12 NO DATA NO DATA 7.91E-07 2.64E-07 CE 141 2.49E-06 1.69E-06 1.91E-07 NO DATA 7.83E-07 4.52E-05 1.50E-05

CE 143 2.33E-08 1.72E-08 1.91E-09 NO DATA 7.60E-09 9.97E-06 2.83E-05 CE 144 4.29E-04 1.79E-04 2.30E-05 NO DATA 1.06E-04 9.72E-04 1.02E-04 PR 143 1.17E-06 4.69E-07 5.80E-08 NO DATA 2.70E-07 3.51E-05 2.50E-05

PR 144 3.76E-12 1.56E-12 1.91E-13 NO DATA 8.81E-13 1.27E-07 2.69E-18 ND 147 6.59E-07 7.62E-07 4.56E-08 NO DATA 4.45E-07 2.76E-05 2.16E-05 W 187 1.06E-09 8.85E-10 3.10E-10 NO DATA NO DATA 3.63E-06 1.94E-05

NP 239 2.87E-08 2.82E-09 1.55E-09 NO DATA 8.75E-09 4.70E-06 1.49E-05

TABLE 3.1-6 (3 of 3)

TABLE 3.1-7* (1 of 3)

INHALATION DOSE FACTORS FOR TEENAGER (MREM PER PCI INHALED)

YUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

H 3 NO DATA 1.59E-07 1.59E-07 1.59E-07 1.59E-07 1.59E-07 1.59E-07 NA 24 1.72E-06 1.72E-06 .1.72E-06 1.72E-06 1.72E-06 1.72E-06 1.72E-06

P 32 2.36E-04 1.37E-05 8.95E-06 NO DATA NO DATA NO DATA 1.16E-05 CR 51 NO DATA NO DATA 1.69E-08 9.37E-09 3.84E-09 2.62E-06 3.75E-07 M'IN 54 NO DATA 6.39E-06 1.05E-06 NO DATA 1.59E-06 2.48E-04 8.35E-06

MN 56 NO DATA 2.12E-10 3.15E-11 NO DATA 2.24E-10 1.90E-06 7.18E-06 FE 55 4.18E-06 2.98E-06 6.93E-07 NO DATA NO DATA 1.55E-05 7.99E-07 FE 59 1.99E-06 4.62E-06 1.79E-06 NO DATA NO DATA 1.91E-04 2.23E-05

CO 58 NO DATA 2.59E-07 3.47E-07 NO DATA NO DATA 1.68E-04 1.19E-05 CO 60 NO DATA 1.89E-06 2.48E-06 NO DATA NO DATA 1.09E-03 3.24E-05 NI 63 7.25E-05 5.43E-06 2.47E-06 NO DATA NO DATA 3.84E-05 1.77E-06

NI 65 2.73E-10 3.66E-11 1.59E-11 NO DATA NO DATA 1.17E-06 4.59E-06 CU 64 NO DATA 2.54E-10 1.06E-10 NO DATA 8.01E-10 1.39E-06 7.68E-06 ZN 65 4.82E-06 1.67E-05 7.80E-06 NO DATA 1.08E-05 1.55E-04 5.83E-06

ZN 69 6.04E-12 1.15E-11 8.07E-13 NO DATA 7.53E-12 1.98E-07 3.56E-08 BR 83 NO DATA NO DATA 4.30E-08 NO DATA NO DATA NO DATA LT E-24 BR 84 NO DATA NO DATA 5.41E-08 NO DATA NO DATA NO DATA LT E-24

BR 85 NO DATA NO DATA 2.29E-09 NO DATA NO DATA NO DATA LT E-24 RB 86 NO DATA 2.38E-05 1.05E-05 NO DATA NO DATA NO DATA 2.21E-06 RB 88 NO DATA 6.82E-08 3.40E-08 NO DATA NO DATA NO DATA 3.65E-15

RB 89 NO DATA 4.40E-08 2.91E-08 NO DATA NO DATA NO DATA 4.22E-17 SR 89 5.43E-05 NO DATA 1.56E-06 NO DATA NO DATA 3.02E-04 4.64E-05 SR 90 1.35E-02 NO DATA 8.35E-04 NO DATA NO DATA 2.06E-03 9.56E-05

SR 91 1.10E-08 NO DATA 4.39E-10 NO DATA NO DATA 7.59E-06 3.24E-05 SR 92 1.19E-09 NO DATA 5.08E-11 NO DATA NO DATA 3.43E-06 1.49E-05 Y 90 3.73E-07 NO DATA 1.OOE-08 NO DATA NO DATA 3.66E-05 6.99E-05

Y 911 4.63E-11 NO DATA 1.77E-12 NO DATA NO DATA 4.OOE-07 3.77E-09 Y 91 8.26E-05 NO DATA 2.21E-06 NO DATA NO DATA 3.67E-04 5.11E-05 Y 92 1.84E-09 NO DATA 5.36E-11 NO DATA NO DATA 3.35E-06 2.06E-05

*Taken from Regulatory Guide 1.109 (Rev. 1)

TABLE 3.1-7 (1 of 3)

Revision 1

TABLE 3.1-7 (2 of 3)

INHALATION DOSE FACTORS FOR TEENAGER (MREM PER PCI INHALED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

Y 93 1.69E-08 NO DATA 4.65E-10 NO DATA NO DATA 1.04E-05 7.24E-05. ZR 95 1.82E-05 5.73E-06 3.94E-06 NO DATA 8.42E-06 3.36E-04 1.86E-05 ZR 97 1.72E-08 3.40E-09 1.57E-09 NO DATA 5.15E-09 1.62E-05 7.88E-05

NB 95 2.32E-06 1.29E-06 7.08E-07 NO DATA 1.25E-06 9.39E-05 1.21E-05 MO 99 NO DATA 2.11E-08 4.03E-09 NO DATA 5.14E-08 1.92E-05 3.36E-05 TC 99M 1.73E-13 4.83E-13 6.24E-12 NO DATA 7.20E-12 1.44E-07 7.66E-07

TC 101 7.40E-15 1.05E-14 1.03E-13 NO DATA 1.90E-13 8.34E-08 1.09E-16 RU 103 2.63E-07 NO DATA 1.12E-07 NO DATA 9.29E-07 9.79E-07 1.36E-05 RU 105 1.40E-10 NO DATA 5.42E-11 NO DATA 1.76E-10 2.27E-06 1.13E-05

RU 106 1.23E-05 NO DATA 1.55E-06 NO DATA 2.38E-05 2.01E-03 1.20E04 AG 110M 1.73E-06 1.64E-06 9.99E-07 NO DATA 3.13E-06 8.44E-04 3.41E-05 TE 125M 6.10E-07 2.80E-07 8.34E-08 1.75E-07 NO DATA 6.70E-05 9.38E-06

TE 127M 2.25E-06 1.02E-06 2.73E-07 5.48E-07 8.17E-06 2.07E-04 1.99E-05 TE 127 2.51E-10 1.14E-10 5.52E-11 1.77E-10 9.10E-10 1.40E-06 1.01E-05 TE 129M 1.74E-06 8.23E-07 2.81E-07 5.72E-07 6.49E-06 2.47E-04 5.06E-05

TE 129 8.87E-12 4.22E-12 2.20E-12 6.48E-12 3.32E-11 4.12E-07 2.02E-07

TE 131M 1.23E-08 7.51E-09 5.03E-09 9.06E-09 5.49E-08 2.97E-05 7.76E-05 TE 131 1.97E-12 1.04E-12 6.30E-13 1.55E-12 7.72E-12 2.92E-07 1.89E-09

TE 132 4.50E-08 3.63E-08 2.74E-08 3.07E-08 2.44E-07 5.61E-05 5.79E-05 I 130 7.80E-07 2.24E-06 8.96E-07 1.86E-04 3.44E-06 NO DATA 1.14E-06 I 131 4.43E-06 6.14E-06 3.30E-06 1.83E-03 1.05E-05 NO DATA 8.11E-07

I 132 1.99E-07 5.47E-07 1.97E-07 1.89E-05 8.65E-07 NO DATA 1.59E-07 I 133 1.52E-06 2.56E-06 7.78E-07 3.65E-04 4.49E-06 NO DATA 1.29E-06 I 134 1.11E-07 2.90E-07 1.05E-07 4.94E-06 4.58E-07 NO DATA 2.55E-09

I 135 4.62E-07 1.18E-06 4.36E-07 7.76E-05 1.86E-06 NO DATA 8.69E-07 CS 134 6.28E-05 1.41E-04 6.86E-05 NO DATA 4.69E-05 1.83E-05 .22E-06 CS 136 6.44E-06 2.42E-05 1.71E-05 NO DATA 1.38E-05 2.22E-06 1.36E-06

CS 137 8.38E-05 1.06E-04 3.89E-05 NO DATA 3.80E-05 1.51E-05 1.06E-06 CS 138 5.82E-08 1.07E-07 5.58E-08 NO DATA 8.28E-08 9.84E-09 3.38E-11 BA 139 1.67E-10 1.18E-13 4.87E-12 NO DATA 1.11E-13 8.08E-07 8.06E-07

TABLE 3.1-7

(2 of 3)

TABLE 3.1-7 (3 of 3)

INHALATION DOSE FACTORS FOR TEENAGER (MREM PER PCI INHALED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

BA 140 6.84E-06 8.38E-09 4.40E-07 NO DATA 2.85E-09 2.54E-04 2.86E-05 BA 141 1.78E-11 1.32E-14 5.93E-13 NO DATA 1.23E-14 4.11E-07 9.33E-14 BA 142 4.62E-12 4.63E-15 2.84E-13 NO DATA 3.92E-15 2.39E-07 5.99E-20

LA 140 5.99E-08 2.95E-08 7.82E-09 NO DATA NO DATA 2.68E-05 6.09E-05 LA 142 1.20E-10 5.31E-11 1.32E-11 NO DATA NO DATA 1.27E-06 1.50E-06 CE 141 3.55E-06 2.37E-06 2.71E-07 NO DATA 1.11E-06 7.67E-05 1.58E-05

CE 143 3.32E-08 2.42E-08 2.70E-09 NO DATA 1.08E-08 1.63E-05 3.19E-05 CE 144 6.11E-04 2.53E-04 3.28E-05 NO DATA 1.51E-04 1.67E-03 1.08E-04 PR 143 1.67E-06 6.64E-07 8.28E-08 NO DATA 3.86E-07 6.04E-05 2.67E-05

PR 144 5.37E-12 2.20E-12 2.72E-13 NO DATA 1.26E-12 2.19E-07 2.94E-14 ND 147 9.83E-07 1.07E-06 6.41E-08 NO DATA 6.28E-07 4.65E-05 2.28E-05 W 187 1.50E-09 1.22E-09 4.29E-10 NO DATA NO DATA 5.92E-06 2.21E-05

NP 239 4.23E-08 3.99E-09 2.21E-09 NO DATA 1.25E-08 8.11E-06 1.65E-05

TABLE 3.1-7 (3 of 3)

TABLE 3.1-8* (1 of 3)

INHALATION DOSE FACTORS FOR CHILD (MREM PER PCI INHALED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

H 3 NO DATA 3.04E-07 3.04E-07 3.04E-07 3.04E-07 3.04E-07 3.04E-07 NA 24 4.35E-06 4.35E-06 4.35E-06 4.35E-06 4.35E-06 4.35E-05 4.35E-06

P 32 7.04E-04 3.09E-05 2.67E-05 NO DATA NO DATA NO DATA 1.14E-05 CR 51 NO DATA NO DATA 4.17E-08 2.31E-08 6.57E-09 4.59E-06 2.93E-07 MN 54 NO DATA 1.16E-05 2.57E-06 NO DATA 2.71E-06 4.26E-04 6.19E-06

MN 56 NO DATA 4.48E-10 8.43E-11 NO DATA 4.52E-10 3.55E-06 3.33E-05 FE 55 1.28E-05 6.80E-06 2.10E-06 NO DATA NO DATA 3.OOE-05 7.75E-07 FE 59 5.59E-06 9.04E-06 4.51E-06 NO DATA NO DATA 3.43E-04 1.91E-05

CO 58 NO DATA 4.79E-07 8.55E-07 NO DATA NO DATA 2.99E-04 9.29E-06 CO 60 NO DATA 3.55E-06 6.12E-06 NO DATA NO DATA 1.91E-03 2.60E-05 NI 63 2.22E-04 1.25E-05 7.56E-06 NO DATA NO DATA 7.43E-05 1.71E-06

NI 65 8.08E-10 7.99E-11 4.44E-11 NO DATA NO DATA 2.21E-06 2.27E-05 CU 64 NO DATA 5.39E-10 2.90E-10 NO DATA 1.63E-09 2.59E-06 9.92E-06 ZN 65 1.15E-05 3.06E-05 1.90E-05 NO DATA 1.93E-05 2.69E-04 4.41E-06

ZN 69 1.81E-11 2.61E-11 2.41E-12 NO DATA 1.58E-11 3.84E-07 2.75E-06 BR 83 NO DATA NO DATA 1.28E-07 NO DATA NO DATA NO DATA LT E-24 BR 84 NO DATA NO DATA 1.48E-07 NO DATA NO DATA NO DATA LT E-24

BR 85 NO DATA NO DATA 6.84E-09 NO DATA NO DATA NO DATA LT E-24 RB 86 NO DATA 5.36E-05 3.09E-05 NO DATA NO DATA NO DATA 2.16E-06 RB 88 NO DATA 1.52E-07 9.90E-08 NO DATA NO DATA NO DATA 4.66E-09

RB 89 NO DATA 9.33E-08 7.83E-08 NO DATA NO DATA NO DATA 5.11E-10 SR 89 1.62E-04 NO DATA 4.66E-06 NO DATA NO DATA 5.83E-04 4.52E-05 SR 90 2.73E-02 NO DATA 1.74E-03 NO DATA NO DATA 3.99E-03 9.28E-O5

SR 91 3.28E-08 NO DATA 1.24E-09 NO DATA NO DATA 1.44E-05 4.70E-05 SR 92 3.54E-09 NO DATA 1.42E-10 NO DATA NO DATA 6.49E-06 6.55E-05 Y 90 1.11E-06 NO DATA 2.99E-08 NO DATA NO DATA 7.07E-05 7.24E-05

Y 91M 1.37E-10 NO DATA 4.98E-12 NO DATA NO DATA 7.60E-07 4.64E-07 Y 91 2.47E-04 NO DATA 6.59E-06 NO DATA NO DATA 7.10E-04 4.97E-05 Y 92 5.50E-09 NO DATA 1.57E-10 NO DATA NO DATA 6.46E-06 6.46E-05

*Taken From Regulatory Guide 1.109 (Rev. 1)

TABLE 3.1-8 (1 of 3)

Revision 1

TABLE 3.1-8 (2 of 3)

INHALATION DOSE FACTORS FOR CHILD (MREM PER PCI INHALED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI

Y 93 5.04E-08 NO DATA 1.38E-09 NO DATA NO DATA 2.01E-05 1.05E-04 ZR 95 5.13E-05 1.13E-05 1.00E-05 NO DATA 1.61E-05 6.03E-04 1.65E-05 ZR 97 5.07E-08 7.34E-09 4.32E-09 NO DATA 1.05E-08 3.06E-05 9.49E-05

NB 95 6.35E-06 2.48E-06 1.77E-06 NO DATA 2.33E-06 1.66E-04 1.00E-05 MO 99 NO DATA 4.66E-08 1.15E-08 NO DATA 1.06E-07 3.66E-05 3.42E-05 TC 99M 4.81E-13 9.41E-13 1.56E-11 NO DATA 1.37E-11 2.57E-07 1.30E-06

TC101 2.19E-14 2.30E-14 2.91E-13 NO DATA 3.92E-13 1.58E-07 4.41E-09 RU103 7.55E-07 NO DATA 2.90E-07 NO DATA 1.90E-06 1.79E-04 1.21E-05 RU105 4.13E-10 NO DATA 1.50E-10 NO DATA 3.63E-10 4.30E-06 2.69E-05

RU106 3.68E-05 NO DATA 4.57E-06 NO DATA 4.97E-05 3.87E-03 1.16E-04 AG110M 4.56E-06 3.08E-06 2.47E-06 NO DATA 5.74E-06 1.48E-03 2.71E-05 TE125M 1.82E-06 6.29E-07 2.47E-07 5.20E-07 NO DATA 1.29E-04 9.13E-06

TE127M 6.72E-06 2.31E-06 8.16E-07 1.64E-06 1.72E-05 4.OOE-04 1.93E-05 TE127 7.49E-10 2.57E-10 1.65E-10 5.30E-10 1.91E-09 2.71E-06 1.52E-05 TE129M 5.19E-06 1.85E-06 8.22E-07 1.71E-06 1.36E-05 4.76E-04 4.91E-05

TE129 2.64E-11 9.45E-12 6.44E-12 1.93E-11 6.94E-11 7.93E-07 6.89E-06 TE131M 3.63E-08 1.60E-08 1.37E-08 2.64E-08 1.08E-07 5.56E-05 8.32E-05 TE131 5.87E-12 2.28E-12 1.78E-12 4.59E-12 1.59E-11 5.55E-07 3.60E-07

TE132 1.30E-07 7.36E-08 7.12E-08 8.58E-08 4.79E-07 1.02E-04 3.72E-05 I 130 2.21E-06 4.43E-06 2.28E-06 4.99E-04 6.61E-06 NO DATA 1.38E-06 I 131 1.30E-05 1.30E-05 7.37E-06 4.39E-03 2.13E-05 NO DATA 7.68E-07

I 132 5.72E-07 1.10E-06 5.07E-07 5.23E-05 1.69E-06 NO DATA 8.65E-07 I 133 4.48E-06 5.49E-06 2.08E-06 1.04E-03 9.13E-06 NO DATA 1.48E-06 I 134 3.17E-07 5.84E-07 2.69E-07 1.37E-05 8.92E-07 NO DATA 2.58E-07

I 135 1.33E-06 2.36E-06 1.12E-06 2.14E-04 3.62E-06 NO DATA 1.20E-06 CS134 1.76E-04 2.74E-04 6.07E-05 NO DATA 8.93E-05 3.27E-05 1.04E-06 CS136 1.76E-05 4.62E-05 3.14E-05 NO DATA 2.58E-05 3.93E-06 1.13E-06

CS137 2.45E-04 2.23E-04 3.47E-05 NO DATA 7.63E-05 2.81E-05 9.78E-07 CS138 1.71E-07 2.27E-07 1.50E-07 NO DATA 1.68E-07 1.84E-08 7.29E-08 BA139 4.98E-10 2.66E-13 1.45E-11 NO DATA 2.33E-13 1.56E-06 1.56E-05

TABLE 3.1-8 (2 of 3)

TABLE 3.1-8 (3 of 3)

INHALATION DOSE FACTORS FOR CHILD (MREM PER PCI INHALED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LII

BA140 2.OOE-05 1.75E-08 1.17E-06 NO DATA 5.71E-09 4.71E-04 2.75E-05 BA141 5.29E-11 2.95E-14 1.72E-12 NO DATA 2.56E-14 7.89E-07 7.44E-08 BA142 1.35E-11 9.73E-15 7.54E-13 NO DATA 7.87E-15 4.44E-07 7.41E-10

LA140 1.74E-07 6.08E-08 2.04E-08 NO DATA NO DATA 4.94E-05 6.10E-05 LA142 3.50E-10 1.11E-10 3.49E-11 NO DATA NO DATA 2.35E-06 2.05E-05 CE141 1.06E-05 5.28E-06 7.83E-07 NO DATA 2.31E-06 1.47E-04 1.53E-05

CE143 9.89E-08 5.37E-08 7.77E-09 NO DATA 2.26E-08 3.12E-05 3.44E-05 CE144 1.83E-03 5.72E-04 9.77E-05 NO DATA 3.17E-04 3.23E-03 1.05E-04 PR143 4.99E-06 1.50E-06 2.47E-07 NO DATA 8.11E-07 1.17E-04 2.63E-05

PR144 1.61E-11 4.99E-12 8.10E-13 NO DATA 2.64E-12 4.23E-07 5.32E-08 ND147 2.92E-06 2.36E-06 1.84E-07 NO DATA 1.30E-06 8.87E-05 2.22E-05 W 187 4.42E-09 2.61E-09 1.17E-09 NO DATA NO DATA 1.11E-05 2.46E-05

NP239 1.26E-07 9.04E-09 6.35E-09 NO DATA 2.63E-08 1.57E-05 1.73E-05

TABLE 3.1-8 (3 of 3)

TABLE 3.1-9* (1 of 3)

INHALATION DOSE FACTORS FOR INFANT (MREM PER PCI INHALED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LII

H 3 NO DATA 4.62E-07 4.62E-07 3.62E-07 4.62E-07 4.62E-07 4.62E-07 NA 24 7.54E-06 7.54E-06 7.54E-06 7.54E-06 7.54E-06 7.54E-06 7.54E-06

P 32 1.45E-03 8.03E-05 5.53E-05 NO DATA NO DATA NO DATA 1.15E-05 CR 51 NO DATA NO DATA 6.39E-08 4.11E-08 9.45E-09 9.17E-06 2.55E-07 M 54 NO DATA 1.81E-05 3.56E-06 NO DATA 3.56E-06 7.14E-04 5.04E-06

HN 56 NO DATA 1.10E-09 1.58E-10 NO DATA 7.86E-10 8.95E-06 5.12E-05 FE 55 1.41E-05 8.39E-06 2.38E-06 NO DATA NO DATA 6.21E-05 7.82E-07FE 59 9.69E-06 1.68E-05 6.77E-06 NO DATA NO DATA 7.25E-04 1.77E-05

CO 58 NO DATA 8.71E-07 1.30E-06 NO DATA NO DATA 5.55E-04 7.95E-06 CO 60 NO DATA 5.73E-06 8.41E-06 NO DATA NO DATA 3.22E-03 2.28E-05 NI 63 2.42E-04 1.46E-05 8.29E-06 NO DATA NO DATA 1.49E-04 1.73E-06

NI 65 1.71E-09 2.03E-10 8.79E-11 NO DATA NO DATA 5.80E-06 3.58E-05 CO 64 NO DATA 1.34E-09 5.53E-10 NO DATA 2.84E-09 6.64E-06 1.07E-05 ZN 65 1.38E-05 4.47E-05 2.22E-05 NO DATA 2.32E-05 4.62E-04 3.67E-05

ZN 69 3.85E-11 6.91E-11 5.13E-12 NO DATA 2.87E-11 1.05E-06 9.44E-06 BR 83 NO DATA NO DATA 2.72E-07 NO DATA NO DATA NO DATA LT E-24 BR 84 NO DATA NO DATA 2.86E-07 NO DATA NO DATA NO DATA LT E-24

BR 85 NO DATA NO DATA 1.46E-08 NO DATA NO DATA NO DATA LT E-24 RB 86 NO DATA 1.36E-04 6.30E-05 NO DATA NO DATA NO DATA 2.17E-06 RB 88 NO DATA 3.98E-07 2.05E-07 NO DATA NO DATA NO DATA 2.42E-07

RB 89 NO DATA 2.29E-07 1.47E-07 NO DATA NO DATA NO DATA 4.87E-08 SR 89 2.84E-04 NO DATA 8.15E-06 NO DATA NO DATA 1.45E-03 4.57E-05 SR 90 2.92E-02 NO DATA 1.85E-03 NO DATA NO DATA 8.03E-03 9.36E-05

SR 91 6.83E-08 NO DATA 2.47E-09 NO DATA NO DATA 3.76E-05 5.24E-05 SR 92 7.50E-09 NO DATA 2.79E-10 NO DATA NO DATA 1.70E-05 1.00E-04 Y 90 2.35E-06 NO DATA 6.30E-08 NO DATA NO DATA 1.92E-04 7.43E-05

Y 91M 2.91E-10 NO DATA 9.90E-12 NO DATA NO DATA 1.99E-06 1.68E-06 Y 91 4.20E-04 NO DATA 1.12E-05 NO DATA NO DATA 1.75E-03 5.02E-05 Y 92 1.17E-08 NO DATA 3.29E-10 NO DATA NO DATA 1.75E-05 9.04E-05

*Taken from Regulatory Guide 1.109 (Rev. 1)

TABLE 3.1-9 (1 of 3)

Revision 1

TABLE 3.1-9 (2 of 3)

INHALATION DOSE FACTORS FOR INFANT (MREiM PER PCI INHALED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LII

Y 93 1.07E-07 NO DATA 2.91E-09 NO DATA NO DATA 5.46E-05 1.19E-04 ZR 95 8.24E-05 1.99E-05 1.45E-05 NO DATA 2.22E-05 1.25E-03 1.55E-05 ZR 97 1.07E-07 1.83E-08 8.36E-09 NO DATA 1.85E-08 7.88E-05 1.00E-04

NB 95 1.12E-05 4.59E-06 2.70E-06 NO DATA 3.37E-06 3.42E-04 9.05E-06 MO 99 NO DATA 1.18E-07 2.31E-08 NO DATA 1.89E-07 9.63E-05 3.48E-05 TC 99M 9.98E-13 2.06E-12 2.66E-11 NO DATA 2.22E-11 5.79E-07 1.45E-06

TC101 4.65E-14 5.88E-14 5.80E-13 NO DATA 6.99E-13 4.17E-07 6.03E-07 RU103 1.44E-06 NO DATA 4.85E-07 NO DATA 3.03E-06 3.94E-04 1.15E-05 RU105 8.74E-10 NO DATA 2.93E-10 NO DATA 6.42E-10 1.12E-05 3.46E-05

RU106 6.20E-05 NO DATA 7.77E-06 NO DATA 7.61E-05 8.26E-03 1.17E-04 AG110M 7.13E-06 5.16E-06 3.57E-06 NO DATA. 7.80E-06 2.62E-03 2.36E-05 TE125M 3.40E-06 1.42E-06 4.70E-07 1.16E-06 NO DATA 3.19E-04 9.22E-06

TE127M 1.19E-05 4.93E-06 1.48E-06 3.48E-06 2.68E-05 9.37E-04 1.95E-05 TE127 1.59E-09 6.81E-10 3.49E-10 1.32E-09 3.47E-09 7.39E-06 1.74E-05 TE129M 1.01E-05 4.35E-06 1.59E-06 3.91E-06 2.27E-05 1.20E-03 4.93E-05

TE129 5.63E-11 2.48E-11 1.34E-11 4.82E-11 1.25E-10 2.14E-06 1.88E-05 TE131M 7.62E-08 3.93E-08 2.59E-08 6.38E-08 1.89E-07 1.42E-04 8.51E-05 TE131 1.24E-11 5.87E-12 3.57E-12 1.13E-11 2.85E-11 1.47E-06 5.87E-06

TE132 2.66E-07 1.69E-07 1.26E-07 1.99E-07 7.39E-07 2.43E-04 3.15E-05 I 130 4.54E-06 9.91E-06 3.98E-06 1.14E-03 1.09E-05 NO DATA 1.42E-06 I 131 2.71E-05 3.17E-05 1.40E-05 1.06E-02 3.70E-05 NO DATA 7.56E-07

I 132 1.21E-06 2.53E-06 8.99E-07 1.21E-04 2.82E-06 NO DATA 1.36E-06 I 133 9.46E-06 1.37E-05 4.OOE-06 2.54E-03 1.60E-05 NO DATA 1.54E-06 I 134 6.58E-07 1.34E-06 4.75E-07 3.18E-05 1.49E-06 NO DATA 9.21E-07

I 135 2.76E-06 5.43E-06 1.98E-06 4.97E-04 6.05E-06 NO DATA 1.31E-06 CS134 2.83E-04 5.02E-04 5.32E-05 NO DATA 1.36E-04 5.69E-05 9.53E-07 CS136 3.45E-05 9.61E-05 3.78E-05 NO DATA 4.03E-05 8.40E-06 1.02E-06

CS137 3.92E-04 4.37E-04 3.25E-05 NO DATA 1.23E-04 5.09E-05 9.53E-07 CS138 3.61E-07 5.58E-07 2.84E-07 NO DATA 2.93E-07 4.67E-08 6.26E-07 BA139 1.06E-09 7.03E-13 3.07E-11 NO DATA 4.23E-13 4.25E-06 3.64E-05

TABLE 3.1-9 (2 of 3)

TABLE 3.1-9 (3 of 3)

INHALATION DOSE FACTORS FOR INFANT (MREM PER PCI INHALED)

NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LII

BA140 4.OOE-05 4.OOE-08 2.07E-06 NO DATA 9.59E-09 1.14E-03 2.74E-05 BA141 1.12E-10 7.70E-14 3.55E-12 NO DATA 4.64E-14 2.12E-06 3.39E-06 BA142 2.84E-11 2.36E-14 1.40E-12 NO DATA 1.36E-14 1.11E-06 4.95E-07

LA140 3.61E-07 1.43E-07 3.68E-08 NO DATA NO DATA 1.20E-04 6.06E-05 LA142 7.36E-10 2.69E-10 6.46E-11 NO DATA NO DATA 5.87E-06 4.25E-05 CE141 1.98E-05 1.19E-05 1.42E-06 NO DATA 3.75E-06 3.69E-04 1.54E-05

CE143 2.98E-07 1.38E-07 1.58E-08 NO DATA 4.03E-08 8.30E-05 3.55E-05 CE144 2.28E-03 8.65E-04 1.26E-04 NO DATA 3.84E-04 7.03E-03 1.06E-04 PR143 1.00E-05 3.74E-06 4.99E-07 NO DATA 1.41E-06 3.09E-04 2.66E-05

PR144 3.42E-11 1.32E-11 1.72E-12 NO DATA 4.80E-12 1.15E-06 3.06E-06 ND147 5.67E-06 5.81E-06 3.57E-07 NO DATA 2.25E-06 2.30E-04 2.23E-05 W 187 9.26E-09 6.44E-09 2.23E-09 NO DATA NO DATA 2.83E-05 2.54E-05

NP239 2.65E-07 2.37E-08 1.34E-08 NO DATA 4.73E-08 4.25E-05 1.78E-05

TABLE 3.1-9 (3 of 3)

TABLE 3.1-10*

EXTERNAL DOSE FACTORS FOR STANDING'ON CONTAMINATED GROND

(mrem/hr per pCi/m2)

Element Total Body Skin

H1-3 0.0 0.0 NA-24 2.50E-08 2.90E-08 P- 32 0.0 0.0 Cr-5i 2.20E-10 2.60E-10 illn-54 5.80E-09 6.80E-09 4n-56 1.10E-08 1.30E-08 Fe-55 0.0 0.0 Fe-59 8.OOE-09 9.40E-09 Co-58 7.OOE-09 8.20E-09 Co-60 1.70E-08 2.OOE-08 Ni-63 0.0 0.0 Nr-65 3.70E-09 4.30E-09 Cu-64 1 .50E-09 1.70E-09 Zn-65 4. 00E-09 4.60E-09 Zn-69 0.0 0.0 Br-83 6. 40E- 11 9.30E-111 Br-84 1.20E-08 1.40E-08 .Br-85 0.0 0.0 Rb-86 6.30E-10 7.20E-10 Rb-88 3.50E-09 4.OOE-09 Rb -89 1.50E-08 1.80E-08 Sr-89 5.60E-13 6.50E- 13 Sr-91 7.10E-09 8.30E-09 Sr-92 9 .00E-09 1.OOE-08 Y-90 2.20E-12 2.60E-12 Y-91M 3.80E-09 4.40E-09 Y-9 1 2. 40E-11 2.70E- 11 Y-92 1.60E-09 1.90E-09 Y-93 5.70E-10 7.80E-10 Zr-95 5.00E-09 5.80E-09 Zr-97 5 .50E-09 6.40E-09Nb-95 5-IOE-09 6.OOE-09 ,40-99 1.90E-09 2.20E-09 Tc-99M 9.60E-10 1. 10E-09 Tc-101 2.70E-09 3-OOE-09 Ru-103 3.60E-09 4.20E-09 Ru-105 4-50E-09 5-10E-09 Ru-106 1.50E-09 1.80E-09 Ag-110M 1.80E-08 2. 10E-08 Te-125M 3.50E-11 4.80E-11 Te-127M 1.10E-12 1.30E-12 Te-127 I-00E-11 .1-10E-11

Te-129M 7.70E-10 9-OOE-10 Te-129 7-10E-10 8.40E-10 .T Taken from Regulatory Guide 1.109 (Rev. 1)

TABLE 3.1-10 (1lof 2)

Revision 1

TABLE 3.1-10 (cont'd) (2 of 2)

EXTERNAL DOSE FACTORS FOR STANDING ON CONTAMINATED GROUND

(mrem/hr per pCi/rn2)

Element Total Body Skin

Te-131M 8.40E-09 9.90E-09 Te-131 2.20E-09 2.60E-06 ,Te-132 1.70E-09 2.OOE-09 1-130 1.40E-08 1.70E-08 1-131 2.80E-09 3.40E-09 1-132 1.70E-08 2.OOE-08 1-133 3.70E-09 4.50E-09 1-134 1.60E-08 1.90E-08 1-135 1.20E-08 1.40E-08 Cs-134 1.20E-08 1.40E-08 Cs-136 1.50E-08 1.70E-08 Cs-137 4.20E-09 4.90E-09 Cs-138 2.10E-08 24E0 Ba-139 2.40E-09 2.70E-09 Ba-140 2. 10E-09 2.40E-09 Ba-141 4.30E-09 4.90E-09 Ba-142 7.90E-09 9.OOE-09 La-140. 1.50E-08 1.70E-08 La-142 1.50E-08 1.80E-08

WCe-141 5.50E-10 6.20E-10 Ce-143 2.20E-09 2.50E-09 Ce-144 3.20E-10 3. 70E- 10 Pr-143 0.*0 0.0 Pr-144 2.OOE-10 2.30E-10 Nd-147 1.00E-09 1.20E-09 W- 187 3.10E-09 3.60E-09 Np-239. 9.50E-10 1.10E-09

Table 3.1-10 (2 of 2)

STABLE ELEMENT TRANSFER DATA

F m(COW) F f Element Milk (d/2~) Meat (d/kg)

H 1.OE-02 1.2E-02 Na 4.OE-02 3.OE-02A p 2.5E-02 4.6E-02 Cr 2.2E-03 2.4E-03 Mn 2.5E-04 8.OE-04 Fe 1.2E-03 4.OE-02 CO 1 .OE-03 1.3E-02 Ni 6.7E-03 5.3E-02 Cu 1.4E-02 8.OE-03 Zn 3.9E-02 3.OE-02 Rb 3.OE-02 3.1E-02 Sr 8.OE-04 6.OE-04

Y1.OE-05 4.6E-03 Zr 5.OE-06. 3.4E-02 Nb 2.5E-03 2. 8E-01 Mo 7 .5E-03 8.OE-03 Tc 2.5E-02 4.OE-O1 Ru 1.OE-06 4.OE-01

Rh1.OE-02 1.5E-03 Ag 5.OE-02 1.7E-02 Te 1.OE-OO 7.7E-02 I 6.OE-O3~ 2.9E-03 Cs 1. 2E-02 4.OE-03 Ba 4.OE-04 3.2E-03 La S.OE-06 2.OE-04 Ce 1.OE-04 1-2E-03 Pr S.OE-06 4. 7E-03 Nd 5.OE-06 3-3E-03 W 5.OE-04 1.3E-03 Np 5.OE-06 2.OE-04

*Taken from Regulatory Guide 1.109 (Rev. 1)

+ F for goat is 6.QE-2. rn

TABLE 3.1-11 (1 of 1)

Revision 1

APPENDIX A

OCONEE NUCLEAR STATION

SITE SPECIFIC INFORMATION

APPENDIX A TABLE OF CONTENTS

PAGE

A1.0 OCONEE NUCLEAR STATION RADWASTE SYSTEMS A-1

A2.0 RELEASE RATE CALCULATION A-3

A3.0 RADIATION MONITOR SETPOINTS A-6

A4.0 DOSE CALCULATIONS A-9

A5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING A-12

A1.0 OCONEE NUCLEAR STATION RADWASTE SYSTEMS

A1.1 LIQUID RADWASTE PROCESSING

The liquid radwaste system at Oconee Nuclear Station (ONS) is used to collect and treat fluid chemical.and radiochemical by-products of unit operation. The systems produce effluents which can be reused in the plant or discharged in small, dilute quantities to the environment. The means of treatment vary with waste type and desired product in the various systems:

A) Filtration - all waste sources are filtered prior to processing except from the Chemical Treatment Ponds, Waste Oil Collection Basin, Turbine Building Sumps, and Backwash Receiving Tank.

B) Adsorption - adsorption of halides and organic chemicals by activated charcoal (Carbon Filter) is used primarily in treating waste in the Laundry and Hot Shower Tank (LHST).

C) Ion Exchange - ion exchange is used to remove radioactive ions from solution. Also, ion exchange is normally used in removing cations (cobalt, manganese) and anions (chloride, flouride) from evaporator distillates in order to purify the distillates for reuse .as makeup water. Distillate from the Waste Evaporator System or the RC Bleed

Evaporator.in the Boron Recycle System can be treated by this method, as well as MWHUT Waste.

D) Gas Stripping - removal of gaseous radioactive fission products is accomplished in all Evaporators.

E) Distillation - production of pure water from the waste by boiling it away from the contaminated solution which originally contained it is accomplished by both evaporators. Proper control of the process will

yield water which can be reused for makeup. Polishing of this product can be achieved by ion exchange as pointed out above.

F) Concentration - in all Evaporators, dissolved chemicals are concentrated in the lower shell as.water is boiled away. In the case of the Waste Evaporator, the volume of water containing waste chemicals and radioactive cations is reduced so that the waste may be more easily and cheaply solidified and shipped for burial.

Figure A1.0-1 is a schematic representation of the liquid radwaste system at Oconee.

A1.2 GASEOUS WASTE PROCESSING

The purpose of the gaseous waste disposal system is to:

(1) Maintain a non-oxidizing cover gas of nitrogen in tanks and equipment that

contain potentially radioactive gas.

(2) Hold up radioactive gas for decay.

A-1

(3) Release gases (radioactive or non-radioactive) to the atmosphere under controlled conditions.

During power operation of the facilities, radioactive materials released to the atmosphere in gaseous effluents include low concentrations of fission product noble gases (krypton and xenon), halogens (mostly iodines), tritium contained in water vapor and particulate material including both fission products and activated corrosion products.

The primary source of gaseous radioactive wastes is from the degassing of the primary coolant during letdown of the cooling water into the various holding tanks. Additional sources of gaseous waste activity include the auxiliary building exhaust, spent fuel area exhaust, the discharge from the steam jet air ejectors, purging of the reactor containment building and ventilation air exhausted from the turbine building.

All components that can contain potentially radioactive gases are vented to a vent header. The vent gases are subsequently drawn from this vent header by one of three waste gas compressors or a waste gas exhauster. The waste gas compressor discharges through a waste gas separator to one of four waste gas tanks. The waste gas tanks and the waste gas exhauster discharge to the unit vent after passing through a filter bank consisting of a prefilter, an absolute filter, and a charcoal filter.

Radioactive gases may be released inside the reactor containment building when components of the primary system are opened to the building atmosphere for operational reasons or where minor leaks occur in the primary system. Prior to access, the reactor containment atmosphere will be monitored for activity and, when necessary, purged through prefilters, high-efficiency particulate air (HEPA) filters and charcoal adsorbers and released to atmosphere through the unit vent. The purge equipment is sized for a flow rate of 50,000 cfm providing approximately 1.5 air changes per hour in the reactor building. Units 1, 2 & 3 will have a separate vent stack from each reactor building.

The gaseous waste handling and treatment systems for the Oconee Nuclear Station are shown schematically in Fig. A1.0-2.

A-2

DRAINS FROM: CONDENSATE MONITOR TANKS CONDENSATE TEST TANK CORE FLOOD BLEED FUEL TRANSFER CANAL LETDOWN COOLER MAKEUP FILTERS QUENCH TANK REACTOR BLDG Mis NORMAL SUMP HOLDUP TAK REACTOR COOLANT BLEED HOLDUP TANKS REACTOR COOLANT PUMP REACTOR VESSEL A BREACTOR SAMPLE SINKSFED ED SPENT FUEL DEMINERALIZER

PRIM4ARY LOOP.

CONDENSATE REACTOR COOLANT CHEMICAL (reactor

TEST 4BLEED --- AND coolant)

TANKS (2) EVAPORATOR rVOLUME

DEBORATING SYSTEM SYSTEM MAKEUP SYSTEM demiE M]SC WASTE EVAPORATORCO(eirazrs PURIFICATION DEMINERALIZER HIGH ACTIVITY WASTE EVAPORATOR SOC TAAI REACTOR BLDG A EMERGENCY SUMP REACTOR COOLANT BLEED SYSTEM SPENT FUEL SYSTEM WASTE GAS SYSTEM L.EAT GENERATOR

SECONDARY.

IWB RK RA-33LOOP

WASTE EVAPORATOR 1- 1 Nf MO-3

AUX BLDG FLOOR -CHEMICAL MIX TANKS DEMINERALIZER COMPONENT COOLING SYSTEM CONDENSATE TEST TANK CONDENSATE DEMINERALIZER LOW ACTIVITY DEMERALIZER

REACTOR COOLANT BLEED WASTE TANK FLUSH HEADER REACTOR COOLANT BLEED EVAPORATOR SPENT FUEL COOLER IX TUBE SIDE VESSEL

7~CONDENSERS.

SOLIDIFICATIONT

INTAKE

DISCHARGE.

LAKE KEOWEE

CONTAMINATED SHOWERS A CHARCOAL B LAUNDRY LHST FILTER LHST LAVATORIES

LIQUID RADWASTE SYSTEM

NO. I CHEMICAL TREATMENT POND jfLQI AWSESSE

NO. 2 CHEMICAL TREATMENT PONDS ISCHARGE "M OCONEE NUCLEAR STATION

SANITARY SEWER RalELL gu YARD DRAINS

VENT VENT RAD HEIGHT

MONITOR 200 ft

DECAY TANKS

GAS HEADER (From Reactor GAS .100f

Coolant Systems) SEPARATOR CHARCOAL CHARCOAL

COMMON TO ADSORBERS ADSORBERS UNITS 1 & 2 COMPRESSOR

SIMILAR HEPA SYSTEM FOR FILTERS HEPA

UNIT 3 FILTERS WASTE GAS LIOUID EXHAUSTER WASTE

SYSTEM PREFILTER FILTER #:3 1154 PREFLE

50,000 cfm REACTOR BUILDING CONTAINMENT PURGE SYSTEM ROOF VENT

TURBINE BUILDING (12 Fans to Roof 69,000 cfm)

UNITS 1, 2, & 3 CONDENSER ALL SIMILAR FOR EJECTOR

IN CONSTRUCTION EM GAS FROM STEAM GENERATOR TUBE LEAK AIR

EJECTOR

HOTWELL SPENT FUEL - POOL AREA ( UNITS 2 & 3 ONLY )

HEPA FILTER

DECONTAMINATION AND FUME HOODS

GASEOUS WASTE SYSTEM

U POWER OCONEE NUCLEAR STATION

Figure A1.0-2

A.2.0 RELEASE RATE CALCULATION

Generic release rate calculations are presented in Section 1.0; these calculations will be used to calculate release rates from Oconee Nuclear Station.

A.2.1 LIQUID RELEASE RATE CALCULATIONS

There are two potential release points at Oconee, the liquid radwaste effluent line to the Keowee Hydroelectric Unit Tailrace and the oil collection basin effluent line to the Keowee River.

A2.1.1 Liquid Radwaste Effluent Line

To simplify calculations for the liquid radwaste effluent line, it is assumed that no activity above background is present in the oil collection basin effluent. This assumption shall be confirmed by radiation monitoring measurements on the turbine building sumps and by periodic analysis of the composite sample collected at.the oil collection basin. For the liquid radwaste effluent line, one of the following calculations shall be performed for discharge flow, (f, or f2), in gpm, depending on hydro-electric unit operation:

n

f, < 1.7E+4 I Ci i=1 MPC.

1 n

f2 < 2.9E+6 I 1 Ci i=1 MPC.

where:

C. = the concentration of radionuclide, 'i', in undiluted effluent as determined by laboratory analyses, in pCi/ml.

MPC. = the concentration of radionuclide, 'i', from 10CFR20,.Appendix B, 1 Table II, Column 2. If radionuclide, 'i', is a dissolved noble gas,

the MPC., = 2E-4pCi/ml. 1

F = the dilution flow available depending on hydroelectric unit operation,. in gpm:

F, = 1.7E+4 gpm (based on a leakage rate of 38 cfs, the minimum flow available)

F2 = 2.9E+6 gpm (when hydro is operating or 6600 cfs)

a = the recirculation factor at equilibrium is 1.0.

The liquid radwaste effluent discharge is located downstream of Keowee Hydroelectric Station, and therefore, has no retoncentration (recirculation) factor associated with it.- (See Section 1.1).

A-3

A2.1.1 Oil Collection Basin Effluent Line

The oil collection basin effluent is normally considered nonradioactive; that is, it is unlikely the effluent will contain measurable activity above background. It is assumed that no activity is present in the effluent until indicated by radiation monitoring measurements on the turbine building sumps and by periodic analyses of the composite sample collected at the oil collection basin. Radiation monitoring alarm/trip setpoints assure that Specification 3.9.1 is not exceeded.

A2.2 GASEOUS RELEASE.RATE CALCULATIONS

The unit vent is the release point for waste gas decay tanks, containment building purges, the condenser air ejector, and auxiliary building ventilation. The condenser air ejector effluent is normally considered nonradioactive; that is, it is unlikely the effluent will contain measureable activity above background. It is assumed that no activity is present in the effluent until indicated by radiation monitoring measurements and by analyses of periodic samples collected on that line. Radiation monitoring alarm/trip setpoints in conjunction with administrative controls assure that release limits are not exceeded; see section on radiation monitoring setpoints.

The following calculations, when solved for flowrate, are the release rates for noble gases and for radioiodines, particulates and other radionuclides with halflives greater than 8 days; the most conservative of release rates calculated in A2.2.1 and A2.2.2 shall control the release rates for a single release point.

A2.2.1 Release rate limit for noble gases:

Ki [( /)^i] < 500 mrem/yr, and

I (L. + 1.1 M.) [(X/Q)Qi] < 3000 mrem/yr i where the terms are defined below:

A2.2.2 Release rate limit for all radioiodines and radioactive materials in particulate form and radionuclides other than noble gases:

I P. [W Q.] < 1500 mrem/yr i 1

where:

K. = The total body dose factor due to gamma emissions for each identified 1 noble gas radionuclide, in mrem/yr per pCi/m 3 from Table 1.2-1.

L. = The skin dose factor due to beta emissions for each identified noble gas radionuclide, in mrem/yr per pCi/m 3 from Table 1.2-1.

A-I4

M= The air dose factor due to gamma emissions for each identified noble gas radionuclide, in mrad/yr per pCi/m 3 from Table 1.2-1 (unit conversion constant of 1.1 mrem/mrad converts air dose to skin dose).

The dose parameter for radionuclides other than noble gases for the inhalation pathway, in mrem/yr per pCi/m 3 and for the food and ground plane pathways in m2 (mrem/yr) per pCi/sec from Table 1.1-2. The dose factors are based on the critical individual organ and most restrictive age group (child or infant).

Q. = The release rate of radionuclides, i, in gaseous effluent from all release points at the site, in pCi/sec.

(X/Q) = 4.1E-7 sec/m 2 . The highest calculated annual average relative concentration for any area at or beyond the unrestricted area boundary.

W = The highest calculated annual average dispersion parameter for estimating the dose to an individual at the controlling location:

W = 4.5E-9 sec/m 3 , for the inhalation pathway. The location is the unrestricted area in the S sector.

W = 2.4E-9 meter 2, for the food and ground plane pathways. The location is the unrestricted area boundary in the SSW sector (nearest residence, cow, and vegetable garden)

Q = kjC.f + k2 = 4.72E+2C.f i 1 1

where:

C. = the concentration of radionuclide, i, in undiluted gaseous effluent, in pCi/ml.

f = the undiluted effluent flow, in cfm

k, = conversion factor, 2.83E4 ml/ft3

k2 = conversion factor, 6El sec/min

A-5

A3.0 RADIATION MONITOR SETPOINTS

Using the generic calculations presented in Section 2.0, radiation monitoring setpoints are calculated for monitoring as required by the Technical Specifications.

All final effluent radiation monitors for Oconee are off-line. These monitors alarm on low flow; the minimum flow alarm level for the liquid monitors is 3 gallons .per minute and for the gas monitors is 7 standard cubic feet per minute. These monitors measure the activity in the liquid or gas volume exposed to the detector and are independent of flow rate if a minimum flow rate is assured.

Radiation monitoring setpoints calculated in the following sections are expressed in activity concentrations; in reality the monitor readout is in counts per minute. The relationship between concentration and counts per minute is established by station procedure using the following relationship:

r c= 2.22 x 10eV

where:

c = the gross activity, in pCi/ml r = the count rate, in cpm

2.22 x 106 = the disintegration per minute per pCi e = the counting efficiency, cpm/dpm v = the volume of fluid exposed to the detector, in.ml.

A3.1 LIQUID RADIATION MONITORS

A3.1.1 Liquid Radwaste Effluent Line

As described in Section A.2.1.1 of this manual on release rate calculations for the waste liquid effluent, the release is controlled by limiting the flow rate of effluent from the station. Although the release rate is flow rate controlled, the radiation monitor setpoint shall be set to terminate the release if the effluent activity should exceed that determined by laboratory analyses and that used to calculate the release rate.

A3.1.2 Turbine Building Sump Discharge Line

As described in Section A.2.1.2 of this manual on release rate calculations for the turbine building sump effluent, the effluent is normally considered nonradioactive; that is, it is unlikely the effluent will contain measurable activity above background. It is assumed that no activity is present in the effluent until indicated by radiation monitoring and-by routine analysis of the composite sample collected at the oil collection basin. Since the system discharges auto

*matically, the maximum system concentration, which also is the radiation monitor setpoint, is calculated to assure compliance with release limits.

A-6

The monitor setpoint and maximum effluent concentration is:

MPG x F_ c < = 3.0E-6 pCi/ml - f

where:

c = the gross activity in undiluted effluent, in pCi/ml.

f = the flow rate of undiluted effluent which may vary from 0-600 gpm, but is assumed to be 600 gpm.

C=MPC= 1E-7 pCi/ml, the MPC for unidentified mixture.

a = 1 (See Section A2.1.1)

F = the flow may vary from 38 to 6,600 cfs, but is conservatively estimated at 38 cfs (1.7E+4 gpm), the minimum flow available.

A.3.2 GAS MONITORS

The following equation shall be used to calculate noble gas radiation monitor set

points based on. Xe-133:

K(X/Q)Q < 500 (See Section A.2.2.1) i

Q. = 4.72E+2 C.f (See Section A.2.2.2) 1 1

C < 8.80E+3/f

where:

C = the gross activity in undiluted effluent, in pCi/ml

f = the flow from the tank or building and varies for various release sources, in cfm

K = from Table 1.2-1 for Xe-133, 2.94E+2 mrem/yr per pCi/m3

(x/Q)= 4.1E-7 sec/m 3, the highest calculated annual average relative concentration for any area at or beyond the unrestricted area boundary for long term releases.

A.3.2.1 Unit Vent

As stated in Section A.2.2, the unit vent is the release point for waste gas decay tanks, containment building purges, the condenser air ejector, and auxiliary building ventilation. Since all of these releases are through the unit vent, the radiation monitor on the unit vent may be used to assure that station release limits are not exceeded. Depending on the stack flow, the radiation monitor setpoint may be:

C < 8.80E+3/f = 9.30E-2pCi/ml

A-7

where:

f = 45,000 cfm (auxiliary building) + 50,000 cfm (containment purge) = 95,000 cfm

or may be:

C < 8.8E3/f = 2.OE-1 pCi/ml

where:

f = 45,000 cfm (auxiliary building ventilation)

A3.2.2 Interim Radwaste Building Ventilation Exhaust

Ventilation exhaust from the interim radwaste building is not released through the unit vent and is considered a separate release point. This exhaust is normally considered non-radioactive; that is, it is possible but unlikely that the effluent will contain measurable activity above background. Since the exhaust is continuous, a maximum concentration of gases in the exhaust, which also is the radiation monitor setpoint, is calculated to assure compliance with release limits. The monitor setpoint is:

C < 8.8E3/f = 6.3E-1 pCi/ml

* where:

f = 1.47E+04 cfm

A- 8

A4.0 DOSE CALCULATIONS

A4.1 FREQUENCY OF CALCULATIONS

Dose contributions to the maximum individual shall be calculated every 31-days, quarterly, semiannually, and annually (as required by Technical Specifications) using the methodology in the generic information sections. This methodology shall also be used for any special reports. Dose projections shall be performed using simplified dose estimates.

Fuel cycle dose calculations shall be performed annually or as required by special reports. Dose contributions shall be calculated using the methodology in the appropriate generic information.sections.

A4.2 DOSE MODELS FOR MAXIMUM EXPOSED INDIVIDUAL

A4.2.1 Liquid Effluents

For dose contributions from liquid radioactive effluent releases, it is assumed that the maximum exposed individual is an adult who consumes fish caught in the discharge area and drinks water from the nearest downstream water supply.

A4.2.2 Gaseous Effluents

A4.2.2.1 Noble Gases

For dose contributions from exposure to beta and gamma radiations from noble gases, it is assumed that the maximum exposed individual is an adult on the site boundary in each meteorological sector.

A4.2.2.2 Radioiodines, Particulates, and Other Nuclides

For dose contributions from radioiodines, particulates, and other radionuclides; it is assumed that the maximum exposed individual is an infant who breathes the air and consumes milk from the nearest goat or cow in each meteorological sector.

A4.3 SIMPLIFIED DOSE ESTIMATES

A4.3.1 Liquid Effluents

For dose projections, a simplified calculation using the assumptions presented in

Section A4.2.1 and operational source term data is presented below. Dose calculations used to evaluate Appendix I to 10CFR50 compliance indicate that the maximum exposed individual is an adult who consumes fish caught in the discharge area and that 90% of the dose is from cesium-134 and cesium-137.

m DWB= 6.38E+5 I (FP)(T9 ) [CCs-134 + 0.59 CCs-1 37

A- 9

where:

aw 6.38E+5 = (1.10)(1.14E5) D + 21 BF. DF. and all terms are as previously

w1

defined except the following:

C = the average concentration of Cs-134 in undiluted effluent, in Cs-134 during the time period considered.

C = The average concentration of Cs-137 in undiluted effluent, in pCi/ml, during the time period considered.

D = dilution factor from the near field area to the potable water intake, 27.5.

F = 1 (assuming C. is the average concentration after initial dilution with 40 cfs)

= QH = 1100 cfs 27.5 W QR 40 efs

where: a = recirculation factor (Sections 1.1 and 3.1.1) F = near field average dilution factor for C (Section 3.1.1)

= dilution factor from near field area to de potable water intake

Q e(Section 3.1.1) Q= average flow past Keowee Dam (110.0 cfs)

Q= average liquid radwaste release rate (40 cfs)

A.4.3.2 Gaseous Effluents

Meteorological data is provided in Table A4.0-1 and A4.0-2.

A4.3.2.1 Noble Gases

For dose projections, simplified dose estimates using the assumptions in A4.2.2.1 and operational source term data are presented below. These calculations further assume that the annual average dispersion parameter is used and that xenon-133 contributes 45% of the dose.

D = 1.02E-11 [ Q ] y Xe-133

D = 3.08E-11 [ QXe-133

where:

X/Q = 4.1E-7 sec/m 3

A-10

1.02E-11 = constant determined from factors previously presented

3.08E-11 = constant determined from factors previously presented

Q Xe-133= the total xenon-133 activity released in pCi

A4.3.2.2. Radioiodines, Particulates, and Other Nuclides

For dose projections, a simplified dose estimate using the assumptions in A4.2.2.2 and operational source term data is presented below. These calculations further assume that the annual (average dispersion) deposition parameter is used and that 95% of the dose is from iodine-131 concentrated in goats milk. The simplified dose estimates to the thyroid of an infant is:

D = 2.llE4w (Q)

where:

w (D/Q) for food and ground plane pathway, in m-2 from Table A4.0-2 for location of nearest real cow or goat.

(Q) = the release of iodine-131 in gaseous effluents, in pCi.

2.11E4 = (1.05)(3.17E-8) RC [D/Q] with the appropriate substitutions in R. [D/Q] for iodine-131. See Section 3.1.2.

A4.3 FUEL CYCLE CALCULATIONS

These calculations shall be performed using models presented in generic information Section 3.3

A-11

TABLE A4.0-1

OCONEE NUCLEAR STATION (1 of 1)

DISPERSION PARAMETER (X/Q) FOR LONG TERM RELEASES > 500 HR/YR OR > 125 H[R/QTR

Distance to the control location, in miles

Sector 0 0-0.5* 0.5-1.0* 1.0-1.5 1.5-2.0 2.0-2.5 2.5-3.0 3.0-3.5 3.5-4.0 4.0-4.5 4.5-5.0

N 6.5E-8 4.8E-8 4.7E-8 4.7E-8 4.7E-8 6.3E-8 5.9E-8 5.6E-8

NNE 1.1E-7 9.3E-8 8.7E-8 8.9E-8 9.2E-8 9.2E-8 7.2E-8 5.9E-8

NE 7.5E-8 7.2E-8 6.8E-8 5.8E-8 6.1E-8 6.4E-8 6.OE-8 5.7E-8

ENE 6.OE-8 6.4E-8 5.9E-8 6.1E-8 5.7E-8 5.7E-8 5.6E-8 5.6E-8

E 4.1E-8 3.7E-8 5.7E-8 4.8E-8 5.2E-8 4.9E-8 4.7E-8 4.5E-8

ESE 3.OE-8 4.OE-8 6.7E-8 5.8E-8 4.3E-8 5.3E-8 4.9E-8 4.7E-8

SE 2.8E-8 2.8E-8 6.OE-8 5.1E-8 4.1E-8 3.7E-8 3.8E-8 3.8E-8

SSE 2.3E-7 2.OE-7 3.2E-7 2.5E-7 3.7E-7 2.9E-7 2.7E-7 2.5E-7

S 2.6E-7 3.OE-7 2.1E-7 2.1E-7 3.6E-7 4.1E-7 3.7E-7 3.6E-7

SSW 3.2E-7 3.1E-7 2.9E-7 2.7E-7 2.OE-7 1.7E-7 1.7E-7 1.7E-7

SW 7.3E-8 7.1E-8 7.1E-8 5.9E-8 3.9E-8 4.4E-8 4.5E-8 4.5E-8

WSW 5.3E-8 5.2E-8 5.3E-8 4.2E-8 4.8E-8 4.3E-8 4.2E-8 4.2E-8

W 2.!E-8 3.2E-8 3.7E-8 3.7E-8. 3.9E-8 3.9E-8 3.7E-8 3.6E-8

WNW 2.3E-8 2.5E-8 3.5E-8 3.5E-8 3.3E-8 3.2E-8 3.OE-8 2.9E-8

NW 3.2E-8 3.7E-8 3.1E-8 3.3E-8 3.OE-8 3.1E-8 2.9E-8 2.8E-8

NNW 6.8E-8 7.7E-8 8.3E-8 7.7E-8 7.8E-8 6.5E-8 6.3E-8 6.2E-8

* Inside Exclusion Area Boundary (EAB)

TABLE A4.0-2

OCONEE NUCLEAR STATION (1 of 1)

DEPOSITION PARAMETER (D/Q) FOR LONG TERM RELEASES > 500 HR/YR OR > 125 HR/QTR

Distance to the control location, in miles

Sector 0 0-0.5* 0.5-1.0* 1.0-1.5 1.5-2.0 2.0-2.5 2.5-3.0 3.0-3.5 3.5-4.0 4.0-4.5 4.5-5.0

N 2.4E-9 1.4E-9 8.7E-10 6.0E-10 4.7E-10 3.6E-10 2.8E-10 2.3E-10

NNE 4.1E-9 2.2E-9 1.4E-9 9.6E-10 7.4E-10 5.7E-10 4.4E-10 3.6E-10

NE 2.7E-9 1.5E-9 9.7E-10 6.6E-10 5.OE-10 3.9E-10 3.1E-10 2.5E-10

ENE 1.5E-9 8.4E-10 5.4E-10 3.7E-10 2.8E-10 2.2E-10 1.7E-10 1.4E-10

E 1.6E-9 8.7E-10 5.6E-10 3.9E-10 3.OE-10 2.3E-10 2.8E-10 1.5E-10

ESE 1.3E-9 7.0E-10 4.5E-10 3.0E-10 2.3E-10 1.8E-10 1.4E-10 1.1E-10

SE 8.OE-10 4.4E-10 2.9E-10 2.OE-10 1.5E-10 1.2E-10 8.9E-11 7.8E-11

SSE 2.7E-9 1.6E-9 1.1E-9 7.SE-10 6.OE-10 4.6E-10 3.6E-10 3.OE-10

S 4.5E-9 2.6E-9 1.7E-9 1.2E-10 9.0E-10 7.OE-10 5.5E-10 4.5E-10

SSW 4.3E-9 2.5E-9 1.6E-9 1.1E-10 8.5E-10 6.5E-10 5.OE-10 4.2E-10

SW .4E-9 8.4E-10 5.5E-10 3.9E-10 3.0E-10 2.3E-10 2.8E-10 1.5E-10

WSW 1.6E-9 9.1E-10 6.OE-10 4.1E-10 3.2E-10 2.5E-10 1.9E-10 1.6E-10

W 1.4E-9 7.9E-10 5.1E-10 3.6E-10 2.7E-10 2.1E-10 1.6E-10 1.3E-10

WNW 7.7E-10 4.4E-10 2.9E-10 2.0E-10 1.5E-10 1.2E-10 9.2E-11 7.4E-11

NW 1.1E-9 5.9E-10 3.8E-10 2.6E-10 2.OE-10 1.6E-10 1.2E-10 9.9E-11

NNW 1.9E-9 1.OE-9 6.6E-10 4.5E-10 3.5E-10 2.7E-10 2.1E-10 1.7E-10

* Inside EAB

A5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING

The radiological environmental monitoring program shall be conducted in accordance with Technical Specification 3/4.12.

The monitoring program locations and analyses are given in Tables A5.0-1 through A5.0-3 and Figure A5.0-1.

Site specific characteristics make ground water sampling, special low-level 1-131 analyses on drinking water, and food product sampling unnecessary. Ground water recharge is from precipitation and the ground water gradient is toward the effluent discharge area; therefore, contamination of ground water from liquid effluents is highly improbable. Special low level 1-131 analyses in drinking water will not be performed routinely since the expected 1-131 dose from this pathway is less than 1 mrem/year. Food products will not be sampled since lake water irrigation of crops is not practiced in the vicinity.

The laboratory performing the radiological environmental analyses shall participate in an interlaboratory comparison which has been approved by the NRC.

A-12

TABLE A5.0-I (I o f 1)

OCONEE RADIOLOGICAL MONITORING PROGRAM SAMPLING LOCATIONS

(TLD LOCATIONS)

SAMPLING LOCATION DESCRIPTION SAMPLING LOCATION DESCRIPTION

020 SITE BOUNDARY (0.2 MILES N) 040 4-5 MILE RADIUS (4.5 MILES E) 021 SITE BOUNDARY (0.2 MILES NNE) 041 4-5 MILE RADIUS (4.0 MILES ESE) 022 SITE BOUNDARY (0.5 MILES NE) 042 4-5 MILE RADIUS (5.0 MILES SE) 023 SITE BOUNDARY (0.9 MILES ENE) 043 4-5 MILE RADIUS (4.0 MILES SSE) 024 SITE BOUNDARY (0.8 MILES E) 044 4-5 MILE RADIUS (4.0 MILES S) 025 SITE BOUNDARY (0.6 MILES ESE) 045 4-5 MILE RADIUS (5.0 MILES SSW) 026 SITE BOUNDARY (0.3 MILES SE) 046 4-5 MILE RADIUS (4.5 MILES SW) 027 SITE BOUNDARY (0.3 MILES SSE) 047 4-5 MILE RADIUS (4.0 MILES WSW) 028 SITE BOUNDARY (0.5 MILES S) 048 4-5 MILE RADIUS (4.0 MIILES W) 029 SITE BOUNDARY (0.6 MILES SSW) 049 4-5 MILE RADIUS (4.0 MILES WNW) 030 SITE BOUNDARY (0.4 MILES SW) 050 4-5 MILE RADIUS (4.0 MILES NW) 031 SITE BOUNDARY (0.2 MILES WSW) 051 4-5 MILE RADIUS (4.5 MILES NNW) 032 SITE BOUNDARY (0.2 MILES W) 052 SPECIAL INTEREST (12.0 MILES ENE) 033 SITE BOUNDARY (0.2 MILES WNW) 053 SPECIAL INTEREST (11.0 MILES E)

034 SITE BOUNDARY (0.2 MILES NW) 054 SPECIAL INTEREST (9.5 MILES ESE) 035 SITE BOUNDARY (0.1 MILES NNW) 055 SPECIAL INTEREST (9.5 MILES SSE) 036 4-5 MILE RADIUS (4.0 MILES N) 056 SPECIAL INTEREST (8.5 MILES SSW) 037 4-5 MILE RADIUS (4.5 MILES NNE) 057 SPECIAL INTEREST (9.0 MILES SW) 038 4-5 MILE RADIUS (4.0 MILES NE) 058 SPECIAL INTEREST (10.0 MILES WSW) 039 4-5 MILE RAIIUS (4.0 MILES ENE) 059 SPECIAL INTEREST (9.0 MILES NW)

All sampling locations are collected quarterly

iitCtNVE4- HA.IOLOGn;IC t:IONiIFi(HOIING I'I(,4;AH- SAxi-iIN f-L; j4AINS

krlo-:u? SiAMPI'4NG. LOCAjIIINS )

W- - '-4h

Sit - y I-)ilihl

H 8)

0-40 -44 in-nI.

-----------------------.---------------------------------------4 --------

06.4 New Grveiivi I Ie WaL.ei IiiLake 161I. (2.5 m41iles NNE~) - W

06, 1 Si Le IBum1,ida y (0.6 i les SSW-) W

(062 Ldk4c Kewoce/flydro I iake (0.7 mi It! ENE) (CONTROL)N

063j laku Hla iWeII - Hlwy 183 Bridge (0.8 mile E~SE) (000.7) NSA

(o.4 Seeca (6.7 mile11s SW) (004.0) (CONTRO~L) -NSA

(065 (Cemson1 (8.1I miles SSE) (006. 1) 1

(4061 Aj44eoo (19.0) i l4c SSE~) (0112) (COJNTROL FORl MILK ONL.Y) 1 1 SN

4)) I ILawrenice Ililscy Br ige , hlwy 27 (4.2 miiles SSE) (005 .2) SA

04,1 If i gli F a I I s 4X,44a Ly P'a I I (2. 0 111 i es W) (CONII() SA

owt) Ptowe It 10.-si Ience (4.5 mi les WNW) (0)02.0 I SN

041/ C (lIisi Da i-Iy 10. 14 1i Ie SSE) (006.:1) - SHI

0 ? Hwy 1:10 1I 11 - i Io s S) 14

(4/A Tamostive Da- S4 ho'I (9.1)0 mi.s NO) KCJIT41,) W4

07/4 K.oo Keye5 cot 0I. 1 iii 1c NNW) W

TABLE A5.0-3 (1 of 1)

OCONEE RADIOLOGICAL MONITORING PROGRAM ANALYSES

ANALYSES SAMPLE MEDIUM ANALYSIS SCHEDULE GAMMA ISOTOPIC TRITIUM LOW LEVEL 1-131 GROSS BEIA I

1. Air Radioiodine and Particulates Weekly X

2. Direct Radiation Quarterly x

3. Surface Water Monthly X 'Quarterly Composite X

4. Drinking Water Monthly X X Quarterly Composite X

5. Shoreline Sediment Semiannually X

6. Milk Semimonthly X X

7. Fish Semiannually X

8. Broadleaf Vegetation Monthly X

073

94 3 052

0 103

8

8 069 0049 4 039

0 35)X E

2 @048 2500Ft. 05'

Excl Area L ILLA E

6 071 0041 UNION 047

058 -'

A OTHR SAPLIN LOCTION

9 04'

(f 93)054 iles

0579 1

056 5 COO

10 Miles

3 071 4066 ANDERSON 19 MILES SSE

* TLD LOCATIONS

A OTHER SAMPLING LOCATIONS

RADIOLOGICAL MONITORING PROGRAM LOCATIONS

KEOEOCONEE NUCLEAR STATION

Figure A5.0-1 (1 of 2)

RADIOLOGICAL MONITORING PROGRAM LOCATIONS OCONEE NUCLEAR STATION

N Figure A5.0-1 2of 2)

130

KEOWEE ESERVOSR

KE E RESERVOiRk

822

)023 KEOWEE 024

SITOR HY O TER H E

0 021

N 032 024

031 KEOWEE

SITE VER B3OUNDARY

026 FENCE

7

NT 028

029 HWY 183

STLD LOCATIONS

s