Text

Attachment B, Off-Site Dose Calculation Manual, Revision 31
	ML091140599
Person / Time
Site:	Crystal River
Issue date:	08/12/2008
From:	; Progress Energy Co, Progress Energy Florida
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML091140599 (166)
	v • d • e

PROGRESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 ATTACHMENT B OFF-SITE DOSE CALCULATION MANUAL REVISION 31

CRYSTAL RIVER UNIT #3 OFF-SITE DOSE CALCULATION MANUAL APPROVED BY: Lee Hughes (signature on file)

Superintendent Environmental & Chemistry DATE: 08/12/2008 REVISION: 31 APPROVED BY: Interpretation Contact Rudy Pinner (signature on file)

Sr. Science and Lab Services Specialist

INTRODUCTION The Off-site Dose Calculation Manual (ODCM) is provided to support implementation of the Crystal River Unit 3 radiological effluent controls.

The ODCM is divided into two parts. Part I contains the specifications for liquid and gaseous radiological effluents and the radiological environmental monitoring program which were relocated from the Technical Specifications in accordance with the provisions of Generic Letter 89-01 issued by the NRC in January, 1989. Part II of the ODCM contains the calculational methods to be used in determining the dose to members of the public resulting from routine radioactive effluents released from Crystal River Unit 3. Part II also contains the methodology used to determine effluent monitor alarm/trip setpoints which assure that releases of radioactive materials remain within specified concentrations.

The ODCM shall become effective after acceptance by the Plant Nuclear Safety Committee and approval by the Plant General Manager in accordance with Technical Specification Section 5.6.2.3. Changes to the ODCM shall be documented and records of reviews performed shall be retained. This documentation shall contain sufficient information to support the change (including analyses or evaluations), and a determination that the change will maintain the level of radioactive effluent control required by the regulations listed in Technical Specification and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

Changes shall be submitted to the NRC in the form of a complete and legible copy of the entire ODCM as part of, or concurrent with, the Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g. month/year) the change was implemented.

TABLE OF CONTENTS PART I - SPECIFICATIONS Section 1.0 DEFINITIONS 1 1.1 Channel Calibration 1 1.2 Channel Check 1 1.3 Channel Functional Test 1 1.4 Degassing 1 1.5 Frequency 2 1.6 Liquid Radwaste Treatment System *2 1.7 Member of the Public 2 1.8 Mode 2 1.9 Offsite Dose Calculation Manual 3 1.10 Operable - Operability 3 1.11 Site Boundary 3 1.12 Source Check 3 1.13 Unplanned Release 3 1.14 Unrestricted Area 3 1.15 Ventilation Exhaust Treatment System 4 1.16 Waste Gas System 4 1.17 Purge - Purging 4 2.0 SPECIFICATIONS 5 2.1 Radioactive Effluent Monitoring Instrumentation 5 2.2 Radioactive Gaseous Effluent Monitoring Instrumentation 10 2.3 Liquid Radwaste Treatment System 17 2.4 Waste Gas System 18 2.5 Liquid Effluents Concentration 19 2.6 Liquid Effluents - Dose 23 Page i

PART I - SPECIFICATIONS (CON'T)

Section 2.0 SPECIFICATIONS (Con't) 2.7 Gaseous Effluents Dose Rate 24 2.8 Dose Noble Gases 28 2.9 Dose 1-131, 1-133, Tritium, and Radioactive Particulates 29 2.10 Total Dose 30 2.11 Radiological Environmental Monitoring 31 2.12 Land Use Census 38 2.13 Interlaboratory Comparison Program 39 2.14 Special Reports 40 2.15 Meteorological Instrumentation 41 2.16 Waste Gas Decay Tank'- Explosive Gas Monitoring 44 Instrumentation 2.17 Waste Gas Decay Tanks 46 2.18 Waste Gas Decay Tank - Explosive Gas Mixture 47 3.0 SPECIFICATION BASES 3.1 Radioactive Effluent Monitoring Instrumentation Basis 48 3.2 Radioactive Gaseous Effluent Monitoring Instrumentation Basis 48 3.3 Liquid Radwaste Treatment System Basis 48 3.4 Waste Gas System Basis 49 3.5 Liquid Effluents Concentration Basis 49 3.6 Liquid Effluents Dose Basis 50 3.7 Gaseous Effluents Dose Rate Basis 50 3.8 Gaseous Effluents Dose Noble Gases Basis 51 3.9 Gaseous Effluents Dose 1-131, 1-133, Tritium, and Radioactive Particulates Basis 51 Page ii

PART I - SPECIFICATIONS (CON'T)

Section Page 3.0 SPECIFICATION BASES (Con't) 3.10 Total Dose Basis 52 3.11 Radiological Environmental Monitoring Program Basis 53 3.12 Radiological Environmental Monitoring Program Land Use Census Basis 53 3.13 Radiological Environmental Monitoring Interlaboratory Comparison Program Basis 53 3.14 Explosive Gas Mixture 54 3.15 Waste Gas Decay Tanks 54 3.16 Waste Gas Decay Tank - Explosive Gas Monitoring 54 3.17 Meteorological Instrumentation 54 PART II - METHODOLOGIES Section Page 1.0 RADIOACTIVE EFFLUENTS MONITOR SETPOINT SPECIFICATIONS 56 1.1 Effluent Monitor Setpoint Specifications 58 1.2 Nuclide Analyses 61 1.3 Pre-Release Calculations 66 1.4 Setpoint Calculations 72 2.0 RADIOACTIVE EFFLUENTS DOSE REDUCTION SPECIFICATIONS 85 2.1 Waste Reduction Specifications 87 2.2 Dose Projection Methodology 89 2.3 Total Dose Specification 91 3.0 RADIOACTIVE EFFLUENTS SAMPLING SPECIFICATIONS 94 3.1-1 Liquid Releases (Batch) 96 3.1-2 Liquid Releases (Continuous) 96 3.1-3 Gaseous Releases (Waste Gas Decay Tanks) 96 3.1.4 Gaseous Releases (RB & AB) 96 Page iii

PART II - METHODOLOGIES (Con't) 3.1-5 Reactor Bldg. with Personnel and Equipment Hatches Open 97 3.1-6 Reactor Bldg. During Integrated Leak Rate Test 97 4.0 RADIOACTIVE EFFLUENTS DOSE CALCULATION SPECIFICATIONS 98 4.1 Dose Specifications 100 4.2 Nuclide Analyses 103 4.3 Dose Calculations 108 4.4 Dose Factors 112 5.0 ENVIRONMENTAL MONITORING 139 6.0 ADMINISTRATIVE CONTROLS 146 6.1 Origin and Purpose of the ODCM 148 6.2 Changes 148 6.3 Review 148 6.4 Unplanned Releases (moved to section 1) 148 6.5 Radioactive Effluent Release Report 148 6.6 Annual Radiological Environmental Operating Report 150 6.7 Voluntary Reporting as a Result of Groundwater Cont. 152 6.8 Pathway Exemptions from Appendix I Calculations 158 6.9 Power Rx Eff. & Direct Rad. Limits in Unrestricted Areas 160 Page iv

PART I LIST OF TABLES Table Page 2-1 Radioactive Liquid Effluent Monitoring Instrumentation 6 2-2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 8 2-3 Radioactive Gaseous Effluent Monitoring Instrumentation 11 2-4 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 15 2-5 Radioactive Liquid Waste Sampling and Analysis Program 20 2-6 Radioactive Gaseous Waste Sampling and Analysis Program 25 2-7 Operational Radiological Environmental Monitoring Program 32 2-8 Reporting Levels for Radioactivity Concentrations in Environmental Samples 34 2-9 Maximum Values for the Lower Limits of Detection 35 2-10 Meteorological Monitoring Instrumentation 42 2-11 Meteorological Monitoring Instrumentation Surveillance Requirements 43 2-12 Waste Gas System Explosive Gas Monitoring Instrumentation 45 Page v

PART II LIST OF TABLES Table I RADIOACTIVE EFFLUENTS MONITOR SETPOINTS 57 II RADWASTE REDUCTION SYSTEM-DOSE PROJECTIONS 86 III GASEOUS AND LIQUID EFFLUENT REPRESENTATIVE SAMPLING 95 IV CUMULATIVE DOSE CALCULATIONS 99 4.4-1 Dose Factors for Exposure to a Semi-Infinite Cloud of Noble Gases 111 4.4-2 Inhalation Dose Factors - Infant 113 4.4-3 Inhalation Dose Factors - Child 114 4.4-4 Inhalation Dose Factors - Teen 115 4.4-5 Inhalation Dose Factors - Adult 116 4.4-6 Ingestion Dose Factors, Grass-Cow-Milk-Infant 119 4.4-7 Ingestion Dose Factors, Grass-Cow-Milk-Child 120 4.4-8 Ingestion Dose Factors, Grass-Cow-Milk-Teen 121 4.4-9 Ingestion Dose Factors, Grass-Cow-Meat-Adult 122 4.4-10 Ingestion Dose Factors, Grass-Cow-Meat-Child 125 4.4-11 Ingestion Dose Factors, Grass-Cow-Meat-Teen 126 4.4-12 Ingestion Dose Factors, Grass-Cow-Meat-Adult 127 4.4-13 Ingestion Dose Factors, Vegetation-Child 130 4.4-14 Ingestion Dose Factors, Vegetation-Teen 131 4.4-15 Ingestion Dose Factors, Vegetation-Adult 132 Page vi

LIST OF TABLES (Continued)

Table 4.4-16 Dose Factors Ground Plane 134 4.4-17 Liquid Effluent Adult Ingestion Dose Factors 136 5.1-1 Environmental Monitoring Station Location 140 5.1-2 Environmental Monitoring Stati6n's Locations, Groundwater Monitoring Wells 141 5.1-3 Ring TLDs (Inner Ring) 142 5.1-4 Ring TLDs (5 Mile Ring) 143 Page vii

PART I SPECIFICATIONS

1.0 DEFINITIONS 1.1 CHANNEL CALIBRATION Refer to Technical Specifications.

1.2 CHANNEL CHECK Refer to Technical Specifications.

1.3 CHANNEL FUNCTIONAL TEST Refer to Technical Specifications.

1.4 DEGASSING DEGASSING, for purposes of hydrogen and oxygen control, means venting of the make-up or reactor coolant systems to the WASTE GAS SYSTEM.

DEGASSING, for purposes of controlling the inventory of radioactive material, means venting of the pressurizer to the WASTE GAS SYSTEM.

DEGASSING does not include sampling.

OFF-SITE DOSE CALCULATION MANUAL Page 1

1.5 FREQUENCY NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 6 months.

R At least once per 18 months.

S/U Prior to each reactor startup.

P Completed prior to each release.

N.A. Not applicable.

NOTE: Surveillance frequencies are met if the surveillance is performed within 1.25 times the interval specified, as measured from the previous performance or as measured from the time a specified condition of the frequency is met.

This is consistent with the convention of ITS 3.0.2.

1.6 LIQUID RADWASTE TREATMENT SYSTEM The LIQUID RADWASTE TREATMENT SYSTEM shall be any available equipment (e.g., filters, evaporators) capable of reducing the quantity of radioactive material, in liquid effluents, prior to discharge.

1.7 MEMBER OF THE PUBLIC MEMBER OF THE PUBLIC means an individual in a controlled or unrestricted area. However, an individual is not a member of the public during any period in which the individual receives an occupational dose.

1.8 MODE Refer to Technical Specifications.

OFF-SITE DOSE CALCULATION MANUAL Page 2

1.9 OFFSITE DOSE CALCULATION MANUAL (ODCM)

The OFFSITE DOSE CALCULATION MANUAL contains the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program.

The ODCM also contains the Radioactive Effluent Controls and Radiological Environmental Monitoring Program and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports.

1.10 OPERABLE - OPERABILITY Refer to Technical Specifications.

1.11 SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee.

1.12 SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

1.13 UNPLANNED RELEASE An UNPLANNED RELEASE is an unintended discharge of liquid-or airborne radioactivity to the environment. The Auxiliary Building ventilation system is designed to handle leakage from various plant components. Leakage of this sort is not considered unplanned unless the magnitude of the leak is significant (i.e. reportable). Minor equipment failures which cause an increase in plant releases are not unplanned as it is expected that minor failures will occur from time-to-time. Human error which results in a release of radioactivity to the environment is considered unplanned.

Examples:

Releasing the wrong waste tank.

Plant leakage which exceeds reporting limits such as those of 10 CFR 50.72 and 10 CFR 50.73.

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1.14 UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area at or beyond the site boundary, access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the site boundary used for residential quarters or industrial, commercial, institutional, and/or recreational purposes.

1.15 VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION 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 release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

1.16 WASTE GAS SYSTEM A WASTE GAS SYSTEM is any equipment (e.g., tanks, vessels, piping) capable of 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.

1.17 PURGE - PURGING PURGE or PURGING 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 required to purify the confinement.

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2.0 SPECIFICATIONS RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 2.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table 2-1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of specification 2.5 are not exceeded.

APPLICABILITY: As shown on Table 2-1 ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required above, without delay suspend the release of radioactive liquid effluents monitored by the affected channel, or change the setpoint so that it is acceptably conservative, or declare the channel inoperable.

b. With one or more radioactive liquid effluent monitoring instrumentation channels inoperable, take the ACTION shown in Table 2-1. For the instrumentation covered by items 1 and 2 of the table, exert best efforts to return the inoperable instrument(s) to OPERABLE status within 30 days. If the affected instrument(s) cannot be returned to OPERABLE status within 30 days, provide information on the reasons for inoperability and lack of timely corrective action in the next Radioactive Effluent Release Report.

SURVEILLANCE REOUIREMENTS 2.1.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 2-2.

OFF-SITE DOSE CALCULATION MANUAL Page 5

TABLE 2-1 RADIOACTIVE LIQUID EFFLUENT AND PROCESS MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE INSTRUMENT OPERABLE MODES ACTIONS

1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Auxiliary Building Liquid Radwaste Effluent Line (RM-L2) 1 ALL MODES 21

b. Secondary Drain Tank Liquid Effluent Line (RM-L7) 1 ALL MODES 22

2. FLOW RATE MEASUREMENT DEVICES

a. Auxiliary Building Liquid Radwaste Effluent Line 1 ALL MODES 23

b. Secondary Drain Tank Liquid Effluent Line 1 ALL MODES 23

3. PROCESS MONITORS

a. Nuclear Services Closed Cooling 1 24 Water Monitor (RM-L3)

b. Decay Heat Closed Cooling Water Monitors (RM-L5 and RM-L6) 1 24

During system operation OFF-SITE DOSE CALCULATION MANUAL Page 6

TABLE 2-1 (Continued)

TABLE NOTATION ACTION 21 With less than the required number of OPERABLE channels, effluent releases via this pathway may continue, provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Specification 2.5.1, and

b. Two qualified persons independently verify the release rate calculations, and

c. Two qualified persons independently verify the discharge valve lineup.

Otherwise, suspend releases of radioactive materials via this pathway.

ACTION 22 With less than the required number of OPERABLE channels, effluent releases via this pathway may continue, provided that grab samples are collected and analyzed for gross radioactivity, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 23 With less than the required number of OPERABLE channels, effluent releases via this pathway may continue, provided that the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

ACTION 24 With no channels OPERABLE, plant operation may continue provided grab samples are collected and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

OFF-SITE DOSE CALCULATION MANUAL Page 7

TABLE 2-2 RADIOACTIVE LIQUID EFFLUENT AND PROCESS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN CHANNEL SOURCE CHANNEL FUNCTIONAL WHICH SURVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED

1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Auxiliary Building Liquid Radwaste Effluent Line (RM-L2) D* P R (1) Q N.A.

b. Secondary Drain Tank Liquid Effluent Line (RM-L7) D* P R (1) Q N.A.

2. FLOW RATE MEASUREMENT DEVICES

a. Auxiliary Building Liquid Radwaste Effluent Line D (2) N.A. R N.A. N.A.

b. Secondary Drain Tank Liquid D (2) N.A. R N.A. N.A.

Effluent Line

3. PROCESS MONITORS

a. Nuclear Services Closed Cooling D N.A. R Q ALL MODES Water Monitor (RM-L3)

b. Decay Heat Closed Cooling Water Monitors (RM-L5 and RM-L6) D N.A. R Q ALL MODES OFF-SITE DOSE CALCULATION MANUAL Page 8

TABLE 2-2 (Continued)

TABLE NOTATION During periods of release.

(1) CHANNEL CALIBRATION shall be performed using:

a. One or more standards traceable to the National Bureau of Standards, or

b. Standards obtained from suppliers that participate in measurement assurance activities with the National Bureau of Standards, or

c. Standards related to previous calibrations performed using (a) or (b) above.

(2) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. A CHANNEL CHECK shall be performed at least once per day on any day that continuous, periodic or batch releases are made.

OFF-SITE DOSE CALCULATION MANUAL Page 9

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 2.2 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 2-3 shall be OPERABLE with the effluent release isolation alarm/trip setpoints set to ensure that the limits of Specification 2.7 are not exceeded.

APPLICABILITY: As shown in Table 2-3 ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required above, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel where applicable, or change the setpoint so that it is acceptably conservative, or declare the channel inoperable.

b. With one or more radioactive gaseous effluent monitoring instrumentation channels inoperable, take the ACTION shown in Table 2-3. For the instruments covered by items 1, 2, and 3 of the table, exert best efforts to return the inoperable instrument(s) to OPERABLE status within 30 days. If the affected instruments cannot be returned to OPERABLE status within 30 days, provide information on reasons for inoperability and lack of timely corrective action in the next Radioactive Effluent Release Report.

SURVEILLANCE REOUIREMENTS 2.2.1 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and frequencies shown in Table 2-4.

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TABLE 2-3 RADIOACTIVE GASEOUS EFFLUENT AND PROCESS MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE OPERABLE MODES ACTION

1. Waste Gas Decay Tank Monitor (RM-A11)

a. Noble Gas Activity Monitor* 1 ALL MODES 24 1 ALL MODES 26

b. Effluent System Flow Rate Monitor

2. Reactor Building Purge Exhaust Duct Monitor (RM-A1)

a. Noble Gas Activity Monitor

i. Operating Range* 1 27 ii. Mid Range# 1 29 iii. High Range# 1 29 1 25

b. Iodine Sampler 1 25

c. Particulate Sampler

d. Effluent System Flow Rate Monitor 1 26

e. Sampler Flow Rate Monitor 1 26

3. Auxiliary Building and Fuel Handling Area Exhaust Duct Monitor (RM-A2)

a. Noble Gas Activity Monitor

i. Operating Range 1 ALL MODES 28 1 1, 2; 3 & 4 29 ii. Mid Range #

iii. High Range # 1 1, 2, 3 & 4 29 Iodine Sampler 1 ALL MODES 25 b.

c. Particulate Sampler 1 ALL MODES 25 1 ALL MODES 26

d. Effluent System Flow Rate Monitor 1 ALL MODES 26

e. Sampler Flow Rate Monitor 1 1, 2, 3, 4

4. Condenser Vacuum Pump Exhaust - Gaseous 30 Activity Monitor (RM-A12)

Provides control room alarm and automatic termination of release.

During periods of reactor building purge, except during fuel movement. During fuel movement surveillance requirements for RM-A1 (operating range) are specified by ITS 3.3.15. Automatic isolation function is not required if moving fuel that has not been recently irradiated. (See ITS B 3.3.15 and B 3.9.3 for discussion). Fuel movement includes preparation for and demobilization from AI-504, Shutdown Condition 3.

There is no isolation setpoint or release termination function for this monitor. Alarm setpoints are determined by the appropriate system procedures.

OFF-SITE DOSE CALCULATION MANUAL Page 11

TABLE 2-3 (Continued)

TABLE NOTATION ACTION 24 With less than the required number of OPERABLE channels, the contents of the Waste Gas Decay Tank may be released to the environment, provided that prior to initiating a release:

1. The Auxiliary Building & Fuel Handling Area Exhaust Duct Monitor ('RM-A2) is OPERABLE with its setpoints set to ensure that the limits of Specification 2.7 are not exceeded. The setpoint shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL, or

2. a. At least two independent samples of the tank's contents are analyzed in accordance with Table 2-6 and

b. Two qualified persons independently verify the release rate calculations, and

c. Two qualified persons independently verify the discharge valve lineup.

Otherwise, suspend releases of radioactive effluents via this pathway.

ACTION 25 RM-A1 With the affected sampler inoperable, operation of the RB purge may continue for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with no auxiliary sampling, provided that RB airborne levels are steady state or declining. If indicators of RB atmospheric activity, such as RM-A6, RCS leakage, or general area air samples, show an increase in RB activity while the sampler is inoperable, then immediately restore the affected sampler, or implement auxiliary sampling, or shut down the purge.

With the affected sampler inoperable, operation of the RB purge may continue for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided that samples (reference Tables 2-6) are continuously taken (except for filter changes) with auxiliary sampling equipment.

Auxiliary sampling equipment includes general area RB air samples or RMA-15. Other sampling regimes are acceptable provided results are representative of plant effluents.

Note: Coordination of sampling during core alterations or fuel movement is required in order to comply with Technical Specifi cati ons.

OFF-SITE DOSE CALCULATION MANUAL Page 12

TABLE 2-3 (Continued)

TABLE NOTATION ACTION 25 RM-A2 (Continued)

With the affected channel inoperable, effluent releases may continue for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with no auxiliary sampling, provided that AB airborne levels are steady state or declining. If indicators of AB atmospheric activity, such as RM-A3, RM-A4, and RM-A8 show an increase in activity then restore the affected sampler, or implement auxiliary sampling, or shut down the release.

With the affected sampler inoperable, effluent releases may continue for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided that samples (reference Table 2-6) are continuously taken (except for filter changes) with auxiliary sampling equipment.

Auxiliary sampling equipment includes 1) RM-A4 and RM-A8 used together 2) general area AB air samples, or 3) RMA-15. Other sampling regimes are acceptable provided results are representative of plant effluents.

ACTION 26 With the number of OPERABLE channels less than required, effluent releases via this pathway may continue, provided flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 27 With the noble gas monitor (operating range) inoperable, operation of the RB purge may continue for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , provided that RB airborne levels are steady state or declining. If indicators of RB atmospheric activity such as RM-A6, RCS leakage, or general area air samples show an increase in RB activity while the monitor is inoperable, then immediately restore the noble gas monitor or shut down the purge.

Note: Coordination of sampling during core alterations or fuel movement is required in order to comply with Technical Specifications.

ACTION 28 With the number of OPERABLE channels less than required, releases via this pathway may continue, provided grab samples are collected at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and analyzed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and either the requirements of ACTION 24 Part 2 are met or Radiation Monitor RM-A11 is OPERABLE prior to releasing the contents of the Waste Gas Decay Tanks.

Gas grabs may be taken from RM-A4 and RM-A8.

OFF-SITE DOSE CALCULATION MANUAL Page 13

TABLE 2-3 (Continued)

TABLE NOTATION ACTION 29 With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirements,

1) Either restore the inoperable Channel(s) to OPERABLE status within 7 days of the event, or

2) Prepare and submit a Special Report to the Commission pursuant to Specification 2.14 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

NOTE: Action Statement 2.2a not applicable.

ACTION 30 With no channels OPERABLE, plant operation may continue provided grab samples are collected and analyzed for noble gases at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

NOTE: The absence of a conversion factor to relate cpm to gpd primary to secondary leakage does not make RM-A12 inoperable; RM-A12 can still be used as an indicator of changes in noble gas concentrations in the condenser off gas. CP-152, Primary to Secondary Leakage Operating Guideline, specifies the actions to take when a leak rate conversion factor is not available for use with RM-A12.

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TABLE 2-4 RADIOACTIVE GASEOUS EFFLUENT AND PROCESS MONITORING INSTRUMENTATION SURVEILLANCE REOUIREMENTS CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED

1. WASTE GAS DECAY TANK MONITOR (RM-A11)

a. Noble Gas Activity Monitor P R(1) Q ALL MODES

b. Effluent System Flow Rate Monitor N.A. R Q ALL MODES

2. REACTOR BUILDING PURGE EXHAUST DUCT MONITOR (RM-Al)

a. Noble Gas Activity Monitor I. Operating Range P R(1) Q ii. Mid Range M R(1) Q iii. High Range M R(1) Q

b. Iodine Sampler N.A. N.A. N.A.

c. Particulate Sampler N.A. N.A. N.A.

d. Effluent System Flow Rate Monitor N.A. R Q

e. Sampler Flow Rate Monitor N.A. R Q

3. AUXILIARY BUILDING & FUEL HANDLING AREA EXHAUST DUCT MONITOR (RM-A2)

a. Noble Gas Activity Monitor I. Operating Range N.A. R(1) Q ALL MODES ii. Mid Range M R(1) Q 1, 2, 3, 4 iii. High Range M R(1) Q 1, 2, 3, 4

b. Iodine Sampler N.A. N.A. N.A. ALL MODES

c. Particulate Sampler N.A. N.A. N.A. ALL MODES

d. Effluent System Flow Rate Monitor N.A. R Q ALL MODES

e. Sampler Flow Rate Monitor N.A. R Q ALL MODES

4. Condenser Vacuum Pump Exhaust - Gaseous D N.A. R Q 1, 2, 3, 4 Activity Monitor (RM-A12)

OFF-SITE DOSE CALCULATION MANUAL Page 15

TABLE 2-4 (Continued)

During periods of Reactor Building Purge, except during fuel movement.

During fuel movement surveillance requirements for RM-A1 (operating range) are specified by ITS 3.3.15. Automatic isolation function is not required if moving fuel that has not been recently irradiated. (See ITS B 3.3.15 and B 3.9.3 for discussion). Fuel movement includes preparation for and demobilization from AI-504, Shutdown Condition 3.

(1) CHANNEL CALIBRATION shall be performed using:

a. One or more standards traceable to the National Bureau of Standards, or

b. Standards obtained from suppliers that participate in measurement assurance activities with the National Bureau of Standards, or

c. Standards related to previous calibrations using (a) or (b) above.

OFF-SITE DOSE CALCULATION MANUAL Page 16

LIQUID RADWASTE TREATMENT SYSTEM 2.3 The LIQUID RADWASTE TREATMENT SYSTEM shall be used, as required, to reduce radioactive materials in liquid wastes prior to their discharge, when projected monthly doses due to liquid effluents discharged to UNRESTRICTED AREAS would exceed the following values:

a. 0.06 mrem whole body;

b. 0.2 mrem to any organ APPLICABILITY: At all times.

ACTION: a. When radioactive liquid waste, in excess of the above limits, is discharged without prior treatment, prepare and submit to the Commission within 30 days, a Special Report pursuant to Specification 2.14, which includes the following information:

1. Identification of inoperable equipment and the reasons for inoperability.

2. Actions taken to restore the inoperable equipment to OPERABLE status.

3. Actions taken to prevent recurrence.

SURVEILLANCE REOUIREMENTS 2.3.1 Doses due to liquid releases shall be projected at least once per 31 days.

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WASTE GAS SYSTEM 2.4 The WASTE GAS SYSTEM shall be used, as required, to reduce the radioactivity of materials in gaseous waste prior to discharge, when projected monthly air doses due to releases of gaseous effluents from the site to areas at or beyond the SITE BOUNDARY would exceed:

1) 0.2 mrad gamma;

2) 0.4 mrad beta; and The VENTILATION EXHAUST TREATMENT SYSTEM shall be used, as required, to reduce the quantity of radioactive materials in gaseous waste prior to discharge, when projected monthly air doses due to release of gaseous effluents from the site to areas at or beyond the SITE BOUNDARY would exceed:

1) 0.3 mrem to any organ APPLICABILITY: At all times.

ACTION:

a. When the WASTE GAS SYSTEM and/or VENTILATION EXHAUST TREATMENT SYSTEM are not used and gaseous waste in excess of the above limits is discharged without prior treatment, prepare and submit to the Commission, within 30 days a Special Report, pursuant to Specification 2.14, which includes:

1) Identification of the inoperable equipment and the reason(s) for inoperability.

2) Actions taken to restore the inoperable equipment to OPERABLE status.

3) Actions taken to prevent recurrence.

SURVEILLANCE REQUIREMENTS 2.4.1 Doses due to gaseous releases from the site shall be projected at least once per 31 days.

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LIQUID EFFLUENTS CONCENTRATION 2.5 The concentration of radioactive material released to UNRESTRICTED AREAS shall be less than or equal to 10 times the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For Xe-133, the concentration shall be < 1 x 10-3 microcuries/ml. For all other dissolved or entrained noble gases, the concentration shall be less than or equal to 2x10- 4 microcuries/ml total activity.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of radioactive materials released to UNRESTRICTED AREAS exceeding the above limits, without delay restore the concentration of radioactive materials being released to UNRESTRICTED AREAS to within the above limits. If the concentration of radioactive materials being released in excess of the above limit is related to a plant-operating characteristic, appropriate corrective measures (e.g., power reduction, plant shutdown) shall be taken to restore the concentration of radioactive materials being released to UNRESTRICTED AREAS to within the above limits.

SURVEILLANCE REQUIREMENTS 2.5.1 Radioactive liquid wastes shall be sampled and analyzed in accordance with the sampling and analysis program of Table 2-5.

2.5.2 The results of the radioactivity analyses shall be used to assure the concentrations of radioactive material released from the site are maintained within the limits of Specification 2.5.

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TABLE 2-5 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit of Liquid Release Sampling Analysis Type of Activity Detection Type Frequency Frequency Analysis (LLD)

(pCi/ml)a A. Batch Waste P P Release Each Batch Each Batch Principal Gamma 5x10-7 Tanksd Emittersf

1. Evaporator 1-131 lx10-6 Condensate Storage Tanks (2)

2. Laundry & P M Dissolved and 1x1O-5 Shower Sump One Batch/M Entrained Gases Tanks (2) (Gamma Emitters

3. Secondary P M H-3 1x10-5 Drain Tank Each Batch Compositeb Gross Alpha lx10-7 P Q Sr-89, Sr-90 5x10-8 Each Batch Compositeb Fe-55 lx10-6 B. Continuous W Principal Gamma Releasese Continuousc Compositec Emittersf 5x10-7

1. Condensate System 1-131 lx10-6 M M Dissolved and 1x10-5 Grab Sample Entrained Gases (Gamma Emitters)

M H-3 1x10-5 Continuousc Compositec Gross Alpha lx10-7 Q Sr-89, SR-90 5x10-8 Continuousc Compositec Fe-55 lx10-6 OFF-SITE DOSE CALCULATION MANUAL Page 20

TABLE 2-5 (Continued)

TABLE NOTATION

a. 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 = 4.66sb/(2.22x106EVYe-XAt)

Where:

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

sb 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 disintegration),

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

2.22x106 is the number of disintegrations 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).

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

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

OFF-SITE DOSE CALCULATION MANUAL Page 21

TABLE 2-5 (Continued)

TABLE NOTATION

b. 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.

c. 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.

d. 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 to assure representative sampling.

e. A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g., from a volume or system that has an input flow during the continuous release.

f. 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. Nuclides which are below the LLD for the analyses shall be reported as "less than" the nuclide's LLD, and shall not be reported as being present at the LLD level for that nuclide.

The "less than" values shall not be used in the required dose calculations.

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LIOUID EFFLUENTS - DOSE 2.6 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to UNRESTRICTED AREAS shall be limited as follows:

a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and less than or equal to 5 mrem to any organ.

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

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission, within 30 days, a Special Report pursuant toSpecification 2.14, which includes:

1. Identification of the cause for exceeding the limit(s);

2. Corrective action taken to reduce the release of radioactive materials in liquid effluents during the remainder of the current calendar quarter an during the remainder of the current calendar year so that the dose or dose commitment to a MEMBER OF THE PUBLIC from this source is less than or equal to 3 mrem total body and less than or equal to 10 mrem to any organ during the calendar year.

SURVEILLANCE REOUIREMENTS 2.6.1 DOSE CALCULATIONS. Cumulative dose contributions from liquid effluents shall be determined at least once per 31 days.

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GASEOUS EFFLUENTS - DOSE RATE 2.7 The dose rate at or beyond the SITE. BOUNDARY, due to radioactive materials released in gaseous effluents, shall be limited as follows:

a. Noble gases: less than or equal to 500 mrem/year total body and less than or equal to 3000 mrem/year to the skin.

b. 1-131, 1-133, Tritium, and radioactive particulates with half-lives of greater than 8 days: less than or equal to 1500 mrem/year to any organ.

APPLICABILITY: At all times ACTION:

a. With dose rate (s) exceeding the above limits, without delay decrease the dose rate to within the above limit(s). If the dose rate at or beyond the SITE BOUNDARY due to radioactive materials in gaseous effluents in excess of the above limits is related to a plant-operating characteristic, appropriate corrective measures (e.g., power reduction, plant shutdown) shall be taken to decrease the dose rate to within the above limits.

SURVEILLANCE REQUIREMENTS 2.7.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits.

2.7.2 The dose rate due to radioactive materials specified above, other than noble gases, in gaseous effluents shall be determined to be within the above limits by obtaining representative samples and performing analyses in accordance with Table 2-6.

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IAbLt- Z-b RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit of Minimum Detection Sampling Analysis (LLD)

Gaseous Release Type Frequency Frequency Type of Activity Analysis (pCi/ml)a A. Waste Gas Decay Tank P P Each Tank Grab Each Tank Principal Gamma Emittersf lx10-4 Sample B. Reactor Building P P Principal Gamma Emittersb,f lx10-4 Purge Exhaust Duct Each Purgec Each Purge Monitor Grab (RM-A1) Sample H-3 lx10-6 C. Auxiliary Building Mc M Principal Gamma Emittersb,f lx10-4 and Fuel Handling Grab Area Exhaust Duct Sample Monitor (RM-A2)

H-3 lx10-6 D. All Release Types as Continuouse Wd 1-131 Listed in A, B, C Charcoal 1X10- 1 2 above Sample Continuouse Wd Principal Gamma Emittersf 1x10- 11 Particulate (1-131, Others)

Sample Continuouse M Gross Alpha Composite 1x10- 11 Particulate Sample Continuouse Q Sr-89, Sr-90 Composite 1x10- 11 Parti cul ate Sample Continuouse Noble Gas Noble Gases Monitor Gross Beta & Gamma lx10-6 OFF-SITE DOSE CALCULATION MANUAL Page 25

TABLE 2-6 (Continued)

TABLE NOTATION

a. 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 = 4.66sb/(2.22x1OGEVYe-&At)

Where:

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

sb 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 disintegration),

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

2.22x10 6 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),

k 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).

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

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

OFF-SITE DOSE CALCULATION MANUAL Page 26

TABLE 2-6 (Continued)

TABLE NOTATION

b. Analyses shall be performed when there is a sustained increase in the noble gas monitor count rate. As sustained increase is one in which the count rate stays above the monitor warning sepoint for at least one hour.

Sampling shall be done within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of warning alarm actuation.

If the associated noble gas monitor (RM-A1 or RM-A2) is out of service during a release, then analyses shall be performed between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following shutdown, startup, or a change in power level exceeding 15%

rated thermal power within one hour.

c. Tritium grab samples shall be taken between 12 and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after flooding the refueling canal and at least once per 7 days thereafter while the refueling canal is flooded.

d. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from sampler).

Sampling and analyses shall be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startup or change in power level exceeding 15% of RATED THERMAL POWER within one hour, unless the Iodine Monitoring Channels in Radiation Monitors RM-A1 and RM-A2 show that the Radionuclide concentration in the Auxiliary Building and Fuel Handling Area or the Reactor Building Purge Exhaust Ducts will lead to a release which is less than the 10 CFR 20, Appendix B, Table II, Column I limits, at or beyond the SITE BOUNDARY.

e. 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 the Specifications 2.7, 2.8, and 2.9.

f. 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. Nuclides which are below the LLD for the analyses shall be reported as "less than" the nuclide's LLD and shall not be reported as being present at the LLD level for that nuclide. The "less than" values shall not be used in the required dose calculations.

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DOSE-NOBLE GASES 2.8 The air dose at or beyond the SITE BOUNDARY, due to radioactive noble gases released in gaseous effluents shall be limited to:

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

b. During any calendar year: less than or equal to 10 mrad gamma and less than or equal to 20 mrad beta radiation.

APPLICABILITY: At all times.

ACTION:

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission, within 30 days, a Special Report, pursuant to Specification 2.14, which includes:

1) Identification of the cause for exceeding the limit(s).

2) Corrective action taken to reduce the release of radioactive noble gases in gases effluents during the remainder of the current calendar quarter and during the remainder of the current calendar year so that the average dose during the calendar year is less than or equal to 10 mrad gamma and 20 mrad beta radiation.

SURVEILLANCE REQUIREMENTS 2.8.1 DOSE CALCULATIONS: Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined at least once per 31 days.

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DOSE 131, 1-133, TRITIUM, AND RADIOACTIVE PARTICULATES 2.9 The dose to a MEMBER OF THE PUBLIC from 1-131, 1-133, Tritium, and radioactive particulates with half-lives greater than 8 days in gaseous effluents released from the site to areas at or beyond the SITE BOUNDARY shall be limited as follows:

a. During any calendar quarter: less than or equal to 7.5 mrem to any organ, and

b. During any calendar year: less than or equal to 15 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of 1-131, 1-133, Tritium, and radioactive particulates with greater than 8 day half-lives, in gaseous effluents, exceeding any of the above limits, prepare and submit to the Commission, within 30 days, a Special Report, pursuant to Specification 2.14, which includes:

1) Identification of the cause for exceeding the limits(s);

2) Corrective action to reduce those releases during the remainder of the current calendar quarter and the remainder of the current calendar year so that the average dose to any organ is less than or equal to 15 mrem.

SURVEILLANCE REQUIREMENTS 2.9.1 DOSE CALCULATIONS: Cumulative dose calculations for the current calendar quarter and current calendar year shall be determined at least once per 31 days.

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TOTAL DOSE 2.10 The calendar year dose or dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity and radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrems).

APPLICABILITY: At all times.

ACTION:

a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specification 2.6a, 2.6b, 2.8a, 2.8b, 2.9a, or 2.9b, calculations should be made, which include direct radiation contributions from the reactor, to determine whether the above limits of Specification 2.10 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Specification 2.14, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR Part 20.2203, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose(s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

SURVEILLANCE REOUIREMENTS 2.10.1 DOSE CALCULATIONS - Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 2.6.1, 2.8.1, and 2.9.1.

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RADIOLOGICAL ENVIRONMENTAL MONITORING 2.11 Theiradiological environmental monitoring program shall be conducted as specified in Table 2-7.

I APPLICABILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 2-7, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

b. With the level of radioactivity, resulting from plant effluents, in an environmental sampling medium exceeding the reporting levels of Table 2-8 when averaged over any calendar quarter, prepare and submit to the Commission, within 30 days of obtaining analytical results from the affected sampling period, a Special Report pursuant to Specification 2.14, which identifies the cause(s) for exceeding the limit(s) and defines corrective actions to be taken to reduce radioactive effluents so that the potential annual dose to a MEMBER OF THE PUBLIC is less than the calendar year limits of Specifications 2.7, 2.8, and 2.9. When more than one of the radionuclides in Table 2-8 are detected in the radionuclides in Table 2-8 are detected in the sampling medium, this report shall be submitted if:

concentration (1) concentration (2) limit level (1) + limit level (2) + ... > 1.0 When radionuclides other than those in Table 2-8 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is greater than or equal to the calendar year limits of Specifications 2.7, 2.8, and 2.9. 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.

c. With milk or fresh leafy vegetation samples unavailable from one or more of the sample locations required by Table 2-7, identify the cause of the unavailability of samples and identify locations for .obtaining replacement samples in the next Annual Radiological Environmental Operating Report. The locations from which samples were unavailable may then be deleted from those required by Table 2-7, provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations.

SURVEILLANCE REOUIREMENTS 2.11.1 The radiological environmental monitoring samples shall be collected pursuant to Table 2-7 from the locations given in the table and Figures 5.1, 5.2, and 5.3 and shall be analyzed pursuant to the requirements of Tables 2-7 and 2-9.

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TABLE 2-7 OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Samples Sampling/ Type/Frequency of and/or Sample and Locations Collection Frequency Analysis 1.AIRBORNE One sample each: Continuous sampler/ Radioiodine canister:

Radioiodine and C07, C18, C40, C41, Weekly collection particulates C46 and Control a) 1-131 analysis Location C47 weekly Particulate sampler:

a) Gross P at > 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s/following weekly filter change.

b) Composite gamma special analysis (by location)/

quarterly. (Gamma Spectral Analysis shall also be performed on individual samples if gross beta activity of any sample is greater than 1.0 pCi/m 3 and which is also greater than ten times the control sample activity.

2.DIRECT RADIATION 1) Si te Boundary: Continuous Gamma exposure C60, C61, C62, placement/Quarterly rate/quarterly C63, C64, C65, collection C66, C67, C68, C69, C41, C70, C27, C71, C72, C73

2) Five Miles:

C18, C03, C04, C74, C75, C76, C08, C77, C09, C78, C14G, C01, C79

3) Control Location: C47 OFF-SITE DOSE CALCULATION MANUAL Page 32

TABLE 2-7 (Continued)

OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Samples Sampling/ Type/Frequency of and/or Sample and Locations Collection Frequency Analysis

3. WATERBORNE One sample each: Grab sample/Monthly Gamma spectral Seawater C14H, C14G Control analysis/monthly Location C13 Tritium analysis on each sample or on a quarterly composite of monthly samples Ground water One sample: Grab Gamma spectral and C40 (Control sampl e/semi annual Tritium analysis/each Location) sample Site Ground One sample each: Grab Gamma spectral and Water CR3-2, CR3-4, CR3-5, sample/quarterly Tritium analysis/each CR3-6S, CR3-6D, CR3-7 sample CR3-8, CR3-9, CR3-10 One sample each:

Drinking water C07, C10, C18 (All Grab Gamma spectral and Control Locations) sample/quarterly Tritium analysis/each sample One sample each:

Shoreline C14H, C14M, C14G Semiannual sample Gamma spectral Sediment Control Location C09 analysis/each sample

4. INGESTION Fish & One sample each: Quarterly: Gamma spectral Invertebrates C29, Control Oysters and analysis on edible Location C30 carnivorous fish portions/each sample Food Products One sample each: Monthly (when Gamma spectral and C48a*, C48b*, available): Sample 1-131 analysis/each Control Location C47 compressed of three sample (3) types of broad leaf vegetation from each location One sample: C19 Annual during Gamma spectral harvest: Citrus analysis/each sample One sample: C04 Annual during Gamma spectral harvest: Watermelon analysis/each sample

Stations C48a and C48b are located near the site boundary for gaseous effluents in the two sectors which yield the highest historical annual average D/Q values.

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TABLE 2-8 RFPflRTTNC I FVFI S FOR RADTAACTTVTTY CONCFNTRATTONS TN FNVTRONNIFNTAI SAMPI FS Water Airborne Particulate Fish Milk Food Products Analysis (pCi/1) or Gases (pCi/m 3) (pCi/Kg, wet) (pCi/1) (pCi/Kg, wet)

H-3 20,000(a)

Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zn-65 300 20,000 Zr-Nb-95(b) 400 1-131 2(c) 0.9 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140(b) 200 300 (a)For drinking water samples. This is 40 CFR Part 141 value. If no drinking water pathway exists, a value of 30,000 pCi/l may be used.

(b)An equilibrium mixture of the parent and daughter isotope which contains the reporting value of the parent isotope.

(c) For drinking water samples only.

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TABLE 2-9 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD) a.d Airborne Particulate Water or Gases Fish Milk Food Products Sediment Analysis (pCi/1) (pCi/mr) (pCi/Kg, wet) (pCi/1) (pCi/Kg, wet) (pCi/Kg, dry) gross beta 0.01 3H 2000b 54Mn 15 130 59 Fe 30 260 5 8 Co 15 130 6 0 Co 15 130 6 5 Zn 30 260 95 Zr-Nb 1 5c 1311 if 0 .0 7 g 1 60 134Cs 15 0 . 0 5e 130 15 60 150 137Cs 18 0 .0 6 e 150 18 80 180 14 0 Ba-La 15c 15c OFF-SITE DOSE CALCULATION MANUAL Page 35

TABLE 2-9 (Continued)

TABLE NOTATION

a. 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 = 4.66sb/(2.22EVYe-XAt)

Where:

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

Sbis 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 disintegration),

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

2.22 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield (when applicable),

k is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between environmental collection, or end of the sample collection period, and time of counting.

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

The LLD is defined as an a priori (before the fact) limit representing the capability of the measurement system and not as an a posteriori (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLD's will be achieved under routine conditions.

Occasionally, background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLD's unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report.

OFF-SITE DOSE CALCULATION MANUAL Page 36

TABLE 2-9 (Continued)

TABLE NOTATION

b. LLD for drinking water. If no drinking water pathway exists, a value of 3000 pCi/l may be used.

c. The specified LLD is for an equilibrium mixture of parent and daughter nuclides which contain 15 pCi/l of the parent nuclide.

d. Other peaks which are measurable and identifiable, together with the radionuclides in Table 2.9, shall be identified and reported.

e. Cs-134, and Cs-137 LLD's apply only to the quarterly composite- gamma spectral analysis, not to analyses of single particulate filters.

f. LLD for drinking water. If no drinking water pathway exists, the LLD of gamma isotopic analysis may be used.

g. LLD for 1-131 applies to a single weekly filter.

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LAND USE CENSUS 2.12 A land use census shall be conducted and shall identify the location of the nearest milk animal, the nearest residence and the nearest garden* of greater than 500 square feet producing fresh leafy vegetables in each of the land based meteorological sectors within a distance of five miles.

APPLICABILITY: At all times.

ACTION:

a. With a land use census identifying a location(s) that yields a calculated dose or dose commitment greater than the values currently being calculated by Specification 2.9.1, identify the new location in the next Annual Radiological Environmental Operating Report.

b. With a land use census identifying a location(s) which yields a calculated dose or dose commitment (via the same exposure pathway) which is at least 20% greater than at a location from which samples are currently being obtained in accordance with Specification 2.11, this location shall be added to the radiological environmental monitoring program within 30 days.

The new sampling location shall replace the present sampling location, which has the lower calculated dose or dose commitment (via the same exposure pathway), after June 30 following this land use census. Identification of the new location and revisions of the appropriate figures shall be submitted with the next Radioactive Effluent Release Report.

Broad leaf vegetation sampling may be performed at the site boundary in the direction sector with the highest D/Q in lieu of the garden census.

SURVEILLANCE REOUIREMENTS 2.12.1 The land use census shall be conducted at least once per 12 months during the growing season by a door-to-door survey, aerial survey, or by consulting local agriculture authorities, using that information which will provide adequate results.

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INTERLABORATORY COMPARISON PROGRAM 2.13 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the Commission. A summary of the results obtained from this program shall be included in the Annual Radiological Environmental Operating Report.

APPLICABILITY: At all times.

ACTION:

a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report.

SURVEILLANCE REOUIREMENTS 2.13.1 No surveillance requirements other than those required by the Interlaboratory Comparison Program.

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ADMINISTRATIVE CONTROLS 2.14 SPECIAL REPORTS Special reports shall be submitted to the Nuclear Regulatory Commission within the time period specified for each report. These reports shall be submitted covering the activities identified below.

A separate Licensee Event Report, when required by 10 CFR 50.73 (a), need not be submitted if the Special Report meets the requirements of 10 CFR 50.73 (b) in addition to the requirements of the applicable referenced Specification.

A. Dose due to radioactive materials in liquid effluents in excess of specified limits, Specification 2.6.

B. Dose due to noble gas in gaseous effluents in excess of specified limits, Specification 2.8.

C. Total calculated dose due to release of radioactive effluents exceeding twice the limits of Specifications 2.6a, 2.6b, 2.8a, 2.8b, 2.9a, or 2.9b (required by Specification 2.10).

D. Dose due to 1-131, 1-133, Tritium, and radioactive particulates with greater than eight day half-lives, in gaseous effluents in excess of specified limits, Specification 2.9.

E. Failure to process liquid radwaste, in excess of limits, prior to release, Specification 2.3.

F. Failure to process gaseous radwaste, in excess of limits, prior to release, Specification 2.4.

G. Measured levels of radioactivity in environmental sampling medium in excess of the reporting levels of Table 2-8, when averaged over any quarterly sampling period, Specification 2.11.

H. Inoperable Mid or High Range Noble Gas Effluent Monitoring Instrumentation, Specification 2.2.

I. Meteorological monitoring channel inoperable for more than 7 days, Specification 2.15.

J. WGDT explosive gas monitoring instrumentation inoperable for more than 30 days, Specification 2.16.

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METEOROLOGICAL INSTRUMENTATION 2.15 The meteorological monitoring instrumentation channels shown in Table 2-10 shall be OPERABLE.

APPLICABILITY: At all times.

ACTION:

a. With one or more required meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 2.14 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel(s) to OPERABLE status.

SURVEILLANCE REQUIREMENTS 2.15.1 Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 2-11.

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TABLE 2-10 METEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM INSTRUMENT LOCATION OPERABLE

1. WIND SPEED Nominal Elev. 33' 1

2. WIND DIRECTION Nominal Elev. 33' 1

3. STABILITY CLASS (DELTA-T OR SIGMA-THETA)

Nominal Elev.

1

33' for sigma-theta. 175'-33' for delta-T.

NOTE: Back up meteorological tower instruments may be used to meet the minimum operability requirement of ODCM specification 2.15.

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TABLE 2-11 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION

1. WIND SPEED Nominal Elev. 33' D SA

2. WIND DIRECTION Nominal Elev. 33' D SA

3. STABILITY CLASS (DELTA-T OR SIGMA-THETA)

Nominal Elev. D SA

33' for sigma-theta. 175' 33' for delta - T OFF-SITE DOSE CALCULATION MANUAL Page 43

WASTE GAS DECAY TANK - EXPLOSIVE GAS MONITORING INSTRUMENTATION 2.16 The Waste Gas Decay Tanks shall have one hydrogen and one oxygen monitoring channel OPERABLE.

APPLICABILITY: During WASTE GAS SYSTEM operation.

ACTION:

a. With the number of OPERABLE channels less than required above, operation of this system may continue, provided grab samples are collected and analyzed:

(1) at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during DEGASSING operations (2) at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during other operations

b. If the affected channel(s) cannot be returned to OPERABLE status within 30 days, submit a special report to the Commission pursuant to Specification 2.14 within 30 days describing the reasons for inoperability and a schedule for corrective action.

SURVEILLANCE REOUIREMENTS 2.16.1 The Waste Gas Decay Tank explosive gas monitoring instrumentation shall be demonstrated operable by performing the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION at the frequencies shown in Table 2-12.

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TABLE 2-12 WASTE GAS SYSTEM EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECK CALIBRATION TEST

1. Hydrogen Monitors D M

2. Oxygen Monitors D M The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

Hydrogen Monitors

a. 1 volume percent hydrogen, balance nitrogen.

b. 4 volume percent hydrogen, balance nitrogen.

Oxygen Monitors

c. 1 volume percent oxygen, balance nitrogen.

d. 4 volume percent oxygen, balance nitrogen.

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WASTE GAS DECAY TANKS 2.17 The quantity of radioactivity contained in each Waste Gas Decay Tank shall be limited to less than or equal to 39000 curies (considered as Xe 133).

APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactivity in any Waste Gas Decay Tank exceeding the above limit, immediately suspend all additions of

radioactive material to that tank, and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within its limit.

SURVEILLANCE REOUIREMENTS 2.17.1 The quantity of radioactive material contained in each Waste Gas Decay Tank shall be determined* to be within the limit at least once per 7 days whenever radioactive materials are being added to the tank, and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during primary coolant system DEGASSING operations.

Determining that each waste gas decay is in compliance with the limit may be done by a method other than direct sampling of the tank provided it is in accordance with an approved procedure.

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WASTE GAS DECAY TANK - EXPLOSIVE GAS MIXTURE 2.18 The concentration of oxygen in any Waste Gas Decay Tank shall be limited to less than or equal to 2% by volume whenever the concentration of hydrogen in that Waste Gas Decay Tank is greater than or equal to 4% by volume.

NOTE: Whenever the concentration of hydrogen in the bulk of the waste gas header, including the cover gas of the reactor coolant bleed tanks, is greater than 4%, or the oxygen concentration is greater than 2%, then consideration should be given to purging the waste gas header with nitrogen.

APPLICABILITY: At all times.

ACTION:

Whenever the concentration of hydrogen in any Waste Gas Decay Tank is greater than or equal to 4% by volume, and:

a. The concentration of oxygen in that Waste Gas Decay Tank is greater than 2% by volume, but less than 4% by volume, without delay begin to reduce the oxygen concentration to within its limit.

b. The concentration of oxygen in that Waste Gas Decay Tank is greater than or equal to 4% by volume, immediately suspend additions of waste gas to that Waste Gas Decay Tank and without delay begin to reduce the oxygen concentration to within its limit.

SURVEILLANCE REOUIREMENTS 2.18.1 The concentrations of hydrogen and oxygen in the in-service Waste Gas Decay Tank shall be continuously monitored with the hydrogen and oxygen monitors required OPERABLE by Specification 2.16 or by sampling in accordance with Specification 2.16 action a.

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3.0 SPECIFICATION BASES 3.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION BASIS The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm/trip setpoints for these instruments shall be calculated in accordance with the procedures in the OFFSITE DOSE CALCULATION MANUAL (ODCM) to ensure that the alarm/trip will occur prior to exceeding the 10 times 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.

3.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION BASIS 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 of gaseous effluents. The alarm/trip setpoints for these instruments are calculated in accordance with the procedures in the OFFSITE DOSE CALCULATION MANUAL (ODCM) to ensure that the alarm/trip will occur prior to exceeding a Site Boundary dose rate of 500 mrem/year to the total body. 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.

3.3 LIQUID RADWASTE TREATMENT SYSTEM BASIS The requirement that these systems be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as reasonably achievable" (ALARA).

This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.

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3.4 WASTE GAS SYSTEM BASIS The requirement that 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 reasonable achievable" (ALARA). This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

3.5 LIQUID EFFLUENTS CONCENTRATION BASIS This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than 10 times the effluent concentration limits (ECLs) specified in 10 CFR Part 20. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within the Section II.A design objectives of Appendix I, 10 CFR 50, to a MEMBER OF THE PUBIC. The concentration limit for Xe-133 was determined by calculating that amount of the isotope, which if present in water, would give a dose rate of 500 mrem/yr at the surface. Typically, over 90% of the noble gas released in liquid effluents at CR-3 is Xe-133. The concentration limit for all other dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

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3.6 LIQUID EFFLUENTS DOSE BASIS This specification is provided to implement the requirements of Sections II.A. Ill-A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statement provides the required operating flexibility and at that same time implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable" (ALARA). The dose calculations in the OFFSITE DOSE CALCULATION MANUAL (ODCM)-.implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the OFFSITE DOSE CALCULATIONAL MANUAL (ODCM) for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

3.7 GASEOUS EFFLUENTS DOSE RATE BASIS This specification is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents will be within the annual dose limits of 10 CFR Part 20, §§ 20.1 - 20.602.

The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, §§ 20.1 - 20.602, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC, either within or outside the SITE BOUNDARY to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)(1)). For a MEMBER OF THE PUBLIC who may at time be within the SITE BOUNDARY, the occupancy of the MEMBER OF THE PUBLIC will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrem/year to the total body or to less than or equal to 3000 mrem/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrem/year.

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3.8 GASEOUS EFFLUENTS DOSE NOBLE GASES BASIS This Specification is provided to implement the requirements of Sections II.B, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as reasonably achievable" (ALARA). The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculations established for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculational of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. The equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

3.9 GASEOUS EFFLUENTS DOSE 1-131, 1-133, TRITIUM, AND RADIOACTIVE PARTICULATE BASIS This specification is provided to implement the requirements of Sections II.C, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluent will be kept "as low as is reasonably achievable" (ALARA). The calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The methods for calculating the dose due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,"

Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for 1-131, 1-133, OFF-SITE DOSE CALCULATION MANUAL Page 51

Tritium, and radioactive particulates with half-life less than eight days are dependent on the existing radionuclide pathways to man, in areas at and beyond the SITE BOUNDARY. The pathways which were examined in the development of these calculations were: 1)

Individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leaf 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.

3.10 TOTAL DOSE BASIS This specification is provided to meet the dose limitations of 40 CFR Part 190 that have now been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR part 190.11 and 10 CFR Part 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed.

The variance only relates to the limits of 40 CFR Part 190 and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 2.5 thru 2.9. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

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3.11 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM BASIS The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of MEMBER OF THE PUBLIC 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 the environmental exposure pathways. Program changes may be initiated based on operational experience.

The LLD's required by Table 2-9 are considered optimum for routine environmental measurements in industrial laboratories. The LLD's for drinking water meet the requirements of 40 CFR 141.

3.12 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LAND USE CENSUS BASIS This specification is provided to ensure that changes in the use of areas at or beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census. Adequate information gained from door-to-door or aerial surveys or through consultation with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50.

Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumption were used: 1) that 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/square meter.

3.13 RADIOLOGICAL ENVIRONMENTAL MONITORING INTERLABORATORY COMPARISON PROGRAM BASIS The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

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BASES 3.14 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the Waste Gas Decay Tanks is maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3.15 WASTE GAS DECAY TANKS Restricting the quantity of radioactivity contained in each waste gas decay tank provides assurance that in the event of a simultaneous uncontrolled release of all the tanks' contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Branch Technical Position ETSB 11-5.

3.16 WASTE GAS DECAY TANK - EXPLOSIVE GAS MONITORING INSTRUMENTATION The OPERABILITY of the Waste Gas Decay Tank explosive gas monitoring instrumentation or the sampling and analysis program required by this specification provides for the monitoring (and controlling) of potentiall.y explosive gas mixtures in the Waste Gas Decay Tanks.

3.17 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures the sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the needs for initiating protective measures to protect the health and safety of the public.

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PART II METHODOLOGIES OFF-SITE DOSE CALCULATION MANUAL Page 55

SECTION 1.0 RADIOACTIVE EFFLUENT MONITOR SETPOINTS SPECIFICATIONS OFF-SITE DOSE CALCULATION MANUAL Page 56

TABLE I - RADIOACTIVE EFFLUENT MONITOR SETPOINTS RELEASE TYPE SETPOINT NUCLIDE ANAL. SETPOINT SETPOINT SPECIFICATION CALCULATION ADJUSTMENT MONITOR BATCH CONT TYPE ** FREQ.

RM-A1 x 1.1-1 1.2-1 P 1.3-1 1.4-1 (Noble Gas)

RM-A1 x 1.1-I 1.2-1 W 1.3-1 1.4-2 (Noble Gas)

RM-A2 x i.i-i 1.2-2 W/P* 1.3-1 1.4-3 (Noble Gas)

RM-A11 x 1.1-1 1.2-3 P 1.3-1 1.4-4 (Noble Gas)

RM-L2 x 1.1-2 1.2-4 P 1.3-2 1.4-5 (Gamma)

RM-L7 x x 1.1-2 1.2-5 W 1.3-2 1.4-6 & 1.4-7 (Gamma)

RM-A1 &

RM-A2 N/A N/A 1.1-3 NA NA 1.3-3 NA (Iodine Channels)

This monitor is used in conjunction with (or instead of) RM-A11 to monitor the release of the waste gas decay tanks. Nuclide analysis and setpoint calculation must be performed for this monitor prior to waste gas decay tank release. At all other times, it is a continuous source monitor and the setpoint is determined weekly.

"For composited samples the results from the most recently completed analysis are used.

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GASEOUS EFFLUENT MONITORS SETPOINT SPECIFICATION 1.1-1 (Monitors RM-A1, RM-A2 and RM-A11)

The dose rate at or beyond the SITE BOUNDARY, due to radioactive materials released in gaseous effluents, is limited as follows:

Noble Gases - 500 mrem/year (total body) 3000 mrem/year (skin) 1-131, 1-133, Tritium and Radioactive 1500 mrem/year (any organ via particulates with the inhalation pathway.)

greater than 8 day half-lives The radioactive gaseous effluent monitors (RM-A1, RM-A2 and RM-A11) shall have their alarm/trip setpoints set to ensure that the above total body, noble gas dose rate limit is not exceeded.

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LIQUID EFFLUENT MONITORS SETPOINT SPECIFICATION 1.1-2 (Monitors RM-L2, RM-L7)

The concentration of radioactive materials in liquid effluents, released to UNRESTRICTED AREAS, is limited to 10 times the effluent concentrations specified by 10 CFR 20, for radionuclides other than noble gases. For-all dissolved or entrained noble gases, except Xe-133, the concentration limit is 2E-4 PCi/ml. For Xe-133 the concentration limit is 1E-3 pCi/ml.

The radioactive liquid effluent monitors (RM-L2 and RM-L7) shall have their alarm/trip setpoints set to ensure that the above gamma emitting concentration limits are not exceeded.

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GASEOUS EFFLUENT MONITORS SETPOINT SPECIFICATION 1.1-3 (Iodine Channels in RM-Al and RM-A2)

Sampling and analyses of the Reactor Building Purge Exhaust, and the Auxiliary Building and Fuel Handling Area Exhaust for radioiodine and other gamma emitters, shall be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days when the Radioiodine concentration in the Auxiliary Building and Fuel Handling Area or the Reactor Building Purge Exhaust Ducts will lead to a release which is greater than or equal to the 10 CFR 20, Appendix B, Table II, Column I limits, at or beyond the SITE BOUNDARY.

The iodine monitoring channels in radiation monitors RM-A1 and RM-A2 shall have their alarm setpoints set to alarm when the above radioiodine concentration limits are exceeded.

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NUCLIDE ANALYSIS 1.2-1 REACTOR BUILDING PURGE EXHAUST NUCLIDE SAMPLE SOURCE LLD(b) (uCi/cc)

A. Principal Gamma Emitters (a)

Mn-54 Fe-59 Co-58 Pre-release grab sample for Batch Co-60 Type release. Weekly Particulate Zn-65 Filter Analysis for continuous(c) 1X10-4 /1x10-1 1 Mo-99 type release.

Cs-134 Cs-137 Ce-141 Ce-144 Kr-87 Pre-release grab sample for Batch Kr-88 type release. Noble Gas monitor Xe-133 during batch and continuous releases lx10-4 Xe-133m Grab sample within 2-6 hr. following Xe-135 startup, shutdown or > 15% RTP Xe-138 change in 1 hr.

B. Iodine 131 Pre-release grab sample for Batch NA/1 x 10-12 type release. Weekly charcoal filter and once per 24 hr for 7 days following startup shutdown or > 15% RTP change in 1 hr if 1-131 concentration at site boundary > 10 CFR 20 limit.

6 C. Tritium Pre-release Grab Sample and within 1x10-12-24 hr following flooding of refueling canal and once per 7 days while canal is flooded.

D. Gross Alpha Monthly Particulate Filter Composite lx10-11 11 E. Sr-89 Quarterly Particulate Filter Composite 1x10 11 F. Sr-90 Quarterly Particulate Filter Composite 1x10 (a) Other identified Gamma Emitters not listed in this table shall be included in dose and setpoint calculations.

(b) The first value refers to the LLD for pre-release grab sample; the second value refers to the LLD for weekly Particulate Filter Analysis.

(c) Reactor Building Purge is considered continuous after a minimum of one Reactor Building volume has been released on a continuous basis (i.e., first volume is a batch type).

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NUCLIDE ANALYSIS 1.2-2 AUXILIARY BUILDING AND FUEL HANDLING AREA EXHAUST NUCLIDE SAMPLE SOURCE LLD(b) (uCi/ml)

A. Principal Gamma Emitters (a)

Mn-54 Fe-59 Co-58 Weekly Particulate Filter Analysis.

Co-60 Zn-65 lxlO-4 /lxlO-11 Mo-99 Cs-134 Cs-137 Ce-141 Ce-144 Kr-87 Monthly Grab Sample and Kr-88 Continuous Noble Gas monitor.

Xe-133 Grab sample within 2-6 hr following 1x10-4 Xe-133m startup, shutdown or > 15% RTP Xe-135 change in 1 hr.

Xe-138 12 B. Iodine 131 Weekly Charcoal Filter analysis and once 1x10-per 24 hr for 7 days following startup shutdown or > 15% RTP change in 1 hr if 1-131 concentration at site boundary > 10 CFR 20 limit.

6 C. Tritium Monthly Grab Sample and within 1x10 12-24 hr following flooding of refueling canal and once per 7 days while canal is flooded.

D. Gross Alpha Monthly Particulate Filter Composite lx10-11 E. Sr-89 Quarterly Particulate Filter Composite lx10-11 F. Sr-90 Quarterly Particulate Filter Composite lx10-11 (a) Other identified Gamma Emitters not listed in this table shall be included in dose and setpoint calculations.

(b) The first value refers to the LLD for pre-release grab sample; the second value refers to the LLD for weekly Particulate Filter Analysis.

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NUCLIDE ANALYSIS 1.2-3 WASTE GAS DECAY TANKS NUCLIDE SAMPLE SOURCE LLD(b) (uCi/ml)

A. Principal Gamma Emitters (a)

Mn-54 Fe-59 Co-58 Co-60 Zn-65 Pre-release Grab sample and Weekly 1x10- 4 /1x10-11 Mo-99 Particulate Filter Sample from RM-A2 Cs-134 Cs-137 Ce-141 Ce-144 Kr-87 Kr-88 Xe-133 Pre-release Grab sample. 1X10-4 Xe-133m Xe-135 Xe-138 12 B. Iodine 131 Weekly Charcoal Filter from RM-A2. 1X10-(a) Other identified Gamma Emitters not listed in this table shall be included in dose and setpoint calculations.

(b) The first value refers to the LLD for pre-release grab sample; the second value refers to the LLD for weekly Particulate Filter Analysis.

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NUCLIDE ANALYSIS 1.2-4 EVAPORATOR CONDENSATE STORAGE TANKS, LAUNDRY AND SHOWER SUMP TANKS, SECONDARY DRAIN TANK NUCLIDE SAMPLE SOURCE LLD(uCi/ml)

A. Principal Gamma Emitters (a)

Mn-54 Fe-59 Co-58 Co-60 Zn-65 Pre-release Grab Sample 5X10-7 Mo-99 Cs-134 Cs-137 Ce-141 Ce-144 6

B. Iodine 131 Pre-Release Grab Sample 1X10-C. Dissolved and Entrained Noble 5

Gases Monthly Grab Sample 1x10-5 D. Tritium Monthly Composite 1x10-7 E. Gross Alpha Monthly Composite lx10-8 F. Sr-89 Quarterly Composite 5x10-G. Sr-90 Quarterly Composite 5x10-8 H. Fe-55 Quarterly Composite lx10-(a) Other identified Gamma Emitters not listed in this table shall be included in dose and setpoint calculations.

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NUCLIDE ANALYSIS 1.2-5 SECONDARY DRAIN TANK AND/OR PLANT CONDENSATE NUCLIDE SAMPLE SOURCE LLD(uCi/ml)

A. Principal Gamma Emitters (a)

Mn-54 Fe-59 Co-58 Co-60 Zn-65 Weekly Composite 5x10-7 Mo-99 Cs-134 Cs-137 Ce-141 Ce-144 6

B. Iodine 131 Weekly Composite 1x10-C. Dissolved and Entrained Noble 5

Gases Monthly Grab Sample 1x10-5 D. Tritium Monthly Composite 1x10-7 E. Gross Alpha Monthly Composite lx10-F. Sr-89 Quarterly Composite 5x10_8 8

G. Sr-90 Quarterly Composite 5x10-6 H. Fe-55 Quarterly Composite 1x10-(a) Other identified Gamma Emitters not listed in this table shall be included in dose and setpoint calculations.

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PRE-RELEASE CALCULATION 1.3-1 GASEOUS RADWASTE RELEASE I. INTRODUCTION Prior to initiating a release of gaseous radwaste, it must be determined that the concentration of radionuclides to be released, and the flow rates at which they are released will not cause the dose rate limitations of Specification 1.1-1 to be exceeded.

II. INFORMATION REQUIRED Results of appropriate Nuclide Analysis from Section 1.2 III. CALCULATIONS Noble Gas Gamma Emissions Dose Rate (Total Body) = Z (X/Q)KiQi mrem/yr. (1.1)

Noble Gas Beta Emissions Dose Rate (Skin) = Y (X/Q)Qi(Li + 1.1Mi) mrem/yr. (1.2)

Iodine 131, Iodine 133, Tritium, Radioactive Particulates Dose Rate (I,T,P) = Z (X/Q)PiQi mrem/yr. (1.3) where:

Ki = The total body dose factor due to gamma emissions for each identified noble gas radionuclide, in mrem/yr per pCi/m 3 . (See Table 4.4-1).

Li = The skin dose factor due to beta emissions for each 3

identified noble gas radionuclide,in mrem/yr per pCi/m .

(See Table 4.4-1).

Mi = The air dose factor due to gamma emissions for each 3

identified noble gas radionuclide, in mrad/yr per pCi/m (unit conversion constant of 1.1 mrem/mrad converts air dose to skin dose). (See Table 4.4-1).

Pi - The dose parameter for radionuclides other than noble 3

gases for the inhalation pathway, in mrem/yr per pCi/m .

(See Table 4.4-3).

Qi = The release rate of radionuclides, i, in gaseous effluent from individual release sources, in pCi/sec (per unit, unless otherwise specified). Qi = Effluent stream nuclide concentration x flow rate.

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Flow Rates (Variable - based on setpoint needs, nominal or maximum values listed below.)

1) Reactor Building Purge Exhaust Duct = 50,000 cfm =

2.4 x 10' cc/sec

2) Auxiliary Building and Fuel Handling Area Exhaust Duct=

156,000 cfm = 7.4 x 107 cc/sec

3) Waste Gas Decay Tank Release Line = 50 cfm max =

2.4 x 104 cc/sec (X/Q) = 2.5 x 10-6 sec/m3 . For all vent releases. The highest calculated annual average relative concentration for any area at or beyond the unrestricted area boundary.

In order for a gaseous release to be within the limits of specification 1.1-1, the Projected Dose Rate Ratio (PDRR) must not exceed 1. The PDRR for each limit is calculated as follows:

PDRRTB = PDRTB / 500 (1.4)

PDRRSK = PDR / 3000 (1.5)

PDRRORG = PDRORG / 1500 (1.6)

PDRTB = Projected Dose Rate to the TOTAL BODY due to noble gas emissions.

PDRSK = Projected Dose Rate to the SKIN due to noble gas emissions.

PDRoRG = Projected Dose Rate to any organ due to inhalation of iodine, tritium and particulates with half-lives greater than 8 days.

500 = The allowable total body dose rate due to noble gas gamma emissions in mrem/yr.

3000 = The allowable skin dose rate due to noble gas beta emissions in mrem/yr.

1500 = The allowable organ dose rate in mrem/y.r.

Equations 1.1, 1.2, and 1.3 are solved for each release type and release point currently releasing or awaiting release. If relationships 1.4, 1.5, and 1.6 are satisfied, the release can be made under the assumed flow rates. If one or more of the relationships 1.4, 1.5 and 1.6 are not satisfied, action must be taken to reduce the the radionuclide release rate prior to initiating a release (or to reduce the radionuclide release rate already in progress).

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The following actions are available to reduce the release rates at the three release points.

1) Waste Gas Decay Tanks a) Release Valve may be throttled b) Tank contents may be diluted c) Release may be delayed for longer decay time.

2) Reactor Building Purge Exhaust Duct a) Dilution flow may be opened to reduce purge rate while maintaining the same flow rate.

3) Auxiliary Building and Fuel Handling Area Exhaust a) Reduce inlet air supply to areas in Auxiliary Building to reduce radioactivity source rate to vent.

b) Identify and isolate the sources of radioactive releases into the Auxiliary Building.

Effluent Monitor LLD Determination The relationship given below may be used to calculate a monitor LLD.

LLD = (4.66-5B)/Slope B = Average monitor background count rate in cpm.

Slope = Slope of monitor calibration curve in cpm/pCi/ml.

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PRE-RELEASE CALCULATION 1.3-2 LIQUID RADWASTE RELEASE I. INTRODUCTION Prior to initiating a release of liquid radwaste, it must be determined that the concentration of radionuclides to be released and the flow rates at which they will be released will not lead to a release concentration greater than the limits of specification 1.1-2 at the point of discharge.

II. INFORMATION REQUIRED Results of appropriate Nuclide Analysis from Section 1.2 III. CALCULATIONS Discharge Concentration= 0.1i1 C + CG CXE-133 Ca CT Cs +/-_CFe I FD+E+

L ECL21 2E-5 1E-4 ECLa ECLT ECLs ECLFeIL E ]

where:

C= The concentration of isotope i, in the gamma spectrum excluding dissolved or entrained noble gases.

CG = Total dissolved or entrained noble gas concentration, excluding Xe-133.

CXE- 133 = XE-133 concentration.

CT Tritium Concentration from most recent analysis.

Ca GCross alpha concentration from most recent analysis.

Cs - Sr-89, 90 concentration from most recent analysis.

Cl-e - Fe-55 concentration from most recent analysis.

E = Effluent Stream Flow Rate D = Dilution Stream Flow Rate (Nuclear Services and Decay Heat seawater flow only)

ECL = IOCFR20 Appendix B, effluent concentration limit.

If Discharge Concentration is less than or equal to 1, the discharge may be initiated. If Discharge Concentration is greater than 1, then release parameters must be changed to assure that Discharge Concentratlon is not greater than 1.

Changes include reducing tank concentration by decay or dilution, reducing the waste stream release rate, or increasing dilution water flow rate.

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PRE-RELEASE CALCULATION 1.3-3 GASEOUS EFFLUENT IODINE MONITORS I. INTRODUCTION In order to determine the setpoints for these monitors, the following assumptions are used.

A. The release rate through the Auxiliary Building and Fuel Handling Area exhaust duct is 7.4 x 10' cc/sec. (156,000 cfm).

B. The release rate through the Reactor Building Purge Exhaust Duct is 2.4 x 107 cc/sec (50,000 cfm).

C. A limitless supply of uniformly concentrated 1-131 is available to supply the Exhaust Ducts.

D. The iodine filter has been installed for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and operating at a constant flow rate of 472 cc/sec (1 cfm). Therefore, total flow through the filter has been 1.36 x 10' cc.

II. CALCULATIONS The limiting concentration of Iodine in the vent which would result in a concentration equal to the 10 CFR 20 limit at the site boundary is calculated as follows:

Cv=Ci/[(X/Q)FK]

where:

Cv = The Concentration of Radioiodine in the vent in pCi/cc.

C1 = The 10 CFR 20 effluent concentration limit for Iodine 131, 2 x 10"0 pCi/cc.

F = The duct flow rate: 2.4 x 10' cc/sec for the Reactor Building Purge Exhaust Duct and 7.4 x 10 7 cc/sec for the Auxiliary Building and Fuel Handling Area Exhaust Duct.

K = Unit conversion constant, 1 x 10- m3/cc X/Q = The highest calculated annual average concentration for any area at or beyond the unrestricted area boundary, 2.5 x 106 sec/m3 .

Solving eqn. 1.7 for the Reactor Building Purge exhaust vent yields:

CV(RB) = 3.33 x 106 pCi/cc Solving eqn. 1.7 for the Auxiliary Building & Fuel Handling Area Exhaust vent yields:

CV(AB) = 1.1 x 106 pCi/cc OFF-SITE DOSE CALCULATION MANUAL Page 70

In order to determine the total quantity of Iodine 131 collected on the filter, the values of C above are multiplied by the volume assumed to have passed through the filter Q, = fkCý (1.8) where:

Q = The total quantity of Iodine 131 collected on the filter, in pCi.

C = The concentration of Iodine 131 in the vent in pCi/cc.

f = The assumed total volume of vent atmosphere that has passed through the filter, 1.36 x 10' cc (1 CFM for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ).

k = The Iodine removal efficiency of the filters: 90%

Solving eqn. 1.8 for the Reactor Building vent yields:

QI(RB) = 40.8 pCi Solving eqn. 1.8 for the Auxiliary Building and Fuel Handling Area vent yields:

Q,(AB) = 13.5 pCi These values are converted to counts per minute for the Iodine monitoring channels through use of the appropriate calibration curve.

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Setpoint Calculation 1.4-1 Reactor Building Purge Exhaust Duct Monitor (RM-Al)

(Batch Type Releases)

INTRODUCTION Following completion of the analyses required by Section 1.2-1 and determination of release rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded.

METHODOLOGY Reactor Building atmosphere is circulated through radiation monitor RM-A6 (containment atmosphere noble gas monitor) and the count rate is observed.

The observed count rate is correlated to a corresponding count rate for RM-A1 (Reactor Building purge exhaust duct monitor), and factors are applied to account for background radiation, and the pressure difference between the detector chambers and exhaust vent. The obtained value establishes the maximum allowable setpoint. The alarm/trip setpoint is adjusted to this or a more conservative value prior to initiating the release. If the concentration of radionuclides to be released is less than the effluent monitor LLD "Net CPM" is obtained from the calibration curve by determining the CPM which corresponds to 2.5E-2 pCi/ml, and PDRR is set equal to 1.

CALCULATION RM_ Al Setpoint (CPM) Net CPM x VF / / CPM)A6 +

29.9-VX (/10CC PDRR 29.9-V6 (/ICi/cc/CPM)AI Bkg where:

Net CPM The observed RM-A6 count rate, in cpm, less background, or obtained from the calibration curve.

VF - The vent fraction; that portion of the total plant gaseous release associated with this vent and discharge type. Value can be set to a number between 0 and 1.

The summation of the vent fractions of RM-A1 and RM-A2 cannot exceed 1.

PDRR The noble gas gamma emission- Projected Dose Rate Ratio calculated in accordance with Section 1.3. This ratio is the actual projected dose rate divided by the allowable dose rate referenced in Section 1.3-1, relationship 1.4.

V6 The actual gauge vacuum reading at RM-A6 at the time of sampling.

VI The actual or average gauge vacuum reading at RM-A1 during normal operation.

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(/i / cc / CPM)A6 pCi/cc per cpm for RM-A6. This is based on an actual sample or derived from the calibration curve.

(JI / cc / CPM)AI - Ci/cc per cpm for RM-A1. This is based on an actual sample or derived from the calibration curve.

Bkg RM-A1 background count rate in cpm.

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Setpoint Calculation 1.4-1A Reactor Building Purge Exhaust Duct Monitor (RM-AI)

(Special Release For Functional Testing of the Reactor Building Purge System)

INTRODUCTION Following completion of the analyses required by Section 1.2-1 and determination of release rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded.

METHODOLOGY Auxiliary Building and Fuel Handling Area atmosphere is continuously passed through radiation monitor RM-A2 and the count rate is observed. The observed count rate is correlated to a corresponding count rate for RM-A1, and factors are applied to account for background radiation and the pressure difference between the detector chambers and exhaust vent. The obtained value establishes the maximum allowable setpoint. The alarm/trip setpoint is adjusted to this or a more conservative value prior to initiating the release.

If the concentration of radionuclides to be released is less than the effluent monitor LLD "Net CPM" is obtained from the calibration curve by determining the CPM which corresponds to 2.5E-2 pCi/ml, and PDRR is set equal to 1.

CALCULATION RM-AI Setpoint (CPM) Net CPM x VF 29.9-V I (/ICi/cc/ CPM)A2]

PDRR 29.9-V2 (/pCi/CC/CPM)Aj where:

NetCPM = The observed RM-A2 count rate, in cpm, less background, or obtained from the calibration curve.

VF - The vent fraction; that portion of the total plant gaseous release associated with this vent and discharge type. VF can be set to a value from 0 and 1. The sum of RM-A1 and RM-A2 vent fractions can not exceed 1.

PDRR The noble gas gamma emission Projected Dose Rate Ratio calculated in accordance with Section 1.3. This ratio is the actual projected dose rate divided by the allowable dose rate referenced in Section 1.3-1, relationship 1.4.

V2 The actual gauge vacuum reading at RM-A2 at the time of sampling.

Vl The actual or average gauge vacuum reading at RM-A1 during normal operation.

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(]Ki / CC / CPM)A2 - Ci/cc per cpm for RM-A2. This is based on an actual sample or derived from the calibration curve.

( //cc/ CPM)AI pCi/cc per cpm for RM-A1. This is based on an actual sample or derived from the calibration curve.

Bkg RM-A1 background count rate in cpm.

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Setpoint Calculation 1.4-1B

.Reactor Building Purge Exhaust Duct Monitor (RM-A1)

(Special Release Following ILRT of Reactor Building)

INTRODUCTION Following completion of the analyses required by Section 1.2-1 and determination of release rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded.

METHODOLOGY Net CPM is obtained from the calibration curve by determining the CPM which corresponds to 2.5E-2 pCi/ml. This value is combined with the monitor background, vent fraction and projected dose rate ratio (PDRR) to arrive at the monitor setpoint. The obtained value establishes the maximum allowable setpoint. The alarm/trip setpoint is adjusted to this or a more conservative value prior to initiating the release.

Shortly, after beginning the purge, new RM-A1 alarm/trip setpoints are determined using the methodology of Setpoint Calculation 1.4-2.

CALCULATION RM -AI Setpoint (CPM) =Net CPM x VF+Bk NetCPM = A value derived from RM-A1 calibration curve.

VF - The vent fraction; that portion of the total plant gaseous release associated with this vent and discharge type. VF can be set to a value from 0 and 1. The sum of RM-A1 and RM-A2 vent fractions can not exceed 1.

PDRR = 1 Bkg = RM-A1 background count rate in cpm.

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Setpoint Calculation 1.4-2 Reactor Building Purge Exhaust Duct Monitor (RM-AI)

(Continuous Type Releases)

INTRODUCTION Following completion of the analyses required by Section 1.2-1 and determination of release rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded.

METHODOLOGY Reactor Building atmosphere is passing through radiation monitor RM-A1 during a continuous type release. Factors are applied to the observed count rate to account for background radiation and vent fraction. The obtained value establishes the maximum allowable setpoint. The alarm/trip setpoint is adjusted to this or a more conservative value weekly during continuous releases. If the concentration of radionuclides to be released is less than the effluent monitor LLD "Net CPM" is obtained from the calibration curve by determining the CPM which corresponds to 2.5E-2 pCi/ml, and PDRR is set equal to 1.

CALCULATION RM -AlI Setpoint (CPM) Ne CP x +/-F Bkg L PDRR where:

NetCPM = The observed RM-A1 count rate, in cpm, less background, or obtained from the calibration curve.

VF - The vent fraction; that portion of the total plant gaseous release associated with this vent and discharge type. Value can be set to a number between 0 and 1.

The summation of the vent fractions of RM-A1 and RM-A2 cannot exceed 1.

PDRR = The noble gas gamma emission Projected Dose Rate Ratio calculated in accordance with Section 1.3. This ratio is the actual projected dose rate divided by the allowable dose rate referenced in Section 1.3-1, relationship 1.4.

Bkg = RM-Al background count rate in cpm.

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Setpoint Calculation 1.4-3 Auxiliary Building & Fuel Handling Area Exhaust Monitor (RM-A2)

(Continuous Type Releases)

INTRODUCTION Following completion of the analyses required by Section 1.2-2 and determination of release rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded.

METHODOLOGY Auxiliary Building and Fuel Handling Area atmosphere is continuously passing through radiation monitor RM-A2. Factors are applied to the observed count rate to account for background radiation and vent fraction. The obtained value establishes the maximum allowable setpoint. The alarm/trip setpoint is adjusted to this or a more conservative value weekly during continuous releases. If the concentration of radionuclides to be released is less than the effluent monitor LLD "Net CPM" is obtained from the calibration curve by determining the CPM which corresponds to 8E-3 pCi/ml, and PDRR is set equal to 1.

CALCULATION RM - A2 Setpoint (CPM) =Net CPM x VF + Bkg L PDRR where:

NetCPM The observed RM-A2 count rate, in cpm, less background, or obtained from the calibration curve.

VF - The vent fraction; that portion of the total plant gaseous release associated with this vent and discharge type. Value can be set to a number between 0 and 1.

The summation of the vent fractions of RM-A1 and RM-A2 cannot exceed 1.

PDRR = The noble gas gamma emission Projected Dose Rate Ratio calculated in accordance with Section 1.3. This ratio is the actual projected dose rate divided by the allowable dose rate referenced in Section 1.3-1, relationship 1.4.

Bkg RM-A2 background count rate in cpm.

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Setpoint Calculation 1.4-4 Waste Gas Decay Tank Monitor (RM-AlI)

(Batch Type Releases)

INTRODUCTION Following completion of the analyses required by Section 1.2-3 and determination of release rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded.

METHODOLOGY Prior to initiating a Waste Gas Decay Tank release, its contents are drawn through radiation monitor RM-A11 and returned to the waste gas header.

Factors are applied to the observed count rate to account for background radiation and vent fraction. The obtained value establishes the maximum allowable setpoint. The alarm/trip setpoint is adjusted to this or a more conservative value weekly during continuous releases. If the concentration of radionuclides to be released is less than the effluent monitor LLD "Net CPM" is obtained from the calibration curve by determining the CPM which corresponds to 20 pCi/ml, and PDRR is set equal to 1.

CALCULATION RM-AII Setpoint(CPM) LNetCPMxVFx24.7

=' PD-RR- x P* j +Bkg a

where:

NetCPM The observed RM-A11 count rate, in cpm, less background, or obtained from the calibration curve.

VF The vent fraction; that portion of the total plant gaseous release associated with this vent and discharge type. Value is equal to 0.5.

PDRR The noble gas gamma emission Projected Dose Rate Ratio calculated in accordance with Section 1.3. This ratio is the actual projected dose rate divided by the allowable dose rate referenced in Section 1.3-1, relationship 1.4.

24.7 The maximum pressure (psia) which RM-A11 detector chamber should be subjected to. This corresponds to a flow of 15 CFM from the release line to the vent.

P Pressure (psia) in RM-A11 at time of obtaining net CPM.

Bkg RM-A11 background count rate in cpm.

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Setpoint Calculation 1.4-5 Plant Discharge Line Monitor (RM-L2)

(Batch Type Releases)

INTRODUCTION Following completion of the analyses required by Section 1.2-4 and determina-tion of release rates and concentration limits in accordance with Section 1.3-2, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded.

METHODOLOGY Evaporator Condensate Storage Tank or Laundry and ShowerSump Tank contents are circulated through radiation monitor RM-L2 and returned to the auxiliary building sump to obtain the actual count rate at RM-L2 for the concentration contained in the tank for release. The observed count rate is adjusted for release flow, background and statistical counting variations, particular to this release flow path. The resulting value is used as the alarm/trip setpoint and RM-L2 is adjusted to this or a more conservative value prior to initiating the release. If the concentration of radionuclides to be released is less than the effluent monitor LLD use setpoint calculation 1.4-8.

CALCULATION RM - L2 Setpoint (CPM) = FNet CPM xAF x (E_+D)

[ZCi/(10xECLL) xE j 1Bkg + 3.3 Bkg where:

NetCPM = The observed RM-L2 count rate, in cpm, less back-ground, or obtained from the calibration curve.

AF = Administration Factor to account for error in setpoint determination. AF = 0.8.

YCi/(10xECLi) = The ratio of the actual gamma emitting concentrations (excluding dissolved and entrained gases) of-the tank contents to be released to 10 times as listed in 10 CFR 20 the Effluent Concentration Limits (ECL).

E = The release flow rate of waste to be discharged in gallons per minute. A maximum flow rate of 100 gpm will be used for the Evaporator Condensate Storage Tanks and 40 gpm for the Laundry and Shower Sump Tanks.

D = The dilution flow from the Nuclear Services and Decay Heat Sea Water system in gallons per minute.

Bkg = RM-L2 background count rate in cpm.

3.3,Bkg = A statistical spread on the background count rate which represents a 99.95% confidence level on monitor counting. This factor is included to prevent inadver-tent high/trip alarms due to random counts on the monitor.

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Setpoint Calculation 1.4-6 Turbine Building Basement Discharge Line Monitor (RM-L7)

(Continuous Type Releases)

INTRODUCTION The activity released through the Turbine Building Basement Discharge Line Monitor RM-L7 is analyzed in accordance with Section 1.2-5. The setpoint is a fixed concentration based on worst case nuclide released at the worst case rate as described in the Methodology Section below. The monitor setpoint is adjusted to ensure isolation of the release pathway if nuclide concentration limits are exceeded.

METHODOLOGY The alarm/trip setpoint determination is based on the worst case assumption that 1-131 is the only nuclide being discharged. This assumption equates all counts on RM-L7 to 1-131 with an ECL of 1E-6 uci/ml. 1-131 has the most conservative ECL of the nuclides available to this release path and "visible" to RM-L7. The setpoint is based on assuring 10 ECLs or less of 1-131 in the discharge canal and is determined by deriving the cpm from the RM-L7 calibration curve which corresponds to a concentration of 1E-5 uci/ml and applying the flow dilution factor, background counts, and statistical counting variations. The resulting value is used as the alarm/trip setpoint and RM-L7 is adjusted to this or a more conservative value to maintain control on release conditions.

CALCULATION RM - L7 Setpoint (CPM) =CPM x(E+D + Bkg+3.3 Bkg where:

CPM = The counts per minute corresponding to 1E-5 uci/ml (10 ECLs 1-131) from the current RM-L7 calibration curve.

E = The maximum release flow rate of water able to be discharged in gallons per minute.

D The dilution flow from the Nuclear Services and Decay Heat Sea Water system in gallons per minute.

Bkg The background count rate at RM-L7 in cpm.

3.3JBk, A statistical spread on the background count rate which represents a 99.95% confidence level on monitor counting.

This factor is included to prevent inadvertent high/trip alarms due to random counts on the monitor.

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Setpoint Calculation 1.4-7 Turbine Building Basement Discharge Line Monitor (RM-L7)

(Batch Type Releases)

INTRODUCTION Following completion of the analyses required by Section 1.2-4 and determination of release rates and concentration limits in accordance with Section 1.3-2, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded.

METHODOLOGY Station Drain Tank (SDT-1) contents are circulated through radiation monitor RM-L7 and returned to the sump to obtain the actual count rate at RM-L7 for the concentration contained in the tank for release. The observed count rate is adjusted for release flow, background and statistical counting variations, particular to this release flow path. The resulting value is used as the alarm/trip setpoint and RM-L7 is adjusted to this or a more conservative value prior to initiating the release. If the concentration of radionuclides to be released is less than the effluent monitor LLD use setpoint calculation 1.4-8.

CALCULATION RM - L7 Setpoint (CP = FNet CPM x AF x (E+ D) ]1 Bkg + 3.31Bkg (cPY) L(C / (10 x ECL)) x E where:

NetCPM The observed RM-L7 count rate, in cpm, less background.

AF Administration Factor to account for error in setpoint determination. AF = 0.8.

ZCi/(10 x ECL) The ratio of the actual gamma emitting concentrations (excluding dissolved and entrained gases) of the tank contents to be released to 10 times the Effluent Concentration Limits (ECL) as listed in 10 CFR 20.

E The release flow rate of waste to be discharged in gallons per minute. A maximum flow rate of 600 gpm will be used.

D The dilution flow from the Nuclear Services and Decay Heat Sea Water system in gallons per minute.

Bkg = RM-L7 background count rate in cpm.

3.3-Bkg 'A statistical spread on the background count rate which represents a 99.95% confidence level on monitor counting.'This factor is included to prevent inadvertent high/trip alarms due to random counts on the monitor.

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Setpoint Calculation 1.4-8 Alternate Setpoints Methodology for RM-L2 and RM-L7 The following method may be employed to establish an upper bound fixed setpoint for RM-L7. Once established, the setpoint need not be changed unless the monitor response or background changes significantly, or there is a significant change in secondary plant activity levels.

This method may also be used to establish setpoints for laundry tanks being released through RM-L2, and for low activity (< monitor LLD) ECSTs.

Setpoint = [(cpm/pCi/ml) x (1E-5 pCi/ml) x DF x RF] + Bkg where:

cpm/iCi/ml = The monitor response (slope) 1E-5 pCi/ml = Worst case effluent concentration limit, for major gamma emitting isotopes in waste stream, multiplied by 10.

DF = The minimum dilution factor based on maximum tank discharge rate and minimum RW dilution; 100 for ECSTs, 240 for LSSTs, 30 for SDT-1 or CD releases through RM-L7.

RF = Release fraction. RF is that fraction of site liquid releases allocated to a particular liquid effluent monitor. The sum of the RFs for each liquid effluent monitor must be < = 1 during periods of simultaneous releases from liquid effluent discharge points. During periods when simultaneous discharges are not made, RF may be set to 1 for each monitor.

Bkg = Monitor background.

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CALCULATION OF INHALATION PATHWAY DOSE FACTOR (Pi) 3 Pi = K' (BR)DFAi mrern / year per uCi / rn where:

K' A constant unit of conversion - 106 pCi/uCi BR = The Breathing Rate of the child age group = 3700 m3 /year DFAi = The maximum organ inhalation dose factor for the child age group for the ith radionuclide, in mrem/pCi. The total body is considered as an organ in the selection of DFA.

NOTE: For the inhalation pathway Pi = Ri, so values of Pi may be taken from Table 4.4-3.

References:

1) NUREG-0133, Section 5.2.1.1

2) Regulatory Guide 1.109, Table E-5, and Table E-9 OFF-SITE DOSE CALCULATION MANUAL Page 84

SECTION 2.0 RADIOACTIVE EFFLUENTS DOSE REDUCTION SPECIFICATIONS OFF-SITE DOSE CALCULATION MANUAL Page 85

TABLE II RADWASTE REDUCTION SYSTEMS - DOSE PROJECTION DOSE PROJECTION SYSTEM SPECIFICATION CALCULATION PROJECTION FREQUENCY FLOW DIAGRAM Waste 2.1-1 2.2-1 M* 2.3-1 Gas Treatment Ventilation 2.1-1 2.2-1 M* 2.3-1 Exhaust Treatment Liquid 2.1-2 2.2-1 M* 2.3-2 Radwaste Treatment

When a Radwaste Reduction System is not available for use.

OFF-SITE DOSE CALCULATION MANUAL Page 86

WASTE REDUCTION SPECIFICATION NO. 2.1-1 The WASTE GAS SYSTEM shall be used, as required, to reduce the radioactivity of materials in gaseous waste prior to discharge, when projected monthly air doses due to releases of gaseous effluents from the site to areas at or beyond the SITEBOUNDARY would exceed:

1) 0.2 mrad gamma/month *

2) 0.4 mrad beta/month AND The VENTILATION EXHAUST TREATMENT SYSTEM shall be used, as required, to reduce the quantity of radioactive materials in gaseous waste prior to discharge, when projected monthly air doses due to release of gaseous effluents from the site to areas at or beyond the SITE BOUNDARY would exceed:

1) 0.3 mrem to any organ/month

Doses due to gaseous releases from the site shall be projected at least once per 31 days.

The limits of the 10CFR5O, Appendix I, paragraph B1 criteria were reduced to 1/4 of the monthly portion of the annual limit as explained in correspondence among AIF, Utilities and the NRC dated December 24, 1981.

References:

1) Plant Procedures

2) Correspondence C.A. Willis (NRC) to S. Pandy (Franklin Research Center) dated 11/20/81 and AIF letter to AIF subcommittee on RETS dated 12/24/81.

OFF-SITE DOSE CALCULATION MANUAL Page 87

WASTE REDUCTION SPECIFICATION NO. 2.1-2 The LIQUID RADWASTE TREATMENT SYSTEM shall be used, as required, to reduce radioactive materials in liquid wastes prior to their discharge, when projected monthly doses due to liquid effluents discharged to UNRESTRICTED AREAS would exceed the following values:

a. 0.06 mrem whole body/month *

b. 0.2 mrem to any organ/month Doses due to liquid releases shall be projected at least once per 31 days.

The limits of the 10CFR50, Appendix I, paragraph A criteria were reduced to 1/4 of the monthly portion of the annual limit as explained in correspondence among AIF, Utilities and the NRC dated 12/24/81.

References:

1) Plant Procedures

2) Correspondence C.A. Willis (NRC) to S. Pandy (Franklin Research Center) dated 11/20/81 and AIF letter to AIF subcommittee on RETS dated 12/24/81.

OFF-SITE DOSE CALCULATION MANUAL Page 88

DOSE PROJECTION METHODOLOGY 2.2-1 GASEOUS RADWASTE I. INTRODUCTION Crystal River Unit 3 operating practices require use of the WASTE GAS SYSTEM (Waste Gas Decay Tanks). The normal release paths for gaseous effluents are via the VENTILATION EXHAUST TREATMENT SYSTEM (HEPA and Charcoal Filters). The operability of the VENTILATION EXHAUST TREATMENT SYSTEM is controlled by Section 2.4 of Part I of the ODCM.

As long as these practices and specifications are maintained, the radwaste reduction requirements of Part I, Section 2 are met, and there is no need to project doses prior to the release of gaseous radwaste.

II. CALCULATIONS Dose projection calculations will be necessary if either system is not available for use.

D = 31Dc/NDQ where:

D = Projected Dose (monthly).

Dc = Current quarter cumulative dose, including projection for release under evaluation.

NDQ = Number of days into quarter, where the quarterly periods are:

January 1 through March 31, April 1 through June 30, July 1 through September 31, October 1 through December 31.

References:

1) FSAR 5.5.1, 5.5.2 OFF-SITE DOSE CALCULATION MANUAL Page 89

DOSE PROJECTION METHODOLOGY 2.2-2 LIQUID RADWASTE I. INTRODUCTION Crystal River Unit 3 operating practices require liquid radwastes (except for Laundry and Shower Sump waste and Secondary Drain Tank waste) to be processed prior to releasing them to the environment.

As long as these practices are maintained the radwaste reduction requirements of Section 2.3 of Part I of the ODCM are met, and there is no need to project doses prior to the release of liquid radwaste.

II. CALCULATIONS Dose projection calculations will be necessary if there is a malfunction of LIQUID RADWASTE TREATEMENT SYSTEM equipment and liquid radwaste must be released without prior treatment.

D = 31Dc/NDQ where:

DP = Projected Dose (monthly).

Dc = Current quarter cumulative dose, including projection for release under evaluation.

NDQ = Number of days into quarter, where the quarterly periods are:

January 1 through March 31, April 1 through June 30, July 1 through September 31, October 1 through December 31.

References:

1) ODCM Part I, Section 2.3 and 3.3.

OFF-SITE DOSE CALCULATION MANUAL Page 90

TOTAL DOSE SPECIFICATION 2.3 (LIQUID AND GASEOUS RELEASES)

The calendar year dose or dose commitment to any member of the public, due to releases of radioactivity and radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, (except the thyroid which shall be limited to less than or equal to 75 mrems).

This specification is satisfied by meeting specifications 4.1-1, 4.1-2, and 4.1-3.

If doses exceed twice the limits of specifications 4.1-1, 4.1-2, and 4.1-3 then an analysis shall be performed to confirm continued compliance with 40CFR190(b).

References:

1) ODCM Part I, Section 2.10

2) Plant Procedures

3) 40 CFR 190 OFF-SITE DOSE CALCULATION MANUAL Page 91

EFFLUENT FLOW DIAGRAM - GASEOUS 2.3-1 To Atmosphere RM A2 RM Al Control Complex Nuc Sx Rm Hoods Prim. Lab Hoods RM A7 RM A4 MU Pumps Evaporator Neut.

Valve Stats Tank MWST RM A3 Waste Gas Area Sample Area Hood CBAST &

CWST Pumps Heat Hx Vault OFF-SITE DOSE CALCULATION MANUAL Page 92

EFFLUENT FLOW DIAGRAM - LIOUID 2.3-2 From Demins From MWe RCE L&S Sump NSSW System RWP 3A RW P ECIT-B 1 ECST-A 2A To RCBT LSST-A LSST-B

-To 2B RW'P 3B To MWST - To MWST RM I Plant Condensate to MWST SDT-1 --- RM L7 to Settling Ponds TB Sump I to Unit 3 Discharge Canal OFF-SITE DOSE CALCULATION MANUAL Page 93

SECTION 3.0 RADIOACTIVE EFFLUENTS SAMPLING SPECIFICATIONS OFF-SITE DOSE CALCULATION MANUAL Page 94

TABLE III GASEOUS AND LIQUID EFFLUENT REPRESENTATIVE SAMPLING RELEASE TYPE REPRESENTATIVE SAMPLING SOURCE OF EFFLUENT METHOD BATCH CONT.

Evaporator x 3.1-1 Condensate Storage Tanks Laundry and x 3.1-1 Shower Sump Tanks Secondary x x 3.1-1, 3.1-2 Drain Tanks Plant Condensate x 3.1-2 Waste Gas x 3.1-3 Decay Tanks Reactor Bldg. x x 3.1-4 Purge Exhaust Auxiliary Bldg. x 3.1-4

& Fuel Handling Area Purge Exhaust Reactor Bldg. x 3.1-5 with Both Personnel and Equipment Hatches Open OFF-SITE DOSE CALCULATION MANUAL Page 95

Representative Sampling Method No. 3.1-1 (Evaporator Condensate Storage Tanks, Laundry & Shower Sump Tanks, Secondary Drain Tank)

To obtain representative samples from these tanks, the contents of the tank to be sampled will be recirculated through two contained volumes and a grab sample will be collected upon completion. No additions of liquid waste will be made to this tank until completion of the release.

Representative Sampling Method No. 3.1-2 (Secondary Drain Tank and/or Plant Condensate)

A representative sample may be obtained via grab sample of the Turbine Building Sump or the Secondary Drain Tank, Plant Condensate, or from the release compositor.

Representative Sampling Method No. 3.1-3 (Waste Gas Decay Tank)

Representative gas, iodine, and particulate samples are drawn from the waste gas decay tank sample lines.

No additions of waste gas is allowed into a tank following sampling until the release has been completed.

Representative Sampling Method No. 3.1-4 (Reactor Building & Auxiliary Building & Fuel Handling Area Exhaust)

Representative gas, iodine, particulate and tritium samples are taken from these ducts at the location of the radiation monitors. The sample for the Reactor Building Purge Duct is taken form radiation monitor RM-A6 prior to a purge and is drawn from radiation monitor RM-A1 during a purge. The sample for the Auxiliary Building and Fuel Handling Area Exhaust Duct is drawn from RM-A2 during venting since this is a continuous release pathway.

If samples cannot be obtained from the ducts of the Reactor or Auxiliary Building, samples can be obtained from areas of these buildings that are considered to be representative of the radionuclide concentrations present throughout the respective buildings. Sampling times and volumes should be established to assure the LLD Limits of Sections 1.2 and 4.2 for the radionuclides can be met.

OFF-SITE DOSE CALCULATION MANUAL Page 96

Representative Sampling Method No. 3.1-5 (Reactor Building With Personnel And Equipment Hatch Opened)

The following requirements do not apply when the Personnel Hatch or Equipment Hatch is closed, or when a structure, such as a wooden door, is used in lieu of either Hatch. By having one of these hatches closed, sustained drafts through the RB are prevented.

Requirements:

The Reactor Building purge exhaust fans are operational and the supply fans are shut down. If the purge exhaust must be shut down then either the personnel hatch or equipment hatch openings must be closed.

Monitor the Reactor Building recirculation system by using RM-A6 or by taking general area air samples.

Considerations:

Run the main purge long enough to assure cleanup of the RB atmosphere.

Degas and depressurize the Reactor Coolant System.

Representative Sampling Method No. 3.1-6 (Reactor Building During Integrated Leak Rate Test)

Due to building overpressure, prepurge samples cannot be taken from RM-A6.

Representative gas, iodine, particulate and tritium samples may be obtained from the Intermediate Building containment sampling apparatus or the Post-Accident Sampling System.

Reference:

Telecon-FPC (Dan Green, Dan Wilder) to NRC (Charles Willis) dated 03/15/85 at 0930;

Subject:

Personnel and Equipment Hatch Openings.

OFF-SITE DOSE CALCULATION MANUAL Page 97

SECTION 4.0 RADIOACTIVE EFFLUENTS DOSE CALCULATIONAL SPECIFICATIONS OFF-SITE DOSE CALCULATION MANUAL Page 98

TABLE IV CUMULATIVE DOSE CALCULATION DOSE NUCLIDE CALCULATION DOSE PATHWAY SPECIFICATION ANALYSIS METHODOLOGY FACTORS Noble Gases 4.1-1 4.2-1, 4.2-2 4.3-1 4.4-1 4.2-3 Radioiodines, Radioactive Parti cul ates 4.1-2 4.2-1, 4.2-2 4.3-2 4.4-2 to 4.4-16 Radionuclides 4.2-3 other than Noble Gases Liquid Effluents 4.1-3 4.2-4, 4.2-5 4.3-3 4.4-17 OFF-SITE DOSE CALCULATION MANUAL Page 99

DOSE SPECIFICATION 4.1-1 (NOBLE GASES)

The air dose at or beyond the SITE BOUNDARY due to radioactive noble gases released in gaseous effluents shall be limited as follows:

1) During any calendar quarter, < 5 mrad gamma, and < 10 mrad beta radiation.

2) During any calendar year, < 10 mrad gamma, and < 20 mrad beta radiation.

Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined at least once per 31 days.

References:

1) ODCM Part I, Section 2.8 OFF-SITE DOSE CALCULATION MANUAL Page 100

DOSE SPECIFICATION 4.1-2 (RADIOIODINE & PARTICULATES)

The dose to a MEMBER OF THE PUBLIC from 1-131, 1-133, Tritium and radioactive particulates with half lives of greater than 8 days in gaseous effluents released from the site to areas at or beyond the SITE BOUNDARY shall be limited as follows:

1) During any calendar quarter, < 7.5 mrem to any organ.

2) During any calendar year, < 15 mrem to any organ.

Cumulative dose calculations for the current calendar quarter and current calendar year shall be determined at least once per 31 days.

References:

1) ODCM Part I, Section 2.9 OFF-SITE DOSE CALCULATION MANUAL Page 101

DOSE SPECIFICATION 4.1-3 (LIQUID EFFLUENTS)

The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to UNRESTRICTED AREAS shall be limited as follows:

1) During any calendar quarter, < 1.5 mrem total body.

2) During any calendar quarter, < 5 mrem any organ.

3) During any calendar year, < 3 mrem total body.

4) During any calendar year, < 10 mrem any organ.

Cumulative dose contributions from liquid effluents shall be determined at least once per 31 days.

References:

1) ODCM Part I, Section 2.6 OFF-SITE DOSE CALCULATION MANUAL Page 102

NUCLIDE ANALYSIS 4.2-1 REACTOR BUILDING PURGE EXHAUST NUCLIDE SAMPLE SOURCE LLD(b) (uCi/ml)

A. Principal Gamma Emitters (a)

Mn-54 Fe-59 Co-58 Batch release particulate filter Co-60 for Batch Releases. Weekly 4 11 Zn-65 Particulate Filter Analysis for 1x10- /1x10-Mo-99 continuous(c) type release.

Cs-134 Cs-137 Ce-141 Ce-144 Kr-87 Pre-release grab sample for Batch Kr-88 type release. Weekly grab sample lx10-4 Xe-133 for continuous type release.

Xe-133m Xe-135 Xe-138 12 B. Iodine 131 Batch release charcoal filter for NA/1 x 10-Batch Releases. Weekly charcoal filter for continuous releases.

6 C. Tritium Pre-release Grab Sample. 1x10-D. Gross Alpha Monthly Particulate Filter Composite lx10-11 I1- 11 E. Sr-89 Quarterly Particulate Filter Composite lx10 lx10-1 1 F. Sr-90 Quarterly Particulate Filter Composite (a) Other identified Gamma Emitters not listed in this table shall be included in dose calculations.

(b) The first value refers to the LLD for pre-release grab sample; the second value refers to the LLD for weekly Particulate Filter Analysis.

(c) Reactor Building Purge is considered continuous after minimum of one Reactor Building volumes have been released on a continuous basis (i.e., first one volume is a batch type).

OFF-SITE DOSE CALCULATION MANUAL Page 103

NUCLIDE ANALYSIS 4.2-2 AUXILIARY BUILDING AND FUEL HANDLING AREA EXHAUST NUCLIDE SAMPLE SOURCE LLD(b) (uCi/ml)

A. Principal Gamma Emitters (a)

Mn-54 Fe-59 Co-58 Weekly Parrticulate Filter Analysis.

Co-60 1x10- 4 /1x10-1 1 Zn-65 Mo-99 Cs-134 Cs-137 Ce-141 Ce-144 Kr-87 Monthly G rab Sample.

Kr-88 lx10-4 Xe-133 Xe-133m Xe-135 Xe-138 B. Iodine 131 Weekly Ch* ircoal Filter Analysis. lx10-12 6

C. Tritium Monthly G rab Sample. 1x10-lx 1 0 -1 1 D. Gross Alpha Monthly P* articulate Filter Composite lx 1100 -1 1 lx E. Sr-89 Quarterly Particulate Filter Composite

-11 F. Sr-90 Quarterly Particulate Filter Composite (a) Other identified Gamma Emitters not listed in this table shall be included in dose calculations.

(b) The first value refers to the LLD for pre-release grab sample; the second value refers to the LLD for weekly Particulate Filter Analysis.

OFF-SITE DOSE CALCULATION MANUAL Page 104

NUCLIDE ANALYSIS 4.2-3 WASTE GAS DECAY TANKS NUCLIDE SAMPLE SOURCE LLD (b) (jCi/ml )

A. Principal Gamma Emitters (a)

Mn-54 Fe-59 Co-58 Co-60 Zn-65 Mo-99 Weekly Particulate Filter sample (from RM-A2) 1x10-4 /1x10 1 1 Cs-134 Cs-137 Ce-141 Ce-144 Kr-87 Kr-88 Xe-133 Pre-release Grab sample lx10-4 Xe-133m Xe-135 Xe-138 12 B. Iodine 131 Weekly Charcoal Filter (from RM-A2) 1x10 (a) Other identified Gamma Emitters not listed in this table shall be included in dose and setpoint calculations.

(b) The first value refers to the LLD for pre-release grab sample; the second value refers to the LLD for weekly Particulate Filter Analysis.

OFF-SITE DOSE CALCULATION MANUAL Page 105

NUCLIDE ANALYSIS 4.2-4 EVAPORATOR CONDENSATE STORAGE TANKS, LAUNDRY AND SHOWER SUMP TANKS, SECONDARY DRAIN TANK NUCLIDE SAMPLE SOURCE LLD(uCi/ml)

A. Principal Gamma Emitters (a)

Mn-54 Fe-59 Co-58 Co-60 Zn-65 Pre-release Grab Sample 5x10-7 Mo-99 Cs-134 Cs-137 Ce-141 Ce-144 6

B. Iodine 131 Pre-Release Grab Sample 1x10-C. Dissolved and Entrained Noble 5

Gases Monthly Grab Sample 1x10-D. Tritium Monthly Composite 5 lx10-E. Gross Alpha Monthly Composite lx10 7 8

F. Sr-89 Quarterly Composite 5x10-8 G. Sr-90 Quarterly Composite 5x10-6 H. Fe-55 Quarterly Composite 1x10-(a) Other identified Gamma Emitters not listed in this table shall be included in dose calculations.

OFF-SITE DOSE CALCULATION MANUAL Page 106

NUCLIDE ANALYSIS 4.2-5 SECONDARY DRAIN TANK AND/OR PLANT CONDENSATE NUCLIDE SAMPLE SOURCE LLD(uCi/ml)

A. Principal Gamma Emitters (a)

Mn-54 Fe-59 Co-58 Co-60 Zn-65 Weekly Composite 5x10-7 Mo-99 Cs-134 Cs-137 Ce-141 Ce-144 6

B. Iodine 131 Weekly Composite 1x10-C. Dissolved and Entrained Noble 5

Gases Monthly Grab Sample 1x10-5 D. Tritium Monthly Composite 1x10-7 E. Gross Alpha Monthly Composite 1x10-8 F. Sr-89 Quarterly Composite 5x10-G. Sr-90 Quarterly Composite 5x10-8 6

H. Fe-55 Quarterly Composite 1x10 (a) Other identified Gamma Emitters not listed in this table shall be included in dose calculations OFF-SITE DOSE CALCULATION MANUAL Page 107

DOSE CALCULATION 4.3-1 (NOBLE GAS)

The air dose at or beyond the SITE BOUNDARY due to noble gases released in gaseous effluents is calculated as follows:

Dy = 3.17 x 10' 1 Mi(X/Q)Qi mrad Do = 3.17 x 10' 8 1 N (X/Q)Qi mrad where:

Dy The air dose at or beyond the SITE BOUNDARY due to gamma emissions from noble' gases in gaseous effluents in mrad/time period.

Do The air dose at or beyond the SITE BOUNDARY due to beta emissions from noble gases in gaseous effluents in mrad/time period.

3.17 x 10' = The number of years in one second, yr/sec.

MI. The air dose factor due to gamma emissions for each identified noble gas radionuclide, in mrad/year per uCi/m 3 .

N1 The air dose factor due to beta emissions for each identified noble gas radionuclide, in mrad/year per uCi/m 3 .

X/Q The highest calculated annual average relative concentration for areas at or beyond the UNRESTRICTED AREA Boundary, 2.5 x 10-6 sec/m'.

Qi Total pCi of isotope i released during the calendar quarter or calendar year, as appropriate.

OFF-SITE DOSE CALCULATION MANUAL Page 108

DOSE CALCULATION 4.3-2 (RADIOIODINES & PARTICULATES)

The dose to an individual at or beyond the SITE BOUNDARY due to 1-131, 1-133, Tritium and radioactive particulates with half lives of greater than 8 days is calculated as follows:

D = 3.17 x 10- YZWRQ mrem where:

D = The radiation dose to an individual at or beyond the UNRESTRICTED AREA BOUNDARY, in mrem.

R. The dose factor for each identified radionuclide, i, in m2 (mrem/year) per uCi/sec or mrem/year per uCi/m 3 .

W X/Q for inhalation pathway, 2.5 x 10-6 sec/m3 the site boundary and 7.5 x 10' sec/m3 at the critical receptor.

W = D/Q for food and ground plane pathway, 1.9 x 10-'m 2 the site boundary and 5.7 x 10- m-2 at the critical receptor.

Total pCi of isotope i released during the calendar quarter or calendar year, as appropriate.

3.17 x 10- = The number of years in one second, yr/sec.

Reference:

NUREG 0133, Section 5.3.1 FSAR, Table 2-20 OFF-SITE DOSE CALCULATION MANUAL Page 109

DOSE CALCULATION 4.3-3 (LIQUID EFFLUENTS)

The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to UNRESTRICTED AREAS is calculated as follows:

D = Air Y_ tkCikFk]

where:

D = The cumulative dose commitment to the total body or any organ, T, from the liquid effluents for the total time period Ytk in mrem.

tk The length of the kth time period over which Cik is averaged for all liquid releases, in hours.

Cik = The average concentration of radionuclide, i, in undiluted liquid effluent during time period tk from any liquid release, in wCi/ml.

Ai The site related ingestion dose commitment factor to the total body or any organ for each identified principal gamma and beta emitter as shown in Table 4.4-17 of this manual, in mrem-ml per hour-pCi.

Fk = Waste flow rate / (Waste flow rate + Dilution flow rate)*

Dilution flow rate is the sum of available circulating water and Nuclear Services and Decay Heat Seawater flow.

Units 1 and 2 circulating water flow may be included.

References:

1) NUREG 0133, Section 4.3.

2) *Telecon/Meeting Summary with C. Willis (USNRC) dated 01/16/85 regarding Fk OFF-SITE DOSE CALCULATION MANUAL Page 110

TABLE 4.4-1 flcw rArTnDC 1fD PVDAnIIDI TO A SEMI-INFINITE CLOUD OF NOBLE GASES Ni Li Mi Ki Nuclide P-Air * (DFiP) P-Skin ** (DFSi) y-Air * (DFic) y-Body ** (DFBi)

Kr-83m 2.88E+2 1. 93E+1 7.56E-2 KR-85m 1.97E+3 1.46E+3 1. 23E+3 1.17E+3 Kr-85 1.95E+3 1.34E+3 1. 72E+1 1.61E+1 Kr-87 1.03E+4 9.73E+3 6.17E+3 5.92E+3 Kr-88 2.93E+3 2 . 37E+3 1. 52E+4 1.47E+4 Kr-89 1.06E+4 1.O1E+4 1. 73E+4 1.66E+4 KR-90 7.83E+3 7.29E+3 1. 63E+4 1.56E+4 Xe-131m 1. 11E+3 4.76E+2 1. 56E+2 9.15E+1 Xe-133m 1.48E+3 9.94E+2 3.27E+2 2 51E+2 Xe-133 1.05E+3 3.06E+2 3.53E+2 2.94E+2 Xe-135m 7.39E+2 7.11E+2 3.36E+3 3.12E+3 Xe-135 2.46E+3 1.86E+3 1.92E+3 1.81E+3 Xe-137 1.27E+4 1.22E+4 1. 51E+3 1.42E+3 Xe-138 4.75E+3 4.13E+3 9.21E+3 8.83E+3 Ar-41 3.28E+3 2. 69E+3 9.30E+3 8.84E+3 3

mrad-m pci -yr 3

mrem-m pCi -yr

References:

1) NUREG 0133

2) USNRC Regulatory Guide 1.109, Table B-1 OFF-SITE DOSE CALCULATION MANUAL Page 111

CALCULATION OF INHALATION PATHWAY DOSE FACTOR (Ri) 3 Ri=K' (BR)DFAi mrem/yearper uCi/m where:

K' = A constant unit of conversion - 106 pCi/uCi BR = The Breathing Rate of the represented age group:

1400 m3 /yr - infant 3700 m3 /yr - child 8000 m3 /yr - teen 8000 m3 /yr - adult DFA, = The maximum organ inhalation dose factor for the represented age group for the ith radionuclide, in mrem/pCi.

References:

1) NUREG-0133, Section 5.3.1.1

2) Regulatory Guide 1.109, Table E-5, and Tables E-7 through E-10 OFF-SITE DOSE CALCULATION MANUAL Page 112

TABLE 4.4-2 Inhalation Dose Factors (Ri) - Infant Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 Cr-51 ND ND 8.95E1 1.32E1 1.32E1 1.28E4 3.57E2 Mn-54 ND 2.53E4 4.98E3 4.98E3 4.98E3 9.95E5 7.06E3 Fe-55 1.97E4 1.17E4 3.33E3 ND ND 8.69E4 1.09E3 Fe-59 1.36E4 2.35E4 9 .48E3 ND ND 1.02E6 2.48E4 Co-58 ND 1.22E3 1.82E3 ND ND 7.77E5 1.11E4 Co-60 ND 8.02E3 1.18E4 ND ND 4.51E6 3.19E4 Ni-63 3.39E5 2.04E4 1.16E4 ND ND 2.09E5 2.42E3 Zn-65 1.93E4 6.26E4 3.11E4 ND 3.25E4 6.47E5 5.14E4 Rb-86 ND 1.90E5 8.82E4 ND ND ND 3.04E3 Sr-89 3.98E5 ND 1.14E4 ND ND 2.03E6 6.40E4 Sr-90 4.09E7 ND 2.59E6 ND ND 1.12E7 1.31E5 Y-91 5.88E5 ND 1.57E4 ND ND 2.45E6 7.07E4 Zr-95 1.15E5 2.79E4 2.03E4 ND 3.11E4 1.75E6 2.17E4 Nb-95 1.57E4 6.43E3 3.78E3 ND 4.72E3 4.79E5 1.27E4 Ru-103 2.02E3 ND 6.79E2 ND 4.24E3 5.52E5 1.61E4 Ru-106 8.68E4 ND 1.09E4 ND 1.07E5 1.16E7 1.64E5 Ag-110m 9.98E3 7.22E3 5.00E3 ND 1.09E4 3.67E6 3.30E4 Te-125m 4.76E3 1.99E3 6.58E2 1.62E3 ND 4.47E5 1.29E4 Te-127m 1.67E4 6.90E3 2.07E3 4.87E3 3.75E4 1.31E6 2.73E4 Te-129m 1.41E4 6.09E3 2.23E3 5.47E3 3.18E4 1.68E6 6.90E4 1-131 3.79E4 4.44E4 1.96E4 1.48E7 5.18E4 ND 1.06E3 Cs-134 3.96E5 7.03E5 7.45E4 ND 1.90E5 7.97E4 1.33E3 Cs-136 4.83E4 1.35E5 5.29E4 ND 5.64E4 1.18E4 1.43E3 Cs-137 5.49E5 6.12E5 4.55E4 ND 1.72E5 7.13E4 1.33E3 Ba-140 5.60E4 5.60E1 2.90E3 ND 1.34E1 1.60E6 3.84E4 Ce-141 2.77E4 1.67E4 1.99E3 ND 5.25E3 5.17E5 2.16E4 Ce-144 3.19E6 1.21E6 1.76E5 ND 5.38E5 9.84E6 1.48E5 Pr-143 1.40E4 5.24E3 6.99E2 ND 1.97E3 4.33E5 3.72E4 Nd-147 7.94E3 8.13E3 5.00E2 ND 3.15E3 3.22E5 3.12E4 OFF-SITE DOSE CALCULATION MANUAL Page 113

TABLE 4.4-3 Inhalation Dose Factors (Ri) - Child I Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 1.12E3 1.12E3 1.12E3 1.12E3 1.12E3 1.12E3 1.12E3 Cr-51 ND ND 1.54E2 8.55E1 2.43E1 1.70E4 1.08E3 Mn-54 ND 4.29E4 9.51E3 ND 1.O0E4 1.58E6 2.29E4 Fe-55 4.74E4 2.52E4 7.77E3 ND ND 1.11E5 2.87E3 Fe-59 2.07E4 3.34E4 1.67E4 ND ND 1.27E6 7.07E4 Co-58 ND 1.77E3 3.16E3 ND ND 1.11E6 3.44E4 Co-60 ND 1.31E4 2.26E4 ND ND 7.07E6 9.62E4 Ni-63 8.21E5 4.63E4 2.80E4 ND ND 2.75E5 6.33E3 Zn-65 4.26E4 1.13E5 7.03E4, ND 7.14E4 9.95E5 1.63E4 Rb-86 ND 1.98E5 1.14E5 ND ND ND 7.99E3 Sr-89 5.99E5 ND 1.72E4 ND ND 2.16E6 1.67E5 Sr-90 1.01E8 ND 6.44E6 ND ND 1.48E7 3.43E5 Y-91 9.14E5 ND 2.44E4 ND ND 2.63E6 1.84E5 Zr-95 1.90E5 4.18E4 3.70E4 ND 5.96E4 2.23E6 6.11E4 Nb-95 2.35E4 9.18E3 6.55E3 ND 8.62E3 6.14E5 3.70E4 Ru-103 2.79E3 ND 1.07E3 ND 7.03E3 6.62E5 4.48E4 Ru-106 1.36E5 ND 1.69E4 ND 1.84E5 1.43E7 4.29E5 Ag-110m 1.69E4 1.14E4 9.14E3 ND 2.12E4 5.48E6 1.00E5 Te-125m 6.73E3 2.33E3 9.14E2 1.92E3 ND 4.77E5 3.38E4 Te-127m 2.49E4 8.55E3 3.02E3 6.07E3 6.36E4 1.48E6 7.14E4 Te-129m 1.92E4 6.85E3 3.04E3 6.33E3 5.03E4 1.76E6 1.82E5 1-131 4.81E4 4.81E4 2.73E4 1.62E7 7.88E4 ND 2.84E3 Cs-134 6.51E5 1.01E6 2.25E5 ND 3.30E5 1.21E5 3.85E3 Cs-136 6.51E4 1.71E5 1.16E5 ND 9.55E4 1.45E4 4.18E3 Cs-137 9.07E5 8.25E5 1.28E5 ND 2.82E5 1.04E5 3.62E3 Ba-140 7.40E4 6.48E1 4.33E3 ND 2.11E1 1.74E6 1.02E5 Ce-141 3.92E4 1.95E4 2.90E3 ND 8.55E3 5.44E5 5.66E4 Ce-144 6.77E6 2.12E6 3.61E5 ND 1.17E6 1.20E7 3.89E5 Pr-143 1.85E4 5.55E3 9.14E2 ND 3.00E3 4.33E5 9.73E4 Nd-147 1.08E4 8.73E3 6.81E2 ND 4.81E3 3.28E5 8.21E4 OFF-SITE DOSE CALCULATION MANUAL Page 114

TABLE 4.4-4 Inhalation Dose Factors (Ri) - Teen I Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 1.27E3 1.27E3 1.27E3 1.27E3 1.27E3 1.27E3 1.27E3 Cr-51 ND ND 1.35E2 7.49E1 3.07E1 2.09E4 3.OOE3 Mn-54 ND 1.70E0 8.40E3 ND 1.27E4 1.98E6 6.68E4 Fe-55 3.34E4 2.38E4 5.54E3 ND ND 1.24E5 6.39E3 Fe-59 1.59E4 3.70E4 1.43E4 ND ND 1.53E6 1.78E5 Co-58 ND 2.07E3 2.78E3 ND ND 1.34E6 9.52E4 Co-60 ND 1.51E4 1.98E4 ND ND 8.72E6 2.59E5 Ni-63 5.80E5 4.34E4 1.98E4 ND ND 3.07E5 1.42E4 Zn-65 3.86E4 1.34E5 6.24E4 ND 8.64E4 1.24E6 4.66E4 Rb-86 ND 1.90E5 8.4OE4 ND ND ND 1.77E4 Sr-89 4.34E5 ND 1.25E4 ND ND 2.42E6 3.71E5 Sr-90 1.08E8 ND 6.68E6 ND ND 1.65E7 7.65E5 Y-91 6.61E5 ND 1.77E4 ND ND 2.94E6 4.09E5 Zr-95 1.48E5 4.58E4 3.15E4 ND 6.74E4 2.69E6 1.49E5 Nb-95 1.86E4 1.03E4 5.66E3 ND 1.00E4 7.51E5 9.68E4 Ru-103 2.1OE3 ND 8.96E3 ND 7.43E3 7.83E5 1.09E5 Ru-106 9.84E4 ND 1.24E4 ND 1.90E5 1.61E7 9.60E5 Ag-110m 1.38E4 1.31E4 7.99E3 ND 2.50E4 6.75E6 2.73E5 Te-125m 4.88E3 2.24E3 6.67E2 1.40E3 ND 5.36E5 7.50E4 Te-127m 1.80E4 8.16E3 2.18E3 4.38E3 6.54E4 1.66E6 1.59E5 Te-129m 1.39E4 6.58E3 2.25E3 4.58E3 5.19E4 1.98E6 4.05E5 1-131 3.54E4 4.91E4 2.64E4 1.46E7 8.40E4 ND 6.49E3 Cs-134 5.02E5 1.13E6 5.49E5 ND 3.75E5 1.46E5 9.76E3 Cs-136 5.15E4 1.94E5 1.37E5 ND 1.10E5 1.78E4 1.09E4 Cs-137 6.70E5 8.48E5 3.11E5 ND 3.04E5 1.21E5 8.48E3 Ba-140 5.47E4 6.70E1 3.52E3 ND 2.28EI 2.03E6 2.29E5 Ce-141 2.84E4 1.90E4 2.17E3 ND 8.88E3 6.14E5 1.26E5 Ce-144 4.89E6 2.02E6 2.62E5 ND 1.21E6 1.34E7 8.64E5 Pr-143 1.34E4 5.31E3 6.62E2 ND 3.09E3 4.83E5 2.14E5 Nd-147 7.86E3 8.56E3 5.13E2 ND 5.02E3 3.72E5 1.82E5 OFF-SITE DOSE CALCULATION MANUAL Page 115

TABLE 4.4-5 Inhalation Dose Factors (Ri) - Adult Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 1.26E3 1.26E3 1.26E3 1.26E3 1.26E3 1.26E3 1.26E3 Cr-51 ND ND 1.00E2 5.95E1 2.28E1 1.44E4 3.32E3 Mn-54 ND 3.96E4 6.30E3 ND 9.84E3 1.40E6 7.74E4 Fe-55 2.46E4 1.70E4 3.94E3 ND ND 7.21E4 6.03E3 Fe-59 1.18E4 2.78E4 1.06E4 ND ND 1.02E6 1.88E5 Co-58 ND 1.58E3 2.07E3 ND ND 9.28E5 1.06E5 Co-60 ND 1.15E4 1.48E4 ND ND 5.97E6 2.85E5 Ni-63 4.32E5 3.14E4 1.45E4 ND ND 1.78E5 1.34E4 Zn-65 3.24E4 1.03E5 4.66E4 ND 6.90E4 8.64E5 5.34E4 Rb-86 ND 1.35E5 5.90E4 ND ND ND 1.66E4 Sr-89 3.04E5 ND 8.72E3 ND ND 1.4E6 3.5E5 Sr-90 9.92E7 ND 6.10E6 ND ND 9.60E6 7.22E5 Y-91 4.62E5 ND 1.24E4 ND ND 1.70E6 3.85E5 Zr-95 1.07E5 3.44E4 2.33E4 ND 5.36E4 1.77E6 1.50E5 Nb-95 1.41E4 7.76E3 4.21E3 ND 7.74E3 5.05E5 1.04E5 Ru-103 1.53E3 ND 6.58E2 ND 5.83E3 5.05E5 1.10E5 Ru-106 6.91E4 ND 8.72E3 ND 1.34E5 9.36E6 9.12E5 Ag-110m 1.08E4 1.00E4 5.94E3 ND 1.97E4 4.63E6 3.02E5 Te-125m 3.42E3 1.58E3 4.67E2 1.05E3 1.24E4 3.14E5 7.06E4 Te-127m 1.26E4 5.77E3 1.57E3 3.29E3 4.58E4 9.60E5 1.50E6 Te-129m 9.76E3 4.67E3 1.58E3 3.44E3 3.66E4 1.16E6 3.83E5 1-131 2.52E4 3.58E4 2.05E4 1.19E7 6.13E4 ND 6.28E3 Cs-134 3.73E5 8.48E5 7.28E5 ND 2.87E5 9.76E4 1.04E4 Cs-136 3.90E4 1.46E5 1.10E5 ND 8.56E4 1.20E4 1.17E4 Cs-137 4.78E5 6.21E5 4.28E5 ND 2.22E5 7.52E4 8.40E3 Ba-140 3.90E4 4.90E1 2.57E3 ND 1.67E1 1.27E6 2.18E5 Ce-141 1.99E4 1.35E4 1.53E3 ND 6.26E3 3.62E5 1.20E5 Ce-144 3.43E6 1.43E6 1.84E5 ND 8.48E5 7.78E6 8.16E5 Pr-143 9.36E3 3.75E3 4.64E2 ND 2.16E3 2.81E5 2.00E5 Nd-147 5.27E3 6.10E3 3.65E2 ND 3.56E3 2.21E5 1.73E5 OFF-SITE DOSE CALCULATION MANUAL Page 116

Calculation of Ingestion Dose Factor Grass-Cow-Milk Pathway Rc[D/Q]= K' LQF(rDFLfwfLp- + (I- fpfs" itu je where: Unit = m2.mrem/yr per pCi/sec Reference Table R.G. 1.109 K' = A constant of unit conversion, 106 pCi/Ci.

QF = The cow's consumption rate, 50 kg/day (wet weight) E-3 Uap = The receptor's milk consumption rate for age (a), E-5 in liters, yr Infant & Child - 330 Teen - 400 Adult - 310 Ys = The agricultural productivity by unit area of pasture E-15 feed grass 0.7 kg/m 2 Ys = The agricultural productivity of unit area of E-15 stored feed 2.0 kg/M 2 Fm = The stable element transfer coefficients, in days/kg. E-1 r = Fraction of deposited activity retained on cow's E-15 feed grass 1.0 radioiodine 0.2 particulates tf = Transport time from pasture to receptor, in sec. E-15 1.73x105 sec (2 days) tl = Transport time from crop field to receptor, in sec. E-15 7.78x106 sec. (90 days)

(DFL)a= The maximum organ ingestion'dose factor for the ith E-11 to radionuclide for the receptor in age group (a), E-14 in mrem/pCi A = The decay constant for the ith radionuclide, in sec A, = The decay constant for removal of activity on leaf and E-15 plant surfaces by weathering 5.73 x 10-7 sec-(corresponding to a 14 day half-life).

fp = Fraction of the year that the cow is on pasture (dimensionless) = 1*.

fs = Fraction of the cow feed that is pasture grass while the cow is on pasture (dimensionless) = 1".

Milk cattle are considered to be fed from two potential sources, pasture grass and stored feeds.

OFF-SITE DOSE CALCULATION MANUAL Page 117

Note: The above equation does not apply to the concentration of tritium in meat. A separate equation is provided in NUREG 0133, section 5.3.1.4 to determine Tritium value.

Reference:

The equation for Rc. (D/Q) was taken from NUREG-0133 Section 5.3.1.3 OFF-SITE DOSE CALCULATION MANUAL Page 118

TABLE 4.4-6 Ingestion Dose Factors (C) I Grass-Cow-Milk Pathway (Infant)

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 2.38E3 2.38E3 2.38E3 2.38E3 2.38E3 2.38E3 2.38E3 Cr-51 ND ND 1.61E5 1.05E5 2.30E4 2.05E5 4.71E6 Mn-54 ND 3.89E7 8.83E6 ND 8.63E6 ND 1.43E7 Fe-55 1.35E8 8.72E7 2.33E7 ND ND 4.26E7 1.11E7 Fe-59 2.26E8 3.94E8 1.55E8 ND ND 1.17E8 1.88E8 Co-58 ND 2.43E7 6.06E7 ND ND ND 6.05E7 Co-60 ND 8.81E7 2.08E8 ND ND ND 2.10E8 Ni-63 3.49E10 2.16E9 1.21E9 ND ND ND 1.07E8 Zn-65 5.55E9 1.90E10 8.78E9 ND 9.24E9 ND 1.61E10 Rb-86 ND 2.23E10 1.1OE1O ND ND ND 5.70E8 Sr-89 ND 1.45E6 9.98E5 ND ND ND 4.93E5 Sr-90 1.22El1 ND 3.1OE1O ND ND ND 1.52E9 Y-91 7.33E4 ND 1.95E3 ND ND ND 5.26E6 Zr-95 6.84E3 1.67E3 1.18E3 ND 1.80E3 ND 8.30E5 Nb-95 5.93E5 2.44E5 1.41E5 ND 1.75E5 ND 2.06E8 Ru-103 8.68E3 ND 2.90E3 ND 1.81E4 ND 1.06E5 Ru-106 1.90E5 ND 2.38E4 ND 2.25E5 ND 1.44E6 Ag-110m 3.86E8 2.82E8 1.87E8 ND 4.03E8 ND 1.46E10 Te-125m 1.51E8 5.04E7 2.04E7 5.07E7 ND ND 7.18E7 Te-127m 4.21E8 1.40E8 5.10E7 1.22E8 1.04E9 ND 1.70E8 Te-129m 5.60E8 1.92E8 8.62E7 2.15E8 1.40E9 ND 3.34E8 1-131 2.72E9 3.21E9 1.41E9 1.05E12 3.75E9 ND 1.15E8 Cs-134 3.65E10 6.80E10 6.87E9 ND 1.75E10 7.18E9 1.85E8 Cs-136 2.03E9 5.96E9 2.22E9 ND 2.37E9 4.85E8 9.05E7 Cs-137 5.15E10 6.02E10 4.27E9 ND 1.62E10 6.55E9 1.88E8 Ba-140 2.41E8 2.41E5 1.24E7 ND 5.73E4 1.48E5 5.92E7 Ce-141 4.34E4 2.64E4 3.11E3 ND 8.16E3 ND 1.37E7 Ce-144 2.33E6 9.52E5 1.30E5 ND 3.85E5 ND 1.33E8 Pr-143 1.49E3 5.56E2 7.37E1 ND 2.07E2 ND 7.85E5 Nd-147 8.86E2 9.10E2 5.57E1 ND 3.51E2 ND 5.77E5 OFF-SITE DOSE CALCULATION MANUAL Page 119

TABLE 4.4-7 Ingestion Dose Factors (c) I Grass-Cow-Milk Pathway (Child)

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 1.57E3 1.57E3 1.57E3 1.57E3 1.57E3 1.57E3 1.57E3 Cr-51 ND ND 1.02E5 5.66E4 1.55E4 1.03E5 5.41E6 Mn-54 ND 2.09E7 5.58E6 ND 5.87E6 ND 1.76E7 Fe-55 1.12E8 5.93E7 1.84E7 ND ND 3.35E7 1.10E7 Fe-59 1.21E8 1.96E8 9.75E7 ND ND 5.67E7 2.04E8 Co-58 ND 1.21E7 3.72E7 ND ND ND 7.08E7 Co-60 ND 4.32E7 1.27E8 ND ND ND 2.39E8 Ni-63 2.96E10 1.59E9 1.01E9 ND ND ND 1.07E8 Zn-65 4.13E9 1.10E10 6.85E9 ND 6.94E9 ND 1.93E9 Rb-86 ND 8.77E9 5.39E9 ND ND ND 5.64E8 Sr-89 6.69E9 ND 1.91E8 ND ND ND 2.59E8 Sr-90 1.12E11 ND 2.83E10 ND ND ND 1.50E9 Y-91 3.91E4 ND 1.04E3 ND ND ND 5.21E6 Zr-95 3.85E3 8.46E2 7.53E2 ND 1.21E3 ND 8.83E5 Nb-95 3.18E5 1.24E5 8.84E4 ND 1.16E5 ND 2.29E8 Ru-103 4.29E3 ND 1.65E3 ND 1.08E4 ND 1.11E5 Ru-106 ,9.24E4 ND 1,.15E4 ND 1.25E5 ND 1.44E6 Pg-110m 2.09E8 1.41E8 1.13E8 ND 2.63E8 ND 1.68E10 Te-125m 7.38E7 2.00E7 9.84E6 2.07E7 ND ND 7.12E7 Te-127m 2.08E8 5.60E7 2.47E7 4.97E7 5.93E8 ND 1.68E8 Te-129m 3.17E8 8.85E7 4.92E7 1.02E8 9.31E8 ND 3.87E8 1-131 1.30E9 1.31E9 7.46E8 4.34E11 2.15E9 ND 1.17E8 Cs-134 2.26E10 3.71E10 7.84E9 ND 1.15E10 4.13E9 2.00E8 Cs-136 1.04E9 2.85E9 1.84E9 ND 1.52E9 2.26E8 1.00E8 Cs-137 3.22E10 3.09E10 4.55E9 ND 1.01E10 3.62E9 1.93E8 Ba-140 1.17E8 1.03E5 6.84E6 ND 3.34E4 6.12E4 5.94E7 Ce-141 2.19E4 1.09E4 1.62E3 ND 4.78E3 ND 1.36E7 Ce-144 1.62E6 5.09E5 8.66E4 ND 2.82E5 ND 1.33E8 Pr-143 7.19E2 2.16E2 3.57E1 ND 1.17E2 ND 7.76E5 Nd-147 4.47E2 3.62E2 2.80E1 ND 1.99E2 ND 5.73E5 OFF-SITE DOSE CALCULATION MANUAL Page 120

TABLE 4.4-8 Ingestion Dose Factors (K.) I Grass-Cow-Milk Pathway (Teen)

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 9.94E2 9.94E2 9.94E2 9.94E2 9.94E2 9.94E2 9.94E2 Cr-51 ND ND 5.00E4 2.78E4 1.09E4 7.13E4 8.40E6 Mn-54 ND 1.40E7 2.78E6 ND 4.18E6 ND 2.87E7 Fe-55 4.45E7 3.16E7 7.36E6 ND ND 2.00E7 1.37E7 Fe-59 5.21E7 1.22E8 4.70E7 ND ND 3.87E7 2.88E8 Co-58 ND 7.95E6 1.83E7 ND ND ND 1.10E8 Co-60 ND 1.64E6 3.70E6 ND ND ND 3.14E7 Ni-63 1.82E10 8.35E8 4.01E8 ND ND ND 1.33E8 Zn-65 2.11E9 7.32E9 3.41E9 ND 4.68E9 ND 3.1OE9 Rb-86 ND 4.73E9 2.22E9 ND ND ND 6.99E8 Sr-89 2.7OE9 ND 7.73E7 ND ND ND 3.22E8 Sr-90 6.61E10 ND 1.63E10 ND ND ND 1.86E9 Y-91 1.58E4 ND 4.24E2 ND ND ND 6.48E6 Zr-95 1.66E3 5.22E2 3.59E2 ND 7.68E2 ND 1.21E6 Nb-95 1.41E5 7.80E4 4.29E4 ND 7.56E4 ND 3.34E8 Ru-103 1.81E3 ND 7.74E2 ND 6.39E3 ND 1.51E5 Ru-106 3.75E4 ND 4.73E3 ND 7.24E4 ND 1.80E6 Ag-110m 9.64E7 9.12E7 5.55E7 ND 1.74E8 ND 2.56E1O Te-125m 3.00E7 1.08E7 4.02E6 8.39E6 ND ND 8.86E7 Te-127m 8.44E7 2.99E7 1.00E7 2.01E7 3.42E8 ND 2.10E8 Te-129m 1.11E8 4.11E7 1.75E7 3.57E7 4.63E8 ND 4.16E8 1-131 5.38E8 7.53E8 4.05E8 2.20E11 1.30E9 ND 1.49E8 Cs-134 9.81E9 2.31E10 1.07E10 ND 7.34E9 2.80E9 2.87E8 Cs-136 4.59E8 1.80E9 1.21E9 ND 9.82E8 1.55E8 1.45E8 Cs-137 1.34E10 1.78E10 6.20E9 ND 6.06E9 2.35E9 2.53E8 Ba-140 4.87E7 5.96E4 3.14E6 ND 2.02E4 4.01E4 7.51E7 Ce-141 8.89E3 5.93E3 6.81E2 ND 2.79E3 ND 1.70E7 Ce-144 6.58E5 2.72E5 3.54E4 ND 1.63E5 ND 1.65E8 Pr-143 2.89E2 1.15E2 1.44E1 ND 6.73E1 ND 9.53E5 Nd-147 1.82E2 1.98E2 1.19E1 ND 1.16E2 ND 7.15E5 OFF-SITE DOSE CALCULATION MANUAL Page 121

TABLE 4.4-9 Ingestion Dose Factors (R) I Grass-Cow-Milk Pathway (Adult)

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 2.63E2 7.63E2 7.63E2 7.63E2 7.63E2 7.63E2 7.63E2 Cr-51 ND ND 2.86E4 1.71E4 6.27E3 3.80E4 7.20E6 Mn-54 ND 8.40E6 1.60E6 ND 2.50E6 ND 2.57E7 Fe-55 2.51E7 1.73E7 4.04E6 ND ND 9.67E6 9.95E6 Fe-59 2.99E7 7.02E7 2.69E7 ND ND 1.96E7 2.34E8 Co-58 ND 4.72E6 1.06E7 ND ND ND 9.51E7 Co-60 ND 1.64E7 3.62E7 ND ND ND 3.08E8 Ni-63 6.73E9 4.66E8 2.27E8 ND ND ND 9.73E7 Zn-65 1.37E9 4.37E9 1.97E9 ND 2.92E9 ND 2.75E9 Rb-86 ND 2.59E9 1.21E9 ND ND ND 5.11E8 Sr-89 1.47E9 ND 4.21E7 ND ND ND 2.35E8 Sr-90 4.69E10 ND 1.15E10 ND ND ND 1.35E9 Y-91 8.60E3 ND 2.29E2 ND ND ND 4.73E6 Zr-95 1.06E3 3.04E2 2.06E2 ND 4.77E2 ND 9.63E5 Nb-95 5.65E5 2.44E5 9.59E3 ND 2.43E5 ND 1.95E9 Ru-103 1.02E3 ND 4.39E2 ND 3.89E3 ND 1.19E5 Ru-106 2.04E4 ND 2.58E3 ND 3.94E4 ND 1.32E6 Ag-110m 5.83E7 5.39E7 3.20E7 ND 1.06E8 ND 2.20E10 Te-125m 1.63E7 5.90E6 2.18E6 4.90E6 6.63E7 ND 6.50E7 Te-127m 4.58E7 1.64E7 5.58E6 1.17E7 1.86E8 ND 1.54E8 Te-129m 6.05E7 2.26E7 9.58E6 2.08E7 2.53E8 ND 3.05E8 1-131 2.97E8 4.24E8 2.43E8 1.39E11 7.27E8 ND 1.12E8 Cs-134 5.65E9 1.34E10 1.10E10 ND 4.33E9 1.44E9 2.35E8 Cs-136 2.69E8 1.06E9 7.65E8 ND 5.92E8 8.11E7 1.21E8 Cs-137 7.38E9 1.01E10 6.61E9 ND 3.43E9 1.14E9 1.95E8 Ba-140 2.70E7 3.39E4 1.77E6 ND 1.15E4 1.94E4 5.55E7 Ce-141 4.85E3 3.28E3 3.72E2 ND 1.52E3 ND 1.25E7 Ce-144 3.58E5 1.50E5 1.92E4 ND 8.87E4 ND 1.21E8 Pr-143 1.94E2 7.79E1 9.62E0 ND 4.49E1 ND 8.50E5 Nd-147 9.49E1 1.10E2 6.56E0 ND 6.41E1 ND 5.26E5 OFF-SITE DOSE CALCULATION MANUAL Page 122

Calculation of Ingestion Dose Factor Grass-Cow-Meat Pathway Rm [D/Q] = K'LQF[UIp]1Ff(rXDFLý)[

LAll,w LYpffj + [ - fPf Yý je-A'f where: Unit = m'.mrem/yr per pCi/sec Reference Table R. G. 1.109 K' = A constant of unit conversion 106 pCi/uCi.

QF = The cow's consumption rate, 50 kg/day (wet weight) E-3 Uap = The receptor's meat consumption rate for age (a), E-5 in kg/yr Infant - 0 Teen - 65 Child - 41 Adult -110 Yp = The agricultural productivity by unit area of pasture E-15 feed grass 0.7 kg/m 2 Ys = The agricultural productivity of unit area of E-15 stored feed 2.0 kg/m 2 Fr = The stable element transfer coefficients, in days/kg. E-1 r = Fraction of deposited activity retained on cow's E-15 feed grass 1.0 radioiodine 0.2 particulates tf = Transport time from pasture to receptor, in sec. E-15 1.73x106 sec (20 days)

= Transport time from crop field to receptor, in sec. E-15 7.78x10 6 sec. (90 days)

(DFLi)a = The maximum organ ingestion dose factor for the ith E-11 to radionuclide for the receptor in age group (a), E-14 in mrem/pCi A[ = The decay constant for the ith radionuclide, in sec

,, = The decay constant for removal of activity on E-15 leaf and plant surfaces by weathering, 5.73 x 10-7 sec -1 (corresponding to a 14 day half-life).

fp Fraction of the year that the cow is on pasture (dimensionless) = 1".

fs = Fraction of the cow feed that is pasture grass while the cow is on pasture (dimensionless) = 1.

OFF-SITE DOSE CALCULATION MANUAL Page 123

Milk cattle are considered to be fed from two potential sources, pasture grass and stored feeds. Following the development in Regulatory Guide 1.109, the values of fp and fs will be considered unity, in lieu of site specific information provided in the annual land census report by the licensee.

Note: The above equation does not apply to the concentration of tritium in meat. A separate equation is provided in NUREG 0133, section 5.3.1.4 to determine Tritium value.

Reference:

The equation deriving R"' (D/Q) was taken from NUREG 0133, Section 5.3.1.4.

tf in NUREG 0133 is equivalent to ts in R.G. 1.109 Table E-15.

OFF-SITE DOSE CALCULATION MANUAL Page 124

TABLE 4.4-10 Ingestion Dose Factors (P7)

Grass-Cow-Meat Pathway (Child)

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 2. 34E2 2. 34E2 2. 34E2 2 . 34E2 2. 34E2 2.34E2 2. 34E2 Cr-51 ND ND 8.82E3 4.89E3 1. 34E3 8.93E3 4.68E5 Mn-54 ND 7.99E6 2. 13E6 ND 2.24E6 ND 6.70E6 Fe-55 4. 57E8 2.42E8 7.50E7 ND ND 1.37E8 4.49E7 Fe-59 3.81E8 6.16E8 3.07E8 ND ND 1.79E8 6.42E8 Co-58 ND 1. 65E7 5.04E7 ND ND ND 9 .60E7 Co-60 ND 6.93E7 2.04E8 ND ND ND 3. 84E8 Ni -63 2.91E10 1. 56E9 9.91E8 ND ND ND 1.05E8 Zn-65 3.76E8 1.00E9 6.22E8 ND 6.30E8 ND 1. 76E8 Rb-86 ND 5.77E8 3.55E8 ND ND ND 3. 71E7 Sr-89 4.92E8 ND 1.40E7 ND ND ND 1.90E7 Sr-90 1.04E10 ND 2. 64E9 ND ND ND 1.40E8 Y-91 1.81E6 ND 4.83E4 ND ND ND 2.41E8 Zr-95 2.69E6 5.91E5 5.26E5 ND 8.46E5 ND 6. 16E8 Nb-95 3.09E6 1.20E6 8.61E5 ND 1. 13E6 ND 2 .23E9 Ru-103 1. 55E8 ND 5.97E7 ND 3.91E8 ND 4. 02E9 Ru-106 4.44E9 ND 5 54E8 ND 5.99E9 ND 6.90E10, Ag-110m 8.41E6 5.68E6 4. 54E6 ND 1.06E7 ND 6. 76E8 Te-125m 5.69E8 1. 54E8 7. 59E7 1.60E8 ND ND 5.49E8 Te-127m 1.77E9 4.78E8 2. 11E8 4.24E8 5.06E9 ND 1.44E9 Te-129m 4.78E9 5.05E8 2.81E8 5.83E8 5.31E9 ND 2. 21E9 1-131 1.66E7 1.67E7 9.49E6 5.52E9 2. 74E7 ND 1.49E6 Cs-134 9.22E8 1. 51E9 3.19E8 ND 4. 69E8 1.68E8 8.16E6 Cs-136 1.73E7 4.74E7 3.07E7 ND 2 .53E7 3.77E6 1.67E6 Cs-137 1. 33E9 1.28E9 1.88E8 ND 4. 16E8 1. 50E8 7.99E6 Ba- 140 4. 39E7 3.85E4 2. 56E6 ND 1.25E4 2.29E4 2.22E7 Ce-141 2.22E4 1.11E4 1.64E3 ND 4.86E3 ND 1.38E7 Ce-144 2. 32E6 7.26E5 1.24E5 ND 4.02E5 ND 1.89E8 Pr-143 3.35E4 1.01E4 1.66E3 ND 5 .45E3 ND 3.61E7, Nd-147 1.18E4 9.60E3 7.43E2 ND 5 .27E3 ND 1.52E7 OFF-SITE DOSE CALCULATION MANUAL Page 125

TABLE 4.4-11 Ingestion Dose Factors (Ru)

Grass-Cow-Meat Pathway (Teen)

Nuclide Bone Liver T. Body Thyroid Kidney Lung_ GI-LLI H-3 1.94E2 1.94E2 1.94E2 1.94E2 1.94E2 1.94E2 1.94E2 Cr-51 ND ND 5.65E3 3.14E3 1.24E3 8.07E3 9.49E5 Mn-54 ND 6.98E6 1. 39E6 ND 2.08E6 ND 1.43E7 Fe-55 2. 38E8 1. 69E8 3.93E7 ND ND 1.07E8 7.30E7 Fe-59 2.15E8 5. 01E8 1.94E8 ND ND 1. 58E8 1.19E9 Co-58 ND 1. 41E7 3.25E7 ND ND ND 1.94E8 Co-60 ND 5. 83E7 1. 31E8 ND ND ND 7.60E8 Ni -63 1. 52E10 1. 07E9 5.15E8 ND ND ND 1.71E8 Zn-65 2. 50E8 8. 69E8 4.06E8 ND 5.56E8 ND 3.68E8 Rb-86 ND 4.06E8 1.91E8 ND ND ND 6.01E7 Sr-89 2.60E8 ND 7.44E6 ND ND ND 3.09E7 Sr-90 8:05E9 ND 1.99E9 ND ND ND 2.26E8 Y-91 9.56E5 ND 2. 56E4 ND ND ND 3.92E8 Zr-95 1. 51E6 4.78E5 3.28E5 ND 7.02E5 ND 1.10E9 Nb-95 1.79E6 9.93E5 5.47E5 ND 9.63E5 ND 4.25E9 Ru-103 8.58E7 ND 3.67E7 ND 3.03E8 ND 7. 17E9 Ru-106 2. 36E9 ND 2.97E8 ND 4.55E9 ND 1.13E1l Ag- 110m 5.07E6 4.80E6 2.92E6 ND 9.15E6 ND 1.35E9 Te-125m 3.03E8 1.09E8 4.05E7 8.47E7 ND ND 8.94E8 Te-127m 9.42E8 3.34E8 1. 12E8 2.24E8 3.82E9 ND 2. 35E9 Te-129m 9.61E8 3.57E8 1. 52E8 3.10E8 4.02E9 ND 3.61E9 1-131 8.97E6 1. 26E7 6.75E6 3.66E9 2.16E7 ND 2.48E6 Cs-134 5.23E8 1.23E9 5. 71E8 ND 3.91E8 1.49E8 1. 53E7 Cs-136 9.96E6 3.92E7 2.63E7 ND 2. 13E7 3.36E6 3.15E6 Cs-137 7. 24E8 9.63E8 3. 36E8 ND 3.28E8 1.27E8 1. 37E7 Ba-140 2.39E7 2.93E4 1. 54E6 ND 9.94E3 1. 97E4 3.69E7 Ce-141 1.18E4 7.88E3 9.05E2 ND 3.71E3 ND 2.25E7 Ce-144 1.23E6 5.08E5 6. 60E4 ND 3.04E5 ND 3.09E8 Pr-143 1.76E4 7.03E3 8. 76E2 ND 4.09E3 ND 5.79E7 Nd-147 6.32E3 6.87E3 4. 12E2 ND 4.04E3 ND 2.48E7 OFF-SITE DOSE CALCULATION MANUAL Page 126

TABLE 4.4-12 Ingestion Dose Factors (g') I Grass-Cow-Meat Pathway (Adult)

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 3.25E2 3.25E2 3.25E2 3.25E2 3.25E2 3.25E2 3.25E2 Cr-51 ND ND 7.06E3 4.22E3 1.56E3 9.37E3 1.78E6 Mn-54 ND 9.16E6 1.75E6 ND 2.72E6 ND 2.80E7 Fe-55 2.93E8 2.02E8 4.72E7 ND ND 1.13E8 1.16E8 Fe-59 2.69E8 6.32E8 2.42E8 ND ND 1.76E8 2.11E9 Co-58 ND 1.83E7 4.10E7 ND ND ND 3.70E8 Co-60 ND 7.52E7 1.66E8 ND ND ND 1.41E9 Ni-63 1.89E10 1.31E9 6.33E8 ND ND ND 2.73E8 Zn-65 3.56E8 1.13E9 5.12E8 ND 7.58E8 ND 7.13E8 Rb-86 ND 4.86E8 2.27E8 ND ND ND 9.59E7 Sr-89 3.08E8 ND 8.83E6 ND ND ND 4.93E7 Sr-90 1.24E10 ND 3.05E9 ND ND ND 3.59E8 Y-91 1.13E6 ND 3.03E4 ND ND ND 6.24E8 Zr-95 1.89E6 6.06E5 4.10E5 ND 9.51E5 ND 1.92E9 Nb-95 2.29E6 1.28E6 6.86E5 ND 1.26E6 ND 7.74E9 Ru-103 1.05E8 ND 4.54E7 ND 4.02E8 ND 1.23E10 Ru-106 2.80E9 ND 3.54E8 ND 5.40E9 ND 1.81E11 Ag-110m 6.70E6 6.19E6 3.69E6 ND 1.22E7 ND 2.53E9 Te-125m 3.59E8 1.30E8 4.81E7 1.08E8 1.46E9 ND 1.43E9 Te-127m 1.12E9 3.99E8 1.36E8 2.85E8 4.53E9 ND 3.74E9 Te-129m 1.15E9 4.28E8 1.82E8 3.94E8 4.79E9 ND 5.78E9 1-131 1.08E7 1.54E7 8.85E6 5.06E9 2.65E7 ND 4.07E6 Cs-134 6.57E8 1.56E9 1.29E9 ND 5.06E8 1.68E8 2.74E7 Cs-136 1.28E7 5.04E7 3.63E7 ND 2.80E7 3.84E6 5.73E6 Cs-137 8.72E8 1.19E9 7.81E8 ND 4.05E8 1.35E8 2.31E7 Ba-140 2.90E7 3.64E4 1.90E6 ND 1.24E4 2.08E4 5.96E7 Ce-141 1.41E4 9.51E3 1.08E3 ND 4.41E3 ND 3.63E7 Ce-144 1.46E6 6.09E5 7.82E4 ND 3.61E5 ND 4.93E8 Pr-143 2.09E4 8.39E3 1.04E3 ND 4.85E3 ND 9.17E7 Nd-147 7.17E3 8.29E3 4.96E2 ND 4.85E3 ND 3.99E7 OFF-SITE DOSE CALCULATION MANUAL Page 127

Calculation of Ingestion Dose Factor Vegetation Pathway

[ ] K=LYv(/Ii,r +q (DFLi)aULa fL e-/itL+Ujsafgelt R//Q]

Rv[D where: Units = m2-mrem/yr per uCi/sec. Reference Table, R.G. 1.109 K' = A constant of unit conversion, 106 pCi/uCi.

L = The consumption rate of fresh leafy vegetation by the E-5 Ua receptor in age group (a), in kg/yr.

Infant 0 Child 26 Teen 42 Adult 64 u= The consumption rate of stored vegetation by the E-5 receptor in age group (a), in kg/yr Infant 0 Child 520 Teen 630 Adult 520 (DFLi)a = The maximum organ ingesting dose factor for the ith E-11 to E-14 radionuclide for the receptor in age group (a),

in mrem/pCi.

fi- = The fraction of the annual intake of fresh leafy E-15 vegetation grown locally. (default 1.0) f9 = The fraction of the annual intake of stored vegetation E-15 grown locally. (default 0.76) tL = The average time between harvest of leafy vegetation E-15 and its consumption, 8.6 x 10'seconds (1 day) th = The average time between harvest of stored vegetation E-15 and its consumption, 5.18 x 106seconds (60 days)

Yv = The vegetation areal density, 2.0 kg/mi E-15

= Fraction of deposited activity retained on the E-15 vegetation 1.0 radioiodine 0.2 particulates 2A = The decay constant for the ith radionuclide, in sec'-

AW = The decay constant for removal of activity on leaf and E-15 plant surfaces by weathering, 5.73 x 10' sec-'

(corresponding to a 14 day half-life).

OFF-SITE DOSE CALCULATION MANUAL Page 128

Note: The above equation does not apply to the concentrations of tritium in vegetation. A separate equation is provided in NUREG 0133, section 5.3.1.5 to determine tritium values.

Reference:

The equation deriving R' (D/Q) was taken from NUREG 0133, Section 5.3.1.5.

OFF-SITE DOSE CALCULATION MANUAL Page 129

TABLE 4.4-13 Ingestion Dose Factors (Rv)

Vegetation Pathway (Child)

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 4.01E3 4.01E3 4.01E3 4.01E3 4.01E3 4.01E3 4.01E3 Cr-51 ND ND 1.18E5 6.54E4 1.79E4 1.19E5 6.25E6 Mn-54 ND 6.61E8 1.76E8 ND 1.85E8 ND 5.55E8 Fe-55 8.00E8 4.24E8 1.31E8 ND ND 2.40E8 7.86E7 Fe-59 4.07E8' 6.58E8 3.28E8 ND ND 1.91E8 6.85E8 Co-58 ND 6.47E7 1.98E8 ND ND ND 3.77E8 Co-60 ND 3.78E8 1.12E9 ND ND ND 2.10E9 Ni-63 3.95E10 2.11E9 1.34E9 ND ND ND 1.42E8 Zn-65 8.13E8 2.17E9 1.35E9 ND 1.36E9 ND 3.80E8 Rb-86 ND 4.52E8 2.78E8 ND ND ND 2.91E7 Sr-89 3.74E10 ND 1.07E9 ND ND ND 1.45E9 Sr-90 1.24E12 ND 3.15E11 ND ND ND 1.67E10 Y-91 1.87E7 ND 5.01E5 ND ND ND 2.49E9 Zr-95 3.92E6 8.63E5 7.68E5 ND 1.23E6 ND 9.00E8 Nb-95 4.10E5 1.60E5 1.14E5 ND 1.50E5 ND 2.95E8 Ru-103 1.54E7 ND 5.92E6 ND 3.88E7 ND 3.98E8 Ru-106 7.45E8 ND 9.30E7 ND 1.01E9 ND 1.16E10 Ag-110m 3.23E7 2.18E7 1.74E7 ND 4.06E7 ND 2.59E9 Te-125m 3.51E8 9.50E7 4.67E7 9.84E7 ND ND 3.38E8 Te-127m 1.32E9 3.56E8 1.57E8 3.16E8 1.94E9 ND 1.07E9 Te-129m 8.58E8 2.40E8 1.33E8 2.77E8 2.52E9 ND 1.05E9 1-131 1.43E8 1.44E8 8.18E7 4.76E10 2.36E8 ND 1.28E7 Cs-134 1.60E10 2.63E10 5.55E9 ND 8.15E9 2.92E9 1.42E8 Cs-136 4.44E8 1.22E9 7.90E8 ND 6.50E8 9.69E7 4.29E7 Cs-137 2.39E10 2.29E10 3.38E9 ND 7.46E9 2.68E9 1.43E8 Ba-140 2.77E8 2.43E5 1.62E7 ND 7.91E4 1.45E5 1.40E8 Ce-141 6.56E5 3.27E5 4.86E4 ND 1.43E5 ND 4.08E8 Ce-144 1.27E8 3.98E7 6.78E6 ND 2.21E7 ND 1.04E10 Pr-143 1.46E5 4.39E4 7.26E3 ND 2.38E4 ND 1.58E8 Nd-147 7.23E4 5.86E4 4.54E3 ND 5.47E1 ND 9.28E7 OFF-SITE DOSE CALCULATION MANUAL Page 130

TABLE 4.4-14 Ingestion Dose Factors (Rý)

Vegetation Pathway (Teen)

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 4.10E3 4.10E3 4.10E3 4.10E3 4.10E3 4.10E3 4.10E3 P-32 1.60E9 9.91E7 6.20E7 ND ND ND 1.34E8 Cr-51 ND ND 6.19E4 3.44E4 1.36E4 8.84E4 1.04E7 Mn-54 ND 4.52E8 8.97E7 ND 1.35E8 ND 9.27E8 Fe-55 3.25E8 2.31E8 _5.38E7 ND ND 1.46E8 9.98E7 Fe-59 1.83E8 4.28E8 1.65E8 ND ND 1.35E8 1.01E9 Co-58 ND 4.38E7 1.01E8 ND ND ND 6.04E8 Co-60 ND 2.49E8 5.60E8 ND ND ND 3.24E9 Ni-63 1.61E10 1.13E9 5.44E8 ND ND ND 1.81E8 Zn-65 4.24E8 1.47E9 6.87E8 ND 9.43E8 ND 6.24E8 Rb-86 ND 2.73E8 1.28E8 ND ND ND 4.04E7 Sr-89 1.57E10 ND 4.50E8 ND ND ND 1.87E9 Sr-90 7.51E11 ND 1.85E11 ND ND ND 2.11E10 Y-91 7.87E6 ND 2.11E5 ND ND ND 3.23E9 Zr-95 1.75E6 5.52E5 3.80E5 ND 8.12E5 ND 1.27E9 Nb-95 1.92E5 1.06E5 5.85E4 ND 1.03E5 ND 4.54E8 Ru-103 6.85E6 ND 2.93E6 ND 2.41E7 ND 5.72E8 Ru-106 3.09E8 ND 3.90E7 ND 5.97E8 ND 1.48E10 Ag-110m 1.52E7 1.44E7 8.76E6 ND 2.75E7 ND 4.04E9 Te-125m 1.48E8 5.34E7 1.98E7 4.14E7 ND ND 4.37E8 Te-127m 5.52E8 1.96E8 6.56E7 1.31E8 2.24E9 ND 1.37E9 Te-129m 3.69E8 1.37E8 5.84E7 1.19E8 1.54E9 ND 1.39E9 1-131 7.70E7 1.08E8 5.79E7 3.15E10 1.86E8 ND 2.13E7 Cs-134 7.10E9 1.67E10 7.75E9 ND 5.31E9 2.03E9 2.08E8 Cs-136 4.65E7 1.83E8 1.23E8 ND 9.96E7 1.57E7 1.47E7 Cs-137 1.01E10 1.35E10 4.69E9 ND 4.59E9 1.78E9 1.92E8 Ba-140 1.39E8 1.71E5 8.97E6 ND 5.78E4 1.15E5 2.15E8 Ce-141 2.83E5 1.89E5 2.17E4 ND 8.90E4 ND 5.41E8 Ce-144 5.27E7 2.18E7 2.82E6 ND 1.30E7 ND 1.33E10 Pr-143 6.99E4 2.79E4 3.48E3 ND 1.62E4 ND 2.30E8 Nd-147 3.66E4 3.98E4 2.39E3 ND 2.34E4 ND 1.44E8 OFF-SITE DOSE CALCULATION MANUAL Page 131

TABLE 4.4-15 Ingestion Dose Factors (Rýv)

Vegetation Pathway (Adult)

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 5.11E3 5.11E3 5.11E3 5.11E3 5.11E3 5.11E3 5.11E3 Cr-51 ND ND 4.66E4 2.79E4 1.03E4 6.18E4 1.17E7 Mn-54 ND 3.11E8 5.94E7 ND 9.27E7 ND 9.54E8 Fe-55 2.09E8 1.45E8 3.37E7 ND ND 8.06E7 8.29E7 Fe-59 1.29E8 3.02E8 1.16E8 ND ND 8.45E7 1.01E9 Co-58 ND 3.09E7 6.92E7 ND ND ND 6.26E8 Co-60 ND 1.67E8 3.69E8 ND ND ND 3.14E9 Ni-63 1.04E10 7.21E8 3.49E8 ND ND ND 1.50E8 Zn-65 3.18E8 1.01E9 4.57E8 ND 6.76E8 ND 6.37E8 Rb-86 ND 2.19E8 1.02E8 ND ND ND 4.32E7 Sr-89 1.03E10 ND 2.96E8 ND ND ND 1.65E9 Sr-90 6.05E11 ND 1.48E11 ND ND ND 1.75E10 Y-91 5.13E6 ND 1.37E5 ND ND ND 2.82E9 Zr-95 1.19E6 3.83E5 2.59E5 ND 6.00E5 ND 1.21E9 Nb-95 1.42E5 7.90E4 4.24E4 ND 7.81E4 ND 4.79E8 Ru-103 4.79E6 ND 2.06E6 ND 1.83E7 ND 5.59E8 Ru-106 1.93E8 ND 2.44E7 ND 3.72E8 ND 1.25E10 Ag-110m 1.06E7 9.78E6 5.81E6 ND 1.92E7 ND 3.99E9 Te-125m 9.66E7 3.50E7 1.29E7 2.90E7 3.93E8 ND 3.86E8 Te-127m 3.49E8 1.25E8 4.26E7 8.93E7 1.42E9 ND 1.17E9 Te-129m 2.56E8 9.55E7 4.05E7 8.79E7 1.07E9 ND 1.29E9 1-131 8.09E7 1.16E8 6.63E7 3.79E10 1.98E8 ND 3.05E7 Cs-134 4.66E9 1.11E10 9.07E9 ND 3.59E9 1.19E9 1.94E8 Cs-136 4.47E7 1.77E8 1.27E8 ND 9.82E7 1.35E7 2.01E7 Cs-137 6.36E9 8.70E9 5.70E9 ND 2.95E9 9.81E8 1.68E8 Ba-140 1.29E8 1.62E5 8.47E6 ND 5.52E4 9.29E4 2.66E8 Ce-141 1.97E5 1.33E5 1.51E4 ND 6.20E4 ND 5.10E8 Ce-144 3.29E7 1.37E7 1.77E6 ND 8.15E6 ND 1.11E10 Pr-143 6.25E4 2.51E4 3.10E3 ND 1.45E4 ND 2.74E8 Nd-147 3.36E4 3.89E4 2.33E3 ND 2.27E4 ND 1.87E8 OFF-SITE DOSE CALCULATION MANUAL Page 132

Calculation of Dose Factors in the Ground Plane Pathway (RG [D/Q])

R. [D/Q]=K'K"(SF)(DFG)[ I-edit) units = m2 mrem/yr per uCi/sec where: Reference TableR.G.1.109 K' A constant unit of conversion, 106 pCi/iCi.

K" A constant unit of conversion, 8760 hr/yr SF = The shielding factor, 0.7(dimensionless) E-15 I = The .decay constant for the ith radionuclide, sec-1 t = The exposure period, 4.73 x 108 sec (15 years)

DFGi The ground plane dose conversion factor for the ith radionuclide (mrem/hr per pCi/m 2 ) E-6

Reference:

The equation deriving RI [D/Q] was taken from NUREG 0133, Section 5.3.1.2.

OFF-SITE DOSE CALCULATION MANUAL Page 133

Table 4.4-16 Dose Factors Ground Plane Pathway (Rý [D/Q])

T. Body Skin Cr-51 4.65E6 5. 5E6 Mn-54 1. 39E9 1.63E9 Fe-55 0 0 Fe-59 2.73E8 3.21E8 Co-58 3.79E8 4.44E8 Co-60 2. 15E10 2. 53E10 Ni-63 0 0 Zn-65 7.47E8 8.57E8 Rb-86 8.98E6 1.02E7 Sr-89 2. 17E4 2.52E4 Y-91 1.07E6 1.21E6 Zr-95 2.45E8 2.84E8 Nb-95 1.41E7 1.66E7 Ru-106 4.22E8 5.07E8 Ag-110m 3.44E9 4.02E9 Te-125m 1.55E6 2.13E6 Te-127m 9. 17E4 1.08E5 Te-129m 1.98E7 2. 31E7 1-131 1.72E7 2.08E7 Cs-134 6.85E9 8.0E9 Cs-136 1.51E8 1.72E8 Cs-137 1.03E10 1.20E10 Ba-140 2.06E7 2 . 35E7 Ce-141 1.37E7 1. 54E7 Ce-144 6.95E7 8.05E7 Pr-143 0 0 Nd-147 8.40E6 1.01E7 Units are m2.mrem/yr per pCi/sec OFF-SITE DOSE CALCULATION MANUAL Page 134

CALCULATION OF LIQUID EFFLUENT ADULT INGESTION DOSE FACTORS Air = 1.14E5 (21BFi+5B1i)DFi Ai = Composite dose parameter for the total body or critical organ of an adult for nuclide i, for all appropriate pathways, mrem/hr per pi/ml 1.14E5 = units conversion factor, 10 6pci/pci x 10ml/kg + 8760 hr/yr BF, = Bioaccumulation factor for nuclide i, in fish, pCi/kg per pCi/L, from Table A-1 of Regulatory Guide 1.109 (Rev. 1) or Table A-8 of Regulatory Guide 1.109 (original draft).

BI = Bioaccumulation factor for nuclide i, in invertebrates, pCi/kg per pCi/L, from Table A-1 of Regulatory Guide 1.109 (Rev. 1) or Table A-8 of Regulatory Guide 1.109 (original draft).

DFi = Dose conversion factor for nuclide i, for adults in pre-selected organ T, in mrem/pCi, from Table E-11 or Regulatory Guide 1.109 (Rev. 1) or Table A-3 of Regulatory Guide 1.109 (original draft).

Reference:

The equation for Saltwater sites from NUREG 0133, Section 4.3.1, where Uw/Dw = 0 since no drinking water pathway exists.

OFF-SITE DOSE CALCULATION MANUAL Page 135

Table 4.4-17 Liquid Effluent - Adult Ingestion Dose Factors (AiT) I Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 2.82E-1 2.82E-1 2.82E-1 2.82E-1 2.82E-1 2.82E-1 2.82E-1 Na-24 4.57E-1 4.57E-1 4.57E-1 4.57E-1 4.57E-1 4.57E-1 4.57E-1 Cr-51 ND ND 5.58E0 3.34E0 1.23E0 7.40E0 1.40E3 Mn-54 ND 7.06E3 1.35E3 ND 2.10E3 ND 2.16E4 Mn-56 ND 1.78E2 3.15E1 ND 2.26E2 ND 5.67E3 Fe-55 5.11E4 3.53E4 8.23E3 ND ND 1.97E4 2.03E4 Fe-59 8.06E4 1.90E5 7.27E4 ND ND 5.30E4 6.32E5 Co-58 ND 6.03E2 1.35E3 ND ND ND 1.22E4 Co-60 ND 1.73E3 3.82E3 ND ND ND 3.25E4 Ni-63 4.96E4 3.44E3 1.67E3 ND ND ND 7.18E2 Ni-65 2.02E2 3.31E1 1.20E1 ND ND ND 6.65E2 Cu-64 ND 2.14E2 1.01E2 ND 5.40E2 ND 1.83E4 Zn-65 1.61E5 5.13E5 2.32E5 ND 3.43E5 ND 3.23E5 Zn-69 3.43E2 6.56E2 4.56E1 ND 4.26E2 ND 9.85E1 Br-83 ND ND 7.25E-2 ND ND ND 1.04E-1 Br-84 ND ND 9.39E-2 ND ND ND 7.37E-7 Br-85 ND ND 3.86E-3 ND ND ND LE-18 Rb-86 ND 6.24E2 2.91E2 ND ND ND 1.23E2 Rb-88 ND 1.79E0 9.49E-1 ND ND ND 2.47E-11 Rb-89 ND 1.19E0 8.34E-1 ND ND ND 6.89E-14 Sr-89 4.99E3 ND 1.43E2 ND ND ND 8.00E2 Sr-90 1.23E5 ND 3.01E4 ND ND ND 3.55E3 Sr-91 9.18E1 ND 3.71E0 ND ND ND 4.37E2 Sr-92 3.48E1 ND 1.51E0 ND ND ND 6.90E2 Y-90 6.06E0 ND 1.63E-1 ND ND ND 6.42E4 Y-91m 5.73E-2 ND 2.22E-3 ND ND ND 1.68E-1 Y-91 8.88E1 ND 2.37E0 ND ND ND 4.89E4 Y-92 5.32E-1 ND 1.56E-2 ND ND ND 9.32E3 Y-93 1.69E0 ND 4.66E-2 ND ND ND 5.35E4 Zr-95 1.59E1 5.11EO 3.46E0 ND 8.02E0 ND 1.62E4 Zr-97 8.81E-1 1.78E-1 8.13E-2 ND 2.68E-1 ND 5.51E4 OFF-SITE DOSE CALCULATION MANUAL Page 136

Table 4.4-17 Liquid Effluent - Adult Ingestion Dose Factors (Ai,)

Nuclide Bone Liver T. Body Thyroid Kidney Lung_ GI-LLI Nb-95 4.47E2 2.49E2 1.34E2 ND 2.46E2 ND 1. 51E6 Mo-99 ND 9.05E-4 1.72E-4 ND 2.05E-3 ND 2. 1OE-3 Tc-99m 1. 30E-2 3.66E-2 4.66E-1 ND 5. 56E-1 1. 79E-2 2. 17E1 Tc-101 1.33E-2 1.92E-2 1.88E-1 ND 3.46E-1 9.81E-3 5.77E-14 Ru-103 1.07E2 ND 4.60E1 ND 4.07E2 ND 1.2 5E4 Ru-105 8.89E0 ND 3.51E0 ND 1.15E2 ND 5.44E3 Ru-106 1. 59E3 ND 2.01E2 ND 3.06E3 ND 1.03E5 Ag-110m 1. 57E3 1.45E3 1.33E1 ND 2.85E3 ND 5.91E5 Sb-124 2.77E2 5.23E0 1.09E2 6.70E1 ND 2 . 15E2 7.83E3 Sb-125. 2 .20E2 2.37E0 4.42E1 1.95E1 ND 2. 30E4 1.94E4 Sb-126 1.13E2 2.31E0 4.09E1 6.95E1 ND 6.95E1 9.27E3 Te-125m 2 .17E2 7.86E1 2.91E1 6.52E1 8.82E2 ND 8.66E2 Te-127m 5.48E2 1.96E2 6.68E1 1.40E2 2.23E3 ND 1.84E3 Te-127 8.90EO 3.20E0 1.93E0 6.60E0 3.63E1 ND 7.03E2 Te-129m 9. 31E2 3.47E2 1.47E2 3.20E2 3.89E3 ND 4.69E3 Te-129 2. 54E0 9.55E-1 6.19E-1 1.95E0 1.07E1 ND 1. 92 EQ Te-131m 1.40E2 6.85E1 5. 71E1 1.08E2 6.94E2 ND 6.80E3 Te-131 1. 59E0 6.66E-1 5.03E-1 1.31E0 6.99E0 ND 2.26E-1 Te-132 2.04E2 1. 32E2 1.24E2 1.46E2 1.27E3 ND 6.24E3 1-130 3.96E1 1.17E2 4.61E1 9.91E3 1. 82 E2 ND 1.01E2 1-131 2. 18E2 3.12E2 1.79E2 1.02E5 5. 35E2 ND 8.23E1 1-132 1.06E1 2.85E1 9.96E0 9.96E2 4.54E1 ND 5. 35E0 1-133 7. 54E1 1.30E2 3.95E1 1.90E4 2 . 26E2 ND 1. 16E2 1-134 5.56E0 1. 51EI 5.40E0 2. 62E2 2. 40E1 ND 1.32E-2 1-135 2 . 32E1 6.08E1 2. 24E1 4.01E3 9.75E1 ND 6.87E1 Cs-134 6.84E3 1.63E4 1. 33E4 ND 5. 27E3 1.75E3 2.85E2 Cs-136 7. 16E2 2.83E3 2.04E3 ND 1. 57E3 2.16E2 3.21E2 Cs-137 8.78E3 1.20E4 7.85E3 ND 4.07E3 1.35E3 2 .32E2 Cs-138 6.07EO 1.20E1 5.94E0 ND 8.81E0 8.70E-1 5.12E-5 Ba-139 7.85E0 5.59E-3 2.30E-1 ND 5.23E-3 3.17E-3 1. 39E1 Ba-140 1.64E3 2.06EO 1.08E2 ND 7.02E-1 1.18E0 3.38E3 Ba-141 3.81E0 3.69E-3 1.29E-1 ND 2.68E-3 1.63E-3 1.80E-9 Ba-142 1.72E0 1.77E-3 1.08E-1 ND 1. 50E-3 1.OOE-3 2.43E-18 La-140 1. 57E0 7.94E-1 2.1OE-1 ND ND ND 5.83E4 La-142 8.06E-2 3.67E-2 9.13E-3 ND ND ND 2.68E2 OFF-SITE DOSE CALCULATION MANUAL Page 137

Table 4.4-17 Liquid Effluent - Adult Ingestion Dose Factors (AlT)

Nuclide Bone Liver T. Body Thyroid Kidney Lung_ GI-LLI Ce-141 3.43E0 2.32E0 2.63E-1 ND 1.08EO ND 8.86E3 Ce-143 6.04E-1 4.46E2 4.94E-2 ND 1.97E-1 ND 1.67E4 Ce-144 1.79E2 7.47E1 9.59E0 ND 4.43E1 ND 6.04E4 Pr-143 5.79E0 2.32E0 2.87E-7 ND 1.34E0 ND 2.54E4 Pr-144 1.90E-2 7.87E-3 9.64E-4 ND 4.44E-3 ND 2.73E-9 Nd-147 3.96E0 4.58E0 2.74E-1 ND 2.68E0 ND 2.20E4 W-187 9.16E0 7.66E0 2.68E0 ND ND ND 2.51E3 Np-239 3.53E-2 3.47E-3 1.91E-3 ND 1.08E-2 Nd 7.11E2 OFF-SITE DOSE CALCULATION MANUAL Page 138

SECTION 5.0 ENVIRONMENTAL MONITORING OFF-SITE DOSE CALCULATION MANUAL Page 139

Table 5.1-1 Environmental Radiological Monitoring Station's Locations DIRECTION APPROX. DISTANCE STATION LOCATION FROM PLANT FROM PLANT (mi)

C04 State Park Old Dam on River ENE 10.6 near road intersection C07 Crystal River Public Water Plant ESE 7.4 C09 Fort Island Gulf Beach S 3.2 C10 Indian Waters Public Water Supply ESE 6.0 C13 Mouth of Intake Canal WSW 4.6 C14H Head of Discharge Canal NW 0.1 C14M Midpoint of Discharge Canal W 1.2 C14G Discharge Canal at Gulf of Mexico W 2.5 C18 Yankeetown City Well N 5.3 C19 NW Corner State Roads 488 & 495 ENE 9.6 C29 Discharge Area W 2.0 C30 Intake Area WSW 3.4 C40 Near E. Site Boundary & well pump E 3.6 station CR-South #5 C41 Onsite meteorological tower SW 0.4 C46 North Pump Station N 0.4 C47 Office of Radiation Control, Orlando ESE 78 C48A Onsite North of CR 4 & 5 N 0.4 C48B Onsite NNE of CR 4 & 5 NNE 0.9 NOTE: Distances are approximate. More than one type of sample media(e.g. air and water) are obtained at some stations. For multi-media stations there may be minor difference in distance for each type of sample.

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Table 5.1-2 Environmental Radiological Monitoring Station's Locations-Groundwater Monitoring Wells DIRECTION APPROX. DISTANCE STATION LOCATION FROM PLANT FROM PLANT(mi)

CR3-2 CR-3 Site Perimeter, Just Outside E 0.1 of Protected Area Fence CR3-4 CR-3 Site Perimeter, Just Outside SSE 0.086 of Protected Area Fence CR3-5 CR-3 Site Perimeter, Just Outside SSW 0.051 of Protected Area Fence CR3-6S CR-3 Site Perimeter, Just Outside W 0.038 of Protected Area Fence CR3-6D CR-3 Site Perimeter, Just Outside W 0.038 of Protected Area Fence CR3-7 CR-3 Site Perimeter, Just Outside WNW 0.060 of Protected Area Fence CR3-8 CR-3 Site Perimeter, Just Outside WNW 0.073 of Protected Area Fence CR3-9 CR-3 Site Perimeter, Just Outside NW 0.1 of Protected Area Fence CR3-10 CR-3 Site Perimeter, Just Outside NNE 0.1 of Protected Area Fence The above listed wells have been included in the REMP as a result of information provided in the groundwater flow study completed January 22, 2007 by EnHydro, LLC.

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TABLE 5.1-3 RING TLDs (INNER RING)

LOCATION DIRECTION APPROX. DISTANCE (Mi.)

C27 W 0.4 C60 N 0.9 C61 NNE 0.9 C62 NE 1.2 C63 ENE 0.9 C64 E 0.8 C65 ESE 0.3 C66 SE 0.4 C67 SSE 0.3 C68 S 0.3 C69 SSW 0.3 C41 SW 0.4 C70 WSW 0.7 C71 WNW 0.6 C72 NW 0.3 C73 NNW 0.7 OFF-SITE DOSE CALCULATION MANUAL Page 142

TABLE 5.1-4 RING TLDs (5 MILE RING)

LOCATION DIRECTION APPROX. DISTANCE (Mi.)

C18 N 5.3 C03 NNE 4.9 C04 NE 6.0 C74 ENE 5.1 C75 E 4.0 C76 ESE 5.6 C08 SE 5.7 C77 SSE 3.4 C09 S 3.2 C78 WSW 4.6 C14G W 2.5 Co1 NW 4.8 C79 NNW 5.0 OFF-SITE DOSE CALCULATION MANUAL Page 143

FIGURE S.1 C19 Environmental Monitoring Sample Station Locations OFF-SITE DOSE CALCULATION MANUAL Page 144

FIGURE 5.2 Environmental Monitoring TLD Locations C60 C61

,,,,t, C62

~1uni 6

,.I Ca3, Tower 0,

C68C63 1

C70 C72* _C6 OFF-SITE DOSE CALCULATION MANUAL Page 145

FIGURE 5.3 Environmental Monitoring TLD Locations (5 mile)

OFF-SITE DOSE CALCULATION MANUAL Page 146

FIGURE 5.4 SECTION 6.0 CR3 Groundwa ter Monitoring Well Locations Deep Wells A re Also Installed at #'s 1, 3, 6 OR",

WE'

" F

~_'-

ýj~i

~?

5j ~i II I

[Il

-y Wells # 1 & 3 are not routinely sampled by the REMP OFF-SITE DOSE CALCULATION MANUAL Page 147

6.1 ORIGIN AND PURPOSE OF THE OFFSITE DOSE CALCULATIONAL MANUAL The Offsite Dose Calculation manual was developed to support the implementation of the Radiological Effluent Technical Specifications required by 10 CFR 50, Appendix I, and 10 CFR 50.36. The purpose of the manual is to provide the NRC with sufficient information relative to effluent monitor setpoint calculations, effluent related dose calculations, and environmental monitoring to demonstrate compliance with radiological effluent controls.

6.2 CHANGES The ODCM shall be changed in accordance with Technical Specifications (ref. ITS 5.6.2.3). In addition, interdepartmental reviews shall be performed as appropriate.

6.3 REVIEW The ODCM and its implementation shall be reviewed every 24 months (ref. FSAR 1.7.1.18) 6.4 UNPLANNED RELEASES This section moved to section 1.

6.5 RAnTNArTTVF 1:9:I:1IHMT RF:IFAZF: QF9:PnDT This report is submitted as required by Technical Specification 5.7.1.1.c to Crystal River Facility Operating License No. DPR-72.

The following information is included:

A summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant as outlined in Regulatory Guide 1.21 (Rev. 1, 1974) with data summarized on a quarterly basis following the format of Appendix B thereof.

OFF-SITE DOSE CALCULATION MANUAL Page 148

An annual summary of hourly meteorological data collected over the previous years. (In lieu of submittal, this data is maintained on-site and is available to the NRC upon request.)

A list and description of unplanned releases to unrestricted areas.

Change to the Process Control Program (PCP)

Changes to the Off-Site Dose Calculation Manual (ODCM)

Significant changes to Radioactive Waste Treatment Systems A list of new Environmental Radiological Monitoring Program dose calculation location changes identified by the land-use census.

Information relating to effluent monitors being inoperable for 30 or more days.

A summary of dose estimates to demonstrate compliance with the dose specifications of the ODCM.

NOTE: The results of the effluent report may reflect different quantities, concentrations, and dose estimates than those of the year end summary from the effluent management system. This is based on a year-end data review and correction of release data (e.g.

removing miss-identified isotopes) and accounting for actual site dilution for liquid releases which are difficult to account for in real time.

OFF-SITE DOSE CALCULATION MANUAL Page 149

6.6 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT This report is submitted as required by Technical Specification 5.7.1.1.b to Crystal River Facility Operating License No. DPR-72.

The following information is included:

- Summaries

- Interpretations

- Unachievable LLDs, and

- An analysis of trends of the results of the radiological environmental studies and previous annual reports.

- An assessment of any observed impact of plant operation on the environment.

NOTE: If harmful effects or evidence of irreversible damage are detected by the monitoring, the Report shall provide an analysis of the problem and a planned course of action to alleviate the problem.

- Summarized and tabulated results, in the format of Regulatory Guide 4.8 (December 1975), of all radiological environmental samples taken during the report period.

NOTE: If some results are not available for inclusion, the report shall note and explain the reason for the missing results. The missing results shall be submitted as soon as possible in a supplementary report.

- A summary description of the REMP.

- A map of all sampling locations keyed to a table giving distances and directions from the reactor.

- Unavailability of milk or fresh leafy vegetable samples required by Table 2-7 of Technical Specifications.

- The results of land-use censuses.

- Results of Interlaboratory Comparison Program.

- Results of any unplanned release or spill of radioactive material that could have the potential to contaminate the groundwater as reported to maintain compliance with the groundwater protection initiative.

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ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT Cont'd OBJECTIVE 2.4 from NEI 07-07 Document all on-site ground water sample results and a description of any significant on-site leaks/spills into ground water for each calendar year in the Annual Radiological Environmental Operating Report (AREOR) for REMP or the Annual Radioactive Effluent Release Report (ARERR) for the RETS as contained in the appropriate reporting procedure, beginning with the report for calendar year 2006.

Acceptance Criteria:

a. The appropriate changes to the ODCM/ODAM or to the appropriate procedures were expected to be completed in a timeframe to support the 2007 report of 2006 performance for plants that were operating or decommissioning when the GPI was adopted. For new plants, appropriate procedures that require inclusion of significant on-site leaks/spills into ground water and all on-site ground water results shall be developed and implemented prior to initial receipt of nuclear fuel.

b. Reporting of on-site ground water sample results shall be as follows:

i Ground water sample results that are taken in support of the GPI but are not part of the REMP program (e.g. samples obtained during the investigatory phase of the Action Plan circa year 2006) are reported in the ARERR required by 10 CFR 50.36a (a)(2).

ii Once the long-term monitoring sample points have been established per Objective 1.3, acceptance criterion d, the results are reported in the AREOR for those sample points that are included in the REMP as described in the ODCM/ODAM. The sample results for those long-term monitoring sample points that are not included in REMP are reported in the ARERR.

c. In addition to 2.4.b, voluntary communications shall be included in an annual report as follows:

i A description of all spills or leaks that were communicated per Objective 2.2 acceptance criterion a shall be included in the ARERR.

ii All on-site or off-site ground water sample results that exceeded the REMP reporting thresholds as described in the ODCM/ODAM that were communicated per section 6.7 acceptance criterion b shall be included in either the ARERR and/or in the AREOR.

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6.7 Voluntary Reportinq as a Result of Groundwater Radioactive Contamination OBJECTIVE 2.2 from NEI 07-07 NOTE The following communication protocol only applies to licensed radioactive materials that are generated as a result of plant operations.

NOTE Baseline radioactivity levels resulting from normal plant effluent releases should be evaluated and considered prior to initiating the following communication protocol (e.g., tritium in rain water or clean systems that are made up from water sources that receive licensed effluent releases).

NOTE Communication time requirements for water samples that exceed reporting criteria start following notification of sample results from an applicable vendor or corporate laboratory to the CR-3 Environmental & Chemistry Section.

A. Unplanned or uncontrolled release of radioactive material from the protected area

1. INVESTIGATE and document the event.

2. INITIATE a Nuclear Condition Report for documentation purposes.

B. Industry groundwater protection voluntary communication protocol for spills or leaks that reach or could potentially reach the groundwater

1. DOCUMENT any event meeting the criteria in this section in the plant 10 CFR 50.75 (g) file.

OFF-SITE DOSE CALCULATION MANUAL Page 152

NOTIFICATION REQUIREMENTS (Cont'd)

NOTE The contact position for the State Official is:

Department of Health Bureau of Radiation Control Environmental Administrator The contact position for the Local Officials are:

Citrus County Emergency Management Director AND Levy County Emergency Management Director NOTE The groundwater beneath CR-3 is too saline and brackish to be a source of potable water and will never be used as a source of drinking water.

Make informal* communication as soon as practicable to appropriate State/Local officials, with follow-up notification to the NRC, as appropriate, regarding significant* on-site leaks/spills into ground water and on-site or off-site water sample results exceeding the criteria in the REMP as described in the ODCM/ODAM.

Acceptance Criteria:

This guidance provides a threshold for voluntary communication. Some States may require different communication thresholds; the licensee shall document any agreements with State/Local officials that differs from Industry guidance.

a. Communication to the designated State/Local officials shall be made before the end of the next business day if an inadvertent leak or spill to the environment has or can potentially get into the ground water and exceeds any of the following criteria:

i. If a spill or leak exceeding 100 gallons from a source containing licensed material, ii. If the volume of a spill or leak cannot be quantified but is likely to exceed 100 gallons from a source containing licensed material, or iii. Any leak or spill, regardless of volume or activity, deemed by the licensee to warrant voluntary communication.

To determine whether a leak or spill would trigger voluntary communication, consider the clarification in the following three text boxes in addition to 2.2.a i to iii above:

LEAK OR SPILL: The "leak or spill" represents an inadvertent event or perturbation in a system or component's performance. This event threshold is intended to ensure that State/Local officials are made aware that there has been an event of interest at the site and to keep them apprised of the licensee's action to contain and, as needed, remediate the event. "Leak or spill" events that meet the criteria shall be communicated regardless of whether or not the on-site ground water is, or could be used as, a source of drinking water.

The quantity of liquid resulting from leaks or spills of solid materials or waste or steam leaks should be evaluated with respect to 2.2.a. i to iii, inclusive.

see glossary at end of section OFF-SITE DOSE CALCULATION MANUAL Page 153

SOURCE CONTAINING LICENSED MATERIAL: A liquid, including steam, for which a statistically valid positive result is obtained when the sample is analyzed to the following a priori lower limits of detection (analytical sensitivity).

The analytical sensitivity for identifying a source containing licensed material is, at a minimum, the licensee's lower limits of detection that are required for radioactive liquid effluents for all isotopes.

POTENTIAL TO REACH GROUND WATER Spills or leaks with the potential to reach ground water:

Spill or leak directly onto native soil or fill, Spill or leak onto an artificial surface (i.e. concrete or asphalt) if the surface is cracked or the material is porous or unsealed, Spill or leak that is directed into unlined or non impervious ponds or retention basins (i.e. water hydrologically connected to ground water).

A spill or leak inside a building or containment unit is generally unlikely to reach ground water, particularly if the building or containment unit has a drain and sump system. However, the sump and drain system should be evaluated as part of the SSC risk assessment.

A spill or leak to a semi-impermeable or impermeable surface that is recaptured or remediated per Objective 1.4 before the close of the next business day does not trigger the voluntary communication protocol.

OFF-SITE DOSE CALCULATION MANUAL Page 154

The licensee shall document any agreement with State/Local officials that differs from this Industry guidance as part of their record. For example, some states or local authorities have indicated that they do not wish leaks/spills to be included in the voluntary communication protocol or that the voluntary communication should be completed in a shorter timeframe.

b. Communication with the designated State/Local officials shall be made before the end of the next business day for a water sample result i Of off-site ground water or surface water that exceeds any of the REMP reporting criteria for water as described in the ODCM/ODAM, or ii Of on-site surface water, that is hydrologically connected to ground water, or ground water that is or could be used as a source of drinking water, that exceeds any of the REMP reporting criteria for water as described in the ODCM/ODAM The licensee shall document the basis for concluding that the on-site ground water is not or would not be considered a source of drinking water. Examples of a defensible basis are documents from the regulatory agency with jurisdiction over ground water use.

c. When communicating to the State/Local officials, be clear and precise in quantifying the actual release information as it applies to the appropriate regulatory criteria (i.e. put it in perspective). The following information should be provided as part of the informal communication:

i A statement that the communication is being made as part of the NEI Ground Water Protection Initiative, ii The date and time of the spill, leak, or sample result(s),

iii Whether or not the spill has been contained or the leak has been stopped, iv If known, the location of the leak or spill or water sample(s),

v The source of the leak or spill, if known, vi A list of the contaminant(s) and the verified concentration(s),

vii Description of the action(s) already taken and a general description of future actions, viii An estimate of the potential or bounding annual dose to a member of the public if available at this time, and ix An estimated time/date to provide additional information or follow-up.

d. Voluntary communication to State and/or Local officials may also require NRC notification under 10 CFR 50.72(b)(2)(xi). Licensees should perform these notifications consistent with their existing program.

e. Contact NEI by email to GW Noticeonei.org as part of a voluntary communication event as described in Objective 2.2.

OFF-SITE DOSE CALCULATION MANUAL Page 1S55

ML091140599

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