ML20081B918

Proposed Radiological Effluent Tech Specs,Incorporating Requirements to Perform Audits of Emergency Preparedness Program & Security Plan

ML20081B918
Person / Time
Site:	Robinson
Issue date:	10/31/1983
From:	CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML14190A733	List: ML20081B918
References
NUDOCS 8310310032
Download: ML20081B918 (272)

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ENCLOSURE 3 PROPOSED RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS H. B. ROBINSON STEAM ELECTRIC PLANT UNIT NO. 2 l

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OCTOBER 1983 8310310032 831025 PDR ADOCK 05000 6 P

TABLE OF CONTENTS Section Title Page TECHNICAL SPECIFICATIONS AND BASES 1.0 Definitions 1-1 2.0 Safety Limits and Limiting Safety System Settings 2.1-1 2.1 Safety Limit, Reactor Core 2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure 2.2-1 2.3 Limiting Safety System Settings, Protective 2.3-1 Instrumentation 3.0 Limiting Conditions for Operation 3.1-1 3.1 Reactor Coolant System 3.1-1 3.1.1 Operational Components 3.1-1 3.1.2 Heatup and Cooldown 3.1-4 3.1.3 Minimum Conditions for Criticality 3.1-11 3.1.4 Maximum Reactor Coolant Activity 3.1-13 3.1.5 Leakage 3.1-16 3.1.6 Maximum Reactor Coolant Oxygen and Chloride 3.1-20 Concentration 3.2 Chemical and Volume Control System 3.2-1 3.3 Emergency Core Cooling System, Auxiliary Cooling Systems, Air Recirculation Fan Coolers, Contain-ment Spray, Post Accident Containment Venting System, and Isolation Seal Water System 3.3-1 3.3.1 Safety Injection and Residual Heat Removal Systems 3.3-1 3.3.2 containment Cooling and Iodine Removal Systems 3.3-5 3.3.3 Component Cooling System 3.3-7 3.3.4 Service Water System 3.3-8 3.3.5 Post Accident Containment Venting System 3.3-9 3.3.6 Isolation Seal Water System 3.3-9 l 3.3.7 Extended Maintenance 3.3-10 3.4 Secondary Steam and Power Conversion System 3.4-1 3.5 Instrumentation Systems 3.5-1 3.5.1 Operational Safety Instrumentation 3.5-1 3.5.2 Radioactive Liquid Effluent Instrumentation 3.5-1 3.5.3 Radioactive Gaseous Effluent Instrumentation 3.5-2 3.6 Containment System 3.6-1 l 3.6.1 Containment Integrity 3.6-1 l 3.6.2 Internal Pressure 3.6-1 3.6.3 Containment Automatic Isolation Trip Valves 3.6-2 3.7 Auxiliary Electric Systems 3.7-1 l 3.8 Refueling 3.8-1 3.9 Radioactive Effluents 3.9-1 3.9.1 Compliance with 10 CFR Part 20 - Radioactive Materials in Liquid Effluents 3.9-1 3.9.2 Compliance with 10 CFR Part 50 - Radioactive Materials in Liquid Effluents 3.9-2 i

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Section Title Page 3.9.3 Compliance with 10 CFR Part 20 - Radioactive Materials in Gaseous Effluents 3.9-3 3.9.4 Compliance with 10 CFR Part 50 - Radionoble

Gases 3.9-4 3.9.5 Compliance with 10 CFR Part 50 - Radioiodines, Radioactive Materials in Particulate Form, and Radionuclides other than Radionoble Gases 3.9-5 3.9.6 Compliance with 40 CFR Part 190 - Radioactive Effluents from Uranium Fuel Cycle Sources 3.9-6 3.10 Required Shutdown Margins, Control Rods, and Power Distribution Limits '

3.10-1 3.10.1 Full Length Control Rod Insertion Limits 3.10-1 3.10.2 Power Distribution Limits 3.10-2 3.10.3 Ouadrant Power Tilt Limits 3.10-7 3.10.4 Rod Drop Time 3.10-8 3.10.5 Deleted 3.10.6 Inoperable Control Rods 3.10-8 3.10.7 Power Ramp Rate Limits 3.10-9 j 3.10.8 Required Shutdown Margins 3.10-9 3.11 Movable In-Core Instrumentation 3.11-1 3.12 Seismic Shutdown 3.12-1 3.13 Shock Suppressors (Snubbers) 3.13-1 3.14 Fire Protection System 3.14-1 3.14.1 Fire Detection Instrumentation 3.14-1 3.14.2 Fire Suppression Water System 3.14-1 3.14.3 CO2 Fire Protection System 3.14-2 3.14.4 Fire Rose Stations 3.14-2a 3.14.5 Fire Barrier Penetration Fire Seals 3.14-3 3.15 Control Room Filter System 3.15-1

! 3.16 Radioactive Waste Systems 3.16-1 3.16.1 Liquid Radwaste Treatment System 3.16-1 3.16.2 Liquid Holdup Tanks 3.16-1 3.16.3 Gaseous Radwaste and Ventilation Exhaust Treatment Systems 3.16-2 3.16.4 Waste Gas Decay Tanks (Hydrogen and Oxygen) 3.16-3 3.16.5 Waste Gas Decay Tanks (Radioactive Materials) 3.16-5 3.16.6 Solidification of Wet Radioactive Waste 3 16-6 3.17 Radiological Environmental Monitoring Program 3.17-1 3.17.1 Monitoring Program 3.17-1 3.17.2 Land Use Census 3.17-3 3.17.3 Interlaboratory Comparison Program 3.17-4 4.0 Surveillance Requirements 4.1-1 4.1 Operational Safety Review 4.1-1 4.2 Primary System Surveillance 4.2-1 4.3 Primary System Testing Following Opening 4.3-1 4.4 Containment Tests 4.4-1 4.4.1 Operational Leakage Rate Tests 4.4-1 4.4.2 Isolation valve Tests 4.4-4 l 4.4.3 Post Accident Recirculation Heat Removal System 4.4-4 l 4.4.4 Operational Surveillance Program 4.4-5 i

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Section Title Page 4.5 Emergency Core Cooling, Containment Cooling and Iodine Removal Systems Tests 4.5-1 4.5.1 System Tests 4.5-1 4.5.2 Component Tests 4.5-2 _

4.6 Emergency Power System Periodic Tests 4.6-1 4.6.1 Diesel Generators 4.6-1 4.6.2 Diesel Fuel Tanks 4.6-2 4.6.3 Station Batteries 4.6-2 4.7 Secondary Steam and Power Conversion System 4.7-1 4.8 Auxiliary Feedwater System 4.8-1 4.9 Reactivity Anomalies 4.9-1 4.10 Radioactive Effluents 4.10-1 4.10.1 Radioactive Liquid Effluents 4.10-1 4.10.2 Radioactive Gaseous Effluents 4.10-2 .

4.10.3 Radionoble Gases 4.10-2 4.10.4 Radioiodines, Radioactive Materials in Particulate Form, and Radionuclides Other Than Radionoble Gases 4.10-3 4.10.5 Radioactive Effluents From Uranium Fuel Cvele Sources 4.10-3 4.11 Reactor Core 4.11-1 4.12 Refueling Filter Systems 4.12-1 4.13 Shock Suppressors (Snubbers) 4.13-1 4.14 Fire Protection System 4.14-1 4.15 Control Room Filter System 4.15-1 4.16 Radioactive Source Leakage Testing 4.16-1 4.17 Systems Integrity 4.17-1 4.18 Radioactive Effluent Instrumentation 4.18-1 4.18.1 Radioactive Liquid Effluent Instrumentation 4.18-1 4.18.2 Radioactive Gaseous Effluent Instrumentation 4.18-1 4.19 Radioactive Waste Systems 4.19-1 4.19.1 Liquid Radwaste Treatment System 4.19-1 4.19.2 Liquid Holdup Tanks 4.19-1 4.19.3 Gaseous Radwaste and Mentilation Exhaust Treatment System 4.19-2 4.19.4 Waste Gas Decay Tanks (Hydrogen and Oxygen) 4.19-2 4.19.5 Waste Gas Decay Tanks (Radioactive Material) 4.19-3 4.19.6 Solidification of Wet Radioactive Waste 4.19-3 4.20 Radiological Environmental Monitoring Program 4.20-1 4.20.1 Monitoring Program 4.20-1 4.20.2 Land Use Census 4.20-1 4.20.3 Interlaboratory Comparison Program 4.20-2 5.0 Design Features 5.1-1 5.1 Site 5.1-1 5.2 Containment 5.2-1 5.2.1 Reactor Containment 5.2-1 5.2.2 Penetrations 5.2-1 5.2.3 Containment Systems 5.2-2 iii

l Section Title Page 5.3 Reactor 5.3-1 5.3.1 Reactor Core 5.3-1 5.3.2 Reactor Coolant System 5.3-2 ~

5.4 Fuel Storage 5.4-1 5.5 Seismic Design 5.5-1 6.0 Administrative Controls 6.1-1 6.1 Responsibility 6.1-1 6.2 Organization 6.2-1 6.3 Facility Staff Qualifications 6.3-1 6.4 Training 6.4-1 6.5 Review and Audit 6.5-1 6.6 Reportable Occurrence Action 6.6-1 6.7 Safety Limit Violation 6.7-1 6.8 Not Used 6.8-1 6.9 Reporting Requirements 6.9-1 6.10 Record Retention 6.10-1 6.11 Radiation Protection Program 6.11-1 6.12 Deleted 6.12-1 6.13 High Radiation Area 6.13-1 6.14 Environmental Qualification 6.14-1 6.15 Process Control Program 6.15-1 6.16 Offsite Ibse Calculation Manual 6.16-1 6.17 Major Changes to Radioactive Liquid, Gaseous and Solid Waste Treatment Systems 6.17-1 iv

LIST OF FIGURES Figure Title Page 1.1-1 Plant Site Boundary and Exclusion Zone 1-8 2.1-1 Safety Limits Reactor Core, Thermal, and Hydraulic Three Loop Operation, 100% Flow 2.1-7 2.1-2 Safety Limits Reactor Core, Thermal, and Hydraulic TWo Loop Operation 2.1-8 3.1-1 Reactor Coolant System Heat Up Limitations -

Applicable for Records Up' to 20 Effective Full Power Years 3.1-21 3.1-2 Reactor Coolant System Cool Down Limitations -

Applicability for Periods Up to 20 Effective Full

. Power Years 3.1-22 3.10-1 Control Group Insertion Limits for Three Loop Operation 3.10-20 3.10-2 Shutdown Margin versus Boron Concentration 3.10-21 3.10-3 Normalized Axial Dependence Factor for FO versus Elevation 3.10-22 6.2-1 'Offsite Organization for H. B. Robinson 2 Management and Technical Support 6.2-3 6.2-2 Conduct of Operations Chart 6.2-4 v

LIST OF TABLES Table TLtie Page 3.5-1 Engineered Safety Feature System Initiation Instrument Setting Limits 3.5-10 3.5-2 Reactor Trip Instrumentation Limiting Operating Conditions 3.5-12 3.5-3 Instrumentation Operating Conditions for Engineered Safety Features 3.5-14 3.5-4 Instrument Operating Conditions for Isolation Functions 3.5-16 3.5-5 Instrumentation to Follow the Course of an Accident 3.5-18 3.5-6 Radioactive Liquid Ef fluent Monitoring Instrumentation 3.5-20 3.5-7 Radioactive Gaseous Ef fluent Monitoring Instrumentation 3.5-24 3.13-1 Safety Related Hydraulic Snubbers 3.13-3 3.14-1 Fire Detection Instrumentation 3.14-4 3.14-2 Hose Stations 3.14-5 3.17-1 Radiological Environmental Monitoring Program 3.17-7 3.17-2 Reporting Levels for Radioactivity Concentrations in Environmental Samples 3.17-11 3.17-3 Maximum Values for the Lower Limits of Detection (LLD) 3.17-12 4.1-1 Minimum Frequencies f or Checks, Calibrations and Test of Instrument Channels 4.1-4 4.1-2 Frequencies f or Sampling Tests 4.1-7 4.1-3 Frequencies for Equipment Tests 4.1-9 4.2-1 Post-Operational Nondestructive Inspections 4.2-17 4.10-1 Radioactive Liquid Waste Sampling and Analysis Program 4.10-5 4.10-2 Radioactive Gaseous Waste Sampling and Analysis Program 4.10-8 vi

LIST OF TABLES (Cont'd)

Table Title Page 4.18-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 4.18-3 4.18-2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 4.18-5 J

A vii

1.6 INSTRUMENTATION SURVEILLANCE 1.6.1 Action Action shall be that part of a specification which prescribes remedial measures required under designated conditions.

1.6.2 Channel Calibration P

Adjustment of channel output such that it responds, with acceptable range and accuracy, to known value of the parameter which the channel l monitors. Calibration shall encompass the entire channel, including the sensor and alarm or trip function, and shall be deemed to include the channel functional test.

1.6.3 Channel Check A qualitative determination of acceptable operability by observation of channel behavior during operation. This determination will include, whenever possible, comparison of the channel with other independent channels measuring the same variable. ',

1.6.4 Channel Functional Test

! Injection of a simulated signal into the channel to verify that it is operable, including alarm and/or trip initiating action.

l 1.6.5 Source Check l

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

l 1.7 CONTAINMENT INTEGRITY l

Containment integrity is defined to exist when:

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a. All non-automatic containment isolation valves not required for' normal operation are closed and blind flanges are properly '

installed where required.

b. The equipment door is properly closed and' seated.

c. At least one door in the personnel air lock is properly closed -

s and sealed.

d. All automatic containment isolation trip valves required to be -

closed during accident conditions are operable or are secured closed except as stated in Specification 3.6.3. Manual valves n

' qualifying as automatic containment isolation valves are secured closed.

e. The uncontrolled containmenc leakage satisfies Specification 4.4.

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1.8 OUADRANT POWER TILT The quadrant power tilt is defined as the ratio of maximum to average of the upper excore detector currents or the lower excore detector currents, whichever is greater. If one exhore is out of service, the three in-service units are used in computing the average.

e 1.9 FIRE SUPPRESSION WATER SYSTEM A fire suppression water system shall concist of : a water source; pumps; and distribution piping with associated sectionalizing control l or isolation valves.

l 1.10 STACCERED TEST BASIS 1 1

A Staggered Test Basis shall consist of: ,

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a. A test schedule for n systems, subsystems, trains or designated components obtained by dividing the specified test interval into n equal subintervals.

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b. The testing of one system, subsystem, train or designated components at the beginning of each subinterval.

1.11 GASEOUS RADWASTE TREATMENT SYSTEM The Gaseous Radgaste Treatment System is the system designed and installed to reduce radioactive gaseous effluents by collecting

. primary coolant's:Ntem off-gases from the primary system and providing for delay c'r holdup for the purpose of reducing the total radioactivity prior to releese to the environment.

1.12 VENTILATION EXHAUST TREATMENT SYSTEM

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The Ventilation Exhaust Treatm'nt e System is the system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluenta by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters prior to their release to the environment. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considared to be Ventilation Exhaust Treatment System components.

1.13 0FFSITE DOSE CALCULATION MANUAL (ODCM)

The Offsite Dose Calculation Manual shall contain the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and the methodology to calculate gaseo'u4 and liquid effluent monitoring alarm / trip setpoints; and, ther requirements of the environmental radiological monitoring program. ,

1.14 DOSE EQUIVALENT I-131 The Dose Equivalent I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and 1-5 w ,, -

..7. ,,,, - _. , __- , , , , . _ .

I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in NRC Regulatory Guide 1.109 Revision 1, October 1977.

1.15 PROCESS CONTROL PROGRAM (PCP)

The Process Control Program (PCP) shall contain the current formulas, sampling, analyses, tests and determinations to be made to ensure that the processing and packaging of solid radiosctive was;es based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with i

10 CFR Part 20,10 CFR Part 71, and Federal ,and State regulations and l other requirements governing the disposal of the radioactive waste.

1.16 SOLIDIFICATION Solidification shall be the conversion of wet radioactive wastes into a form that meets shipping and burial ground requirements.

1.17 PURGE - PURGING Purge or purging is the controlled process of discharging air or gas I

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 ccnfinement.

1.18 VENTING Venting is the c'nt <o ) < trocess of discharging air or gas from a l confinement to ma33tain tapperature, pressure, humidity, l concentration or other operating condition, in such a manner that l replacement air or gas is not provided or required during venting. I l

Vent, used in system names, does not imply a venting process.

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1.19 SITE BOUNDARY 4

The site boundary shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee, as defined by Figure 1.1-1.

I 1.20 MEMBER (S) 0F THE PUBLIC l

Member (s) of the public shall include all individuals who by virtue of their occupational status have no formal association with the plant. This category shall include non-employees of the licensee who are permitted to use portions of the site for recreational, occupa-tional or other purposes not associ*ated with plant functions. This category shall not include non-employees such as vending machine servicemen, or postmen who, as part of their formal job function, occasionally enter an area that is controlled by the licensee for the

, purposes of protection of individuals from exposure to radiation and radioactive materials.

1.21 UNRESTRICTED AREA Unrestricted area shall be any area at or beyond the Site Boundary to which access 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 l

quarters or for industrial, commercial, institutional, and/or recreational purposes.

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3.5 INSTRUMENTATION SYSTEMS 3.5.1 Operational Safety Instrumentation Applicability Applies to plant operational safety instrumentation systems.

Objective To provide for automatic initiation of the Engineered Safety Features in the event that principal process variable limits are exceeded, and to delineate the conditions of the plant instrumentation and safety circuits necessary to ensure reactor safety.

Specification 3.5.1.1 The Engineered Safety Features initiation instrumentation setting limits shall be as stated in Table 3.5-1.

I 3.5.1.2 For on-line testing or in the event of a subsystem instrumentation channel failure, plant operation at rated power 3 hall be permitted to continue in accordance with Tables 3.5-2 through 3.5-5.

l 3.5.1.3 In the event the number of channels of a particular subsystem in service falls below the limits given in the column entitled Minimum Operable Channels, or Minimum Degree of Redundancy cannot be achieved, operation shall be limited according to the requirement shown in Column 3 of Tables 3.5-2 through 3.5-4 and l Column 2 of Table 3.5-5.

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3.5-1

I 3.5.2 Radioactive Liquid Effluent Instrumentation Applicability Applies to the radioactive liquid effluent instrumentation system. l l

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Objective To define the operating requirements for the radioactive liquid effluent instrumentation system.

l Specification 1

3.5.2.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.5-6 shall be operable with their alarm / trip setpoints set to ensure that the limits of Specification 3.9.1.1 are not exceeded. The alarm / trip setpoints shall be determined in accordance with the ODCM.

3.5.2.2 With a radioactive liquid monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive liquid effluent monitored by the affected channel, change the setpoint so it is acceptably conservative, or declare the channel not operable.

3.5.2.3 With less than the minimum number of radioactive liquid effluent monitoring instrumentation operable, take the action shown in Table 3.5-6.

3.5.2.4 The provisions of Specifications 3.0 and 6.9.2.b(2) are not applicable.

3.5-2

3.5.3 Radioactive Gaseous Effluent Instrumentation Applicability Applies to the radioactive gaseous effluent instrumentation system.

Objective To define the operating requirements for the radioactive gaseous effluent instrumentation system.

Specification 3.5.3.1 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.5-7 shall be operable with their alarm / trip setpoints set to ensure that the limits of Specification 3.9.3.1 are- not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the ODCM.

3.5.3.2 With a radioactive effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive gaseous effluents, change the setpoint so it is acceptably conservative, or declare the channel not operable.

3.5.3.3 With less than the minimum number of radioactive effluent f monitoring instrumentation channels operable take the action shown in Table 3.5-7.

3.5.3.4 The provisions of Specification 3.0 and 6.9.2.b(2) are not applicable.

l 3.5-3

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Basis Operational Safety Instrumentation Instrumentation has been provided to sense accident conditions and to initiate operation of the Engineered Safety Features.(I)

Safety Injection System Actuation e

Protection against a Loss-of-Coolant or Steam Break accident is brought about by automatic actuation of the Safety Injection System which provides emergency cooling and reduction of reactivity.

The Loss-of-Coolant Accident is characterized by depressurization of the Reactor Coolant System and rapid lose of reactor coolant to the containment.

The Engineered Safety Features have been designed to sense these effects of the Loss-of-Coolant Accident by detceting low pressurizer pressure and generate signals actuating the SIS active phase.

The SIS active phase is also actuated by a high containment pressure signal (Hi-Level) brought about by loss of high enthalpy coolant to the containment. This actuation signal acts as a backup to the low pressurizer pressure signal actuation of the SIS and also adds diversity to protection against loss of coolant.

Signals are also provided to actuate the SIS upon sensing the effects of a steam line break accident. Therefore, SIS actuation following a steam line break is designed to oce.ur upon sensing high differential steam pressure between the steam header and steam generator line or upon sensing high steam line flow in coincidence with low reactor coolant average temperature or low steam line pressure.

The increase in the extraction of RCS heat following a steam line break results in reactor coolant temperature and pressure reduction. For this reason, protection against a steam line break accident is also provided by low pressurizer pressure signals actuating safety injection.

3.5-4

_ _ ._. _ ___ ._. = _ . . . . . _ _ _ _ . _ ___ ._.

Protection is also provided for a steam line break in the containment by actuation of SIS upon high containment pressure.

EIS actuation injects highly borated fluid into the Reactor Coolant System in order to counter the reactivity insertion brought about by cooldown of the reactor coolant which occurs during a steam line break accident.

Containment Spray The Engineered Safety Features also initiate containment spray upon sensing a high containment pressure signal (Hi-Ri Level). The containment spray acts to reduce containment pressure in the event of a loss of coolant or steam line break accident inside the containment, in order to reduce containment pressure. The containment spray cools the containment directly and limits the release of fission products by absorbing iodine should it be released to the containment. .

! Containment spray is designed to be actuated at a higher containment pressure I

(approximately 50% of design containment pressure) than the SIS (10% of design). Since spurious actuation of containment spray is to be avoided, it is initiated only on coincidence of Hi-Ri Level containment pressure sensed by both of the two sets of containment pressure signals provided for its

actuation.

Steam Line Isolation i

Steam line isolation signals are initiated by the Engineered Safety Features clocing all steam line stop valves. In the event of a steam line break, this i action prevents continuous, uncontrolled steam release from more than one

! steam generator by isolating the steam lines on high containment pressure (Hi-Hi-Level) or high steam line flow. Protection is afforded for breaks inside

'or outside the containment even when it is assumed that there is a single failure in the steam line isolation system.

3.5-5

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Feedwater Line Isolation The feedwater lines are isolated upon actuation of the Safety Injection System in order to prevent excessive cooldown of the reactor coolant system. This mitigates the effect of an accident such as a steam break which, in itself, causes excessive coolant temperature cooldown.

Feedwater line isolation also reduces the consequences of a steam line break inside the containment, by stopping the entry of feedwater.

Setting Limits

a. The Hi-Level containment pressure limit is set at about 10% of design containment pressure. Initiation of Safety Injection protects against loss-of-Coolant (2) or ,g,,,

line break (3) accidents as discussed in the safety analysis.

b. The Hi-Hi Level containment pressure limit is set at about 50% of design containment pressure. Initiation of Containment Spray and Steam Line Isolation protects against large Loss-of-Coolant (2) or steam line break accidents,(3) as discussed in the safety analysis.

c. The pressurizer low pressure limit is set substantially below system operating pressure limits. However it is sufficiently high to protect against a Loss-of-Coolant Accident as shown in the safety analysis.(2) l d. The steam line high differential pressure limit is set ,

in the event of a large steam line break accident, as shown  !

in the safety analysis.(3)

e. The high steam line flow limit is set at approximately 40%  !

of the steam flow from no load to 20% and at 110% of t full steam flow at full load, with the steam flow differential pressure measurement linearly programmed between 3.5-6

l 20% load and 100% load in order to protect against

, large steam line break accidents.(4) The coincident low T setting limit for SIS and steam line isolation initia-ave tion is set below its hot shutdown value. The coincident steam line pressure setting limit is set below the full load operating pressure. The safety analysis shows that these settings provide protection in the event of a large steam line break.(3)

Instrument Operating Conditions _

During plant operations, the complete instrumentation systems will normally be in service. Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established limits. Safety is not comprised, however, by continuing operation with certain instrumentation channels out of service since pro-visions were made for thia in the plant design. This specification outlines limiting conditions for operation necessary to preserve the effectiveness of the Reactor Control and Protection System when any one or more of the channels is out of service.

Almost all reactor protection channels are supplied with sufficient redundancy to provide the capability for channel calibration and test at power. Excep-tions are backup channels such as reactor coolant pump breakers. The removal of one trip channel on process control equipment is accomplished by placing that channel bistable in a tripped mode; e.g., a two-out-of-three circuit becomes a one-out-of-two circuit. The nuclear instrumenation system channels are not intentionally placed in a tripped mode since the test signal is

superimposed on the normal detector signal to test at power. Testing of the NIS power range channel requires (a) bypassing the Dropped Rod protection from l NIS, for the channel being tested, (b) defeating the A T/T av8 protection CHANNEL SET that is being fed from the NIS channel, and (c) defeating the f

f power mismatch section of T av8 e ntr 1 channels when the appropriate NIS l channel is being tested. However, the Rod Pbsition System and remaining NIS l channels still provide the dropped-rod protection. Testing does not trip the system unless a trip condition exists in a concurrent channel.

3.5-7 i.___ _ _ . _ _ _ _ _ _.._________._ -

Instrumentation to Access Plant Conditions During and Following an Accident The operability of the accident monitoring instrumentation ensures that suffi-cient information is available on selected plant parameters to monitor and assess these variables during and following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," July 1979.

Radioactive Liquid Effluent Instrumentation The radioactive liquid effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The operability and use of this instrumentation are consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

1 3.5-8

Radioactive Gaseous Effluent Instrumentation The radioactive gaseous effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.

The alarm / trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The operability and use of this instrumentation are consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

References (1) FSAR Section 7.5 (2) FSAR Section 14.3 (3) FSAR Section 14.2.5 (4) CP&L Letter to the Directorate of. Licensing dated October 23, 1973.

3.5-9

TABLE 3.5-1 ENGINEERED SAFETY FEATURE SYSTEM' INITIATION INSTRUMENT SETTING LIMITS NO. FUNCTIONAL UNIT CHANNEL ACTION SETTING LIMIT

1. High Containment Pressure (HI Level) Safety Injection

• j( 5 psig

2. High Containment Pressure (HI-HI Level) a. Containment Spray ** j[ 25 psig

b. Steam Line Isolation

3. Pressurizer Low Pressure Safety Injection * > 1700 psig

4. High Differential Pressure Safety Injection

• j[ 150 psi Between any Steam Line and the Steam Line Header

5. High Steam Flow in 2/3 a. Safety Injection

• j[ 40% (at zero load) u Steam Lines *** b. Steam Line Isolation of full steam flow En

< 40% (at 20% load) d-

~~ of full steam flow

~~< 110% (at full load) of full steam flow Coincident with Low T r ***

av8 >

> 541*F T,yg*

s team line Low Steam Line Pressure _

600 psig pressure

6. Loss of Power

a. 480 V Emerg. Bus Undervoltage Trip Normal Supply Breaker 328 Volts + 1 Volt (Loss of Voltage) Time Delay

.75 + .25 sec.

t

TABLE 3.5-1 (Ccntinued)

ENGINEERED SAFETY FEATURE SYSTEM INITIATION INSTRUMENT SETTING LIMITS NO. FUNCTIONAL UNIT CHANNEL ACTION SETTING LIMIT

6. b. 480V Emerg. Bus Undervoltage Trip Normal Supply Breaker 412 Volts ;t 1 Volt (Cont'd) (Degraded Voltage) Time Delay 10.0 second delay +_

0.5 sec.

7. Containment Radioactivity High Ventilation Isolation __

< 2 X reading at the time the alarm is set with known plant conditions Y'

Initiates also containment isolation (Phase A), feedwater line isolation, and starting of all containment fans.

• • Initiates also containment isolation (Phase B).

• • • Derived f rom equivalent AP measurements.

• • • • These setting limits shall be greater than or equal to 524*F and 450 psig when operating under reduced temperature conditions described in the Novenber 11, 1981 license submittal.

l

TABLE 3.5-2 REACTOR TRIP INSTRUMENTATION LIMITING OPERATING CONDITIONS 1 2 3 MINIMUM OPERATOR ACTION MINIMUM DEGREE IF CONDITIONS OF OPERABLE OF COLUMN 1 OR 2 NO. FUNCTIONAL UNIT CHANNELS REDUNDANCY CANNOT BE MET

1. Manual 1 0 Maintain hot shutdown

2. Nuclear Flux Power Range

• 3 2 Maintain hot shutdown

3. Nuclear Flux Intermediate Range 1 0 Maintain hot shutdown **

p> 4. Nuclear Flux Source Range 1 0 Maintain hot ja shutdown ***

C

5. Overtemperature at 2 1 Mhintain hot shutdown

6. Overpower ur 2 1 Maintain hot shutdown

7. Low Pressurizer Pressure - 2 1 Maintain hot shutdown

8. Hi Pressurizer Pressure 2 1 Maintain hot shutdown

9. Pressurizer-Hi Water Level 2 1 Maintain hot shutdown

10. Low Reactor Coolant Flow 2/ operable loop 1/ operable loop Maintain hot shutdown l 11. Turbine Trip 2 1 Maintain less than 10% R.P.

i i TABLE 3.5-2 (Cont'd)

REACTOR TRIP INSTRIMENTATION LIMITING OPERATING CONDITIONS 1 2 3

MINIMUM OPERATOR ACTION IF CONDITIONS OF MINIMUM DEGREE l OPERABLE OF COLUMN 1 OR 2 NO. FUNCTIONAL UNIT CHANNELS REDUNDANCY CANNOT BE MET

, 12. Lo Lo Steam Generator 2 1 Maintain Hot Water Level Shutdown

13. Underf requency 4 KV System 2 1 Maintain Hot Shutdown

14. Undervoltage on 4 KV System 2 1 Maintain Hot Shutdown

F 15. Control Rod Misalignment i

f Ma ni tor * * *

• C

, a. Rod Position Deviation 1 0 Log individual rod position once/ hour, and after a load change >10% or j af ter >30 inches of

! control rod motion i

b. Quadrant Power Tilt 1 0 Log individual upper j Monitor (upper and and lower ion cham-lower ox-core neutron ber currents once/

detectors) hour and af ter a los change >10% or after

>30 inches of contro rod motion

• For zero power physics testing, it is permissible to take one channel out of service.

j ** When two of four power channels are greater than 10% full power, hot shutdown is not required.

l *** When one of two intermediate range channels is greater than IE-10 amps, hot shutdown is not required.

• • • • If both rod misalignment monitors (a and b) are inoperable for two hours or more, the nuclear overpower trip shall be reset to 93 percent of rated power in addition to the increased surveillance noted.

R.P. = Rated Power

TABLE 3.5-3 I

INSTRUMENTATION OPERATING CONDITIONS FOR ENGINEERED SAFETY FEATURES i

1 2 3 MINIMUM OPERATOR ACTION MINIMUM DEGREE IF CONDITIONS OF CHANNELS OF COLIMN 1 OR 2 NO. FUNCTIONAL UNIT OPERABLE REDUNDANCY CANNOT BE MET 1 SAFETY INJECTION

a. Manual 1 0 Cold Shutdown

't

b. High Containment Pressure 2 1 Cold Shutdown

]1 (Hi lovel)

I

c. High Differential Pressure 2 1 Cold Shutdown ***

l between any Steam and the l ,w Steam Line Header

!Y g d. Pressurizer Iow Pressure 2 1 Cold Shutdown ***

e. High Steam Flow in 2/3 Steam 1/ Steam Line *****

Lines Coincident with Low T,yg 2T SiE nals 1 Cold Shutdown ***

' 8vE or Low Steam Pressure 2 Pressure Signals 1 i

i i

I

)

t l

l

i TABLE 3.5-3 (Continurd)

INSTRUMENTATION OPERATING CONDITIONS FOR ENGINEERED SAFETY FEATURES 3

1 2 OPERATOR ACTION MINIMUM MINIMUM IF CONDITIONS OF I

CHANNELS DEGREE OF COLUMN 1 OR 2 j NO. FUNCTIONAL UNIT OPERABLE RCDUNDANCY CANNOT BE MET t

l 2. CONTAI! MENT SPRAY l

j a. Manual

• 2 0** Cold Shutdown j b. High Containment Pressure

• 2/ set 1/ set Cold Shutdown i

1 (Hi-Hi Invel)

3. LOSS OF POWER u a. 480V Emerg. Bus 2/ bus (a) 1/ bus (b) Maintain Hot Shutdown j 'u Undervoltage (Ioss

' d.

u of Voltage)

b. 480V Emerg. Bus 2/ bus 1/ bus Maintain Hot Shutdown (c) l Undervoltage

(Degraded Voltage) i i

j

• Also initiates a Phase B containment isolation, j ** Must actuate two switches simultaneously.

{ *** When primary pressure is less than 2000 psig, channels may be blocked l' **** When primary temperature is less than 547'F, channels may be blocked.(d)

• • • • • In this case, the 2/3 high steam flow is already in the trip mode.

) (a) During testing and maintenance of one channel, may be reduced to 1/ bus.

i (b) Ikaring testing and maintenance of one channel, may be reduced to 0/ bus.

1 (c) The reactor may remain critical below the power operating conditions I with this feature inhibited f or the purpose of starting reactor coolant Pumps.

(d) When operating under the reduced temperature conditions described in the November 11, 1981 license j

' submittal, the channels may be blocked when primary temperature is less than 530*F.

i -

i

i TABLE 3.5-4 INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FUNCTIONS 1 2 3 MINIMUM OPERATOR ACTION MINIMUM DEGREE IF CONDITIONS OF OPERABLE OF COLIMN 1 OR 2 NO. FUNCTIONAL UNIT CHANNELS REDUNDANCY CANNOT BE MET

1. CONTAIMENT ISOLATION

a. Phase A

1. Safety Injection See Item No. 1 of Table 3.5-3 Cold Shutdown

11. Manual 1 0 Hot Shutdown

b. Phase B See Item No. 2 of Table 3.5-3

. c. Ventilation Isolation 5 1. High Containment Activity 1 0 Containment shall not be purged.

11. Phase A See Item No. 1.a of Table 3.5-4 l

l

TABLE 3.5-4 (Continued)

INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FUNCTIONS I 2 3 MINIMUM OPERATOR ACTION

. MINIMUM DEGREE IF CONDITIONS OF l OPERABLE OF COLIMN 1 OR 2 NO. ~ FUNCTIONAL UNIT CHANNELS REDUNDANCY CANNOT BE NET

! 2. STEAM LINE ISOLATION

a. High Steam Flow in 2/3 Steam Lines See Item No. 1 of Table 3.5-3 Cold Shutdown Coincident with Low T r Low 8vE j Steam Pressure

b. High Containment Pressure See Item No. 1 of Table 3.5-3 Cold Shutdown

." c. Manual 1/Line 0 Hot Shutdown Y

Z 3. FEEINATER LINE ISOLATION

~

. a. Safety Injection See Item No. 1 of Table 3.5-3 Cold Shutdown l

.I f

1 i

1 TABLE 3.5-5 (THIS TABLE APPLIES WHEN THE RCS IS > 350*F) l INSTRIMENTATION TO FOLLOW THE COURSE OF AN ACCIDENT 2

1 OPERATOR ACTION l

HINEMUM IF CONDITIONS OF CHANNELS COLUMN 1 CANNOT NO. INSTRUMENT OPERABLE BE MET 1 Pressurizer Level 2 See Item 9 Table 3.5-2 2 Auxiliary Feedwater Flow Indication Note 1 (Primary Indication)

SD AFW Pump 1 per S/G l w MD AW Pump 1 per S/G l Y' j g 3 Reactor Coolant System Subcooling Monitor 1 Note 2 4 PORV Position Indicator (Primary) 1 Note 3 5 PORV Blocking Valve Position Indicator 1 Note 3 (Primary) 1 6 Safety Valve Position Indicator (Primary) 1 Note 3 l

~

Note 1: The three AFW lines f rom the MD AFW pumps and the three AFW lines f rom the SD AFW pump each contain one primary flow indicator (2 AFW flow paths per steam generator for a total of 6 AFW lines). These

) primary indicators are backed up by the narrow range steam generator level indications. If one or j more of the direct AFW flow indicators becomes inoperable when the RCS is > 350*F, restore the indicator (s) i to an operable status within 7 days, or prepare and submit a special report to the NRC within the j following 14 days detailing the cause(s) of the inoperable indicator (s), the actions being taken to l restore the indicator (s) to an operating status, the estimated date for completion of the repairs, and j any compensatory action being taken while the indicator (s) is inoperable. The action required when any j of the backup indications of AFW flow are inoperable, is described in Table 3.5-2.

J (Notes 2 & 3 -- see next page)

I

4

TABLE 3.5-5 (Continued)

I INSTRUMENTATION TO FOLLOW THE COURSE OF AN ACCIDENT i

I Note 2 If both channels of the RCS subcooling monitor become inoperable when ?.he RCS is >350*F, restore at least one channel to an operable status within 7 days, or prepare and submit a special report to i the NRC within the following 14 days detailing the cause(s) of the inoperable channels, the actions 4 being taken to restore at least one channel to an operable status, the estimated date for completion of the repairs, and any compensatory action being taken while both channels are inoperable.

] .

No te: 3 The Pzr PORVs and Pzr PORV blocking valves both incorporate limit switches f or the direct (primary) means of position indication. The backup method of position indication consists of PRT pressure

} and a temperature element in a common line downstream of the valves. The Pzr safety relief valves j incorporate a vibration monitoring system as the ' primary method of valve position indication. The 1 backup method of position indication consists of a temperature element downstream of each valve

! w and PRT pressure. If the primary method of position indication for either the Par PORVs, Pzr PORV y blocking valves, or Pzr safety relief valves becomes inoperable when the RCS is >350*F, restore the

primary method to an operable status within 7 days, or prepare and submit a special report to the NRC within the following 14 days detailing the cause of the inoperable primary position indication method, the actions being taken to restore it to an operable status, the estimated date for com-pletion of the repairs, and any compensatory action being taken while the primary position indication j method is inoperable. If any of the backup methods of position indication for these valves becomes

inoperable, it is to be repaired as soon as plant conditions permit.

I l

I i

i u

l 1

I i

TABLE 3.5-6 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Re lease Pathway / Instrumentation MCO* Required Action

1. Liquid Radwaste Ef fluent Discharge Line

a. Monitor (RMS-18) 1 With the number of channels operable less than the MCO requirements:

provides automatic a. Exert best efforts to return the instruments to operable status termination of within 30 days and, if unsuccessful, explain in the next Semi-release upon exceeding annual Radioactive Effluent Release Report why the inoperability alarm / trip setpoint was not corrected in a timely manner in accordance with Specification 6.9.1.d.4 and,

b. Ef fluent releases via this pathway may continue provided that l

prior to initiating a release:

1. TWo independent samples are analyzed in accordance with the l

P' Surveillance Requirements of Specification 3.9.1.1 and;

! j" Tuo members of the f acility staf f independently verify the y 2.

release rate calculations and the discharge line valving.

j

b. Flow rate measurement 1 With the number of channels operable less than the MCO requirement:

device a. Exert best ef f orts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Semi-I annual Radioactive Ef fluent Release Report why the inoperability was not corrected in a timely manner in accordance with Specification 6.9.1.d.4 and, l

1

• b. Ef fluent releases via this pathway may be continued, provided that the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump performance curves generated "in situ" and tank volumes may be used to estimate flow.

I

*MCO - Minimum Channels Operable -

l 5

4

TABLE 3.5-6 (Continu:d)

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Release Pathway / Ins t rumentation MCO* Required Action

2. Steam Generator Blowdown Ef fluent Line

a. Monitor (RMS-19) 1 With the number of channels operable less than the MCO requirement:

provides automatic a. Exert best efforts to return the instruments to operable status termination of blow- within 30 days and, if unsuccessful, explain in the next Semi-down from all three annual Radioactive Effluent Release Report why the inoperability Steam Generators upon was not corrected in a timely manner in accordance with exceeding alarm / trip Specification 6.9.1.d.4 and, setpoint.

b. Ef fluent releases via this pathway may continue provided that grab samples are analyzed for gross radioactivity (beta or gamma) with a lower limit of detection of at least 1.0E-07 p Ci/ml or are analyzed f or principle gamma emitters consistent with Table 4.10-1; y 1. Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of the j

t secondary coolant is j[ 0.01 p Ci/ml Dose Equivalent I-131, or;

2. Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activity of the secondary coolant is > 0.01 p Ci/ml Dose Equivalent I-131.

b. Flow rate measurement 1 per With the number of channels operable less than the MCO requirement:

devices - each Steam S/G a. Exert best efforts to return the instruments to operable status Generator has its own within 30 days and, if unsuccessful, explain in the next Semi-blowdown flow rate annual Radioactive Ef fluent Release Report why the inoperability measuring device. was not corrected in a timely manner in accordance with Specification 6.9.1.d.4 and,

b. Ef fluent releases via this pathway may continue provided that the flow rate for the affected blowdown line(s) is estimated at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• MCO - tu ntmum Channels Operable 1

TABLE 3.5-6 (Continutd) i RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 1

Re lease Pathway / Ins trumentation MCO* Required Action

3. Discharge Canal Flow Note 1 With the number of channels operable less than the MCO requirement suspend effluent release via this pathway

4. Tank Level Indicating With the ntmaber of channels operable less than the MCO requirement:

Devices a. Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Semi-

a. Refueling Water Storage 1 annual Radioactive Ef fluent Release Report why the inoperability Ta nk was not corrected in a timely manner in accordance with Specification 6.9.1.d.4 and,

b. Monitor Tanks Tank A 1 b. Liquid additions to the affected tank (s) may continue provided Tank B that the. liquid level for the affected tanks is estimated during all liquid additions to the affected tank (s).

l i

Y' c. Waste Condensate Tanks

,~ O Tank C 1 Tank D 1 Tank E 1

d. Temporary Tanks (Note 2) I per Tank J

1 i,

TABLE 3.5-6 (Continuid) 1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION i

Release Pathway / Ins trumentation MCO* Required Action

5. Containment Fan Cooling Water Monitor (Service Water Effluent Line)

a. Monitor (RMS-16) does 1 With the number of channels operable less than the MCO requirement:

not provide automatic a. Exert best efforts to return the instruments to operable status termination of release within 30 days and, if unsuccessful, explain in the next Semi-upon exceeding alarm annual Radioactive Effluent Release Report why the inoperability setpoint. was not corrected in a timely manner in accordance with Specification 6.9.1.d.4 and,

b. Ef fluent releases via this pathway may continue provided that, once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, grab samples are collected and analyzed for j gross radioactivity (beta or gamma) with a lower limit of l

u, detection of at least 1.0E-07 p C1/ml or are analyzed for l

u principal gamma emitters consistent with Table 4.10-1. .

6. Composite Sampler for Settling Pondo 1 With the number of channels operable less than the MCO requirement:

a. Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Semi-

, annual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner in accordance with Specification 6.9.1.d.4 and, i

b. Ef fluent releases via this pathway may continue provided that, i once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, grab samples are collected and analyzed for gross radioactivity (beta or gamma) with lower limit of

, detection of at least 1.0E-07 p Ci/a1 or are analyzed for principal gamma emitters consistent with Table 4.10-1.

l,

• MCO - Minimum Channels Operable

. NOTE TO TABLE 3.5-6 l Note 1 - Pump curves for Unit 2 operating circulating water pumps .amy be used to satisfy this MCO.

j If no Unit 2 circulating water pumps are operating the pump curves for circulating water i pumps operating in Unit 1 may be used to satisfy this MCO.

j Note 2 - A temporary tank is defined as any tank having a capacity of > 100 gallons used for the

~~

2 receipt or transfer of radioactive liquids.

\

l TABLE-3.5-7 -

l l RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION l

l j Release Pathway / Instrumentation MCO* Required Action

1. Plant Vent

a. Radionoble gas monitor 1 With the number of channels operable less than the MCO requirements:

(RMS-14) provides automatic a. Exert best efforts to return the instruments to operable status termination of Waste Gas Decay within 30 days and, if unsuccessful, explain in the next Semi-l Tank releases upon exceeding annual Radioactive Effluent Release Report why the inopera-

! alarm / trip setpoint. bility was not corrected in a timely manner in accordance with Specification 6.9.1.d.4 and, i b. Ef fluent releases via this pathway may continue provided that

prior to initiating a release

IF 1. Tko independent samples are analyzed in accordance with iY the Surveillance Requirements of Specification 3.9.3.1 and; Y l- ' - ,

j

2. Two members of 'the f acility s'taff indepe'ndently verify.the ,,

release rate calculations an'd the' discharge line 9alving.-l '

b. Radionoble gas monitors cf With the number of channels operable less than the MCO requirementir -

l RMS-14 and RMS-34' monitor the a. Exert best efforts to return the instruments to operable all effluents from two status within 30 days and;'ll unsuccessful, explain in the Auxiliary Building ,- moni- next Semiannual P.adioactive Ef fluen'c Release .P. aport why the i Ventilation System tors inoperability was not corrected in a timely manner in i

without providing' accordance with Specilication 6.9.1.d.4 and, l automatic termination l of release upon exceed- b. Ef fluent releases via this pathway may continue provided that s ing their respective grab samples are collected once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and are analyzed alarm setpoints. for radionoble gases once per 24 houre.

i

• MCO - Minimum Channels Operable , p.

1 i

i i

l i -

y ~- .

.i TABLE 3.5-7 (Continued) f RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION .

Re lease Pathway / Instrumentation MCO* Required Action 1.. Plant Vent (Continued)

c. Radioiodine Sampler 1 With the ntaber of channels operable less than the MCO requirement:

^_~ (RMS-34) a. Exert best efforts to return the instruments to operable status

. within 30 days and, if unsuccessful, explain in the next Sesi '

i - " ' - ~

annual Radioactive Effluent Release Report why the inoperability i , was not corrected in a timely manner i~n accordance with

'/ ,

Specification 6.9.1.d.4 and,.

b.~ Ef fluent release via this pathway may continue provided that a continuous sample is collected utilizing auxiliary sampling '

~

equipment as provided by Table 4.10-2. .

l y' d. Particulate Sampler 1 With the number of channels operable less than the MCO requirement:

]U -

.(RMS-34) a. Exert best efforts to return the instruments to operabl,e statais within 30 days and, if unsuccessful, explain in the-next Senti-annual Radioactive Ef fluent Release Report why the inoperability was not corrected in a timely manner in accordance with Specification 6.9.1.d.4 and,

b. Ef fluent releases via this pathway may continue provided that a continuous sample is collected utilizing auxiliary sampling equipment as required by Table 4.10-2.

e. Sampler flow rate monitor 1 of With the number of channels operable less than the MCO requirement:

(RMS-34) and Vacute the a. Exert best efforts to return the instruments to operable status

gauge (RMS-34) two within 30 days and, if unsucessessful, explain in the next Semi-4 moni- annual Radioactive Ef fluent Release Report why the inoperability j tors was not corrected in a timely manner in accordance with Specification 6.9.1.d.4 and,

b. Ef fluent releases via this pathway may continue provided the flow rate is estimated once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

i i

i i

i f

l TABLE 3.5-7 (Continued)

RADIOACTIVE GASEOUS EFFLUENT NONITORING INSTRIMENTATION i

Release Pathway / Instrumentation MCO* haquired Action l

1. Plant Vent (Continued) l l f. Plant Vent flow rate 1 With the nusber of channels operable less than the MCO requirement:  :

l monitor a. Exert best efforts to return the instruments to operable status ,

, within 30 days and, if unsuccessful, explain in the next Semi-i annual Radioactive Effluent Release Report why the inoperability i was not corrected in a timely manner in accordance with Specification 6.9.1.d.4 and, I

l'

b. Ef fluent releases via this pathuey may continue provided that flow rate is estimated once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

f. 2. Waste Gas Holdup System 1 With the number of channels operable less than the HC0 requirement:

g Explosive Gas Monitoring a. Exert best efforts to return the instruments to operable status i j os System within 14 days and, if unsuccessful, explain in the next Semi-i annual Radioactive Ef fluent Release Report why the inoperability was not corrected in a timely manner in accordance with

}

l Specification 6.9.1.d.4 and, i

b. When continuous monitoring is out of service daily grab samples will be taken and analyzed during normal operations and once per 4 hsors during degassing operations.

t

} 3. Containment Vessel via l j Plant Vent j a. Radionoble gas monitor 1 With the number of channels operable less than the MCO requirement:

i (RMS-12) provides automatic a. Exert best efforts to return the instruments to operable status termination of Containment within 30 days and. .sf unsuccessful, explain in the next Semi- +

Vessel releases upon annual Radioactive Efiluent Release Report why the inoperability exceeding alare/ trip was not corrected in a timely manner in accordance with i Setpoint. Specification 6.9.1.d.4 and, ,

l i I.

TABLE 3.5-7 (Osntinued)

RADICACTIVE GASEOUS EFFLUENT MONITORING INSTRIMENTATION Release Pathway / Instrumentation MCO* Required Action

3. Containment Vessel Via b. Ef fluent releases via this pathway may continue provided that Plant Vent (Continued) either of the Plant Vent Radionoble Gas Monitors (RMS-14 or RMS ~

is operable; otherwise, suspend all releases via this pathway. ,

I

b. Radioparticulate Monitor 1 With the number of channels operable less than the MCO requirement:

(RMS-II) provides automatic a. Exert best efforts to return the instruments to operable status i termination of containment within 30 days and, if unsuccessful, explain in the next Semi- t vessel releases exceeding annual Radioactive Ef fluent Release Report why the inoperability alare/ trip setpoints was not corrected in a timely menner in accordance with Specification 6.9.1.d.4 and,

b. Ef fluent releases via this pathway may continue provided that w either of the Plant Vent Radionob1Le Gas Monitors (RMS-14 or RMS *

'v, is operable; otherwise, suspend all releases via this pathway.

h

" c. Sampler flow rate monitor 1 With the number of channels operable less than the MCO requirement:

(RMS-ll) a. Exert best efforts to return the instrinnents to operable status i

within 30 days and. if unsuccesaful, explain in the next Semi-annual Radioactive Ef fluent Release Report why the inoperability was not corrected in a timely manner in accordance with '

Specification 6.9.1.d.4 and,

b. Ef fluent releases via this pathway may continue provided that the '

flow rate is estimated once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4. Condenser Vacuum Pump Vent

a. Radionoble gas monitor 1 With the number of channels operable less than the MCO requirement:

(RMS-15) diverts effluents a. Exert best efforts to return the instruments to operable status

! f rom Condenser Vacuta Pump within 30 days and, if unsuccessful, explain in the next Sesi-i Vent to the Plant Vent upon annus1 Radioactive Effluent Release Report why the inoperability i exceeding alare/ trip was not corrected in a timely manner in accordance with setpoint. Specification 6.9.1.d.4 and, l

TABLE 3.5-7 (Continued)

! RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRLMENTATION I

Release Pathway / Instrumentation MCO* Required Action

, 4. Condenser Vacuum Pump Vent (Continued)

! b. Ef fluent releases via this pathway may continue provided that;

1. Grab samples are collected occe per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and are j analyzed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> f or radionoble gases, or; i

2. The effluent is diverted to the Plant Vent and RMS-14 is operable.

l

b. Flow rate measuring devices 1 With the number of channels operable less than the NGO requirement:

(one for each Vacuum for a. Exert best efforts to return the instruments to operable status l Pump). each within 30 days and, if unsuccessf ul, explain in the next Semi-

." pump annual Radioactive Ef fluent Release Report why the inoperability l Y in was not corrected in a timely manner in accordance with

! o$ service Specification 6.9.1.d.4 and, I

b. Ef fluent releases via this pathway may continue provided the flow rate is estimated once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

} 6

5. Fuel Handling Building Lower l Level Exhaust Vent j

• a

a. Radionoble gas monitor 1 With the nusber of channels operable less than the MCO requirement:

(RMS-20) trips the exhaust a. Exert best efforts to return the instruments to operable status i j and supply f ans for the within 30 days and, if unsuccessful, explain in the next Semi-lower level of the Fuel annual Radioactive Effluent Release Report why the inoperability Handling Building upon was not corrected in a timely manner in accordance with exceeding alarm / trip ' Specification 6.9.1.d.4 and, s et point.

b. Ef fluent releases via this pathway may continue provided that grab samples are taken once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and analyzed for gross ,

i activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. [

i i

• MCO - Minimum Channels Operable I i 1 + l

TABLE 3.5-7 (Contirued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRLMENTATION Re lease Pathway / Instrumentation MCO* Baquired Action

5. Fuel Handling Building lower Levc1 Exhaust Vent (Continued)

b. Sampler flow rate monitor 1 With the number of channels operable less than the MCO requirement:

(RMS-20) a. Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Semi-annual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner in accordance with

. Specification 6.9.1.d.4 and,

b. Ef fluent releases via this pathway may continue provided the

." flow rate is estimated once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

y' U$

6. Fuel Handling Building Upper Level Exhaust Vent

a. Radionoble gas monitor 1 With the number of channels operable less than the MCO requirement:

(RMS-21) trips the exhaust a. Exert best efforts to return the instruments to operable status and supply f ans for the within 30 days and, if unsuccessful, explain in the next Semi-upper level of the Fuel annual Radioactive Effluent Release Report why the inoperability Handling Building upon was not corrected in a timely manner in accordance with exceeding alare/ trip Specification 6.9.1.d.4 and, setpoint.

b. Ef fluent releases via this pathway may continue provided that:

1. The Plant Vent Radionoble Gas Monitor (RMS-14) is operable,

2. Crab samples are collected once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and are analyzed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for radionoble gases.

• MCO - Minimum mannels Operable

'l

, TABLE 3.5-7 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRIMENTATION Release Pathway /Instrtamentation MCO* Required Action

6. Fuel Handling Building Upper Level Exhaust Vent (Continued)

, b. Sampler flow rate monitor 1 With the ntsaber of channels operable less than the MCO requirement:

(RMS-21) a. Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Semi-annual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner in accordance with Specification 6.9.1.d.4 and,

b. Ef fluent releases via this pathway asy continue provided the

." flow rate is estimated once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Y 8

• MCO - Minimum Qiannels Operable

3.9 RADIOACTIVE EFFLUENTS 3.9.1 Compliance With 10 CFR Part 20 - Radioactive Materials in Liquid Ef fluents

, Applicability Applies to radioactive material in liquid effluents released f rom the site to unrestricted areas.

Objective To define the concentration limits of 10CFR20 for radioactive material in liquid effluents released to unrestricted areas.

i Specification 3.9.1.1 The concentration of radioactive material in liquid effluents released at any time from the site to unrestricted areas (see Figure 1.1-1) shall be limited to the concentrations specified in i

10CFR20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to

~4 u CL/ml total activity.

2 x 10 3.9.1.2 With the concentration of radioactive material in liquid effluents released from the site to unrestricted areas exceeding the above limits, without delay restore the concentration to l within the above limits. In addition, a prompt notification must be made to the Commission in accordance with Specification 6.9.2(a)(10).

3.9.1.3 In the event that the immediate action required by 3.9.1.2 above cannot be satisfied, the facility shall be placed in hot shutdown l

l 3.9-1

within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, and entry into the power operating condition shall not be made unless Specification 3.9.1.1 is 'nat.

3.9.1.4 The provisions of Specifications 3.0 and 6.9.2.b(2) are not applicable.  ;

3.9.2 Compliance With 10 CFR Part 50 - Radioactive Materials in Idquid Effluents Applicability Applies to radioactive materials in liquid effluents released from the site to unrestricted areas.

Objective To define the calculated dose limits of 10CFR50 for radioactive materials in liquid effluents released to unrestricted areas.

Specification 3.9.2.1 The dose commitment at all times to a member of the public from radioactive materials in liquid effluents released to unrestricted areas (See Figure 1.1-1) shall be limited:

a. During any calendar quarter to <1.5 arem to the total body and to jg5 mrem to any organ, and

b. During any calendar year to <3 mrem to the total body and to 110 area to any organ.

3.9.2.2 With the calculated dose commitment from the release of radioactive materials in liquid effluents exceeding any of the limits prescribed by Specification 3.9.2.1 above, prepare and l submit a report to the Commission in accordance with l

l Specification 6.9.3.2.

3.9-2

3.9.3 Compliance With 10 CFR Part 20 - Radioactive Materials in Gaseous Effluents Applicability Applies to radioactive materials in gaseous effluents released from the site to unrestricted areas.

Objective To define the dose rate limits for radioactive materials in gaseous effluents released to unrestricted areas.

Specification 3.9.3.1 The dose rate due to radioactive materials in gaseous effluents released from t!.a site boundary (see Figure 1.1-1) shall be limited to the following:

a. For radionoble gases: <500 mrem /yr to the total body, 13000 mrem /yr to the skin, and

b. For all I-131, and tritium, and for all radioactive materials j in particulate form, inhalation pathway only, with half lives greater than 8 days: <1500 arem/yr to any organ.

l 3.9.3.2 With the dose rate (s) exceeding the above limits, without delay decrease the release rate to within the above limits. In addition, a prompt notification must be made to the Commission in accordance with Specification 6.9.2(a)(10).

I 3.9.3.3 In the event that the immediate action required by 3.9.3.2 above cannot be satisfied, the facility shall be placed in hot shutdown 3.9-3

within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, and entry into the power operating condition shall not be made until Specification 3.9.3.1 is met.

3.9.4 Compliance With 10 CFR Part 50 - Radionoble Gases Applicability Applies to radionoble gases released in gaseous effluents to unrestricted areas.

Objective To define the air dose limits of 10CFR50 for radionoble gases released in gaseous effluents to unrestricted areas.

] Specification r

3.9.4.1 The air dose commitment due to radionoble gases released in gaseous effluents to areas at and beyond the site boundary (See Figure 1.1-1) shall be limited, at all times, to the following:

s. During any calendar quarter, to f,5 mrad for gamma radiation and fl0 mrad for beta radiation; l

b. During any calendar year,'to j_10 mrad for gamma radiation and

<20 mead for beta radiation.

l r

3.9.4.2 With the calculated air dose commitment from radioactive noble gases in gaseous effluents exceeding any of the limits, prescribed by Specification 3.9.4.1 above, prepare and submit a report to the Commission in accordance with Specification 6.9.3.2.

e 3.9-4

3.9.5 Compliance With 10 CFR Part 50 - Radioiodines, Radioactive Materials in Particulate Form, and Radionuclideo Other Than Radionoble Gases Applicability 1 Applies to radioiodines, radioactive materials in particulate form, and radionuclides other than radionoble gases released from the site to unrestricted areas.

Objective To define the dose limits of 10CFR50 for radioiodines, radioactive materials in particulate form, and radionuclides other than radionoble gases released from the site to unrestricted areas.

Specification 3.9.5.1 The dose to a member of the public from I-131, tritium and radioactive materials in particulate form, with half-lives greater than 8 days in gaseous effluents released to unrestricted areas (See Figure 1.1-1), inhalation pathway only, shall be limited, at all times, to the following:

I

a. During any calendar quarter, <7.5 mrem to any organ and, i

b. During any calendar year, <15 mrem to any organ.

3.9.5.2 With the calculated dose commitment from the release of I-131, tritium and radioactive materials in particulate form, with half lives greater than 8 days, in gaseous effluents exceeding any of the limits prescribed by Specification 3.9.5.1 above, prepare and submit a report to the Commission in accordance with Specification 6.9.3.2.

3.9-5

I 3.9.6 Compliance With 40 CFR Part 190 - Radioactive Effluents From Uranium Fuel Cycle Sources Applicability Applies to radioactive effluents from uranium fuel cycle sources.

Objective To define the dose limits of 40CFR190 for radioactive effluents from uranium fuel cycle sources.

Specifications 3.9.6.1 The dose commitment to any member of the public, due to releases of licensed materials and radiation, from uranium fuel cycle sources shall be limited to ;gl5 mrem to the total body or any organ except the thyroid, which shall be limited to < 75 mrem over 12 consecutive months. This specification is applicable to Robinson Unit 2 only for the area within a five mile radius around the Robinson Plant.

3.9.6.2 With the calculated doses from the release of the radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specification 3.9.2.1.a, 3.9.2.1.b, 3.9.4.1.a.

3.9.4.1.b, 3.9.5.1.a or 3.9.5.1.b, calculations should be made including direct radiation contributions from the reactor units and from outside storage tanks to determine whether the above limits of Specification 3.9.6.1 have been exceeded. If such is the case in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.3.2.d, 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 20.405c, shall include an 3.9-6

l l

l l

I 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 same request is complete.

3.9.6.3 The provisions of Specifications 3.0 and 6.9.2.b(2) are not applicable.

Basis Compliance With 10 CFR Part 20 - Radioactive Materials in Liquid Effluents This specification is provided to ensure that the concentration of radioactive materials in liquid effluents released from the site to unrestricted areas will be less than the concentrations specified in 10 CFR Part 20, Appendix B, Table II. This limitation provides the additional assurance that the concentrations of radioactive materials in bodies of water outside the site will result in exposures within the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Ke-135 is the controlling radionuclide and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs).

Detailed discussion of the LLD, and other detection limits can be found in 3.9-7

HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A. , " Limits for Qualitative Detection and Quantitative Determination ~ Application to Radiochendstry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K.,

" Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

Compliance With 10 CFR Part 50 - Radioactive Materials in Liquid Effluents This specification is provided to implement the requirements of Sections II.A, and 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.A of Appendix I. The action statement provides the required operating flexibility and at the same time implements the guides set forth in Section IV.A of Appendix I of 10 CFR Part 50 to assure that the release of radioactive material in liquid effluents will be kept "as low as is reasonably achievable." The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculative procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.

The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in the Regulatory Guide 1.109, t " 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.

Compliance With 10 CFR Part 20 - Radioactive Materials in Gaseous Effluents I This specification is provided to ensure that the dose rate at any time at the site boundary from gaseous effluents from H. B. Robinson Unit No. 2 will be

, within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20 Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents 3.9-8

will result in the exposure of individuals outside the site boundary, to annual average concentrations within the limits specified in Appendix B Table II of 10 CFR Part 20, (10 CFR Part 20.106(b)). For individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary unrestricted area. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rate equivalents above background to an individual in unrestricted areas to ;[500 arem/ year to the total body or to <3000 mrem / year to the skin.

Compliance With 10 CFR Part 50 - Radionoble Gases This specification is provided to implement the requirements of Section II.B, III.A and IV.A of Appendix I, 10 CFR Part 50. The limiting condition for operation implementing the guides provides the required operating flexibility and at the same time implements 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 is reasonably achievable". The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculative procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.

The methods established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in the 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. The ODCM equations provided for determining the air dose commitments at the site boundary are based upon historical average atmospheric conditions.

3.9-9

Compliance With 10 CFR Part 50 - Radioiodines, Radioactive Materials in Particulate Form, and Radionuclides Other Than Radionoble Gases This specification is provided to implement the requirements of Section II.C, 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.C of Appendix I. The action statement provides the required operating flexibility and at the same time implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials as gaseous effluents will be kept "as low as reasonably achievable." The surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculative procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The methods i

established in the ODCM for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in

~

Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors",

Revison 1, J uly 1977. The ODCM equations provided for determining the commitment are based upon historical average atmospheric conditions.

Compliance With 40 CFR Part 190 - Radioactive Effluents From Uranium Fuel Cycle Sources This specification is provided to meet the dose limitations of 40 CFR Part 190 that have 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 generated radioactive effluents and direct radiation exceed 25 mrems to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. It is highly unlikely that the resultant dose to a member of the public will exceed dose limits of 40 CFR Part 190 if the reactor remains within twice the dose design objectives of Appendix I, and if direct radiation doses from the 3.9-10

reactor unit and outside storage tanks are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a member of the public to within the 40 CFR Part 190 limits.

For the purposes of the Special Report, it may be assumed that the dose commitment to the member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the done 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 3.9.1.1 and

~

3.9.3.1. An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

l 3.9-11

5 3.16 RADIOACTIVE WASTE SYSTEMS 3.16.1 Liquid Radwaste Treatment System Applicability Applies to the liquid radwaste treatment system.

Objective To define the operating requirements for the liquid radwaste treatment system.

Specification

~

3.16.1.1 The appropriate portions of the Liquid Radwaste Treatment System shall be maintained and used to reduce the concentrations of radioactive materials in liquid wastes prior to their discharge when the projected dose commitments, due to the release of radioactive liquid effluents to unrestricted areas (See Figure 1.1-1) when averaged over a calendar quarter, would exceed 0.2 mrem to the total body or 0.6 mrem to any organ.

3.16.1.2 With radioactive liquid wastes being discharged without treatment while in excess of the limits of Specification 3.16.1.1 above, prepare and submit a report to the Commission in accordance with Specification 6.9.3.2.b.

3.16.2 Liquid Holdup Tanks

• Applicability Applies to the liquid holdup tanks.

• Tanks included in this Specification are those outdoor tanks that are not surrounded by liners, dykes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to l the liquid radwaste treatment system. Tanks classed as " Seismic Class 1" are excluded from this Specification.

3.16-1

Objective To define the operating requirements for the liquid holdup tanks.

Specification 3.16.2.1 The quantity of radioactive material contained in each of the following tanks shall at all times be limited to 110 curies, excluding tritium and dissolved or entrained noble gases.

a. A monitor tank

b. B monitor tank

c. C Waste Condensate tank

d. D Waste Condensate tank

e. E Waste Condensate tank

f. Any Outside temporary tank

• 3.16.2.2 With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit, and the event should be described in the Semiannual Radioactive Effluent

! Release Report, Specification 6.9.1.d.

l 3.16.2.3 If Specification 3.16.2.2 is not completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> a prompt notification with written followup is required as per l Specification 6.9.2.a(10).

3.16.3 Gaseous Radwaste and Ventilation Exhaust Treatment Systems Applicability Applies to the gaseous radwaste and ventilation exhaust treatment systems.

l I

1

!

• A temporary tank is defined as any tank having a capacity of > 100 gallons ,

j used for the receipt or transfer of radioactive liquids.

3.16-2

Objective To define the operating requirements for the gaseous radwaste and ventilation exhaust treatment systems.

Specificat j_

3.16.3.1 The appropriate portions of the Gaseous Radwaste Treatment System and the Ventilation Exhaust Treatment System shall be maintained and used to reduce the concentrations of radioactive materials in gaseous wastes prior to their discharge when the projected dose commitments due to the release of gaseous effluents to unrestricted areas (See Figurn 1.1-1) when averaged over a calendar quarter would exceed:

a. 0.6 mrem for gamma radiation and 1.3 mrem for beta radiation due to radionoble gases or,

b. 1.0 mrem to any organ due to radioiodines, radioactive materials in particulate form, and radionuclides other than radionoble gases.

3.16.3.2 With the Gaseous Radwaste Treatment System and/or the Ventilation Exhaust Treatment System not operable and with radioactive gaseous wastes being discharged without treatment while in excess of the limits of Specification 3.16.3.1 above, prepare and submit a report to the Commission in accordance with Specification 6.9.3.2.b.

3.16.4 Waste Gas Decay Tanks (Hydrogen and Oxygen)

Applicability Applies to the volumetric hydrogen and oxygen concentration limits for the four Waste Gas Decay Tanks.

3.16-3

Objective To define operating requirements for the Waste Gas Decay Tanks.

Specification 3.16.4.1 The oxygen concentration in the four Waste Gas Decay Tanks should be limited to i 4% by volume when the hydrogen concentration in the same tank exceeds 4% by volume. The hydrogen concentration in the four Waste Gas Decay Tanks should be limited to i 4% by volume when the oxygen concentration in the same tank exceeds 4%

by volume.

3.16.4.1.a When the concentration of oxygen in a Waste Gas Decay Tank is

> 4% but i 6% by volume and the hydrogen concenti.stion in the same tank fa > 4% by volu'me, or the concentration of hydrogen in a Waste Gas Decay Tank is > 4% but i 6% by volume and the oxygen concentration in the same tank is > 4% by volume, restore one or j both to i 4% by volume within 48 hrs.

3.16.4.1.b When the concentration of oxygen in a Waste Gas Decay Tank is

> 6% by volume and the hydrogen concentration in the same tank is

> 4% by volume, or the concentration of hydrogen in a Waste Gas Decay Tank is > 6% by volume and the oxygen concentration in the same tank is > 4% by volume, immediately suspend all additions of j waste gas to the affected tank and immediately commence efforts l to lower the concentratica of one or both to i 4% by volume.

3.16.4.2 If the requirements of paragraph 3.16.4.1.a cannot be met within l the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> limit, submit a special report to the NRC within the 1

following 14 days which outlines the cause of the occurrence, i corrective actions taken to date, corrective actions which will

! be taken, and any compensatory actions being taken to minimize the potential hazard.

i l 3.16-4

3.16.4.3 If the actions taken to comply with paragraph 3.16.4.1.b do not reduce the concentration of hydrogen and/or oxygen in the affected tank to ;[ 6% by volume within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a prompt notification with written followup is required per specification 6.9.2.a(10). Once the concentration of hydrogen and/or oxygen in the affected tank is 16% by volume', paragraphs 3.16.4.1.a and 3.16.4.2 apply.

3.16.5 Waste Gas Decay Tanks (Radioactive Materials)

Applicability Applies to the four Waste Gas Decay Tanks.

Objective To define the operating requirements for the Weste Gas Decay Tanks.

Specification 3.16.5.1 The quantity of radioactivity contained in each Waste Gas Decay Tadk shall at all times be limited to j( 6.0E5 curies noble gases (considered as Xe-133).

l 3.16.5.2 With the quantity of radioactive materials in any Waste Gas Decay Tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit.

3.16.5.3 If Specification 3.16.5.2 is not completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, a prompt notification with written follow-up is required as per Specification 6.9.2.a(10).

1 3.16-5

3.16.6 Solidification of Wet Radioactive Waste Applicability Applies to the solidification of wet radioactive waste.

Objective To define the requirements for the solidification of wet radioactive waste.

Specification 3.16.6.1 The Solid Radwaste System shall be used in accordance with a Process Control Program (PCP) to process wet radioactive waste to meet shipping and burial ground requirements.

3.16.6.2 With the provisions of the PCP not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive waste from the site.

3.16.6.3 If any test specimen, as required by the PCP, fails to verify solidification, the solidification of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative solidification parameters can be determined i

in accordance with the PCP, and a subsequent test verifies l

i solidification. The PCP shall be modified as required in accordance with Section 6.15, and solidification of the batch may l

j then be resumed using alternative solidification parameters as

( determined by the PCP.

l Bases l

Liquid Radwaste Treatrent System l

l The requirements that the appropriate portions of this system be maintained and used when specified provides assurance that the releases of radioactive l

l 3.16-6

materials in liquid effluents will be kept "as low as reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and 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 the dose design objective set forth in Section II.A of Appendix I,10 CFR Part 50, for liquid effluents.

Liquid Holdup Tanks The tanks listed in this Specification include all those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank l

contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.

Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area.

I

! Gaseous Radwaste and Ventilation Exhaust Treatment Systems The requirements that the appropriate portions of these systems be maintained and used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and 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 the dose design objectives set forth in Section II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

l l

3.16-7

Waste Gas Decay Tanks (Hydrogen and Oxygen)

This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of hydrogen and oxygen. (Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen, automatic diversion to recombiners, or injection of dilutants to reduce the concentration below the flammability limits.) 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.

Waste Gas Decay Tanks (Radioactive Materials)

The tanks included in this specification are those tanks for which the

/

quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification to a quantity that is less than the quantity that provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to a member of the public at the nearest site boundary will not exceed 0.5 rem in an event of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> duration.

'estricting the quantity of redioactivity contained in cach ga

stor:ge tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to a member of the public at the nearest site boundary will not exceed 0.5 rem. This is consistent with Branch Technical Bosition ETSB 11-5 in NUREG-0800, July 1981.

Solidification of Wet Radioactive Waste This specification ensures that the packaging of wet radioactive wastes meets the requirements of 10 CFR Part 20 and 10 CFR Part 71 prior to their shipment from the site for disposal.

3.16-8

3.17 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 3.17.1 Monitoring Program Applicability .

Applies to the radiological environmental monitoring program.

Objective i

To define the requirements for implementation of the radiological environmental monitoring program.

Specification l

l 3.17.1.1 The Rad'iological Environmental Monitoring Program shall be conducted as specified in Table 3.17-1.

3.17.1.2 With the radiological environmental monitoring program not being conducted as specified in Table 3.17-1, in lieu of a Licensee Event Report, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.e, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

3.17.1.3 With the level of radioactivity as the result of plant effluents i in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.17-2 when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to

\

Specification 6.9.3.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive effluents so that the l

3.17-1

t potential annual dose

• to a member of the public is less than the calendar year limits of Specifications 3.9.2.1, 3.9.4.1, and 3.9.5.1. When more than one of the radionuclides in Table 3.17-2 are detected in the sampling medium, this report shall be I submitted if:

concentration (1) , concentration (2) + . . . > 1.0 i reporting level (1) reporting level (2)

When radionuclides other than those in Table 3.17-2 are detected and are the result of plant affluents, this report shall be

! submitted if the potential annual dose

• to a member of the public

! is equal to or greater than the calendar year limits of 2 Specifications 3.9.2.1, 3.9.4.1, and 3.9.5.1. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

3.17.1.4 With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 3.17-1, identify locations for obtaining replacement samples and add them to the rad'iological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program. In lieu of a Licensee Event Report and pursuant to Specification 6.9.1.d, identify the cause of the unavailability of samples and identify the new location (s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s).

l l

• The methodology and parameters used to estimate the potential annual dose to a member of the public shall be indicated in this report.

l 3.17-2 l

a 3.17.1.5 The provisions of Specifications 3.0 and 6.9.2.b(2) are not applicable.

3.17.1.6 Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, or to asifunction of automatic sampling equipment. If the latter,sev'ery effort shall be made to complete corrective action prior to the end of the next sampling period.

s 3.17.2 Land' Use Census I

s 4

Applicability L

Applies to the land use census.

Objective To define the requirements for the conduct of the land use census.

Specification z 3.17.2.1 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 16 meteorological sectors within a distance of five miles.

3.17.2.2 With a land use census identifying a location (s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.10.4.1, in lieu of a Licensee Event Report, identify the new location (s) in the next Semiannual Radioactive Effluent Release report, pursuant to Specification 6.9.1.d.6.

3.17-3

1 I

i 3.17.2.3 With the land use census identifying a location which yields an annual calculated dose or dose commiteent of a specific pathway which is 20% greater than that at a current sampling location:

(a) add the new location (s) to the radiological environmental monitoring program within 30 days and, (b) if desired, delete the sampling location having the lowest N calculated dose or dose commitments via the same exposure pathway, excluding the control station location, from the ,

monitoring program after October 31 of the year in which the Isod use census was conducted, and (c) identify the new location (s) in the next Semiannual '

Radioactive Effluent Release Report, Specification 6.9.1.d.4,

[(

l_, including a revised figure (s) and table for the ODCM reflecting the new location (s).

+

3.17.3 Interlaboratory Comparison Program Applicability N l. '

Applies to the inter 1Aboratory comparison program of like media.

i, f Objective l' >

To ensure precision and accuracy of laboratory analyses.

t l C

! Specification 3.17.3.1 Analyses shall be pe'rformed on radioactive materials supplied by y

j .

EPA as a part of an Interlaboratory Comparison Program of like media within' the ' environmental prograta as per Table 3.17-1.

i I

I l 3.17.3.2 With analyses not being performed as required above, report the 1

I corrective actions taken to prevent a recurrence to the

! 3.17-4 L . _ _ - _ _ _ _ _ _ . . _ _ _ _ _ . _ _ _ _ _ _ _ _,_ _ __, _ _

Commission in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.e.

3.17.3.3 The provisions of Specificationa 3 0 and 6.9.2.b(2) are not applicable.

3.17.3.4 The Interlaboratory Comparison Program shall be described in the ODCM. A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.e.

Basis Monitoring Program The radiological environmental monitoring program required by this specification provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides

, that lead to the highest potential radiation exposures of members of the i public resulting from the station operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the  ;

measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical R)sition

, on Environmental Monitoring. The initally specified monitoring program will be effective for at least the first three years of commercial operation.

Following this period, program changes may be initiated based on operational experience.

The required detection capabilities it . environmental sample analyses are ta' lated in terms of the lower limits of detection (LLD). The LLDs required by Table 3.17-3 are considered optimur for routine environgental measurements in-industrial laboratories. It should be recognized that the LLD is defined 3.17-5

- -.w-<_,-e- - , . - ., , - . . . ,--,--.~,-y . _ - , . , .~ . . , - -

-,c---e , ,-. ..,--n m-,. . . - - - - -

as an a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (af ter the fact) limit for a particular measurement.

Detailed discussion of the LLD, and other detection limits, can be found in NASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K.,

i " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

Land Use Census This specification is provided to ensure that changes in the use of areas at and beyond the Site Boundary are identified and that modifications to the monitoring program are made if required by the results of the census. 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, Revision 1 for consumption by a child. To determine this minimum garden size, the following assumptions 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.

! Interlaboratory Comparison Program The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring *in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.

3.17-6

1 TABLE 3.17-1 d

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM i

Number of Re presentative Exposure Pathway Samples and Sampling and Type and Frequency and/or Sample Sample Locations Collection Frequency of Analysis S

1. DIRECT RADIATION 33 routine monitoring stations Quarterly Gamma dose quarterly.

I with two or more dosimeters or with one instrtament f or

measuring and recording dose .

j rate continuously, placed

as follows

j an inner ring of stations, one in each of the 16 meteorological sectors

.in the general area of the site boundary;

.'d an outer ring of stations, one in

, O each of the 16 meteorological sectors L in the 6- to 8-km range from the site; b

area to serve as a contro1 station.

2. AIRBORNE Radiciodine and Samples f rom 5 locations Continuous sampler Radioiodine Cannister:

Particulates operation with sample 1-131 analysis weekly.

. 3 samples f rom close collection weekly, or to the 3 site boundary more frequently if locations, in different sectora, required by dust Particulate Sampler:

of the highest calculated loading. Gross beta radioactivity annual average analysis following

' groundlevel D/Q. filter change;c Gamma isotopic analysis d I sample from the vicinity of composite (by j of a commanity having the location) quarterly.

' highest calculated annual average groundlevel D/Q.

! I sample from a controi b

, location, as f or example 15-30 km

, distant and in the least preva-

) lent wind direction.

1

TABLE 3.17-1 (Continued)

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

3. WATERBORNE

a. Surface

• I sample upstream control Compositesamp}eover Gamma isotopic analysis d location 1-month period monthly. Composite for 1 sample downstream tritium analysis quarterly,

b. Groundg 2 samples Quarterly Camma isotopied and tritium analysis quarterly.

c. Sediment I sample from downstream area Semiannually Gamma isotopic analysis d from with existing or potential semiannually.

shoreline recreational value F 4. INGESTION O a. Milk I sample from milking animals Semimonthly when Gamma'isotopie d and I-131

$ within 5 km distance having animals are on analysis semimonthly when i the highest dose potential, pasture, monthly at animals are on pasture; If there are none, then, other times monthly at other times.

I sample from milking animals between 5 to 8 km distant where doses are calculated to be greater than 1 mrem per year h, I sampling from milkigg animals at a control location 15-30 km distant and in the least prevalent wind direction.

b. Fish I sample of each recreationally Semiannually Gamma isotopic analysis d important species in vicinity on edible portions of plant discharge area. semiannually.

I sample of same species in areas not influenced by plant discharge to serve as control location.b I

l

TABLE 3.17-1 (Continued)

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

c. Food 1 sample of each principal At time of harvest i

Gamma isotopic analyses d Products class of food products from on edible portion any area that is irrigated by water in which liquid plant wastes have been discharged.

d Samples of 3 different kinds Monthly when Gamma isotopie and 1-131 of broad leaf vegetation grown available analysis.

nearest each of two different offsite locations of highest g predicted annual average ground-

+

level D/Q if milk sampling is y not performed.

e d

1 sample of each of the similar Manthly when Gamma isotopie and I-131 broad leaf vegetation grown available analysis.

15-30 km distant in the least prevalent wind direction if milk -

sampling is not performed.

TABLE NOTATION eOne or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this

, table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation.

b The purpose of this sample is to obtain background information.

! TABLES 3.17-1 (Continued) l TABLE NOTATION

, cAirborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more af ter campling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples i is greater than ten times the yearly mean of control samples, gamma isotopic analysis shall be performed en the individual samples.

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

• The " upstream sample" shall be taken at a distance beyond significant influence of the discharge. The i " downstream" sample shall be taken in an area beyond but near the mixing zone.

1 f

A cosposite sample is one in which the quantity (aliquot) of liquid sampled is proportional to the quantity cf flowing liquid and in which the method of sampling employed results in a specimen that is representative l of the liquid flow. In this program cosposite sample aliquots shall be collected at time intervals that are very short (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining

. a representative sample.

i ss

' L EG roundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas

! where the hydraulic gradient or recharge properties are suitable for contamination.

! h The dose shall be calculated f or the maximum organ and age group, using the methodology and parameters in i

the ODCM.

I If harvest occurs more than once a year, sampling shall be performed during each discrete harvest. If harvest occurs continuously, sampling shall be monthly. Attenion shall be paid to including samples of tuborous and root food products.

i, i

1 l

1 f

TABLE 3.17-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIROIMENTAL SAMPLES Water Airborge Fish Milk Food Products Radionuclide (pCi/1) (pCi/a ) (pCi/ Kg, wet) (pCi/1) (pCi/ Kg,we t)

H-3 3E+04 Mn-54 IE+03 2+04 Fe-59 4E+02 1E+04 I Co-58 1E+03 2+04 Co-60 3E+02 1E+04 Zn-65 2+02 2E+04 w

Zr-Nb-95 4E+02 w

I-131 2E+00 9E-01 E+00 1E+02 Cs-134 3E+01 1E+01 1E+03 6E+01 1E+03 Cs-137 5E+01 2E+01 2E+03 7E+01 2E+03 Ba-La-140 2E+02 3E+02 1

TABLE 3.17-3 MAXIMIM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)a Water Airborge Fish Milk Food Products Sediment Analysis (pCi/1) (pCi/m ) (pCi/ Kg, wet) (pC1/1) (pC1/ Kg, wet) (pCi/ Kg, dry) gross beta 4E+00 IE-02 H-3 3E+03 l

141- 5 4 1.SE+01 1. 3E+02 Fe-59 3E+01 2.6E+02 Co-58,60 1. SE+01 1.3E+02 1

,u Zn-65 3E+01 2.6E+02 w

Y Zr-Nb-95 1.5E+01 b

I-131 IE+00 7E-02 IE+00 6E+01 Cs-134 1.5E+01 SE-02 1.3E+02 1.5E+01 6E+01 1.5E+02 Cs-137 1.8E+01 6E-02 1. SE+02 1. 8E+01 8E+01 1.8E+02 Ba-La-140 1.5E+01 1.5E+01 l

1 TABLE 3.17-3 (Continued)

TABLE NOTATION  ;

"The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of f alsely concluding that a blank observation represents a "real" signal.

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

4.66 s b E +

V 2.22 + Y + exp(- A a t)

Where:

LLD is the "a priori" lower limit of detection as defined above, as picoeuries 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, 2.22 is the ntaber of disintegrations per minute per picocurie, Y is the fractional radiochemical yield, when applicable, A is the radioactive decay constant for the particular radionuclide, and i.c for environmental samples is the elapsed time between sample collection, or and of the sample collection period, and time of counting Typical values of E, V, Y, and 4t should be used in the calculation.

It should be recognized that the LLD is defined as a priori (before the fact) limit representing the capability of a measure _ ment system and not as an a posteriori (af ter the f act) limit f or a particular measurement.

AnsTysis shall be performed in such a manner that the stated LLDs 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 LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.e.

b LLD for drinking water samples. If no drinking water pathway exists, the LLD of ganana isotopic analysis any be used.

3.17-13

4 t

TABLE 4.1-1 (Continued)

I Channel Description Check Calibrate Test Remarks i

! 9. Analog Rod Position S (1,2) R M (1) With step counters (2) Following rod motion in

excess of six inches

! when the computer is out of service ,

10. Rod Position Bank Counters S (1,2) N.A. N.A. (1) Following rod motion in ex-cess of six inches when the

computer is out of service (2) With analog rod position

11. Steam Generator Level S R M

. t.

y 12. Charging Flow N.A. R N.A.

ui

13. Residual Heat Removal Pump Flow N.A. R N.A.

14. Boric Acid Tank Level D (1) R N.A. (1) Bubbler tube rodded weekly

15. Refueling Water Storage W R N.A.

Tank Level i 16. Boron Injection Tank Level W R N. A .

i 17. Volume Control Tank Level . N.A. R N.A.

i

18. Containment Pressure D R B/W (1) (1) Containment isolation valve

. signal i 19. Deleted by Amendment No.

20. Boric Acid Hakeup Flow Channel N.A. R N.A.

i f

4.10 RADIOACTIVE EFFLUENTS 4.10.1 Radioactive Liquid Effluents Applicability Applies to the monitoring of radioactive liquid efflueats.

Objective To ascertain that radioactive liquid effluent releases are being maintained as low as reasonably achievable and within allowable limits.

Specification 4.10.1.1 The radioactivity content of each batch of radioactive liquid waste to be discharged shall be determined prior to release by sampling and analysis in accordance with Table 4.10-1. The results of pre-release analyses shall be used with the calculative methods in the ODCM to assure that the concentration at the point of release to the unrestricted area is maintained within the limits of Specification 3.9.1.1.

4.10.1.2 Analyses of samples composited from batch releases shall be performed in accordance with Table 4.10-1. The results of the post-release analyses shall be used with the calculative methods l in the ODCM to assure that the concentrations at the point of release were maintained within the limits of Specification 3.9.1.1.

. 4.10.1.3 The concentration of radioactive materials in liquid effluents discharged from continuous release points shall be determined by collection and analysis of samples in accordance with Table 4.10-1. The results of the analyses shall be used with the 4.10-1

l calculative methods in the ODQd to assure that the concentrations at the point of release are maintained within the limits of Specification 3.9.1.1.

4.10.1.4 Dose Calculations: Cumulative dose commitments for the current calendar quarter and calendar year from liquid effluents shall be determined in accordance with the ODCM once per 31 days.

4.10.2 Radioactive Gaseous Effluents Applicability Applies to the monitoring of radioactive gaseous effluents.

Objective To ascertain that radioactive gaseous effluent releases are being maintained as low as reasonably achievable and within allowable limits.

Specifications 4.10.2.1 The dose rate due to radioactive materials in gaseous effluents shall be determined to be within the limits of

! Specification 3.9.3.1 in accordance with the methods and procedures of the ODCH by obtaining representative samples and t

performing analyses in accordance with the sampling and analysis I program specified in Table 4.10-2.

I 1

4.10.3 Radionoble Gases Applicability Applies to the determination of cumulative doses from radionoble gases.

(

4.10-2 i

l l

I Objective l

To ascertain that cumulative doses from radionoble gases are being maintained as low as reasonably achievable and within allowable limits.

Specification 4.10.3.1 Cumulative dose commitments for the current calendar quarter and current calendar year shall be determined in accordance with the ODCM once per 31 days.

4.10.4 Radioiodines, Radioactive Materials in Particulate Form, and Radionuclides Other Than Radionoble Gases Applicability Applies to the determination of cusslative doses from radioiodines, radioactive materials in particulate form, and radionuclides other than radionoble gases.

Objective To ascertain that cumulative doses from radioiodines, radioactive materials in particulate form, and radionuclides other than radionoble gases are maintained i as low as reasonably achievable and within allowable limits.

l Specification 4.10.4.1 Cumulative dose contributions for the current calendar quarter and current calendar year for iodine-131, tritium, and radionuclides in particulate form with half lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

4.10-3 l

4.10.5 Radioactive Effluents From Uranium Fuel Cycle Sources Applicability Applies to the determination of cumulative doses from radioactive effluents from uranium fuel cycle sources.

Objective To ascertain that cumulative doses from radioactive effluents from uranium fuel cycle sources are maintained as low as reasonably achievable and within allowable limits.

Specification ,

4.10.5.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 3.9.2.1, 3.9.4.1, and 3.9.5.1 in accordance with the methodology and parameters in the ODCM. For the purposes of this Surveillance Requirement , it may be assumed that fuel cycle sources are negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered. In addition, an individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation which is part of the nuclear fuel cycle.

4.10.5.2 Cumulative dose contributions from direct radiation from the reactor units and from radwaste storage tanks shall be determined in accordance with the methodology and parameters in the ODCM.

This requirement is applicable only under conditions set forth in Specification 3.9.6.2.

i l

l 4.10-4

TABLE 4.10-1 (continued)

TABLE NOTATION

a. The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of f alsely concluding that a blank observation represents a "real" signal.

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

4.66 a b LLD =

E

• V

• 2. 22 x 10 6 , y , exp (- A At)

Where:

LLD is the "a priori" lower limit of detection as defined above, as microcuries 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. 22 x 106 is the number of disintegrations per minute per microcurie, l

Y is the f ractional radiochemical yield, when applicable,

, A is the radioactive decay constant f or the particular radionuclide, and l

at for plant effluents is the elapsed time between the midpoint of i

sample collection and time of counting.

4.10-6

TABLE 4.10-1 (continued)

TABLE NOTATION Typical values of E, V, Y, and at should be used in the calculation.

It should be recognized that 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 (af ter the f act) limit f or a particular measurement.

b. A batch release is the discharge of liquid wastes of a discrete volume.

Prior to sampling for analyses each batch shall be isolated and thoroughly mixed whenever possible, to assure representative sampling. Residual liquids in systems such as f eedwater heaters and lines cannot be thoroughly mixed for representative samples of their respective system.

Grab samples f rom these systems will be accepted as representative of their respective system.

c. The principal gamma emitters f or 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.

d. 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.

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

f. Grab sample of continuous flows taken f or compositing purposes will be taken in volumes proportional to the existing flow rate of the system in a manner described in the ODCM.

4.10-7

TABLE 4.10-2 RADICACTIVE GASEOUS WASTE SAWLING AND ANALYSIS A0 GRAM Ty pe of Samp l i ng Mi nimum Analysis Required Activity Required LLDs l Release Frequency Frequency Ana lysl s u Cl/mi !

Weste Gas P P Principal Gamma De cay Ta nks Emi tterse IE-04 Contal rune nt We We Principai Gamne Pressure Grab Saglob on Grab Sample Emi tterse IE-04 Ro ll ef s a nd Co ntal nne nt Purgos Tri tI um 1E-06 Continuous Me , g, h Me Prlnctpai Gamme Re leases Grab Sagle on Grab Sample Emitterse IE-04 f or Redlonoble 1 Plant Geses and Vent Tri tlun Tel tl un I E-06

2. Condenser Co nti nuousd Wf I-131 lE-12 Air Ejector Radi ol odi ne on Sample vent I f S/G Sanc i o Activity is

> l x 10-4 Conti nuousd Wf Principal Gamne ICl/cc Parti culate on Sanple Emi tterse IE-11 condenser Samp l e off gas is rated to Conti nuousd Q plant vent Pa rti cui ate on Cong ost to Sr-89, Sr-90 1E-11 Senples to be Conp osited M

on Conp osi te Alpha 1E-11 Co nti nuous Noble Gas Noble Gases Gross Z-5 Monitor Beta and Gamme p Cl/cc 4.10-8

l i

TABLE 4.10-2 (Continued)

TABLE NOTATION

a. Lower Limit of Detection (LLD) is an "a priori" limit representing the capability of a measurement system. LLD is calculated in accordance with methodology established in the ODCM and Table 4.10-1, Note a.

b. Containment pressure reliefs and purges can be made during the week without sampling by correcting the weekly sample analysis results with the ratio of the Containment Radionoble Gas Monitor (RMS-12) and the Containment Particulate Monitor (RMS-11) readings at the time of sampling to the desired time of the pressure relief.

c. The principal gamma emitters for which the LLD specification applies l exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xa-138 for gaseous emissions, I-131 for halogen emissions, and Mn-54, Fe-59, Co-58, Co-60, 2n-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 measureable and identifiable, together with the above nuclides, shall also be identified and reported.

d. 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.

In addition, these continuous samples are not required for the Condenser Vacuta Pump Vent.

e. Sampling and analysis shall also be performed f ollowing shutdown, startup, or a power change exceeding 15 percent of rated power within one hour unless (1) analysis shows that the dose equivalent I-131 concentration in the primary coolant has not increased more than a f actor of 3; and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3.

I 4.10-9

TABLE 4.10-2 (continued)

TABLE NOTATION

f. Samples shall be changed once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> af ter changing (or af ter removal f rom sampler) . Sampling and analyses shall also be performed once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for 7 days f ollowing shutdown, start-up or thermal power level change exceeding 15%

of rated thermal power in one hour and if I-131 Dose Equivalent in the RCS is greater than 0.1 pCL/cc. When samples collected f or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLD's may be increased by a f actor of 10. The analyses shall be performed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

g. Tritium grab samples shall be taken at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the refueling canal is flooded.

h. Tritium grab samples shall be taken at least once per 7 days f rom the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool.

4.10-10

! . _ _ _ . _ _ . _ .__ _ . _ . _ _ . _ - _ - . - . _ _.- _ . - - _ _ - - - _ - - - - - . . . - - . ~ . - - . -

4.18 RADIOACTIVE EFFLUENT INSTRUMENTATION 4.18.1 Radioactive Liquid Effluent Instrumentation Applicability Applies to the radioactive liquid effluent instrumentation system.

Objective To ascertain that the radioactive liquid effluent instrumentation system is functioning properly in order to accurately monitor radioactive lic,uid effluent releases.

Specification 4.18.1.1 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 4.18-1.

4.18.2 Radioactive Gaseous Effluent Instrumentation Applicability Applies to the radioactive gaseous effluent instrumentation system.

Objective To ascertain that the radioactive gaseous effluent instrumentation system is functioning properly in order to accurately monitor radioactive gaseous effluent releases.

4.18-1

Specification 4.18.2.1 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 at the frequencies shown in Table 4.18-2.

9 4.18-2

TABLE 4.18-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS mannel Source Channel mannel Pathway / Instruments Check Check Calibratiort Functional Test

1. Liquid Radwaste Ef fluent Line

a. Monitor (RMS-18) D P R (Note 3) Q (Note 4)

b. Flow rate measurement device (Note 1) N. A. R N.A.

2. Steam Generator Blowdown Effluent Line

a. Monitor (RMS-19) D M R (Note 3) Q (Note 4)

b. Flow rate measurement devices for measuring flew of sample to RMS-19 (Note 2) N.A. N.A. N.A.

g c. Flow rate measuring devices for each L steam generator blowdown line (Note 2) N. A. R N.A.

3. Containment Fan Cooling Water Monitor (Service Water Ef fluent Line)

a. Fbnitor (RMS-16) D M R (Note 3) Q (Note 5)

4. Tank Level Indicating Devices

a. Refueling Water Storage Tank D* N.A. R Q

b. Monitor Tanks A & B D* N.A. R Q

c. Waste Condensate Tanks C D&E D* N.A. R Q

• During liquid additions to the tank t

NOTES TO TABLE 4.18-1 Note 1 - The channel check shall consist of verifying indication of flow at least once during each batch type release or shall consist of verifying indication of flow at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for continuous type releases.

No te 2 - The channel check shall consist of verifying indication of flow at least once during each batch type release or shall consist of verifying indication of flow at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> f or continuous . releases, except during steam generator drain at cold shutdown.

No te 3 - The channel calibration shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained f rom suppliers that participate in measurement assurance activities or otherwise NBS traceable.

No te 4 - The Channel Functional Test shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the f ollowing conditions exists:

1. Instrument indicates measured levels above the alarm / trip setpoint.

2. Power f.silure.

3. Instrument controls not set in operate mode.

Note 5 - The Channel Functional Test shall also demonstrate that control room alarm annunciation occurs if any of the f ollowing conditions exists:

1. Instrument indicates measured levels above the alarm setpoint.

2. Power failure.

3. Instrument indicates a downscale f allure.

4. Instrument controls not set in operate mode.

NOTATION P Completed prior to making a radioactive materials release l D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> W At least once per 7 days N.A, Not applicable M At least once per 31 days R At least once per 18 months Q At least once per 92 days l

1 4.18-4

l. . _ _ _ - _-_ - _ _ _ _ _ ~ _ _ . _ _ _. . - - , . _ - _ . _ _ _ ..- ---. - ,_ - - ~ _ _ . - - - _ _

TABLE 4.18-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMEKIS Channel Source Channel Channel Pathway / Instruments Check Check Calibration Functional Test

1. Plant Vent

a. Radionoble gas monitor (RMS-14) P P (Note 5) R Q (Note 6)

b. Radionoble gas monitor (RMS-34) D M R(Note 2) Q

c. Radioiodine monitor (RMS-34) W M R(Note 3) Q

d. Radioparticulate monitor (RMS-34) W M R Q

e. Sampler flow rate monitor (RMS-34) D(Note 1) N.A. R Q

f. Plant Vent flow rate monitor D(Note 1) N.A. R N.A.

, 2. Containment Vessel via Plant Vent I

a. Radioparticulate Monitor (RMS-II) D D R(Note 2) Q 4

b. Radionoble gas monitor (RMS-12) D P (Note 4) R(Note 2) Q

c. Sampler flow rate monitor (RMS-12) D N. A. R Q j 3. Condenser Vacuum Pump Vent

a. Radionoble gas monitor (RMS-15) D M R(Note 2) Q (Note 6)

b. Flow rate measuring devices - one for each Vacutan Pump D(Note 1) N.A. N.A. N.A.

4. Fuel Handling Building Lower Level i Exhaust Vent

i. a. Radionoble gas monitor (RMS-20) D M R(Note 2) Q 4

b. Sampler flow rate monitor (RMS-20) D(Note 1) N.A. N.A. N.A.

TABLE 4.18-2 (Continutd)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRIMENTATION SURVEILLANCE REQUIREMENTS Channel Source Channel Channel Pathway / Instruments Check Check Calibration Functional Test

5. Fuel Handling Building Upper Level Exhaust Vent

a. Radionoble gas monitor (RMS-21) D M R(Note 2) Q

b. Sampler flow rate monitor (RMS-21) D(Note 1) N.A. N.A. N.A.

6. Waste Gas Holdup System

a. Hydrogen Monitor D N.A. Q (Note 8) N. A.

. b. Oxygen Monitor D N.A. Q (Note 8) N.A.

5 a

l l

I NOTES TO TABLE 4.18-2 Note 1 -

The channel check shall consist of verifying indication of flow whenever plant conditions dictate that flow is supposed to be present.

No te 2 - The channel calibration shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that hava been obtained f rom suppliers that participate in measurement assurance activities or otherwise NBS traceable.

I Note 3 - The channel calibration shall consist of changing the filter and cartridge at the frequency indicated and performing appropriate analyses with NBS traceable calibrated analytical equipment.

Note 4 - Prior to each containment release.

J No te 5 - Prior to each Waste Gas Decay Tank release.

Note 6 - The Channel Functional Test shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm / trip setpoint.

2. Power f ailure.

3. Instrument controls not set in operate mode.

Note 7 - The Channel Functional Test shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm setpoint.

2. Power failure.

3. Instrument indicates a downscale f allure.

4. Instrument controls not set in operate mode.

l l No te 8 - The Channel Calibration shall include the use of standard gas ,

l samples containing a nominal 3% oxygen, balance nitrogen and 4%

hydrogen, balance nitrogen or as recommended by manufacturer.

l l

4.18-7

4.19 RADIOACTIVE WASTE SYSTEMS 4.19.1 Liquid Radwaste Treatment System Applicability, Applies to the liquid waste treatment system.

Objective y To ascertain that the concentration of radioactive materials in the liquid ,

waste treatment system is maintained as low as reasonably achievable and s within allowable limits.

Specification 't ,

4.19.1.1 Dose commitments from liquid releases shall be projected at least once per 31 days, in accordance with the.0DCM to ensure the provisions of Specification 3.16.1.1 are satisfied when the Liquid Radwaste Treatment'$ystem is not in use.

4.19.2 Liquid Holdup Tanks

• Applicability Applies to liquid holdup tanks.

\

Objective

- s g

'\

To ascertain that the quantity of radioactive material contained in the liquid holdup tanks is maintained as low as reasonably achievable and within allowable limits. .

N

• Tanks included in this Specification are those outdoor tanks that are not l surrounded by liners, dykes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system. Tanks classed as " Seismic Class 1" are excluded from this Specification. ,-

s 4.19-1 .

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. _ _ . - - ~ , __ _... _ _ . . , ._ _ , . _ __ _. _ _ _ _ _ _ _ _ , _t

4 1

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Specification; 4.19.2.1 The quantity of radioactive material contained in each of the tanks listed in Specification 3.16.2.1 shall be determined to be within:the limits'specified in Specification 3.16.2.1 by either analyzing a representative sample of the tank's content at least

\\

once per 7 days when radioactive materials are being added to the tank or\ sampling the evaporator output when adding it to the tank. \

s s 4.19.3 Gaseous Radwaste and Ventiliation Exhaust Treatment System Applicability Applies to the gaseous radwaste and ventilation exhaust treatment system.-

~, Objective ,,

i To ascertain that the concentration of radioactive materials in the gaseous radwaste and ventilation exhaust treatment systems is maintained as low as reasonably achievable and within allowable limits.

• Specification i 4.19.3[1 Dose commitments due to gaseous releases shall be projected at s

'\- least once per 31 days, in accordance with the ODCH to ensure the

!, provisions of Specification 3.16.3.1 are satisfied.

f \ N l

4.19.4 Waste Gas Decay Tanks (Hydrogen and Oxygen)

I Applicability l

Applies to the Waste Gas Decay Tanks.

6 4.19-2 l

l ..

_ 2% , . _ t- ~ _ _ -- _ ._ . .

T Objective To ascertain that the concentration of hydrogen and oxygen in the Waste Gas Decay Tanks is maintained as low as reasonably achievable and within allowable limits. -

Specification 4.19.4.1 The concentration of hydrogen and oxygen in the Waste Gas Decay Tanks shall be determined to be within the limits specified in Specification 3.16.4.1 by monitoring the waste gases in the Waste Gas Decay Tanks with, the hydrogen and oxygen monitors or monitoring procedures required operable by Table 3.5-7 of Specification 3.5.3.1.

4.19.5 Waste Gas Decay Tanks (Radioactive Material)

Applicability Applies to the Waste Gas Decay Tanks.

Objective To ascertain that the quantity of radioactive material in the Waste Gas Decay Tanks is maintained as low as reasonably achievable and within allowable limits.

Specification 4.19.5.1 With the primary coolant activity >100 u C1/mi the quantity of radioactive material contained in each Waste Gas Decay Tank shall be determined to be within the limit specified in Specification 3.16.5.1 once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when radioactive materials are being  :

added to the tank.

4.19-3

4.19.6 Solidification of Wet Radioactive Waste Applicability Applies to the solidification of wet radioactive waste.

Objective To ascertain that wet radioactive vaste is solidified to meet the requirements of 10CFR20, 10CFR71, and burial ground requirements.

Specification 4.19.6.1 The PCP shall be used to verify the solidification of one representative test specimen from every tenth batch of wet radioactive waste.

4.19.6.2 If the initial test specimen from a batch of waste fails to verify solidification, the Process Control Program shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least 3 consecutive initial test specimens demonstrate solidification. The Process Control Program shall be modified as required, as provided in Specification 6.15, to assure .

6 solidification of subsequent batches of waste.

i i

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(

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4.19-4 L

4.20 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 4.20.1 Monitoring Program Applicability Applies to the radiological environmental monitoring program.

Objective To ascertain that radiological environmental monitoring samples are collected and analyzed in accordance with the radiological environmental monitoring program.

Specification 4.20.1,.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.17-1 from the locations defined in the ODCM and shall be analyzed pursuant to the requirements of Tables 3.17-2 and 3.17-3.

4.20.2 Land Use Census Applicability Applies to the land use census.

Objective To ascertain that the land use census is conducted in accordance with the radiological environmental monitoring program.

4.20-1

Specification 4.20.2.1 The land use census shall be conducted once per 12 months during the growing season, by door-to-door survey, aerial survey, by consulting local agriculture authorities or by broad leaf vegetation sampling of at least three different kinds of vegetation. This sampling may be performed at the site boundary in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census. Specifications for broad leaf vegetation sampling in Table 3.11-1.4c shall be followed, including analysis of control samples.

4.20.3 Interlaboratory Comparison Program Applicability Applies to the Interlaboratory Comparison Program of like media.

Objective To ensure precision and accuracy of laboratory analyses.

j Specification l

t 1

1 4.20.3.1 Analyses shall be performed on radioactive materials supplied by EPA as a part of Interlaboratory Ccmparison Program of like media within the environmental program as per Table 3.17-1 and pursuant to Specifications 3.17.3.2, 3.17.3.3, and 3.17. 3.4 l

i I

i 4.20-2

r 1 1

6.5 REVIEW AND AUDIT 6.5.1 The license organization's review and approval process shall assure that the nuclear safety of the facility is maintained.

6.5.1.1 Procedures, Tests. and Experiments 6.5.1.1.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below: )

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Rev. 2, February 1978.

b. Refueling operations.

c. Surveillance and test activities of safety-related equipment.

d. Security Plan ^ implementing procedures.

e. Emergency Plan implementing procedures.

f. Fire Protection Program implementing procedures.

g. Radiological Environmental Monitoring Program implementing procedures.

h. Offsite Dose Calculation Manual implementing procedures.

1. Process Control Program implementation procedure.

J. Quality Assurance Program for effluent and enviror. mental monitoring (using the guidance in Regulatory Guide 4.15, December 1977).

6.5-1

1 6.5.1.1.2 A safety analysis shall be prepared for all procedures, tests, and experiments covering the activities identified in 6.5.1.1.1 and procedures that affect nuclear safety. The analysis shall include a written determination of whether or not the procedure.

test, or experiment is a change in the facility as described in the FSAR, involves a change to the Technical Specification, or constitutes an unreviewed safety question as defined in 10CFR50.59(a)(2). This analysis constitutes a first party safety review and may be accomplished by the individual who prepared the document.

l I

l r

l 6.5-la

6.5.1.6.5 A quorum of the PNSC shall consist of the Chairman, and three members, of which two may be alternates.

6.5.1.6.6 The PNSC activities shall include the following:

a. Perform an overview of Specifications 6.5.1.1. and 6.5.1.2 to assure that processes are effectively maintained.

b. Performance of special reviews, investigations, and reports thereon requested by the Manager - Corporate Nuclear Safety.

c. Annual review of the Security Plan and Emergency Plan.

d. Perform reviews of Specifications 6.5.1.1.6, 6.5.1.2.4, 6.5.1.3.1, and 6.5.1.4.1.

e. Perform review of all events requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to the NRC.

f. Review of facility operations to detect potential nuclear safety hazards.

g. Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recommendations and dispostion of the corrective action to prevent recurrences to the Vice President - Nuclear Operations, Manager - Corporate Nuclear Safety and the Manager - Corporate Ouality Assurance.

h. Review of changes to the Process Control Program and the Offsite Dose Calculation Manual.

6.5-7

I 6.5.1.6.7 In the event of disagreement between the recommendations of the Plant Nuclear Safety Committee and the actions contemplated by the General Manager, the course determined by the General Manager to be more conservative will be followed. The Vice President -

Nuclear Operations and the Manager - Corporate Nuclear Safety will be notified within the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the disagreement and subsequent actions.

6.5.1.6.8 The PNSC shall maintain written minutes of each meeting that, at a minimum, document the results of all PNSC activities performed under the provisions of these Technical Specifications; and copies shall be provided to the Vice President - Nuclear Operations, and to the Manager - Corporate Nuclear Safety.

l i

6.5-7a

(4) The verification of compliance and implementation of the requirements of the Ouality Assurance Program to meet the criteria of Appendix B, 10CFR50, at least once per 24 months.

(5) The Emergency P1,an and implementing procedures at least once per 12 months.

(6) The Security Plan and implementing procedures at least once per 12 months.

(7) The Facility Fire Protection Program and implementing procedures at least once per 24 months.

(8) Any other area of facility operation considered appropriate by the Corporate Ouality Assurance Performance Evaluation Unit; the Executive Vice President - Power Supply and Engineering & Construction; or the Senior Vice President -

Power Supply.

(9) The Radiological Environmental Monitoring Program and the results thereof at least once per 12 months.

(10) The Offsite Dose Calculation Manual and implementing procedure at least once per 24 months.

(11) The Process Control Program and implementing procedures for solidification of radioactive wastes at least once per 24 months.

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

e. Distribute reports and other records to appropriate managers.

6.5-14

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

6.5.3.3 a. Audit personnel shall be independent of the area audited.

Selection for auditing assignments is based on experience or training that establishes that their qualifications are commensurate with the complexity or special nature of the activities to be audited. In selecting auditing personnel, consideration shall be given to special abilities, specialized technical training, prior pertinent experience, personal characteristics, and education.

l

b. Qualified outside consultants or other individuals independent from those personnel directly involved in plant operation shall be used to augment the audit teams when necessary. Individuals ' performing the audits may be members of the audited organization; however, they shall not audit l activities for which they have immediate responsibility, and while performing the audit, they shall not report to a management rep'resentative who has immediate responsibility for the activity audited.

6.5.3.4 Results of plant audits are approved by the Principal OA Specialist - Performance Evaluation Unit, and transmitted to the Executive Vice President - Power Supply and Engineering &

Construction; the Senior Vice President - Power Supply; Vice President - Nuclear Operations; General. Manager; and the Vice President - Corporate N'e' u ear Safety & Research; and others, as l appropriate within 30 days after the completion of the audit.

i l 6.5.3.5 The Corporate Ouality Assurance Audit Program shall be conducted t

! in accordance with written, approved procedures.

1 l

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j 6.5-15

c. Monthly Operating Report Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis. The report formats set forth in Appendices B, C, and D to Regulatory Guide 1.16 shall be completed in accordance with the instructions provided. The completed forms should be submitted by the tenth of the month following the calendar month covered by the report to the Director, Office of Management and Program Analysis, U. S.

Nuclear Regulatory Commission, Washington, D. C. 20555, with a copy to the appropriate NRC regional Office.

d. Semiannual Radioactive Effluent Release Report Routine radioactive effluent release reports covering the operation of the unit during the previous 6 months shall be submitted within 60 days af ter January 1 and July 1 of each year. Those portions of the report due within 60 days of January l', and J'u ly 1, shall include:

1. A summary of the quantities of radioactive liquid and I gaseous effluent and solid waste released from the Unit as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants" (Revision 1, June, 1974) with l data summarized on a quarterly basis following the format of Appendix B thereof.

! 2. The Radioactive Effluent Release Report to be submitted f within 60 days af ter January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind

! speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint l

6.9-3

l frequency distributions of wind speed, wind direction, and atmospheric stability.* This same report shall include an assessment of the radiation doses due to the radioactive i liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary (Figure 1.1-1) during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time and location shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. [For ors: approximate and conservative approximate methods are acceptable.] The assessment of radiation doses shall be performed in accordance with the 4 methodology and parameters in the Offsite Dose Calculation Manual (ODCM).

i t

The Radioactive Effluent Release Report to be submitted 60  !

3.

days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed l

member of the public from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operation.

l l

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l

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

6.9-4 l

l 4

4. The Radioactive Effluent Release Reports shall include the following information for each class of solid waste (as defined by 10 CFR Part 61) shipped offsite during the report period:

i a. Container volume,

b. Total curie quantity (specify whether determined by measurement or estimate),

I

c. Principal radionuclides (specify whether determined by measurement or estimate),

d. Source of waste and processing employed (e.g. ,

j dewatered spent resin, compacted dry waste, evaporator bottoms),

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

f. Solidification agent or absorbent (e.g. , cement, urea i formaldehyde).

5. The Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to j l unrestricted areas of radioactive materials in gaseous and l c

l liquid effluents made during the reporting period. l t

i.

6. The Radioactive Effluent Release Reports shall include any  !

changes made during the reporting period to the Process

Control Program (PCP) and to the Offsite Dose Calculation i

Manual (ODCM), as well as a listing of new locations for l

dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.17.2.2.

i 6.9-5

7. Changes to the radioactive waste systems (liquid, gaseous, and solid) shall be reported to the Commission in the i

Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the Plant Nuclear Safety Committee (PNSC).* The discussion of each change shall contain:

a. A summary of the evaluation that led to the determination that the change could be made in  !

I

. accordance with 10 CFR Part 50.59; {

l

b. Sufficient detailed information to totally support the j reason for the change without benefit of additional or supplemental information;

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

i

d. An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous -

effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto; t  !

l l e. An evaluation of the change, which shows the expected maximum exposures to an individual in the unrestricted area and to the general population that differ from l'

l those previously estimated in the license application l and amendments thereto; 1

!

• The licensee may chose to submit the information called for in this l

Specification as part of the annual FSAR update.

i 6.9-6 i I

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

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

h. Ibcumentation of the fact that the change was reviewed and found acceptable by the PNSC.

8. Changes to the radioactive waste systems (liquid, gaseous, and solid) shall become effective upon review and acceptance by the PNSC.

e. Annual Radiological Environmental Operating Report Routine radiological environmental operating reports e overing the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. With the radiological environmental monitoring program not being conducted as specified in Table 3.17-1, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence i shall be included.

The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and analysis of trends of the l

results of the radiological environmental surveillance l activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate),

i and previous environmental surveillance reports and an assessment of the observed impacts of the plant operations on l

the environment. The reports shall also include the results of i 8

l

! land use censuses required by Specification 3.17.2. <

i l 6.9-7 l

I - .-- . . _= , . - , , _ . . - . - - . _ . _ . - =-, ~ , --. - , - - . - - . - - -

The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the Table and Figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment , Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the l report, the report shall be submitted noting and explaining the  !

reasons for the missing results. The missing data chall be submitted as soon as possible in a supplementary report.

The reports shall also include the following: a summary description of the radiological environmental monitoring program:

at least two legible maps

• covering all sampling locations keyed to a table giving distances and directions from the centerline of the reactor, the results of licensee participation in the Interlaboratory Comparison Program, required by Specification 3.17.3; discussion of all deviations from the sampling schedule of Table 3.17-1; and discussion of all analyses in which the LLD required by Table 3.17-3 was not achievable.

i

• 0ne map shall cover stations near the site boundary; a second shall be the more distant stations.

6.9-8

6.9.2 Reportable Occurrences The Reportable Occurrences of Specifications 6.9.2.a and 6.9.2.b below, including corrective actions and measures to prevent recurrence, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution of the occurrence. In case of corrected or supplemental rep' orts, a Licensee Event Report (LER) shall be completed and reference made to the original report date.

a. Prompt Notification With Written Followup The types of events listed below shall be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Regional Administrator of the NRC Regional Office or his designee no later than the first working day following the event, with a written followup report within 14 days. The written followup report shall include as a minimum, a completed copy of the LER form.

Information provided on the LER shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.

(1) Failure of the reactor protection system, or other systems subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reaches the setpoint specified as the limiting safety system setting in the Technical Specifications or failure to complete the required protective function.

Note: Instrument drift discovered as a result of testing need not be reported under this item (but see 6.9.2.a(5), 6.9.2.a(6), and 6.9.2.b(1) below.

6.9-9

(2) Operation of the unit or affected systems when any parameter or operation subject to a LCO is less ,

I conservative than the least conservative aspect of the LCO established in the Technical Specifications.

Note: If specified action is taken when a system is found to be operating between the most conversative and least conservative aspects of a LCO listed in the Technical Specifications, the LCO is not considered to have been violated and no report need be submitted under this section (but see 6.9.2.b(2) below).

(3) Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary or primary containment.

Note: Leakage of valve packing or gaskets within the limits for identified leakage set forth in Technical Specifications need not be reported under this section.

(4) Reactivity anomalies involving disagreement with predicted value of reactivity balance under steady state conditio6s during power operation greater than or equal to 1% ak/k; a calculated reactivity balance indicating a shutdown margin less conservative than specified in the Technical Specifications; short-term reactivity increases that correspond to a reactor startup rate greater than 5 dpm, or if suberitical, an unplanned reactivity insertion of more than 0.5% Ak/k; or any unplanned criticality.

(5) Failure or malfunction of one or more components which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to l cope with accidents analyzed in the SAR.

l I

1 6.9-10

1 (6) Bersonnel error or procedural inadequacy which prevents or could prevent,' by itself, the fulfillment of the functional requirements of systems required to cope

, with accidents analyzed in the SAR.

b Note: For 6.9.2.a(5) and 6.9.2.a(6) reduced redundancy that does not result in loss of system function need not be reported under this section (but see 6.9.2.b(2) and 6.9.2.b(3) below).

(7) Conditions arising from natural or man-made events that, as a direct result of the event, require plant 1 shutdown, operation of safety systems, or other protective measures required by Technical Specifications.

(8) Errors' discovered in the transient or accident analyses or in the methods used for such analyses as described i in the SAR or in the bases for the Technical Specifications that have or could have

' permitted reactor operation in a manner less conservative than assumed in the analyses.

(9) Berformance of structures, systems or components that require remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the FSAR or Technical Specifications bases or discovery during '

plant life of conditions not specifically considered in

, the FSAR or Technical Specifications that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.

Note: This item is intended to provide for reporting of potentially generic problems.

6.9-11

(10) Offsite releases of radioactive materials in liquid and gaseous effluents which exceed the limits of Specifications 3.9.1.1, 3.9.3.1, and for tank contents which exceed the limits of Specifications 3.16.2.1 and 3.16.4.3.

b. Thirty-day Written Reports. The reportable occurrences discussed below shall be the subject of written reports to the Regional Administrator of the NRC Regional Office within thirty days of occurrence of the event. The written report shall include, as a minimum, a completed copy of the LER form, used for entering data into the NRC's computer-based file of information concerning licensee events. Information provided on the LER form shall be supplemented, as needed, by additonal narrative material to provide complete explanation of the circumstances surrounding the event.

(1) Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those established by the Technical Specifications but which do not prevent the fulfullment of the functional requirements of affected systems (but see 6.9.2.a(1) and 6.9.2.a(2) above).

(2) Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation (but see 6.9.2.a(2) above).

Note: Routine surveillance testing, instrument l

calibration or preventive maintenance which require system configurations described in 6.9.2.b(1) and 6.9.2.b(2) above need not be reported except where test results themselves reveal a degraded mode as described above.

6.9-12

U (3) Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature system (but see 6.9.2.a(6) above).

(4) Abnormal degradation of systems other than those specified in 6.9.2.a(3) above designed to contain radioactive material resulting from the fission process.

Note: Sealed sources or calibration sources are not i included under this item. Leakage of valve packing or gaskets within the limits for identified leakage set forth in Technical Specifications need not be reported under this item.

6.9.3 Special Reports 6.9.3.1 Special reports shall be submitted to the Regional Administrator of the NRC Regional Office of within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

Area Reference Submittal Date l

a. Containment Leak 4.4 Upon completion of Rate each test l

6.9-13

I

b. Containment Sample 4.4 Upon completion of i Tendon Surveillance the inspection at 25 ,

l years of operation

c. Post-operational 4.4 Upon completion of Containment the test at 20 years Structural Test of operation

d. Fire Protection 3.14 As specified by System limiting condition for operation

e. Overpressure Pro- 3.1.2.1.e Within 30 days of tection System operation Operation

f. Auxiliary Feedwater 3.4 Within 30 days af ter Pumps becoming inoperable 6.9.3.2 Special Radiological Ef fluent Reports The special radiological effluent reports discussed below shall be the subject of written reports to the Regional Administrator of the NRC Regional Of fice within thirty days of the occurrence of the event in lieu of a Licensee Event Report (LER).

a. Exceeding any of the limits prescribed by Specification 3.9. 2.1, 3.9.4.1, and/or 3.9. 5.1. This report shall include the f ollowing information(1):

1. The cause for exceeding the limit (s)

2. The corrective action (s) to be taken to reduce the ,

I releases of radioactive materials in the affected j ef fluents (i.e. liquid, radionoble gas, and/or  !

i 6.9-14 f

radioiodines, particulates, etc.) within the Specification and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

3. If any of the limits of Specification 3.9.2.1 were exceeded, the report must include a statement that no drinking water source exists that could be affected or include the results of radiological impact on finished drinking water supplied with regards to the requirements of 40CFR141 Safe Drinking Water Act. '

b. Exceeding any of the limits prescribed by Specification l 3.16.1.1 and/or 3.16.3.1. This report shall include the  !

following information:

1. Identification of equipment or subsystem that rendered the affected radwaste treatment system not operable.

I

2. The corrective action (s) taken to restore the affected radwaste treatment system to an operable status.

3. A summary description of the action (s) taken to prevent a similar recurrence.

I

c. Exceeding the reporting level for environmental sample media as specified in Specifications 3.17.1.3. This report shall include the following information:

1. An evaluation of any environmental factor, release i

condition or other aspect which may have caused the j reporting level to be exceeded.

}

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2. A description of action (s) taken or planned to reduce  !

the levels of licensed materials in the affected j environmental media to below the reporting level.  !

{

}

6.9-15 I

O

d. Exceeding the limits prescribed by Specification 3.9.6.1.

This report shall be made in lieu of any other report and shall include the following information:

1. The corrective action (s) to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits prescribed by Specification 3.9.6.1.

2. An analysis which estimates the dose commitment to a member of the general public from uranium fuel cycle source including all effluent pathways and direct radiation, for a 12 month period that includes releases covered by this report.

3. If the release conditions resulting in violation of 40CFR190 has not already been corrected, include a request for a variance in accordance with the provisions of 40CFR190 and include the specified information of 40CFR190.11(b).

I 6.9-16

b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.

c. Records of facility radiation and contamination surveys.

d. Records of radiation exposure for all individuals entering radiation control areas.

l

e. Records of gaseous and liquid radioactive material released to the environs.

f. Records of transient or operational cycles for those facility components designed for a limited number of transients or cycles.

g. Records of training and qualification for current members of the plant staff.

h. Records of in-service inspections performed pursuant to these Technical Specifications.

i. Records of Quality Assurance activities required by the QA program.

j. Records of review performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10CFR50.59.

k. Records of meetings of the PNSC and of the independent reviews performed by the Corporate Nuclear Safety Section.

1. Records of data results required by the radiological environmental monitoring program.

6.10-2

6.14 ENVIRONMENTAL QUALIFICATION 6.14.1 By no later than June 30, 1982 all safety-telated electrical equipment in the facility shall be qualified in accordance with the provisions of: Division of Operating Reactors " Guidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588

" Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment" December 1979. Copies of these documents are attached to the Order for Modification of License No. DPR-23 dated October 24, 1980.

6.14.2 By no later than December 1,1980, complete and auditible records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NUREG-0588. Thereaf ter, such records should be updated and maintained current as equipment is replaced, further tasted, or otherwise further qualified.

6.14-1

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s 6.15 PROCESS CONTROL PROGRAM (PCP)

\ s s s 6.15.1 The PCP shall be approved by the Commission prior to >

implementation.

6.15.2 Licensee initiated changes to the PCP:

A. Shall be submitted to the Commission in the Annual Report for the period in which the change (s) was/were made. This subriittal shall contain: s i

1. Sufficiently detailed information' to totally support the rationale for the change without benefit of M

. - w additional or supplemental information;

• s p

2. A determination that the change did not reduce the 3 w

overall conformance of the solidified waste 'prociuct to existing criteria for solid wastes. j

, , ,N

3. Documentation of the fact that the change has been t reviewed and found acceptable by the PNSC.

B. Shall become effective upon review and acceptance by the PNSC.

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-6.15-1

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6.16 0FESITE DOSE CALCUATION MANU,AL i

'U ' 6.16.1 The ODCM shall be approved by the Commission prior to implementation.

t J 6.16.2 Licensee initiated changes to the ODCM:

\d l

' A '. - Shall te submitted to the Commission in the Semiannual i i

Radioactive Effluent Release Report for the period in which, ,

the change (s) was made effective. This submittal shall contain:

1. Sufficiently detailed information to totally support the rationale for the change without benefit of .

additional or supplemental information. Information i

submitted should consist of a package of those pages of I the ODCM to be changed with each page numbered and {

provided with an approval and date box, together with '

appropriate analyses or evaluations justifying the l

charge (s);  ;

l I

,' 2. A determination that the change will not reduce the .

l accuracy or reliability of dose calculations or l

^ setpoint determinations; and i I

3. Documentation of the fact that the change has been reviewed and found acceptable by the PNSC.

l B. Shall become effective upon review and acceptance by the i PNSC.

6.16-1 c,.,.j -

, . - _ - ,v-. -g y,-m _ .- -- . - . ,----w - ,,.

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7 MAJOR CHANGES TO RADI0 ACTIVE LIQUID, GASEOUS, AND SOLID WASTE TREATMENT SYSTEMS

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

1. Shall be reported to the Commission in the Seniannual Radioactive Effluent Release .Roport for the period in which the evaluation was reviewed by the PNSC. The discussion of each change shall contain:

a. A summary of the evaluation that led to the determination that the change could be made in'accordance with 10CFR50.59.

b. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental inf ormation;

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

d. An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ f rom those previously predicted in the license application and amendments thereto;

e. An evaluation of the change, which shows the expected maximum exposures to an individual in the unrestricted area and to the general population that differ f rom those previously estimated in the license application and amendments thereto; 6.17-1

m i

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

i

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

h. Documentation of the fact that the change was reviewed and found acceptable by the PNSC.

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

B I

.-= _

• Licensee may chose to submit the information called for in this Specification as.part of the annual FSAR Update.

6.17-2  !

ENCLOSURE 4 0FFSITE DOSE CALCULATION MANUAL I

1 H. B. ROBINSON STEAM ELECTRIC PLANT UNIT NO. 2 OCTOBER 1983

- _= . ,

H.B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 0FF-SITE DOSE CALCULATIONAL MANUAL (ODCM)

Revision 1 l

DOCKET NO. 50-261 l

CAROLINA POWER & LIGHT COMPANY October 6, 1983

TABLE OF CONTENTS Section Title Page 1

TABLE OF C0NTENTS...........................................

LIST OF TABLES.............................................. 11 LIST OF FIGURES............................................. iv

1.0 INTRODUCTION

................................................ 1-1 2.0 LIQUID EFFLUENTS............................................. 2-1 2.1 Monitor Alarm Setpoint Determination................. 2-1 2.2 Compliance with 10CFR20 (Liquids).................... 2-11 2.3 Compliance with 10CFR50 (Liquids).................... 2-16 3.0 GASEOUS EFFLUENTS........................................... 3-1 3.1 Monitor Alarm Setpoint Determination................. 3-1 3.2 Compl i ance wi th 10CFR20 (Gaseou s) . . . . . . . . . . . . . . . . . . . . 3-12 3.3 Compliance with 10CFR50 (Gaseous).................... 3-22 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM............... 4-1 5.0 INTERLABORATORY COMP ARISON STUDIES . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1 0bjective............................................ 5-1 S.2 Program.............................................. 5-1 6.0 TOTAL DOSE (40CFR190 CONFORMANCE)........................... 6-1 6.1 Compliance with 40CFR190............................. 6-1 6.2 Cciculations Evaluating Conformance with 40CFR190.... 6-1 6.3 Cal cul ations of Total Body Dose. . . . . . . . . . . . . . . . . . . . . . 6-2 6.4 Thyroid Dose......................................... 6-3 APPENDIX A - Meteorological Dispersion Factor Computations......................................... ,A-1 APPENDIX B - Dose Parameters for Radioiodines, P a rti cul ates , and Tri ti um. . . . . . . . . . . . . . . . . . . . . . . . . . . . B-1 APPENDIX C - Lower Limit of Detectability................... C-1 l

1

I LIST OF TABLES No. Title Page I l

l 2.3-1 Aj, Values for the Adult for the H.B. Robinson Steam Electric Plant....................... 2-22 2.3-2 Values of e IP for Liquid Dose Calculations............ 2-24 3.1-1 Gaseous Source Terms..................................... 3-10 3.1-2 Dose Factors and Constants............................... 3-11 3.2-1 Releases from H.B. Robinson Unit No. 2................... 3-18 3.2-2 Distance to Special Locations for the H.B. Robinson Plant (Mi.)................................ 3-19 3.2-3 Dose Factors for Noble Gases and Daughters............... 3-20 3.2-4 P9 Values for an Infant for the H.B. Robinson Unit No. 2............................................... 3-21 3.3-1 thru R Values for the H.B. Robinson Steam Electric Plant...... 3-31 3-19 thru 3-49 4.0-1 H.B. Robinson Radiological Environmental 4-2 Monitoring Program.......................................

A-1 X/QValuesforLong-TermGrgundLevelReleases at Special Locations (sec/m )............................ A-4 A-2 DepletedX/QValuesforLong-TermGrgundLevel Releases at Special Locations (sec/m ) ....... . .. ......... A-5 A-3 D/Q Values for Long-Term Ground Level Releases at Special Locations (m-2)............................... A-6 A-4 X/Q Values for Long-Tenn Grognd Level Releases at Standard Distances (sec/m )........................... A-7 A-5 Depleted X/Q Values for Long-Term Grognd Level Releases at Standard Distances (sec/m ).................. A-8 A-6 D/QValuesforLong-TermgroundLevelReleases at Standard Distances (m~ ).............................. A-9 A-7 X/Q Value for Short-Term Grgund Level Releases at Special Locations (sec/m )............................ A-10 A-8 Depleted X/Q Values for Short-Term Ggound Level Releases at Special Locations (sec/m )................... A-11 11

LIST OF TABLES (continued)

No. Title Page

. A-9 D/QValuesforShort-Terg)GroundLevelReleases at Special Locations (m- ............................... A-12 A-10 X/Q Values for Long-Tenn Miged Mode Releases at Speci al Locati ons ( sec/m ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-13 A-11 Depleted X/Q Values for Long-Term Miged Mode Releases at Special Locations (sec/m )................... A-14 A-12 D/Q Values for Long-Term Mixed Mode Releases at Special Locations (m-2)............................... A-15 A-13 X/QValuesforLong-TermMixgdModeReleases at Standard Distances (sec/m )........................... A-16 A-14 Depleted X/Q Values for Long-Term Mixgd Mode Releases at Standard Distances (sec/m ).................. A-17 A-15 D/QValuesforLong-TermgixedModeReleases at Standard Distances (m~ ).............................. A-18 A-16 X/QValuesforShort-TermMjxedModeReleases at Speci al Locati ons ( sec/m ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-19 1

l A-17 DepletedX/QValuesforShort-TermMjxedMode l

Releases at Special Locations (sec/m )................... A-20 A-18 D/QValuesforShort-Terg)MixedModeReleases at Special Locations (m' ............................... A-21 1

A-19 Robinson Plant Site Information to be Used for Ground Level Calculations with NRC i "X0QD0Q" Program......................................... A-22 A-20 Robinson Plant Site Information to be Used for Mixed Mode Release Calculations with NRC

! "X0QD0Q" Program......................................... A-25 B-1 Parameters for Cow and Goat Milk Pathways... ... . .. . .. .... 8-15 B-2 Parameters for the Meat Pathway.......................... B-16 B-3 Parameters for the Vegetable Pathway..................... B-17 D-1 Liquid Process Monitors.................................. D-1 1

! D-2 Gaseous Process Monitors................................. D-2 iii

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LIST OF FIGURES Title Page No.

4-1 Radiological Sample Locations Near Site.................. 4-10 4-2 Radi ol ogi cal Sampl e Di stant Locations . . . . . . . . . . . . . . . . . . . . 4-11 D-1 H.B. Robinson Liquid Radwaste Effluent System... . .. . .. .. . D-3 D-2 H.B. Robinson Gaseous Radwaste Effluent System........... D-4 l

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l iv

1.0 INTRODUCTION

The Off-Site Dose Calculation Manual (0DCM) provides the information and methodologies to be used by H. B. Robinson Steam Electric Plant Unit 2 (HBR) to assure compliance with Specification 3.9.1, 3.9.2, 3.9.3, 3.9.4, 3.9.5, and 3.9.6 of the H. B. Robinson Technical Specification. These portions are those related to liquid and gaseous radiological effluents. They are intended to show compliance with 10CFR20, 10CFR50.36a, Appendix 1 of 10CFR50, and 40CFR190.

The ODCM is based on " Radiological Effluent Technical Specifications for PWRs (NUREG 0472, Rev. 3, Draft 6), " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants" (NUREG 0133), and guidance from the United States Nuclear Regulatory Consission (NRC). Specific plant procedures for implementation of this manual are presented in H. B. Robinson Unit 2 Plant Operating Manual and the H. B. Robinson Unit 2 Standing Orders. These pro-cedures will be utilized by the operating staff of HBR to assure compliance with technical specifications.

The vDCM has been prepared as generically as possible in order to minimize the need for future revisions. However, some changes to the ODCM will be expected in the future. Any such changes will be properly reviewed and approved as j

indicated in the Administrative Control Section, Specification 6.16.2, of the HBR Technical Specifications.

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1-1 Rev. 1

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2.0 LIQUID EFFLUENTS 2.1 MONITOR ALARM SETPOINT DETERMINATION This procedure determines the monitor alarm setpoint that indicates if the concentration of radionuclides in the liquid effluent released from the site to unrestricted areas exceeds the concentrations speci-fied in 10CFR20, Appendix B, Table II, Column 2, for radionuclides other than dissolved or entrained noble gases or exceeds a concentra-tion 2 x 10~4 C1/ml for dissolved or entrained noble gases. The methodology described in Section 2.1.2 provides an alternate means to determine monitor high alarm setpoints that may be used when an analysis is performed prior to release.

2.1.1 Setpoint Based on an Unidentified Radionuclide Mix The following method applies to liquid releases via the discharge canal when determining the alarm / trip setpoint for the Waste Disposal System Effluent Monitor (RMS-18) and the Steam Generator Blowdown Monitor (RMS-19) during all operational conditions when the radwaste l discharge flow rate is maintained constant. This methodology com-plies with Specification 3.9.1.1 of the RETS by satisfying the fol-lowing equation:

cf

'T S C where:

C = The effluent concentration limit (Specification 3.9.1.1) imple-menting 10CFR20 for the site in C1/ml.

l t

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2-1 Revi i .. -_- - - - _ . _ - . - -

c = The setpoint, in pC1/ml, of the radioactivity monitor measuring the radioactivity concentration in the effluent line prior to dilution and subsequent release; the setpoint represents a value which, if exceeded, would result in concentrations exceeding the limits of 10CFR20 in the unrestricted area.

f = The waste effluent flow rate in gpm.

F = The dilution water flow rate in gpm.

2.1.1.1 Determine c (the effluent monitor setpoint) in Ci/ml for each of the dilution water flow rates.

where: c=

C=1 x 10-7 pCi/ml, the effluent concentration limit based on

- 10CFR20, Appendix B, for an unknown radionuclide mixture.

F = Dilution water flow rate (gpm).

t 1

= 160,000 gpm from one circulating water pump , Unit 2.

1

= 250,000 gpm from two circulating water pumps , Unit 2.

1

= 400,000 gpm from three circulating water pumps , Unit 2.

or 2

= 50,000 gpm from one circulating water pump , Unit 1.

2 l

= 80,000 gpm from two circulating water pumps , Unit 1. ,

f = The maximum acceptable discharge flow rate prior to dilution (gpm).

l

= 60 gpm for the Waste Disposal System Liquid Effluent Monitor3, 2-2 Rev. 1

) y

= 80 gpm for the Steam Generator Blowdown Monitor during normal operation3,

= 300 gpm for the Steam Generator Blowdown Monitor while draining a steam generator3, 2.1.1.2 Determine CR (calculated monitor count rate in corrected counts per

. minute [cepm]). Attributed to the radionuclides for each of the dilution water flow rates.

CR = (c) (E)

E = The applicable affluent monitor efficiency located in the Plant Operating Manual, Volume 15, Curve Book. Use the radioactivity concentration "c" to find CR.

2.1.1.3 Determine SP (the monitor alarm / trip setpoint including background

[ cpm] for each of the dilution water flow rates.

SP = (T,CR + Background) where: T, = Fraction of the radioactivity from the site that may be re-leased via the monitored pathway to ensure that the site bound-ary limit is not exceeded due to simultaneous releases from several pathways.

l = .50 for the Waste Disposal System Liquid Effluent Monitor (RMS-18) .

i r

= .50 for the Steam Generator Blowdown Monitor (RMS-19).

l 2.1.2 Setpoint Based on an Analysis of Liquid Prior to Discharge l

The following method applies to liquid releases via the discharge canal when determining the alarm setpoint for the Waste Disposal System Liquid Effluent Monitor (RMS-18) and the Steam Generator

v. & J _ .- - -

Blowdown Monitor (RMS-19) when an analysis of the actitivity of the principal gamma emitters has been made prior to each batch rel' eased.

2.1.2.1 Determine D (the minimum acceptable dilution factor):

D =S[jg j Cj = Radioact.ivity concentration of radionuclide "i" in the liquid effluent prior to dilution ( C1/ml) from analysis of the liquid effluent to be released.

MPCj = The liquid effluent radioactivity limit for radionuclide "i"

( Ci/ml) from 10CFR20, Appendix B.

S = 2, A safety factor used as a conservatism to assure that the radionuclide concentrations are less than the limits specified in 10CFR20, Appendix B, at the point of discharge.

2.1.2.3 Determine c (the monitor setpoint concentration [ Ci/ml] attributed to the radionuclides for the dilution water flow rate available during the release.

c =([g Cg ) (D f) (Tm) where:

Cg = The total radioactivity concentration of gamma-emitting radio-nuclides in liquid effluent prior to dilution (pC1/ml),

f = The maximum approved discharged flow rate prior to dilution (gpm). ,

3

= 60 gpm for the Waste Disposal System Liquid Effluent Monitor ,

= 80 gpm for the Steam Generator Blowdown Monitor during normal i operation3 ,

2-4 Rev. 1

)  ?

= 300 gpm for the Steam Generator Blowdown Monitor while drain-ing a steam generator.

F = Dilution water flow rate (gpm).

1

= 160,000 gpm from one circulating water pump , Unit 2.

= 250,000 gpm from two circulating water pumps l

, Unit 2.

= 400,000 gpm from three circulating water pumps , Unit 2.

l or 2

= 50,000 gpm from one circulating water pump , Unit 1.

2

= 80,000 gpm from two circulating water pumps , Unit 1.

T,

= Fraction of the radioactivity from the site that may be re-leased via the monitored pathway to ensure that the site boundary limit is not exceeded due to simultaneous releases from more than one pathway.

i = .50 for the Waste Disposal System Liquid Effluent Monitor (RMS-18).

= .50 for the Steam Generator Blowdown Monitor (RMS-19).

If it is determined that h < 1 , the release cannot be made.

Reevaluate the discharge flow rate prior to dilution and/or the ,

dilution flow rates.

If h > 1, the release may be made.

2.1.2.4 Determine SP (the monitor alarm setpoint [cepm].

SP = (c) (E,) t background.

where:

__________m_.L

E, = The applicable effluent monitor efficiency based on "c," from the efficiency curves located in the Plant Operating Manual, Volume 15, Curve Book.

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(Left blank intentianally for future revisions)

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(Left blank intentionally for future revisions)

I 2-9

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SECTION

2.1 REFERENCES

1. Carolina Power & Light Company Drawing Number G-190825. Using the System Q-H Curve for Emergency Low Water Level.

2. Carolina Power & Light Company, Darlington County S.E. Plant. 1960-182 MW Installation, Unit 1. SYSTEM HEAD CURVES Unit 1 Circulating Water System Draining Quosig.

3. H.B. Robinson Electric Plant Unit 2. Updated Final Safety Analysis Re-port.

2.2 COMPLIANCE WITH 10CFR20 (LIQUIDS)

Liquid effluents from H.B. Robinson Unit 2 (HBR) will occur both continuously and on a batch basis. The following sections discuss the methodology which will be utilized by the ER to show compliance with 10CFR20.

2.2.1 Continuous Releases Steam generator blowdown is continuously released from 2R. Each operational working day grab samples will be taken of steam generator blowdown. These samples are composited at the rate of 100 ml/sgr.

An aliquot of the SG composite is analyzed each week for I-131 and various other fission, activation, and corrosion products, as out-

. lined in Table 4.10-1 of the technical specification for HBR. Sam-ples are to be maintained until the end of the quarter and analyzed for strontium. Steam generator volumes are based on blowdown rates. In addition, a monthly analysis will be performed to deter-mine the activity levels of tritium and dissolved and entrained gases. Compliance with 10CFR20 during actual release is established through the steam generator blowdown effluent monitor alarm set-point. This setpoint is based upon a given radionuclide mix as noted l in Section 2.1. However, if a continuous release should occur in which the effluent monitor alarm setpoint is exceeded, then actual compliance with 10CFR20 may be determined utilizing the actual radio-nuclide mix and the following equation:

/

C V ic c Concj = (2.2-1) dc where:

Concj = Concentration of radionuclide "i" at the unrestricted area, Ci/ml; 2-11

C ic

= Concentration of radionuclide "1 " in the continuous re-lease, pC1/ml; V

e

= Volume of continuous effluent released, gal; V

dc

= Volume of dilution flow during release, gal.

2.2.2 Batch Releases Batch releases will occur during normal operation. When this does occur at MR, a continuous release will usually be occurring at the same time. However, during certain shutdown conditions, only batch releases may occur at 2R. Therefore, both situations are treated here to provide the methodology to show compliance with 10CFR20.

2.2.2.1 Prerelease The radioactivity content of each. batch release will be determined prior to release in accordance with Table 4.10-1 of the technical specifications for ER. M R will show compliance with 10CFR20 in the following manner:

For the case where only a batch release is to occur, the concentra-l tion of the various radionuclides in the batch release, determined in accordance with Table 4.10-1 of the technical specifications for 2R, is multiplied by the ratio of the maximum release rate of the poten-tial batch release to the dilution flow rate to obtain the concentra-l tion at the unrestricted area. This calculation is shown in the l following equation:

l C R ib b Concj = D (2.2-2) i fr r

where:

2-12 l.. -. . . - . .. .- .-.-- ,_-.__._ - - _.- ..-.-- . - -. - - - __.

Concj = Concentration of radionuclide "1 " at the unrestricted ,

area, C1/ml; C

ib

= Concentration of radionuclide "1 " in the potential batch release, pCi/ml; Rb = Release rate of the potential batch release, gpm; D fp = The dilution flow rate based upon the number of circulating water pumps in service during the release, gpm.

l The concentration in the unrestricted area is compared to the con-centrations in Appendix B, Table II, Column 2, of 10CFR20. Before release may occur, the mixture of radionuclides released must be of such concentration that Equation 2.2-3 is met.

{j (Concj /MPC j )I1 (2.2-3) where:

MPC j = Maximum permissible concentration of radionuclide "1" from l Appendix B, Table II, Column 2 of 10CFR20, pC1/ml.

For those cases where batch releases may be occurring at the same time that continuous releases are occurring, the concentration in the unrestricted area will be calculated by the following equation:

l Concj =

ib Rb+Cic g

R c (2.2-4) fr where:

R e

a Maximum continuous liquid effluent release rate, gpm.

l 2-13

The mixture of radionuclides released must be of such concentrations that Equation 2.2-3 must be met.

For ER, the liquid radwaste effluent line discharges to the circu-lating water system. Therefore, the dilution flow rate (Dfp) is a function of the number of circulating water pumps operating. Unit 2 of the H.B. Robinson Steam Electric Plant has three circulating water pumps. Pump curves show that with three pumps operating, the circu-lating water flow is 400,000 gpm, with two pumps--250,000 gpm, and with one pump--160,000 gpm. Unit 1 of the H.B. Robinson Steam Elec-tric Plant has two circulating water pumps. The circulating water flow is 50,000 gpm with one pump and 80,000 gpm with two pumps. At

, least one circulating water pump must be operating during any liquid waste discharge.

Batch releases from the ER liquid radwaste system may occur from the waste condensate tanks, the monitor tanks, and the steam gener-ators. The maximum release rate (R b

) is 300 gpm for the .iteam gener-ators and 60 gpm from the monitor and waste condensate tanks.

l 2.2.2.2 Pestrelease The Steam Generation Blowdown Monitor (RMS-19) and the Waste Disposal System Liquid Monitor (RMS-18) setpoint will each be limited to 50 percent of the 10CFR20 limits. These setpoints will ensure that 10CFR20 limits are met. However, because they are based upon a given mix, the possibility exists that the alarm trip setpoints may be exceeded, while 10CFR20 limits are not exceeded. The following methodology is provided to determine whether actual releases exceeded 10CFR20 limits.

The concentration of each radionuclide in the unrestricted area following release from a batch tank will be calculated in the follow-ing manner:

2-14

For the case where only batch releases are occurring, the total activity of radionuclide "1" released is divided by the actual dilu-tion flow to obtain the concentration in the unrestricted area. This calculation is shown in the following equation:

Concik , C ikb Y kb (2.2-5)

Vkd where:

Concik = The concentration of radionuclide "i" at the unrestricted area during release k, pC1/ml; C

ikb = Concentration of radionuclide "i" in the batch release k,pC1/ml; V

kb = Volume of batch release k, gal; V

kd = Actual volume of dilution flow during release k, gal.

To show compliance with 10CFR20, the following relationship must hold:

{ (Conc ik /MPC9 ) < 1 (2.2-6)

The actual dilution volume during release k (Vkd) is calculated by the following equation:

Y kd " 60 I k (Dfr)*k (2.2-7) where:

60 = Conversion factor, min /hr; tk = Duration of release k, hr; 2-15 _ _ _ _

D fp = Dilution flow rate from circulating water pumps during release k, gpm.

The circulating water pump flow rates were given in Section 2.2.2.1 above.

For the case where a batch release is occurring at the same time that a continuous release is occurring, the compliance with 10CFR20 limits may be determined by the following equation:

C Concik , ikb Ykb + CikeYkc (2.2-8)

V kd where:

C ike = Concentration of radionuclide "1 " in continuous releases during release period k, pCi/ml; Ykc = Volume of continuous release du.-ing period k, gal.

2.3 COMPLIANCE WITH 10CFR50 2.3.1 Cumulation of Doses The dose contribution from the release of liquid effluents will be calculated once per month, and a cumulative summation of these total body and any organ doses should be maintained for each calendar quarter. The dose contribution for all batch releases will be calcu-lated using the following equation:

O sb "Sk Ei

^I, t kb C

ikb F

kb' (2.3-1) 2-16

where:

D 4 = The cumulative dose commitment to the total body or any organ t, from batch liquid effluents, mrem; tkb = The length of time of batch release k over which Cikb and Fkb are averaged for each batch liquid release, hours; Cikb = The average concentration of radionuclide "i" in undiluted batch liquid effluent during batch release k, pCi/ml; A,9

= The site-related ingestion dose connitment factor to the total body or any organ 5 for each identified principal gamma and beta emitter, arem-m1 per hr- Ci; l \

A 9

= Radiological decay constant of radionuclide "i", hr-I-

= 0.693/(t 1/2)9 (t 1/2)j = Radiological half-life of radionuclide "i", hr; tp = average transport time to reach the point of exposure, hr;

= 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the fish pathway from Equation A-3 of Regula-tory Guide 1.109, Revision 1.

Fkb = The near-field average dilution factor for C ikb during any batch liquid effluent release k. Defined as the ratio of the volume of undiluted liquid waste released to the product of the dilution volume from the site discharge structure to unrestricted receiving waters times 1.0.

(1.0 is the site-specific applicable factor for the mixing effect of the HBR discharge structure as defined in NUREG-0133, October 1978).

t 2-17 l_. ~ e _o_._

.N .

h.

Y

, kb -

Ykd x 1.0 -

Where Ykb and Ykd are as defined in Equation 2.2-5.

The dose factor Aj was calculated for an adult for each isotope using thd following equation:

s A jy w 1.14 x 105 (21BFj ) DFj , (2.3-2) where: I "

. n 1.14 x 105 = 100hx10 hx 87 hr 21 = Adult fish consumption rate from Table E-5 of Regulatory Guide 1.109, Revision 1, kg/yr; BFj = Bioaccumulation factor f6r radionuclide "i" in fish from Table A-1 of Regulatory Guide 1.109, Revision 1, pCi/kg per pCi/1; \.

\

DF j = Dose convers' ion factor for radionuclide "i" for adults for a_ pirticular* organ x from Table E-11 of Regulatory Guide 1.109, Revision 1, mrem /pC1.

l The potable water pathway does not exist either within Lake Robinson l or downstream of the Lake Robinson dam. Therefore, the potable water tenn was excluded from the calculation of A js values. Table 2.3-1 presents Aj s values for an adult at HBR. Values of exp (-A t ) are jp l presented in Table 2.3-2 as a function of radionuclide.

As noted in Section 2.2.2, steam generator blowdown is continuously released from HBR. The dose from continuous releases will be calcu-lated using the following equation:

2-18

l

)  !

l

- l

\

' -A t D

sc ={k 1A,tke 1 j ike ke ')

C F

• E where:

D = The cumulative dose commitment to the total body or any c

organ 5,. from liquid effluents for continuous releases, mrem; - -

tkc = The length of time of continuous release period k over which C ike and F ke are averaged for all continuous liquid re-leases, hours; Cike = The average concentration of radionuclide "1" in undiluted liquid effluent during continuous release period k from any continuous liquid release, pCi/ml; F

ke = The near-field average dilution factor for C ike during continuous liquid effluent release k. Defined as the ratio of the volume of undiluted liquid waste released to the product of the dilutiori volume from the site discharge structure to unrestricted receiving water times 1.0. (1.0 is the site-specific applicable factor for the mixing effect s

o'f the }BR discharge structure as defined in NUREG-0133, October 1978). -

Y F

ke

= kc Ykd x 1.0 s Where Vke and Vkd are, as defined in Equation 2.2-5, only now distin-

) guished for continuous releases.

g<

'i The sum of the cumulative dose frcm all batch and continuous releases for a quarter are compared to one half the design objectives for

' total body and any organ. The sum of the cumulative doses from all batch and continuous releases for a calendar year are compared to the l design objective doses. The following relationships should hold for  !

~

a in

HBR to show compliance with Technical Specification 3.9.2.1 of the technical specifications for H.B. Robinson Unit 2.

For the calendar quarter, D 1 1.5 mrem total body (2.3-4)

D,15 mrem any organ (2.3-5)

For the calendar year, D 1 3 mrem total body (2.3-6)

D, 1 10 mrem any organ (2.3-7) where:

0 = Cumulative total dose to any organ 5 or the total body from continuous and batch releases, mrem;

= 0 + D b sc l The quarterly limits given above represent one half the annual design objective of Section II.A of Appendix I of 10CFR50. If any of the limits in Expressions 2.3-4 through 2.3-7 are exceeded, a special report pursuant to Technical Specification 6.9.3.2 must be filed with l

the NRC. This report complies with Section IV.A, of Appendix I of 10CRF50.

2.3.2 Projection of Doses l Doses resulting from the release of liquid effluents will be projec-i ted once per month. The doses will be projected using Equa-tions 2.3-1 and 2.3-3 with Fkb and Fke now based on minimum dilution flow rate (Dfp), as given in Equation 2.2-2, rather than dilution volume (Vkd) and based upon the maximum release rates (Rb and R c ), as i

2-20

given in Equation 2.2-4, rather than actual release volume (Vb and

Vg ). C ikb and Cike are based on the projected releases for tha remainder of the calendar quarter.

I 4

l l

l l

l l

2-21

l l

TABLE 2.3-1 Aj, VALUES FOR THE ADULT FOR THE H.B. RW INSON STEAM ELECTRIC PLANT (MEM/R PER MICRO-Ci/ML) 1 Nuclide Bone Liver T. Body Thyroid Kidney pang, GI-LLI H-3 0.00E-01 2.26E 01 2.26E-01 2.26E-01 2.26E-01 2.26E-01 2.26E-01 C-14 3.13E+04 6.26E 03 6.26E+03 6.26E+03 6.26E+03 6.26E+03 6.26E+03 Na-24 4.07E+02 4.07E 02 4.07E+02 4.07E+02 4.07E+02 4.07E+02 4.07E+02 P-32 4.62E+07 2.87E 06 1.79E 06 0.00E-01 0.00E-01 '0.00E-01 5.19E 06 Cr-51 0.00E-01 0.00E+01 1.27E 00 7.61E-01 2.81E-01 1.69E 00 3.20E 02 Mn-54 0.00E-01 4.38E 03 8.35E 02 0.00E-01 1.30E 03 0.00E-01 1.34E 04 Mn-56 0.00E+01 1.10E 02 1.95E 01 0.00E-01 1.40E 02 0.00E-01 3.51E 03 Fe-55 6.58E 02 4.55E 02 1.06E 02 0.00E-01 0.00E-01 2.54E 02 2.61E 02 Fe-59 1.04E 03 2.44E 03 9.36E 02 0.00E-01 0.00E-01 6.82E 02 8.14E 03 Co-58 0.00E-01 8.92E 01 2.00E 02 0.00E-01 0.00E-01 0.00E-Oi 1.81E 03 Co-60 0.00E-01 2.56E 02 5.65E 02 0.00E-01 0.00E-01 0.00E-01 4.81E 03 Ni-63 3.11E 04 2.16E 03 1.04E 03 0.00E-01 0.00E-01 0.00E-01 4.50E 02 Ni-65 1.26E 02 1.64E 01 7.49E 00 0.00E-01 0.00E-01 0.00E-01 4.17E 02 Cu-64 0.00E-01 9.97E 00 4.68E 00 0.00E-01 2.51E 01 0.00E-01 8.50E 02 Zn-65 2.32E 04 7.37E 04 3.33E 04 0.00E-01 4.93E 04 0.00E-01 4.64E 04 Zn-69 4.93E 01 9.43E 01 6.56E 00 0.00E-01 6.13E 01 0.00E-01 1.42E 01 B r-83 0.00E-01 0.00E-01 4.04E 01 0.00E-01 0.00E-01 0.00E-01 5.82E 01 B r-84 0.00E-01 0.00E-01 5.24E 01 0.00E-01 0.00E-01 0.00E-01 4.11E-04 Br-85 0.00E-01 0.00E-01 2.15E 00 0.00E-01 0.00E-01 0.00E-01 1.01E-15 Rb-86 0.00E-01 1.01E 05 4.71E 04 0.00E-01 0.00E-01 0.00E-01 1.39E 04 -

Rb-88 0.00E-01 2.90E 02 1.54E 02 0.00E-01 0.00E-01 0.00E-01 4.00E-09 Rb-89 0.00E-01 1.92E 02 1.35E 02 0.00E-01 0.00E-01 0.00E-01 1.12E-11 Sr-89 2.21E 04 0.00E-01 6.35E 02 0.00E-01 0.00E-01 0.00E-01 3.55E 03 Sr-90 5.44E 05 0.00E-01 1.34E 05 0.00E-01 0.00E-01 0.00E-01 1.57E 04 Sr-91 4.07E 02 0.00E-01 1.64E 01 0.00E-01 0.00E-01 0.00E-01 1.94E 03 Sr-92 1.54E 02 0.00E-01 6.68E 00 0.00E-01 0.00E-01 0.00E-01 3.06E 03 Y-90 5.76E-01 0.00E-01 1.54E-02 0.00E-01 0.00E-01 0.00E-01 6.10E 03 Y-91M 5.44E-03 0.00E-01 2.11E-04 0.00E-01 0,00E-01 0.00E-01 1.60E-02 Y-91 8.44E 00 0.00E-01 2.26E-01 0.00E-01 'O.00E-01 0.00E-01 4.64E 03 Y-92 5.06E-02 0.00E-01 1.48E-03 0.00E-01 0.00E-01 0.00E-01 8.86E 02 Y-93 1.60E-01 0.00E-01 4.43E-03 0.00E-01 0.00E-01 0.00E-01 5.09E 03 Zr-95 2.40E-01 7.70E-02 5.21E-02 0.00E-01 1.21E-01 0.00E-01 2.44E 02 ,

Zr-97 1.33E-02 2.68E-03 1.22E-03 0.00E-01 4.04E-03 0.00E-01 8.30E 02 i Nb-95 4.47E 02 2.48E 02 1.34E 02 0.00E-01 2.46E 02 0.00E-01 1.51E 06 Mo-99 0.00E-01 1.03E 02 1.96E 01 0.00E-01 2.34E 02 0.00E-01 2.39E 02 j

- - . _ _ . _ _ w n e i

l TABLE 2.3-1 (continued)

Nuclide Bone Liver TJBody Thyroid Kidney ljygg GI-LLI Tc-99M 8.87E-03 2.51E-02 3.19E-01 0.00E-01 3.81E-01 1.23E-02 1.48E+01

. Tc-101 9.12E-03 1.31E-02 1.29E-01 0.00E-01 2.37E-01 6.72E-03 3.95E-14 Ru-103 4.43E+00 0.00E-01 1.91E+00 0.00E-01 1.69E+01 0.00E-01 5.17Et02 Ru-105 3.69E-01 0.00E-01 1.46E-01 0.00E-01 4.76E+00 0.00E-01 2.26E+02 I

Ru-106 6.58E+01 0.00E-01 3.33E+00 0.00E-01 1.27E+02 0.00E-01 4.26E+03 Ag-110M 8.81E-01 8.15E-01 4.84E-01 0.00E-01 1.60E 00 0.00E-01 3.33E 02 l

, Te-125M 2.57E 03 9.30E 02 3.44E 02 7.72E 02 1.04E 04 0.00E-01 1.02E 04 Te-127M 6.48E 03 2.32E 03 7.90E 02 1.66E 03 2.63E 04 0.00E-01 2.17E 04 Te-127 1.05E 02 3.78E+01 2.28E 01 7.80E 01 4.29E 02 0.00E-01 8.31E 03 ,

Te-129M 1.10E 04 4.11E 03 1.74E 03 3.78E 03 4.60E 04 0.00E-01 5.54E'04 Te-129 3.01E 01 1.13E 01 7.33E 00 2.31E 01 1.26E 02 0.00E-01 2.27E 01 Te-131M 1.66E 03 8.10E 02 6.75E 02 1.28E 03 8.21E 03 0.00E-01 8.04E 04 Te-131 1.89E 01 7.88E 00 5.96E 00 1.55E 01 8.26E 01 0.00E-01 2.67E 00 Te-132 2.41E 03 1.56E 03 1.47E 03 1.72E 03 1.50E 04 0.00E-01 7.38E 04 I-130 2.71E 01 8.01E 01 3.16E 01 6.79E 03 1.25E 02 0.00E-01 6.89E 01 1-131 1.49E 02 2.14E 02 1.22E 02 7.00E 04 3.66E 02 0.00E-01 5.64E 01 1-132 7.29E 00 1.95E 01 6.82E 00 6.82E 02 3.11E 01 0.00E-01 3.66E 00 I-133 5.10E 01 8.87E 01 2.70E 01 1.30E 04 1.55E 02 0.00E-01 7.97E 01 1-134 3.81E 00 1.03E 01 3.70E 00 1.79E 02 1.64E 01 0.00E-01 9.01E-03 l I-135 1.59E 01 4.17E 01 1.54E 01 2.75E 03 6.68E 01 0.00E-01 4.70E 01 Cs-134 2.98E 05 7.09E 05 5.79E 05 0.00E-01 2.29E 05 7.61E 04 1.24E 04 Cs-136 3.12E 04 1.23E 05 8.86E 04 0.00E-01 6.85E 04 9.38E 03 1.40E 04 Cs-137 3.82E 05 5.22E 05 3.42E 05 0.00E-01 1.77E 05 5.89E 04 1.01E 04 Cs-138 2.64E 02 5.22E 02 2.59E 02 0.00E-01 3.84E 02 3.79E+01 2.23E-03 Ba-139 9.29E-01 6.62E-04 2.72E-02 0.00E-01 6.19E-04 3.75E-04 1.65E 00 Ba-140 1.94E 02 2.44E-01 1.27E 01 0.00E-01 8.30E-02 1.40E-01 4.00E 02 Ba-141 4.51E-01 3.41E,04 1.52E-02 0.00E-01 3.17E-04 1.93E-04 2.13E-10 Ba-142 2.04E-01 2.10E-04 1.28E-02 0.00E-01 1.77E-04 1.19E-04 2.87E-19 La-140 1.50E-01 7.54E-02 1.99E-02 0.00E-01 0.00E-01 0.00E-01 5.54E 03 La-142 7.66E-03 3.48E-02 8.68E-04 0.00E-01 0.00E-01 0.00E-01 2.54E 01 Ce-141 2.24E-02 1.52E-02 1.72E-03 0.00E-01 7.04E-03 0.00E-01 5.79E 01

~

Ce-143 3.95E-03 2.92E 00 3.23E-04 0.00E-01 1.29E-03 0.00E-01 1.09E 02 Ce-144 1.17E 00 4.88E-01 6.27E-02 0.00E-01 2.90E-01 0.00E-01 3.95E 02 Pr-143 5.51E-01 2.21E-01 2.73E-02 0.00E-01 1.27E-01 0.00E-01 2.41E 03 Pr-144 1.80E-03 7.48E-04 9.16E-05 0.00E-01 4.22E-04 0.00E-01 2.59E-10 Nd-147 3.76E-01 4.35E-01 2.60E-02 0.00E-01 2.54E-01 0.00E-01 2.09E 03 W-187 2.96E 02 2.47E 02 8.65E 01 0.00E-01 0.00E-01 0.00E-01 8.10E 04 Np-239 2.85E-02 2.80E-03 1.54E-03 0.00E-01 8.74E-03 0.00E-01 5.75E 02

-- . . . - . - - - . - - ._.. - -. . - . - - - - - - . - _ _ .- . r L .

TABLE 2.3-2 Values of e IP For Liquid Dose Calculations h

• At*

9p Radionuclide A1 (hr'I) e iP Radionuclide Ai (hr-1) e H-3 6.43E-6 1.00 Zr-95 4.44E-4 9.89E-1 C-14 1.38E-8 1.00 Zr-97 4.08E-2 3.76E-1 F-18 3.75E-1 1.23E-4 Nb-95 8.25E-4 9.80E-1 Na-24 4.62E-2 3.30E-1 Mo-99 1.03E-2 7.81E-1 P-32 2.02E-3 9.53E-1 Tc-99m 1.16E-1 6.18E-2 Cr-51 1.04E-3 9.75E-1 Tc-101 2.97 0.00 Mn-54 9.53E-5 9.98E-1 Ru-103 7.29E-4 9.83E-1 Mn-56 2.69E-1 1.57E-3 Ru-105 1.56E-1 2.37E-2

Fe-55 3.04E-5 9.99E-1 Ru-106 7.87E-5 9.98E-1 l Fe-59 6.42E-4 9.85E-1 Ag-110m 1.14E-4 9.97E-1 Co-58 4.05E-4 9.90E-1 56-124 4.81E-4 9.89E-1 Co-60 1.50E-5 1.00 Te-125m 4.98E-4 9.88E-1 Ni-63 8.60E-7 1.00 Te-127m 2.65E-4 9.94E-1 Ni-65 2.71E-1 1.50E-3 Te-127 7.37E-2 1.71E-1 l Cu-64 5.42E-2 2.72E-1. Te-129m 8.49E-4 9.8CE-1 Zn-65 1.18E-4 9.97E-1 Te-129 6.03E-1 5.19E-7 Zn-69 7.29E-1 2.52E-8 Te-131m 2.31E-2 5.74E-1 Br-83 2.89E-1 9.72E-4 Te-131 1.66 4.99E-18 Br-84 1.31 2.22E-14 Te-132 8.89E-3 8.08E-1 Br-85 1.39E+1 0.00 I-130 5.59E-2 2.61E-1 I Rb-86 1.55E-3 9.63E-1 1-131 3.59E-3 9.17E-1 Rb-88 2.33 5.18E-25 I-132 3.01E-1 7.29E-4 Rb-89 2.70 0.00 I-133 3.30E-2 4.53E-1 Sr-89 5.55E-4 9.87E-1 I-134 8.00E-1 4.59E-9 Sr-90 2.80E-6 1.00 I-135 1.03E-1 8.44E-2 Sr-91 7.17E-2 1.79E-1 Cs-134 3.86E-5 9.99E-1 Sr-92 2.56E-1 2.15E-3 Cs-136 2.22E-3 9.48E-1 Y-90 1.08E-2 7.72E-1 Cs-137 2.63E-6 1.00 Y-91m 8.32E-1 2.13E-9 Cs-138 1.29 3.58E-14 Y-91 4.91E-4 9.88E-1 Ba-139 5.02E-1 5.86E-6 Y-92 1.96E-1 9.06E-3 Ba-140 2.26E-3 9.4 7E-1 Y-93 6.80E-2 1.96E-1 Ba-141 2.31 8.37E-25 Ba-142 3.78 0.00 Pr-143 2.12E-3 9.50E-1 l

La-140 1.72E-2 6.62E-1 Pr-144 2.41 0.00 La-142 4.52E-1 1.94E-5 Nd-147 2.60E-3 9.40E-1 Ce-141 8.75E-4 9.79E-1 W-187 2.90E-2 4.99E-1 Ce-143 2.09E-2 6.06E-1 Np-239 1.23E-2 7.44E-1 Ce-144 1.02E-4 9.98E-1

• Note: All values less than 1E-25 are reported as 0.

I l .

l l

l 3.0 GASEOUS EFFLUENTS 3.1 MONITOR ALARM SETPOINT DETERMINATION This procedure determines the monitor alarm setpoint that indicates if the dose rate in the unestricted areas due to noble gas radionuclides in the gaseous effluent released from the site to areas at and beyond the site boundary exceeds 500 mrem / year to the whole body or exceeds 3000 mrem / year to the skin.

The methodology described in Section 3.1.2 provides an alternative means to determine monitor alarm setpoints that may be used when an analysis of batch releases is performed prior to release.

3.1.1 Setpoint Based on Conservative Radionuclide Mix (Ground and Mixed Mode o Releases Releases through the steam generator flash tank vent can only occur through this vent when significant primary-to-secondary leakage exists within the steam generators and the plant is operating below 30 percent power. Detection of primary-to-secondary leakage is accomplished most effectively by continu-ously monitoring the condenser vacuum pump vent (RMS-15). Steam generator blowdown is continuously monitored by RMS-19 as a liquid pathway.

The following method applies to gaseous releases via the plant vent and con-denser vacuum pump vent when determining the high alarm setpoint for the plant vent gas monitor (RMS-14) and condenser vacuum pump vent gas monitor (RMS-15) during the following operational conditions:

. Continuous release via the plant vent.

. Continuous release via the condenser vacuum pump vent.

. Batch release of containment purge via the plant vent.

I t _ _ - _ _ - _ - . _ _ _ . _ _ _ _ . -

3-1_ . _ .

. Batch release for containment pressure relief via the plant vent.

. Batch release of waste gas decay tanks via the plant vent.

3.1.1.1 Determine the " mix" (noble gas radionuclides and composition) of the _.

gaseous effluent.

a. Determine the gaseous source terms that are representative of the

" mix" of the gaseous effluent. Gaseous source terms are the no51e gas activities in the effluent.

Gaseous source terms can be obtained from Table 3.1-1 or from analysis of the gaseous effluent.

b. Determine Sj (the fraction of the total noble gas radioactivity in the gaseous effluent comprised by noble gas radionuclide "1")

for each individual noble gas radionuclide in the gaseous efflu-ent.

A I

S j =

(3.1-1)

EA i j Aj = The radioactivity of noble gas radionuclide "1" in the gaseous effluent from Table 3.1-1 or from analysis of gaseous effluent to be released.

3.1.1.2 Determine the Q, (the maximum acceptable total release rate of all noble gas radionuclides in the gaseous effluent [pCi/sec]) based upon the whole body exposure limit of 500 mrem / year by:

500 Q, =

(3.1-2)

(TM) I, x is, .

3-2 l

i

~

(IM) = The highest calculated annual average relative dispersion factor for any area at or beyond the unrestricted area boundary for all sectors (sec/m3 ),

=

8.1 E-5 sec/m3 (Continuous Ground Release) from Table A-1, Appendix A.

=

9.9 E-7 sec/m3 (Continuous Mixed Mode Release) from Table A-10, Appendix A only with upper wind speed > 9 nph.

= 5.1 E-5 sec/m 3 (Batch Ground Release) from Ta-ble A-7, Appendix A.

=

2.9 E-6 sec/m3 (Batch Mixed Mode Release) from Ta-ble A-16, Appendix A only with upper wind speed

?,_ 9 mph.

Kj = The total whole body dose factor due to gamma emissions from noble gas radionuclide "1 "

(mrem /yr/ C1/m 3 ) from Table 3.1-2.

3.1.1.3 Determine Q, (the maximum acceptable release rate of all gas radio-l nuclides in the gaseous effluent [pci/sec]) based upon the skin expo-sure limit of 3000 mrem /yr by:

000 Q, =

(3.1-3)

(TM)[9[(Lj + 1.1 M9 ) 59 ]

L5 + 1.1M9

= The total skin dose factor due to emissions from 3

noble gas radionuclide "1" (mrem /yr/pC1/m ) from Table 3.1-2.

3-3

_ _ _.__._ _______ ___ _ _ a _JL__

3.1.1.4 Determine C, (the maximum acceptable total radioactivity concentra-tion of all noble gas radionuclides in the gaseous effluent

[pC1/cc]). ,

2.12 E-3 Q" l C, =

7 (3.1-4)

NOTE: Use the lower of the Q, values obtained in Sections 3.1.1.2 and 3.1.1.3. This will protect both the skin and total body from being exposed to the limit.

t where:

F = The maximum acceptable effluent flow rate at the point of release (cfm).

= 60,000 cfm for plant vent.

= 45 cfm for the condenser vacuum pump vent.

2.12 E-3 = Unit conversion constant to convert pCi/sec/cfm to pC1/cc.

3.1.1.5 Determine CR (the calculated monitor count rate above background attributed to the noble gas radionuclides [ccpm]) by:

CR =

{$ (C,) (E,)

E, = Obtained from the applicable effluer.t monitor efficiency curve located in the Plant Operating Manual, Volume 15, Curve Book. Use the radioac-tivity concentration "C m" to find CR.

3.1.1.6 Determine the HSP (the monitor high alarm setpoint inc!uding back-ground [ cpm]) by:

=

HSP T,CR + background (cpm) (3.1-5) 3-4

where:

TM = Fraction of the radioactivity from the site that may be released via the monitored pathway to en-sure that the site boundary limit is not exceeded due to simultaneous releases from several path-ways.

= 0.90 percent for Plant Vent Gas Monitor (RMS-14).

= 0.01 percent for the Condenser Vacuum Pump Vent Monitor (RMS-15).

= 0.09 for the Fuel Handling Basement Exhaust Moni-tor (RMS-20).

3.1.2 Gaseous Effluents Analyzed Prior to Release The follwing method applies to gaseous releases via the plant vent when determinia the maximum acceptable effluent flow rate at the point of release and the assodated high-alarm setpoint based on this flow rate for the plant vent gas monits (RMS-14) during the following operational conditions:

l . Batch release of containment purge.

Batch release of containment pressure relief.

. Batch release of waste gas decay tanks.

3.1.2.1 Determine Rj (the noble gas release rate [ C1/sec] for radionuclide "1"):

R

=

472 (C9 ) (F) i 3-5

where:

472 =

A conversion factor to convert cfm to cc/sec.

Cj = The radioactivity concentration of noble gas radionuclide "i" in the gaseous effluent (pC1/cc) from the analysis of the gaseous effluent to be released.

F = The maximum acceptable effluent flow rate at the point of release (cfm).

= 45 for the condenser.

= S0,000 for the containment purge.

=

2 E6 ( ^

I4 *7

)(T C

)

for pressure relief.

t 525 (1 ,7 ) (2730) c -

=

g fgp a gas decay tank release where:

=

2 E6 and 525 are the volumes (ft ) of the contain-ment and decay tank rdspectively, and T, c T te a Pc , and a P g are the respective temperature and change in pressure (psig) following the release of the containment and decay tank. .

t = Length of release (min).

3-6 Rev. 1

) 't

3.1.2.2 Determine the monitor alarm setpoint based on total body dose rate:

a. Determine CRt (the monitor count rate per mrem /yr, total body).

CR t "

(T/(T){j KR j$

where:

C = The count rate of the monitor corresponding to the radioactivity concentration in the analyzed sample (C = Cj [x the monitor efficiency]) in cpm.

T/IT = The highest calculated annual average relative dispersion factor for any area at or beyond the unrestricted area boundary for all sectors (sec/m3 ) from Appendix A.

3 from

= 5.1 E-5 sec/m (Batch Ground Release)

Table A-7, Appendix A.

= 2.9 E-6 sec/m 3 (Batch Mixed Mode Release) from Table A-16, Appendix A only with upper wind speeds of > 9 mph.

Kj = The total whole body dose factor due to gamma emissions from noble gas radionuclide "1" 3

(mrem /yr/pCi/m ) from Table 3.1-2.

b. Determine St (the count rate of the gaseous effluent noble gas monitor at the alarm setpoint based on total body dose rate

[cepm]):

St = SF x T, x Dt x CR t 3-7 , ,

where:

SF = An engineering factor used to provide a margin of safety for cumulative uncertainties of measure-ments;

= .5; Dt = 500 mrem /yr (the total body dose rate limit);

T, = Fraction of the radioactivity from the site that may be released via the monitored pathway to en-sure that the site boundary limit is not exceeded due to simultaneous releases from several path-ways;

= 0.9 for the Plant Vent Gas Monitor (RMS-14).

3.1.2.3 Determine the monitor alarm setpoint based on the skin dose rate:

a. Determine CRs (the monitor count rate per mrem /yr, skin):

CR s

• l T/7{9(L9 + 1.1 Mg ) (R$ )

where:

L1 + 1.1 Mj= The total skin dose factor due to emissions from 3 ) from noble gas radionuclide "1" (mrem /yr/pC1/m Table 3.1-2.

l b. Determine S 3 (the count rate of the gaseous effluent noble gas monitor at the alarm setpoint based on the dose rate to the skin

[cepm]):

SS = SF x T,x Ds x CR s 3-8

where:

D s

= 3000 mrem /yr (the dose rate to the skin limit).

3.1.2.4 Determine the actual gaseous monitor setpoint:

The setpoints that were determined based on the dose rate limits to the total bo@ (S t) and to the skin (S ) sare compared and the lesser value is used as the actual setpoint.

l l

l 3-9

- - - - _ _m _ _ _. __

TABLE 3.1-1 GASEOUS SOURCE TERMS

• Condenser Vacuum Containment Purge I 2 Plant Vent Release _

Puse Vent or Presure Relief Gas Decay Tanks Radlonuclide A, (CI/yr) s, A, (CI/yr) s, A, (CI/yr) s, A, (Cl/yr) s, Kr-85m 2.0E0 5.26E-2 1.0E0 4.35c-2 0.00 0.00 0.00 0.00 Kr-85 0.00 0.00 0.00 0.00 0.00 0.00 1.6E2 8.00E-1 Kr47 1.0E0 2.63E-2 0.00 0.00 0.00 0.00 0.00 0.00 Kr-88 3.0E0 7.89E-2 2.0E0 8.70E-2 1.0E0 2.90E-3 0.00 0.00 Xe-131m 0.00 0.00 0.00 0.00 1.0E0 2.90E-3 9.0E0 4.50E-2 ,

Xe-133m 0.00 0.00 0.00 0.00 4.0E0 1.16E-2 0.00 0.00 Xe-133 2.8El 7.37E-1 1.8E-1 7.83E-1 3.lE2 8.99E-l 3.lEl I.55E-l 135 4.0E0 1.05E-1 2.0E-1 8.70E-2 4.0E0 1.16E-2 0.00 0.00 Ar-41 0.00 0.00 0.00 0.00 2.5El 7.25E-2 0.00 0.00 TOTAL 3.8E1 2.3El 3.45E2 2.0E2

• Source terms are based upon GALE Code and not actual releases from the evalu-aticn of H.B. Robinson Unit 2 to demonstrate conformance to the design objec-tives of 10CFR50, Appendix I, Table 2-4. These values are only for routine releases and not for a complete inventory of gases in an emergency.

1These values are used to determine the monitor alarm setpoints for the Plant Vent Gas Monitor (RMS-14) and Fuel Handling Basement Exhaust Monitor (RMS-20).

2Th'ese values are used to determine the monitor alarm setpoint for the Conden-ser Vacuum Pump Vent Monitor (RMS-15).

l 3-10

- u

TABLE 3.1-2 00SE FACTORS AND CONSTANTS

• Total Whole Body Total Skin Dose Factor Dose Factor (Kj ) (Lj + 1.1 Mj)3 Radior.uclide (arem/yr/nCi/m3 ) meem/yr/uci/m )

Kr-83m 7.56E-2 2.12E1 Kr-85m 1.17E3 2.81E3 Kr-85 1.61El 1.36E3 Kr-87 5.92E3 1.65E4 Kr-88 1.47E4 1.91E4 Kr-89 1.66E4 2.91E4 Kr-90 1.56E4 2.52E4 Xe-131m 9.15El 6.48E2 Xe-133m 2.51E2 1.35E3 Xe-133 2.94E2 6.94E2 Xe-135m 3.12E3 4.41E3 Xee135 1.81E3 3.97E3 Xe-137 1.42E3 1.39E4 Xe-138 8.83E3 1.43E4 Xe-139 0.00 0.00 Ar-41 8.84E3 1.29E4

• Regulatory Guide 1.109, October 1977, Table 8-1 times (1.0 E6 pCi/ Ci).

3-11

3.2 COMPLIANCE WITH 10CFR20 (GASE0US) 3.2.1 Noble Gases The gaseous effluent monitors setpoints are utilized to show compliance with 10CFR20 for noble gases. However, because they are based upon a conservative mix of radionuclides, the possibility exists that the setpoints could be exceeded and yet 10CFR20 limits may actually be met. Therefore, the following methodology has been provided in the event that if the alarm trip setpoints are exceeded, a determination may be made as to whether the actual releases have exceeded 10CFR20.

The dose rate in unrestricted areas resulting from noble gas effluents is limi';ed to 500 mrem / year to the total body and 3000 mrem / year to the skin.

Based upon NUREG 0133, the following are used to show compliance with 10CFR20.

[j Kj [ (TN), h jy + (TM), hj ,] 1 500 mrem /yr (3,2-1)

(3.2-2) lj (Lj + 1.1 Mj ) [(IM) hjy + (IM) h ,]j 3 3000 mrem /yr where:

(TN), = Annual average relative dilution for plant vent releases at the site boundary, sec/m ,3

= From Table A-1 for ground level releases.

= From Table A-10 for mixed mode releases only with

( upper wind speed of > 9 mph.

(TN), = Annual average relative dilution for condenser vacuum pump vent releases at the site boundary, sec/m3 ,

i

= From Table A-1 for ground level releases.

l j 3-12

Kj = The total body dose factor due to gamma emissions for noble gas radionuclide "1," mrem / year per pC1/m3 ,

Lj = The skin dose factor due to beta emissions for noble gas radionuclide "1," mrem / year per pCi/m3 ,

= The air dose factor due to gamma emissions for Mj noble gas radionuclide "1," mrad / year per pC1/m3 ,

1.1 = The ratio of the tissue to air absorption coeffi-cients over the energy range of the photon of interest, mrem / mrad (reference, NUREG 0133, October 1978).

h, j = The release rate of noble gas radionuclide "1" in gaseous effluents from the condenser vacuum pump vent pC1/sec.

• hjy = The release rate of noble gas radionuclide "i" in gaseous effluents from the plant vent pC1/sec.

The determination of limiting location for implementation of 10CFR20 for noble gases is a function of the radionuclide mix, isotopic release rate, and the

, meteorology.

l The radionuclide mix was based upon source terms calculated using the NRC GALE Code. They were calculated based upon the present operating mode of HBR.

They are presented in Table 3.2-1 as a function of release point.

The X/Q value utilized in the equations for implementation of 10CFR20 is based .

upon the maximum long-term annual average (T/4) in the unrestricted area.

Table 3.2-2 presents the distances from 2R to the nearest area for each of the 16 sectors as well as to the nearest residence, vegetable garden, ccW, goat, and beef animal. Long-term annual average (T/7) values for the ER

release points to the special locations in Table 3.2-2 are presented in Appen-dix A. A description of their deriviation is also provided n this appendix.

3-13

l l

l To select the limiting location, the highest annual average M value for the ground level releases and the mixed mode releases was used. Since mixed mode l releases may not necessarily decrease with distance (i.e., the site boundary may not have the highest M value), long-term annual average (M) values, calculated at the midpoint of 10 standard distances as given in Appendix A were also considered. For IBR, mixed mode release X/Q values decrease with distance for all directions except the WW, NW, and NNW so that the maximum site boundary X/Q is usually greater at the site boundary than at distances greater than the site boundary. In addition, the maximum site boundary X/Q for both the ground level and mixed mode releases occurs at the SSE site boun-dary. Therefore, the limiting location for implementation of 10CFR20 for noble gases is the SSE site boundary.

Values for Kj , Lj, and M j, which were used in the determination of the limiting location and which are to be used by IBR in Expressions 3.2-1 and 3.2-2 to show compliance with 10CFR20, are presented in Table 3.2-3. These values were taken from Table B-1 of MC Regulatory Guide 1.109, Revision 1.

The values have been multiplied by 1.0 E6 to convert microcuries to picoeuries for use in Expressions 3.2-1 and-3.2-2.

3.2.2 RadiJiodines and Particulates The dose rate in unrestricted area resulting from the release of radioiodines, l tritium, and particulates with half-lives > 8 days is limited to 1500 mrem /yr ~

l to any organ. Based upon NUREG 0133, the following 1s used to show compliance with 10CFR20.

~

[j P 9 [ (77i!), Qj , + ( M ), Q 4,] + (P9 +P 9 ) [ ( R ), Q , + ( M ), Q ,3 +

9 9 w

+P T IEI < 1500 mrem /yr (3.2-3)

(P T I M

)v OTV

• ITI leOT] e, 3-14

_ _ . _ _ . _ _ .__ __________.____P_ t__.

where:

=

(TM), Annual average relative dilution for plant vcat releases at the site boundary, sec/m ,3

= From Table A-1 for ground-level releases.

~

! = From Table A-10 for mixed mode releases to be used only with upper wind speeds > 9 mph.

=

(TM), Annual nyerage relative dilution for condenser vacuum pump vent releases at site boundary, sec/m3 ,

= From Table A-1 for ground-level releases.

(UM), = Annual average deposition factor for plant vent releases at site boundary, m-2,

= From Table A-3 for ground-level releases.

l

= From Table A-12 for mixed mode releases to be used only with upper wind speeds > 9 mph.

(UM), = Annual average deposition factor for condenser l vacuum pump vent releases at the site boundary, m -2 l

= from Table A-3 for ground-level releases.

P = Dose parameter for radionuclide "1" for the inha-9 I lation pathway, mrem / year per pC1/m3 ,

l l

l

, 3-15

}

P j

= Dose parameter for radionuclide "1" for the ground G plane pathway, mrem / year per C1/sec per m-2, P = Dose parameter for radionuclide "i" for either the 9

M cow milk or goat milk pathway, mrem / year per pC1/sec per m-2, P = Dose parameter for tritium for the inhalation T

I pathway, mrem / year per pCu/m3 ,

h,9

= Release rate of radionuclide "1" from the plant vent,pC1/sec.

Q = Release rate for tritium from the plant vent, T

V pC1/sec.

hTe = Release rate of tritium from the condenser vacuum pump vent, pC1/sec.

h,j = Release rate of radionuclide "1 " from the main condenser vacuum pump vent pC1/sec.

l In the calculation to show compliance with 10CFR20, only the inhalation, i ground plane, cow milk, and goat milk pathways are considered. Equation 3.2-3 is evaluated first at the limiting site. boundary. If the 1500 mrem /yr limit is exceeded at the limiting, site boundary when all pathways are considered present at this site boundary but the inhalation pathway contributed < 1500 meem/yr, then Equation 3.2-3 is evaluated at the limiting site boundary is 0.26 miles SSE, and the limiting relay pathway location is the cow milk path-way 4.2 miles E.

The determination of limiting location for implementation of 10CFR20 for radiotodines and particulates is a function of the same parameters as for noble gases plus a fourth, actual receptor pathway. The radionuclide mix was again based upon the source terms calculated using the GALE Code. The mix and the source terms are presented in Table 3.2-1 as a function of release point.

l l

3-16

The determination of the controlling site boundary location was based upon the highest site boundary D/Q value. The determination of actual receptor lim-iting location was based upon the milk pathway D/Q value and the Pj value for the respective milk pathway. Values for Pj were calculated for an infant for various radionuclides for the inhalation, ground plane, cow milk, and goat milk pathways using the methodology of NUREG 0133. The Pj values are pre-sented in Table 3.2-4. A description of the methodology used in calculating the Pj values is presented in Appendix B. The values of Pj reflect, for each radionuclide, the maximum Pj value for any organ for each individual pathway

of exposure. The goat milk pathway is present near HBR, as is the cow milk pathway. However, the cow milk pathway Pj values were utilized in the deter-mination of the controlling location because the product of the maximum cow milk pathway D/Q and P4 were greater than those for the goat. In addition, the goM. milk is not used for human consumption. For the case of an infant being present at the site at the site boundary or at the real pathway loca-tion, the ground plane pathway is not considered as a reasonable exposure pathway (i .e., Pj = 0). However, Pj values are presented in Table 3.2-4 for completeness.

The annual average D/Q values at the special locations, which will be utilized

in Equation 3.2-3, are obtained from the tables presented in Appendix A. The

! X/Q values which will be utilized in Equation 3.2-3 are also obtained from the tables presented in Appendix A. ' A description of the derivation of the X/Q and D/Q values is provided in Appendix A.

l l

l i

l I

3-17

TABLE 3.2-1 RELEASES FROM H. B. ROBINSON UNIT NO. 2*

(C1/yr)

Condenser Vacuum Plant Vent Pump Vent Isotope (Qo) (03) Total Kr-85m 2.0E0 1.0E0 3.0E0 Kr-85 1.6E2 0.00 1.6E2 Kr-87 1.0E0 0.00 1.0E0 Kr-88 4.0E0 2.0E0 6.0E0 Xe-131m 1.0E1 0.00 1.0E1 Xe-133m 4.0E0 0.00 4.0E0 Xe-133 3.7E2 1.8E1 3.9E2 Xe-135 8.0E0 2.0E0 1.0E1 I-131 3.6E-2 2.3E-2 5.9E-2 I-133 5.4E-2 3.4E-2 9.8E-2 l

l Mn-54 4.7E-3 0.00 4.7E-3 Fe-59 1.6E-3 0.00 1.6E-3 Co-58 1.6E-2 0.00 1.6E-2 Co-60 7.3E-3 0.00 7.3E-3 Sr-89 3.4E-4 0.00 3.4E-4 Sr-90 6.3E-5 0.00 6.3E-5 Cs-134 4.7E-3 0.00 4.7E-3 Cs-137 7.8E-3 0.00 7.8E-3 i

• Calculations based upon GALE Code and do not reflect actual release data from the Evaluation Comformance to the Design Objectives of 10CFR50, Appendix I.

These values are only for routine releases and not for a complete inventory of gases in an emergency.

l l

l 3-18 l

TABLE 3.2-2 DISTANCE TO SPECIAL LOCATIONS FOR THE H. B. RGLINSON PLANT (MILES)

Site Milk Milk Meat Nearest Nearest Sector Boundary Cow Goat Animal Resident Garden NNE 1.26 - - 1.65 1.3 1.4 NE 1.01 - - 1.16 1.2 1.3 ENE 0.86 - - 2.41 0.9 2.2 E 0.61 4.2 - 3.12 0.8 2.8 ESE 0.50 - -

1.99 0.6 0.6 SE 0.29 - - - 0.3 0.3 SSE 0.26 - - - 0.3 0.3 S 0.28 - - 2.32 0.3 0.4 SSW 0.29 - - 2.08 0.3 0.5 SW 0.36 - 2.5* 2.27 0.4 0.5 WSW 0.36 - - 2.69 0.4 0.6

, W 0.50 - -

3.97 0.6 0.6

! WNW 0.55 - - 4.07 0.7 0.9 NW 1.23 - - 1.60 1.3 1.3 NNW 1.89 - - 2.84 2.9 3.0 N 1.94 - - 2.93 2.9 2.9

• Milk is not presently used for human consumption.

3-19

TABLE 3.2-3 DOSE FACTORS FOR NGlLE GASES AND DAUGHTERS

• Total Body Skin Gamma Air Beta Air Dose Factor Dose Factor Dose Factor Dose Factor Kj Lj Mj Nj (mres/yr (mrem /yr (mrad /yr (mrad /yr -

i Radionuclide per pC1/m3 ) per pC1/m3 ) per uC1/m3 ) per uC1/m3 )

Kr-83m 7.56E-02 --- 1.93E+01 2.88E+02 Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1.34E+03 1.72E+01 1.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 1.03E+04 Kr-88 1.47E+04 2.37E+03 1.52E+04 2.93E+03 Kr-89 1.66E+04 1.01E+04 1.73E+04 1.06E+04 Kr-90 1.56E+04 7.29E+03 1.63E+04 7.83E+03 Xe-131m 9.15E+01 4.76E+02 1.56E+02 1.11E+03 Xe-133m 2.51E+02 9.94E+02 3.27E+02 1.48F+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03 Xe-135m 3.1EE+03 7.11E+02 3.36E+03 7.39E+02 Xe-135 1.81E+03 1.86E+03 1.92E+03 2.46E+03 Xe-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04 Xe-138 8.33E+03 4.13E+03 9.21E+03 4.75E+03 Ar-41 8.84E+03 2.69E+03 9.30E+03 3.28E+03

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

l l

3-20

TABLE 3.2-4 Pg VALUES FOR AN INFANT FOR THE H. B. R(BINSON UNIT NO. 2*

Isotope Inhalation Ground Plane Cow Milk Goat Milk H-3 6.47E2 0.00 2.38E3 4.86E3 P-32 2.03E6 0.00 1.60E11 1.93E11 Cr-51 1.28E4 6.67E6 4.79E6 5.65ES Mn-54 1.00E6 1.09E9 3.89E7 4.68E6 Fe-59 1.02E6 3.92E8 3.93 8 5.11E6 Co-58 7.77E5 5.29E8 6.06t/ 7.28E6 Co-60 4.51E6 4.40E9 2.10E8 2.52E7 Zn-65 6.47ES 6.89E8 1.90E10 2.29E9 Rb-86 1.90E5 1.28E7 2.22E10 2.67E9 Sr-89 2.03E6 3.16E4 1.27E10 2.66E10 Sr-90 4.09E7 0.00 1.21E11 2.55E11 Y-91 2.45E6 1.52E6 5.26E6 6.32E5 Zr-95 1.75E6 3.48E8 8.28E5 9.95E4 Mb-95 4.79ES 1.95E8 2.06E8 2.48E7 Ru-103 5.52E5 1.55E8 1.05E5 1.27E4 Ru-106 1.16E7 2.99E8 1.44E6 1.73E5 l

Ag-110m 3.67E6 3.14E9 1.46E10 1.75E9 Te-127m 1.31E6 1.18E5 1.04E9 1.24E8 Te-129m 1.68E6 2.86E7 1.40E9 1.68E8 Cs-134 7.03E5 2.81E9 6.79E10 2.04E11 Cs-136 1.35ES 2.13E8 5.76E9 1.73E10 Cs-137 6.12E5 1.15E9 6.02E10 1.81E11 Ba-140 1.60E6 2.94E7 2.41E8 2.89E7 Ce-141 5.17E5 1.98E7 1.37E7 1.65E6 Ce-144 9.84E6 5.84E7 1.33E8 1.60E7 I-131 1.48E7 2.46E7 1.06E12 1.27E12 I-132 1.69E5 1.78E6 1.39E2 1.64E2 I-133 3.56E6 3.54E6 9.80E9 1.18E10 1-135 6.96ES 3.67E6 2.27E7 2.68E7 l

• Units are pres /yr per C1/m3for H-3 and the inhalation pathway and mrem /yr per pC1/sec per m-2 for the food and ground plane pathways.

3-21

r 3.3 COMPLIANCE WITH 10CRF50 (GASEOUS)

, , p. .

3.3.1 Noble Gases - -

3.3.1.1 Cumulation of Doses Based upon NUREG 0133, the air dose in the unrestricted area due to noble gases released in gaseous effluents can be detemined by the following equations: '

Dy = 3.17 x 10-8 1M9 9

[ ( M )y Tjy + ( M )y 49 ,+ ( M ),3 9,3 (3.3-1)

\

D, = 3.17 x 104 { 9, N, & (M)y1 9 , + ( M ), q9, + ( S ), V 9

,] (3.3-2) s where:

D y

= The air dose from gamma radiation, mrad.

i Dp = The air dose from beta radiation, mrad.

M9

= The air dose ; factor due to gamma emissions for each identifiedi noble gas radionuclide "1,"

mrad / year per pCi/m3 .

N9

= The air dcse factor due to beta emissions for each identified noble gas radionuclide "i," mrad / year per pC1/m3 ,

1 ,

(M), = The annual average dilution for areas at or beyond the unrestricted area boundary for long-term plant vent releases (> 500 hrs / year), sec/m3 ,

= From Table A-1 for ground level releases.

l 3-22

= From Table A-10 for mixed mode releases to be used only with upper wind speeds > 9 mph. _

(TR), =

The. lution for areas at or beyond the unre-stricteif area boundary for short-term vent releases (< 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> / year), sec/m3 .

-= From Table A-7 for ground level releases.

= From Table A-16 for mixed mode releases.

(TM), = Annual average relative dilution for condenser vacuum pump vent releases at the site boundary.

(> 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> / year), sec/m3 ,

=5 From Table A-1 for ground level releases; qiy = The average release of noble gas radionuclide "1" in gaseous releases for short-term plant releases

(< 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> / year), pC1; Qj, = The average release of noble gas radionuclide "i"

, in gaseous releases for long-term condenser vacuum l

pump vent releases (> 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> / year), pCi; t

Qj, = The average release of noble gas radionuclide "i" in gaseous effluents for long-term vent releases

(> 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> / year), pC1; 3.17 x 10-8 = The inverse of the number of seconds in a year (sec/ year)-1 At MIR the limiting location is 0.26 miles SSE. Based upon the tables pre-sented in Appendix A, substitution of the appropriate X/Q values into Equation 3.3-1 would yield an equation with the short-term X/Q value being less than the long-term value. Therefore, for this document, only the long-term annual average TMyalues (i.e., the more conservative values) will be used.

3-23

• ^

The determination of the limiting location for implementation of 10CFR50 is a function of parameters such as radionuclide mix, isotopic release, and meteo-rology. The radionuclide r,ix was based upon source terms calculated using the PRC GALE Code and is presented in Table 3.2-1 as a function of release point.

The only source of short-term releases from the plant vent are containment purges.

To select the limiting location, the highest annual average T/IT value for ground level and mixed mode releases and the highest short-term X/Q value for ground ' level and mixed mode releases were considered. Since mixed mode re-leases may increase and then decrease with distance (i.e., the site boundary m4y not have the highest X/Q value), long-term X/Q values were calculated at the midpoint of 10 standard distances as given in Appendix A. The calculated values decreased with the distance for all but the WNW, NW, and NNW sectors.

The values for these sectors were not found to be limiting such that the maximum site boundary X/Q for both long-tcrm and short-term ground level and mixed mode releases occurred at the SSE site boundary. The limiting location for implementation of 10CFR20 for noble gases is the SSE site boundary.

Values for Mj and Nj , which are utilized in the calculation of the gamma air and beta air doses in Equation 3.3-1 to show compliance with 10CFR50, were presented in Table 3.2-3. These values originate from NUREG 0472, Revision 0, and were taken from Table B-1 of the NRC Regulatory Guide 1.109, Revision 1.

The values have been multiplied by 1.0 E6 to convert from picocuries to micro-curies.

The following relationship should hold for ER to show compliance with 2R's Technical Specification 3.9.4.1.

For the calendar quarter:

l l

D 1 5 mrad (3.3-3) 0 1 10 mrad (3.3-4)

I 3-24

l l

For the calendar year:

D 1 10 mrad (3.3-5)

D, 1 20 mrad (3.3-6)

The quarterly limits given above represent one-half of the annual design ob-jectives of Section II.B.1 of Appendix I of 10CFR50. If any of the limits of Equations 3.3-3 through 3.3-6 are exceeded, a special report pursuant to Tech-nical Specification 6.9.4.a must be fileo with the NRC. This report complies with Section IV.A of Appendix I of 10CFR50.

3.3.1.2 Projection of Doses Doses resulting from the release of gaseous effluents will be projected once l per month. The doses will be projected using Equations 3.3-1 and 3.3-2.

3.3.2 Radiciodine and Particulates l

3.3.2.1 Cumulation of DosesSection II.C of Appendix I of 10CFR50 limits the release of radiciodines and radioactive material in particulate form from each reactor such that estimated dose or dose commitment to an individual in an unrestricted area from all pathways of exposure is not in excess of 15 mrem to any organ. Based upon l NUREG 0133, the dose to an organ of an individual from radioiodines, tritium, I

and particulates with half-lives > 8 days in gaseous effluents released to unrestricted areas can be determined by the following equation:

1 D

=3.17x10'0{$ R

$ [(M)y Q9 , + (X/q),q9, + (X/Q), Q9 , ] +

(R 9 +R9 ) [ (TM), Q9, (lT/ii)y q$y + ( M), Q9 ,] + (3.3-7)

(R T g

+RTy ) E (TN)v OTv + (TIii)v 9Tv + ITN)e OTe3 l

3-25 , ,

where:

D5 = Dose to any organ t from radiciodines and partic-ulates , mrem.

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

(sec/ year)-1 (TN), = Annual average relative concentration for plant vent releases (> 500 hrs /yr) sec/m3 ,

= From Table A-1 for ground level releases.

= From Table A-10 for mixed mode releases only to be used with wind speeds > 9 mph.

l (TM), = Annual average dilution for condenser vacuum pump vent releases (> 500 ho'urs/yr) sec/m3 ,

= From Table A-1 for ground level releases.

(LIM), = Annual average deposition factor for plant vent releases (> 500 hrs /yr) m-2,

= From Table A-3 for ground level releases.

! = Fron Table A-12 for mixed node releases only to be used with upper wind speeds > 9 mph.

(D/q)y = Relativ6 deposition facter f9.e short-term plant vent releases (< 500 hrs /yr), m-2,

= From Table A-9 for ground level releases.

= From Table A-18 for mixed mode releases only to be used with upper wind speeds > 9 mph.

3-26

(D/Q), = Annual average relative deposition factor for condenser tacuum pump vent releases (> 500 hrs /

yr), m-2, i

= From Table A-3 for ground level release:;.

Qj, = Release of radionuclide "1" in gaseous effluents for long-term condenser vacuum pump vent releases

(> 500 hrs /yr), C1. ,

Qjy = Release of radionuclide "i" in gaseous effluents

, for long-term plant tent releases (> 500 hrs /yr),

pC1.

qjy = Release of radionuclide "i" in gaseous effluents for short-term plant vent releases (1500 hrs /yr),

pC1.

Rj = he facw & an wgan & ramnucWe 9" &

6 the ground plane exposure pathway, mrem /yr per pC1/sec per m-2, ,

Rg

= Dose factor for an organ for radionuclide "1" for the inhalation pathway, mrem /yr per pC1/m3 ,

Rj = Dose factor for an organ for radionuclide "i" for the vegetable pathway, mrem /yr per Ci/m-2, RT

= Dose factor for an organ for tritium for the vege-table pathway, mrem /yr per pC1/m3 ,

R = Dose factor for an organ for tritium for the inha-Tg lation pathway, mrem /yr per pC1/m3 ,

QTV

= Release of tritium in gaseous effluents for long-term vent releases (> 500 hrs /yr), pCi.

3-27

1 1

Qe T

= Release of tritium in gaseous effluents for long-term condenser vacuum pump releases -(> 500 hrs /yr), pC1.

qTV

= Release of tritium in gaseous effluents for short-term plant vent releases (< 500 hrs /yr), C1.

To show compliance with 10CFR50, Equation 3.3-7 is evaluated at the limiting pathway location. At IBR this location is the vegetable garden 0.3 miles in the SSE sector. The critical receptor is a child. Substitution of the appro-priate X/Q and D/Q values from trbles in Appendix A into Equation 3.3-7 would yield an equation with the short-term X/Q and D/Q values being less than the long-term values. Therefore, for this document, only long-term annual X/Q and D/Q values (i.e., more conservative values) are used.

The determination of a limiting location for implementation of 10CFR50 for radiofodines and particulates is a function of:

1. Radionuclide mix and isotopic release

2. Meteorology

3. Exposure pathway

4. Receptor's age In the determination of the limiting location, the radionuclide mix of radio-iodines and particulates was based upon the source terms calculated using the GALE Code. This mix is presented in Taole 3.2-1 as a function of release point. The only source of short-term releases from the plant vent is contain-ment purges.

In the determination of the limiting location, all of the exposure pathways, as presented in Table 3.2-2, were evaluated. These include cow milk, goat milk, beef and vegetable ingestion, and inhalation and grout:d plane expo-sure. An infant was assumed to be present at all milk pathway locations. A child was assured to be present at all vegetable garden and beef animal loca-tions. The ground plane exposure' pathway was not considered a viable pathway for an infant. Naturally, the inhalation pathway was present everywhere an individual was present.

3-28

I 2R Technical Specification 4.20.2.1 requires that a land-use census survey be conducted on an annual basis. The age groupings at the various receptor locations are also determined during this survey; a new limiting location and receptor age group can result.

For the determination of the limiting location, the highest D/Q values for the vegetable garden, cow milk, and goat milk pathways were selected. The thyroid dose was calculated at each of these locations using the radionuclide mix and releases of Table 3.2-1. Based upon these calculations, it was determined that the limiting receptor pathway is the vegetable / child pathway.

In the determination of the limiting location, annual average D/Q and X/Q values are used. A description of the derivation of the various X/Q and D/Q values is presented in Appendix A.

Short-term and long-term X/Q and D/Q values for ground level releases and for long-term mixed mode releases are provided in tables in Appendix A. They may be utilized if an additional special location arises different from those presented in the special locations of Table 3.2-2.

[

Tables 3.3-1 through 3.3-19 present Rj values for the total body, GI-tract, bone, liver , kidney, thyroid, and lung organs for the ground plane, inhala-tion, cow milk, goat milk, vegetable, and meat ingestion pathways for the infant, child, teen, and adult age groups as appropriate to the pathways.

These values were calculated using the methodology described in NUREG 0133 using a grazing period of eight months. A description of the methodology is presented in Appendix B.

The following relationship should hold for ISR to show compliance with HBR Technical Specification 3.9.5.1.

For the calendar quarter:

D < 7.5 mrem (3.3-8) l 3-29

For the calendar year:

0 < 15 mrem (3.3-9)

The quarterly limits given above represent one-half the annual design objec-tives of Section II.C of Appendix I of 10CFR50. If any of the limits of Equations 3.3-8 or 3.3-9 are exceeded, a special report pursuant to Technical l

Specification 6.9.4.a must be filed with the NRC. This report complies with Section IV.A of Appendix I of 10CFR50.

3.3.2.2 Projection of Doses Doses resulting from release of radiofodines and particulates will be pro-jected once per month using Equation 3.3-7.

3-30

TABLE 3.3-1 R W.ES FGt TIE H. B. RSINSON STEAM ELECTRIC PLANT

• PAT}ARY = Ground Ibc11de T.Bo@ GI-Tract Bone Liver Kihey Thyroid M Skin Cr-51 4.66E 06 4.66E 06 4.66E 06 4.66E 06 4.66E 06 4.66E 06 4.66E 06 5.51E 06 Mn-54 1.34E 09 1.34E 09 1.34E 09 1.34E 09 1.34E 09 1.34E 09 1.34E 09 1.57E 09 Fe-59 2.75E 08. 2.75E 08 2.75E 08 2.75E 08 2.75E 08 2.75E 08 2.75E 08 3.23E 08 Co-58 3.79E 08 3.79E 08 3.79E'08 3.79E 08 3.79E 08 3.79E 08 3.79E 08 4.44E 09 Co-60 2.15E 10 2.15E 10 2.15E 10 2.15E 10 2.15E 10 2.15E 10 2.15E 10 2.52E 10 Zn-65 7.49E 08 7.49E 08 7.49E 08 7.49E 08 7.49E 03 7.49E 08 7.49E 08 8.61E 08

Rb-86 8.99E 06 8.99E 06 8.99E 06 8.99E 06 8.99E 06 8.99E 06 8.99E 06 1.03E 07 Sr-89 2.23E 04 2.23E 04 2.23E 04 2.23E 04 2.23E 04 2.23E 04 2.23E 04 2.58E 04 Y-91 1.08E 06 1.08E 06 1.0E 06 1.08E 06 1.08E 06 1.08E 06 1.08E 06 1.22E 06 Zr-95 2.49E 08 2.49E 08 2.49E 08 2.49E 08 2.49E 08 2.49E 08 2.49E 08 2.89E 08 Nb-95 1.36E 08 1.36E 08 1.36E 08 1.36E 08 1.36E 08 1.36E 08 1.36E 08 1.60E 08 Ru-103 1.09E 08 1.09E 08 1.09E 08 1.09E 08 1.09E 08 1.09E 08 1.09E 08 1.27E 08 Ru-106 4.19E 08 4.19E 08 4.19E 08 4.19E 08 4.19E 08 4.19E 08 4.19E 08 5.03E 08 Ag-110M 3.48E*09 3.4E 09 3.48E 09 3.48E 09 3.48E 09 3.48E 09 3.48E 09 4.06E 09 Te-127M 9.15E 04 9.15E 04 9.15E 04 9.15E 04 9.15E 04 9.15E 04 9.15E 04 1.08E 05 Te-129M 2.00E 07 2.00E 07 2.00E 07 2.00E 07 2.00E 07 2.00E 07 2.00E 07 2.34E 07 I-131 1.72E 07 1.72E 07 1.72E 07 1.72E 07 1.72E 07 1.72E 07 1.72E 07 2.09E 07 I-132 1.24E 06 1.24E 06 1.24E 06 1.24E 06 1.24E 06 1.246 06 1.24E 06 1.46E 06 I-133 2.47E 06 2.47E 06 2.47E 06 2.47E 06 2.47E 06 2.47E 06 2.47E 06 3.00E 06 I-135 2.56E 06 2.56E 06 2.56E 06 2.56E 06 2.56E 06 2.56E 06 2.56E 06 2.99E 06 Cs-134 6.32E 09 6,82E 09 6.82E 09 6.82E 09 6.82E 09 6.82E 09 6.82E 09 7.96E 09 Cs-136 1.49E 08 1.49E 08 1.49E 08 1.49E 08 1.49E 08 1.49E 08 1.49E 08 1.69E 08 Cs-137 1.03E 10 1.03E 10 1.03E 10 1.03E 10 1.03E 10 1.03E 10 1.03E 10 1.20E 10 Ba-140 2.05E 07 2.05E 07 2.05E 07 2.05E 07 2.05E 07 2.05E 07 2.05E 07 2.34E 07 Ce-141 1.36E 07 1.36E 07 1.36E 07 1.36E 07 1.36E 07 1.36E 07 1.36E 07 1.53E 07 Ce-144 6.95E 07 6.95E 07 6.95E 07 6.95E 07 6.95E 07 6.95E 07 6.95E 07 8.03E 07

• R Values in units f mrem /yr per micro-C1/m-3 for inhalation and tritium, and in units of M~ mrem /yr per micro-Ci/sec for all others.

3-31

TABLE 3.3-2 g

R MLES FOR llE H. B. RGlINSON STEAM ELECTRIC PUU(T*

PATHWAY = Veget AE Gt0UP = Adult hclide T.Bo@ GI-Tract Bone Liver Kihey Thyroid h Skin H-3 2.28E 03 2.28E 03 0.00E 01 2.28E 03 2.28E 03 2.28E 03 2.28E 03 2.28E 03 P-32 5.91E 07 1.72E 08 1.53E 09 9.51E 07 0.00E 01 0.00E 01 0.0E 01 0.0E 01 Cr-51. 4.60E 04 1.16E 07 0.00E 01 0.00E 01 1.01E 04 2.75E 04 6.10E 04 0.00E 01 m-54 5.83E 07 9.36E 08 0.0E 01 3.05E 08 9.09E 07 0.0 % 01 0.00E 01 0.00E 01 Fe-59 1.12E 08 9.75E 08 1.24E 08 2.93E 08 0.00E 01 0.00E 01 8.17E 07 0.00E 01 Co-58 6.71E 07 6.07E 08 0.00E 01 2.99E 07 0.00E 01 0.00E 01 0.00E 01 0.00E 01 Co-60 3.67E 08 3.12E 09 0.00E 01 1.66E 08 0.00E 01 0.00E 01 0.00E 01 0.00E 01 Zn-65 5.77E 08 8.04E 08 4.01E 08 1.28E 09 8.54E 08 0.0E 01 0.00E 01 0.00E 01 Rb-86 1.03E 08 4.36E 07 0.00E 01 2.21E 08 0.00E 01 0.00E 01 0.00E 01 0.00E 01 Sr-89 2.87E 08 1.6 E 09 1.00E 10 0.00E 01 0.00E 01 0.00E 01 0.00E 01 0.00E 01 Sr-90 1.64E 11 1.93E 10 6.7E 11 0.00E 01 0.00E 01 0.00E 01 0.00E 01 0.00E 01 Y-91 1.34E 05 2.76E 09 5.01E 06 0.0E 01 0.0E 01 0.0E 01 0.00E 01 0.0E 01 Zr-% 2.51E 05 1.17E 09 1.16E 06 3.71E 05 5.82E 05 0.00E 01 0.00E 01 0.00E 01 Nb-95 4.19E 04 4.73E 08 1.40E 05 7.79E 04 7.70E 04 0.00E 01 0.00E 01 0.00E 01 Ru-103 2.04E 06 5.53E 08 4.74E 06 0.00E 01 1.81E 07 0.00E 01 0.00E 01 0.0E 01 l

Ru-106 2.46E 07 1.26E 10 1.94E 08 0.0E 01 3.75E 08 0.0E 01 0.0E 01 0.0E 01 Ag-110M 6.23E 06 4.28E 09 1.13E 07 1.05E 07 2.06E 07 0.00E 01 0.00E 01 0.00E 01 TG-127M 6.12E 07 1.68E 09 5.02E 08 1.80E 08 2.04E 09 1.28E 08 0.00E 01 0.00E 01 Te-129M 4.71E 07 1.50E 09 2.98E 08 1.11E 08 1.24E 09 1.02E 08 0.00E 01 0.0% 01 1-131 6.61E 07 3.04E 07 8.07E 07 1.15E 08 1.98E 03 3.78E 10 0.00E 01 0.00E 01 I-132 5.21E 01 2.80E 01 5.57E 01 1.49E 02 2.37E 02 5.21E 03 0.00E 01 0.0% 01 I-133 1.12E 06 3.30E 06 2.11E 06 3.67E 06 6.40E 06 5.39E 08 0.0E 01 0.0E 01 1-135' 3.91E 04 1.20E 05 4.05E 04 1.06E 05 1.70E 05 7.00E 06 0.00E 01 0.0% 01 Cs-134 8.83E 09 1.89E 08 4.54E 09 1.08E 10 3.49E 09 0.00E 01 1.16E 09 0.0 % 01

( Cs-136 1.19E 08 1.88E 07 4.19E 07 1.66E 08 9.21E 07 0.00E 01 1.26E 07 0.0E 01 Cs-137 5.94E 09 1.76E 08 6.63E 09 9.07E 09 3.08E 09 0.00E 01 1.02E 09 0.00E 01 Ba-140 8.40E 06 2.64E 08 1.28E 08 1.61E 05 5.47E 04 0.0E 01 9.22E 04 0.0E 01 Ce-141 1.48E 04 4.99E 08 1.93E 05 1.315 05 6.07E 04 0.0E 01 0.0E 01 0.00E 01 l Ce-144 1.69E 06 1.05E 10 3.15E 07 1.32E 07 7.80E 06 0.0 % 01 0.0E 01 0.0% 01 i

• R Values in units of mrem /yr per micro-C1/m-3 for inhalation and tritium, and in units of M-2 mrem /yr per micro-Ci/sec for all others.

l l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ - - - - - - . . -

TABLE 3.3-3 R MLES F(R TE H. 8. R(BINSON STEAM E!ICTRIC PUUE*

i PATHWAY = Yeget AE GROUP = Teen kclide T.Bo@ GI-Tract Bone Liver Kidney Thyroid p_ng Skin H-3 2.61E 03 2.61E 03 0.'00E 01 2.61E 03 2.61E 03 2.61E 03 2.61E 03 2.61E 03 P-32 6.80E 07 1.47E 08 1.75E 09 1.09E 08 0.00E 01 0,00E 01 0.0E 01 0.00E 01 Cr-51 6.11E 04 1.03E 07 0.00E 01 0.00E 01 1.34E 04 3.39E 04 8.72E 04 0.00E 01 Mn-54 8.79E 07 9.09E 08 0.00E 01 4.4 E 08 1.32E 08 0.0E 01 0.00E 01 0.00E 01 Fe-59 1.60E 08 9.78E 08 1.77E 08 4.14E 08 0.00E 01 0.0E 01 1.30E 08 0.00E 01 Co-58 9.79E 07 5.85E 08 ,0.00E 01 4.25E 07 0.0E 01 0.00E 01 0.00E 01 0.00E 01 Co-60 5.57E 08 3.22E 09 0.00E 01 2.47E 08 0.00E 01 0.00E 01 0.00E 01 0.00E 01 Zn-65 8.68t 08 7.88E 08 5.36E 08 1.86E 09 1.19E 09 0.00E 01 0.00E 01 0.00E 01

~

Rb-86 1.30E 08 4.09E 07 0.00E 01 2.76E 08 0.00E 01 0.00E 01 0.00E 01 0.00E 01 Sr-89 4.36E 08 1.81E 09 1.52E 10 0.00E 01 0.0E 01 0.0E 01 0.00E 01 0.00E 01 Sr-90 2.05E 11 2.33E 10 8.32E 11 0.00E 01 0.00E 01 0.00E 01 0.00E 01 0.0E 01 f-91 2.06E 05 3.15E 09 7.68E 06 0.00E 01 0.00E 01 0.0E 01 0.00E 01 0.00E 01 Zr-95 3.68E 05 1.23E 09 1.69E 06 5.35E 05 7.86E 05 0.00E 01 0.00E 01 0.00E 01 2-95 5.77E 04 4.48E 08 1.89E 05 1.05E 05 1.02E 05 'O.00E 01 0.00E 01 0.00E 01 Ru-103 2.90E 06 5.66E 08 6.78E 06 0.00E 01 2.39E 07 0.00E 01 0.00E 01 0.00E 01 Ru-106 3.93E 07 1.5E 10 3.12E 08 0.00E 01 6.02E 08 0.00E 01 0.0E 01 0.00E 01 Ag-1191 9.39E 06 4.34E 09 1.63E 07 1.54E 07 2.95E 07 0.00E 01 0.00E 01 0.00E 01 Te-127M 9.44E 07 1.98E 09 7.93E 08 2.81E 08 3.22E 09 1.89E 08 0.00E 01. 0.00E 01 l

Te-129M 6.79E 07 1.61E 09 4.29E 08 1.59E 08 1.79E 08 1.38E 08 0.00E 01 0.00E 01 I-131 5.77E 07 2.13E 07 7.68E 07 1.07E 08 1.85E 08 3.14E 10 0.00E 01 0.00E 01 I-132 4.72E 01 5.72 01 5.02E 01 1.31E 02 2.07E 02 4.43E 03 0.00E 01 0.00E 01 I-133 1.01E 06 2.51E 03 1.96E 06 3.32E 06 5.83E 06 4.64E 08 0.00E 01 0.00E 01 I-135 3.49E 04 1.04E 05 3.65E 04 9.42E 04 1.49E 05 6.06E 06 0.00E 01 0.00E 01 Cs-134 7.54E 09 2.02E 08 6.90E 09 1.62E 10 5.16E 09 0.0E 01 1.97E 09 0.0E 01 Cs-136 1.13E 08 1.35E 07 4.28E 07 1.68E 08 9.16E 07 0.00E 01 1.44E 07 0.00E 01 Cs-137 4.90E 09 2.00E 08 1.06E 10 1.41E 10 4.78E 09 0.00E 01 1.86E 09 0.00E 01 Ba-140 8.88E 06 2.12E 08 1.38E 08 1.69E 05 5.72E 04 0.00E 01 1.14E 05 0.00E 01 Ce-141 2.12E 04 5.29E 08 2.77E 05 1.85E 05 8.70E 04 0.00E 01 0.0E 01 0.00E 01 Ce-144 2.71E 06 1.276 10 5.04E 07 2.09E 07 1.25E 07 0.0E 01 0.00E 01 0.00E 01

• R Values in units of mrem /yr per micro-Ci/m-3 for inhalation and tritium, and in units of M-2 mrem /yr per micro-Ci/sec for all others.

TABLE 3.3-4 R MLES FGt TE H. B. R(BINSON STEAM ELECTRIC PLANT

• l 1

PATINAY = Veget AGE (ROUP = Child gelide T.Bo@ GI-Tract Bone Liver Kidney Thyroid Lyng, _ Skin H-3 i.04E03 4.04E 03 0.00E 01 4.04E 03 4.04E 03 4.04E 03 4.04E 03 4.04E 03 P-32 1.42E 08 1.01E 08 3.67E 09 1.72E 08 0.00E 01 0.0E 01 0.0% 01 0.00E 01 Cr-51 1.16E 05 6.15E 06 0.00E 01 0.00E 01 1.76E 04 6.44E 04 1.18E 05 0.0E 01 Mn-54 1.73E 08 5.44E 08 0.00E 01 6.49E 08 1.82E 08 0.0E 01 0.0% 01 0.00E 01 Fs-59 3.17E 08 6.62E 08 3.93E 08 6.36E 08 0.0% 01 0.00E 01 1.84E 08 0.0E 01 Co-58 1.92E 08 3.66E 08 0.0E 01 6.27E 07 0.0E 01 0.0% 01 0.00E 01 0.00E 01 Co-60 1.11E 09 2.08E 09 0.00E 01 3.76E 08 0.00E 01 0.0E 01 0.00E 01 0.00E 01 Zn-65 1.7E 09 4.81E 08 1.03E 09 2.74E 09 1.73E 09 0.0 % 01 0.0E 01 0.00E 01 Rb-86 2.81E 08 2.94E 07 0.0E 01 4.56E 03 0.00E 01 0.00E 01 0.00E 01 0.00E 01 Sr-89 1.03E 09 1.4 E 09 3.62E 10 0.0E 01 0.00E 01 0.00E 01 0.00E 01 0.0E 01 Sr-90 3.49E 11 1.86E 10 1.38E 12 0.00E 01 0.0% 01 0.00E 01 0.00E 01 0.00E 01 Y-91 4.89E 05 2.44E 09 1.83E 07 0.00E 01 0.00E 01 0.0E 01 0.00E 01 0.00E 01 Zr-95 7.44E 05 8.71E 08 3.80E 06 8.35E 05 1.2% 06 0.00E 01 0.0% 01 0.00E 01 Nb-95 1.12E 05 2.91E 08 4.04E 05 1.57E 05 1.48E 05 0.0% 01 0.00E 01 0.0E 01 Ru-103 5.86E 06 3.94E 08 1.52E 07 0.00E 01 3.84E 07 0.00E 01 0.00E 01 0.0E 01 Ru-106 9.38E 07 1.17E 10 7.52E 08 0.0E 01 1.02E 09 0.00E 01 0.0E 01 0.00E 01 Ag-110M 1.87E 07 2.78E 09 3.46E 07 2.34E 07 4.35E 07 0.00E 01 0.00E 01 0.00E 01 Te-127M 2.26E 08 1.54E 09 1.90E 09 5.12E 08 5.42E 09 4.55E 08 0.0E 01 0.00E 01 Te-129M 1.55E 08 1.22E 09 9.98E 08 2.79E 08 2.93E 09 3.22E 08 0.0 E 01 0.0E 01 I-131 8.16E 07 1.23E 07 1.43E 08 1.44E 08 2.36E 08 4.75E 10 0.0E 01 0.00E 01 I 132 7.53E 01 1.93E 02 8.91E 01 1.64E 02 2.51E 02 7.60E 03 0.00E 01 0.0E 01 1-133 1.67E 06 1.78E 06 3.57E 06 4.42E 06 7.36E 06 8.21E 08 0.00E 01 0.00E 01 I-135 5.54E 04 8.92E 04 6.5% 04 1.17E 05 1.79E 05 1.04E 07 0.0% 01 0.0E 01 Cs-134 5.40E 09 1.38E 08 1.56E 10 2.56E 10 7.93E 09 0.0E 01 2.84E 09 0.00E 01 Cs-136 1.43E 08 7.77E 06 8.04E 07 2.21E 08 1.18E 08 0.00E 01 1.76E 07 0.0E 01 Cs-137 3 52E 09 1.5E 08 2.49E 10 2.39E 10 7.78E 09 0.0E 01 2.80E 09 0.00E 01 Ba-140 1.61E 07 1.40E 08 2.76E 08 2.42E 05 7.87E 04 0.00E 01 1.44E 05 0.00E 01 Ce-141 4.75E 04 3.99E 08 6.42E 05 3.20E 05 1.40E 05 0.00E 01 0.00E 01 0.00E 01 CI-144 6.49E 06 9.94E 09 1.22E 08 3.81E 07 2.11E 07 0.00E 01 0.0E 01 0.00E 01 I

• R Values in units of mrem /yr per micro-C1/m-3 for inhalation and tritium, and in units of M-2 mrem /yr per micro-C1/sec for all others.

l TABLE 3.3-5 I

R MLES F(R TE H. B. R(BINSON STEAM ELECTRIC PUUE*

PATitRY = Meat

! AGE GROUP = Adult l

lhc11de T. Bob GI-Tract Sone Liver Kidney Thyroid M Skin H-3 3.27E 02 3.27E 02 0.00E 01 3.27E 02 3.27E 02 3.27E 02 3.27E 02 3.27E 02 P-32 1.18E 08 3.43E 08 3.05E 09 1.89E 08 0.00E 01 0.0 E 01 0.00E 01 0.00E 01 Cr-51 4.27E 03 1.08E 06 0.00E 01 0.00E 01 9.42E 02 2.56E 03 5.67E 03 0.00E 01 Mn-54 1.06E 06 1.71E 07 0.00E 01 5.57E 06 1.6E 06 0.00E 01 0.0E 01 0.00E 01 Fs-59 1.43E 08 1.25E 09 1.59E 08 3.74E 08 0.00E 01 0.00E 01 1.04E 08 0.00E 01 Co-58 2.43E 07 2.2 E 08 0.00E 01 1.08E 07 0.00E 01 0.0E 01 0.0E 01 0.00E 01 Co-60 1.03E 08 8.76E 08 0.00E 01 4.66E 07 0.00E 01 0.00E 01 0.00E 01 0.00E 01 Zn-65 3.58E 08 4.9E 08 2.49E 08 7.91E 08 5.29E 08 0.00E 01 0.00E 01 0.00E 01 Rb-86 1.42E 08 6.00E 07 0.00E 01 3.04E 08 0.00E 01 0.00E 01 0.00E 01 0.00E 01 Sr-89 5.23E 06 2.92E 07 1.82E 08 0.00E 01 0.00E 01 0.00E 01 0.00E 01 0.0E 01 Sr-90 2.02E 09 2.38E 08 8.22E 09 0.00E 01 0.00E 01 0.00E 01 0.00E 01 0.00E 01

(-91 1.80E 04 3.71E 08 6.75E 05 0.00E 01 0.00E 01 0.00E 01 0.00E 01 0.00E 01 Zr-95 2.43E 05 1.14E 09 1.12E 06 3.59E 05 5.64E 05 0.00E 01 0.00E 01 0.00E 01 Nb-95 4.12E 05 4.65E 09 1.38E 06 7.66E 05 7.5E 05 0.00E 01 0.00E 01 0.00E 01 Rr-103 2.72E 07 7.38E 09 6.32E 07 0.00E 01 2.41E 08 0.00E 01 0.00E 01 0.00E 01 Ru-106 2.19E 08 1.12E 11 1.73E 09 0.00E 01 3.35E 09 0.00E 01 0.00E 01 0.00E 01 Ag-11(N 2.34E 06 1.61E 09 4.27E 06 3.95E 06 7.76E 06 0.00E 01 0.00E 01 0.00E 01 l

Ts-127M 1.00E 08 2.76E 09 8.22E 08 2.94E 08 3.34E 09 2.10E 08 0.00E 01 0.00E 01 Te-129M 1.17E 08 3.73E 09 7.40E 08 2.76E 08 3.09E 09 2.54E 08 0.00E 01 0.00E 01 l I-131 5.77E 06 2.66E 06 7.04E 06 1.01E 07 1.73E 07 3.30E 09 0.00E 01 0.00E 01 I-133 1.51E-01 4.46E-01 2.85E-01 4.96E-01 8.66E-01 7.29E 01 0.00E 01 0.00E 01 I-135 6.07E-17 1.86E-16 6.28E-17 1.64E-16 '2.64E-16 1.08E-14 0.00E 01 0.00E 01 Cs-134 7.81E 08 1.67E 07 4.01E 08 9.55E 08 3.09E 08 0.00E 01 1.G3E 08 0.00E 01 Cs-136 2.14E 07 3.3 I 06 7.53E 06 2.97E 07 1.65E 07 0.0E 01 2.27E 06 0.00E 01 Cs-137 4.99E 08 1.47E 07 5.57E 08 7.61E 08 2.58E 08 0.00E 01 8.59E 07 0.00E 01 Ba-140 1.20E 06 3.77E 07 1.83E 07 2.3E 04 /.82E 03 0.00E 01 1.32E 04 0.00E 01 Ce-141 6.46E 02 2.18E 07 8.42E 03 5.69E OJ 2.65E 03 0.00E 01 0.00E 01 0.00E 01 Ce-144 4.70E 04 2.96E 08 8.75E 05 3.66E 05 2.17E 05 0.00E 01 0.0E 01 0.00E 01

• R Values in units of mrem /yr per micro-C1/m-3 for inhalation and tritium, and in units of M-2 mrem /yr per micro-C1/sec for all others.

TABLE 3.3-6 R MLES FGt TE H. B. RQlINSON STEAM ELECTRIC PUUE* 1 j

PATMAY = Meat l

AGE Gt00P = Teen lbc11de T.8o@ GI-Tract Bone Liver Kidne Thyroid M Skin H-3 1.95E 02 1.95E 02 0.00E 01 i.95E02 1.95E 02 1.95E 02 1.95E 02 1.95E 02 P-32 9.98E 07 2.16E 08 2.58E 09 1.60E 08 0.00E 01 0.0E 01 0.00E 01 0.0% 01 Cr-51 3.42E 03 5.75E 05 0.00E 01 0.00E 01, 7.49E 02 1.90E 03 4.88E 03 0.00E 01 Mn-54 8.43E 05 8.72E 06 0.00E 01 4.25E 06 1.27E 06 0.00E 01 0.00E 01 0.00E 01 Fe-59 1.15E 08 7.02E 08 1.27E 08 2.97E 08 0.00E 01 0.00E 01 9.36E 07 0.00E 01 Co-58 1.93E 07 1.15E 08 0.00E 01 8.36E 06 0.00E 01 0.00E 01 0.00E 01 0.00E 01 Co-60 8.15E 07 4.71E 08 0.00E 01 3.62E 07 0.00E 01 0.00E 01 0.00E 01 0.00E 01 Zn-65 2.83E 08 2.57E 08 1.75E 08 6.07E 08 3.89E 08 0.00E 01 0.00E 01 0.00E 01 Rb-86 1.19E 08 3.76E 07 0.00E 01 2.54E 08 0.00E 01 0.00E 01 0.00E 01 0.00E 01 Sr-89 4.40E 06 1.83E 07 1.54E 08 0.0E 01 0.00E 01 0.0E 01 0.0 % 01 0.0% 01 Sr-90 1.31E 09 1.49E 08 5.32E 09 0.00E 01 0.00E 01 0.00E 01 0.00E 01 0.0E 01 Y-91 1.52E 04 2.33E 08 5.68E 05 0.00E 01 0.00E 01 0.0E 01 0.00E 01 0.0 % 01 Zr-95 1.95E 05 6.53E 08 8.97E 05 2.83E 05 4.16E 05 0.00E 01 0.00E 01 0.0E 01 j Nb-95 3.29E 05 2.55E 09 1.08E 06 5.97E 05 5.79E 05 0.00E 01 0.0(E 01 0.00E 01 Ru-103 2.20E 07 4.30E 09 5.15E 07 0.00E 01 1.82E 08 0.0E 01 0.00E 01 0.00E 01 Ru-106 1.84E 08 7.00E 10 1.46E 09 0.00E 01 2.81E 09 0.0E 01 0.00E 01 0.0% 01 Ag-110M 1.86E 06 8.59E 08 3.23E 06 3.06E 06 5.83E 06 0.00E 01 0.00E 01 0.0E 01 Te-127M 8.25E 07 1.73E 09 6.94E 08 2.46E 08 2.81E 09 1.65E 08 0.0E 01 0.00E 01 l

TG-129M 9.81E 07 2.33E 09 6.20E 08 2.30E 08 2.59E 09 2.00E 08 0.00E 01 0.0E 01 I-131 4.40E 06 1.62E 06 5.85E C6 8.20E 06 1.41E 07 2.39E 09 0.00E 01 0.00E 01 I-133 1.23E-01 3.06E-01 2.39E-01 4.05E-01 7.10E-01 5.65E 01 0.00E 01 0.0E 01 I-135 4.88E-17 1.46E-16 5.11E-17 1.32E-16 2.08E-16 8.46E-15 0.00E 01 0.00E 01 Cs-134 3.4E 08 9.34E 06 3.19E 08 7.51E 08 2.39E 08 0.0E 01 9.11E 07 0.0% 01 Cs-136 1.55E 07 1.86E 06 5.87E 06 2.31E 07 1.26E 07 0.00E 01 1.98E 06 0.0% 01 Cs-137 2.14E 08 8.75E 06 4.62E 08 6.15E 08 2.09E 08 0.00E 01 8.13E 07 0.0E 01 Ba-140 9.76E 05 2.34E 07 1.51E 07 1.86E 04 6.29E 03 0.0E 01 1.25E 04 0.00E 01 Ce-141 5.42E 02 1.35E 07 7.07E 03 4.72E 03 2.22E 03 0.00E 01 0.00E 01 0.00E 01 Ce-144 3.96E 04 1.85E 08 7.37E 05 3.05E 05 1.82E 05 0.00E 01 0.00E 01 0.00E 01

• R Values in units of mrem /yr per micro-C1/m-3 for inhalation and tritium, and in units of M-Z mrem /yr per micro-C1/sec for all others.

l l TABLE 3.3-7 R Wu.ES FGt TE H. B. RSINSON STEAM ELECTRIC PLANT

• l l

l PAT}6&Y = Meat l AE Gt0VP = Child l

Nuclide T. Body GI-Tract Bone Liver Kidney _ Thyroid M Skin 11 - 3 2.36E 02 2.36E 02 0.00E 01 2.36E 02 2.36E 02 2.36E 02 2.36E 02 2.36E 02 l

P-32 1.87E 08 1.34E 08 4.86E 09 2.27E 08 0.0E 01 0.00E 01 0.00E 01 0.00E 01 Cr s1 5.33E 03 2.83E 05 0.00E 01 0.00E 01 8.09E 02 2.96E 03 5.40E 03 0.00E 01 Mn-54 1.3E 06 4.0E 06 0.00E 01 4.86E 06 1.36E 06 0.0E 01 0.00E 01 0.00E 01 Fe-59 1.82E 08 3.80E 08 2.25E 08 3.65E 08 0.00E 01 0.00E 01 1.06E 08 0.00E 01 Co-53 2.99E 07 5.70E 07 0.00E 01 9.76E 06 0.00E 01 0.0E 01 0.0 % 01 0.00E 01 Co-60 1.27E 08 2.3E 08 0.00E 01 4.30E 07 0.00E 01 0.00E 01 0.00E 01 0.0E 01 Zn-65 4.35E 08 1.23E 08 2.62E 08 6.99E 08 4.40E 08 0.00E 01 0.00E 01 0.00E 01 Rb-86 2.21E 08 2.32E 07 0.00E 01 3.60E 08 0.00E 01 0.00E 01 0.00E 01 0.00E 01 Sr-89 8.31E 06 1.13E 07 2.91E 08 0.00E 01 0.00E 01 0.0E 01 0.00E 01 0.0E 01 Sr-90 1.74E 09 9.26E 07 6.87E 09 0.00E 01 0.00E 01 0.00E 01 0.00E 01 0.00E 01 Y-91 ' 2.87E 04 1.43E 08 1.07E 06 0.00E 01 0.00E 01 0.00E 01 0.0E 01 0.0E 01 Zr-95 3.12E 05 3.65E 08 1.59E 06 3.50E 05 5.01E 05 0.00E 01 0.00E 01 0.00E 01 Nb-95 5.17E 05 1.34E 09 1.86E 06 7.23E 05 6.80E 05 0.00E 01 0.00E 01 0.0E 01 Ru-103 3.58E 07 2.41E 09 9.31E 07 0.00E 01 2.34E 08 0.00E 01 0.CE 01 0.0% 01 Ru-106 3.43E 08 4.27E 10 2.75E 09 0.0E 01 3.71E 09 0.0E 01 0.00E 01 0.00E 01 Ag-1104 2.89E 06 4.30E 08 5.36E 06 3.62E 06 6.74E 06 0.00E 01 0.00E 01 0.00E 01

(

Te-127M 1.55E 08 1.06E 09 1.31E 09 3.52E 08 3.73E 09 3.13E 08 0.00E 01 0.00E 01 Te-129M 1.81E 08 1.42E 09 1.17E 09 3.26E 08 3.43E 09 3.77E 08 0.00E 01 0.00E 01 I-131 6.20E 06 9.72E 05 1.09E 07 1.09E 07 1.79E 07 3.61E 09 0.0% 01 0.00E 01 I-133 2.07E-01 2.21E-01 4.43E-01 5.48E-01 9.13E-01 1.02E 07. 0.00E 01 0.00E 01

! I-135 7.87E-17 1.27E-16 9.25E-17 1.66E-16 2.55E-16 1.47E-14 0.0E 01 0.00E 01 l Cs-134 1.95E 08 4.93E 06 5.63E 08 9.23E 08 2.86E 08 0.00E 01 1.03E 08 0.0E 01 l Cs-136 1.80E 07 9.78E 05 1.01E 07 2.78E 07 1.48E 07 0.00E 01 2.21E 06 0.00E 01 Cs-137 1.20E 08 5.1E 06 8.51E 08 8.15E 08 2.65E 08 0.00E 01 9.55E 07 0.00E 01 Ba-140 1.63E 06 1.42E 07 2.80E 07 2.45E 04 7.97E 03 0.00E 01 1.46E 04 0.00E 01 Ce-141 9.86E 02 8.28E 06 1.33E 04 6.64E 03 2.91E 03 0.00E 01 0.00E 01 0.00E 01 Cs-144 7.42E 04 1.14E 08 1.39E 06 4.36E 05 2.41E 05 0.00E 01 0.00E 01 0.00E 01

• R Values in units of mrem /yr per micro-Ci/m-3 for inhalation and tritium, and in units of M-2 mrem /yr per micro-C1/sec for all others.

I TABLE 3.3-8 l R MLES FGt ilE H.B. RSINSON STEAM ELECTRIC PLANT

• l' PATHWAY = Cow Milk AE GROUP = Adult Nuclide T.8o@ GI-Tract Bone Liver Kidney Thyroid Lung, Skin H-3 7.69E 02 7.69E 02 0.00E-01 7.69E 02 7.69E 02 7.69E 02 7.69E 02 7.69E 02 P-32 4.32E 08 1.26E 09 1.12E 10 6.95E 08 0.00E-01 0.0E-01 0.00E-01 0.0E-01 Cr-51 1.73E 04 4.36E 06 0.00E-01 0.00E-01 3.82E 03 1.04E 04 2.30E 04 0.00E-01 Mn-54 9.76E 05 1.57E 07 0.00E-01 5.11E 06 1.52E 06 0.0E-01 0.00E-01 0.0E-01 Fe-59 1.60E 07 1.39E 08 1.77E 07 4.17E 07 0.00E-01 0.00E-01 1.17E 07 0.00E-01 Co-58 6.28E 06 5.6E 07 0.00E-01 2.8E 06 0.00E-01 0.0E-01 0.00E-01 0.00E-01 Co-60 2.24E 07 1.91E 08 0.00E-01 1.02E 07 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Zn-65 1.38E 09 1.92E 09 9.59E 08 3.05E 09 2.04E 09 0.0E-01 0.00E-01 0.00E-01 Rb-86 7.54E 08 3.19E 08 0.00E-01 1.62E 09 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Sr-89 2.50E 07 1.4 E 08 8.7E 08 0.00E-01 0.00E-01 0.0E-01 0.00E-01 0.00E-01 Sr-90 7.59E 09 8.94E 08 3.09E 10 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Y-91 1.37E 02 2.81E 06 5.11E 03 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Zr-% 1.22E 02 5.71E 05 5.62E 02 1.80E 02 2.83E 02 0.00E-01 0.00E-01 0.00E-01

(

Nb-95 1.48E 04 1.67E 08 4.95E 04 2.75E 04 2.72E 04 0.00E-01 0.00E-01 0.00E-01 Ru-103 2.63E 02 7.14E 04 6.11E 02 0.00E-01 2.33E 03 0.00E-01 0.00E-01 0.00E-01 Ru-106 1.60E 03 8.17E 05 1.26E 04 0.0E-01 2.44E 04 0.00E-01 0.00E-01 0.0E-01 Ag-110M 2.04E 07 1.40E 10 3.71E 07 3.44E 07 6.76E 07 0.00E-01 0.00E-01 0.00E-01

, Te-127M 4.11E 06 1.13E 08 3.37E 07 1.21E 07 1.37E 08 8.62E 06 0.00E-01 0.0E-01 Te-129M 6.19E 06 1.97E 08 3.91E 07 1.46E 07 1.63L 08 1.34E 07 0.00E-01 0.00E-01 i I-131 1.59E 08 7.32E 07 1.94E 08 2.77E 08 4.76E 08 9.09E 10 0.00E-01 0.0E-01 I-132 1.03E-01 5.51E-02 1.10E-01 2.93E-01 4.67E-01 1.03E 01 0.00E-01 0.00E-01 1-133 1.40E 06 4.13E 06 2.64E 06 4.59E 06 8.01E 06 6.75E 08 0.00E-01 0.00E-01 I-135 9.03E 03 2.76E 04 9.34E 03 2.45E 04 3.92E 04 1.61E 06 0.00E-01 0.00E-01 Cs-134 6.71E 09 1.44E 08 3.45E 09 3.21E 09 2.66E 09 0.0E-01 8.82E 08 0.00E-01 Cs-136 4.73E 08 7.46E 07 1.66E 08 6.5E 08 3.65E 08 0.00E-01 5.01E 07 0.00E-01 l Cs-137 4.22E 09 1.25E 08 4.71E 09 6.44E 09 2.19E 09 0.00E-01 7.27E 08 0.0d-01 Ba-140 1.12E 06 '3.53E 07 1.71E 07 2.15E 04 7.32E 03 0.00E-01 1.23E 04 0.00E-01 Ce-141 2.23E 02 7.52E 06 2.91E 03 1.97E 03 9.14E 02 0.00E-01 0.00E-01 0.0E-01 Ce-144 1.15E 04 7.26E 07 2.15E 05 8.97E 04 5.32E 04 0.00E-01 0.00E-01 0.00E-01 i

I

• R Values in units of mrem /yr per micro-Ci/m-3 for inhalation and tritiun, and in units of l

M-2 mrem /yr per micro-C1/sec for all others.

l i

3-38

TABLE 3.3-9 i R idWES FGt TIE H.B. R(BINSON STEAM ELECTRIC PLANT

• PATW AY = Cow Milk AE GROUP = Teen Nuclide T.8ody GI-Tract Bone Liver Kidney Thyroid tyng Skin H-3 1.00E 03 1.00E 03 0.00E-01 1.00E 03 1.00E 03 1.00E 03 1.00E 03 1.0E 03 P-32 8.00E 08 1.73E 09 2.06E 10 1.2F2 09 0.00E-01 0.0E-01 0.00E-01 0.00E-01 Cr-51 3.02E 04 5.08E 06 0.00E-01 0.00E-01 6.63E 03 1.68E 04 4.32E 04 0.00E-01 2-54 1.69E 06 1.75E 07 0.00E-01 8.52E 06 2.54E 06 0.00E-01 0.00E-01 0.00E-01 Fe-59 2.79E 07 1.71E 08 3.1E 07 7.23E 07 0.00E-01 0.00E-01 2.28E 07 0.00E-01 Co-58 1.09E 07 6.50E 07 0.00E-01 4.72E 06 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Co-60 3.88E 07 2.25E 08 0.00E-01 1.72E 07 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Zn-65 2.38E 09 2.16E 09 1.47E 09 5.11E 09 3.27E 09 0.0E-01 0.00E-01 0.00E-01 Rb-86 1.39E 09 4.37E 08 0.00E-01 2.95E 09 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Sr-89 4.59E 07 1.91E 08 1.60E 09 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Sr-90 1.08E 10 1.23E 09 4.37E 10 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.0E-01 Y-91 2.52E 02 3.85E 06 9.4E 03 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Zr-95 2.13E 02 7.16E 06 9.83E 02 3.10E 02 4.56E 02 0.00E-01 0.00E-01 0.00E-01 Nb-95 2.58E 04 2.0 E 08 8.45E 04 4.68E 04 4.54E 04 0.00E-01 0.0E-01 0.0E-01 Ru-103 4.65E 02 9.08E 04 1.09E 03 0.00E-01 3.83E 03 0.00E-01 0.00E-01 0.00E-01 Ru-106 2.9E 03 1.11E 06 2.32E 04 0.00E-01 4.48E 04 0.00E-01 0.00E-01 0.00E-01 Ag-110M 3.53E 07 1.63E 10 6.14E 07 5.81E 07 1.11E 08 0.00E-01 0.00E-01 0.00E-01 Te-127M 7.39E 06 1.55E 08 6.22E 07 2.21E 07 2.52E 08 1.48E 07 0.0E-01 0.00E-01 Te-129M 1.13E 07 2.69E 08 7.15E 07 2.65E 07 2.99E 08 2.31E 07 0.00E-01 0.00E-01 I-131 2.65E 08 9.75E 07 3.52E 08 4.93E 08 8.48E 08 1.44E 11 0.00E-01 0.00E-01 I-132 1.83E-01 2.22E-01 1.94E-01 5.09E-01 8.02E-01 1.71E 01 0.00E-91 0.00E-01 I-133 2.49E 06 6.19E 06 4.82E 06 8.18E 06 1.43E 07 1.14E 09 0.00E-01 0.00E-01 i 1-135 1.5E 04 4.74E 04 1.66E 04 4.27E 04 6.75E 04 2.75E 06 0.00E-01 0.00E-01 Cs-134 6.54E 09 1.75E 08 5.99E 09 1.41E 10 4.48E 09 0.00E-01 1.71E 09 0.00E-01 Cs-136 7.48E 08 8.97E 07 2.83E 08 1.11E 09 6.07E 08 0.00E-01 9.56E 07 0.00E-01 Cs-137 3.96E 09 1.62E 08 8.54E 09 1.14E 10 3.87E 09 0.00E-01 1.50E 09 0.00E-01 Ba-140 1.99E 06 4.77E 07 3.09E 07 3.79E 04 1.28E 04 0.00E-01 2.55E 04 0.00E-01 Ce-141 4.09E 02 1.02E 07 6.33E 03 3.56E 03 1.68E 03 0.00E-01 0.00E-01 0.00E-01 Ce-144 2.12E 04 9.93E 07 3.95E 05 1.63E 05 9.76E 04 0.00E-01 0.00E-01 0.00E-01 l

• R Values in units of mrem /yr per micro-Ci/m-3 for inhalation and trititsn, and in units of M~2 mrem /yr per micro-Ci/sec for all others.

3-39

TABLE 3,3-10 R W.ES FGt TE H.B. REINSON STEAM ELECTRIC PLANT

• I PATHWAY = Cow Milk AGE Gt0VP = Child Ibclide T.Bo@ GI-Tract Bone Liver Kichey Thyroid Luna 3,in H-3 1.58E 03 1.58E 03 0.00E-01 1.58E 03 1.58E 03 1.58E 03 1.58E 03 1.58E 03 P-32 1.96E 09 1.41E 09 5.09E 10 2.38E 09 0.0E-01 0.0E-01 0.0E-01 0.00E-01 Cr-51 6.17E 04 3.27E 06 0.00E-01 0.00E-01 9.36E 03 3.42E 04 6.25E 04 ~0.00E-01 Mn-54 3.39E 06 1.07E 07 0.0E-01 1.27E 07 3.57E 06 0.0E-01 0.00E-01 0.0E-01 l Fe-59 5.79E 07 1.21E 08 7.18E 07 1.16E 08 0.00E-01 0.00E-01 3.37E 07 0.00E-01 Co-58 2.21E 07 4.20E 07 0.00E-01 7.21E 06 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Co-60 7.90E 07 1.48E 08 0.00E-01 2.68E G7 0.00E-01 0.00E-01 0.00E-01 0.00E-01 l Zn-65 4.79E 09 1.35E 09 2.89E 09 7.7E 09 4.85E.09 0.00E-01 0.00E-01 0.00E-01 Rb-86 3.36E 09 3.52E 08 0.00E-01 5.47E 09 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Sr-89 1.13E 08 1.54E 08 3.97E 09 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Sr-90 1.87E 10 9.95E 08 7.38E 10 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Y-91 6.21E 02 3.09E 06 2.52E 04 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.0E-01 Zr-95 4.47E 02 5.23E 05 2.28E 03 5.02E 02 7.18E 02 0.00E-01 0.00E-01 0.00E-01 ft-95 5.31E 04 1.37E 08 1.91E 05 7.42E 04 G.98E 04 0.00E-01 0.00E-01 0.00E-01 Ru-103 9.88E 02 6.65E 04 2.57E 03 0.00E-01 6.47E 03 0.00E-01 0.00E-01 0.00E-01 Ru-106 7.14E 03 8.90E 05 5.72E 04 0.0E-01 7.72E 04 0.00E-01 0.00E-01 0.00E-01 Ag-110M 7.19E 07 1.07E 10 1.33E 08 9.00E 07 1.68E 08 0.00E-01 0.00E-01 0.00E-01 l Te-127M 1.82E 07 1.24E 08 1.53E 08 4.13E 07 4.37E 08 3.66E 07 0.00E-01 0.00E-01 Te-129M 2.74E 07 2.15E 08 1.76E 08 4.92E 07 5.18E 08 5.68E 07 0.00E-01 0.00E-01 I-131 4.88E 08 7.64E 07 8.54E 08 8.59E 08 1.41E 09 2.84E 11 0.00E-01 0.0E-01 I-132 3.89E-01 9.95E-01 4.60E-01 8.45E-01 1.29E 00 3.92E 01 0.00E-01 0.00E-01 i I-133 5.48E 06 5.84E 06 1.17E 07 1.45E 07 2.41E 07 2.69E 09 0.00E-01 0.00E-01 I-135 3.35E 04 5.39E 04 3.93E 04 7.07E 04 1.08E 05 6.26E 06 0.00E-01 0.00E-01 Cs-134 4.78E 09 1.22E 08 1.38E 10 2.27E 10 7.03E 09 0.00E-01 2.52E 09 0.00E-01 Cs-136 1.14E 09 6.17E 07 6.39E 08 1.76E 09 9.36E 08 0.00E-01 1.40E 08 0.00E-01 Cs-137 2.91E 09 1.23E 08 2.06E 10 1.97E 10 6.42E 09 0.00E-01 2.31E 09 0.00E-01 Ba-140 4.36E 06 3.7EE 07 7.47E 07 6.54E 04 2.13E 04 0.00E-01 3.90E 04 0.00E-01 l Ce-141 9.73E 02 8.17E 06 1.31E 04 6.55E 03 2.87E 03 0.00E-01 0.00E-01 0.00E-01 Ce-144 5.20E 04 7.96E 07 9.74E 05 3.05E 05 1.69E 05 0.00E-01 0.00E-01 0.00E-01 l

l l

• R Values in units of mrem /yr per micro-C1/m-3 for inhalation and tritium, and in units of i

M-2 mrem /yr per micro-C1/sec for all others.

1

, 3-40 l

l t

f TABLE 3.3-11 R Wu.ES F01 TIE H.B. REINSON STEAM ElICTRIC PLANT

• r PATHWAY = Cow Milk AE GROUP = Infant Mac11de T.Bo@ GI-Tract Bone Liver Kidney Thyroid M Skin H-3 2.40E 03 2.40E 03 0.00E-01 2.4E 03 2.4E 03 2.4E 03 2.40E 03 2.40E 03 P-32 4.06E 09 1.42E 09 1.05E 11 6.17E 09 0.0E-01 0.00E-01 0.00E-01 0.0E-01 Cr-51 9.77E 04 2.85E 06 0.00E-01 0.00E-01 1.39E 04 6.38E 04 1.24E 05 0.00E-01 Mn-54 5.37E 06 8.71E 06 0.0E-01 2.37E 07 5.25E 06 0.00E-01 0.00E-01 0.0E-01 Fe-59 9.23E 07 1.12E 08 1.34E 08 2.34E 08 0.00E-01 0.00E-01 6.92E 07 0.00E-01 Co-58 3.60E 07 3.59E 07 0.00E-01 1.44E 07 0.0E-01 0.0E-01 0.00E-01 0.0E-01 Co-60 1.29E 08 1.30E 08 0.00E-01 5.47E 07 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Zn-65 6.14E 09 1.12E 10 3.88E 09 1.33E 10 6.45E 09 0.0E-01 0.00E-01 0.00E-01 Rb-86 6.8cE 09 3.55E 08 0.00E-01 1.39E 10 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Sr-89 2.17E 08 1.55E 08 7.55E 09 0.00E-01 0.0E-01 0.0E-01 0.00E-01 0.0E-01 Sr-90 2.05E 10 1.00E 09 8.04E 10 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Y-91 1.16E 08 3.12E 06 4.36E 04 0.00E-01 0.00E-01 0.0E-01 0.0E-01 0.0E-01 Zr-95 7.01E 02 4.92E 05 4.05E 03 9.88E 02 1.06E 03 0.00E-01 0.00E-01 0.0E-01 Hb-95 8.48E 04 1.24E 08 3.56E 05 1.47E 05 1.05E 05 0.00E-01 0.00E-01 0.0E-01 Ru-103 1.74E 03 6.33E 04 5.21E 03 0.00E-01 1.08E 04 0.00E-01 0.00E-01 0.00E-01 Ru-106 1.47E 04 8.95E 05 1.18E 05 0.00E-01 1.39E 05 0.0E-01 0.0E-01 0.00E-01 Ag-110M' 1.19E 08 9.32E 09 2.46E 08 1.80E 08 2.57E 08 0.00E-01 0.00E-01 0.00E-01 Te-127M 3.75E 07 1.25E 08 3.1E 08 1.03E 08 7.64E 08 8.96E 07 0.00E-01 0.0E-01 Te-129M 5.57E 07 2.16E 08 3.62E 08 1.24E 08 9.05E 08 1.39E 08 0.00E-01 0.00E-01 l

I-131 9.23E 08 7.49E 07 1.78E 09 2.10E 09 2.45E 09 6.90E 11 0.00E-01 0.00E-01 1-132 6.90E-01 1.57E-00 9.55E-01 1.94E 00 2.16E 00 9.09E 01 0.00E-01 0.00E-01 1-133 1.05E 07 6.09E 06 2.47E 07 3.60E 07 4.23E 07 6.55E 09 0.00E-01 0.0E-01 I-135 5.93E 04 5.83E 04 8.17E 04 1.63E 05 1.81E 05 1.46E 07 0.00E-01 0.00E-01 Cs-134 4.19E 09 1.13E 08 2.23E 10 4.15E 10 1.07E 10 0.0E-01 4.38E 09 0.00E-01 Cs-136 1.37E 09 5.58E 07 1.25E 09 3.67E 09 1.46E 09 0.00E-01 2.99E 08 0.00E-01 Cs-137 2.72E 09 1.20E 08 3.28E 10 3.84E 10 1.03E 10 0.00E-01 4.18E 09 0.00E-01 Ba-140 7.91E 06 3.77E 07 1.54E 08 1.54E 05 3.65E 04 0.0CE-01 9.43E 04 0.00E-01 Ce-141 1.87E 03 3.21E 06 2.60E 04 1.59E 04 4.90E 03 0.00E-01 0.00E-01 0.00E-01 Ce-144 7.82E 04 8.01E 07 1.40E 06 5.71E 05 2.31E 05 0.00E-01 0.00E-01 0.00E-01

• R Values in units of mrem /yr per micro-Ci/m-3 for inhalation and tritium, and in units of M-2 mrem /yr per micro-Cf/sec for all others.

3-41

TNILE 3.3-12 R MLES FGt TE H.B. RGlINSON STEAM ELECTRIC PUUR*

PATE RY = Goat Milk AE 9t0VP = Adult Ibclide T.Bo@ GI-Tract Bone Liver Kidney Thyroid M Skin H-3 1.57E 03 1.57E 03 0.00E-01 1.57E 03 1.57E 03 1.57E 03 1.57E 03 1.57E 03 P-32 5.19E 08 1.51E 09 1.34E 10 8.34E 08 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Cr-51 2.0E 03 5.23E 05 0.00E-01 0.00E-01 4.58E 02- 1.24E 03 2.76E 03 0.00E-01 Mn-54 1.17E 05 1.8E 06 0.00E-01 6.14E 05 1.83E 05 0.00E-01 0.00E-01 0.00E-01 Fe-59 2.08E 05 1.81E 06 2.31E 05 5.42E 05 0.00E-01 0.00E-01 1.51E 05 0.00E-01

,Co-58 7.54E 05 6.82E 06 0.00E-01 3.36E 05 0.00E-01 0.00E-01 0.0E-01 0.0E-01 Co-60 2.69E 06 2.29E 07 0.00E-01 1.22E 06 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Zn-65 1.65E 08 2.31E 08 1.15E 08 3.66E 08 2.45E 08 0.0E-01 0.00E-01 0.00E-01 Rb-06 9.05E 07 3.83E 07 0.00E-01 1.94E 08 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Sr-89 5.24E 07 2.93E 08 1.83E 09 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Sr-90 1.59E 10 1.88E 09 6.49E 10 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Y-91 1.64E 01 3.37E 05 6.13E 02 0.0E-01 0.0E-01 0.00E-01 0.00E-01 0.00E-01 Zr-95 1.46E 01 6.85E 04 6.74E 01 2.12 01 3.39E 01 0.00E-01 0.00E-01 0.00E-01 16-95 1.78E 03 2.01E 07 5.94E 03 3.31E 03 3.27E 03 0.00E-01 0.00E-01 0.00E-01 l

l Ru-103 3.16E 01 8.56E 03 7.33E 01 0.00E-01 2.80E 02 0.00E-01 0.00E-01 0.00E-01 1

Ru-106 1.92E 02 9.81E 04 1.52E 03 0.0E-01 2.93E 03 0.00E-01 0.00E-01 0.0E-01 Ag-110M 2.45E 06 1.68E 09 4.46E 06 4.12E 06 8.11E 06 0.00E-01 0.00E-01 0.00E-01 Te-127M 4.93E 05 1.36E 07 4.05E 06 1.45E 06 1.64E 07 1.03E 06 0.0E-01 0.0E-01 Te-129M 7.43E 05 2.36E 07 4.69E 06 1.75E 06 1.96E 07 1.61E 06 0.00E-01 0.00E-01 I-131 1.91E 08 8.7E 07 2.33E 08 3.33E 08 5.71E 08 1.09E 11 0.0E-01 0.0E-01 I-132 1.23E-01 6.61E-02 1.32E-01 3.52E-01 5.61E-01 1.23E 01 0.00E-01 0.00E-01 I-133 1.6E 06 4.95E 06 3.17E 06 5.51E 06 9.61E 06 8.10E 08 0.0E-01 0.00E-01 I-135 1.08E 04 3.32E 04 1.12E 04 2.94E 04 4.71E 04 1.94E 06 0.00E-01 0.00E-01 l Cs-134 2.01E 10 4.31E 08 1.03E 10 2.46E 10 7.97E 09 0.00E-01 2.65E 09 0.0E-01 Cs-136 1.42E 09 2.24E 08 4.99E 08 1.97E 09 1.1T 09 0.00E-01 1.50E 08 0.00E-01 Cs-137 1.27E 10 3.74E 08 1.41E 10 1.93E 10 6.56E 09 0.00E-01 2.18E 09 0.00E-01 Ba-140 1.35E 05 4.23E 06 2.06E 06 2.58E 03 8.78E 02 0.00E-01 1.48E 03 0.0E-01 Ce-141 2.68E 01 9.03E 05 3.49E 02 2.36E 02 1.10E 02 0.00E-01 0.00E-01 0.00E-01 Ce-144 1.38E 03 8.71E 06 2.5E 04 1.08E 04 6.39E 03 0.00E-01 0.00E-01 0.00E-01

• R Values in units of mrem /yr per micro-Ci/m-3 for inhalation and tritium, and in units of M-2 ntrem/yr per micro-C1/sec for all others.

3-42

TABLE 3.3-13 i R MLES F(R TE H.B. RIBINSON STEAM ELECTRIC PLANT

• PATERY = Goat Milk AGE Gt00P = Teen hclide T.Bo@ GI-Tract Bone Liver Kidney Thyroid M Skin H-3 2.04E 03 2.04E 03 0.00E-01 2.04E 03 2.04E 03 2.04E 03 2.04E 03 2.04E 03 P-32 9.60E 08 2.08E 09 2.48E 10 1.53E 09 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Cr-51 3.63E 03 6.10E 05 0.90E-01 0.00E-01 7.95E 02 2.02E 03 5.18E 03 0.00E-01 Mn-54 2.03E 05 2.10E 06 0.01-01 1.02E 06 3.05E 05 0.00E-01 0.00E-01 0.00E-01 Fe-59 3.63E 05 2.22E 06 4.03E 05 9.40E 05 0.00E-01 0.00E-01 2.96E 05 0.00E-01 Co-58 1.30E 06 7.80E 06 0.00E-01 5.66E 05 0.00E-01 0.0E-01 0.00E-01 0.0E-01 Co-60 4.66E 06 2.69E 07 0.00E-01 2.07E 06 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Zn-65 2.86E 08 2.6E 08 1.77E 08 6.13E 08 3.93E 08 0.00E-01 0.00E-01 0.00E-01 Rb-86 1.66E 08 5.24E 07 0.00E-01 3.54E 08 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Sr-89 9.65E 07 4.01E 08 3.37E 09 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Sr-90 2.27E 10 2.58E 09 9.18E 10 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Y-91 3.02E 01 4.62E 05 1.13E 03 0.00E-01 0.00E-01 0.00E-01 0.0E-01 0.00E-01 Zr-95 2.56E 01 8.59E 04 1.18E 02 3.72E 01 5.47E 01 0.00E-01 0.00E-01 0.00E-01 It-95 3.09E 03 2.40E 07 1.01E 04 5.62E 03 5.45E 03 0.00E-01 0.00E-01 0.00E-01 Ru-103 5.58E 01 1.09E 04 1.30E 02 0.00E-01 4.60E 02 0.00E-01 0.00E-01 0.00E-01 Ru-106 3.51E 02 1.34E 05 2.79E 03 0.00E-01 5.38E 03 0.00E-01 0.00E-01 0.0E-01 Ag-11(N . 4.24E 06 1.96E 09 7.37E 06 6.97E 06 1.33E 07 0.00E-01 0.00E-01 0.00E-01 Te-127M 8.87E 05 1.86E 07 7.46E 06 2.65E 06 3.02E 07 1.77E 06 0.00E-01 0.00E-01 Te-129M 1.36E 06 3.22E 07 8.58E 06 3.19E 06 3.59E 07 2.77E 06 0.00E-01 0.00E-01 1-131 3.18E 08 1.17E 08 4.22E 08 5.91E 08 1.02E 09 1.73E 11 0.00E-01 0.0E-01 1-132 2.19E-01 2.66E-01 2.33E-01 6.11E-01 9.62E-01 2.06E 01 0.00E-01 0.00E-01 1-133 2.99E 06 7.43E OG .5.79E 06 9.81E 06 1.72E 07 1.37E 09 0.00E-01 0.00E-01 I-135 1.90E 04 5.63E 04 1.99E 04 5.13E 04 8.10E 04 3.30E 06 0.00E-01 0.00E-01 Cs-134 1.96E 10 5.26E 08 1.80E 10 4.23E 10 1.34E 10 0.00E-01 5.13E 09 0.00E-01 Cs-136 2.25E 09 2.69E 07 8.50C 08 3.34E 09 1.82E 09 0.00E-01 2.87E 08 0.00E-01 l

Cs-137 1.19E 10 4.85E 08 2.56E 10 3.41E 10 1.16E 10 0.00E-01 4.51E 09 0.00E-01 Ba-140 2.39E 05 5.72E 06 3.71E 06 4.55E 03 1.54E 03 0.00E-01 3.06E 03 0.00E-01 Ce-141 4.91E 01 1.22E 06 6.4E 02 4.27E 02 2.01E 02 0.0E-01 0.00E-01 0.00E-01 Ce-144 2.55E 03 1.19E 07 4.74E 04 1.96E 04 1.17E 04 0.00E-01 0.00E-01 0.00E-01 l

t 1 Values in units of mrem /yr per micro-C1/m-3 for inhalation and tritium, and in units of M-2 mrem /,yr per micro-C1/see for all others.

3-43

TABLE 3.3-14 R ELES F(R TE HJ. R(BINSON STEAM ELEClRIC PLANT

• PATMdAY = Goat Milk AGE Gt00P = Child Ibclide, T.80@ GI-Tract Bone Liver Kidney Thyroid Lung Skin H-3 3.23E 03 3.23E 03 0.00E-01 3.23E 03 3.23E 03 3.23E 03 3.23E 03 3.23E 03 P-32 2.35E 09 1.69E 09 6.11E 10 2.E6E 09 0.00E-01 0.00E-01 0.00E-01 0.0E-01 Cr-51 7.40E 03 3.93E 05 0.00E-01 0.00E-01 1.12E 03 4.11E 03 7.50E 03 0.00E-01 Mn-54 4.07E 05 1.28E 06 0.00E-01 1.53E 06 4.29E 05 0.00E-01 0.00E-01 0.0E-01 Fe 7.52E 05 1.57E 06 9.34E 05 1.51E 06 0.00E-01 0.00E-01 4.38E 05 0.00E-01 Co-58 2.65E 06 5.05E 06 0.00E-01 8.65E 05 0.00E-01 0.0E-01 0.00E-01 0.00E-01 l Co-60 9.48E 06 1.78E 07 0.00E-01 3.21E 06 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Zn-65 5.74E 08 1.62E 08 3.47E 08 9.24E 08 5.82E 08 0.00E-01 0.00E-01 0.00E-01 Rb-86 4.04E 08 4.22E 07 0.00E-01 6.57E 08 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Sr-89 2.38E 08 3.23E 08 8.34E 09 0.0E-01 0.00E-01 0.0E-01 0.00E-01 0.00E-01 Sr-90 3.93E 10 2.09E 09 1.55E 11 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Y-91 7.45E 01 3.71E 05 2.79E 03 0.0E-01 0.00E-01 0.0E-01 0.0E-01 0.00E-01 Zr-95 5.36E 01 6.28E 04 2.74E 02 6.02E 01 8.62E 01 0.00E-01 0.00E-01 0.00E-01 Nb-95 6.37E 03 1.65E 07 2.29E 04 8.91E 03 8.37E 03 0.0E-01 0.00E-01 0.0E-01 Ru-103 1.19E 02 7.98E 03 3.09E 02 0.00E-01 7.77E 02 0.00E-31 0.00E-01 0.00E-01 Ru-106 8.5EE 02 1.07E 05 6.86E 03 0.00E-01 9.27E 03 0.0E-01 0.00E-01 0.00E-01

. Ag-110M 8.63E 06 1.28E 09 1.60E 07 1.08E 07 2.01E 07 0.00E-01 0.00E-01 0.00E-01 l Te-127M 2.1E 06 1.49E 07 1.84E 07 4.95E 06 5.24E 07 4.40E 06 0.00E-01 0.00E-01 Te-129M 3.28E 06 2.58E 07 2.12E 07 5.91E 06 6.21E 07 6.82E 06 0.00E-01 0.00E-01 1-131 5.85E 08 9.17E 07 1.02E 09 1.03E 09 1.69E 09 3.41E 11 0.00E-01 0.0E-01 1-132 4.67E-01 1.19E 00 5.52E-01 1.01E 00 1.55E 00 4.71E 01 0.00E-01 0.00E-01 1-133 6.58E 06 7.00E 06 1.41E 07 1.74E 07 2.9% 07 3.232 09 0.00E-01 0.00E-01 I-135 4.01E 04 6.47E 04 4.72E 04 4.49E 04 1.30E 05 7.52E 06 0.00E-01 0.00E-01 Cs-134 1.43E 10 3.67E 08 4.14E 10 6.80E 10 2.11E 10 0.00E-01 7.56E 09 0.0E-01 Cs-136 3.41E 09 1.8SE 06 1.92E 09 5.27E 09 2.81E 09 0.00E-01 4.19E 08 0.00E-01 Cs-137 8.72E 09 3.70E 08 6.17E 10 5.91E 10 1.93E 10 0.0E-01 6.93E 09 0.0E-01 Ba-140 5.23E 05 4.54E 05 8.96E 06 7.85E 03 2.56E 03 0.00E-01 4.68E 03 0.00E-01 Ce-141 1.17E 02 9.81E 05 1.53E 03 7.36E 02 3.45E 02 0.0E-01 0.00E-01 0.0E-01 Ce-144 6.24E 03 9.55E 06 1.17E 05 3.66E 04 2.03E 04 0.00E-01 0.00E-01 0.00E-01 l

• R Values in units of mrem /yr per micro-Ci/m-3 for innalation and tritiisn, and in units of l

M-2 mrem /yr per micro-Ci/sec for all others.

3-44

TABLE 3.3-15 R 1RLES FGt TE H.B. RGlINSON STEAM ELECTRIC PLANT *

/

PATM AY = Goat Milk AE GROUP = Infant hclide T. Body GI-Tract Sone Liver K1 4ey Thyroid gng Skin H-3 4.90E 03 4.90E 03 0.00E-01 4.90E 03 4.90E 03 4.90E 03 4.90E 03 4.90E 03

) P-32 4.88E 09 1.7E 09 1.26E 11 7.40E 09 0.00E-01 0.00E-01 0.00E-01 0.0E-01 Cr-51 1.17E 04 3.42E 05 0.00E-01 0.00E-01 1.67E 03 7.65E 03 1.49E 04 0.00E-01 2-54 6.45E 05 1.04E 06 0.00E-01 2.84E 06 6.30E 05 0.00E-01 0.00E-01 0.00E-01 Fe-59 1.20E 06 1.45E 06 1.74E 06 3.04E 06 0.00E-01 0.00E-01 9.00E 05 0.00E-01 Co-58 4.31E 06 4.31E 06 0.00E-01 1.73E 06 0.0E-01 0.00E-01 0.00E-01 0.00E-01 Co-60 1.55E 07 1.56E 07 0.00E-01 6.56E 06 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Zn-65 7.36E 08 1.35E 09 4.66E 08 1.6E 09 7.74E 08 0.00E-01 0.00E-01 0.00E-01 Rb-06 8.23E 08 4.26E 07 0.00E-01 1.67E 09 0.00E-01 0.00E-01 0.00E-01 0.00E-01 I

Sr-89 4.55E 08 3.26E 08 1.59E 10 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.0E-01 Sr-90 4.30E 10 2.11E 09 1.69E 11 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Y-91 1.39E 02 3.75E 05 5.23E 03 0.00E-01 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Zr-95 8.41E 01 5.90E 04 4.85E 02 1.19E 02 1.28E 02 0.0E-01 0.00E-01 0.00E-01 2-95 1.02E 04 1.48E 07 4.27E 04 1.76E 04 1.26E"04 0.0E-01 0.0E-01 0.00E-01 Ru-103 2.09E 02 7.60E 03 6.25E 02 0.00E-01 1.30E 03 0.00E-01 0.00E-01 0.00E-01 Ru-106 1.77E 03 1.07E 05 1.41E 04 0.00E-01 1.67E 04 0.00E-01 0.00E-01 0.00E-01 l Ag-110M 1.43E 07 1.12E 09 2.95E 07 2.16E 07 3.08E 07 0.00E-01 0.00E-01 0.00E-01 l Te-127M 4.51E 06 1.5E 07 3.72E 07 1.23E 07 9.16E 07 1.08E 07 0.00E-01 0.0E-01 Te-129M 6.69E 06 2.59E 07 4.34E 07 1.49E 07 1.09E 08 1.67E 07 0.00E-01 0.00E-01 I-131 1.11E 09 8.99E 07 2.14E 09 2.52E 09 2.94E 09 8.28E 11 0.00E-01 0.00E-01 l I-132 8.28E-01 1.88E 00 1.15E 00 2.33E 00 2.59E 00 1.09E 02 0.00E-01 0.0E-01 l

I-133 1.27E 07 7.31E 06 2.97E 07 4.32E 07 5.08E 07 7.86E 09 0.00E-01 0.0E-01 I-135 7.11E 04 7.06E 04 9.81E 04 1.95E 05 2.17E 05 1.75E 07 0.00E-01 0.00E-01 Cs-134 1.26E 10 3.38E 08 6.68E 10 1.25E 11 3.21E 10 0.00E-01 1.31E 10 0.00E-01 Cs-136 4.11E 09 1.67E 08 3.75E 09 1.1E 10 4.39E 09 0.00E-01 8.98E 08 0.00E-01 Cs-137 8.17E 09 3.61E 08 9.85E 10 1.15E 11 3.10E 10 0.00E-01 1.25E 10 0.00E-01

Ba-140 9.5E 05 4.53E 06 1.84E 07 1.84E 04 4.38E 03 0.00E-01 1.13E 04 0.00E-01 Ce-141 2.24E 02 9.85E 05 3.13E 03 1.91E 03 5.88E 02 0.00E-01 0.0E-01 0.0E-01 Ce-144 9.39E 03 9.61E 06 1.67E 05 6.86E 04 2.77E 04 0.00E-01 0.00E-01 0.00E-01 i

1 r

• R Values in units of mrem /yr per micro-Ci/m-3 for inhalation and tritium, and in units of j M-2 mres/yr per micro-Ci/sec for all others. '

3-45

TABLE 3.3-16 ,

R MLES RR TE H.B. NBINSON STEAM El.ECTRIC PUUK* l e '

PAT)WAY = Inhal AGE Gt00P = Adult Nuclide T.Bo@ GI-Tract Bone Liver Kidw Thyrofd M Skin H-3 1.2E 03 1.2E 03 0.00E-01 1.2E 03 1.2E 03 1.2E 03 1.2E 03 1.2E 03 P-32 5.00E 04 8.63E 04 1.32E 06 7.70E 04 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Cr-51 9.99E 01 3.32E 03 0.00E-01 0.00E-01 2.28E 01 5.94E 01 1.44E 04 0.00E-01 Mn-54 6.29E 03 7.72E 04 0.00E-01 3.95E 04 9.83E 03 0.00E-01 1.40E 06 0.00E-01 Fe-59 1.05E 04 1.88E 05 1.1E 04 2.7K 04 0.00E-01 0.00E-01 1.01E 06 0.00E-01 Co-58 2.0E 03 1.0E 05 0.00E-01 1.58E 03 0.00E-01 0.00E-01 9.27E 05 0.00E-01 Co-60 1.4E 04 2.84E 05 0.00E-01 1.15E 04 0.00E-01 0.00E-01 5.9E 06 0.00E-01 Zn-65 4.65E 04 5.34E 04 3.24E 04 1.03E 05 6.89E 04 C.00E-01 8.63E 05 0.00E-01 Rb-86 5.89E 04 1.6E 04 0.00E-01 1.35E 05 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Sr-89 8.71E 03 3.49E 05 3.04E 05 0.00E-01 0.00E-01 0.00E-01 1.40E 06 0.00E-01 Sr-90 6.09E 06 7.21E 05 9.91E 07 0.00E-01 0.00E-01 0.00E-01 9.59E 06 0.00E-01 Y-91 1.24E 04 3.84E 05 4.62E 05 0.00E-01 0.00E-01 0.00E-01 1.70E 06 0.00E-01 Zr-95 2.32E 04 1.50E 05 1.0E 05 3.44E 04 5.41E 04 0.00E-01 1.77E 06 0.00E-01 b 95 4.20E 03 1.04E 05 1.41E 04 7.80E 03 7.72E 03 0.00E-01 5.04E 05 0.0CE-01 Ru-103 6.5E 02 1.1 E 05 1.51 03 0.00E-01 5.82E 03 0.00E-01 5.04E 05 0.00E-01 Ru-106 8.71E 03 9.11E 05 6.90E 04 0.00E-01 1.33E 05 0.00E-01 9.35E 06 0.00E-01 Ag-1101 5.94E 03 3.02E 05 1.0E 04 9.99E 03 1.9E 04 0.0E-01 4.63E 06 0.00E-01 Te-127M 1.57E 03 1.49E 05 1.2E 04 5.7E 03 4.5E 04 3.28E 03 9.59E 05 0.00E-01 Te-129M 1.58E 03 3.82 05 9.75E 03 4.67E 03 3.65E 04 3.44E 03 1.1E 06 0.00E-01 I-131 2.05E 04 6.27E 03 2.52E 04 3.57E 04 6.12E 04 1.19E 07 0.00E-01 0.00E-01 I-132 1.1E 03 4.0E 02 1.1E 03 3.25E 03 5.18E 03 1.14E 05 0.00E-01 0.0E-01 I-133 4.51E 03 8.87E 03 3.63E 03 1.48E 04 2.58E 04 2.15E 06 0.00E-01 0.0E-01 l I-135 2.5E 03 5.24E 03 2.6E 03 6.97E 03 1.11E 04 4.4 E 05 0.00E-01 0.00E-01 Cs-134 7.2K 05 1.04E 04 3.72E 05 8.47E 05 2.87E 05 0.00E-01 9.75E 04 0.00E-01 i Cs-136 1.1E 05 1.17E 04 3.90E 04 1.4E 05 8.5E 04 0.00E-01 1.2 2 04 0.00E-01 Cs-137 4.27E 05 8.39E 03 4.78E 05 6.20E 05 2.22E 05 0.00E-01 7.51E 04 0.00E-01 Ba-140 2.5E 03 2.1E 05 3.90E 04 4.90E 01 1.67E 01 0.00E-01 1.2K 06 0.00E-01 Ce-141 1.53E 03 1.20E 05 1.99E 04 1.35E 04 6.25E 03 0.00E-01 3.61E 05 0.00E-01

)

Ce-144 1.84E 05 8.15E 05 3.41 06 1.4I 06 8.4 E 05 0.00E-01 7.7E 06 0.0E-01 i

• R Values in units of mrem /yr per micro-Cf/m-3 for inhalation and trittun, and in units of M-2 mrem /yr per micro-C1/sec for all others.

3-46 boi

TABLE 3.3-17 R IN.ES FGt TE H.B. IEBINSON STEAM ELECTRIC R. ANT

• PATHWAY = Inhal AE GROUP = Teen Nuclide T.Bostr GI-Tract Bone Liver Kidney 1hyroid g g H-3 1.2E 03 1.2E 03 0.00E-01 1.27E 03 1.27E 03 1.2 E 03 1.2 K 03 1.2 E 03 P-32 7.15E 04 9.2K 04 1.89E 06 1.09E 05 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Cr-51 1.3E 02 3.0E 03 0.00E-01 0.00E-01 3.0X 01 7.49E 01 2.09E 04 0.00E-01 Hn-54 8.39E 03 6.67E 04 0.00E-01 5.1E 04 1.27E 04 0.00E-01 1.96E 06 0.00E-01 Fe-59 1.41 04 1.7E 03 1.59E 04 3.69E 04 0.00E-01 0.00E-01 1.5I 06 0.00E-01 Cc-58 2.77E 03 9.51E 04 0.00E-01 2.0K 03 0.00E-01 0.0E-01 1.34E 06 0.00E-01 Co-@ 1.90E 04 2.59E 05 0.00E-01 1.51E 04 0.00E-01 0.00E-01 8.71E 06 0.00E-01 Zn-65 6.23E 04 4.6E 04 3.85E 04 1.33E 05 8.63E 04 0.00E-01 1.24E 06 0.00E-01 Rb-86 8.39E 04 1.7E 04 0.00E-01 1.9E 05 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Sr-89 1.25E 04 3.71E 05 4.34E 05 0.00E-01 0.00E-01 0.00E-01 2.41E 06 0.00E-01 Sr-90 6.67E 06 7.64E 05 1.0E 08 0.00E-01 0.00E-01 0.00E-01 1.6E 07 0.00E-01 Y-91 1.77E 04 4.00E 05 6.EE 05 0.00E-01 0.00E-01 0.00E-01 2.93E 06 0.00E-01 Zr-95 3.15E 04 1.49E 05 1.4E 05 4.58E 04 6.71 04 0.00E-01 2.68E 06 0.00E-01 2-95 5.6E 03 9.67E 04 1.85E 04 1.03E 04 9.99E 03 0.00E-01 7.50E 05 0.00E-01 j

Ru-103 8.95E 02 1.09E 05 2.1 E 03 0.00E-01 7.42E 03 0.00E-01 7.82E 05 0.00E-01 Ru-106 1.24E 04 9.59E 05 9.83E 04 0.00E-01 1.90E 05 0.00E-01 1.61E 07 0.00E-01 Ag-1101 7.9E 03 2.72E 05 1.33E 04 1.31E 04 2.50E 04 0.00E-01 6.74E 06 0.00E-01 Te-127M 2.1E 03 1.59E 05 1.80E 04 8.15E 03 6.53E 04 4.38E 03 1.65E 06 0.00E-01 Te-121M 2.24E 03 4.04E 05 1.39E 04 6.57E 03 5.18E 04 4.5E 03 1.97E 06 0.00E-01 I-131 2.64E 04 6.48E 03 3.54E 04 4.90E 04 8.39E 04 1.4E 07 0.00E-01 0.00E-01 I-132 1.57E 03 1.27E 03 1.59E 03 4.3E 03 6.91E 03 1.51E 05 0.00E-01 0.00E-01 I-133 6.21E 03 1.03E 04 1.21E 04 2.05E 04 3.59E 04 2.92E 06 0.00E-01 0.00E-01 I-135 3.48E 03 6.94E 03 3.69E 03 9.4I 03 1.49E 04 6.2E 05 0.00E-01 0.00E-01 Cs-134 5.4E 05 9.75E 03 5.02E 05 1.13E 06 3.75E 05 0.00E-01 1.4E 05 0.00E-01 Cs-136 1.37E 05 1.09E 04 5.14E 04 1.9 I 05 1.1E 05 0.00E-01 1.77E 04 0.00E-01 l Cs-137 3.11E 05 8.48E 03 6.69E 05 8.47E 05 3.04E 05 0.00E-01 1.21E 05 0.00E-01 Ba-140 3.51E 03 2.2E 05 5.4E 04 6.69E 01 2.28E 01 0.00E-01 2.03E 06 0.0E-01 Ce-141 2.1E 03 1.2E 05 2.84E 04 1.89E 04 8.87E 03 0.00E-01 6.13E 05 0.00E-01 l Ce-144 2.62E 05 8.63E 05 4.88E 06, 2.0E 06 1.21E 06 0.00E-01 1.33E 07 0.00E-01

• R Values in units of mren/yr per micro-Cf/m-3 for inhalation and tritiun, and in units of M-2 area /yr per micro-C1/sec for all others.

3-47 l h1

TABLE 3.3-18 R RES FGt T}E H.B. RGIINSON STEAM ELECTRIC R. ANT

• f PATlWAY = Inhal AE GROUP = Child GI-Tract Liver Kidney Thyroid M mac11de T. Bog Bone Skin ]

H-3 1.12E 03 1.12E m 0.00E-01 1.12E 03 1.12E 03 1.12E 03 1.12E 03 1.12E 03 P-32 9.8E 04 4.21E 04 2.S)E 06 1.14E 05 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Cr-51 1.54E 02 1.0E 03 0.00E-01 0.00E-01 2.4I 01 8.5 I 01 1.7% 04 0.00E-01 Mn-54 9.50E 03 2.29E 04 0.00E-01 4.29E 04 1.00E 04 0.00E-01 1.57E 06 0.00E-01 ;

Fe-59 1.67E 04 7.0E 04 2.0E 04 3.34E 04 0.00E-01 0.00E-01 1.2E 06 0.00E-01 !

3.1E 03 Co-58 3.43E 04 0.00E-01 1.7E 03 0.00E-01 0.00E-01 1.10E 06 0.0E-01 Co-8) 2.2E 04 9.61E 04 0.00E-01 1.3E 04 0.00E-01 0.00E-01 7.Gui uo 0.00E-01 l Zn-65 7.02E 04 1.63E 04 4.25E 04 1.13E 05 7.11 04 0.00E-01 9.94E 05 0.00E-01 1 Rb-86 1.14E 05 7.9E 03 0.0E-01 1.9E 05 0.00E-01 0.00E-01 0.00E-01 0.00E-01 i Sr-89 1.72E 04 1.67E 05 5.99E 05 0.00E-01 0.00E-01 0.00E-01 2.15E 06 0.00E-01 Sr-90 6.4I 06 3.42 05 1.01E 08 0.00E-01 0.00E-01 0.0E-01 1.4 E 07 0.00E-01 Y-91 2.43E 04 1.84E 05 9.13E 05 0.00E-01 0.00E-01 0.00E-01 2.62E 06 0.00E-01 Zr-95 3.69E 04 6.1E 04 1.90E 05 4.1E 04 5.95E 04 0.00E-01 2.21 06 0.00E-01 ,

16-95 6.54E 03 3.69E 04 2.35E 04 9.1E 03 8.61E 03 0.00E-01 6.13E 05 0.00E-01 j Ru-103 1.0E m 4.47E 04 2.79E 03 0.00E-01 7.02E 03 0.00E-01 6.61E 05 0.00E-01 I

Ru-106 1.69E 04 4.29E 05 1.3E 05 0.00E-01 1.84E 05 0.00E-01 1.43E 07 0.00E-01 Ag-1104 9.11 03 1.00E 05 1.6E 04 1.1E 04 2.12E 04 0.00E-01 5.4?E 06 0.00E-01 Te-127M 3.01E 0:3 7.13E 04 2.48E 04 8.53E 03 6.35E 04 6.0E 03 1.48E 06 0.00E-01 Te-1291 3.04E 03 1.81E 05 1.92E 04 6.84E 03 5.02E 04 6.32E 03 1.7E 06 0.00E-01 1-131 2.72E 04 2.84E 03 4.80E 04 4.80E 04 7.87E 04 1.62E 07 0.00E-01 0.00E4)1 )

I-132 1.8K 03 3.20E 03 2.11E 03 4.0E 03 6.24E 03 1.93E 05 0.0E-01 0.00E-01 I-133 7.6BE 03 5.47E 03 1.6E 04 2.03E 04 3.3E 04 3.84E 06 0.00E-01 0.00E-01 1-135 4.14E 03 4.41 03 4.91E 03 8.72E 03 1.34E 04 7.91E 05 0.00E-01 0.00E-01 Cs-134 2.24E 05 3.84E 03 6.50E 05 1.01E 06 3.30E 05 0.00E-01 1.21E 05 0.0E-01 Cs-136 1.1E 05 4.17E 03 6.5E 04 1.71E 05 9.5I 04 0.00E-01 1.45E 04 0.0E-01 Cs-137 1.28E 05 3.61E 03 9.05E 05 8.24E 05 2.82E 05 0.00E-01 1.04E 05 0.00E-01 Ba-140 4.32E 03 1.02E 05 7.39E 04 6.47C 01 2.11E 01 0.00E-01 1.74E 06 0.00E-01 Ce-141 2.89E 03 5.65E 04 3.92E 04 1.95E 04 8.53E 03 0.00E-01 5.43E 05 0.00E-01 Ce-144 3.61E 05 3.8E 05 6.7E 06 2.11E 05 1.1E 06 0.00E-01 1.19E 07 0.00E-01

• R Values in units of mrem /yr per micro-C1/m-3 for inhalatica and tritiun, and in units of M-2 wWyr per micro-C1/sec for all others.

1 3-48

. _ _ _ _ _ . - - -- - - - - _ - - _ _ _-__ __ _- . 2 :L .

TABLE 3.3-19 R VALES FOR TE H.B. REINSON STEAM U.ECTRIC R#rT*

PATHWAY = Inhal AE GROUP = Infant naclide T.Bo# GI-Tract Bone Liver Kidney Thyroid Lung Stin H-3 6.4 E 02 6.4 E 02 0.00E-01 6.4 E 02 6.4E 02 6.4 E 02 6.4 E 02 6.4E 02 P-32 7.73E 04 1.61E 04 2.03E 06 1.12E 05 0.00E-01 0.00E-01 0.00E-01 0.00E-01 Cr-51 8.93E 01 3.5E 02 0.00E-01 0.00E-01 1.32E 01 5.7E 01 1.28E 04 0.00E-01 Mn-54 4.98E 03 7.0E 03 0.00E-01 2.53E 04 4.98E 03 0.00E-01 9.98E 05 0.00E-01 Fe-59 9.4E 03 2.47E 04 1.3 E 04 2.3E 04 0.0E-01 0.0E-01 1.01E 06 0.00E-01 Co-58 1.82E 03 1.11E 04 0.00E-01 1.22E 03 0.00E-01 0.00E-01 7.75E 05 0.00E-01 Co-60 1.18E 04 3.19E 04 0.00E-01 8.01E 03 0.00E-01 0.0E-01 4.50E 06 0.0E-01 In>65 3.1E 04 5.13E 04 1.93E 04 6.25E 04 3.24E 04 0.00E-01 6.4E 05 0.00E-01 Rb-86 8.81E 04 3.03E 03 0.00E-01 1.9E 05 0.0E-01 0.00E-01 0.0E-01 0.0E-01 Sr-89 1.14E 04 6.39E 04 3.97E 05 0.00E-01 0.00E-01 0.00E-01 2.03E 06 0.00E-01 Sr-90 2.59E 06 1.31E 05 4.08E 07 0.00E-01 0.00E-01 0.00E-01 1.12E 07 0.0E-01 Y-91 1.57E 04 7.02E 04 5.87E 05 0.00E-01 0.00E-01 0.00E-01 2.45E 06 0.00E-01 Z:-95 2.03E 04 2.17E 04 1.15E 05 2.78E 04 3.1 % 04 0.00E-01 1.75E 06 0.0E-01 M>-95 3.77E 03 1.27E 04 1.57E 04 6.42E 03 4.71E 03 0.00E-01 4.78E 05 0.00E-01 Ru-103 6.78E 02 1.6?E 04 2.01E 03 0.00E-01 4.24E 03 0.0E-01 5.51E 05 0.0E-01 l Ru-106 1.09E 04 1.64E 05 8.67E 04 0.00E-01 1.0E 05 0.00E-01 1.15E 07 0.00E-01 Ag-1104 4.99E 03 3.3 E 04 9.97E 03 7.21E 03 1.09E 04 0.0E-01 3.6E 06 0.00E-01 l Te-127M 2.07E 03 2.73E 04 1.6E 04 6.89E 03 3.75E 04 4.8E 03 1.31E 06 0.00E-01

! Te-129M 2.22E 03 6.89E 04 1.41E 04 6.08E 03 3.17E 04 5.47E 03 1.68E 06 0.0E-01 I-131 1.9E 04 1.0E 03 3.79E 04 4.43E 04 5.17E 04 1.48E 07 0.00E-01 0.00E-01 1-132 1.2E 03 1.90E 03 1.69E 03 3.54E 03 3.94E 03 1.69E 05 0.0E-01 0.00E-01 I-133 5.59E 03 2.15E 03 1.32E OA 1.92E 04 2.24E 04 3.55E 06 0.00E-01 0.0E-01 I-135 2.77E 03 1.83E 03 3.8E C3 7.59E 03 8.4E 03 6.95E 05 0.0E-01 0.00E-01 Cs-134 7.44E 04 1.33E 03 3.9E 05 7.02E 05 1.90E 05 0.00E-01 7.95E 04 0.00E-01 Cs-136 5.28E 04 1.43E 03 4.82E 04 1.34E 05 5.63E 04 0.00E-01 1.17E 04 0.00E-01 Cs-137 4.54E 04 1.3% 03 5.48E 05 6.11E 05 1.72E 05 0.00E-01 7.12E 04 0.00E-01 Ba-140 2.89E 03 3.83E 04 5.55E 'J4 5.59E 01 1.34E 01 0.0E-01 1.59E 06 0.0E-01 Ce-141 1.99E 03 2.1 E 04 2.77E 04 1.6E 04 5.24E 03 0.00E-01 5.1E 05 0.00E-01 Ce-144 1.7E 05 1.48E 05 3.19E 06 1.21E 06 5.37E 05 0.00E-01 9.83E 06 0.00E-01

• R Values in units of mren/yr per mien >-Cf/m-3 for inhalation and trititan, and in units of M-2 mrem /y, per micro-Cf/sec for all others.

l 3-19 g T2 -.1