Text

Proposed Tech Spec Revs,Incorporating New Requirements for Radiological Effluent Monitoring,Recording & Reporting in Accordance w/NUREG-0473.Description & Justification for Changes & Significant Hazards Consideration Encl
	ML20081J154
Person / Time
Site:	Browns Ferry
Issue date:	10/27/1983
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML20081J143	List: ML20081J141; ML20081J154; ML20081J171;
References
	RTR-NUREG-0473, RTR-NUREG-473 TAC-63022, TAC-63023, TAC-63024, NUDOCS 8311080327
	Download: ML20081J154 (185)
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. 73 f60 ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION REVISIONS BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3

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8311080327 831027 PDR ADOCK 05000259 P PDR

m T APPENDIX A TECHNICAL SPECIFICATIONS BROWNS FERRY UNITS 1 AND 2 L _ _ _ _ . _

T UNIT 1 TABLE OF C0hTEhTS e

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3 "ABLE OF CONTENTS Section Pace No.

. Introduction . . . . . . . . . . . . . . . . . . . 1 1.0 Definitions ................... 2 SAFETY LIMITS AND LIMIT.TNG SAFETY SYSTEM SETTINGS 1.1/2.1 Puel Cladding Integrity ...... . . . . . . . 8 1.2/2.2 Reactor Coolant System Integrity . . . . . . . . . 27 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3 1/4.1 Reacter Protection System ............ 31 3 2/4.2 Protective Inctrumentation . . . . . . . . . . . . 50 A. Primary Containment and Reactor Building Isolation Functions . . . . . . . . . . . . . 50 B. Core and Containment Cooling Systems -

< Initiation and Control. . . . . . . . . . . . 50 C. Control Rod Block Actuation . . . . . . . . . 51 D. Radioactive Liquid Effluent Monitoring Instrumentation . . . . . . . . . ... . . . . 51 E. Drywell Leak Detection. . . . . . . . . . . 52 4

F. Surveillance Instrumentation. . . . . . . . . 52 G. Control Room Isolation. . . . . . . . . . . . 52 l

l H. Flood Protection. . . . . . . . . . . . . . . 53 I. Meteorological Monitoring Instrumentation . . 53 J. Seismic Monitoring Instrumentation. . . . . . 54 i

K. Radioactive Gaseous Effluent Monitoring Instrumentation . . . . . . . . . . . . . . . 54 i

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m Section Pare No.

120 3 3/4.3 Reactivity Control . . . . . . . . . . . . . . . .

A. Reactivity Limitations. . . . . . . . . . . . 120 B. Control Rods. . . . . . . . . . . . . . . . . 121 C. Scram Insertion Times . . . . . . . . . . . . 124 D. Reactivity Anomalies ............ 125 E. Reactivity Control ............. 126 F. Scram Discharge Volume ........... 126 3 4/4.4 Standby Liquid Control System .......... 135 A. Normal System Availability ......... 135 B. Operation with Inoperable Components .... 136 C. Sodium Pentaborate Solution . . . . . . . . . 137 3 5/4.5 Core and Containment Cooling Systems . . . . . . . 143 A. Core Spray System (CSS) . . . . . . . . . . . . 143 B. Residual Heat Removal System (RHRS)

(LPCI and Containment Cooling) ....... 145 C. RER Service Water System and Emergency Equipment Cooling Water System (EECWS) ............... 151 D. Equipment Area Coolers ........... 154 E. High Pressure Coolant Injection System (HPCIS) . . . . . . . . . . . . . . . . . . . 154 F. Reactor Core Isolation Cooling System (RCICS) . . . . . . . . . . . . . . . . . . . 156 G. Automatic Depressurization System (ADS) . . . 157 H. Maintenance of Filled Discharge Pipe .... 158 I. Average Planar Linear Heat Generation Rate . 159 J. Linear Heat Generation Rate . . . . . . . . . 159 11

Section Pare No.

K. Minimum Critical Power Ratio (MCPR) ..... 160 L. Reporting Requirements ........... 160a i

3.6/4.6 Primary System Boundary ............. 174 A. Thermal and Pressurization Limitations ... 174 B. Coolant Chemistry . . . . . . . . . . . . . . 176 C. Coolant Leakage . . . . . . . . . . . . . . . 180 D. Relief Valves . . . . . . . . . . . . . . . . 181 E. Jet Pumps . . . . . . . ........... 181 F. Recirculation Pump Operation ........ 182 G. Structural Integrity ............ 182 H. Seismic Restraints, Supports, and Snubbers . 185 3 7/4.7 Containment Systems ............... 227 A. Primary Containment . . . . . . . . . . . . . 227 B. Standby Gas Treatment System ........ 236 C. Secondary Containment . . . . . . . . . . . . 240 D. Primary Containment Isolation Valves .... 242 E. Control Room Emergency Ventilation ..... 244 F. Primary Containment Purge System ...... 246 G. Containment Atmosphere Dilution System (CAD). 248 H. Containment Aar.osphere Monitoring (CAM)

System H2 Analyzer ............ 249 3 8/4.8 Radioactive Materials .............. 281 A. Liquid Effluents .............. 281 B. Airborne Effluents ............. 284 C. Radioactive Effluents - Dose ........ 287 iii

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P2ge No.

Section I-D. Mechanical Vacuum Pumps . . . . . . . . . . . 287 j E. Miscellaneous Radioactive Materials Sources . 288 F. Solid Radwaste ............... 290 3 9/4.9 Auxiliary Electrical System ........... 292 A. Auxiliary Electrical Equipment ....... 292 B. Operation with Inoperable Equipment . . . . . 295 C. Operation in Cold Shutdown ......... 298 3 10/4.10 Core Alterations . . . . . . . . . . . . . . . . . 302 A. Refueling Interlocks ............ 302 B. Core Monitoring . . . . . . . . . . . . . . . 305 C. Spent Fuel Pool Water . . . . . . . . . . . . 305 D. Reactor Building Crane. ........... 307 E. Spent Fuel Cask . . . . . . . . . . . . . . . 307 F. Spent Fuel Cask Handling-Refueling Floor. . . 308 3 11/4.11 Fire Protection Systems ............. 315 A. High Pressure Fire Protection System .... 315 B. CO2 Fire Protection System ........ 319 C. Fire Detectors ............... 320

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D. Roving Fire Watch . . . . . . . . . . . . . . 321 E. Fire Protection Systems Inspection ..... 322 F. Fire Protection Organization ........ 32E G. i'* Masks and Cylinders . . . . . . . . . . . 323 H. Continuous Fire Watch . . . . . . . . . . . . 323 I. Open Flames, Welding, and Bu ting in the Cable Spreading Room .......... 323 3.12/4.12 (RESERVED) . . . . .................

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S etion 3 13/4.13 Radiological Environmental Monitoring . . . . . . 329A A. Monitoring Program ............. 329A B. Land Use Census . . . . . . . . . . . . . . . 3290 C. Interlaboratory Comparison Program . . . . . 329E D. Deviations from Sampling Schedule . . . . . . 329F E. Exemption . . . . . . . . . . . . . . . . . . 329F 5.0 Major Design Features .............. 330 5.1 Site Features . . . . . . . . . . . . . . . . 330 5.2 Reactor . . . . . . . . . . . . . . . . . . . 330 5.3 Reactor Vessel ............... 330 5.4 Containment . ................ 330 l 5.5 Fuel Storage ................ 330 5.6 Seismic Design ............... 331 l

6.0 Administrative Controls . . . . . . . . . . . . . 332 6.1 Organization . . . . . . . . . . . . . . . . 332 6.2 Review and Audit . ............. 333 63 Procedures . . . . . . . . . . . . . . . . . 338 6.4 Actions to be Taken in the Event of a Reportable Occurrence in Plant Operation . . 346 6.5 Actions to be Taken in the Event a Safety

' Limit is Exceeded ............. 346 6.6 Station Operating Records . . . . . . . . . 346 f

f 6.7 Reporting Requirements . . . . . . . . . . . 349 6.8 Minimum Plant Staffing . . . . . . . . . . . 358 6.9 Environmental Qualification . . . . . . . . 358 6.10 Integrity of Systems outside containment . . 359 6.11 Iodine Monitoring ............. 359 Y

LIST OF TABLES j Table . Title Page No.

3 1.A Reactor Protection System (SCRAM) Instrumentation l Requirements . . . . . . . . . . . . . . . . . . . 33 4.1.A Reactor Protection System (SCRAM) Instrumencation Functional Tests Minimum Functional Test Frequencies for Safety Instrumentation and Control Circuits . . . . . . . . . . . . . . . . . .

37 4.1.B Reactor Protection System (SCRAM) Instrument Calibration Minimum Calibration Frequencies for Reactor Protection Instrument Channels . . . . . . 40 3 2.A Primary Containment and Reactor Building Isolation Instrumentation ................. 55 3 2.B Instrumentation that Initiates or Controls the Core and Containment Cooling Systems ......... 62 3 2.C Instrumentation that Initiates Rod Blocks .... 73 3 2.D Radioactive Liquid Effluent Monitoring Instrumentation .. ............... 76 3 2.E Instrumentation that Monitors Leakage Into Drywell . .................... 77 3.2.F Surveillance Instrumentation . . . . . . . . . . .- 78 3 2.G Control Room Isolation Instrumentation . . . . . . 81 3 2.H Flood Protection Instrumentation . . . . . . . . . 82 3 2.I Meteorological Monitoring Instrumentation .... 83 84 3 2.J Seismio Monitoring Instrumentation . . . . . . . .

3.2.K Radioactive Gaseous Effluent Monitoring Instrumentation ................. 84A 4.2.A Surveillance Requirements for Primary Containment and Reactor Building Isolation Instrumentation . . 85 4.2.B Surveillance Requirements for Instrumentation that Initiate or Control the CSCS . . . . . . . . . . . 96 4.2.C Surveillance Requirements for Instrumentation that Initiate Rod Blocks ............... 102 vi

LIST OF TABLES (Continu::d)

Table Title Pace No.

4.2.D Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirement . .... 103 .

l 4.2.E Minimum Test and Calibration Frequency for Drywell Leak Detection Instrumentation . . . . . . 104 4.2.F Minimum Test and Calibration Frequency for Surveillance Instrumentation . . . . . . . . . . . 105 4.2.G Surveillance Requirements for Control Room l Isolation Instrumentation ............ 106 l 4.2.H Minimum Test and Calibration Frequency for Floor Protection Instrumentation . . . . . . . . . 107 4.2.J Seismic Monitoring Instrument Surveillance . . . . 108 4.2.K Radioactive Gaseous Effluent Instrumentation Surveillance . . . . . . . . . . . . . . . . . . . 108A 3 5-1 Minimum RERSW and EECW Pump Assignment . ..... 152a 3 5.I MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE . . . . . . 171 4.6.A Reactor Coolant System Inservice Inspection Schedule . . . . . . . . . . . . . . . . . . . . . 209 3 7.A Primary Containment Isolation Valves . . . . . . . 250 3 7.B Testable Penetrations with Double 0-Ring Seals . . 256 3 7.C Testable Penetrations with Testable Bellows ... 257 3 7.D Primary Containment Testable Isolation valves .. 258 3 7.E Suppression Chamber Influent Lines Stop-Check Globe Valve Leakage Rates ............ 263 3 7.F Check Valves on Suppression Chamber Influent Lines ...................... 263 3 7.G C6eck Valves on Drywell Influent Lines . . . . . . 264 l

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3 7.H Testable Electrical Penetrations . . . . . . . . . 265 4.8.A Radioactive Liquid Waste Sampling and Analysis . . 287 4.8.B Radioactive Gaseous Waste Sampling and Analysis. . 288 vii

LIST OF TABLES (Cont'd,)

Table Title Page No.

4.9.A.4.C Voltage Relay Setpoints/ Diesel Generator Start . . 298a 3 11.A Fire Protection System Hydraulic Requirements. . . 324

'3.13.A Radiological Environmental Monitoring Program. . . 329G 3 13.B Pa.ximum values for the Lower Limits of Detection (LLD) . . . . . . . . . . . . . . . . . 329J 3 13.C Reporting Levels for Radioactivity Concentrations in Environmental Samples ............ 329M 6.8.A Minimum Shift Crew Requirements . . . . . . . . . 360 viii

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LIST OF ILLUSTRATIONS Figure Title Pare No.

2.1.1 APRM Flow Reference Scram and APRM Rod Block Settings . . . .. . . ............. 13 2.1-2 APRM Flow Bias Scram Vs. Reactor Core-Flow ... 26 4.1-1 Graphic Aid in the Selection of an Adequate Interval Between Tests ............ . 49 4.2-1 System Unavailability . . . . . .. ........ 119 3.4-1 Sodium Pentaborate Solution Volume Concentration Requirements . .. . ... ........... 138 3.4-2 Sodium Pentaborate Solution Temperature Requirements . . . . . ............. 139 3.5.K-1 MCPR Limits . . . . . . . . . . . . . . . . . . . 172b 3.5.2 Kg Factor . . . . . . . ............. 173 3.6-1 Minimum Temperature of Above Change in Transient T emperat ure . . . . . . . . . . . . . . . . . . . 194 3.6-2 Change in Charpy V Transition Temperature Vs.

Neutron Exposure . . .. ............ 195 4.8.1 Site Boundary . . . . . . . . . . . . . . . . . . 290G 6.1-1 TVA Office of Power Organization for Operation of Nuclear Power Plants . . . . . . . . . . . . . 361 6.1-2 Functional Organization . . . . . . . .-. . . . . 362 6.2-1 Review and Audit Function . . . . . . . . . . . . 363 6.3-1 In-Plant Fire Program Organization ....... 364 ix

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r 1 TABLE OF CONTENTS Section Pace No.

Introduction ~. . . . . . . . . . . . . . . . . . . I 1.0 Definitions ................... 2

' SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

. 1/2.1 Fuel Cladding Integrity ... ... ....... 8 1.2/2.2 Reactor Coolant System Integrity . . . . . . . . . 27 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMEh"fS 3 1/4.1 Reactor Protection System ............ 31 3.2/4.2 Protective Instrumentation . . . . . . . . . . . . 50 A. Primary Containment and Reactor Building Isolation Functions . . . . . . . . . . . . . 50 B. Cor; and Containment Cooling Systems -

Initiation and Control. . . . . . . . . . . . 50 C. Control Rod Block Actuation . . . . . . . . . 51 D. Radioactive Liquid Effluent Monitoring Instrumentation . . . . . . . . . . . . . . . 51 E. Drywell Leak Detection. . . . . . . . . . . . 52 F. Surveillance Instrumentation. . . . . . . . . 52 G. Control Room Isolation. . . . . . . . . . . . 52 H. Flood Protection. . . . . . . . . . . . . . . 53 I. Meteorological Monitoring Instrumentation . . 53 J. Sieismic Monitoring Instrumentation. . . . . . 54 K. Radioactive Gaseous Effluent Monitoring Instrumentation . . . . . . . . . . . . . . . 54 i

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1 Prge Ns.

Section E. Minimum Critical Power Ratio (MOPR) . . . . . 160 L. Reporting Requirements ........... 160a Primary System Boundary 174 3.6/4.6 .............

A. Thermal and Pressurization Limitations ... 174 B. Coolant Chemistry . . . . . . . . . . . . . . 176 C. Coolant Leakage . . . . . . . . . . . . . . . 180 D. Relief Valves . . . . . . . . . . . . . . . . 181 E. Jet Pumps . . . . . . . . . . . . . . . . . . 181 F. Recirculation Pump Operation ........ 182 G. Structural Integrity ............ 182 H. Seismic Restraints, Supports, and Snubbers . 185 3 7/4.7 Containment Systems ............... 227 A. Primary Centainment . . . . . . . . . . ... 227 B. Standby Gas Treatment System ........

236 C. Secondary Containment . . .......... 240 D. Primary Containment Isolation Valves .... 242 E. Control Room Emergency Ventilation ..... 244 F. Primary Containment Purge System ...... 246 G. Containment Atmosphere Dilution System (CAD). 248 H. Containment Atmosphere Monitoring (CAM)

System H2 Analyzer ............ 249 3 8/4.8 Radioactive Materials .............. 281 A. Liquid Effluents .............. 281 B. Airborne Effluents ............. 284 C. Radioactive Effluents - Dose ........ 287 iii

7 Section Parc Wo.

D. Mechanics.1 Vacuum Pumps . . . . . . . . . . . 287 E. Miscellaneous Radioactive Materials Sources . 238 t

F. Solid Radwaste ............... 290 3 9/4.9 Auxiliary Electrical System ........... 292 A. Auxiliary Electrical Equipment ....... 292 B. Operation with Inoperable Equipment . . ... . 295 C. Operation in Cold Shutdown ......... 298 3 10/4.10 Core Alterations . . . . . . . . . . . . . . . . . 302 A. Refueling Interlocks ............ 302 B. Core Monitoring . . . . . . . . . . . . . . . 305 C. Spent Fuel Pool Water . . . . . . . . . . . . 305 D. Reactor Building Crane. . . . . . . . . . . . 307 E. Spent Fuel Cask . . . . . . . . . . . . . . . 307 F. Spent Fuel Cask Handling-Refueling Floor. . . 308 3 11/4.11 Fire Protection Systems ............. 315 A. High Pressure Fire Protection System .... 315 B. CO2 Fire Protection System ........ 319 C. Fire Detectors. ............... 320 D. Roving Fire Watch . . . . . . . . . . . . . . 321 E. Fire Protection Systems Inspection ..... 322 F. Fire Protection Organization ........

322 G. Air Masks and Cylinder's . . . . . . . . . . . 323 H. Continuous Fire Watch . . . . . . . . . . . . 323 I. Open Flames, Welding, and Burning in the Cable Spreading Room .......... 323 3.12/3.12 (RESERVED) . . ....................

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S7etion

-3 13/4.13' Radiological Environmental Monitoring . . . . . . 329A A. Monitoring Program ............. 329A B. Land Use Census . . . . . . . . . . . . . . . 329C C. Interlaboratory Comparison Program . . . . . 329E D. Deviations from Sampling Schedule . . . . . . 329F E. Exemption . . . . . . . . . . . . . . . . . . 329F 5.0 Major Design Features .............. 330 5.1 Site Features . . . . . . . . . . . . . . . . 330 5.2 Reactor . . . . . . . . . . . . . . . . . . . 330 53 Reactor Vessel ............... 330 5.4 containment . ................ 330 5.5 Fuel Storage ................ 330 5.6 Seismic Design ............... 331 6.0 Administrative Controls . . . . . . . . . . . . . 332 i 6.1 organization . . . . . . . . . . . . . . . . 332 333 6.2 Review and Audit . .............

6.3 Procedures . . . . . . . . . . . . . . . . . 338 6.4 Actions to be Taken in the Event of a Reportable Occurrence in Plant Operation . . 346 l

6.5 Actions to be Taken in the Event a Safety l Limit is Exceeded ............. 346

6.6 Station Operating Records . . . . . . . . . 346 6.7 Reporting Requirements . . . . . . . . . . . 349 6.8 Minimum Plant Staffing . . . . . . . . . . . 358 6.9 Environmental Qualification . . . . . . . . 358 6.10 Integrity of Systems outside Containment . . 359 l

6.11 Iodine Monitoring ............. 359 l-Y

7-LIST OF TABLES Table Title Page No.

3 1.A Reactor Protection System (SCRAM) Instrumentation Requirements . . . . . . . . . . . . . ...... 33 4.1.A Reactor Protection System (SCRAM) Instrumentation Functional Tests Minimum Functional Test Frequencies for Safety Instrumentation and Control Circuits . . . . . . . . . . . . . . . . . 37 4.1.B Reactor Protection System (SCRAM) Instrument Calibration Minimum Calibration Frequencies for Reactor Protection Instrument Channels . . . . . . 40 3 2.A Primary Containment and Reactor Building Isolation Instrumentation .... ............. 55 3 2.B Instrumentation that Initiates or Controls the Core and Containment Cooling Systems ......... 62 3 2.C Instrumentation that Initiates Rod Blocks .... 73 3.2.D Radioactive Liquid Effluent Monitoring Instrumentation ................. 76 s

3.2.E Instrumentation that Monitors Leakage Into Drywell ..................... 77 3 2.F Surveillance Instrumentation . . . . . . . . . . . 78 3 2.G Control Room Isolation Instrumentation . . . . . . 81 3 2.H Flood Protection Instrumentation . . . . . . . . . 82 3 2.I Meteorological Monitoring Instrumentation .... 83 3 2.J Seismic Monitoring Instrumentation . . . . . . . . 84 3 2.K Radioactive Gaseous Effluent Monitoring Instrumentation . ... ............. 84A 4.2.A Surveillance Requirements for Primary Con'ainment and Reactor Building Isolation Instrumentation . . 85 4.2.B Surveillance Requirements for Instrumen*ation that Initiate or Control the CSCS . . . . . . . . . . 89 4.2.C Surveillance Requirements for Instrumentation that Initiate Rod Blocks . .............. 102 vi

LIST OF TABLES (Continu2d)

Table Title Page No.

4.2.D Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirement . .... 103 4.2.E Minimum Test and Calibration Frequency for Drywell Leak Detection Instrumentation . . . . . . 104 4.2.F Minimum Test and Calibration Frequency for Surveillance Instrumentation . . . . . . . . . . . 105 4.2.G Surveillance Requirements for Control Room Isolation Instrumentation ............ 106

_ 4.2.H Minimum Test and Calibration Frequency for Floor Protection Instrumentation . . . . . . . . . 107 4.2.J Seismic Monitoring Instrument Surveillance . . . . 108

4. 2.K ' Radioactive Gaseous Effluent Instrumentation Surveillance . . . . . . . . . . . . . . . . . . . 10BA 3 5-1 Minimum RHRSW and EECW Pump Assignment . . . . . . 152a 3 5.I MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE . . . . . . 171 4.6.A Reactor Coolant System Inservice Inspection Schedule . . . . . . . . . . . . . . . . . . . . . ^209 3 7.A Primary Containment Isolation valves . . . . . . . 250 3 7.B Testable Penetrations with Double 0-Ring Seals . . 256 3 7.C Testable Penetrations with Testable Bellows ... 257 3 7.D Primary Containment Testable Isolation valves .. 258 i

3 7.E Suppression Chamber Influent Lines Stop-Check Globe Valve Leakage Rates ............ 263 3 7.F Check Valves on Suppression Chamber Influent l

Lines ...................... 263 3 7.0 Check valves on Drywell Influent Lines . . . . . . 264 3 7.H Testable Electrical Penetrations . . . . . . . . . 265 4.8.A Radioactive Liquid Waste Sampling and Analysis . . 287 4.8.B Radioactive Gaseous Waste Sampling and Analysis. . 288 vii

LIST OF TABLES (Cont'd)

Table Title Page No.

4.9.A.4.C Voltage Relay Setpoints/ Diesel Generator Start . . 29Ba 3 11.A Fire Protection System Hydraulic Requirements. . . 324 3 13.k Radiological Environmental Monitoring Program. .. 329G 3 13.B Haximum values.for the Lower Limits of Detection (LLD) . . . ... . . . . . . . . . . . . 329J 3 13.C Reporting Levels for Radioactivity Concentrations in Environmental Samples ............ 329M 6.8.A Minimum Shift Crew Requirements . . . . . . . . . 360 l

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2.1.1 APRM Flow Reference Scram and APRM Rod Block Settings ............. ....... 13 2.1-2 APRM Flow Bias Scram Vs. Reactor Core-Flow . . . 26 4.1-1 Graphic Aid in the Selection of an Adequate Interval Between Tests . ............ 49 4.2-1 System Unavailability . . . . . . . . . . . . . . 119 3.4-1 Sodium Pentaborate Solution Volume Concentration Requirements ..... . ............ 133 3 4-2 Sodium Pentaborate Solution Temperature Requirements ........... ....... 139 3 5.K-1 MCPR Limits . . . . . . . ............ 172a 3 5.2 K r Factor . . . . . . . . . . . . . . . . . . . . 173 3.6-1 Minimum TemperatureC F Above Change in Transient Temperature . .................. 194 3.6-2 Change in Charpy V Transition Temperature Vs.

Neutron Exposure .. .... ........... 195 4.8.1 Site Boundary . . .. . . ........... 290G 6.1-1 TVA Office of Power Organization for Operation of Nuclear Power Plants . . . . . . . . . . . . . 361 6.1-2 Functional Organization . . . . . . . . . . . . . 362 6.2-1 Review and Audit Function . . . . . . . . . . . . 363 6.3-1 In-Plant Fire Program Organization ....... 364 s

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1.0 ' DEFINITIONS (Cont'd)

10. Logie - A logic is an arrangement of relays, contacts, and other components that produces a decision output.

(a) Initiating - A logic that receive signals from channels and produce decision outputs to the actuation logic.

(b) Actuation - A logic that receives signals (either from initiation logic or channels) and produces decision outputs to accomplish a protective action.

11. Channel Calibration - Shall be the adjustment, as necessary, of the channel output such that it responds with necessary range and accuracy to known values of the parameters which the channel monitors. The channel calibration shall encompass the entire channel including alarm and/or trip functions and shall include the channel functional test. The channel calibration may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated. Non-calibratable components shall be excluded from this requirement, but will included in channel functieaal test and source check.

12. Channel Functional Test - Shall be :

a. Analog Channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

b. Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

13. Source Check - Shall be the qualitative assessment of channsi response when the channel sensor is exposed to a radioactive source or multiple of sources.

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1.0 DITINITIONS (Cont'd)

W. Functional Tests - A functional test is the manual operation or initiation of a system, subsystem, or component to verify that it functions within design tolerances (e.g., the manual start of a core spray pump to verify that it runs and that it pumps the required volume of water).

S. Shutdown - The reactor is in a shutdown condition when the reactor mode switch is in the shutdown mode position and no core alterations are being performed.

Y. Engineered Safeguard - An engineered safeguard is a safety system the actions of which are essential to a safety action required in response to accidents.

Z. Solidification - Shall be the conversion of radioactive wastes into a form that meets shipping and burial ground ,

, requirements.

AA. Offsite Dose Calculation Manual (ODCM) - Shall be a manual describing the environmental monitoring program and the methodology and parameters used in the calculation of release rate limits and offsite doses _due to radioactive gaseous and liquid effluents. The ODCM will also provide the plant with guidance for establishing alarm / trip setpoints to ensure technical specifications sections 3.8.A.1 and 3.8.B.1 are not exceeded.

BB. Purge or nurging - The controlled process of discharging air

, or gas from the primary containment to maintain tenperature, pressure, humidity, concentration, or other operating condition in sr.ch a manner that replacement air or gas is required to purify the containment.

CC. -Process Control Program - Shall contain the sampling, analysis, and formulation determination by which SOLIDIFICATION of radioactive wastes from liquid systems is assured.

DD. Venting - The controlled process of discharging air or gas from the primary containment to maintain temperature, pressure, humidity, concentration, or other operating condition in such a manner that replacement air or gas is not

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

7A

1.0 DEFINITIONS (Cent'd)

EE. Site Boundary - Shall be that line beyond which the land is not ownus, leased, or otherwise controlled by TVA.

IT. Unrestricted Area - 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 industrial, commercial, institutional, or recreational purposes.

GG. Dose Eauivalent I-131 - The DOSE EQUIVALENT I-131 shall be the concentration of I-131 (in pCi/gs) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factor used for this calculation shall be those listed in Table III of TID-14844 " Calculation of Distance Factors for Power and Test Reactor Sites".

HH. Gaseous Waste Treatment System - The charcoal adsorber vessels installed on the discharge of the steam jet air ejector to provide delay to a unit's offgas activity prior to release.

II. Members 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, occupational, or other purposes not associated 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 restricted areas.

JJ. Surveillance - Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS or other conditions specified for individual limiting conditions for operation unless otherwise L stated in an individual Surveillance Requirements. Each Surveillance Requirement shall be performed within the specified time interval with, (1) A maximum allowable extention not to exceed 25% of the surveillance interval, but (2) The combined time entered for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval Performance of a Surveillance Requirement within the specified time interval shall constitute compliance and OPERABILITY requirements for a limiting condition for operation and associated action statements unless otherwise required by these specifications. Surveillance requirements do not have to be performed on inoperable equipment.

7B

Table 1.1 SURVEILLANCE FREQUENCY NOTATION NOTATION FREQUENCY S (Shift) At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ,

D (Daily) At least once per normal calendar 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months day (midnight to midnight).

W (Weekly) At least once per 7 days.

M (Monthly) At least once per 31 days.

Q (Quarterly) At least once per 3 months or 92 days.

SA (Semi-Annually) At least once per 6 months or 184 days.

Y_ (Yearly) At least once per year or 366 days.

R (Refueling) At least once per operating cycle.

S/U (Start-Up) Prior to each reactor startup.

N.A. Not applicable.

P (Prior) Completed prior to each release.

7C

!.IMITING CONDITIONS FOR OPERAT20N SURVEII.IANCE RE013IREENTS 3.2.B Core and Containment Cooling 4.2.B Core and Containment Cooling Systems - Initiation & Control Systems - Initiation & Control are required to be cperable shall be considered operable if they are within the required surveillance testing frequency and there is no reason to suspect that they are <

inoperable.

C. Control Rod Block Actuation C. Control Rod Block Actuation The limiting conditions of Instrumentation shall be function-operation for the instrumentation ally tested, calibrated, and that initiates control rod block checked as indicated in Table 4.2.C.

are given in Table 3.2.C.

System logic shall be functionally tested as indicated in Table 4.2.C.

3.2.D Radioactive Liquid Effluent 4.2.D Radioactive Liquid Effluent Monitoring instrumentation Monitoring Instru:ssntation

1. The radioactive liquid 1. Each of the radioactive effluent monitoring liquid effluent monitoring instrumentation listed in instruments shall be Table 3.2.D shall be demonstrated operable by operable with the performance of test in applicability as shown in accordance with Table 4.2.D.

Tables 3.2.D/4.2.D. Alarm /

trip setpoints will be set in accordance with guidance given in the ODCM to ensure that the limits of specifi-cation 3.8.A.1 are not exceeded.

2. The' action required when the number of operable channels is less than the minimum channels operable requirement is specified in the notes for Table 3.2.D. Exert best efforts to return the instru-ment (s) to OPERADLE status within 30 days and if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was act corrected in a timely manner.

51 l

l

. I LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMEh7S l l

3.2.D Radioactive Liouid Effluent 4.2.D Radioactive Liquid Effluent (Con't) (Con't)

3. With a radioactive liquid effluent monitoring channel alare/ trip setpoint less conservative than required by these specifications, 4

suspend the release without delay, declare the channel inoperable, or adjust the alare/ trip setpoint to establish the conservatism required by these specifi-cations.

4. The provisions of specifi-cations 1.0.C and 6.7.2 are not applicable.

E. Drywell Leak Detection E. Drywell Leak Detection The limiting conditions of opera- Instrumentation shall be tion for the instrumentation that calibrated and checked as monitors drywell leak detection indicated in Table 4.2 E.

are given in Table 3.2.E.

4 F. Surveillance Instrumentation F. Surveillance Instrumentation The limiting conditions for the Instrumentation shall be instrumentation that provides calibrated and checked as surveillance information readouts indicated in Table 4.2.F.

are given in Table 3.2.F.

G. Control Room Isolation G. Control Room Isolation The limiting conditions for Instrumentation shall be instrumentation that isolates the calibrated and checked as control room and initiates the indicated in Table 4.2.G.

l control room emergency pressurization systems are given j in Table 3.2.G.

t 52 l

t

LIMITING CONDITIONS FCR OPERATION SURVEILLANCE REQUIREMENTS -3.2.J Seismic Monitoring Instrumentation 4.2.J Seismic Monitorine Instrumentation

1. The seismic monitoring 1. Each of the seismic monitoring instruments listed in shall be demonstrated operable table 3.2.J shall be operable by performance cf tests at the at all times. frequencies listed in table 4.2.J.

2. With the number of seismic monitoring instruments less 2. Data shall be retrieved ftbm than the number listed in all seismic instruments table 3.t.J, restore the actuated during a seismic inoperable instrument (s) to event and analyzed to operable status within 30 days. determine the magnitude of the vibratory ground motion.

3. With one or more of the A Special Report shall be instruments listed in table submitted to the Coce.tission 3.2.J inoperable for more than pursuant to specification 30 days, submit a Special Report 6.7.3.D within 10 days to the Commission pursuant to describing the magnitude, specification 6.7.3.C within frequency spectrum, and the next 10 days describing the resultant effect upon plant .

cause of the malfunction and features important to safety.

plans for restoring the instruments to operable status.

~

3.2.K Radioactive Gaseous Effluent 4.2.K Radioactive Gaseous Effluent Monitoring Instrumentation Monitoring Instrumentation

1. The radioactive gaseous 1. Each of the radioactive effluent monitoring gaseous effluent monitoring instruments listed in instruments shall be demon-Table 3.2.K shall be strated operable by perfor-operable with the mance of tests in accordance applicability as shown in with Table 4.2.K.

Tables 3.2.K/4.2.K. Alarm /

trip setpoints will be set in accordance with guidance given in the ODCM to ensure that the limits of specification 3.8.B.1 are not exceeded.

t 54

SURVEILLANCE REQUIREMENTS l LIMITING CONDITIONS FOR OPERATION 3.2.K Radioactive Gaseous Effluent 4.2.K Radioactive Gaseous Effluent Monitoring Instrumentation Monitoring Instrumentation (Cca't) (Con't)

2. The action required when the number of operable channels is less than the Minimum Channels Operable requirement is specified .

in the notes for Table 3.2.K.

Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Semiannual l Radioactive Release Report why the inoperability was

.not corre:ted in a timely manner.

3. With a radioactive gaseous effluent monitoring channel alarm / trip setpoint less conservative than required by these specifications, suspend the release without delay, declare the channel inoperable, or adjust the alarm / trip setpoint to establish .the conservatism required by these specifications.

4. The provisions of specifi-cations 1.0.C and 6.7.2 are not applicable.

1 54A i

a f

e

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

i TABLE 3.2.D 4

Radioactive Liquid Effluent Monitorina Instrumentation Hinimum Channels j

Instrument (p) Operable gplicability Action

. 1. LIQUID RADWASTE EFFLUENT 1 ** A, B HONITOR (RN-90-130)

2. RilR SERVICE WATER HONITOR 1 *** C (RH-90-133, -134) y 3. RAW COOLING WATER HONITOR 1 ** D e (RH-90-132)

{ 4. LIQUID RADWASTE EFFLUENT 1 ** E

{ FLOW RATE (77-60 loop excluding fixed in line rotometer) 4 r

l i

1 e

__ _ _ _ _ - . _ _-

l l

NOTES FOR TABLE 3.2.D

At all times

- During releases via this pathway

- - During cperation of an RHR loop having RHR system pressure greater than service water pressure ACTION A During release of radioactive wastes from the radwaste processing system, the following shall be met (1) liquid waste activity and flowrate shall be continuously monitored and recorded during release and shall be set to alarm and automatically close the waste discharge valve before exceeding the limits i specified in 3.8.A.1, (2) if this cannot be met, two independent samples of the tank being discharged shall be analyzed in accordance with Table 4.8.A; and two qualified station personnel shall independently check valving before the discharge. Otherwise, suspend release via this pathway.

ACTION B Vith a radioactive liquid effluent monitoring channel / alarm trip setpoint less conservative than required by these specifications, suspend release via this pathway without delay, declare the channel inoperable, or adjust the alarm / trip setpoint to establish the censervatism referred by these specifications.

ACTION C During operation of an RER loop having RHR system pressure greater than service water pressure, the effluent from that unit's service water shall be continuously monitored. If an installed monitoring system is not available, either a temporary monitor is installed or grab samples taken every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed for radioactivity with an LLD of IE-7 pCi/ml (gross) or less than the applicable MPC ratio (y isotopic).

ACTION D With the number of channels OPERABLE less than required by the Minimu:n Channels Operable requirement, effluent releases via this pathway may continue provided that a temporary monitor is installed or, at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , grab samples are collected and analyzed for radioactivity with an LLD cf IE-7 pCi/ml.

(gross) or.( applicable MPC ratio (y isotopic). ,

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

ACTION F Alarm / trip setpoints will be calculated in accordance with the guidance given in the Offsite Dose Calculation Manual (ODCM).

76A

Table 3 2.I Meteorological Monitoring Instrumentation INSTRUMENT HINIMUM

. ACCURACY OPERABLE INSTRUMENT

1. WIND SPEED

a. Channel A Elevation 620 MSL Note #1 1

b. Channel B Elevation 737 MSL Note #1 1

c. Channel C Elevation 887 MSL Note #1 1

2. WIND DIRECTION ,

a. Channel A Elevation 620 MSL + 50 1

b. Channel B Elevation 737 HSL i 50 1 887 MSL 1 50 1

c. Channel C Elevation 3 AIR TEMPERATURE - DELTA T

a. Channel A Elevation 620-737 MSL 0.10C 1

b. Channel B Elevation 620-887 MSL 0.1 0C 1

h. DEW POINT

~

a. Elevation 620 IGL 0.1"C l' Channel A_

h. Channel B Elevation 620 MSL 0.1 C 1 NOTE #1 - Starting speed or anemometer shall be <1 mph. Accuracy is within i 1 percent of sph reading or 0.15 mph, whichever is greater.

83

TABLE 3.20K Radioactive Gascons Effluent Honitoring Instrumentation Hinimum Channels /

Instrument Devices Operable Applicability ,

Action

1. STACK (RN-90-147A & B)

A/C

a. Noble Gas Monitor (1)

(1)

B/C

b. Iodine Cartridge (1)

B/C

c. Particulate Filter

D

d. Sampler Flow Abnormal (1)

Stack Flow (FT, FM, (1)

D c.

FI-90-271)

2. REACTOR / TURBINE BLDG VENTILATION (RH-90-250) *

a. Noble Gas Honitor (1) A/C Iodine Sampler (1)

B/C b.

Particulate Sampler (1)

B/C c.

D

d. Sampler Flowmeter (1)

3. TURBINE BLDG EXIIAUST (RM-90-249, 251)

- A/C

a. Nobile Gas Honitor (1)

- B/C

b. Iodine Sampler (1)

Pa'rticulte Sampler (1) ** B/C

c. **

(1) D

d. Sampler Flowmeter

4. HADWASTE BLDG VENT (RH-90-252) *

a. Noble Gas Honitor (1) A/C

D/C

b. lodine Sampler (1)

B/C

c. Particulate Sampler (1)

(1)

D

d. Sampler Flowrate

5. OFF GAS IlYDROGEN ANALYZER

- - E (ll A 2 e ll 2H) (1)

6. OFF GAS POST TREATHENT

a. Noble Gas Activity Honitor

F (101-90-265, 266) (1)

b. Sample Flow Abnormal

D (PA-20-262) (1)

NO~ES FOR TAEE 3.2.K

At all times

- During releases via this pathway

- - During main condenser offgas treatment system operation ACTION A With the number of channels OPERABE less than required by the Minimum Channels Operable requirement, effluent releases via the affected pathway may continue provided a temporary monitoring system is installed or grab samples are taken and analyzed at least once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

ACTION B Vith the number of channels OPEPJLBLE less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue provided samples are continuously collected with auxiliary sampling equipment for periods on the order of seven (7) days and analyzed in accordance with Table 4.8.B within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after-the end of the sampling period.

ACTION C A monitoring system may be out of service for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for functional testing, calibration, or repair without providing or initiating grab sampling.

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

ACTION E With the number of channels OPERABLE less than required by the Minimum Channels Operable requirement, operation of main condenser offgas treatment system may continue provided that a temporary monitor is installed or grab samples are collected at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION F With the number of channels OPERABLE less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

84B

TABLE 4.2.D Radioactive Liquid Effluent Honitoring Instrumentation Surveillance Requirements Channel Funcational Instrument Instrument Check Source Check Calibration Test

1. LIQUID RA0 WASTE EFFLUENT D(4) H R(5) Q(1)

HONITOR (RH-90-130)

2. HilR SERVICE WATER HONITOR D(4) H R(5) Q(2)

(RN-90-133, -134)

3. RAW COOLING WATER HONITOR D(4) H R(5) Q(2)

(RH-90-132)

4. LIQUID RADWASTE EFFLUENT D(4) NA R Q(3)

Fl.0W RATE (77-66 loop) 103 ,

NOTES FOR TABLE 4.2.D (1) The channel functional test shall also demonstrate that automatic isolation of this pathway and control room annunciatica occurs if any of the following conditions exist:

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

b. Instrument indicates an inoperative /downscale failure

c. Instrument controls not set in operate mode (2) The channel functional test shall also demonstrate that control room annunciation occurs if any of the following conditions exist:

a. Instrument indicates measured levels above; the alarm setpoint

b. Instrument indicates an inoperative /downscale failure

c. Instrument controls not set in operate mode (3) This functional test shall consist of measuring rate of tank decrease over a period of time and comparing this value with flow rate instrument reading.

(4) INSTRIMENT CHECK shall consist of verifying indication during periods of release. INSTRUMENT CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days which continous, periodic, or batch releases are made.

(5) The CHANNEL CALIBRATION shall include the use of a known (traceable to National Bureau of Standards Radiation Measurement System) radioactive source (s) positioned in a reproducible geometry with respect to the sensor 'or using standards that have been obtained from suppliers that participate in measurement assurance activities witn the National Bureau of-Standards (NBS).

103A

--ee. ----yae-w-ns +-m

-g y ,-nm , - y ,-y- - - -g y y,- ,y 4 -,e y- y mm- -

i TABl.E 4.2.K Radioactive Gaseous Effluent Instrumentation Surveillance Functional Channel Source Check Calibration Test Instrument Instrument Check

1. STAM Noble Gas Monitor (5) D H R(1) Q(2) j a. NA W NA NA

b. Iodine Cartridge NA NA

c. Particulate Filter W. NA D NA R Q l d. Sampler Flowmeter NA R Q

e. Stack Flowmeter D

2. REACT 0H/TURilINE BLDG g T R(1)' Q(2)

a. Noble Gas Monitor D H NA NA NA

b. Iodine Sampler W NA NA NA

c. Particulate Sampler W ,

D NA - R Q

> d. Sampler Flowmeter ,

3. TURBINE Bl.DG EXIIAllST Nobile Gas Honitor f0) D H R( Q i a. NA NA Iodine Sampler W NA

b. NA W NA NA

c. Particulte Sampler i

D NA R Q i d. Sampler Flowmeter

! 4. RADWASTE BLDG VENT R(1) Q(2)

Noble Gas Monitor (6) D H

a. NA W NA NA j b. Iodine Sampler NA NA W NA

c. Particulate Sampler R Q l D NA

d. Sampler Flowrate i

i 5. OFF GAS 11YDROGEN ANALYZER NA R(3) qW (11 N

] 2 ^ II 20) l 6. OFF FAS POST THEATt!ENT R(1)

a. Noble Gas Activity Honitor D H R

Q((2)

Q 4)

D NA

b. Sample Flow Abnormal

,l i

l 108A a

NOTES FOR TABLE 4.2.N (1) The CHANNEL CALBIRATION shall include the use of a known (traceable to the National Bureau of standards radiation measurement system) radioactive source (s) positioned in a reproducible geometry with respect to the sensor or using standards that have obtained from suppliers that participate in measurement assurance activities with the National Bureau of Standards.

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:

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

b. Instrument indicates an inoperable /downscale failure.

c. Instrument controls not set in operate mode (stack only).

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

a. Zero volume percent hydrogen (compressed air) and,

b. One volume percent hydrogen, balance nitrogen.

(4) The channel functional test shall demonstrate that automatic isolation of this pathway and control room annunciation occurs if any of the following

. conditions exists:

a. Instrument indicates measured levels above the alar =/ trip setpoint.

b. Instrument indicates an inoperative /downscale failure.

c. Instrument controls not set in operate mode.

The two channels are arranged in a coincidence logic such that 2 upscale, or 3 downscale and I upscale or 2 downscale will isolate the offgas line.

(5) The noble gas monitor shall have a LLD of IE-5 (Xe 133 Equivalent).

(6) The noble gas monitor shall have a LLD of IE-6 (Xe 133 Equivalent).

108B

T 3.2 BASES Tor each parameter monitored, as listed in Table 3.2.F, there are two channels of instrumentation except as noted. By comparing readings l between the two channels, a near continuous surveillance of instrument performance is available. Any deviation in readings will initiate an early recalibration, thereby maintaining the quality of the instrument

-adings.

Instrumentation is provided for isolating the control room and initiating a pressurizing system that processes outside air before supplying it to the control room. An accident signal that isolates primary containment will also automatically isolate the control room and initiate the emergency pressurization system. In additica, there are radiation monitors in the normal ventilation system that will isolate the control room and initiate the emergency pressuri:aiton system. Activity required to cause automatic actuation is about one mrem /hr.

Because of the constant surveillance and control exercised by TVA over the Tennessee River, flood levels of large mangitudes can be predicted in advance of their actual occurrence. In all cases, full advantage will be taken of advance warning to take appropriate action whenever reservoir levels above normal pool are predicted; however, the plant flood protection is always in place and does not depend in any way on advanced warning. Therefore, during flood conditions, the plant will be permitted to operate until water begins to run across the top of the pumping station at elevation 565. Seismically qualified, redundant level switches each powered from a separate division of power are provided at the pumping station to give main control room indication of this condition. At that time an orderly shutdown of the plant will be initiated, although surges even to a depth of several feet over the pumping station deck will not cause the loss of the main condenser circulating water pumps.

The operability of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation dose to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public.

The operability of the seismic instrumentaiton ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for Browns Ferry Nuclear Plant. The instrumentation provided is consistent with specific portions of the recommendations of Regulatory Guide 1.12

" Instrumentation for Earthquakes."

115

y l

The radioactive gaseous effluent instrumentation is provided to monitor and centrol, as applicable, the releases of radioactive materials in gaseous effluents during actual er potential releases of gaseous effluents. The alarm / trip setpoints for these instruments will be calculated in accordance with guidance provided in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisicas for monitoring the concentration of potentially explosive gas mixtures in the offgas holdup system. The operability and use of this instrumentation is consistent with the requirements of General Design-Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid I

effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with guidance provided in the ODCM to ensure i that the alarm / trip will occur prior to exceeding the limits of 1 10 CFR Part 20 Appendix B, Table II, Column 2. The OPERABILITY and use of this instrumentation is consistent with the requirem-nts of General ;

Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

i 9

9 e

0 115A

4.2 BASES there is no true minimum. The curve does have a definite knee and very little reduction in system unavailability is achieved by testing at a shorter interval than computed by the question for a single channel.

The best test procedure of all those examined is to perfectly stagger the tests. That is, if the test interval is four months, test one or the other channel every two months. This is shown in Curve No. 5. The difference between Cases 4 and 5 is negligible. There may be other arguments, however, that more strongly support the perfectly staggered tests, including reductions in human error.

The conclusions to be drawn are these:

1. A 1 out of n system may be treated the same as a single channel in terms of choosing a test interval; and

2. more than one channel should not be bypassed for testing at any one time. ,

'The radiation monitors in the refueling area ventilation duct which initiate building isolation and standby gas treatment operation are arranged in two 1 out of 2 logic systems. The bases given for the rod blocks apply here also and were used to arrive at the functional testing frequency. The off-gas post treatment monitors are connected in a 2 out of 2 logic arrangement. Based on experience with instruments of similar design, a testing interval of once every three months has been found adequate.

The automatic pressure relief instrumentation can be considered to be a 1 out of 2 logic system and the discussion above applies also.

The criteria for ensuring the reliability and accuracy of the radioactive gaseous effluent instrumentation is listed in Table 4.2.K.

The crit'eria for ensuring the reliability and accuracy of the radioactive liquid effluent instrumentation is listed in Table 4.2.D.

118

LIMITING CONDITIONS FOR OPERATION SURVEILI.ANCE REQUIREMEh"IS 3.6 PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY

6. Wenever the reactor is 6. Additional coolant critical, the limits on samples shall be taken activity concentrations in whenever the reactor the reactor coolant shall activity exceeds one not exceed the equilibrium percent of the equili-value of 3.2 pc/gm of dose brium concentration equivalent I-131. specificed in 3.6.B.6 and one of the following This limit may be exceeded conditions are met:

following power transients for a maximum of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . a. During startup During this activity b. Following z transient the iodine concen- significant trations shall not exceed power change **

26 pCi/gm whenever the c.

reactor is critical. The Following an,he increase in t reactor shall not be operate'd equilibrium off-more than 5 percent of its gas level yearly power operation under exceeding 10,000 this exception for the pCi/sec (at the equilibrium activity limits. steam jet air If the iodine concentraiton ejector) within in the coolant exceeds a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period.

26 pCi/gm, the reactor shall d. Whenever the equi-be shut down, and the steam librium iodine limit line isolation valves shall specified in 3.6.B.5 be closed immediately. is exceeded.

The additional coolant liquid samples shall be taken at 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months intervals for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or until a stable iodine concen-tration below the limit-ing value (3.2 pCi/ge) is established. However, at least 3 consecutive

- For the purpose of this section samples shall be taken on sampling frequency, a in all cases. An isoto-

, significant power exchange is pic analysis shall be defined as a change exceeding performed for each sample, 15% of rated power in less than and quantitative measure-I hour. ments made to determine the dose equivalent I-131 concentration. If the total iodine activity of the sample is below 0.32 pCi/gm, an isotopic analysis to determine 4 equivalent I-131 is not required.

179

I.IMITING CONDITIONS FOR OPERATION SURVEILLANCE PIQUIREMENTS 3.8 RADIOACTIVE MATERIALS 4.8 RADIOACTIVE MATERIALS Applicability Applicability Applies to the release of Appl 1es to the periodic test radioactive liquids and gases and record requirements and from the facility. sampling and monitoring methods used for facility Objective effluents.

To define the limits and Objective conditions for the release of radioactive effluents to the To ensure that radioactive environs to assure that any liquid and gaseous releases from the facility are radioactive releases are as low as reasonably achievable maintained within the limits and within the limits of specified by Specifications 10 CFR Part 20. The specifi- 3.8.A and 3.8.B.

cations except for 3.8.A.1 and 3.8.B.1 are exempt from the Specification requirements of definition 1.0.C (Limiting Condition for A. Licuid Effluents Operation). Facility records shall 1.

Specification be maintained of

. radioactive concentrations A. Liquid Effluents _ _ and volume before

. dilution of each batch of liquid effluent

1. The concentration of released, and of the radioactive material average dilution flow released at any time from the site to and length of time over l

which each discharge unrestricted areas (see Figure 4.8-1) occurred.

I shall be limited to 2. Radioactive liquid waste l the concentrations sampling and. activity specified in 10 CFR analytis of each liquid Part 20, Appendix B, Table II, Column 2 waste batch to be discharged shall be for radionuclides other performed prior to than dissolved or entrained noble gases. release in accordance For dissolved or with Table 4.8.A.

entrained noble gases, the concentration shall 3. The operation of the be limited to 2E-4 pCi/ml automatic isolation valves and discharge total activity, tank selection valves ~

2. If the limits of 3.8.A.1 shall be checked annually.

are exceeded, appropriate l action shall be initiated l

without delay to bring -

l the release within 281

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

. ~

3.6 RADIOACTIVE MATERIALS 4.8 RADIOACTIVE MATERIALS limits. Provide prompt 4. The results of the analysis notification to the NRC of samples collected from pursuant to section release points shall be used 6.7.2.A. with the calculational methodology in the ODCM to

3. The dose or dose commit- assure that the concentrations ment to a member of the at the point of release are public from radioactive maintained within the limits materials in liquid of specification 3.8.A.I.

effluents released from each unit to unrestricted 5. Cummulative quarterly and areas (See Figure 4.8-1) yearly dose contributions shall be limited: from liquid effluents shall be determined as specified

a. During any calendar in the ODCM at least once quarter to (1.5 mrem to every 31 days.

the total body and to (5 mrem to any organ and, 6. Doses due to liquid releases to unrestricted areas shall

b. During any calendar be projected at lease once year to (3 mrem to the per 31 days, in accordance total body and (10 with the ODCM.

mrem to any organ.

4. 'If the limits specified in 3.8.A.3 a & b above are exceeded, prepare and sub-mit Special Report pursuant to Section 6.7.2.B.2.

5. The liquid radwaste system shall be used to reduce the radioactive .

materials in liquid wastes prior to their discharge from the site when the projected monthly dose would exceed 0.06 mrem to the total body or 0.21 mrem to any organ per unit (see Figure 4.8.1).

6. With radioactive liquid waste being discharged for more than 31 days without treatment and when the projected dose is in 282 i

I,IMITING CONDITIONS FOR OPERATION St.7NEILLANCE REOUIRESTS 3.8 RADIOACTIVE MATERIALS 4.8 RADI0 ACTIVE MATERIALS excess of limits specified in 3.8.A.5, prepare and submit the Special Report pursuant to Section 6.7.2.B.2.

7. The maximum activity to be contained in one liquid radwaste tank or temporary radwaste storage tank that can be discharged directly to the environs shall not exceed 10 curies excluding tritium and dissolved / entrained noble gas.

8. With radioactive liquid waste exceeding 3.8.A.7 limits, without delay suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the activity of tank contents to within the limit.

Y 28)

- LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8 RADI0 ACTIVE MATERIALS , 4.8 RADIOACTIVE MATERIALS

3. Airborne Effluents 3. Airborne Effluents

1. The dose rate at any 1. The gross $/y and time in the unrestricted particulate activity of areas (see Figure 4.8-1) gaseous wastes released due to radioactivity to the environment released in gaseous shall be monitored and effluents from the site recorded.

shall be limited to the following values: a. For effluent streams having continuous

a. The dose rate limit monitoring capability, for noble gases shall the activity shall be (500 mrem /yr to be monitored and the total body and flow rate evaluated (3000 mrem /yr to the and recorded to enable skin, and release rates of gross radioactivity to be

b. The dose rate limit determined at least for I-131, H-3, and once per shift using particulates with instruments specified greater than eight in table 3.2.E.

day half-lives shall be<1500 mrem /yr to b. For' effluent streams any organ, without continuous monitoring capability,

2. If the limits of 3.8.3.1 the activity shall be are exceeded, appropriate monitored and recorded corrective action shall be and the releases immediately initiat:d to through these streams bring the releases within controlled to within limits. Provide prompt ,

the licits specified notification to the NRC in 3.8.B.

pursuant to section 6.7.2.A. 2. Radioactive gaseous waste sampling and activity analysis shall be performed in accordance

~

with Table 4.8.B.

284.

I,IMITING CONDITIONS FOR OPERATION SURVEII1ANCE REQUIFJMENTS

3. The air dose in 3. Cumulative quarterly unrestricted areas (see and yearly dose Figure 4.8-1) due to contributions from noble gases released in gaseous releases shall gaseous effluents per be deterr.ined using unit shall be limited to methods contained in the following: the ODCM at least once every 31 days.

a. During any calendar quarter, to f5 mrad 4. Doses due to gaseous for gamma radiation releases to unrestricted and {10 mrad for areas shall be projected beta radiation; in accordance with the ODCM at least once

b. During any calendar per 31 days.

year, to {10 mrad for gamma radiation and 5. Samples of offgas system (20 mrad for beta effluents shall be Eadiation. analyzed an least weekly to determine the

4. If the calculated air identity and quantity dose exceeds the limits of the principal specified in 3.8.B.3 radionuclides being above, prepare and submit released.

a special report pursuant to section 6.7.2.B.2. -

6. In accordance with the methods and procedures of

5. The dose to a member of the ODCM a release rate the public from radio- limit methodology for iodines, radioactive noble gases in gaseous materials in particulate effluents shall be used form, and radionuelides ensure compliance with other than noble gases the limits specified in with half lives greater specification 3.8.B.1.

than 8 days in gaseous effluent released per ,

unit to unrestricted areas (see Figure 4.8-1) shall be limited to the -

following:

a. To any organ during any calendar quarter

! to$7.5mren; _

b. To any organ during any calendar year to {15 mres; 4

285 1

-,, - e-c---e-evvy--r - y-+-- -------,-w e w - - - - - - - ----r y --

1 LIMITING CONDITIONS FOR OPERATION _ SURVEILLANCE REQUIPIMENTS L . <

6. If the calculated doses i

1 exceed the limits of 3.8.B.5 above, prepare -

and submit a special 3 report pursuant to l

section 6.7.2.B.2.

7. During operation above 1 25% power the discharge of the SJAE must be '

i routed through the , ,

charcoal adsorbers when' ,

the projected gaseous ;

effluent releases to ,

unrestricted aream (see Figure 4.S-1; when averaged over 31 days would exceed 0.2 mrad '

for gamma radiation ,

! and 0.4 mrad for beta s e,,

radiation per unit. ,

8. With gaseous vastes being discharged for more than N s

31 days without treatment and when the projected ~4 s, dose is in excess of the limits of 3.S.B.7 above, ..

prepare and submit a g special report pursuant ,.

to section 6.7.2.B.2. s s g % \

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~,,

LIMITING CONDITIONS F R OPERATION SURVEILLANCE REQUIREMENTS

% 3.S.C Radioactive Effluents - Dose 4.8.C Radioactive Effluents - Dose

-- l . The dose or dose commitment 1. Cumulative dose contributions to,a real individual from all from liquid and gaseous uranium fuel cycle sources effluents shall be determined is limited to { 25 m:em to in accordance with specifi-the total body or any organ cations 3.8.A.3, 3.8.B.3, (except the thyroid, which and 3.8.B.5 and the methods is limited to { 75 mrem) in the ODCM.

over a period of one ,

calendar year. i

2. .Vith the calculated dose from the release of

' radioactive materials in liquid or gaseous effluents exceeding twice the limits of specification 3.8.A.3, 3.8.B.3, or 3.8.B.5, prepare and submit a

- _ . Special Report to the Commission pursuant to ,.

specification 6.7.2.B.2 and limit the subsequent releases such that the limits of 3.8.C.1 are not exceeded.

3.8.D Mechanical Vacuum Pump 4.8.D Mechanical Vacuum Pitmp

1. The mechaaical vacuum pump At least once during each shall be capable of being operating cycle verify automatic automatically isolated and securing and isolation of the secured on a signal or high mechanical vacuum pump.

radioactivity in the steam lines whenever the main steam isolation valves are i open.

s N. 2. If the limits of 3.8.C.1 are

, 'not met, the vacuum pump shall be isolated.

287 I ,

' i _ _ _ _ _ _ _ _

s.

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIPIMENTS 3,8 RADIOACTIVE MATERIALS 4.8.C RADI0 ACTIVE MATERIALS I. Miscellaneous Radioactive E. Miscellaneous Radioactive Materials Sources Materials Sources

1. Source Leakare Test 1. Surveillance Reouirement 7

Each sealed source Tests for leakage and/or containing radioactive contamination shall be

/ material either in performed by the licensee excess of 100 microcuries or by other persons of beta and/or gamma specifically authorized emitting material or by the Coc=ission or an 5 microcuries of alpha agreement State, as emitting material shall follows:

be free of > 0.005 microcurie of removable a. Sources In Use contamination. Each sealed source with Each sealed source,

.' removable contamination excluding startup in excess of the above sources and flux limit shall be detectors previously immediately withdrawn subjected to core from use and (a) either flux, containing decontaminated and radioactive material, repaired, or (b) disposed other than Hydrogen 3, of in accordance with with a half-life Commission regulations. greater than thirty days and in any form other than gas shall 2

't be tested for leakage

. and/or contamination

,' jtz at least once per six months. The leakage

test shall be capable 1 of detecting the presence of 0.005 microcurie of radioactive material on the test sample.

l 1

0 288

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4

LIMITING CONDITIONS FOR JPERATION SURVEILLANCE RFQUIREMENTS 4.8.E Miscellaneous Radioactive Materials Sources

1. Surveillance Recuirements

b. Stored Sources Not In Use Each sealed source and fission detector not previously subjected to core flux shall be tested prior to use or transfer to another licensee unless tested within the previous six months.

Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to use.

c. Startup Sources and Fission Detectors Each sealed startup source and fission detector shall be tested prior to being subjected to core flux and following repair or maintenance to the source.

2. Reports A report shall be prepared and submitted to the Commission on an annual basis if sealed sources or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcuries of removable contamins. tion.

e 289

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS i

3.8 RADIOACTIVE MATERIALS 4.8 RADIOACTIVE MATERIALS F. Solid Radwaste F. Solid Radwaste

1. The solid radwaste system 1. The Process Control shall be operated in Program shall accordance with a process include surveillance control program, for the checks necessary to solidification and packag. demonstrate compliance ing of wet radioactive with 3.8.F.1.

wastes to ensure meeting the requirements of 10 CFR 20 and 10 CFR 71 and burial ground -

requirements prior to shipment of radioactive wastes from the site.

2. With the packaging requirenents of 10 CFR 20 or burial ground require-ments and/or 10 CFR 71 not satisfied, suspend shipments of defectively packaged solid radioactive wastes from the site.

290

I TABLE 4.8.A

^

RADI0 ACTIVE LIQUID WASTE SAHPLING AND ANALYSIS PROGRAM SYSTEM DESIGN CAPABILITY HINIHUM ANALYSIS TYPE OF ACTIVITY LOWER LIMIT OF DETECTION LIQUID RELEASE SAMPLING (pC1/el)

FREQUENCY ANALYSIS (LLD)

TYPE FREQUENCY Each Batch Each Batch Prior Principayanma SE-7( }

BatchWag to Release Emitters Heleases One Batch Honthly Dissolved anel ~

Entrained Cases (5) per Month Tritium 1 E-5 Monthly Honthl:-

Proportional I E-7 Composite Gross a (2)

Sr-89, Sr-90 5 E-8 Quarterly Proportional Quarterly 1 E-6 Composite Fe-55 (2) 290A

TABLE NOTATION - TABLE 4.8.A (1) A batch release is the discharge of liquid wastes of a discrete volume.

The discharge shall be throughly mixed prior to sampling.

_ (2) A proportional composite sample is one in which the quantity of liquid

. sampled is, proportional to the quantity of liquid waste discharged from the plant and is representative of the liquid discharged.

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

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

4.66 s

"* E - V 1.22 x 10*

bY exp (-Aat)

Where:

LLD is the "a priori" lower Ibnit of detection as defined above (as microcuries per unit mass or volume),

~

i s is the standard deviation of the background counting rate or of h

ue 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 10s is the number of disintegrations per minute per microcurie, Y is the :ractional radiochemical yield (when applicable),

A is the radioactive decay constant for the particular radionuclide, and At for plant effluents is the elapsed time between the cidpoint of

, sample collection and time of counting.

Typical values of E, V, Y, and at should be used in the 4

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 (after the fact) limit for a particular measurement.

LLD applicability begins within six months after the Nuclear Data or 1 equivalent systems becomes operational. ,

290B

(4) The principal gre:ma emitters for which the LLD specification will apply are exclusively the following radionuclides: Zn65, Co60, Cs137, Mc54, Cc58,-Cs134,.Ce141, CE144, Mo99, and Fe59 for liquid releases. This list does not mean that only these nuclides are to be detected and reported. Other nucludes detected within a ~95*/. confidence level, together with the above nuclides, shall also be identified and reported as being present. Nuclides which are below the LLD for the analysis may not be reported as being present at the LLD Level for that nuclide.

I-131 shall have a LLD of 31 E-6.

(5) Gam:ca Emitters Only.

e e

290C

TAllLE 4.8.B '

RADI0 ACTIVE CAdEOUS WASTE SAMPl.ING AND ANALYSIS PROGRAM I

I SYSTEM DESIGN CAPABILITY SAMPLING HINIMUN ANALYSIS TYPE OF ACTIVITY LOWER LIMIT OF DETECTION CASEOUS RELEASE TYPE FREQUENCY ANALYSIS (pCi/ml)

FREQUENCY A. Containment Prior to Each Prior to Each Purge Principa Emitters {3 gamma IE b Purge Purge Grab Sample II-3 IE-6 H. 1. Stack Principal IE-4(

g Emitters (gjuuna

2. Building Grab Sample Monthly 11 - 3 IE-6 Ventilation

a. Reactor /

Turbine

b. Turbine Exhaust

c. Radwaste ,

C. Continuous I-131 IE-12( )

All Release Charcogg) Sample Points Listed Sampler Weekly in 11. Above Continuous ParticyjgteSample Principa IE-l bs Weekly Enuniters {3 gamma and Sampler I-131 IE-12(2)

Continuous Composite Particulate Gross Alpha IE-11 Sampler Sample Monthly l Composite Particulate Sr-89, Sr 90 1E-11 Coutinuous Sampler Sampic Quarterly 290D

TAST.I NOTATION - 4.8.B (1) _ The LID is defined, for the purposes of these specifications at the smallest concentration of radioactive material in a sample that will yield a new count (above system background) that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

4.66 sb E V 2.22 x 106 Y exp (-Aat)

Where:

LLD is the "a priori" lower limit of detection as defined above (as microcuries per unit mass or volume),

s is the standard deviation of the background counting rate or of b

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

A is the radioactive decay constant for the particular radionuclide, 3 and

. At for plant effluents is the elapsed time between the midpoint of sample collection and time of counting.

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 (after the fact) limit for a particular measurement.

LLD applicability begins within six months after the Nuclear Data or equivalent system becomes operational.

(2) When samples are taken more often that that shown, the minimum detectable concentrations can be correspondingly higher.

290E

(3) The principal samma e=itters for which the LLD specification will apply are exclusively the following radionuclides: Kr-87, Kr-SS, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide.

(4) Analysis shall also be performed if the radiation monitor alarm exceeds the setpoint value.

1 I

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'3.8- BASES Radioactive waste release levels to unrestricted areas should be kept "as low as reasonably achievable" and are not to exceed the concentration limits specified in 10 CFR Part 20. At the same time, these specifications' permit the flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided a dependable source of power under unusual operating conditions which may temporarily result in releases higher than design objectives but still within the concentration limits specified in 10 CFR Part 20. It is expected that by using this operational flexibility and exerting every effort to keep levels of radioactive materials released as low'as reasonably achievable in accordance with criteria established in 10 CFR 50' Appendix I, the annual releases will not exceed a small fraction of the annual average concentration limits specified in 10 CFR Part.20.

3.8.A LIQUID EFFLUENTS

-Specification 3.8.A.1 is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendir B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within (1) the Section 11.A design objectives of Appendix I, 10 CFR Part 50, to an individual and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

Specification 3.8.A.3 is provided to implement the dose requirements of Sections II.A, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth the Section 11.A of Appendix I.

f f Specification 3.8.A.4 action statements provides the required operating j flexibility and at the same time implement the guides set forth in l .Section IV.A of Appendix I to assure that the releases of radioactive material l in liquid effluents will be kept "as low as is reasonably achievable". Also, for fresh water sites with drinhing water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculations in the ODCM implement the requirements in Section III.A of Appendix"I that conformance with the guides of Appendix I be shown by l

l i

291 9 *

- _ , _, __ -.__ ,_ ___ . _ _ , ... _ . , ~ . _ . . _ . _ . . . . _ . , _ . , _ _ _ _ . . . _ , . . . . - , , _ -- .,,

3.8.A LIQUID EFFIIENTS (cont'd) emiculational 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 will be censistent 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 Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I,"

April 1977. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.113.

Specification 3.8.A.5 requires that the appropriate portions of the liquid radwaste treatment system be used when specified. This provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and design objective Section II.D of Appendix I to 10 CFR Part 50. -The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the guide set forth in Section II. A of Appendix I,10 CFR Part 50, for liquid effluents.

Specification 3.8.A.6 requires submittal of a special report if the limiting values of Specification 3.8.A.5 are exceeded and unexpected failures of non-redundant radwaste processing equipment halt waste treatment.

3.8.B' AIRBORNE EFFLUENTS Specification 3.8.B.1 is provided to ensure that the dose rate at anytime at the exclusion boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas.

The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a member of the public in an unrestricted area, either within or outside the exclusion area boundary, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CIR Part 20 (10 CFR Part 20.106(b)). For members of the public who may at times be within the exclusion area boundary, the occupancy of the members of the public will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the exclusion area boundary.

291A

_ _ _ _ __ ._ _ ~ . _ __ _ _ __. _.____-- _ . _ , - . , -._._, _ _ _,

e 3.S.B AIR 30RNE EFFLUENTS (cont'd)

The.specified release rate limits restrict, at all times, the corresponding i gamma and beta dose rates above background to an individual at or beyond the exclusion area boundary to { 500 mrem / year to the total body or to f 3000 mrem / year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to an infant via.the cow-milk-infant pathway to { 1500 mrem / year for the nearest cow to the plant.

Specification 3.8.B.2 requires that appropriate correction action (s) be taken to reduce gaseous effluent releases if the limits of 3.8.B.1 are exceeded.

Specification 3.8.B.5 dose limits is provided to implement the requirements of Section II.C, III. A, and IV of Appendix I,10 CFR Part 50. The limiting conditions for operation are the guides set forth in Section II.C of Appendix I.

Specification 3.8.B.6 action statement provides the required operating flexibility and at the same time Dnplement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Secticn III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methods used fer calculating the doses due to the actual release rates of the subject materials are required to be consistent with the methodology provided in Regulatory Guide 1.109,

" Calculating of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,"

Revision I, October 1977, NUREG/CR-1004, "A Statistical Analysis of Selected i Parameters for Predicting Food Chain Transport and Internal Dose of

+

Radionuclides", October 1979, and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases.from Light-Water-Cooled Reactors," Revision 1, July 1977.

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

Spe'cification 3.8.B.6 action statement requires that a special report be prepared and submitted to explain violations of the limiting doses contained in Specification 3.8.B.5.

291B i

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_ , . , _ . , . - , n - ,_. n , ., , . - - - _ . . - -, , , , , - , , . - - _ , --,n ..,- , , - .

AIRBORNE EFFLUENTS Specification 3 8.B.7 requires that the offgas charcoal adsorber beds be used when specified to treat gaseous effluents prior to their release to the environment. This provides reasonable assurance that the releases of radioacive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50 36a, General Design Criterion 60 of Appendix A to 10 CFR part 50, and design objective Section II.D of Appendix I to 10 CFR Part 50.

The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous effluents.

Specification 3 8.B.8 requires that a special report be prepared and submitted to explain reasons for any failure to comply with Specification 3 8.B.7 Specification 3 8.B.3 is provided to implement the requirements of Section II.B, III.A. and IV.A of Appencix I,10 CFR Part 50. The Limiting Condition for Operation implements the guide set forth in Section II.C of Appendix I.

Specification 3 8.B.4 action statement provides the required operating flexibility and at the same time implement the guides set forth in Sec-tion IV.A of Appendix I to assure that the releases of radioative material in gaseous effluents will be kept "as low as is reasonably achieveable."

The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of a member of the public through the appropriate pathways is unlikely to be substantially underestimated. The dose calculations estab-lished in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents will be 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, NUREG/R-1004, "A Statistical Analysis of Selected Parameters for Predicting Food Chain Transport and Internal Dose of Radio-nuclides," October 1979 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977 The ODCM equations provided for determining the air doses at the exclusion area boundary will be based upon the historical average atmospheric conditions.

NUREG-0133 provides methods for dose calculations consistent with Regula-tory Guides 1.109 and 1.111. Specifications 3 8.B.4 requires that a special report be prepared and submitted to explain violations of the limiting doses contained in Specification 3 8.B.3 4.8.A and 4.8.B BASES The surveillance requirements given under Specification 4.8.A and 4.8.B provide assurance that liquid and gaseous wastes are properly controlled and monitored during any release of radioactive materials in the liquid and 291C

~ . .- . . . . . - . - . -

4.8.A and 4.8.B BASES (cont'd)-

gaseous effluents.L These surveillance requirements provide the data for the licensee and the Commission to evaluate the' station's, performance '

relative to radioactive wastes released to the environment. Reports on the quantities of radioactive materials released in effluents shall be furnished to the Comission on the basis of Section 6 of these technical specifications. On the basis of such reports and any additional informa-tion the Comission may obtain from the licensee or others, the Comission may from time to time require the licensee to take such actions as the Commission deems-appropriate.-

3.8.C and 4.8'.C BASES This specification is provided to awt the dose limitations of 40 TR 190.

The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites con-taining up to four reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. . The l

Special Report will c'escribe a course of action which should result in the limitation of dose to a member of the public for the calendar year to be within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assused that the dose comitment 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 five miles must be considered.

3.8.D and 4.8.D ECHANICAL VACUUM PUMP

~ The purpose of isolating the mechanical vacuum pump line is to limit the release of activity from the main condenser. During an accident, fission-products would be transported from the reactor through the main steam lines to the condenser. The fission product radioactivity would be sensed by the main. steam line radioactivity monitors which initiate isolation.

3.'8.E and 4.8.E BASES .

The limitations on removable contamination for sources requiring leak-l testing, including alpha emitters, based on 10 CFR 70 39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allcwable intake values. . Sealed sources are classified into three groups according to their use, with surveillance requirements comensurate with the proba-bility of damage to a source in that group. Those sources which are fre-quently handled are required to be tested more often than those which are 4 not.~ Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron j

measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

291D i

-r- ,n,-w- . , . e,,-, , ,, a w---%-wwy-,-e ,==w %- , y- ,m.-,.y-w -----,ey .-w--,,7-,,*,s--%,,-+y-,.v-n-. y ,--m -w- +me- =yyw

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.13 RADIOLOGICAL ENVIRONMENTAL MONITORHG 4.13 RADIOLOGICAL ENVIRONMENTAL MONITORING A. Monf toring Program Monitorine Program A.

1. The radiological environ-The radiological mental monitoring **"i" ""**E*1 * *it #i"8 program shall be conducted sa=ples shall be collected as specified in Table 3.13.A.

pursuant to Table 3.13.A frem the locations given in the table and figure in the ODCM

2. With the radiological and shall be analyzed environmental monitoring program not being Pursuant to the requirement conducted as specified f Tables 3.13.A and the detection capabilities in Table 3.13.A, in lieu of a LER, prepare required by Table 3.13.B.

and submit to the Commission, in the Annual Radiological Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

3. With the level of radioactivity in an environmental sampling medium exceeding the reporting levels of Table 3.13.C when averaged over any calendar quarter, in lieu of a LER, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter, a report which identifies the cause(s) for exceeding the limit (s) and defines the corrective action to be taken to reduce '

radioactive effluents so that the potential annual dose to a member of the public is less than the calendar year limits of .

329A

LIMITING CONDITIONS FOR OPERATION SUPNEIL'.ANCE REQUIPIMENTS specification 3.8.A.3, 3.8.B.3, and 3.8.B.S.

When more than one of the radionuclides in Table 3.13.C are detected in the sampling medium, this report shall be submitted if:

Conc (1)

Limit (1) + Conc (2) Limit (27***1

4. When radionuclides other than those in Table 3.13.C are detected and are result of plant effluents, the report in 3.13.A.3 shall be submitted if the potential annual dose to a merber of the public is equal to or greater than the calendar year limits of specification 3.8.A.3, 3.8.3.3, and 3.8.B.S.

5. The report in specifi-cation 3.13.A.3 is not required if the measured level of radioactivity was not the result of plant effluents; 1 however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

6. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 3.13.A, identify locations for obtaining /

replacement samples, if available, and add them to the radiological 329B

LIMITING CONDITIONS FOR OPERATION SUPIEII. LANCE PIQUIPIMEh7S 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 LER identify the cause of the unavail-ability of samples and identify the new loca-tion (s), if available, for obtaining replacement samples in the next Annual Radiological Environmental Operating Report and also include a revised figure (s) and table (s) for the ODCM reflecting the new locations.

E. Land Use Census

1. A land use census shall be conducted and shall identify the location of the-nearest milk animal, the nearert residence and the nearest garden

of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meterological sectors within a distance of five miles. (For elevated releases as defined in Regulatory Guide 1.111, Revision 1, July 1977, the land use census shall also identify the locations of all milk animals and all gardens of greater than 500 square feet producing fresh leafy vegetables in each.of the 16 meteorological sectors within a distance of three miles.)

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

329C

LIMITING CONDITIONS FOR OPERATIOT SURVEILLANCE REOLUREMENTS

2. With a land use census ,

identifying a location (s) which yields a calculated dose or dose commitment greater than the maximum value currently being

.. calculated in specifi-cation 4.8.B.4, in lieu cf-a LER, identify the new locations in the next Annual Radiclogical Environmental Operating

! Report.

3. With a land use census identifying a location (s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent 3

greater than at a location from which samples are currently being obtained in accordance with i

specification 3.13.A, add the new location (s) to the radiological

-environmental monitoring programs within 30 days i

if the owner consents.

The sampling location (s),

excluding the control station location, having j

the lowest calculated dose or dose commitment (s)

(via the same exposure pathway) may be deleted from this monitoring program after October 31 of the year in which this land use census-was conducted. In lieu

, of a LER, identify the

' new location (s) in the next Annual Radiological Environmental Operating Report and provide a f revised figure (s) and I table for the ODCM reflecting the new location (s).

,329D

I LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

4. The land use census shall be conducted at least once per calendar year between the dates of April 1 and October 1 using the following techniques: *

a. Within a 2 mile radius from the plant or within the 15 mrem per year isodose line, whichever is larger, enumeration by a door-to-door or equivalent counting technique.

b. Within a 5 mile radius from the plant, enumeration by using appropriate techniques such as door-to-door survey, mail survey, tele-phone survey, aerial survey, or infor-mation from local agricultural authorities or other reliable sources.

C. Interlaboratory Comparison .

Program

1. Analyses shall be performed on radioactive materials supplied as part of an 2nterlaboratory Comparison l Program which has been approved by the Commission.

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

329E L

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

3. A summary of the results obtained as part of the above required Inte laboratory Comparison Program and in accordance with the ODCM (or -

participants in the EPA .

cross check program shall provide the EPA program code designation for the unit) shall be included in the Annual Radiological Environmental Operating Report.

D. Deviations From Samoline Schedule Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability or malfunction of automatic sampling equipment. If the latter, every effort shall be made to complete corrective action prior to the end of the next sampling period.

All deviations from the sampling schedule shall be reported in the Annual Radiological Environmental Operating Report.

E. Exemition The provisions of Specification 1.0.C are not applicable.

329F

TABLE 3.13.A RADIOLOGICAL ENVIRONMENTAL HONITORING PROGRAM

Number of Samples d

Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample' Locations

Collection Frequency of Analysis

1. AIRBORNE Hadiolodine and Hinimum of 5 locations Continuous operation of ,

Radiciodine canister.

Particulates sampler with sample Analyze at least once j collection as required by per 7 days for I-131.

dust loading but at least ,

once per 7 days. particulate sampler.

Analyze for gross beta radioactivity } 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following filter change.

Perform gamma isotopic i analysis on each sample when gross beta activity 4 is greater than 10 times the average of control samples. Perform gamma isotopic analysis on composite (by location) sample at least once per 92 days.

2. DIRECT RADIATION At least 40 locations At least once per 92 days. Gamma dose. At least once with } 2 dosimeters per 92 days.

at each location.

"Samtile locations are given in the ODCH.

4 329G 1

i

TAllLE 3.13. A (Continued)

RADIOLOGICAL ENVIRONMENTAL HONITORING PHOGRAN Number of Samples i Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations

CollectJon Frequency of Analysis

. 3. WATER 110RNE I

a. Surface 2 locations Composite' sample collected Gamma isotopic analysis over a period of { 31 days. of each composite sample.

Tritium analysis of com-posite sample at least once per 92 days.

h. Drinking Hinimum of I downstream Composite' sample collected Gross beta and gamma location, or all water over a period of { 31 days, isotopic analysis of

. supplies within 10 miles each composite sample.

downstream which are Tritium analysis of taken from the Tennessee composite sample at least River. once per 92 days.

l c. Sediment Minimum of I location At least once per 184 days. Gamma isotopic analysis of each sample.

d. Ground'I i

i "S.smple locations are shown in the ODCil. .

I h

, composite samples shall be collected by collecting an aliquot at intervals not exceeding 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

" Composite samples shall be collected over a period of { 14 days for 1311 if drinking water is obtained within 3 miles downstream of the plant.

d Groun.1 water movement in the area has been determined to be from the plant site toward the Tennessee River.

Since no drinking water wells exist lietween the plant and the river, ground water will not be monitored. .

32911

1 TABf.E 3.13.A (Centinurd)

RADIOLOGICAL ENVIRONMENTAT. HONITORING PROGRAN

- Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample I.ocations* . Collection Frequency of Analysis

4. INGESTION

a. Hilk 3 locations At.least once per 15 days I-131 analysis of each when animals are on pasture; sample. Gasuna isotopic at least once per 31 days analysis at least once at other times. per 31 days,

b. Fish 2 samples One sample in season, or at Ganuna isotopic analysis

' least once per 184 days if on edible portions.

not seasonal. One sample .

of conunercial and game species.

3

c. Food Products

2 locations At Icast once per year Gamma isotopic analysis at time of harvest. . on edible portion.

t l

t 4

" Sample locations are sliown in the OI)Cll.

"Since water from the Tennessee R!ver in the immediate area downstream is not used for irrigation ' purposes, the sampling of food products (primarily broad leaf vegetation) is not required unless milk sampling is not performed.

3291 i

i

TABLE 3.13.B NAXINilN VAttlES FOR Tile LOWER I.INITS OF DETECTION (LLD)*'"

Airborne Particulate or Gas Fisti Hilk Food Products Sediment Water (pci/kg, dry)

Analysis (pci/1) (pci/m3 ) (pCi/kg, wet) (pci/1) (pci/kg, wet) i 1 x 102 N.A. N.A. N.A. N.A.

gross tieta 4 N.A. N.A. N.A. N.A.

11 - 3 2000 N.A.

N.A. 130 N.A. N.A.

N.A.

Hn-54 15 N.A. 260 N.A. N.A. N.A.

Fe-59 30 N.A. 130 N.A. N.A. N.A.

Co-58, 60 15 H.A 260 N.A. N.A. -N.A.

Zn-65 30 N.A. N.A. N.A. N.A. N.A.

Zr-95 30 N.A. N.A. N.A. N.A. N.A.

Nie-95 15 60 N.A.

I-131 I"I 7 x 10 2 N.A. 1 2 60 150 Cs-134 15 5 x 10 130 15 80 180 l Cs-137 18 6 x 102 150 18 N.A. N.A. 60 N.A. N.A. I Ila-140 60 N.A. N.A. 15 N.A. N.A.

I.a-140 15 329J I

TABLE 3.13.B (Continued)

TABLE NOTATION

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

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

4.66 s b E V 2.22 Y exp (-Aat)

'Where:

LLD is the "a priori" lower limit of detection as defined above-(as picoeurie per unit mass or volume),

s is the standard deviation of the background counting rate or of 3

t5e counting rate of a blank sample as appropriate (as counts per

! minute),

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

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

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

A is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environmental l samples, not plant effluent samples).

l l It should be recogni=ed that the LLD is defined as a a priori (before the l

fact) limit representing the capability of a measurement system and not as l

a,posteriori (after the fact) limit for a particular measurement.

l l

t 329K

. ~, . -_ _

~

s s

J s N t

~ .

TA212 3.13.B (Continued)

,. TABII K0TATION b.- The LLD for-analysis of drinking water and surface water samplis shall be performed by' gamma spectroscopy at approximately 15 pCi/L. If' levels greater than 15 pC1/L are identified in surface water samples downstream from the plant, or in ,the event of an unanticipated release of I-131, drinking water samples will be analyzed at an LI3 of 1.0 pCi/L for I-131.

c. Other peaks which are measurable and identifiable, together with the radionuclides in Table 3.13.C, shall be identified and reported.

4.

f s.

' g'l d

=

s' s

t 1 s s i ,

x ..

(1 ,

( 4 t _.

~ ~

. -g t s N,

, ,- 4 5

.s . ~

329L T

\ -s

._ A.________.__._._.- _

s- ff e . 1 ?-

, -r [ -

TABLE 3.13.C /

, ,fr. , s ,

?

l REPORTING LEVELS FOR RAD 10ACTWIIT CONCENTRATIONS IN ENT'IRONilENTAL SAHPLES ' \

~ , f f ~ \

'e

,- r,.., .

' - j

, * */ ', - i f , , , Reporting Levelsi

'r /

p

j Water r Airborn'e Pn(aculate Fish , Hilk Food Products Analysis (pCi/1T or Gases ('hCi/m ) 3 (pCi/Kg, wet)I fpci/1) (pci/Kg, w5t) '

~ '

, m ,',,, ,

, 11 - 3 -

2 x 10 4f*) N.A. N.A. N.A. N . A. ,.

f v'

Hn-54 1 x 10 3 N.A. 3 x 10 4 N.A. N.A. -

~ l 1 x 30 4 4 x 102 y,4, - N.A. N .'A .

Fe-59 N.A. 3 x 10 4 N.A. N.A. I Co-58 1 x 103 1 x 10 4 N.A. N.A.

f Co-60 3 x 10 2 N.A.

l~ .

N.A. N.A.

' Zn-65 3 x 10 2 N.A. , 2 x 10 4 -

N.A. N.A. N.A. N.A.

Zr-Nb-95 4 x 102 2 0.9 j N.A. 3 1 x 10 2 1-131 I x 10 3 60 1 x 10 3 Cs-134 30 10 20 2 x 10 3 70 2 x 10 8 Cs-137 50 Ba-La-140 2 x 10 2 N.A. N.A. 3 x 10 2 N'.A.

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

329H

~

BASES Monitorine Program -

1.

The r$31ological monitoring program required by this specification provides measurements of radiation andLof radioactive materials in those exposure pathways and for those radionuefides, which lead to the highest cotential_ radiation exposures of individuals "esulting from the station operation. This monitoring program thereby supplements the radiological

" effluent monitoring program by verifying that the measureable #

-. concentrations of radioactive materials and levels of radiation are not

? ' higher than expected on the basis of the effluent measurements and modeling of the evironmental exposure pathways.

The detection capabilities required by Table 313.B are state-of-the-art for routine environmental measurements in industrial laboratories. It

'y should be recognized that the LLD is defined as an a oriori (before

< the fact) limit representing the capability of the measurement system and not as an a posteriori (after the fact) licit for a particular measurement. Analyses shall be performed in such a manner that the

? stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of f interfering nuclides, or other uncontrollable circumstances may render

, these LLLs unachievable. In suah cases, the contributing factors will

be identified and described in the Annual Radiological Environmental Operating Report.

s

2. Land Use Census L

i s This specification is provided to ensure that changes in the use of unrestriced areas are identified and that modifications to the monitoring program are made if required by the results of this census.

The best survey information from the door-to-door, mail, telephone, aerial or consulting with local agricultural authorities shall be used.

This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 500 sq'uare feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of i this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetation assumed in Regulatory Guide 1.109 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 2kg/ square meter.

3 Interlaboratory Comparison Program The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision

.and accuracy of the measurements of radiorctive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

329N l

L

6.0 ADMINISTRATIVE CONTROLS

h. The radiological environmental monitoring program and the results thereof at least once per 12 months.

1. The performance of activities required by the Quality Assurance Program to meet the criteria of Regulatory Guide 4.15, December 1977 or Regulatory Guide 1.21, Rev. 1, 1974 and Regulatory Guide 4.1, 1975 at least once per 12 months.

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

n. The Process Control Program and implementing procedures for solidification of wet radioactive wastes at least once per 24 months.

9. AUTHORITY The NSRB shall report to and advise the Manager of Power on those areas of responsibility specified in Sections 6.2.A.7 and 6.2.A.8.

10. RECORDS Records of NSRB activities shall be prepared, approved and distributed as indicated below:

a. Minutes of each NSRB meeting shall be prepared, approved and forwarded to the Manager of Power within 14 days following each meeting.

b. Reports of reviews encompassed by Section 6.2.A.7 above, shall be prepared, approved and forwarded to the Manager of Power within 14 days following completion of the review.

c. Audit reports encompassed by Section 6.2.A.8 above, shall be forwarded to the Manager of Power and to the management positions responsible for the areas audited within 30 days after completion of the audit.

s 334A

6.0 ' ADMINIS""RATI\T CONTROI.S

j. Review adequacy of empicyee training program and i recommend change. '

k. Review adequacy of the Process Control Program and Offsite Dose Calculation Manual at least once every 24

, months.

1. Review changes to the radwaste treatment systems,

m. Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recommendation, and deposition of the corrective action to prevent recurrence to the Director, Nuclear Power and to the Nuclear Safety Review Board.

5. Authority The PORC shall be advisory to the plant sr; :rintendent.

6. Records Minutes shall be kept for all PORC meetings with copies sent to Director, Nuclear Power; Assistant Director of Nuclear Power (Operations); Chairman, NSRB.

7. Procedures Written administrative procedures for committee operation

-shall be prepared and maintained describing the method for submission and content of presentations to the committee, review and approval by members of committee actions, dissemination of minutes, agenda and scheduling of meetings.

337

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

6.0 ADMINISTRATIVE CohTROLS 6.3 Procedures A. Detailed written procedures, including applicable chechoff lists . covering items listed below shall be prepared, approved and adhered to.

1. Normal startup, operation and shutdown of the reactor and of all systems and components involving nuclear scfety of the facility.

2. Refueling operations.

3. Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms, suspected primary system leaks and abnormal reactivity changes.

4. Emergency conditions involving potential or actual release of Radioactivity.

5. Preventive or cerrective maintenance operations which could have an effect on the safety of the reactor. ,

6. Surveillance and testing requirements.

7. Radiation control procedure.

8. Radiological Emergency Plan bnplementing procedures.

9. Plant security program implementing procedures.

10. Fire protection and prevention procedures.

11. Process Control Program (PCP).

12. Offsite Dose Calculation Manual.

B. Written procedures pertaining to those items listed above shall be reviewed by PORC and approved by the plant superintendent prior to implementation. Temporary changes to a procedure may be made by a member of the plant staff knowledgeable in the area affected by the procedure except that temporary changes to those items listed above except item 5 require the additional approval of a member of the plant staff who holds a Senior Reactor Operator license on the unit affected. Such changes shall be documented and subsequently reviewed by PORC and approved by the plant superintendent.

338 l

6.0 ADMINISTRATnT CONTROLS 6.3 Precedures

-E. Process Control Pro: ram (PCP)

1. The PCP shall be approved by the Com=ission prior to implementation.

2. Changes to the PCP shall be sub=itted to the Com=ission in the semi-annual Radioactive Effluent Release Report for the period in which the change (s) was made. This sub=ittal shall contain:

a. Sufficiently detailed information to totally support the change.

b. A determination that the change did not change the overall conformance of the solidified product to existing criteria.

3. Changes to the PCP shall become effective upon review and l acceptance by PORC. '

l T.

Ouality Assurance Procedures - Effluent and Environmental l

l . Monitoring Quality Assurance procedures shall be established, implemented, and maintained for effluent and environmental l monitoring, using the guidance in Regulatory Guide 1.21, j rev. 1, June 1974 and Regulatory Guide 4.1, rev. 1, April 1975 or Regulatory Guide 4.15, Dec. 1977.

l l

340

6.0 ADMINISTRATIVE CONTROLS 6.4 Actions to be Taken in the Event of a Reoortable Occurrence in Plant Doeration (Ref. Section 6.7)

A. Any reportable occurrence shall be promptly reported to the Manager, Nuclear Production and shall be promptly reviewed by PORC. This committee shall prepare a separate report for each reportable occurrence. This report shall include an evaluation of the cause of the occurrence and recommendations for appropriate action to prevent or reduce the probability of a repetition of the occurrence.

B. Copies of all such reports shall be submitted to the Manager, Nuclear Production, the Manager of Power and the Chair =an of the NSRB for their review.

C. The plant superintendent shall notify the NRC as specified in Specification 6.7 of the circu= stances of any reportable occurrence.

6.5 ' Action to be Taken in the Event a safetv Limit is Exceeded If a safety limit is exceeded, the reactor sball be shut down and reactor operation shall not be resumed until authorized by the NRC. A prcmpt report shall be made to the Manager, Nuclear

~ Production, and the Chairman of the NSRB. A complete analysis of the circumstances leading up to and resulting from the situation, together with recommendativns to prevent 2 recurrence, shall be prepared by the PORC. This report shall be submitted to the Manager, Nuclear Productica, the Manager of Power, and the NSRB.

Notification of such occurrences will be made to the NRC by the plant superintendent within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

6.6 Station Operating Records A. Records and/or logs shall be kept in a manner convenient for review as indicated below: (Items 1-6, 5 years; Items 7-17,

duration of the operating license).

. 1. All normal plant operation including such ite=s as power level, fuel exposure,. and shutdowns

2. Principal maintenance activities

3. Reportable occurrences

4. Radioactive shipments

5. Test results, in units of microcuries, for leak tests l

performed pursuant to Specification 3.8.E i

i l

l l 346

6.0 ADMINISTRATIVE CONTROI.S

6. Record of annual physical inventory verifying accountability of sources on record

7. Chechs, inspections, tests, and calibrations of components and systems, including such diverse items as source leakage

8. Reviews of changes made to the procedures or equipment or reviews of tests and experiments to comply with 10 CFR 50.59

9. Gaseous and liquid radioactive waste released to the enirons

10. Off-site environmental monitoring surveys

11. Fuel inventories and transfers

12. Plant radiation and contamination surveys

13. Radittion exposures for all plant personnel

14. Updated, corrected, and as-build drawings of the plant

15. Reactor coolant system inservice inspection

16. Minutes of meetings of the Nuclear Safety Review Board

17. Design fatigue usage evaluation

a. Monitoring, recording, evaluating, and reporting requirements contained in 17.b, below will be met for various portions of the reactor coolant pressure boundary (RCPB) for which detailed fatigue usage evaluation per the ASHI Boiler and Pressure Vessel Code Section III was performed 1 for the conditions

-defined in the design specification. In this 1- See paragraph N-415.2, ASHI Section III, 1965 Edition.

347

'6.0 ADMINISTRATIVE CONTROLS 6.7.1 Routine Reports

b. Annual Operating Report A tabulation en an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man rem exposure according to work and job functions, e.g. ,

reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), vaste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions,

c. Monthlv Operating Report Routine reports of operating statistics and shutdown 3

experience shall be submitted on a monthly basis to the Office of Inspection and Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the appropriate Regional Office, to be submitted no later than the 15th of each month following the calendar month covered by the report. A narrative summarv of operating experience shall be submitted in the above schedule.

d. Radiological Environmental Monitoring Routine Reporting

1. Routine Annual Radiological Environmental Operating Reports covering operation of the plant during the previous calendar year shall be submitted prior to May 1 of each year.

2. The Annual Radiological Environmental Operating Reports shall include swnmaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational contro1r (as appropriate), and previous environmental surveillance reports and an assessment of the, observed impacts of the plant operation on the environment. The repo::ts shall also include the results of land use censuses required by Specification 3.13.B.I. If harmful effects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problems and a planned course of action to alleviate ;he problem.

350 L

3. The annual radiological environ = ental operating reports shall include summarized and tabulated results in the format of Regulatory G.:!de 4.8, December 1975 of all radiological environmestal samples taken during the report period. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report. !

4. The reports shall also include the following: a summary description of the radiclogical environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions from one reactor; and the results of licensee participation in the Interlaboratory Co=parison Program, required by Specification 3.13.C.

6.7.2 Reoortable Occurrences Reportable occurrences, including corrective actions and measures to prevent reoccurrence, shall be reported to the NRC.

Supplemental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplementel reports, a licensee event report shall be co=pleted and reference shall be made to the original report date.

9 351

6.0 ADMINISTpATIVE CONTp0I.S

a. Pro =ut Notification with Written Followne (9) Performance of structures, systems, or co=ponents that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analysis in the cafety analysis report or technical specifications bases; or discovery during plant life of conditions not specifically considered in the safety analysis report 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.

(10) The concentration of radioactive material in liquid effluents released to unrestricted area exceeds the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. Concentration of dissolved or entrained noble gases exceeds 2 x 10

pCi/ml totcl activity.

(11) Offsite releases of fadioactive material in gaseous effluents which exceed the following limits:

(a) The dose rate for nobles gases equals or exceeds 500 mrem /yr to the total body or 3000 mrem /yr to the skin.

(b) The dose rate of all radiciodines, for all radioactive materials in particular form, and for radionuclides other than nobles gases with half lives greater than 8 days exceeds 1500 mrem /yr to any organ.

9 i

353A

6.0 /MINISTRATIVE CONTROI.S 6.7.2 Reportable Occurrences

b. Thirtv-Day Written Reoorts. The reportable occurrences discussed below shall be the subject of written reports to the Director of the appropriate Regional office within thirty days of occurrence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.

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

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

Note: Routine surveillance testing, instrument calibration, or preventative maintenance which require system configurations as described in items 2.b. (1) and 2.b. (2) need not be reported except where test results themselves reveal a degraded mode as described above.

(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 systems.

T e T o

354

. . - - _ .. . = - - - .. .. . _ _ - -

J 1

6.0 ADMINISTRATIIT CONTROLS 6.7.2.b Thirty Day Written Reports (4) Abnormal degradation of systems other than those specified in item 2.a(3) above designed to contain radioactive material resulting from the fission process.

Note: Sealed sources or calibration sources orc not 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.

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

1. A description of the event of equipment involved.

. 2. Causc(s) for the unplanned release.

3. Actions taken to prevent recurrence.

4. Consequences of the unplanned release.

4 9

4 i

a 354A

\

.6.0 AD".IKISTRATIVE CONTROLS 6.7 Reoorting Reouirements

\

3. Unioue Reportinr Reouirements A. Radioactive Effluent Release Report A report on the radioactive discharges released from the site during the previous 6 months of operation shall be submitted to the Director of the Regional Office of Inspection and Enforcement within 60 days after January I and July 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents released and solid waste shipped from the plant as delineated in Regulatory Guide 1.21, Revision 1, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Coeled Nuclear Power Plants,"

with data summarized on a quarterly basis following the format of Appendix B thereof.

The report shall include a summary of the meteorological conditions concurrent with the release of gaseous effluents during each quarter as outlined in Regulatory Guide 1.21, Revision 1, with data summarized on a quarterly basis following the format of Appendix B thereof. Calculated offsite dose to members of the public resulting from the release of liquid and gaseous effluents and their subsequent dispersion in the river and atmosphere shall be reported as recommended in Regulatory Guide 1.21, Revision 1. The Radioactive Effluent Release Report shall includa the following information for each type of solid was- 4pped offsite during the report period (a) container (b) total curie quantity, (specify whether dete;- . ay measurement or estimate, (c) principal radionuclx. (specify whether determined by measurement or estimate), (d) sources of waste and processing employed (e.g. dewatered spent resins, compacted dry waste, etc.), (e) type of container (e.g., LSA, Type A, Type B, large. quantity), and (f) solidification agent or absorbant (e.g. concrete, urea formaldehyde, etc.).

355

6.0 ADMIh*ISTRATIVE C0h"I'ROLS B. Source Tests Results of required leak tests performed on sources if the tests reveal the presence of 0.005 microcurie or more of removable contamination.

C. Special Reports (in writing to the Director of Regional Office of Inspection and Enforcement).

1. Reports on the following areas shall be submitted as noted:

a. Secondary Containment 4.7.C Within 90 days Leak Rate Testing (5) of completion of each test.

b. Fatigue Usage Evaluation 6.6 Annual Operating Report

c. Seismic Instrumentation 3.2.J.3 Within 10 days Inoperability after 30 days of inoperability

d. Relief Valve Tailpipe 3.2.F Within 30 days Instrumentation after inoperability of thermocouple and acoustic monitor on one valve.

e. Meteorological Monitoring 3.2.I.2 Within 10 days Instrumentation after 7 days of Inoperability inoperability D. Special Report (in writing to the Director of Regional

. Office of. Inspection and Enforcement)

Data shall be retrieved from all seismic instruments actuated during a seismic event and analyzed to determine the magnitude of the vibratory ground motion. A Special Report

'shall be submitted within 10 days af ter the event describing i the magnitude, frequency spectrum, and resultant effect upon plant features important to safety. ,

E. Special Reports: Radiological Environmental Monitoring If measured levels of radioactivity in an environmental sampling medium are determined to exceed the reporting level values of Table 3.13.C when averaged over any calendar quarter sampling period, a report shall be submitted to the Commission pursuant to Specification 3.13.A.3.

356

APPENDIX B TECHNICAL SPECIFICATIONS BROWNS FERRY UNITS 1 AND 2 e

f i

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\ . _ . - _ - . ,_ . _. _

ENVIRONMENTAL TECHNICAL SPECIFICATIONS FOR BROWNS FERRY NUCLEAR PLANT TABLE OF CONTENTS Page No.

DEFINITIONS . ..................... Deleted 1.0 Deleted 2.0 LIMITING CONDITIONS FOR OPERATION . . . . . . . . . . .

2.1 The rmal Discharge Limits . . . . . . . . . . . . . Deleted 2.2 Chemical . . . . . . . . . . . . . . . . . . . . .

Deleted 2.2.1 - Makeup Water Treatment Plant Spent Demineralizer Regerants . . . . . . . . . Deleted 2.2.2 Chlorine ................. Deleted

. Deleted 30 DESIGN FEATURES AND OPERATING PRACTICES . . . . . . .

31 Chemical Usage . ................. Deleted Oils and Hazardous Materials. . . . . . . . Deleted 3 1.1 Deleted 4 3 1.2 Other Chemicals . .............

3.2 Land Management. ................- Deleted Deleted 3 2.1 Power Plant Site. . . . . . . . . . . . . .

3 2.2 Transmission Line Right-of-Way Maintenance. 1 33 Onsite Meteorological Monitoring . . . . . . . . . Deleted Deleted 4.0 ENVIRONMENTAL SURVEILLANCE. ... ...........

4.1 Ecological Surveillance. . . . . . . . . . . . . . Deleted 4.1.1 Abiotic . ...... ........... Deleted 4.1.2 B io ti c . . . . . . . . . . . . . . . . . . . Deleted 4.1 3 Special Studies . . . . . . . . . . . . . . Deleted 4.2 Radiological Environmental Monitoring Program. .. Deleted i

l l

h . . . . -

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t 50 ADMINISTRATIVB CONTROLS . . . . . ........... Deleted l

51 Res ponsibility . . . . . . . . . . . . . . . . . . Deleted 52 Organization . . . . . . . . . . . . . . . . . . . Deleted 53 Review and Audit . . . . . . ........... Deleted 54 Action to be Taken if an Environment LCO is Exceeded . ................... Deleted 5.5 Procedura. . . . . . . . . . . . . . . . . . . . . Deleted 5.6 Reporting Requirements . . . . ... . . . . . . . . Deleted 57 Environmental Records. . . . . . . . . . . .,. . . Deleted Tables . . . . . . . . . . . . . . . . . . . . . . . . . . . Deleted Figures ........................... Deleted t

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, - , -. - - , - - . ,.--w. e .-, ,,,---vr- -, , - - , --- , , en,e ,w -,>.m,-e, , . , - - -- , , -

3 2.2 Transmission Line Right-of-Way Maintenance Objective The sole purpose of this section is to provide reporting requirements (to USNRC) on herbicide usage, if any, for purposes of right-of-way maintenance regarding only those transmissicn lines under USNRC's jurisdiction for the Browns Ferry Nuclear Plant.

Specification A statement as to vhether or not herbicides have been used in maintaining rights-of-way for those transmission lines associated with the Browns Ferry Nuclear Plant shall be provided. If herbicides have been used, a description of the types, volumes, concentrations, manners and frequencies of application, and miles or rights-of-way that have been treated shall be included.

Reporting Requirements Information as specified above shall be provided in the Annual Operating Report (Appendix A. Section 6.7.1.(b)).

Bases Vegetation growth on a transmission line right-of-way must be controlled in such a manner that it will neither interfere with safe and reliable operation of the line or impede restoration of service when outages occur.

Vegetation growth is controlled by mechanical cutting and the limited use of herbicides. Selected chemicals approved by EPA for use as herbicides are assigned (by EPA) label instructions which provide guidance on and procedures for their use.

-1

APPENDIX A TECHNICAL SPECIFICATIONS UNIT 3 l

l l

MHT 3 TAELE OF CONTENTS i

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TABLE OF CONTENTS l

Page No.

Section 1

IntrCduction . . . . . . . . . . . . . . . . . . . l 1.0 Definitions ................... 2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1.1/2.1 Fuel Cladding Integrity ............. 9 1.2/2.2 Reactor Coolant System Integrity . . . . . . . . . 26 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3 1/4.1 Reactor Protection System ............ 31 3 2/4.2 Protective Instrumentation . . . . . . . . . . . . 49 A. Primary Containment and Reactor Building Isolation Functions . . . . . . . . . . . . . 49 1

B. Core and Containment Cooling Systems -

Initiation and Control. . . . . . . . . . . . 50 C. Control Rod Block Actuation . . . . . . . . . 50 D. Radioactive Liquid Effluent Monitoring Instrumentation . . . . . . . . . . . . . . . 51 E. Drywell Leak Detection. . . . . . . . . . . . 53 F. Surveillance Instrumentation. . . . . . . . . 53 G. Control Room Isolation. . . . . . . . . . . . 53 H. Flood Protection. .'. . . . . . . . . . . . . 54 ,

I. Meteorological Monitoring Instrumentation . . 54 J. Seismic Monitoring Instrumentation. . . . . . 56 K. Radioactive Gaseous Effluent Monitoring Instrumentation . . . . . . . . . . . . . . . 56A i

_. . ._ . , _ _ - _ _ . _ . _ . _ _ _ _ ~ _ . _ _ _ _ ..__,_ _ _ _ ___ _ __ , _ __ __ _ . _ __ _ _ __

Section Page No.

3 3/4.3 Reactivity Control . . . . . . . . . . . . . . . . 118 A. Reactivity Limitations. . . . . . . . . . . . 118 B. Control Rods. . . . . . . . . . . . . . . . . 122 C. Scram Insertion Times . . . . . . . . . . . . 128 D. Reactivity Anomalies ............ 129 ,

i E. Reactivity Control ............. 129 F. Scram Discharge Volume ........... 129 3 4/4.4 Standby Liquid Control System .......... 137 A. Normal System Availability ......... 137 B. Operation with Inoperable Components .... 139 C. Sodium Pentaborate Solution . . . . . . . . . 139 3.5/4.5 Core and Containment cooling Systems . . . . . . . 146 A. Core Spray System . . . . . . . . . . . . . . 146 B. Residual Heat Removal System (RHRS)

(LPCI and Containment Cooling) ....... 149 C. RHR Service Water System and Emergency Equipment Cooling Water .

System (EECWS) ............... 155 D. Equipment Area Coolers ........... 153 E. High Pressure Coolant Injection System (HPCIS) . . . . . . . . . . . . . . . . . . . 159' F. Reactor Core Isolation Cooling System (RCICS) . . . . . . . . . . . . . . . . . . . 160 G. Automatic Depressurization System (ADS) . . . 161 l

H. Maintenance of Filled Discharge Pipe .... _163 I. Average Planar Linear Heat Generation Rate . 165 J. Linear Heat Generation Rate . . . . . . . . . 166 ii i

{

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Section Pree No. ,

F. Minimum Critical Power Ratio (MOPR) . . . . . 167 i

L. seperting Requirements . . . . . . . . . . . 167a Primary System Boundary 184 3 6/4.6 .............

A. Thermal and Pressurizstion Limitations . . . 184

3. Coolant Chemistry . . . . . . . . . . . . . . 187 C. Coolant Leakage . . . . . . . . . . . . . . . 191 E. Relief Valves . . . . . . . . . . . . . . . . 192 E. Jet Pumps . . . . . . . . . . . . . . . . . . 193 F. Recirculation Pump Operation . . . . . . . . 195 G. Structural Integrity ............ 196 H. Shock Suppressors (Snubbers) . . . . . . . . 198 Containment Systems . .............. 231 3 7/4.7 A. Primary Containment . . . . . . . . . . . . . 231 B. Standby Gas Treatment System . . . . . . . . 247 C. Secondary Containment . . . . . . . . . . . . 251 D. Primary Containment Isolation Valves . . . . 254 E. Control Room Emergency Ventilation . . . . . 256 F. Primary containment Purge System . . . . . . 258 l

l G. Containment Atmosphere Dilution System (CAD). 260 r

l H. Containment Atmosphere Monitoring (CAM) 261 l

System H2 and 02 Analyzer . . . . . . . .

Radioactive Materials .............. 299 j

3.8/4.8 i

A. Liquid Effluents .............. 299 B. Airborne Effluents ............. 302 C. ' Radioactive Effluents - Dose . . . . . . . . 305 f

iii i

t i

i s

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Section Pare No.

D. Mechanical Vacuum Pumps . . . . . . . . . . . 305 E. Miscellaneous Radioactive Materials Sources . '306 F. Solid Radwaste ............... 303 3 9/4.9 Auxiliary Electrical System ........... 316 A. Auxiliary Electrical Equipment ....... 316 B. Operation with Inoperable Equipment . . . . . 322 C. Operation in Cold Shutdown ......... 326 3.10/4.10 Core Alterations . . . . . . . . . . . . . . . . . 331 A. Refueling Interlocks ............ 331 B. Core Monitoring . . . . ........... 336 C. Spent Fuel Pool Water . . . . . . . . . . . . 337 D. Reactor Building Crane. . . . . . . . . . . . 338 E. Spent Fuel Cask . . . . . . . . . . . . . . . 339 F. Spent Fuel Cask Handling-Refueling Floor. . . 339 3 11/4.11 Fire Protection Systems .............

347 A. High Pressure Fire Protection System .... 347 B. CO2 Fire Protection System ........ 350 C. Fire Detectors ............... 352 D. Roving Fire Watch . . . . . . . . . . . . . . 353 E. Fire Protection Systems Inspection ..... 354 F. Fire Protection Organination ........ 354 G. Air Masks and Cylinders . . . . . . . . . . . 355 H. Continuous Fire Watch . . . . . . . . . . . . 355 I. Open Flames, Welding, and Burning in the Cable Spreading Room .......... 355 iv k

pnge No.

Seation 3 13/4.13 Radiological Environmental Monitoring ...... 359A A. Monitoring Program ............. 359A B. Land Use Census . . . . . . . . . . . . . . . 359C C. Interlaboratory Comparison Program ..... 359E ,

D. Deviations from Sampling Schedule . . . . . . 359F E. Exemption . . . . . . . . . . . . . . .'. . . 359F 5.0 Major Design Features .............. 360 5.1 Site Features . . . . . . . . . . . . . . . . 360 5.2 Reactor . . . . . . . . . . . . . . . . . . . 360 53 Reactor Vessel ............... 360 5.4 Containment . . . .............. 360 5.5 Fuel Storage ................ 360 5.6 Seismic Design ............... 361 6.0 Administrative controls . . . . . . . . . . . . . 362

~

6.1 Organization . . . . . . . . . . . . . . . . 362 6.2 Review and Audit . ............. 362 6.3 Procedures . . . . . . . . . . . . . . . . . 368 6.4 Actions to be Taken in the Event of a Reportable Occurrence in Plant Operation . . 376 6.5 Actions to be Taken in the Event a Safety Limit is Exceeded ............. 376 l

6.6 Station Operating Records ......... 376 6.7 Reporting Requirements . . . . . . . . . . . 379 l 6.8 MinLaum Plant Staffing . . . . . . . . . . . 388 I

6.9 Environmental Qualification ........ 388 6.10 Integrity of Systems Outside containment . . 389 t

6.11 Iodine Monitoring ............. 389 v .

l l

l

LIST OF TABLES Table Title Page No.

3 1.A Reactor Protection System (SCRAM) Instrunentation Requirements . . . . . . . . . . . . . . . . . . . 32 4.1.A Reactor Protection System (SCRAM) Instrumentation Functional Iests Minimum Functional Test Frequencies for Safety Instrumentation and Control Circuits . . . . . . . . . . . . . . . . . 36 4.1.3 Reactor Protection System (SCRAM) Instrument Calibration Minimum Calibration Frequencies for Reactor Protection Instrument Channels . . . . . . 39 3 2.A Primary Containment and Reactor Building Isolation Instrumentation ................. 57 3 2.B Instrumentation that Initiates or Controls the Core and Containment Cooling Systems ......... 64 3.2.C Instrumentation that Initiates Rod Blocks .... 76 3.2.D Radioactive Liquid Effluent Monitoring Instrumentation ................. 79 3 2.E Instrumentation that Monitors Leakage Into Drywell ..................... 80 3 2.F Surveillance Instrumentation . . . . . . . . . . . 81 3 2.G Control Room Isolation Instrumentation . . . . . . 84 3 2.H Flood Protection Instrumentation . . . . . . . . . 85 3 2.I Meteorological Monitoring Instrumentation .... 86 3 2.J seismic Monitoring Instrumentation . . . . . . . . 87 3J.K Radioactive Gaseous Effluent Monitoring Instrumentation ................. 87A 4.2.A Surveillance Requirements for Primary Containment and Reactor Building Isolation Instrumentation . . 88 4.2.B Surveillance Requirements for Instrumentation that Initiate or Control the CSCS . . . . . . . . . . . 92 4.2.C Surveillance Requirements for Instrumentation that Initiate Rod Blocks ............... 99 vi

LIST OF TABLES (Continu d) s

' Table Title - Pare No.

4.2.D Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirement . . . . 100 N

4.2.E Minimum Test and Calibration Frsquency for Drywell Leak Detection Instrumentation . . . . . . 101 4.2.F Minimum Test and Calibration frequency for Surveillance Instrumentation . . . . . . .. . . . . 102 4.2.0 Surveillance Requirements for Control Room Isolation Instrumentation ............ .

103 S s .

4.2.H Minimum Test and Calibration Frequency for Floor Protection Instrumentation . . . . . . . . . 104 4.2.J Seismic Monitoring Instrument Surveillance . . . . 105 4.2.K Radioactive Gaseous Effluent Instrumentation Surveillance . . . . . . . . . . . . . . . . . . . 1051 3 5-1 Minimum RHRSW and EECW Pump Assignment . . . . . . 156a 3 5.I MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE . . . . . . 181 4.6.A Reactor Coolant System Inservice Inspection Schedule . . . . . . . . . . . . . . . . . . . . . 209 l

l c 3 7.A Primary Conta.. went Isolation Valves . . . . . . . 262 3 7.B Testable Penetrations with Double 0-Ring Seals . . 268 l 3 7.C Testable Penetrations with Testable Bellows . . . 269 3 7.D Primary Containment Testable Isolation Valves . . 270 3'.7.E Suppression Chamber Influent Lines Stop-Check Globe Valve L'eakage Rates ............ 279 .

3 7.F Check Valves on Suppression Chamber Influent .

Lines . . .................... 280 1

3 7.H Testable Electrical Penetrations . . . . . . . . . 283 4.8.A Radioactive Liquid Waste Sampling and Analysis . . 309 4.8.B Radioactive Gaseous Waste Sampling and Analysis. . 310 vii r

(

3 , LIST OF TABLES (Cont'd)

.- 5i Table Title Page No.

4 9.A.4.c voltage Relay Setpoints/ Diesel Generator Start . . 325a 3 11.A Fire Protection System Hydraulic Requirements. . . 355 3 13.A Radiological Environmental Monitoring Program. . . 359G 3 13.B Mavimum values for the Lower Limits of.

Detection (LLD) . . . . . . . . . . . . . . . . . 359J 3 13.C Reporting Levels for Radioactivity Concentrations in Environmental Samples ............ 359M 6.8.A Minimum Shift Crew Requirements . . . . . . . . . 390 l

I I

{ viii

LIST OF ILLUSTRATIOMS Figure Title Pace Co.

2.1.1 APRM Flow Reference Scram and APRM Rod Block Settings . . . . . . . . ............ 14 2.1-2 APRM Flow Bias Scram Vs. Reactor Core-Flow ... 25 4.1-1 Graphic Aid in the Selection of an Adequate Interval Between Tests . ............ 4S 4.2-1 System Unavailability . . . . . . . . . . . . . . 117 3 4-1 Sodium Pentaborate Solution Volume Concentration Requirements . . . . . . ............ 141 3.4-2 Sodium Pentaborate Solution Temperature Requirements . . .. . . ............ 142 3 5.K-1 MCPR Limits . . . . . . . . . . . . . . . . . . . 182b 3.5.2 Kg Factor vs. Percent Core Flow . ........ 183 3 6-1 Temperature-Pressure Limitations ........ 207 3.6-2 Change in charpy V Transition Temperature vs.

Neutron Exposure ... . ............ 208 4.8.1 Site Boundary . . . . . . . . . . . . . . . . . . 310C 6.1-1 TVA Office of Power Organization for Operation of Nuclear Power Plants . . . . . . . . . . . . . 391 6.1-2 Functional Organization . . . . . . . . . . . . . 392 6.2-1 Review and Audit Function . . . . . . . . . . . . 393 6.3-1 In-Plant Fire Program Organization ....... 394 ix

BROWS FERki EUCLEAR PLANT - UNIT 3 RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS 4

e

a protective trip function. A trip system may require one or

more instrument chennel trip signals related to one or more plant parameters in order to initiate trip system action.

Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip

- systems.

> 7. Protective Action - An action initiated by the protection system when a limit is reached. A protective action can be at a channel or system level.

8. Protective Function - A system protective action which results from the protective action of the channels monitoring a particular plant condition.

9. Simulated Automatic Actuation - Simulated automatic actuation means applying a simulated signal to the sensor to actuate the 4

circuit in question.

10. Logic - A logic is an arrangement of relays, contacts, and other components that produces a decision output.

(a) Initiating - A logic that receive signals from channels and produce decision outputs to the actuation logic.

(b) Actuation - A logic that receives signals (either from

' initiation logic or channels) and produces decision outputs to accomplish a protective action.

11. Channel Calibration - Shall be the adjustment, as necessary,

- of the channel output such that it responds with necessary i range and accuracy to known values of the parameters which the channel monitors. The channel calibration shall encompass the entire channel including alarm and/or trip functions and shall include the channel functional test. The channel calibration may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated. Non-calibratable components shall be excluded from this requirement, but will included in channel functional test and source check.

12. Channel Functional Test - Shall be :

a. Analog Channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

b. Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

13. Source Check - Shall be the qualitative assessment of channel response when the chann.1 sensor is exposed to a radioactive source or multiple of sources.

s 7

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1.0 DEFINITIONS (Cont'd) k' . Functional Tests - A functional test is the manual cperation l

or initiation of a systec, subsystem, or component to verify that it functions within design tolerances (e.g., the manual start of a core spray pump to verify that it runs and that it pumps the required volume of water).

X. Shutdown - The reactor is in a shutdown condition when the reactor mode switch is in the shutdown mode position and no core alterations are being performed.

Y. Engineered Safeguard - An engineered safeguard is a safety system the actions of which are essential to a safety action required in response to accidents.

Z. Solidification - Shall be the conversion of radioactive wastes into a form that meets shipping and burial ground requirements.

AA. Offsite Dese Calculation Manual (ODCM) - Shall be a manual describing the environmental monitoring program and the methodology and parameters used in the calculation of release rate limits and offsite doses due to radioactive gaseous and liquid effluents. The ODCM will also provide the plant with guidance for establishing alarm / trip setpoints to ensure technical specifications sections 3.8.A.1 and 3.8.3.1 are not exceeded.

BB. Purge or purging - The controlled process of. discharging air or gas from the primary containment to maintain temperature, pressure, humidity, concentration, or other operating condition in such a manner that replacement air or gas is required to purify the containment.

CC. Process Control Program - Shall contain the sampling, analysis, and formulation determination by which SOLIDIFICATION of radioactive wastes from liquid systems is assured.

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

7A

1.0 D FINITIONS (Cont'd)

EE. Site Boundary - Shall be that line beyond which the land is not owned, leased, or otherwise controlled by TVA.

FF. Unrestricted Area - Any area at or bevond 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 industrial, commercial, institutional, or recreational I purposes.

-GG. Dose Ecuivalent I-131 - The DOSE EQUIVALENT I-131 shall be the concentration of I-131 (in pCi/gm) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factor used for this calculation shall be those listed in Table III of TID-14844 " Calculation of Distance Factors for Power and Test Reactor Sites".

l HH. Gaseous Waste Treatment System - The charcoal adsorber vessels installed on the discharge of the steam jet air ejector to

[

[. provide delay to a unit's offgas activity prior to release.

1

[

II. Members 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 parmitted to use portions of the site for recreational, occupational, or other purposes not associated with plant functions. This category shall not include

! non-employees such a< vending machine servicemen or postmen I

who, as part of their formal job function, occasionally enter restricted areas.

JJ. Surveillance - Surveillance Requirements shall be cet during the OPERATIONAL CONDITIONS or other conditions specified for individual limiting conditions for operation unless otherwise stated in an individual Surveillance Requirements. Each Surveillance Requirement shall be performed within the specified time interval with, (1) A maximum allowable extention not to exceed 25% of the surveillance interval, but (2) The combined time entered for any 3 etssecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval Performance of a Surveillance Requirement within the specified time interval shall constitute compliance and OPERABILITY requirements for a limiting condition for operation and associated action statements unless otherwise required by these specifications. Surveillance requirements do not have to be performed on inoperable equipment.

73 4

Table 1.1 SURVEIIIANCE FREQUENCY NOTATION NOTATION FREQUENCY S (Shift) At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D (Daily) At least once per normal calendar 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months day (midnight to mienight).

W (Weekly) At least once per 7 days.

M (Monthly) At least once per 31 days.

Q (Quarterly) At least once per 3 months or 92 days.

SA (Semi-Annually) At least once per 6 months or 184 days.

Y (Yearly) At least once per year or 366 days.

R (Refueling) At least once per operating cycle.

S/U (Start-Up) Prior to each reactor startup.

N.A. Not applicable.

P (Prior) Completed prior to each release.

i 7C

LIMITING' CONDITIONS FOR OPERATION SURVEILLANCE REQUIPIMENTS 3.2 PROTECTIVE INSTRLF.:.NTATION 4.2 PROTECTn'E INSTRUMENTATION 3.2.D RADIOACTIVE LIQUID EFFLUENT 4.2.D RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MONITORING INSTRUMEhTATION I. The radioactive liquid 1. Each of the radioactive effluent instrumentation liquid effluent monitoring listed in Table 3.2.D shall instruments shall be be operable with the demonstrated operable by applicability as shown in performance of test in Table 3.2.D/4.2.D. Alarm / accordance with Table 4.2.D.

trip setpoints will be set in accordance with guidance given in the ODCM to ensure that the limits of l specification 3.8.A.1 are not exceeded.

2. The action required when the number of operable channels is less than the minimum channels operable requirements is specified in the notes for Table 3.2.D

3. With a radioactive liquid effluent monitoring channel alarm / trip setpoint less conservative than required by these specifications, suspend the release without delay, declare the channel inoperable, or adjust the alarm / trip setpoint to '

establish the conservatism required by these specifi-cations.

4. The provisions of specifi-cations 1.0.C and 6.7.2 are not applicable. .

51

M s page deleted I

1 I

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l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.2.K Radioactive Gaseous Effluent 3.2.K Radioactive Gaseous Effluent Monitoring Instrumentation Monitoring Instrumentation

1. The radioactive gaseous 1. Each of the radioactive effluent monitoring gaseous effluent monitoring instruments listed in instruments shall be demon-Table 3.2.K shall be strated operable by perfor-operable with the mance of tests in accordance applicability as shown in with Yable 4.2.K.

Tables 3.2.K/4.2.K. Alarm /

trip setpoints will be set in accordance with guidance given in the ODCM to ensure that the limits of specification 3.8.B.1 are not exceeded.

2. The action required when the number of operable channels is less than the Minimum Channels Operable requirement is specified in the notes for Table 3.2.K.

Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Release Report why the inoperability was not corrected in a timely manner.

3. With a radioactive gaseous effluent monitoring channel alarm / trip setpoint less censervative than required by these specifications, suspend the release without delay, declare the channel inoperable, or adjust the alarm / trip setpoint to establish the conservatism required by these specifications.

4. The provisions of specifi-cations 1.0.C and 6.7.2 are not applicable.

56A

r-l TABLE 3.2.D Hadioactive Liquid Effluent Honitoring Instrumentation Hinimum Channels Instrument (F) Operable A y plicability Action

- A, B

1. LIQUID RADWASTE EFFLUENT 1 HONITOR (RM-90-130)

- - C

2. HilR SERVICE WATER HONITOR 1 (RH-90-133, -134)

- D-

3. RAW COOLING WATER HONITOR 1 (101-90-132)

- E

4. LIQUID RADWASTE EFFLUENT 1 FLOW RATE (77-60 loop excluding fixed in line rotometer) l 79

-___ = _ _ _ _ _ _

NOTES FOR TABLE 3.2.D

At all times

- During releases via this pathway

- During operation of an RER loop having RHR system pressure greater than service water pressure ACTION A During release of radioactive wastes from the radwaste processing system, the following shall be met-(1) liquid waste activity and flowrate shall be continuously monitored and recorded during release and shall be set to alarm and automatically close the waste discharge valve before exceeding the limits specified in 3.8.A.1, (2) if this cannot be met, two independent samples of the tank being discharged shall be analyzed in accordance with Table 4.S.A; and two qualified station personnel shall independently check valving before the discharge. Otherwise, suspend release via this pathway.

ACTION B With a radioactive liquid effluent monitoring channel / alarm trip se point less conservative than required by these specifications, suspend release via this pathway without delay, ieclare the channel inoperable, or adjust the alarm / trip setpoint to establish the conservatism referred by these specifications.

ACTION C l

During operation of an RER loop having RER system pressure greater than service water pressure, the effluent from that unit's service water shall be continuously monitored. If an installed monitoring system is not available, either a. temporary monitor or grab samples taken every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed for radioactivity with an LLD of lE-7 pCi/ml (gross) or less than applicable MPC ratio (y isotopic).

ACTION D ,

With the number of channels OPERABLE less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue provided that a temporary monitor is installed or, at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , grab samples are collected and analyzed for radioactivity with an LLD of lE-7 pCi/ml (gross) or ( applicable MPC ratio (y isotopic).

ACTION E With the number of channels OEPRABLE less than required by the Mini =um Channels j Operable requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

Pump curves may be used to estimate ilow.

ACTION F Alarm / trip setpoints will be calculated in accordance with the guidance given in the Offsite Dose Calculation Manual (ODCM).

79A

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

i Table 3.2.I Meteorological Monitoring Instrumentation INSTRUMENT HINIMUH.

INSTRUMENT ' ACCURACY OPERABLE

1. WIND SPEED

a. Channel A Elevaticn 620 MSL Note #1 1

b. Channel B Elevation 737 MSL Hote #1 1

c. Channel C Elevation 887 tiSL Note #1 1 l

2. WIND DIRECTION

a. Channel A Elevation 620 MSL + 50 1

b. Channel D Elevation 737 tiSL + 50 1

c. Channel C Elevation 887 MSL i 50 1 3 AIR TEl1PEllATURE - DELTA T

a. Channel A Elevation 620-737 MSL 0.10C 1 l b. Channel B Elevation 620-887 MSL 0.1 0C 1 II . DEW POINT 0

a. Channel A Elevation 620 ftSL 0.1 C l'

b. Channel B Elevation 620 MSL 0.10C 1 i

!!0TE #1 - Starting speed of anemometer shall be <1 mph. Accuracy is within 2 1 percent of mph reading or 0.15 mph, whichever is greater. l 86

TABLE 3.2.K lladiorcLiva Gaseous Effluent H:nitoring Instrumentetien Hinimum Channels /

Instrument Devices Operable Applicability Action

1. STACK (RH-90-147A & B)

Noble Gas Honitor (1) A/C

a. *

(1) B/C

b. Iodine Cartridge Particulate Filter (1) B/C c.

D

d. Sampler Flow Abnormal (1)

D

e. Stack Flow (FT, FM, (1)

FI-90-271)

2. REACTOR /TURHINE DLDG VENTILATION (RM-90-250)

A/C

a. Noble Gas Monitor (1) *

(1) B/C

b. Iodine Sampler Particulate Sampler (1) D/C c.

D

d. Sampler Flowmeter (1)

3. TURBINE BLDG EXilAUST (RM-90-249, 251) **

(1) A/C

a. Nobile Gas Honitor ** B/C

b. Iodine Sampler (1) **

Particulte Sampler (1) B/C c.

- D

d. Sampler Flowmeter (1)

4. HADWASTE BLDG VENT (RM-90-252)

A/C

a. Noble Gas Honitor (1)

Iodine Sampler (1) H/C

b. *

B/C

c. Particulate Sampler (1)

D

d. Sampler Flowrate (1)

5. OFF GAS IlYDROGEN ANALYZER *** E (II2 6. II 28) (1)

6. OFF GAS POST TREATHENT

a. Noble Gas Activity Honitor

F

,. (RM-90-265, 266) (1)

b. Sample Flow Abnormal

D (PA-20-262) (1) 87A

N0'"ES FOR TABE 3.2.K

At all times

- During releases via this pathway

- - During main condenser offgas treatment system operation ACTION A With the number of channels OPERABLE less than required by the Minimum Chanrels Operable requirement, effluent releases via the affected pathway may continue provided a temporary monitoring system is installed or grab samples are taken and analyzed at least once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

ACTION B With the number of channels OPERABLE less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue provided samples are continuously collected with auxiliary sampling equipment for periods on the order of seven (7) days and analyzed in accordance with Table 4.8.B within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after the end of the sampling period.

ACTCN C A monitoring system may be ,out of service for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for functional testing, calibration, or repair without providing or initiating grab sampling.

ACTION D e

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

ACTION E With the number of channels OPERABLE less than required by the Minimum Channels Operable requirement, operation of main condenser offgas treatment system may continue provided that a temporary monitor is installed or grab samples are collected at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION F With the number of channels OPERABLE less than required by the Mini =um Channels Operable requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gross activity within 24 houra.

87B

TAllLE 4.2.1)

Radioactive Liquid Effluent Honitoring Instrumentation Siirveillance Requirements Channel Funcational Source Check Calibration Test Inst rtiment Instrument Check D(4) H R(5) Q(1)

1. LIQUID RADWASTE EFFLUENT HONITOR (RH-90-130)

D(4) H R(5) Q(2) 4 2. RilR SERVICE WATER HONITOR (RN-90-133, -134)

D(4) H R(5) Q(2)

3. RAW COOLING WATER HONITOR (RH-90-132)

D(4) NA R Q(3)

4. LIQUID RADWASTE EFFLUENT FLOW RATE (77-66 loop) t l

i l

l I

l 100 i

=- ,

NOTES FOR TASI.E 4.2.D f

(1) The channel functional test shall also demonstrate that automatic isolation of this pathway and control room annunciation occurs if any of the following conditions exist:

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

b. Instrument indicates an inoperative /downscale failure

c. Instrument controls not set in operate mode (2) The channel functional test shall also demonstrate that control room annunciation occurs if any of the following conditions exist:

a. Instrument indicates measured levels above the alarm setpoint

b. Instrument indicates an inoperative /downscale failure

c. Instrument controls not set in operate mode (3) This functional test shall consist of measuring rate of tank decrease over 4 period of time and comparing this value with flow rate instrument reading.

(4) INSTRUENT CECK shall consist of verifying indication during periods of release. INSTRUENT CECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days which continous, periodic, or batch releases are made.

(5) The CHANEI, CALIBRATION shall include the use of a known (traceable to National Bureau of Standards Radiation Measurement System) radioactive source (s) positioned in a reproducible geometry with respect to the sensor or using standards that have been obtained from suppliers that participate in measurement assurance activities with the National Bureau of Standards (NBS).

l l

i f

I 100A

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

TABLE 4.2.K Radioactive Gaseous Effluent Instrumentation Surveillance Channel Functional Instrument Check Source Check Calibration Test Instrument .

I* A D H R(1) Q(2)

a. Noble Gas Monitor (5) NA NA NA

b. Iodine Cartridge W NA NA NA

c. Particulate Filter W NA R Q

d. Sampler Flowmeter D NA R Q

e. Stack Flowmeter D

2. REACTOR /TURHINF. BLDG g T H R O) Q(2)

Noble Gas Honitor D

a. NA NA NA )

b. Iodine Sampler W NA NA NA l Particulate Sampler W

c. NA R Q Sampler Flowmeter D l

. d.

3. TURRINE ilLDG EX11AUST D H R(I Q

a. Nobile Gas Monitor (6) NA NA NA ,

b. Iodine Sampler W NA NA NA

c. Particulte Sainpler W Q NA R Sampler Flowmeter D

d. l f

4. RADWASTE BLDG VENT D !! R(1) Q(2) j

a. Noble Gas Monitor (6) NA NA NA !

b. Iodine Sampler W NA !

l NA NA Particulate Sampler W Q j

c. NA R

d. Sampler Flowrate D

5. OFF GAS IlYDROGEN ANALYZER W NA R(3) Q(4) f (II2 A,If2 B) 2)

6. OFF GAS POST TREATIIENT H R(1) Qg(4)

a. Noble Gas Act.ivity Honitor D NA R Q Sample Flow Abnormal D

b. ,

105A I

. .. 1

s NOTES FCR TABLE 4.2.K (1) The CHANNEL CAL 3IRATION shall include the use of a known (traceable to the National Bureau of standards radiation measurement system) radioactive source (s) positioned in a reproducible geometry with respect

.to the sensor or using standards that have obtained from suppliers _ that _ ,

participate in measurement assurance activities with the National Bureau of Standards.

(2) The CHANKEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:

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

b. Instrument indicates an inoperable /downscale failure,

c. Instrument controls not set in operate mode (stack only).

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

a. Zero volume percent hydrogen (compressed air) and,

b. One volume percent hydrogen, balance nitrogen.

(4) The channel functional test shall demonstrate that automatic isolation of this pathway and control room annunciation occurs if any of the following conditions exists:

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

b. Instrument indicates an inoperative /downscale failure.

c. Instrument controls not set in operate mode.

The two channels are arranged in a coincidence logic such that 2 upscale, or 1 downscale and 1 upscale or 2 downscale will isolate the offgas line.

- .(5) The noble gas monitor shall have a LLD of lE-5 (Xe 133 Equivalent).

(6h The noble gas monitor shall have a LLD of lE-6 (Xe 133 Equivalent).

105B

The operability of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seisnic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for Browns Ferry Nuclear Plant. The instrumentation provided is consistent with specific portions of the recommendations of Regulatory Guide 1.12 " Instrumentation for Earthquakes."

The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments will be calculated in accordance with guidance provided in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring the concentration of potentially explosive gas mixtures in the offgas holding system. The operability and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with guidance provided in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20 Appendix B, Table II, Column 2. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

l 1

I 113

.. . - _ ._ . ~

- The most likely cause would be to stipulate that one channel be bypassed, tested, and restored, and then immediately following, the second channel be bypassed, tested, and restored. This is shown by Curve No. 4. Note that there is no true minimum. The curve does have a definite knee and very little red uction in system unavailability is achieved by testing at a shorter

- interval than computed by the equation for a single channel.

The best test procedure of all those examined is to perfectly stagger- the

, tests. That is, if the test interval is four months, test one or the other channel every two months. This is shown in Curve No. 5. The difference between Cases 4 and 5 is negligible. There may be other arguments, however,

.that more strongly support the perfectly staggered tests, including reductions

. in human error.

The conclusions to be drawn are these:

1. A 1 out of n system may be treated the same as a single channel in terms of choosing a test interval; and

2. more than one channel should not be bypassed for testing at any one time.

The radiation monitors in the refueling area ventilation duct which initiate building isolation and standby gas treatment operation are arranged in two 1 out of'2 logic systems. The bases given for the rod blocks apply here also and were used to arrive at the functional testing frequency. The off gas post treatment monitors are connected in a 2 out of 2 logic arrangement. Based on experience with instruments of similar design, a testing interval of once every three months has been found adequate.

The automatic pressure relief instrumentation can be considered to be a 1 out of 2 logic system and the discussion above applies also.

The criteria for ensuring the reliability and accuracy of the radioactive gaseous effluent instrumentation in listed in Table 4.2.K.

The criteria for ensuring the reliability and accuracy of the radioactive liquid effluent instrumentation is listed in Table 4.2.D.

l l

I i

116

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIP M NTS 3.6 PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY

3. 'At steaming rates 3. Additional coolant

( greater than 100,000 samples shall be 1b/hr, the reactor taken whenever the water quality may reactor activity exceed specification exceeds one percent 3.6.3.2 only for the of the equilibrium time limits specified concentration below. Exceeding specified in 3.6.B.5 these time limits of and one of the the following maximum following conditions quality limits shall are met:

becaus- for placing the reactor in the a. During startup cold shutdown b. Following a condition. significant power change **

a. Conductivity c. Following an time above increase in the 2 pmho/cm @ 25'C equilibrium off-4 we'aks/ year. gas level exceed-Maximum Limit ing 10,000 uci/see 10 pmho/cm @ 25'C (at the steam)

b. Chloride jet air ejector) concentration time within a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months above 0.2 ppm - period.

4 weeks / year. d. Whenever the Maximum Limit - equilibriu:n 0.5 ppm. iodine limit specified in 3.6.B.5.in exceeded.

- For the purpose of this section on sampling frequency, a significant power exchange is defined as a change exceeding 15%

of rated power in less than I hour.

1 J

188 1

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY 4.6 PRI!iARY SYSE1 BOUhTARY

5. klenever the reacter is

critical, the limits on activity concentrations in the reactor coolant shall not exceed the equilibrium value of 3.2 pc/gm of dose equivalent I-131. This limit may be exceeded following power transients for a maximum of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

During this activity transient the iodine concentrations shall not exceed the equilibrium values by a factor of more than 10 whenever the reactor is critical.

The reactor shall not be operated more than 5 percent of its yearly power operation under this exception for the equilibrium activity limits. If the iodine concentration in the coolant exceeds the equilibrium limit by a factor of ten, the reactor shall shutdown, and the steam line isolation valves shall be closed immediately.

190

LIMITING CONDITIONS FOR OPERATION SURVEILI).NCE P2QUIREMENTS 3.8 RADIOACTIVE MATERIALS 4.8 RADI0 ACTIVE MATERIALS Applicability Applicability Applies to the release of Applies to the periodic test radioactive liquids and gases and record requirements and from the facility. sampling and monitoring methods used for facility Objective effluents.

To define the limits and Objective conditions for the release of radioactive effluents to the To ensure that radioactive environs to assure that any liquid and gaseous releases radioactive releases are as from the facility are low as reasonably achievable maintained within the limits and within the limits of specified by Specifications 10 CFR Part 20. The specifi- 3.8.A and 3.8.B.

cations except for 3.8.A.1 and 3.8.B.1 are exempt from the Specification requirements of definition 1.0.C (Limiting Condition for A. Lionid Effluents Operation).

1. Facility records shall Snecification be maintained of radioactive concentrations A. Liquid Effluents and volume before dilution of each batch

1. The concentration of of liquid effluent radioactive material released, and of the released at any time average dilution flow from the site to and length of time over

. unrestricted areas which each discharge (see Figure 4.8-1) occurred.

shall be limited to i the concentrations 2. Radioactive liquid waste specified in 10 CFR sampling and activity Part 20, Appendix B, analysis of each liquid Table II, Column 2 waste batch to be for radionuclides other discharged shall be performed prior to l than dissolved or I entrained noble gases. release in accordance For dissolved or with Table 4.8.A.

I entrained noble gases, the concentration shall 3. The operation of the j

be 1:Lmited to 23-4 pCi/ml automatic isolatiCn l

total activity. valves and discharge tank selection valves

. 2. If the limits of 3.8.A.I shall be checked are exceeded, appropriate annually.

action shall be initiated .s without delay to bring .,

the release within -

299

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8 RADIOACTIVE MATERIAI.S 4.8 RADIOACTIVE MATERIALS limits. Provide prompt 4. The results of the analysis of notification to the NRC samples collected from release pursuant to section Points shall be used with the calculational methodology 6.7.2.A.

in the OCDM to assure that

3. The dose or dose commit- the concentrations at the ment to a member raf the Point of release are public from radioactive maintained within the materials in liquid limits of specification effluents released from 3.8.A.I.

each unit to unrestricted areas (See Figure 4.8-1) 5. Cummulative quarterly and shall be limited: yearly dose contributions from liquid effluents shall be determined as specified

a. During any calendar quarter to (1.5 mrem to in the ODCM at least once the total body and to (5 every 31 days.

mrem to any organ and,

6. Doses due to liquid

b. During any calendar releases to unrestricted year to (3 mrem to the areas shall be projected total body and (10 at least once per 31 days, mrem to any organ. in accordance with the ODCM.

4. If the limits specified in 3.8.A.3 a & b above are exceeded, prepare and sub-mit Special Report pursuant to Section 6.7.2.B.2.

5. The liquid radwaste

, system shall be.used to

! reduce the radioactive i materials in liquid l wastes prior to their j discharge from the site when the projected l

monthly dose would l exceed 0.06 mrem to l the total body or 0.21

! mrem to any organ per unit (see Figure 4.8.1).

6. With radioactive liquid waste being discharged for

- more than 31 day's without treatment and when the projected dose is in 300

N +*6 Ne IIIIITING CONDITIONS FOR OPERATION SURVEII.IANCE REQUIRE!!EhTS 3.8 RADIOACTIVE IfATERIALS 4.8 RADIOACTIVE MATERIALS excess of limits specified in 3.8.A.5, prepare and submit the Special Report pursuant to Section 6.7.2.B.2.

7. - The maximum activity to be contained in one liquid radwaste tank or temporary radwaste storage tank that can i be discharged directly to the environs shall .

not exceed 10 curies excluding tritium and dissolved / entrained noble gas.

8. With radioactive liquid waste exceeding 3.8.A.7 limits, without delay suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the activity of tank contents to within the limit.

G a

9 1

301

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IIMITING CONDITIONS FOR OPERATION SURVEILLANCE PIOUIRE.ENTS 3.8 RADIOACTIVE MATERIALS 4.8 RADIOACTIVE MATFlIALS B. Airborne Effluents B. Airborne Effluents

1. The dose rate at any 1. The gross E/y and time in the unrestricted particulate activity of areas (see Figure 4.8-1) gaseous wastes released due to radioactivity to the encironment released in gaseous shall be conitored and effluents from the site recorded, shall be limited to the following values: a. For effluent stresas having continuous

a. The dose rate limit monitoring capability, for noble gases shall the activity shall be (500 mrem /yr to be monitored and the total body and flow rate evaluated (3000 mrem /yr to the and recorded to enable skin, and release rates of gross radioactivity to be

b. The dose rate limit determined at least for I-131, H-3, and once per shift using particulates with instruments specified greater than eight in table 3.2.K. .

day half-lives shall be<1500 mrem /yr to b. For effluent streams any organ. without continuous monitoring capability,

2. If the limits of 3.8.B.1 the activity shall be are exceeded, appropriate monitored and recorded corrective action shall be and the releases immediately initiated to through these streams bring the releases within controlled to within limits. Provide prompt the limits specified notification to the NRC in 3.S.B.

pursuant to section 6.7.2.A. 2. Radioactive gaseous waste sampling and activity

- analysis shall be performed in accordance with Table 4.8.B.

302

I.IMITING CONDITIONS FOR OPERATION SURVEIIIANCE REQUIREMENTS

3. The air dose in 3. Cumulative quarterly unrestricted areas (see and yearly dose Figure 4.8-1) due to contributions.from

noble gases released in gaseous releases shall gaseous effluents per be determined using unit shall be limited to methods contained in the following: the ODCM at least once every 31 days.

a. During any calendar quarter, to {5 mrad 4. Doses due to gaseous for gamma radiation releases to unrestricted and {10 mrat', for areas shall be projected beta radiation; in accordance with the ODCM at least once

b. During any calendar per 31 days.

year, to {10 mrad for gamma radiation and 5. Samples of offgas system

{20 mrad for beta effluents shall be radiation. analyzed at least weekly to deterr.ine the

4. If the .alculated air identity and quantity dose exceeds the limits of the principal specified in 3.8.B.3 radionuclides being above, prepare and submit released.

a special report pursuant to section 6.7.2.B.2. 6. .In accordance with the methods and procedures

5. The dose to a member of of the ODCM a release the public from radio- rate limit methodology iodines, radicactive for noble gases in gaseous materials in particulate effluents shall be used form, and radionuclides to ensure compliance other than noble gases with the limits specified with half lives greater in specification 3.8.B.1. ,

than 8 days in_ gaseous effluent released per unit to unrestricted areas (see Figure 4.8-1) shall be 1imited to the following: .

a. To any organ during any calendar quarter to {7.5 mrem;

b. To any organ during any calendar year to {15 mrem; .

303 w - - w-- .yp.- r>+ -t--e,--g-rw , ,g y9,%_ , 3 -g,,-p+- --

y9 -e . . - -,97,y- pw- y-w7- w--- y -, - - -9 3,cw -r-r -'7'T---f w

r-W v=1.g---- -1 wv- - -"'r--N

LIMITING CONDITIONS FOR OPERATION SURVEII.I.ANCE REOUIREMEh"IS

6. If the calculated doses exceed the limits of 3.8.B.5 abovt, prepare and sub=it a sper.ial report pursuant to section 6.7.2.S.2.

7. During operation above 25*4 power the discharge of the SJAE must be routed through the charcoal adsorbers when the projected gaseous effluent releases to unrestricted areas (see Figure 4.8-1) when averaged over 31 days would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation per unit.

8. With gaseous wastes being discharged for more than 31 ' days without treatment and when the projected . -

dose is in excess of the limits of 3.8.B.7 above, prepare and submit a special report pursuant to section 6.7.2.B.2.

4 l

304 l

l

I.IMITING CONDITIONS y0R OPERATION SURVEILIANCE REQUIREMENTS 3.8.C Radioactive Effluents - Dose 4.8.C Radioactive Effluents - Dose

1. The dose or dose commitment 1. Cumulative dose contributions to a real individual from all from liquid and gaseous uranium fuel cycle sources effluents shall be determined is limited to f 25 mrem to in accordance with specifi-the total body or any organ cations 3.8.A.3, 3.8.B.3, (except the thyroid, which and 3.8.B.5 and the methods is limited to f 75 mrem) in the ODCM. s over a period of one calendar year.

2. k'ith the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of specification 3.8.A.3, 3.8.B.3, or 3.8.B.5, prepare and submit a Special Report to the Commission pursuant to specification 6.7.2.B.2 and limit the subsequent releases such that the limits of 3.8.C.1 are not exceeded.

3.6.D Mechanical Vacuum Pump 4.8.D Mechanical Vacuum Pump

1. The mechanical vacuum pump At least once during each shall be capable of being operating cycle verify automatic automatically isolated and securing and isolation of the -

secured on a signal or high mechanical vacuum pump.

radioactivity in the sveam lines whenever the main steam isolation valves are open.

2. If the limits of 3.8.C.1 are not met, the vacuum pump shall be isolated.

i 305

..-r - .- e- -.----., 4 % .---

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e-~, p--.--e, , - - - - w-- ----. yw, , - - - - - y- -.- . --

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.8 RADIOACTIVE MATERIALS 4.8 RADIOACTIVE MATERIALS E. Miscellaneous Radioactive E. Miscellaneous Radioactive Materials Sources haterials Sources

1. Source Leakare Test 1. Surveillance Reonirement Each sealed source Tests for leakage and/or containing radioactive contamination shall be material either in performed by the licensee excess of 100 microcuries or by other persons of beta and/or gamma specifically authorized emitting material or by the Co==ission or an 5 microcuries of alpha agreement State, as emitting material shall follows:

be free of > 0.005 microcurie of removable a. Sources In Use contamination. Each sealed source with Each sealed source, removable contamination excluding startup in excess of the abeve sources and flux limit shall be dettetors previously immediately withdrawn subjected to core

- from use and (a) either flux, containing decontaminated and radioactive material, repaired, or (b) disposed other than Hydrogen 3, of.in accordance with with a half-life Commission reguletions. greater than thirty days and in any form other than gas shall be tested for leakage and/or contamination at least once per six months. The leakage test shall be capable of detecting the presence of 0.005 microcurie of

. radioactive material on the test sample.

306

EIMITING CONDITIONS FOR OPERATION SUDNEILIJ.NCE REQUIREMEh"IS 4.8.E Miscellaneous Radioactive Materials Sources

1. Surveillance ~Reauirements

b. Stored Sources Not In Use Each sealed source and fission detector not previously subjected to core flux shall be tested prior to use or transfer to another licensee unless tested within the previous six months.

Sealed sources and fission detectors transferred without-a certificate indicating the last test date shall be tested prior to use.

c. Startup Sources and Fission Detectors Each sealed startup source and fission detector shall be tested prior to being subjected to core flux and following repair or maintenance to the source.

2. Reports A report shall be prepared and submitted to the Commission on an annual basis if sealed sources or fission detector leakage tests reveal the presence of greater than or equal tc 0.005 microcuries of removable contamination.

T 307

=

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8 RADIOACTIVE MATERIALS 4.8 RADIOACTIVE MATERIALS F. Solid Radwaste F. Solid Radwaste

1. The solid radvaste system 1. The Process Control shall be operated in Program shall accordance with a process include surveillance control program, for the checks necessary to solidification and packag- demonstrate compliance ing of wet radioactive with 3.8.F.1.

vastes to ensure meeting the requirements of 10 CFR 20 and 10 CFR 71 and burial ground requirements prior to shipment of radioactive wastes from the site.

2. With the packaging requirements of 10 CFR 20 or burial ground require-ments and/or 10 CFR 71 not satisfied, suspend shipments of defectively packaged solid radioactive wastes from the site.

i l

1 l

308

TABLE 4.8.A RADIDACTIVE QQUID WASTE SAMPLING AND ANALYSIS PROGRAM SYSTEM DESIGN CAPABILITY SAMPLING HINIMUM ANALYSIS TYPE OF ACTIVITY LOWER LIMIT OF DETECTION LIQUID RELEASE (pci/ml)

TYPE FREQUENCY FREQUENCY ANAT.YSIS (LLD)

Each Batch Each Batch Prior Principa SE-7(3 DatcliWag to Helease Emitters {4jamma Releases One Batch Honthly Dissolved and ~

per Month Entrained Gases (5)

Monthly Honthly Tritium 1 E-5 Proportional Composite Gross a 1 E-7 (2)

Quarterly Sr-89, Sr-90 5 E-8 Proportional Quarterly Composite Fe-55 1 E-6 (2) 1 309 ,

_-_._m __-_______m

TABLE NOTATIOK - TABLE 4.8.A (I) A batch release is the discharge of liquid wastes of a discrete volume.

The discha rge shall be throughly mixed prior to sampling.

(2) A proportitnal composite sample is one in which the quantity of liquid sampled is ,iroportional to the quantity of liquid waste discharged from the plant at' is representative of the liquid discharged .

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

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

4.66 s b E V 2.22 x 10' + Y exp (-Mt)

Where:

LLD is the "a priori" lower limit of detection as deficed above (as microcuries per unit mass or volume),

s b

is the standard deviation of the b'ackground 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 los is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),

A is the radioactive decay constant for the particular radionuclide, and .

At for plant effluents is the elapsed time between the midpoint of sample collection and time of counting.

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 (after the fact) limit for a particular measurement.

LLD applicability begins within six months after the Nuclear Data or equivalent systems becomes operational.

~

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g. ,.-- -,-. , .,- yy,s -, ,,g,_ pp,. , g.w_.g,--o- ,,p w y . %-,_.o- ,.oo,.--os.

(4) The principal gamma emitters for which the IID specification will apply are exclusively the following radionuelides: Zn65, Co60, Cs137, Mn54, CoS8, Csl34, Ce141, CE144, Mo99, and Fe59 for liquid releases. This list does not mean that only these nuclides are to be detected and reported. Other nucludes detected within a ~95", confidence level, together with the above nuclides, shall also be identified and reported as being present. Nuclides which are below the IID for the analysis may not be reported as being present at the IID Level for that nuclide.

I-131 shall have a IID of <1 E-6.

(5) Gamma Emitters Only.

m 9

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TABLE 4.8.B RADIOACTIVE GASEOUS WASTE SAMPLINd AND ANALYSIS PROGRAM SYSTEM DESIGN CAPABILII 2 GASEOUS RELEASE SAMPLING HINIMUN ANALYSIS TYPE OF ACTIVITY LOWER L1 HIT OF DETECTIC

TYPE FREQUENCY FREQUENCY ANALYSIS (pCi/ml)

A. Containment Prior to Each Prior to Each Purge Principa Emitters {3 gamma IE-b Purge Purge Grab Sample l

H-3 IE-6 B. 1. Stack Principal IE-4( }

{i Emitters (gjema Grab Sample Monthly ) H-3 1E-6

2. Building

Ventilation ,

- a. Reactor /

I Turbine

b. Turbine Exhaust j c. Radwaste C. Continuous I-131 1E-12( }

All Release - Charcog) Sample Points Listed Sampler Weekly in B. Above IE-11 g

Continuous Principa j PartieggteSample Emmiters{3 gamma

! Sampler Weekly I-131 1E-12(2) l IE-11

vontinuous Composite Particulate Gross Alpha Sampler Sample Monthly

. Continuous Composite Particulate Sr-89, Sr 90 IE-11

> Sampler Sample Quarterly I 310 i

t TABLE NOTATION - 4.8.B i

(1) The LLD is defined, for the purposes of these specifications as the smallest concentration of radioactive material in a sample that will yield a new count (above system background) that will be detected with 95% probability with only 5% probability of falsely concluding that a

, blank observation represents a "real" signal.

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

4.66 ab E V 2.22 x 106 Y - erp (-Aat) .

Where:

LLD is the "a priori" lower limit of detection as defined above (as microcuries per unit mass or volume),

ss is the standard deviation of the background counting rate or of

le counting rate of a blank sample as appropriate (as counts per j

minute),

.i E is the counting efficiency (as counts per disintegration),

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

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

A is the radioactive decay constant for the particular radionuclide, ,

and i

At for plant effluents is the elapsed time between the midpoint of sample collection and time of counting.

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

l 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 i

not as an a posteriori (after the fact) limit for a particular j measurement.

LLD applicability begins within six months after the Nuclear Data or equivalent system becomes operational.

(2) When samples are taken more often that that shown, the minimum detectable i

concentrations can be correspondingly higher.

1 310A

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(3) The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Ma-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the IID for the analyses should not be rg..,rd as being present at the LLD level for that nuclide.

(4) Analysis shall also be performed if the radiation monitor alara exceeds the setpoint value.

n 310B

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J 3.8 BASES Radioactive waste release levels to unrestricted 'artas should be kept "as low as reasonably achievable and are not to exceed the concentration limits specified in 10 CFR Part 20. At the same time, these specifications permit 4 the flexibility of operation, compatible with considerations of health and ,

safety, to assure that the public is provided a dependable source of power under unusual operating conditions which may temporarily result in releases higher than design objectives but still within the concentration limits specified in 10 CFR Part 20. It is expected that by using this operational flexibility and exerting every effort to keep levels of radioactive materials releared as low as reasonably achievable in accordance with criteria established in 10 CFR 50 Appendix I, the annual releases will not exceed a small. fraction of the annual average concentration limits specified in 10 CIR Part 20.

3.8.A LIQUID EFFLUENTS Specification 3.8. A.1 is provided to ensure that the concentration of

^

radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within (1) the Section 11.A design objectives of Appendix I, 10 CFR Part 50, to an individual and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

Specification 3.8.A.3 is provided to implement the dose requirements of Sections II.A, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth the Section 11.A of Appendix I. .

Specification 3.8.A.4 action statements provides the required operating i flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is. reasonably achievable". Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by 311 l

m . _ . . - _ . _ , _ . _ _ _ _ -._,,_.....___._,,..--,,.,~._._.m,_. _ _ _ . - . _ _ _ _ _ . _ . - - , , _ _ , , _ , _ . . _ _ _ _ _ , - _ , _ . . - . . _ , . ~ _

(

1 1 - 3.8.A I.IQUID EFFLUENTS (cont'd) calculational 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 will be 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 Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I,"

April 1977. NUREG-0133 provides methods for dose calculations consistent with i ~ Regulatory Guides 1.109 and 1.113.

Specification 3.8.A.5 requires that the appropriate portions of the liquid radwaste treatment system be used when specified. This provides assurance

- that the releases of radioactive ~ materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the

. requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and design objective Section II.D of Appendix I to 10 CFR

! Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the guide set forth in Section II.A of Appendix I,10 CFR Part 50, for liquid effluents.

Specification 3.8.A.6 requires submittal of a special report if the limiting values of Specification 3.8.A.5 are exceeded and unexpected failures of non-redundant radwaste processing equipment halt waste treatment.

3.8.B AIRBORNE EFFLUENTS Specification 3.8.B.1 is provided to ensure that the dose rate at anytime at the exclusion boundary from gaseous effluents from all units on the site will

be within the annual dose limits of 10 CFR Part 20 for unrestricted areas.

The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide i reasonable assurance that radioactive material discharged in gaseous effluents l will not result in the exposure of a member of the public in an unrestricted area, either within or outside the exclusion area boundary, to annual average l- concentrations exceeding the limits specified in Appendix B, Table II of i 10 CFR Part 20 (10 CFR Part 20.106(b)). For members of the public who say

at times be within the exclusion area boundary, the occupancy of the members of the public will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the exclusion area boundary.

1 312 I

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~3.8.E AIRBORKE EFFI.UENTS (cont'd)

The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the exclusionareaboundaryto{'500mram/yeartothetotalbodyorto f 3000 mrem / year to the skin. . These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to an infant via the cow-milk-infant pathway to { 1500 mrem / year for the nearest cow to the

. plant.

4 Specification 3.8.B.2 requires that appropriate correction action (s) be taken to reduce gaseous effluent releases if the limits of 3.8.B.1 are exceeded.

Specification 3.8.B.5 dose limits is provided to implement the requirements of Section II.C, III.A, and IV of Appendix I, 10 CFR Part 50. The limiting conditions for operation are the guides set forth in Section II.C of

, Appendix I. .

Specification 3.8.B.6 action statement provides the required operating flexibility and at the same time implement the guides set forth in '

Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an e individual through appropriate pathways is unlikely to be substantially 1 underestimated. The ODCM calculational methods used for

calculating the doses due to the actual release rates of the subject materials are required to be consistent with the methodology provided in Regulatory Guide 1.109, " Calculating of Annual Doses to Man from Routine Releases of

! Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR i Part 50, Appendix I," Revision I, October 1977, NUREG/CR-1004, "A Statistical Analysis of Selected Parameters for Predicting Food Chain Transport and Internal Dose.of Radionuclides", October 1979, and Regulatory Guide 1.111,

" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous

' Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1,

July 1977. These equations also provide for determining the actual doses i based upon the historical average atmospheric conditions. The release rate specifications for radiciodines, radioactive material in particulate form and

- radionuclides other than noble gases are dependent on the existing l radionuclide pathways to man, in the unrestricted area. The pathways which are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto l

[

green leafy vegetation with subsequent consumption by man, 3) deposition onto l

grassy areas where milk animals and meat producing animals graze with

consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

Specification 3.8.B.6 action statement requires that a special report be prepared and submitted to explain violations of the limiting doses contained in Specification 3.8.B.5. i 313 l '

I l

i AIRBORNE EFFLUENTS Specification 3 8.B.7 requires that the offgas charcoal adsorber beds be used when specified to treat gaseous effluents prior to their release to the environment. This provides reasonable assurance that the releases of radioacive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50 36a, General Design Criterion 60 of Appendix A to 10 CFR part 50, and desi 6n objective Section II.D of Appendix I to 10 CFR Part 50.

l The specified limite hoverning the use of appropriate portions of the

' systems were specified as a suitable fraction of the guide set forth in Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous effluents.

Specification 3 8.B.8 requires that a special report be prepared and submitted to explain reasons for any failure to comply with Specification 3 8.B.7.

l Specification 3 8.B.3 is provided to implement the requirements of Section II.B III.A, and IV.A of Appencix I,10 CFR Part 50. The Limiting Condition for Operation implements the guide set forth in Section II.C of Appendix I. _

Specification 3 8.B.4 action statement provides the required operating flexibility and. at the same time implement the guides set forth in Sec-tion IV.A of Appendix I to assure that the releases of radioative material in gaseous effluents will be kept "as low as is reasonably achieveable."

The Surveillance Requirements implement the requirements in Section III.A l of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of a member of the public through the appropriate pathways is unlikely to be substantially underestimated. The dose calculaticas estab-lished in the ODCM for calculating the doses dua to the actual relcase rates of radioactive noble gases in gaseous effluents will be consistent i

.with the methodology previded in Regulatory Guide 1.109, " Calculation of l 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, NUREG/R-1004, "A Statistical Analysis of Selected

, Parameters for Predicting Food Chain Transport and Internal Dose of Radio-nuclides," October 1979 and Regulatory Guide 1.111, " Methods for Estimating ,

Atmospheric Transport and Dispersion of Gaseous Effluents in Routine

! Releases from Light-Water-Cooled Reactors," Revision 1, July 1977 The

+ ODCM equations provided for determining the air doses at the exclusion area '

j boundary will be based upon the historical average atmospheric conditions.

NUREG-0133 provides methods for dose calculations consistent with Regula-tory Guides 1.109 and 1.111. Specifications 3.8.B.4 requires that a special report be prepared and submitted to explain violations of the limiting doses contained in Specification 3 8.B.3 4.8.A and 4.8.B BASES The surveillance requirements given under Specification 4.8.A and 4.8.B

provide assurance that liquid and gaseous wastes are properly controlled and monitored during any release of radioactive materials in the liquid and I

314

-,...,n- - , . - y.-.i-w . ,- ,..=m ~,-w,.,.., , - , , . , - - . , , ,,....y%.,,. ,.77_. ,.,.,.m%,_,.y%..,,.,%mmer-,,-.+,,w-m. u mw w w y y--m .~ - -,

4.8.A and 4.8.B BASES (cont'd) gaseous effluents. These surveillance reouirements provide the data for the licensee and the Commission to evaluate the station's performance relative to radioactive wastes released to the environment. Reports on the quantities of radioactive materials released in effluents shall be furnished to the Commission on the basis of Section 6 of these technical specifications. On the basis of such reports and any additional informa-tion the Commicsion may obtain from the licensee or others, the Com::11ssion may from time to time require the licensee to take such actions as the

Commission deems appropriate.

3.8.C and 4.8.C BASES This specification is provided to meet the dose limitations of 40 TR 190.

The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites con-taining up to four reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action which should result in the limitation of dose to a member of the public for the calendar year to be within the 40 CFR 190 limits. For the purposes of the Special Repcrt, it may be assumed that the dose comitment to the r. ember 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 five miles must be considered.

3.8.D and 4.8.D PECHANICAL VACUUM PUMP The purpose of isolating the mechanical vacuum pump line is to limit the release of activity from the main condenser. During an accident, fission products would be transported from the reactor through the main steam lines to the condenser. The fission product radioactivity would be sensed by the main steam line radioactivity monitors which initiate isolation.

3.8.E and 4.8.E BASES The limitations on removable contamination for sources requiring leak testing, including alpha emitters, based on 10 CFR 70 39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use, with surveillance requirements comensurate with the proba-bility of damage to a source in that group. Those sources which are fre-j quently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

1 315 i

1

, -,,.,-c,--,..---,,.--,-,-,,,.,,,-a,nn,-- ,,n,,---,,--,,r-~,,-,,,,,--,n-_,nn.---, -

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.13 RADIOLOGICAL E RONMENTAL 4.13 RADIOLOGICAL ENVIRONMENTAL MONITORING MONITORING A. Menitering Program A. Monitorine Progra=

1. The radiological environ- The radiological mental monitoring environmental monitoring program shall be conducted samples shall be collected as specified in pursuant to Table 3.13.A from Table 3.13.A. the locations given in the table and figure in the ODCM

2. With the radiological and shall be analyzed environmental monitoring pursuant to the requirement program not being of Tables 3.13.A and the conducted as specified detection capabilities in Table 3.13.A, in required by Table 3.13.B.

lieu of a LER, prepare and submit to the Coc: mission, in the Annual Radiological Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

3. With the level of radioactivity in an environmental sampling medium exceeding the reporting levels of Table 3.13.C when averaged over any calendar quarter, in lieu of a LER, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter, a report which identifies the cause(s) for exceeding the limit (s) and defines ~

the corrective action to be taken to reduce radioactive effluents so that the potential annual dose to a member of the public is less than the calendar year limits of 359A

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQliIREMENTS specification 3.8.A.3, 3.8.B.3, and 3.8.B.S.

When more than one of the radionuclides in Table 3.13.C are detected in the sampling medium, this report shall be submitted if:

Cone (1) Cone (2)

Limit (1) Limit (2$***1

4. When radionuclides other than those in Table 3.13.C are detected and are result of plant effluents, the report in 3.13.A.3 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 specification 3.8.A.3, 3.8.B.3, and 3.8.B.5.

5. The report in specifi-cation 3.13.A.3 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.

l 6. With, milk or fresh leafy

. vegetable samples unavailable from one or more of the sample locations required by' Table 3.13.A, identify locations for obtaining replacement samples, if available, and add them to the radio-359B .

-- , - -,,----.g--,--- .,,w ,p,,e , nw ,,,n.nm v..-- ,. ,--y, _ _- ,,m ,--- ,,,,,- m _ w,-n-,,,ng,,-an,.e.,g.,,,_,-,,-r,--

I 1

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS logical 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 LER, identify the cause of the unavaila-bility of samples and identify the new location (s),

if available, for obtaining replacement sa=ples in the aext Annual Radiological Environmental Operating Report and also include a revised figure (s) and table (s) for the ODCM reflecting the new locations.

B. Land Use Census

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 meterological sectors within a distance of five miles. (For elevated releases as defined in Regulatory Guide 1.111, Revision 1, July 1977, the land use census shall also identify the locations of all milk animals and all gardens of greater than SCO square feet producing fresh leafy vegetables in each of the meteorological sectors within a distance of three miles.)

Broad leaf vegetation sampling may be perfo'rmed at the site boundary in the direction section yith the highest D/Q in lieu of the garden census.

359C

LIMITING CONDITIONS FOR OPERATION SURW.ILLANCE REQUIPIMEh"rS

2. With a land use census identifying a location (s) which yields a calculated .

dose or dose commitment greater than the maximum value currently being calculated in specifi-cation 4.8.B.4, in lieu of a LER, identify the new locations in the next Annual Radiological .

Environmental Operating Report.

3. With a land use census identifying a location (s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent.

greater than at a location from which samples are currently being obtained in accordance with specification 3.13.A, add the new location (s) to the radiological environmental monitoring programs within 30 days if the owner consents.

The sampling location (s),

excluding the control station location, having the lowest calculated dose or dose commitment (s)

(via the same exposure pathway) may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted. In lieu 4 of a LER, identify the new location (s) in the next Annual Radiological Environmental Operating Report and revised figures and tables for the ODCM reflecting the new location (s).

359D

~

i i

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

4. The land use census shall be conducted at least once per calendar year between the dates of April 1 and October 1 using the following techniques: ,

a. Within a 2 mile radius from the plant or within the 15 mrem per year ,

isodose line, whichever is larger, enumeration by a door-to-door or equivalent counting technique.

b. Within a 5 mile radius from the plant, enumeration by using appropriate techniques such as door-to-door survey, mail survey, tele-phone survey, aerial survey, or infor-cation from local agricultural authorities or other reliable sources.

i C. Interlaboratory Cornarison Program i 1. Analyses shall be performed l on radioactive materials I supplied as part of an

! Interlaboratory Conparison Program which has been j approved by the Commission.

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

I 359E l

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

3. A summary of the results obtained as part of the above required Inter-laboratory Comparison Program and in accordance with the ODCM (or participants in the EPA cross check program shall provide the EPA program code designation for the unit) shall be included in the Annual Radiological Environmental Operating Report.

D. Deviations From Samuling Schedule Deviations are permitted from the required sampling schedule if specimens are unobtainable due to harardous conditions, seasonal unavailability or malfunction of automatic sampling equipment. If the latter., every effort shall be made to complete corrective action prior to the end of the next sampling period.

All deviations from the sampling schedule shall be reported in the Annual Radiological Environmental Operating Report.

E. Exemption The provisions of Specification 1.0.C are not applicable.

359F

TABLE 3.13.A RADIOLOGICAL ENVIRONHENTAL HONITORING PROGRAM Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations

Collection Frequency of Analysis

1. AlHBORNE 4

Radioiodine and Minimum of 5 locations Continuous operation of Radiolodine canister.

Particulates sampler with sample Analyze at least once collection as required by per 7 days for I-131.

dust loading but at least once per 7 days. Particulate sampler, i Analyze for gross beta radioactivity}24 hours i

following filter change.

Perform gamma isotopic analysis on each sample when gross beta activity -

! is greater than 10 times the average of control 3

samples. Perform gamma l isotopic analysis on composite (by location) sample at least once per 92 days.

1 2. 1)IRECT RADIATION At least 40 locations At least once per 92 days. Ganuna dose. At least once with } 2 dosimeters per 92 days.

at each location.

1 I

i i

" Sample locations are given in the ODCH.

1 3590 1

TABI.E 3.13.A (C:ntinu d)

HADIOLOGICAL ENVIRONMENTAL HONITORING PROGRAM 1

Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations

Collection Frequency of Analysis

3. WATERBORNE b Gamma isotopic analysis

a. Surface 2 locations Composite nample collected over a period of f 31 days. of each composite sample.

Tritium analysis of com-d posite sample at least I

once per 92 days.

1

b. Drinking Hinimum of I downstream Compositeb sample collected

Gross beta and gamma over a period of f 31 days.

j location, or all water isotopic analysis of supplies within 10 miles each composite sample.

downstream which are Tritium analysis of

composite sample at least 1 taken from the Tennessee River. once per 92 days.

l I c. Sediment Hinimum of I location At least once per 184 days. Camma isotopic analysis of each sample.

I

d. Ground" i

" Sample locations are shown in the ODCH. .

j I' Composite samples s' all be collected by collecting an aliquot at intervals not exceeding 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

s 181 1 if drinking water is obtained within l ' Composite samples shall be collected over a period of { 14 days for

! 3 miles downstream of the plant. .

Ground water movement 'in the area has been determined to be from the plant site toward the Tennessee River.

j Since no drinking water wells exist between the plant and the river, ground water will not be monitored.

l I

l 35911 1

TAHLE 3.13. A (Continued)

RADIOLOGICAL ENVIRdNNENTAL HONITORING PROGRAM Number of Samples Exposure Pathway and Sampling and Type and Frequency.

and/or Sample Sample Locations

Collection Frequency of Analysis I

k 4. INGESTION I a. Hilk 3 locations At least once per 15 days 1-131 analysis of each when animals are on pasture; sample. Gamme isotopic at least once per 31 days analysis at least once per 31 days.

at other times.

b. Fish 2 samples One sample in season, or at Gamma isotopic analysis least once per 184 days if on edible portions.

not seasonal. One sample of commercial and game species.

i

c. Food Products" 2 locations At Icast once per year Gamma isotopic analysis

] at time of harvest, on edible portion.

i

)

i i

a

" Sample locati ms are shown in the 011Cll.

.i

[ "Since water from the Tennessee River in the immediate area downstream is not used for irrigation : purposes, the

! sampling of food products (primarily broad Icaf vegetation) is not required unless milk sampling is not performed.

i 1

3591

TABLE 3ol3.B HAXIHUH VALilES FOR TIIE IJMJER L1 HITS OF DETECTION (LLD)"'"

Airborne Particulate Water or Gas Fish Hilk Food Products Sediment Analysis (pC1/1) a (pCi/m ) (pCi/kg, wet) (pCi/1) (pCi/kg, vet) (pci/kg, dry) gross beta 4 1 x 10 2 N.A. N.A. N.A. N.A.

II-3 2000 N.A. N.A. N.A. N.A. N.A.

Hn-54 15 N.A. 130 N.A. N.A. N.A.

1 Fe-59 30 N.A. 260 N.A. N.A. N.A.

i Co-58, 60 15 N.A. 130 N.A. N.A. N.A.

2n-65 30 N.A. 260 N.A. N.A. N.A.

Zr-95 30 N.A. N.A. N.A. N.A. N.A.

i

! Nin-95 15 N.A. N.A. N.A. N.A. N.A.

I 2 N.A. 60 N.A.

) .I-131 I' 7 x 10 I Cs-134 15 5 x 10 2 130 15 60 150 Cs-137 18 6 x 102 150 18 80 , 180 l

i Ila-140 60 N.A. N.A, 60 N.A. H.A.

I.a- 140 15 H.A. N.A. 15 N.A. N.A.

i

+

t l 359J l

1 i

e

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

TABLE 3.13.B (Continued)

TABLE NOTATION

a. The LLD " the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

4.66 s b E V 2.22 Y exp (-Aat)

Were:

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

^

picoeurie per unit mass or volume),

sg is the standard deviation of the background counting rate or of t!ie counting rate of a blank sample as appropriate (as counts per minute),

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

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

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

A is the radioactive decay' constant for the particular radionuclide, and l

At is the elapsed time between sample collection (or end of the l

sample collection period) and time of counting (for environmental samples, not plant effluent samples).

It should be recognized that the LLD is defined cs a a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement.

i a %

g k

f4 ig- w 359K q ..

% g '*p t i

'+

\ - - -

y

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

TABI.E 3.13.E (Continued)

TABLE NOTATION

b. The LLD for analysis of drinking water and surface water samples shall be performed by gamma spectroscopy at approximately 15 pCi/L. If levels greater than 15 pCi/I, are identified in surface water samples downstream from the plant, or in the event of an unanticipated release of I-131, drinking water samples will be analyzed at an LLD of 1.0 pCi/L for I-131.

c. Other peaks which are measurable and identifiable, together with the radienuclides in Table 3.13.B shall be identified and reported.

I e

.. = . . . . - . - _ . . .-

4 TABLE 3.13.C REPORTING LEVELS FOR RADIDACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SA Reporting Levels Water .Mrborne Particulate Fish Milk Food Produ

& cts (pci/Kg, w

] Analysis (pCi/1) or Cases (pCi/m 8) (pci/Kg, wet) _(pCI/1) l Set) iI H-3 2 x 10 4I'} N.A. ,

N.A. N.A. N.A.

l Hn-54 1 x 10 3 N.A. 3 x 104 N.A. N.A.

! Fe-59 4 x 102 N.A. 1 x 10 4 N.A. N.A.

Co-58 1 x 10 8 N.A. '

3 x 10 4 N.A. N.A.

O Co-60 3 x 10 2 N.A. I x 10 4 N.A. N.A.

Zn-65 3 x 10 2 N.A. 2 x 10 4 N.A. N.A.

Zr-Nb-95 4 x 102 N.A. N.A. N.A. N.A.

]

I 1-131 2 0.9 N.A. 3 1 x 10

! s2 Cs-134 30 to 1 x 10 8 60 1 x 10 1

~

$3 i

Cs-137 50 20 2 x 10 8 70 2 x 10

)

J g3 Ba-La-140 2 x 102 N.A. N.A. 3 x 10 8 N.A.

j d

(a)For drinking water sampics. This.is 40 CFR Part 141 value.

]

a

1 BASES

1. Monitoring Program The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides, which lead to the highest potential radiation exposures of individuals resulting from the station operation. This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measureable' concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the evironmental exposure pathways.

4 The detection capabilities required by Table 313.B are state-of-the-art

, for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of the measurement system i and not as an a posteriori (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

2. Land Use Census 2

This specification is provided to ensure that changes in the use of unrestriced areas are identified and that modifications to the monitoring program are made if required by the results of this census.

The best survey information from the door-to-door, mail, telephone, aerial or consulting with local agricultural authorities shall be used.

This census satisfies the requirements of Section IV.B.3 of Appendix I r 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 vegetation assumed in Regulatory Guide 1.109 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 j broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a l vegetation yield of 2kg/ square meter.

3 _Interlsboratory Comparison Program The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

359N f

i l

l

[0 ADMINISTRATIVE CONTROLS

k. The radiological environmental monitoring program and the results thereof at least once per 12 months.

1. The performance of activities required by the Quality Assurance Program to meet the criteria of Regulatory Guide 4.15, December 1977 or Regulatory Guide 1.21, Rev. 1, 1974 and Regulatory Guide 4.1, 1975 at least once per 12 months.

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

n. The Process Control Program and implementing procedures for solidification of wet radioactive wastes at least once per 24 months.

9. AUTHORITY The NSRB shall report to and advise the Manager of Power on those areas of responsibility specified in Sections 6.2.A.7 and 6.2.A.8.

10. RECORDS Records of NSRB activities shall be prepared, approved and distributed as indicated below:

a. Minutes of each NSRB meeting shall be prepared, approved and forwarded to the Manager of Power within 14 days following each meeting.

b. Reports of reviews encompassed by Section 6.2.A.7 above, shall be prepared, approved and forwarded to the Manager of Power within 14 days following completion of the review.

c. Audit reports encompassed by Section 6.2.A.8 above, shall be forwarded to the Manager of Power and to the management positions responsible for the areas audited within 30 days after completion of the audit.

4 364A

. , _ _ , _ , - - . - - - _ . - . . . _ , . , _ , , , _ , , , , . _ . . - - . . . _ , , , . y _ _ _ _ _ _ _ _ , ,_ ,_y _ - .

- . _ u -

6.0 ADMINISTRATIVE C0hTROLS

j. Review adequacy of employee training program and reco:nmend change.

h. Review adequacy of the Process Control Program and Offsite Dose Calculation Manual at least once every 24 months.

1. Review changes to the radwaste treatment systems.

m. Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recommendation, and deposition of the corrective action to prevent recurrence to the Director, Nuclear Power and to the Nuclear Safety Review Board.

5. Authority The PORC shall be advisory to the plant superintendent.

6. Records Minutes shall be kept for all PORC meetings with copies sent to Director, Nuclear Power; Assistant Director of Nuclear Power (Operations); Chairman, NSRB.

~

7. Procedures Written administrative procedures for committee operation shall be prepared and maintained describing the method for submission and content of presentations to the committee, review and approval by members of committee actions, dissemination of minutes, agenda and scheduling of meetings.

t i

367

- - - - _ =_. ...

6.0 ADMINISTRATn'E C0hTROLS 6.3 Procedures A. Detailed written procedures, including applicsble checkoff lists covering items listed below shall be prepared, approved and adhered to.

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

2. Refueling operations.

3. Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms, suspected primary system leaks and abnormal reactivity changes.

4. Emergency conditions involving potential or actual release of Radioactivity.

5. Preventive or corrective maintenance operation's which could have an effect on the safety of the reactor.

6. Surveillance and testing requirements.

7. Radiation control procedure.

8. Radiological Emergency Plan implementing procedures.

9. Plant security program implementing procedures.

10. Fire protection and prevention procedures.

11. Process Control Program (PCP).

-12.. Offsite Dose Calculation Manual.

B. Written procedures pertaining to those items listed above shall be reviewed by PORC and approved by the plant superintendent prior to implementation. Temporary changes to a procedure may be made by a member of the plant staff knowledgeable in the area affected by the procedure except that temporary changes to those items listed above except item 5 require the additional approval of a member of the plant staff who holds a Senior Reactor Operator license on the unit j affected. Such changes shall be documented and subsequently

, reviewed by PORC and approved by the plant superintendent.

368 i

6.0 ADMINISTRATIVE CONTROLS 6.3 Procedures E. Process Control Program (PCP)

1. The PCP shall be approved by the Commission prior to implementation.

2. Changes to the PCP shall be submitted to the Co= mission in the semi-annual Radioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:

a. Sufficiently detailed information to totally support the change.

b. A determination that the change did not change the overall conformance of the solidified product to existing criteria.

3. Changes to the PCP shall become effective upon review and acceptance by PORC.

F.

Quality Assurance Procedures - Effluent and Environmental Monitoring Quality Assurance procedures shall be established, implemented,- and maintained for effluent and environmental monitoring, using the guidance in Regulatory Guide 1.21, rev. 1, June 1974 and Regulatory Guide 4.1, rev. 1, April 1975 or Regulatory Guide 4.15, Dec. 1977.

a o

G 9

e 370

6.0 ADMINISTRATIVE CONTROLS 6.4 Actions to be Taken in the Event of a Reportable Occurrence in Plant Operation (Ref. Section 6.7) _ , , , ,

A. Any reportable occurrence shall be promptly reported to the Manager, Nuclear Production and shall be promptly reviewed by PORC. This committee shall prepare a separate report for each s reportable occurrence. This report shall include an evaluation of the cause of the occurrence and recom=endations for appropriate action to prevent or reduce the probability of a repetition of the occurrence.

B. Copies of all such reports shall be submitted to the Manager, Nuclear Production, the Manager of Power and the Chairman of the NSR3 for their review.

C. The plant superintendent shall notify the NRC as specified in Specification 6.7 of the circumstances of any reportable occurrence.

6.5 Action to be Taken in the Event a Safetv Limit is Exceeded If a safety limit is exceeded, the reactor shall be shut down and .

reactor operation shall not be resumed until authorized by the NRC. A prompt report shall be made to the Manager, Nuclear Production, and the Chairman of the NSR3. A complete analysis of the circumstances leading up to and resulting from the situation, together with recommendations to prevent a recurrence, shall be prepared by the PORC. This report shall be submitted to the Manager, Nuclear Production, the Manager of Power, and the NSRB.

Notification of such occurrences will b2 made to the NRC by the plant superintendent within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

6.6 Station Operating Records A. Records and/or logs shall be kept in a manner convenient for review as indicated below: (Items 1-6, 5 years; Items 7-17, duration of the operating license).

1. All normal plant operation including such ite=s as power ,

level, fuel exposure, and shutdowns e

2. Principal maintenance activities

3. Reportable occurrences

4. Radioactive shipments

5. Test results, in' units of microcuries, for leak tests performed pursuant to Specification 3.8.E 376

6.0 ADMINISTRATIVE C0hTROI,5

6. Record of annual physical inve'ntory verifying au.ountability of sources on record

7. Checks, inspections, tests, and calibrations of components and systems, including such diverse items as source leakage

6. Reviews of changes made to the procedures or equipment or reviews of tests and experiments to comply with 10 CFR 50.59

9. Gaseous and liquid radioactive waste released to the enirons

10. Off-site environmental monitoring surveys

11. Fuel inventories and transfers

12. Plant radiation and contamination :urveys

- 13. Radiation exposures for all plant personnel

14. Updated, corrected, and as-build drawings of the plant

15. Reactor coolant system inservice inspection

16. Minutes of meetings of the Nuclear Safety Review Board

17. Design fatigue usage evaluation I

a. Monitoring,. recording, evaluating, and reporting requirements contained in 17.b, below will be met for various portions of the reactor coolant pressure boundary (RCPB) for which detailed fatigue usage evaluation per the ASME Boiler and Pressure Vessel Code Section III was performed 1 for the conditions defined in the design specification. In this l

t r

l 1 See paragraph N-415.2, ASME Section III, 1965 Edition.

377

6.0 ~ AT!IUISTRATIVE CONTROLS 5.7.1 Routine Reoorts

b. Annual Ooeratinc Report A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem / year and their associated man rem exposure accordin5 to work and job function, e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignment to various duty functions

'may be estimates based on pocket dosimeter, TLD, or file badF e ceasurements. Small exposures totalling less than 20 % of the individual total dose need not be accounted for. In the aggregate, at least 80 % of the total whole body dose received from external sources shall be assigned to specific major work functions,

c. Monthlv Oserating Report Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Office of Inspection and Enforcement, U.S. Nuclear Regulatory Com=ission, Ucshington, D.C. 20555, with a copy to the appropriate Regional Office, to be submitted no later than the 15th of each month following the calendar month covered by the report. A narrative summary of operating experience shall be submitted in the above schedule.

d. Annual Radiological Environmental Monitoring Recorts

1. Routine Annual Radiological Environmental Operating Reports covering the operation of the plant during the previous calendar year shall be submitted prior to Ma3 1 of each year.

2. The Annual Radiological Environmental Operating Reports shall include sumnaries, interpretations,.and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by Specification. 3 13 3.1. If harmful effects or evidence of irreversible damage are detected by the monitorin6, the report shall provide an analysis of the problems and a planned course of action to alleviate the problem.

1 381

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3 The .a.nnual Radiological Envirennenta) operating Reports shall include su=narized and tabulated results in the for at of Regulatory Guide t.8, Decenber 1975 of all radiological environmental samples taken during the report period. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

4. The reports shall also include the following: a su=r.ary description of the "adiologietl environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions from one reactor; and the results of licensee participation in the Interlaboratory Comparison Program required by Sepcifications 3.13.c.

6.7.2 Rencrtable Occurrences Reportable occurrences, including coarective actions and measures to prevent reoccurrence, shall be reported to the NRC.

Supplemental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date.

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6.0 ADF.INISTRATIVE CONTROLS

a. Promet Notification with Written Fol'owup (9) Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analysis in the safety analysis report or technical specifications bases; or discovery during plant life of conditions not specifically considered in the safety analysis report 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.

(10) The concentration of radioactive material in liquid effluents released to unrestricted area exceeds the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radiomtelides other than dissolved or entrained noble gases. Concentration of dissolved or entrained noble gases exceeds 2 x 10 ~4 pCi/ml total cctivity.

(11) Offsite releases of radioactive material in gaseous effluents which exceed the following limits:

(a) The dose rate for nobles gases equals or exceeds 500 mrem /yr to the total body or 3000 mrem /yr to the skin.

(b) The dose rate of all radioiodines, for all radioactive materials in particular form, and for radionuclides other than nobles gases with half lives greater than 8 days exceeds 1500 mrem /yr to any organ.

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f 6.0 ' ADMINISTRATIVE CONTROLS 6.7.2 Reportable Occurrences

b. Thirty-Day Written Reports. The reportable occurrences discussed below shall be the subject of written reports to the Director of the appropriate Regional Office within thirty days of occurrence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.

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

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

Note: Routine surveillance testing, instrument calibration, or preventative maintenance which require system configurations as described in items 2.b.(1) and.2.b.(2) need not be reported except where test results themselves reveal a degraded mode as described above.

(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 systems, i

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6.0- ADMINISTRATIVE CONTROLS 6.7.2.b Thirty Day Written Reports (4) Abnormal degradation of systems other than those specified in item 2.a(3) above designed to contain radioactive material resulting from the fission process.

Note: Sealed sources or calibration sources are not i.ncluded 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.

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

1. A description of the event of equipment involved.

2. Cause(s) for the unplanned release.

3. Actions taken to prevent recurrence.

4. Consequences of the unplanned release.

384A

6.0 ADMINISTRATIVE CONTROLS 6.7 Reporting Recuirements

3. Unione Reoorting Recuirements A. Radioactive Effluent Release Report A report on the-radioactive discharges released from the site during the previous 6 months of operation shall be submitted to the Director of the Regional Office of Inspection and Enforcement within 60 days after January 1 and July 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gasecus effluents released and solid waste shipped from the plant as delineated in Regulatory Guide 1.21, Revision 1, " Measuring,

, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants,"

with data summarized on a quarterly basis following,the format of Appendix B thereof.

The report shall include a summary of the meteorological conditions concurrent with the release of gaseous effluents durine each quarter as outlined in Regulatory Guide 1.21, Revision 1, with data summarized on a quarterly basis following the format of Appendix B thereof. Calculated offsite dose to members of the public resulting from the release of liquid and gaseous effluents and their subsequent dispersion in the river and atmosphere shall be reported as recommended in Regulatory Guide 1.21, Revision 1. The Radioactive Effluent Release Report shall include the following information for each type of solid waste shipped offsite during the report period (a) container volume, (b) total curie quantity, (specify whether determined by measurement or estimate, (c) principal radionuclides (specify whether determined by measurement or estimate), (d) sources of waste and processing employed (e.g. dewatered spent resins, compacted dry waste, etc.), (e) type of container (e.g., LSA, Type A, Type B, large quantity), and '

(f) solidification agent or absorbant (e.g. concrete, urea formaldehyde, etc.).

e 385 s

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6.0 ADMINISTRATIVE CONTROLS E. Special Resorts: Radiological Environmental Moninoring If measured levels of radioactivity in an environnental sampling medium are determined to exceed the reporting level values of Table 3.13.C when averaged over any calendar quarter sampling period, a report shall be submitted to the Commission pursuant to Specification 3.13.A.3.

FOOTNOTES

1. A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.

2. The term " forced reduction in power" is normally defined in the electric power industry as the occurrence of a component failure or other condition which requires that the load on the unit be reduced for corrective action immediately or up to and including the very next weekend. Note that routine preventive maintenance, surveillance, and calibration activities requiring power reductions are not covered by this section.

3. The term " forced outage" is normally defined in the electric power industry as the occurrence of a component failure or other condition l

which requires that the unit be removed from service for corrective action immediately or up to and including the very next weekend.

4. This tabulation supplements the requirements of $20.407 of 10 CFR Part 20.

l 5. Each integrated leak rate test of the secondary containment shall be the I subject of a summary technical report. This report should include data on the wind speed, wind direction, outside and inside temperatures during the test, concurrent reactor building pressure, and emergency ventilation flow rate. The report shall also include analyses and interpretations of those data which demonstrate compliance with the specified leak rate limits.

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, h Nm-APPENDIX B TECHNICAL SPECIFICATIONS BROWNS FERRY UNIT 3

ENVIRONPdNTAL TECHNICAL SPECIFICATIONS FOR I EROWNS FERRY NUCLEAR PLANT TABLE OF CONTENTS Page No.

..... . Deleted 1.0 DEFINITIONS . . . . . . . . . . . . .. . . .

Deleted 2.0 LIMITING C0 EDITIONS FOR OPERATION . . . . . . . . . . .

2.1 Thermal Discharge Limits . . . . . . . . . . . . .

Deleted Deleted 2.2 Chemical . . . . . . . . . . . . . . . . . . . . .

2.2.1 Hakeup Water Treatment Plant Spent Demineralizer Regerants . . ....... Deleted Deleted 2.2.2 Ch2orine .................

...... Deleted 30 DESIGN FEATURES AND OPERATING PRACTICES . .

Deleted 31 Chemic al U s a ge . . . . . . . . . . . . . . . . . .

Deleted 3 1.1 Oils and Hazardous Materials. . . . . . . .

.... ...... Deleted 3 1.2 Other Chemicals . . . .

Deleted 32 Land Management. . . . . . . . . . . . . . . . . .

Power Plant Site. .............. Deleted 3 2.1 1

3 2.2 Transmission Line Right-of-Way Maintenance.

33 Onsite Meteorological Monitoring . . . . . . . . . Deleted

.......... Deleted 4.0~ ENVIRONMENTAL SURVEILLANCE. . . . .

Deleted 4.1 Ecological Surveillance. . . . . . . . . . . . . .

4.1.1 Abiotic . ................. Deleted 4.1.2 Biotic. .................. ' Deleted Deleted 4.1.3 Special Studies . ... .... ......

4.2 Radiological Environmental Monitoring Program. . . Deleted

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L 50 ADMINISTRATIVE CONTROLS . . . . . ........... Deleted 51 Responsibility . . . . . . . . . . . . . . . . . . Deleted 52 organization . . . . . . . . . . . . . . . . . . . Deleted 53 Review and Audit . . . . . . ........... Deleted 5.4 Action to be Taken if an Environment LCO is Exceeded . . . . . . . . . ........... Deleted 55 Procedure. . . . . . . . . . . . . . . . . . . . . Deleted 5.6 Reporting Requirements . . . . ... . ....... Deleted 57 Environmental Records. . . . . . . . . . . . . . . Deleted Tabl es . . . . . . . . . . . . . . . . . . . . . . . . . . . Deleted Figures . ... ...................... Deleted 9

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4 3 2.2 Transmission Line Right-of-Way Maintenance i

Objective ,

The sole purpose of this section is to provide reporting requirements (to USNRC) on herbicide usage, if any, for purposes of '

right-of-way maintenance regarding only those transmission lines under USNRC's jurisdiction for the Browns Ferry Nuclear Plant.

Specification.

A statement as to whether or not herbicides have been used in maintaining rights-of-way for those transmission lines associated with the Browns Ferry Nuclear Plant shall be provided. If -

herbicides have been used, a description of the types, volumes, i concentrations, manners and frequencies of application, and miles or rights-of-way that have been treated shall be included.

Reporting Requirements Information as specified above shall be provided in the Annual Operating Report (Appendix A, Section 6.7.1.(b)).

Bases Vegetation growth on a transmission line right-of-way must be controlled in such a manner that it will neither interfere with safe and reliable operation of the line or impede restoration of service when outages occur.

Vegetation growth is controlled by mechanical cutting and the limited use of herbicides. Selected chemicals approved by EPA for use as herbicides are assigned (by EPA) label instructions which provide guidance on and procedures for their use.

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ENCLOSURE 2 DESCRIPTION OF CHANGES AND JUSTIFICATION TVA BFNP TS 192 BROWNS' FERRY NUCLEAR PLANT Description of Proposed Technical Specification Changes The proposed changes to Appendix A add new limiting conditions for operation and corresponding surveillance requirements for the monitoring, recording, and reporting of radiological effluents. Requirements for radiological environmental monitoring, presently contained in Appendix B are proposed for incorporation into Appendix A. The applicable portions of Appendix B are therefore deleted.

Justification of Proposed Changes The proposed technical specifications for radiological effluent monitoring are based on NUREG-0473 and in a meeting between TVA and the NRC staff held at Browns Ferry Nuclear Plant in August of 1982. The majority of the proposed changes are submitted in response to NUREG-0473 as requested by the NRC staff. Several changes involve clarification of existing specifications. Such clarification should reduce the possibility of misinterpretation of the technical specifications.

ENCLOSURE 3 SIGNIFICANT HAZARDS CONSIDERATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES DESCRIPTION OF AMENDMENT REQUEST This amendment would make changes to the technical specifications by adding new limiting conditions for operation and corresponding surveillance requirements to conform to NRC requirements fcr radiological effluent technical specifications. These changes were requested by NRC.

SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The proposed specifications incorporate new requirements for radiological effluent monitoring, recording, and reporting. Based on that the probability or consequences of an accident previously evaluated would not be increased.

2. Does the proposed amendment create the probability of a new or different kind of accident from any accident previously evaluated?

No. The new requirements being added for monitoring radiological effluents do not create any probability for a new or different kind of accident from those previously evaluated.

3 Does the proposed amendment involve a significant reduction in a margin of safety?

No. Because new and more restrictive monitoring requirements for effluents are being added the margin of safety would conceivably be increased. No reduction in any margin of safety can be envisioned from these proposed changes.

BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The Commission hac provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples (48 FR 14870). The examples of actions involving no significant hazards consideration include, ". .

.(vii) A change to make a license conform to changes in the regulations, where the license change results in very monor changes to facility operations clearly in keeping with the regulations." Another example of a change involving no significant hazards consideration is: "(ii) A change that constitutes an additional limitation, restriction, or control not presently included in the technical specifications: for example, a more stringent surveillance requirement." The changes proposed in the application for amendment are encompassed by these examples in that the proposed change would add limiting conditions for operation and surveillance requirements for environmental radiological monitoring.

Therefore, since the application for amendment involves a proposed change that is similar to examples for which no significant hazards consideration exists, the staff has made a proposed determination that the application involves no significant hazards consideration.

ENCLOSURE 4 PROPOSED JUSTIFICATIONS FOR TAKING EXCEPTIONS TO CERTAIN REQUIREMENTS OF NUREG-0473 ~

BROWNS FERRY NUCLEAR PLANT

1. TVA takes exception to Sectica Table 3.7.11.1 of NUREG-0473 R3. It is not our intention to provide radiation monitors for the three turbine building f.tation sumps at Browns Ferry. There is one sump for each of the'three units located on elevation 557 feet of the turbine building. These sumps are located adjacent to their respective units raw cooling water pumps. These pumps are

. physically isolated from other plant equipment (see attached drawings). Besides the floor drainage from the raw water pumps, the sumps have piped inputs from the raw service water and raw cooling water pumps. Other principal inputs are:

A. Miscellaneous air handling system drains B. Miscellaneous clean shop drains C. Fire protection system flush drain D. Air compressor drains The three turbine building station sumps are listed and monitored as

, outfalls on the National Pollution Discharge Elimination System Permit.

2. TVA takes exception to Section 3.11.2.7 and 3.3.7.12 of NUREG-0473 R3. It is not our intention to request addition of the SJAE radiation monitor located on the discharge of the steam jet air ejector to the technical specifications. Deletion of this monitor and its recommended alarm / trip (329,300 UCI/SEC) point is justified because:

In the unlikely event of a break in the integrity of the SJAE discharge piping inside the turbine building, the amount of activity released prior to operations awareness is presently j limited to a small fraction of the 329,300 UCI/SEC limit proposed by the RETS.

Continuous air monitor 90-250 receives approximately 37% of its sample flow for the turbine building exhaust. The monitor setpoint for noble gas is approximately 10% of technical specification limit or 0.013 E+6 UCI/SEC. If no activity is assumed to reach the CAM from other sources, then the turbine building release rate would be 0.035 E+6 UCI/SEC before the CAM alarmed. Maximum flow is assumed in the turbine building duct (1.2 E+8 CC/SEC). The allowable

-activity divided by the flow rate yields an allowable concentration of 6.13 E-4 UCI/CC in the building exhaust. Assuming all this activity came from a 50 SCFM leak in the offgas system, then the concentration of the activity in the offgas system would have to be 1.49 E+0 UCI/CC. Plant data indicates that an offgas release rate (at SJAE discharge) of 100,000 UCI/SEC corresponds to a concentration of approximately 4 E+0 UCI/CC. Therefore, if 329,300 UCI/SEC were reached, the concentration would be 1.28 E+1 UCI/CC.

As shown above, a concentration of 1.49 I+0 is sufficient to cause the continuous air monitor to alar =. This is S.6 times lower than the concentration per=itted by RETS. Even if the setpoint on the CAM was increased to 100% of the technical specification limit (unlikely because of cultiple release points), the monitor would still alarm at concentrations below those permitted by RETS (1.28 E+1 UCI/CC). Based on these calculations and our response to Bulletin 78-03 concerning early detection offgas hydrogen leaks, we conclude that any significant leak in the offgas system would be detected and stopped quickly even at levels below those allowed by RETS.

3. It is not our intention to include the requirement for sa=pling and analyzing charcoal and particulate filters based on unit operational transients. Our justification for this position is as follows:

With the exception of the main stack, all airborne effluent continuous air monitors have halogen and particulate instrument channels associated with the installed filters. Our plant procedures require that the chenical laboratory retrieve and analyze any installed filter following receipt of its associated instrument channel alarm. Alarm setpoints are maintained at a 4

fraction of the technical specification halogen / particulate release limit for conservatism. The need for increased sampling and analysis of filters usually has resulted from the development of system leaks. Past operating experience for our three units does not support the need for an increased sampling frequency based on operational transients. The present laboratory staffing may not be adequate to support the proposed requirement. Additionally, the laboratory does not have instrumentation to monitor unit status. Imposition of a sampling program tied to unit status would have to be implemented administratively. Each failure to correctly respond to any operational transients on any of our three units will require submittal of an licensee event report.

4. -TVA takes exception to footnote E of Table 4.11-2 of NUREG-0473 R3.

It is not our intention to take the additional tritium grab samples specified in the footnote. Our justification for this exception is given below:

Since the first refueling outage at Browns Ferry Nuclear Plant, there has always been spent fuel stored onsite. Therefore, the effect of the proposed footnote would be to permanently increase the present monthly sampling frequency to weekly. A review of data taken since 1973 indicates that tritium is not present in large concentrations at Browns Ferry Nuclear Plant. Typically tritium concentrations are much lower than 1 E-8 UCI/CC. Any variations in concentration levels have not been directly related to fuel handling activities.

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5. TVA takes exceptica to the requirement for the IE-6 UCI/C IID on l the stack noble gas monitor (ref. W.dG Table 4.11-2)..

TVA proposes to specify an IID of lE-5 UCI/CC for the continuous radioactive effluent monitor located in the stack. The inputs to the stack are from known sources of activity (G.E. offgas systems and steam packing exhauster). Since there is always some slight release of activity, there is no. need to specify an IID for this instrument. Low II.D values are needed to monitor effluents having no or.very small concentrations of activity.

The results of actual analysis on stack effluents indicate that the monitors are capable of detecting concentrations below IE-5 UCI/CC.

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v t e Letter Sequence Other
TAC:63022, (Open) TAC:63023, (Open) TAC:63024, (Open)
Initiation Request Acceptance... Administration Questions Answers Results Approval... Other: ML20081J154
MONTHYEAR1983-10-27 [Table View]

ML20081J154

Text

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