Text

Draft Radiological Effluent Tech Specs
	ML20063N338
Person / Time
Site:	North Anna
Issue date:	09/16/1982
From:	; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	;
Shared Package
ML20063N333	List: ML20063N332; ML20063N338; ML20063N339;
References
	NUDOCS 8209200219
	Download: ML20063N338 (232)
	v • d • e

ATTACle!ENT 1 NORTH ANNA UNITS 1 AND 2 RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS 8209200219 820916 PDR ADOCK 05000338 P PDR

INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS Action............................................................. 1-1 Axial Flux Difference ............................................. 1-1 Channel Calibration................................................ 1-1 Channel Check...................................................... 1-1 Channel Functional Test............................................ 1-1 Containment Intergrity............................................. 1-1 Controlled Leakage................................................. 1-2 Core Alteration.................................................... 1-2 I

Dose Equivalent I-131.............................................. 1-2 E-Average Disintegration Energy.................................... 1-2 Engineered Safety Feature Response Time............................ 1-3 Frequency Notation................................................. 1-3 l

Caseous Radwaste Treatment System.................................. 1-3 Identified Leakage................................................. 1-3 I

I l Member (s) of the Public............................................ 1-3 Offsite Dose Calculation Manual (0DCM)............................. 1-4 Operable - Operability............................................. 1-4 Operational Mode - Mode............................................ 1-4 b

Physics Tests...................................................... 1-4 Pre s su re Bound ry Leaka ge . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 Process Control Program (PCP)...................................... 1-4 Purge-Purging...................................................... 1-4 Quadrant Power Tilt Ratio.......................................... 1-5 NORTH ANNA - UNIT I I

l INDEX (Cont'd.)

l PAGE Rated Thermal Power................................................ 1-5 Reactor Trip System Response Time.................................. 1-5 Reportable Occurrence.............................................. 1-5 Shutdown Margin.................................................... 1-5 Site Boundary...................................................... 1-5 Solidification..................................................... 1-5 Source Check....................................................... 1-5 Staggered Test Basis............................................... 1-6 Thermal Power...................................................... 1-6 Unidentified Leakage............................................... 1-6 Unrestricted Area.................................................. 1-6 Ventilation Exhaust Treatment System. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 Venting............................................................ 1-6 i

b NORTH ANNA - UNIT 1 II

INDEX I

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION g

2.1 SAFETY LIMITS Reactor Core...................................................... 2-1 Reactor Coolant System Pressure................................... 2-1 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip Setpoints............................................ 2-5 BASES SECTION Pace 2.1 SAFETY LIMITS

, Reactor Core...................................................... B 2-1 Reactor Coolant System Pressure................................... B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip Setpoints............................................ B 2-3 NORTH ANNA - UNIT 1

~., - , , - - . w,. . - .,,.. . , e- .,,,.. --._,.., ,.e,., , , . --,,,4, ,. , -, . - .- .m.. , ,, .. - , , . . ,-- ,,,,, y

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY ............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - T,yg >200*F........................... 3/4 1-1 Shutdown Margin - T,yg <200*F........................... 3/4 1-3 Boration Dilution - Reactor Coolant Flow ................. 3/4 1-4 Baron Dilution - Valve Position .......................... 3/4 1-5 Moderator Temperature coefficient ........................ 3/4 1-6 Minimum Temperature for Criticality ...................... 3/4 1-7 3/4.1.2 B0 RATION SYSTEMS Flow Paths - Shutdown .................................... 3/4 1-8 Flow Paths - Operating ................................... 3/4 1-9 Charging Pump - Shutdown ................................. 3/4 1-11 Charging Pumps - Operating ............................... 3/4 1-12 Boric Acid Transfer Pumps - Shutdown ..................... 3/4 1-13 Boric Acid Transfer Pumps - Operating .................... 3/4 1-14 Borated Water Sources - Shutcrvn ......................... 3/4 1-15 Borated Water Sources - Operating ........................ 3/4 1-16 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height ............................................. 3/4 1-18 Position Indicator Channels - Operating .................. 3/4 1-21 Position Indicator Channels - Shutdown ................... 3/4 1-22 Rod Drop Time ............................................ 3/4 1-23 Shutdown Rod Insertion Limit ............................. 3/4 1-24 Control Rod Insertion Limits ............................. 3/4 1-25 NORTH ANNA - UNIT 1 IV

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS I 3 /4 . ." Axial Flux Difference .................................. 3/4 2-1 3/4.2.2 Heat Flux Hot Channel Factor ............................ 3/4 2-5 3/4.2.3 Nuclear Enthalpy Hot Channel Factor ..................... 3/4 2-9 3/4.2.4 Quadrant Power Tilt Ratio ............................... 3/4 2-12 l 3/4.2.5 D N B Pa rame t e r s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 2-14 3/4.2.6 Axial Powe r Dis tribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 2-16 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION ..................... 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION .................................. ....... 3/4 3-15 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring .................................... 3/4 3-35 Movable Incore Detectors ................................ 3/4 3-39 Seismic Instrumentation ................................. 3/4 3-40 Meteorological Instrumentation .......................... 3/4 3-43 Auxiliary Shutdown Panel Monitoring Instrumentation ..... 3/4 3-46 Accident Monitoring Instrumentation ..................... 3/4 3-49 Fire Detection Instrumentation .......................... 3/4 3-52 Axial Power Distribution Monitoring System .............. 3/4 3-54 Loose Parts Monitoring System ........................... 3/4 3-56 Radioactive Liquid Effluent Monitoring Instrumentation .. 3/4 3-58 Radioactive Gaseous Effluent Monitoring Instrumentation . 3/4 3-65 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR CCOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation ............................. 3/4 4-1 H o t S t andb y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 4 -2 i

j S hu t d o wn . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 4 -3 I s o la t e d Lo o p . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 4 -4 i

Isolated Loop Startup ................................... 3/4 4-5 NORTH ANNA - UNIT 1 V

INDEX a

4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i

SECTION PAGE 3/4.4.2 SAFETY VALVES - SHUTDOWN ................................ 3/4 4-6 3/4.4.3 SAFETY VALVES - OPERATING ............................... 3/4 4-7 SAFETY and RELIEF VALVES - OPERATING Safety Valves ........................................... 3/4 4-7 Relief Valves ........................................... 3/4 4-7a 3/4.4.4 PRESSURIZER ............................................. 3/4 4-8 3/4.4.5 STEAM GENERATORS ........................................ 3/4 4-9 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems ............................... 3/4 4-16 Operational Leakage ..................................... 3/4 4-17 3/4.4.7 CHEMISTRY ............................................... 3/4 4-19 3/4.4.8 SPECIFIC ACTIVITY ....................................... 3/4 4-22 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System .................................. 3/4 4-29 Pressurizer ............................................. 3/4 4-30 Overpressure Protection Systems ......................... 3/4 4-31 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1, 2 and 3 Components ................... 3/4 4-33 Steam Generator Supports ................................ 3/4 4-35

3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS ............................................ 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T, 3,350*F .......................... 3/4 5-3 i 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 350*F .......................... 3/4 5-6 i

h/4.5.4 BORON INJECTION SYSTEM Boron Injection Tank .................................... 3/4 5-7 Heat Tracing ............................................ 3/4 5-8 3/4.5.5 REFUELING WATER STORAGE TANK . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-9 NORTH ANNA - UNIT 1 VI

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 CONTAINMENT Containment Integrity ................................... 3/i 6-1 Containment Leakage ..................................... 3/. 6-2 Containment Air Locks ................................... 3/4 6-4 Internal Pressure ....................................... 3/4-6-6 Air Temperature ......................................... 3/4 6-8 Containment Structural Integrity ........................ 3/4 6-9 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Quench Spray System ......................... 3/4 6-10 Containment Recirculation Spray System .................. 3/4 6-12 Chemical Addition System ................................ 3/4 6-13 l 3/4.6.3 CONTAINMENT ISOLATION VALVES ............................ 3/4 6-15 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers ...................................... 3/4 6-33 Electric Hydrogen Recombiners ........................... 3/4 6-34 Waste Gas Charcoal Filter System ........................ 3/4 6-35 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM Steam Jet Air Ejector ................................... 3/4 6-37 f

a i

f j

NORTH ANNA - UNIT 1 VII

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves ........................................... 3/4 7-1 Auxiliary Feedwater System .............................. 3/4 7-5 Emergency Condensate Storage Tank ....................... 3/4 7-7 Activity ................................................ 3/4 7-8 Main Steam Trip Valves .................................. 3/4 7-10 Steam Turbine Assembly .................................. 3/4 7-14 Overspeed Protection .................................... 3/4 7-15 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION ......... 3/4 7-16 3/4.7.3 COMP,JENT COOLING WATER SUBSYSTEM ....................... 3/4 7-17 3/4.7.4 SERVICE WATER SYSTEM .................................... 3/4 7-18 3/4.7.5 ULTIMATE HEAT SINK ...................................... 3/4 7-19 3/4.7.6 FLOOD PROTECTION ........................................ 3/4 7-20 3/4.7.7 CONTROL ROOM EMERGENCY HABITABILITY SYSTEMS ............. 3/4 7-21

{ 3/4.7.8 SAFEGUARDS AREA VENTILATION SYSTEM ...................... 3/4 7-24 3/4.7.9 RESIDUAL HEAT REMOVAL SYSTEM T y 1. 350*F ............................................ 3/4 7-26 T,yg <350*F............................................ 3/4 7-27 3/4.7.10 SNUBBERS ................................................ 3/4 7-28 3/4.7.11 SEALED SOURCE CONTAMINATION ............................. 3/4 7-68 3/4.7.12 SETTLEMENT OF CLASS 1 STRUCTURES ........................ 3/4 7-70 3/4.7.13 CROUNDWATER LEVEL-SERVICE WATER RESERVOIR ............... 3/4 7-73 i

6 I

NORTH ANNA - UNIT 1 VIII I

INDEX j LIMITING CONDITIONS FOR OPERATION AND SURVEILLA4CE REQUIREMENTS SECTI_ON PAGE 4

3/4.7.14 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System .......................... 3/4 7-75 Low Pressure CO Systems ...............................

2 3/4 7-79 High Pressure CO Systems .............................. 3/4 7-80 2

Halon Systems .......................................... 3/4 7-81 Fire Hose Stations ..................................... 3/4 7-82 3/4.7.15 PENETRATION FIRE BARRIERS .............................. 3/4 7-84 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A. C. SOURCES operating .............................................. 3/4 8-1 Shutdown ............................................... 3/4 8-5 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A. C. Distribution - Operating ......................... 3/4 8-6 A. C. Distribution - Shutdown .......................... 3/4 8-7 D. C. Distribution - Operating ......................... 3/4 8-8 D. C. Distribution - Shutdown .......................... 3/4 8-10 6

NORTH ANNA - UNIT 1 IX

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION .................................... 3/4 9-1 3/4.9.2 INSTRUMENTATION ........................................ 3/4 9-2 3/4.9.3 DECAY TIME ............................................. 3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS ...................... 3/4 9-4 3/4.9.5 COMMUNICATIONS ......................................... 3/4 9-5 3/4.9.6 MANIPULATOR CRANE OPERABILITY .......................... 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT ............... ....... 3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION All Water Levels ....................................... 3/4 9-8 Low Water Level ........................................ 3/4 9-8a 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM ......... 3/4 9-9 3/4.9.10 WATER LEVEL-REACTOR VESSEL ............................. 3/4 9-10 3/4.9.11 SPENT FUEL PIT WATER LEVEL ............................. 3/4 9-11 3/4.9.12 FUEL BUILDING VENTILATION SYSTEM ....................... 3/4 9-12 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN ......................................... 3/4 10-1 3/4.10.2 GROUP HEIGHT INSERTION AND POWER-DISTRIBUTION ........... 3/4 10-2 3/4.10.3 PHYSICS TEST ............................................ 3/4 10-3 3/4.10.4 REACTOR COOLANT LOOPS ................................... 3/4 10-4 3/4.10.5 POSITION INDICATOR CHANNELS-SHUTDOWN .................... 3/4 10-5 e

d

NORTH ANNA - UNIT 1 X

l l

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 1

3/4.11.1 LIQUID EFFLUENTS Concentration............................................. 3/4 11-1 Dose...................................................... 3/4 11-5 Liquid Radwaste Treatment................................. 3/4 11-6 Liquid Holdup Tanks....................................... 3/4 11-7 3/4.11.2 GASEOUS EFFLUENTS Dose Rate................................................. 3/4 11-8 Dose-Noble Gases.......................................... 3/4 11-13 Dose-Iodine-131, Tritium, and Radionuclides in Particulate Fo rm. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-14 Gaseous Radwaste Treatment................................ 3/4 11-15 Explosive Gas Mixture..................................... 3/4 11-16 Gas Storage Tanks......................................... 3/4 11-17 3/4.11.3 SOLID RADIOACTIVE WASTE................................... 3/4 11-18 3/4.11.4 TOTAL D0SE................................................ 3/4 11-19 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM...................................... 3/4 12-1 1 3/4.12.2 LAND USE CENSUS......................................... 3/4 12-13 1

3/4.12.3 INTERLABORATORY COMPARISON.............................. 3/4 12-14 NORTH ANNA - UNIT 1 XI

INDEX .

o BASES SECTION PAGE 3/4.0 APPLICABILITY................................................ B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L........................................... B 3/4 1-1 3/4.1.2 BO RATION SYST EMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 1-2 3/4.1.3 MOVAB LE CONTROL ASS EMB LI ES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE...................................... B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACT 0RS............................................ B 3/4 2-4 3/4.2.4 QUADRANT POWER TILT RATI0.................................. B 3/4 2-5 3/4.2.5 DNB PARAMETERS............................................. B 3/4 2-6 3/4.2.6 AXIAL POWER DISTRIBUTION................................... B 3/4 2-6 l-b i

i P

t .- .

NORTH ANNA - UNIT 1 XII

INDEX BASES SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION ............................... B 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION ................ B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION ............................... B 3/4 3-1 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS .................................... B 3/4 4-1 3/4.4.2 and 3/4.4.3 SAFETY VALVES................................. B 3/4 4-2 3/4.4.4 P RES S URI ZER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 4 -2 3/4.4.5 ST EAM G ENE RATORS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 4 -3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE ........................... B 3/4 4-4 3/4.4.7 CH EMI STRY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 4 -5 3/4.4.8 SPECIFIC ACTIVITY ........................................ B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS .............................. B 3/4 4-6 3/4.4.10 STRUCTURAL INTEGRITY ..................................... B 3/4 4-11 NORTH ANNA - UNIT 1 XIII

INDEX

)

BASES SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 AC CUMU LATO R S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS................................ B 3/4 5-1 3/4.5.4 BORON INJECTION SYSTEM..................................... B 3/4 5-2 3/4.5.5 REFUELING WATER STORAGE TANK (RWST) . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS .

3/4.6.1 PRIMARY CONTAINMENT.......................................... B 3/4 6-1 i 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS......................... B 3/4 6-3 3/4.6.3 CONTAINMENT ISO LATION VALVES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 6-3 3.4/6.4 COMBUSTIBLE GAS CONTR0L...................................... B 3/4 6-4 3/4.6.5 SUBSTMOSPNERIC PRESSURE CONTROL SYSTEM....................... B 3/4 6-4 NORTH ANNA - UNIT 1 XIV l

INDEX BASES SECTION 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ........................................... B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION ......... B 3/4 7-4 3/4.7.3 COMPONENT COOLING WATER SUBSYSTEM ....................... B 3/4 7-4 3/4.7.4 SERVICE WATER SYSTEM .................................... B 3/4 7-4 3/4.7.5 ULTIMATE HEAT SINK ...................................... B 3/4 7-5 3/4.7.6 FLOOD PROTECTION ........................................ B 3/4 7-5 3/4.7.7 CONTROL ROOM EMERGENCY HABITABILITY ..................... B 3/4 7-5 3/4.7.8 SAFEGUARDS AREA VENTILATION SYSTEM ...................... B 3/4 7-5 3/4.7.9 RESIDUAL HEAT REMOVAL SYSTEMS ........................... B 3/4 7-6 3/4.7.10 HYDRAULIC SNUBBERS ...................................... B 3/4 7-6 3/4.7.11 SEALED SOURCE CONTAMINATION ............................. B 3/4 7-7 3/4.7.12 SETTLEMENT OF CLASS 1 STRUCTURES ........................ B 3/4 7-7a 3/4.7.13 GROUNDWATER LEVEL - SERVICE WATER RESERVOIR ............. B 3/4 7-9 3/4.7.14 FIRE SUPPRESSION SYSTEMS ................................ B 3/4 7-9 3/4.7.15 PENETRATION FIRE BARRIERS ............................... B 3/4 7-10 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A. C. SOURCES ........................................... B 3/4 8-1 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS ....................... B 3/4 8-1 3/4.9 REFUELING OPERATIONS f

3/4.9.1 BORON CONCENTRATION ..................................... B 3/4 9-1 3/4.9.2 INSTRUMENTATION ......................................... B 3/4 9-1 A

f NORTH ANNA - UNIT 1 XV

a INDEX BASES SECTION PAGE 3/4.9.3 D ECAY TIM E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 9 - 1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS ....................... B 3/4 9-1 3/4.9.5 COMMUNICATIONS .......................................... B 3/4 9-1 3/4.9.6 MANIPULATOR CRANE OPERABILITY ........................... B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT ........................... B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION . . . . . . . . . . . B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST' ISOLATION SYSTEM .......... B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL-REACTOR VESSEL AND SPENT FUEL PIT .......................................... B 3/4 9-3 3/4.9.12 FUEL BUILDING VENTILATION SYSTEM ........................ B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN ........................................ B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION L IM ITS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 10 - 1 3/4.10.3 P HY S I C S TE STS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 10- 1 3/4.10.4 REACTOR COOLANT LOOPS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 10-1 3/4.10.5 POSITION INDICATOR CRANNELS - SHUTDOWN ................. B 3/4 10-1 b

NORTH ANNA - UNIT 1 XVI

+ "

3 INDEX BASES SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS........................................ B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS....................................... B 3/4 11-2 3/4.11.3 SOLID RADIOACTIVE WASTE................................. B 3/4 11-5 3/4.11.4 TOTAL D0SE.............................................. B 3/4 11-5 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM...................................... B 3/4 12-1 3/4.12.2 LAND USE CENSUS......................................... B 3/4 12-2 1

3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM...................... B 3/4 12-2 i

.I NORTH ANNA - UNIT 1 XVII

i INDEX_

, DESIGN FEATURES SECTION PAGE 5.1 SITE Exclusion Area .............................................. 5-1 Low Population Zone ......................................... 5-1 Map Defining UNRESTRICTED AREAS For Radioactive Gaseous and 4

Liquid Effluents ............................................ 5-2 5.2 CONTAINMENT Configuration ............................................... 5-1 Design Pressure and Temperature ............................. 54 5.3 REACTOR CORE Fuel Assemblies ............................................. 5-4 Control Rod Assemblies ...................................... 5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature ............................. 5-4 1

Volume ...................................................... 55 5.5 METEOROLOGICAL TOWER LOCATION ............................... 5-5 5.6 FUEL STORAGE Criticality ................................................. 5-5 Drainage .................................................... 56

Capacity .................................................... 5-6 k

5.7 COMPONENT CYCLE OR TRANSIENT LIMIT .......................... 5-6 b

I l

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J NORTH ANNA - UNIT 1 XVIII

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY ............................................... 6-1 6.2 ORGANIZATION Offsite ...................................................... 6-1 Facility Staff ............................................... 6-1 Safety Engineering Staff ..................................... 6-la Shift Technical Advisor ...................................... 6-la 6.3 FACILITY STAFF QUALIFICATIONS ................................ 6-5 6.4 TRAINING ..................................................... 6-5 6.5 REVIEW AND AUDIT 6.5.1 STATION NUCLEAR SAFETY AND OPERATING COMMITTEE (SNSOC)

Function ................................................... 6-5 Composition ................................................ 6-5 Alternates ................................................. 6-5 Meeting Frequency .......................................... 6-6 Quorum ..................................................... 6-6 Responsibilities ........................................... 6-6 Authority .................................................. 6-7 Records .................................................... 6-7 6.5.2 SAFETY EVALUATION AND CONTROL (SEC)

Function ................................................... 6-7 Composition ................................................ 6-8 h

1

)

O NORTH ANNA - UNIT 1 XIX

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE Consultants ................................................. 6-8 Meeting Frequency ........................................... 6-8 Review ...................................................... 6-8 Authority ................................................... 6-9 Records ..................................................... 6-9 6.5.3 QUALITY ASSURANCE DEPARTMENT Function .................................................... 6-10 Authority ................................................... 6-11 Records ..................................................... 6-11 6.6 REPORTABLE OCCURRENCE ACTION ................................ 6-12 6.7 SAFETY LIMIT VIOLATION ...................................... 6-12 6.8 PROCEDURFS AND PROGRAMS ..................................... 6-12 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS AND REPORTABLE OCCURRENCES ................ 6-14 6.9.2 SPECIAL REPORTS ........................................... 6-21 6.10 RECORD RETENTION ........................................... 6-22 6.11 RADIATION PROTECTION PROGRAM ............................... 6-23 6.12 HIGH RADIATION AREA ........................................ 6-23 6.13 ENVIRONMENTAL QUALIFICATION ................................ 6-24 6.14 PROCESS CONTROL PROGRAM (PCP) .............................. 6-25 6.15 0FFSITE DOSE CALCULATION MANUAL (ODCM) ..................... 6-25 b

NORTil ANNA - UNIT 1 XX

1.0 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.

ACTION i

1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.

AXIAL FLUX DIFFERENCE 1.2 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals, expressed in % of RATED THERMAL POWER between the top and bottom halves of a two section excore neutron detector.

CHANNEL CALIBRATION 1.3 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRA-TION shall encompass the entire channel including the sensor and 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.

CHANNEL CHECK 1.4 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indica-tions and/or status derived from independent instrumentation channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST d

1.5 A 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.

CONTAINMENT INTEGRITY 1.6 CONTAINMENT INTEGRITY shall exist when:

,1.6.1 All penetrations required to be closed during accident conditions are either:

NORTH ANNA - UNIT 1 1-1

1.0 DEFINITIONS (Continued)

a. Capable of being closed by an OPERABLE containment auto-matic isolation valve system, or

b. Closed by manual valves, blind flanges, or deactivated auto-matic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.1, 1.6.2 All equipment hatches are closed and sealed, 1.6.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, 1.6.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and 1.6.5 The sealing mechanism associated with each penetration (e.g.

welds, bellows or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.7 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION 1.8 CORE ALTERATION shall be the movement.or manipulation of any com-ponent within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative pbsitiou. ,

~

DOSE EQUIVALENT I-131

1.9 The DOSE EQUIVALENT I-131 shall be char concentration. oL I-131 . . .

(microcurie / gram) which alone would produce the same thyroid dose as .the ,

quantity and isotopic mixture of I-131, I-132, 1-133, I-134 and I-135 actually

~

present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distence Factors for Power and Test Reactor Sites." .

E - AVERAGE DISINTEGRATION ENERGY _ . ,

1.10 E shall be the average (weighted in proportion to the concentration

~

of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegrstion (in MeV) for isotopes, other than iodit.es, with half lives ' greater than 15 minu,tes, makirg up at least 95% of'the total non-iodine activity in the coolant. .

p " w

.O f

NORTH ANNA - UNIT 1 1-2 '.

N 1

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1

,, n- - , - , .-,-r----.- , . , , , , - --- ., - , ~ . n._ * ..-

1.0 DEFINITIONS (Continued)

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.11 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of

, performing its safety function (i.e., the valves travel to their re-quired positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.12 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

GASEOUS RADWASTE TREATMENT SYSTEM I

1.13 A CASE 0US RADWASTE TREATMENT SYSTEM is the system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity ~ prior to release to the environment. The system is composed of the waste gas decay tanks, regenerative heat exchanger, waste gas charcoal filters, process vent blowers, waste gas surge tenks and waste gas diaphram compressor.

IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such I as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or F b. Leakage into the containment atmosphere from sources that are f' both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor coolant system leakage through a steam generator to the secondary system.

HEMBER(S) 0F THE PUBLIC l

1.15 MEMBER (S) 0F THE PUBLIC shall include all individuals who by virtue of I

their occupational status have no formal association with the plant. This category shall include non-employees of the licensee who are permitted to use i

portions of the site for recreational, occupational, or other purposes not I

associated with plant functfans. This category shall not include non-employees

( + such as vending machine servicemen or postmen who, as part of their formal job function, ocassionally enter an area that is controlled by the licensee for

purposes of protection of individuals from exposure to radiation and radio-active materials.

i NORTH ANNA - UNIT 1 1-3

.~ <

+

l -

1.0 DEFINITIONS (Continued) 0FFSITE DOSE CALCULATION MANUAL (ODCM) 1.16 The OFFSITE DOSE CALCULATION MANUAL shall contain the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints and the specific monitoring locations of the environmental radiological monitoring program.

OPERABLE - OPERABILITY 1.17 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL MODE - MODE 1.18 An OPERATIONAL MODE (i.e. , MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.1.

PHYSICS TESTS 1.19 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.20 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

PROCESS CONTROL PROGRAM (PCP),

1.21 The PROCESS CONTROL PROGRAM shall contain the current formula, sampling, unalyses, tests and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 71 and Federal and State regulations and other requirements governing the disposal of the radioactive waste.

PURdE-PURGING 1.22 PURGE or PURGING is the controlled process of discharging air .or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

NORTH ANNA - UNIT 1 1-4

1.0 DEFINITIONS (Continued)

QUADRANT POWER TILT RATIO 1.23 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.24 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2775 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.25 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE OCCURRENCE 1.26 A REPORTABLE OCCURRENCE shall be any of those conditions specified in Specification 6.9.1.6 and 6.9.1.9.

SHUTDOWN MARGIN 1.27 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be suberitical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1.28 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased or otherwise controlled by the licensee.

SOLIDIFICATION 1.29 SOLIDIFICATION shall be the conversion of wet wastes into a solid form that meets shipping and burial ground requirements.

SOURCE CHECK 1.30 A SOURCE CHECK shall be the qualitative assessment of channel response

~

when the channel sensor is exposed to radiation. This applies to installed radiation monitoring systems.

NORTH ANNA - UNIT 1 1-5 i

1.0 DEFINITIONS (Continued)

STACCERED TEST BASIS 1.31 A STAGGERED TEST BASIS shall consist of:

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

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

THERMAL POWER 1.32 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

UNIDENTIFIED LEAKAGE 1.33 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UNRESTRICTED AREA 1.34 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY where access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM 1.35 A VENTILATION EXHAUST TREATMENT SYSTEM is the system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing fodines or

! particulates from the gaseous exhaust stream prior to the release to the

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

VENTING 1.36 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not

^ provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

NORTH ANNA - UNIT 1 1-6

TABLE 1.1 OPERATIONAL MODES REACTIVITY % RATED AVERAGE COOLANT MODE C0FDITION, K THERMAL POWER

TEMPERATURE ff

1. POWER OPERATION 2.0.99 > 5% 2.350*F

2. STARTUP 2;0.99 ji 5% 2; 350*F

3. HOT STANDBY < 0.99 0 2;350*F

4. HOT SHUTDOWN < 0.99 0 350*F > T

> 200*F ##E

5. COLD SHUTDOWN < 0.99 0 ;L200*F

6. REFUELING ** < 0.95 0 < 140*F

Excluding decay heat.

- Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

NORTH ANNA - UNIT 1 1-7

TABLE 1.2 FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

R At least once per 18 months.

S/U Prior to each reactor startup.

P Completed prior to each release.

N.A. Not applicable.

e NORTH ANNA - UNIT 1 1-8 s

I e

, __~ y , - - - - --

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY l

LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding Specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

,3.0.2 Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour ACTION shall be initiated to place the unit in a MODE in which the Specification does not apply by placing it, as applicable, in:

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual Specifications.

This Specification is not applicable in MODES 5 or 6.

3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual Specifications.

3.0.5 When a system, subsystem train, component or device is determined to be

' inoperable solely because its eme gency power source is inoperable or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding subsystem (s),

train (s), component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of this Specification. Unless both conditions (1) and (2) are satisfied, the unit shall be placed in at least HOT STANDBY within I hour, in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in at least COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

This Specification is not applicable in MODES 5 or 6.

NORTH ANNA - UNIT 1 3/4 0-1

INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm /

trip setpoints of these channels shall be determined and adjusted in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILIIT: At all times.

ACTION:

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

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, for reasons other than a above, take the ACTION shown in Table 3.3-12. Exert best efforts to return the instruments to OPERABLE status within 30 days and, if unsuccessful, explain in the next Seminannual Radioactive Effluent l

Release Report why the inoperability was not corrected in a timely manner.

c. The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1.9.b are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE

CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-12.

NORTH ANNA - UNIT 1 3/4 3-58 1

TABLE 3.3-12 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 5

5 MINIMUM CHANNELS h ,

INSTRUMENT OPERABLE ACTION 5

e g 1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND I y AUTOMATIC TERMINATION OF RELEASE

,?6

~

a. Liquid Radwaste Effluent Line (1)

2. GROSS BETA OR GAMMA RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE (1) 26

a. Service Water System Effluent Line (1) 29

b. Circulating Water System Effluent Line Y

g 3. CONTINUOUS COMPOSITE SAMPLERS AND SAMPLER FLOW MONITOR (1) 26

a. Clarifier Effluent Line

4. FLOW RATE MEASUREMENT DEVICES (1) 27

a. Liquid Radwaste Effluent Line I

l

.. .. . .. . __ _ _ _ _ _ _ _1

5 . TABLE 3.3-12 (Continued)

E

. RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION E

I MINIMUM CHANNELS g ACTION INSTRUMENT OPERABLE q

5. TANK LEVEL INDICATING DEVICES *

a. Refueling Water Storage Tank (1) 28

b. Casing Cooling Storage Tank (1) 28

c. PG Water Storage Tanks ** (1) 28

d. Boron Recovery Test Tanks ** (1) 28 Y

Tanks included in this Specification are those outdoor tanks that are not surrounded by liners, dikes.

I or walls capable of holding the tank contents and do not have tank overflows and surrounding area i drains connected to the liquid radwaste treatment system.

- This is a shared system with Unit 2.

1 l

t

TABLE 3.3-12 (Continued)

TABLE NOTATION ACTION 26 - With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and analyzed for gross radioactivity (petaorgamma)atalowerlimitofdetection of at least 10 microcurie / gram or an isotopic 7radioactivity at a lower limit of detection of at least 5x10 microcuries/ gram.

ACTION 27 - 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. Design capacity performance curves generated in situ may be used to estimate flow.

ACTION 28 - With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, liquid additions to this tank may continue provided the tank liquid level is estimated during all liquid additions to the tank.

ACTION 29 - With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, make repairs as soon as possible. Grab samples cannot be obtained via this pathway.

6 NORTH ANNA - UNIT 1 3/4 3-61

$ TABLE 4.3-12

^

RADI0AdTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

, CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL CHECK CHECK CALIBRATION TEST i h

H INSTRUMENT

- 1. CROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Liquid Radwaste Effluent Line D D R Q(1)

2. GROSS BETA OR CAMMMA RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE y a. Service Water System Effluent Line D M R Q(2)

C

b. Circulating Water System Effluent Line D M R Q(2)

3. CONTINUOUS COMPOSITE SAMPLERS AND SAMPLER FLOW MONITOR

a. Clarifier Effluent Line N.A. N.A. R N.A.

TABLE 4.3-12 (Continued) a h

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

I CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL g

y INSTRUMENT CHECK CHECK CALIBRATION TEST

4. FLOW RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line D(3) N.A. R Q

5. TANK LEVEL INDICATING DEVICES ***

a. Refueling Water Storage Tank D* N.A. R Q y b. Casing Cooling Storage Tank D* N.A. R Q

c. PG Water Storage Tanks ** D* N.A. R Q

d. Boron Recovery Storage Tanks ** D* N.A. R Q

During liquid additions to the tank.

- This is a shared system with Unit 2.

- - Tanks included in this Specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.

i

(

l t

TABLE 4.3-12 (Continued)

TABLE NOTATION (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:

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

2. Instrument controls not set in operate mode.

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

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

2. Instrument controls not set in operate mode.

(3) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL 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 on which continuous, periodic, or batch releases are made.

NORTH ANNA - UNIT 1 3/4 3-64

INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.11 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The j

alarm / trip setpoints of these channels shall be determined and adjusted in accordance with the ODCM.

APPLICABILITY: As shown in Table 3.3-13 4 ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the l above Specification, without delay suspend the release of radioactive

] gaseous effluents monitored by the affected channel or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, for reasons other than a above, take the ACTION shown in Table 3.3-13. Exert best efforts

to return the instruments to OPERABLE status within 30 days and, if l unsuccessful, explain in the next Semiannual Radioactive Effluent

Release Report why the inoperability was not corrected in a timely manner.

c. The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1.9.b are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.11 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE

CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-13.

1 NORTH ANNA - UNIT 1 3/43 65

TABLE 3.3-13 g

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION z

MINIMUM CHANNELS

' OPERABLE APPLICABILITY ACTION INSTRUMENT E

U l. PROCESS VENT SYSTEM w

a. Noble Gas Activity Monitor -

Providing Alarm and Automatic

31 Termination of Release (1)

31

b. Ic~.ine Sampler (1)

31

c. Particulate Sampler (1) w 30 d. Process Vent Flow Rate Measuring Device (1)

30 92

e. Sampler Flow Rate Measuring Device (1)

30

2. WASTE GAS EdLDUP SYSTEM EXPLOSIVE GAS MONITORING SYSTEM (Shared with Unit 2)

(1) ** 32

a. Hydrogen Monitor

- 32

b. Oxygen Monitor (1)

I l

f$ -

TABLE 3.3-13 (Continued)

RADI0 ACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION I MINIMUM CHANNELS g INSTRUMENT OPERABLE APPLICABILITY ACTION n

3. CONDENSER AIR EJECTOR SYSTEM w

31

a. Cross Activity Monitor (1)

b. Flow Rate Monitor (1)

30

4. VENTILATION VENT SYSTEM (Shared with Unit 2 )

a. Noble Gas Activity Monitor (1)*

31 T'

$ b. Iodine Sampler (1)*

31

c. Particulate Sampler (1)*

31

d. Flow Rate Monitor (1)*

30

e. Sampler Flow Rate Monitor (1)*

30 i

0ne per vent stack.

TABLE 3.3-13 (Continued)

TABLE NOTATION

At all times.

- During process vent system operation (treatment for primary system

, offgases).

ACTION 30 - 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 31 - 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 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for gross activity and gamma isotopic activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 32 - With the number of channnels OPERABLE one less than required by the minimum channels OPERABLE requirement, operation of this system may continue for up to 14 days provided grab samples are taken and analyzed daily. With both channels inoperable, operation may continue provided grab samples are taken and analyzed: (1) every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during degassing operations and (2) daily during other operations.

s b

a O

D e

NORTH ANNA - UNIT 1 3/4 3 68 1

5 TABLE 4.3-13 s .

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE

, g CHECK CALIBRATION TEST REQUIRED INSTRUMENT CHECK q

~ 1. PROCESS VENT SYSTEM

a. Noble Gas Activity Monitor -

Providing Alarm and Automatic Termination of Release D P R Q(1)

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

N.A. N.A.

w c. Particulate Sampler W N.A. '

5

w d. Process Vent Flow Rate D N.A. R Q en Measuring Device N.A. *

e. Sampler Flow Rate Monitor D(5) N.A. R

2. WASTE GAS HOLDUP SYSTEM EXPLOSIVE GAS MONITORING SYSTEM N.A. M **

a. Hydrogen Monitor D Q(3)

N.A. M **

b. Oxygen Monitor D Q(4) l i

i

2 TABLE 4.3-13 (Continued)

O .

y RADIOACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

> CHANNEL MODES IN WHICH

' CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE

@ INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED Q

~ 3. CONDENSER AIR EJECTOR SYSTEM M *

a. Noble Gas Activity Monitor D R Q(2)

b. Flow Rate Monitor D N.A. R Q *

4. VENTILATION VENT SYSTEM (Shared with Unit 2 )

a. Noble Gas Activity Monitor D M R Q(2) y b. Iodine Sampler W N.A. N.A. N.A. *

c. Particulate Sampler W N.A. N.A. N.A. *

d. Flow Rate Monitor D N.A. R Q *

e. Sampler Flow Rate Monitor D(5) N.A. R N.A.

l l

I

TABLE 4.3-13 (Continued)

TABLE NOTATION

At all times other than when the line is valved out and/or locked.

- During process vent system operation (treatment for primary system offgases).

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

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

2. Instrument controls not set in operate mode.

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

1. Instrument indicates measured levels above the alarm setpoint.

2. Instrument controls not set in operate mode.

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

1. One volume percent hydrogen, balance nitrogen, and

2. Four volume percent hydrogen, balance nitrogen.

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

1. One volume percent oxygen, balance nitrogen, and

2. Four volume percent oxygen, balance nitrogen.

(5) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL 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 on which continuous, periodic or batch releases are made.

NORTH ANNA - UNIT 1 3/4 3-71

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

I t 1 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS CONCENTRATION I

LIMITING CONDITION FOR OPERATION 3.11.1.1 The concentration of radioactive material released in liquid

effluents to UNRESTRICTED AREAS (see Figure 5.1-1) shall be limited to the con-centrations specified in 10 CFR Part 20, Appendix B. Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For digsolved or entrained noble gases, the concentration shall be limited to 2 x 10 microcuries/ml.

APPLICABILITY: At all times.

ACTION:

With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, without delay restore the concentration to within the above limits.

SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.11-1.

4.11.1.1.2 The results of the radioactivity analyses shall be used in accordance with the methods in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1.

i 2

T e

NORTH ANNA - UNIT 1 3/4 11-1 i

i

-, . _ _ , , . - _ , _. _-. _- m., . - .- _

TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum of Detection Liquid Release Sampling Analysis Type of Activity (LLD)a Type Frequency Frequency Analysis (pCi/ml)

A. Batch P P Each Batch Each Batch Principal Gamma 5x10

-7 Releages '8 Tanks Emitters"

-6 I-131 1x10

-5 P M Dissolved and 1x10 One Batch /M Entrained Gases (Gamma Emitters)8 P M -

1x10 d

Each Batch Composite

~7 Gross Alpha 1x10

-8 P Q Sr-89, Sr-90 5x10 d

Each Batch Composite

-6 Fe-55 1x10

~

B. Continuogs W Principa} Gamma 5x10 f f Releases Continuous Composite Emitters

-6 I-131 1x10 Dissolved and 1x10

-5 Entrained Cases (Gamma Emitters)8

-5 M H-3 1x10 f f Continuous Composite

_7 Gross Alpha 1x10 h

-8 Q Sr-89, Sr-90 5x10 f f Continuous Composite -6 Fe-55 1x10 NORTH ANNA - UNIT 1 3/4 11-2 1

i

TABLE 4.11-1 (Continued) l TABLE NOTATION The (LLD) is defined, for purposes of these Specifications, as the smallest concentration of radioactive material in a sample that will yield a net count (above system background) that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

4.66 s b LLD =

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),

. shis the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),

I

~

E is the counting efficiency (as counts per disintegration).

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

6 2.22 x 10 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),

I A is the radioactive decay constant for the particular radionuclide, and, f at for plant' effluents is the elapsed time between the midpoint of t sample collection and time of counting.

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

l .

NORTH ANNA - UNIT 1 3/4 11-3 i

TABLE 4.11-1 (Continued)

TABLE NOTATION It should be recognized that the LLD is defined as an JL priori (before the fact) limit representing the capability of a measurement system and not as an JL posteriori (after the fact) limit for a particular measurement.

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

Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mix as the situation permits, to assure representative sampling.

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

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

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

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

g For certain mixtures of gamma emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the same sample in much greater concentrations.

Under these circumstances, it will be more appropriate to calculate the concentrations of such radionuclides using measured ratios with those radionuclides which are routinely identified and measured.

NORTH ANNA - UNIT 1 3/4 11-4

4 RADIOACTIVE EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION l

3.11.1.2 The dose or dose commitment to the maximum exposed MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each reactor unit, to UNRESTRICTED AREAS (see Figure 5.1-1) shall be limited:

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

and

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

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of

, a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.2 Dose Calculations. Cumulative dose contributions from liquid effluents shall be determined in accordance with the ODCM at least once per 31 days.

b i

! *The maximum exposed MEMBER OF THE PUBLIC and the critical organ will be l addressed in the ODCM.

L 9

NORTH ANNA - UNIT 1 3/4 11-5

RADIOACTIVE EFFLUENTS LIQUID RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.1.3 The liquid radwaste ion exchanger system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent, from each reactor unit, to UNRESTRICTED AREAS (see Figure 5.1-1) would exceed 0.06 mrem to the total body or 0.2 mrem to the critical organ

in a 31 day period.

APPLICABILITY: At all times.

ACTION:

a. With radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report that includes the following information:

1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or sub-systems, and the reason for the inoperability,

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

3. Summary description of ACTION (s) taken to prevent a recurrence.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liqu'id releases shall be projected at least once per 31 days in accordance with the ODCM.

The critical organ will be addressed in the ODCM.

NORTH A!a. UNIT 1 3/4 11-6

I RADIOACTIVE EFFLUENTS ,

LIQUID HOLDUP TANKS .

T ;

1

'N. i LIMITING CONDITION FOR OPERATION ,

3.11.1.4 The quantity of radioactive material contained in' eahh of the following unprotected outdoor tanks shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained no' ole gases.

a. Refueling Water Storage Tank

b. Casing Cooling Storage Tank .

c. PG Water Storage Tank * + ,

d. Boron Recovery Test Tank * ,i

e. Any Outside Temporary Tank ** -

N-'

APPLICABILITY: At all times. - ~. .

s ACTION: ,

a. Withthequantityofradioatk0ematerial'inhnyoftheabovelisted tanks exceeding the above limitr immediately 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 tank contents to within the limit. >

b. The provisions of Specifications 3.0.3 and 3.0.4 are not tpplicable.

/

m SURVEILLANCE REQUIREMENTS I

4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by anslyzing a _ '

representative sample of the tank's coritents at least once per month when radioactive materials are being added to the tank.

This is a shared system with Unit 2.

l

- Tanks included in this Specification are those outdoor tanks that are not l surrounded by liners, dikes, or walls capable of holding the tank contents i and that do not have tank overficws and surrounding area drains connected to l the liquid radwaste ion exchanger systen. -

l NORTH ANNA - UNIT 1 3/4 11-7 W

i 4

9 RADkOACTIVEEFFLUENTS 3/4.11.2 CASEOUS EFFLUENTS DOSE RATE 4

, LIMITING CONDITION FOR OPERATION 4

e 3.11.2.1 The dose rate due to radioactive materials released in gaseous i effluents from the site to areas at and beyond the SITE BOUNDARY (see J Figure 5.1-1) shall be limited to the following:

a. For noble gases: Less than or equal to 500 mrems/yr to the total body and less than or equal to 3000 mtems/yr to the skin, and

-s ~

b.

For lodine-131, for tritium, and for all radionuclides in particulate form with half lives greater than 8 days: Less than or equal to 1500 mrems/yr to the critical organ.*

APPLICABILITY: At all times.

ACTION:

4 With the dose rate (s) exceeding the above limits, without delay restore the release rate to within the above limit (s).

SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined continuously to be within the above limits in accordance with the methods and procedures of the ODCM.

4.11.2.1.2 The dose rate due to iodine-131, tritium, and all radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents

shall be determined to be within the above limits in accordance with the

' methods and procedures of the ODCM by obtaining representative samples and perforcing analyses in accordance with the sampling and analysis program

,specified in Table 4.11-2.

F -

4

The critical organ is defined in the ODCM.

NORTH ANNA - UNIT 1 3/4 11-8

TABLE 4.11-2 g

W .

$ RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM E

Minimum Lower Limit of

' Sampling Analysis Type of Detection (LLD)*

E Gaseous Release Type Frequency Frequency Activity Analysis (uCi/ml)

U P P -4 A. Waste Gas Storage Each Tank Each Tank Principal Gamma Emitters 1x10 Tank" Grab Sample P P b -4 B. Containment PURGE Each PURGE Each PURGE Principal Gamma Emitters 1x10 Grab Sample H-3 1x10-C. Process Vent M M" Principal Gamma Emitters 1x10-

.-. Vent. Vent A Grab Y Vent. Vent B Sample e

-6 H-3 lx10

-12 D. All Release Types Continuous W" I-131 1x10 as listed in A, B, Charcoal C above. Sample Continuous W" Principal Gamma Emitters lx10 '

Particulate (I-131, Others)

Sample

-1 Continuous M Gross Alpha 1x10 Composite Particulate Sample

-ll Continuous Q Sr-89, Sr-90 lx10 Composite Particulate Sample Continuous Noble Gas Noble Cases 1x10-Monitor Gross Beta or Gamma

TABLE 4.11-2 (Continued) z -

@ RADIOACTIVE CASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Y -

E

' Minimum Lower Limit of

@ Sampling Analysis Type of Detection (LLD)*

y Gaseous Release Type Frequency Frequency Activity Analysis (uCi/ml)

E. Condynser Air Ejector W W Principle Gamma Emitters 1x10~

Vent Grab Sample -6 H-3 1x10 Steam Generatgr Blowdown Vent w

h d

b f

3

TABLE 4.11-2 (Continued)

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

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

4.66 s b LLD =

E

V

2.22 x 106

Y

exp (-Aet)

Where:

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

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

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

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

2.22 x 100 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, b

NORTH ANNA - UNIT 1 3/4 11-11

TABLE 4.11-2 (Continued)

TABLE NOTATION 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 (Edter the fact) limit for a particular measurement.

The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, Xe-135m, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-134 Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks that are measurable and identifiable, at levels exceeding the LLD together with the above nuclides, shall also be identified and reported, i

" Sampling and analysis shall also be performed following shutdown, j startup, and whenever a THERMAL POWER change exceeding 15 percent of the i RATED THERMAL POWER occurs within a one hour period, if (1) analysis i shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant is greater than 1.0 uCi/ga; and (2) the noble gas activity monitor shows that effluent activity has increased by more than a factor of 3.

d The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2 and 3.11.2.3.

Samples shall be changed at least once per 7 days and analyses shall be $

i completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from sampler).

Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER in one hour and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement applies if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant is greater than 1.0 uCi/gm and; (2) the noble gas monitor shows that effluent activity has increased more than a factor of 3.

f For certain mixtures of gamma emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the same sample in much greater concentrations.

Under those circumstances, it will be more appropriate to calculate the l concentrations of such radionuclides using measured ratios with those radionuclides which are routinely identified and measured.

[

I -

l l

NORTH ANNA - UNIT 1 3/4 11-12 l

1 _ - -_ . -. ._

RADIOACTIVE EFFLUENTS DOSE - NOBLE CASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each reactor unit, from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) shall be limited to the following:

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

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

APPLICABILITY: At all times.

ACTION

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.2 Dose Calculations Cumulative dose contributions for noble gases for the current calendar quarter and current calendar year shall be determined in accordance with tha ODCM at least once per 31 days.

s F-l NORTH ANNA - UNIT 1 3/4 11-13

RADIOACTIVE EFFLUENTS DOSE - IODINE-131. TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM LIMITING CONDITION FOR OPERATION l

l 3.11.2.3 The dose to the maximum exposed MEMBER OF THE PUBLIC from I fodine-131, from tritium, and from all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each react.r unit, from the site to UNRESTRICTED AREAS (see Figure 5.1-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrems to the critical organ

and,

b. During any calendar year: Less than or equal to 15 mrems to the critical organ *.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of iodine-131, tritium, and radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.3 Dose Calculations Cumulative dose contributions for the current calendar quarter and current calendar year for iodine-131, tritium and l

radionuclides in particulate form with half lives greater than 8 days shall be determined in accordance with the ODCM at least once per 31 days.

I

The critical organ is addressed in the ODCM.

1 NORTH ANNA - UNIT 1 3/4 11-14 l

1 I

I RADIOACTIVE EFFLUENTS CASEOUS RADWASTE TREATMENT LIMITING CONDITION FOR OPEPATION 3.11.2.4 The CASE 0US RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases, from each reactor unit, from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation over 31 days. The VENTILATION EXHAUST TREATMENT SYSTEH shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases, from each reactor unit, from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) would exceed 0.3 mrem to the critical organ

over 31 days.

APPLICABILITY: At all times.

ACTION:

a. With gaseous waste being discharged without treatment and in excess of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specifica-tion 6.9.2, a Special Report that includes the following information:

1. Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or sub-systems, and the reason for the inoperability,

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

3. Summary description of ACTION (s) taken to prevent a recurrence.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from the site shall be projected at least once per 31 days in accordance with the ODCM.

l l

l l *The critical organ is addressed in the ODCM.

1 NORTH ANNA - UNIT 1 3/4 11-15

RADIOACTIVE EFFLUENTS l

I EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the waste gas decay tanks shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume or is less than 96% by volume.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of oxygen in the waste gas decay tanks greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ,

b. With the concentration of oxygen in the waste gas decay tanks greater than 4% volume and the hydrogen concentretion greater than 2% by volume, immediately suspend all additions of waste gases to the system and reduce the concentration of oxygen to less than or equal to 2% by volume without delay.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentrations of hydrogen and oxygen in the waste gas decay tanks shall be determined to be within the above limits by continuously monitoring the waste gases in the waste gas decay tanks with the hydrogen and oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.11.

6 NORTH ANNA - UNIT 1 3/4 11-16

RADI0 ACTIVE EFFLUENTS GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained _a each gas storage tank shall be limited to less than or equal to < 25,000 curies noble gases l (considered as Xe-133).

APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and 9tchin 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

4 SURVEILLANCE REQUIREMENTS 4

4.11.2.6 The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank.

i NORTH ANNA - UNIT 1 3/4 11-17

RADIOACTIVE EFFLUENTS 3/4.11.3 SOLID RADIOACTIVE WASTE LIMITING CONDITION FOR OPERATION 3.11.3 SOLIDIFICATION shall be conducted in accordance with a PROCESS CONTROL PROGRAM.

APPLICABILITY: At all times.

ACTION:

a. With the provisions of the PROCESS CONTROL PROGRAM not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.

b. The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1.9.b are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.3.1 The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive waste (e.g., filter sludges, ,

spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions).

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

b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the

, collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least 3 consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.13, to assure SOLIDIFICATION of subsequent batches of waste.

I NORTH ANNA - UNIT 1 3/4 11-18

RADIOACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION l

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

1 APPLICABILITY: At all times.

ACTION:

a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifica-tion 3.11.1.2.a. 3.11.1.2.b, 3.11.2.2.a. 3.ll.2.2.b, 3.11.2.3.a. or 3.11.2.3.b, calculations should be made to determine whether the above limits of Specification 3.11.4 have been exceeded. If such is the case in lieu of a Licensee Event Report, prepare and submit to the Commission within 3G days, pursuant to Specification 6.9.2, a Special Report that defines the corrective action to be taken to reduce sub-sequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR Part 20.405c, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report.

It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose (s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.4 Dose Calculations Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the ODCM.

The critical organ is addressed in the ODCM.

NORTH ANNA - UNIT 1 3/4 11-19

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 4.12-1.

APPLICABILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 4.12-1, in lieu of a Licensee Event Report, prepare and submit to the Commission, in the Annual Radio-logical Environmental Operating Report required by Specification 6.9.1.11, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence,

b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 4.12-2 when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare and "abmit to the Commission within 30 days, pursuant to Specification 6.1.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose to A MEMBER OF THE PUBLIC is less than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2, and 3.11.2.3. When more than one of the radionuclides in Table 4.12-2 are detected in the sampling medium, this report shall be submitted if:

I concentration (1) concentration (2) * ***

! reporting level (1)

+

reporting level (2) 1.0 When radionuclides other than those in Table 4.12-2 are detected and l are the result of plant effluents, this report shall be submitted if I

the potential annual dose to a MEMBER OF THE PUBLIC is equal to or l

greater than the calendar year limits of Specifications 3.11.1.2, i

3.11.2.2 and 3.11.2.3. This report is not required if the measured l 1evel of radioactivity was not the result of plant effluents; however, i in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

c. With milk or fresh leafy vegetable samples unavailable from one or l . more of the sample locations required by Table 4.12-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be l NORTH ANNA - UNIT 1 3/4 12-1

1 l

l

, l RADIOLOGICAL ENVIRONMENTAL MONITORING deleted from the monitoring program. In lieu of a Licensee Event Report and pursuant to Specification 6.9.1.12, identify the cause of the unavailability of samples and identify the new location (s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s).

d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS s 4.12.1.1 The radiological environmental monitoring samples shall be collected pursuant to Table 4.12-1 from the specific locations given in the table and figure (s) in the ODCM and shall be analyzed pursuant to the requirements of Table 4.12-1, the detection capabilities required by Table 4.12-3, and the guidance of the Radiological Assessment Branch Technical Position on Environmental Monitoring dated November, 1979, Revision No. 1.

NORTH ANNA - UNIT 1 3/4 12-2

(

TABLE 4.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM

Y -

j Number of Samples

> Exposure Pathway and Sampling and Type and Frequency a

8 and/or Sample Sample Locations Collection Frequency of Analysis h 1. DIRECT RADIATION 36 routine monitoring Quarterly Gamma dose H stations either with two or quarterly.

L more dosimeters or with one instrument for measuring and recording dose rate continuously to be placed as follows: 1) an inner ring of stations, one in each meteorological sector within the SITE BOUNDARY; an outer ring of ti stations, one in each meteorological o sector within 8 km range from

[; the site; the balance of the 3 stations to be placed in special interest areas such as population centers, nearby residences, schools, and in 1 or 2 areas to serve as control stations.

2. AIRBORNE Radiciodine and Samples from 5 locations: Continuous sampler Radiciodine Cannister:

Particulates operation with sample I-131 analysis weekly,

a. 3 samples from close collection weekly or to the 3 SITE BOUNDARY more frequently if locations (in different required by dust Particulate Sampler:

sectors) of the highest loading." Gross beta radioactivity calculated historical analysis folloging annual average ground- filter change; level D/Q. Gamma isotopic analysis of composite (by

The number, media, frequency, and location of samples may vary from site to site. This table presents an acceptable minimum program for a site at which each entry is applicable. Local site characteristics must be examined to determine if pathways not covered by this table may significantly contribute to an individual's dose and should be included in the sampling program.

i 2 _ TABLE 4.12-1 (Continued)

O y .

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM

> Number of Samples 8 Exposure Pathway and Sampling and Type and Frequency a of Analysis g and/or Sample Sample Locations Collection Frequency E

-. b. I sample from the vicinity location) quarterly.'

i of a community having the highest calculated annual average groundlevel D/Q.

c. I sample from a control location 15-30 km w distant and in the 37 least prevalent wind g direction.

S.

3. WATERBORNE

a. Surface a. I sample circulating Sample off upstream, Camma isotopic analysis' water discharge downstream and monthly. Composite for cooling lagoon. tritium analysis quarterly.

Grab Monthly.

i

b. Ground Samples from 1 or 2 sources Grab Gamma isotopic

and tritium only if likely to be affected. Quarterly analysis quarterly.

c.. Sediment I sample from downstream area Semiannually Gamma isotopic analysis

with existing or potential semiannually.

recreational value.

1 l

TABLE 4.12-1 (Continued)

O y ,

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM f, Exposure Pathway Number of Samples and Sampling and Type and Frequency and/or Sample Sample Locations, Collection Frequency of Analysis c:

3

-i

~

. 4. INGESTION

a. Milk a. Samples from milking animals Monthly at all Gamma isotopic

and I-131 in 3 locations within 5 km times, analysis monthly .

distance having the highest dose potential. If there u,

are none, then, I sample from milking animals in each 3:

,. of 3 areas between 5 to 8 km y distant where doses are cal-u culated to be greater than 1 mrem per yr 8.

b. I sample from milking animals at a control location (15-30 km i distant and in the least pre-valent wind direction).

b. Fish and a. I sample of commercially Annually. Gamma isotopic analysis Invertebrates and recreationally on edible portions.

important species (bass, sunfish, catfish) in vicinity of plant discharge area.

b. I sample of same species in areas not influenced by plant discharge.

i

2 . TABLE 4.12-1 (Continued)

O y . RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM h Number of Samples

, Exposure Pathway and Sampling and Type and Frequency a

and/or Sample Sample Locations Collection Frequency of Analysis 2

H

- c. Food a. Samples of an ediable Monthly if available, Ganssa isotopice and I-131 Products broad leaf vegetation or at harvest. analysis.

(cont'd) grown nearest each of two different offsite locations of highest predicted historical annual average ground-w level D/Q if milk samp-2 ling is not performed.

U Ganuna isotopic e and I-131 4 b. I sample of broad leaf Monthly if available, vegetation grown 15-30 or at harvest, analysis.

km distant in the least prevalent wind direction if milk sampling is not performed.

z TABLE 4.12-1 (Continued)

O .

$ TABLE NOTATION l a Specific parameters of distance and direction sector from the centerline of one reactor, and additional description where pertinent, shall be provided for each and every sample location in Table 4.12-1 in a g table and figure (s) in the ODCM. Refer to NUREG-0133 " Preparation of Radiological Effluent Technical

- Specifications for Nuclear Power Plants, " October 1978, and to Radiological Assessment Branch Technical

Positions, Revision 1. November 1979. Deviations are permitted from the required sampling schedule if

~

specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, 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 documented in the Annual Radiologi-cal Environmental Operating Report pursuant to Specification 6.9.11.1. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the radiological environmental monitoring program. In lieu of a Licensee Event Report and pursuant to Specification 6.9.1.12, W identify the cause of the unavailability of samples for that pathway and identify the new location (s) for obtaining replacement samples in the next Semisnnual Radioactive Effluent Release Report and also include U in the report a revised figure (s) and table for the ODCM reflecting the new location (s).

O One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter-(TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. The 40 stations is not an absolute number. The number of direct radiation monitoring stations may be reduced according to geographical limitations; e.g., at an ocean site, some sectors will be over water so that the number of dosimeters may be reduced accordingly. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.

2 O TABLE 4.12-1 (Continued) c-a

= .

TABLE NOTATION g

E Iodine collection efficiency of the charcoal filters shall be determined, n

~

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

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

I A composite sample is one in which the quantity (aliquot) of liquid sampled is proportional to the quantity of flowing liquid and in which the method of sampling employed results in a specimen that is M represnetative of the liquid flow. In this program composite sample aliquots shall be collected at

" time intervals that are very short (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample,

{

cm EThe dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM.

h If milk sampling cannot be performed, use item 4.c.

jI lillII l1 f1l j\l llll llllllll s) tt ce d

uw 0 0 0 o , 0 0 0 Pk rg 1 0, 0,

/ 1 2 di oC op F(

S E

L P

M A

S )

kt 3 0 0 0 L l/

6 7 0 A ii 3

T MCp N

E (

M N

O R

I V

N E

N )

t I

e S w 0 N 0 0 0 0 0 0 O hsg , 0 0 0 0 0 0 0 I

s 0, 0, 0, 0, 0, 0, 0, T ik 0 2 A l F/ 0 0 0 0 1 2 R e i 3 1 3 1 2

- T v C 2 N e (

p 1 E L

. C 4 N g O n E C i L t B Y r A T o T I p V e I R e T t) g C

A l m O u/

I ci D iC A t p R r( 9 a 0 0 0 R Ps O e 1 2 F es na S rG L o E br V ro E i L A G

N I

T R

O P )

E 1 0 0 0 0 0 0 0 2 0 0 0 R r/ 0 0 0 0 0 0 0 3 5 0 4 2 ei tC 0, 0, 4 0, 3 3 ap 0 1 1 W( 3 0

s 5 4 i 9 1 s - 4 7 -

y 4 9 8 0 5 b 1 3 3 a l 5 5 5 6 6 N 3 1 1 L a 3 - - - - - - 1 -

s a

n - n e o o n r - s A H M F C C Z Z I C C B

5Qm g> ' @ ~ R* [E i ! ,ll' ll ll 1 \

$ TABLE 4.12-3

-, . a

DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS

> LOWER LIMIT OF DETECTION (LLD)b i

Fish Milk Food Products Sediment E Water (pCi/kg. dry)

Airborne or Gas Partieg)

(pC1/m late (pCi/kg, wet) (pCi/t) (pCi/kg, wet)

[i Analysis (pCi/t) gross beta 4 0.01 H-3 3000 Mn-54 15 130 Fe-59 30 260 Co-58,60 15 130 w

3 260

- Zn-65 30 Y

El Zr-Nb-95 15

60 I-131 10 0.07 1 0.05 130 15 60 150 Cs-134 15 0.06 150 18 80 180 Cs-137 18 15 Ba-La-140 15

The LLD for gamma isotopic analysis shall be used.

TABLE 4.12-3 (Continued)

TABLE NOTATION a

This list does not mean that only these nuclides are to be considered.

Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.11.

b The LLD is defined, for purposes of these Specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95%

probability with only 5% probability of falsely concluding that a blank observation represents a "rcal" signal.

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

LLD =

E -

V . 2.22 -

Y . exp( Aat)

Where:

LLD is the "a priori" lower limit of detection as defined above, as picoeuries per unit mass or volume, sg is the standard deviation of the background counting rate or of

~

the counting rate of a blank sample as appropriate, as counts per minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 is the 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 for environmental samples is the elapsed time between sample collection, or end of the sample collection period, and time of counting Typical valves of E, V, Y and at should be used in the calculation.

NORTH ANNA - UNIT 1 3/4 12-11

TABLE 4.12-2 (Continued)

TABLE NOTATION It should be recognized that the LLD is defined as an JL priori (before the fact) limit representing the capability of a measurement system and not as an jt posteriori (after the fact) limit for a particular I measurement. Analysis shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally i

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 pursuant to Specification 6.9.1.11.

b NORTH ANNA - UNIT 1 3/4 12-12

RADIOLOGICAL ENVIRONMENTAL MONITORING j 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 3.12.2 A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectorsofthenearestmilkgnimal,tgenearestresidenceandthenearest garden

of greater than 50 m (500 ft ) producing broad leaf vegetation.

APPLICABILITY: At all times.

ACTION:

a. With a land use census identifying a location (s) that yields a calcu-lated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, in lieu of a Licensee Event Report, identify the new location (s) in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.12.

b. With a land use census identifying a location (s) that yields a calculated dose or dose commitment (via the same exposure pathway) 25 percent greater than at a location from which samples are cur-rently being obtained in accordance with Specification 3.12.1, add the new location (s) to the radiological environmental monitoring program within 30 days. 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 of a Licensee Event Report

, and pursuant to. Specification 6.9.1.12, identify the new location (s) in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s).

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.2 The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting l

local agriculture authorities. The results of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.11.

Broad leaf vegetation sampling of at least three different kinds of vegetation l may be performed at the site boundary in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census. Specifications for broad leaf vegetation sampling in Table 4.12-1.4c shall be followed, including analysis of control samples.

l l

NORTH ANNA - UNIT 1 3/4 12-13

RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on radioactive materials (which contain nuclides produced at nuclear power stations) supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission.

APPLICABILITY: At all times.

ACTION:

a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.11.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.3 A summary of the results obtained as part of the above required Interlaboratory Comparison Program and in accordance with the methodology and parameters in the ODCM shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.11.

i i

P-NORTH ANNA - UNIT 1 3/4 12-14 4

INSTRUMENTATION BASES 3/4.3.3.9 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 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 and adjusted in accordance with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50. The purpose of tank level indicating devices is to assure the detection and control of leaks that if not controlled could potentially result in the transport of radioactive materials to UNRESTRICTED AREAS.

3/4.3.3.10 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 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 shall be calculated and adjusted in accordance with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring (and centrolling) the concentrations of potentially explosive gas mixtures in the waste gas holdup system. The OPERABILITY and use of this instrumentation is consistent with che requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

[

l i

1 NORTH ANNA - UNIT 1 B 3/4 3-4

3/4.11 RADIOACTIVE EFFLUENTS y V BASES 1

l 3/4.11.1 LIQUID EFFLUENTS ~

~m !

3/4.11.1.1 CONCENTRATION ~

This specification is provided to ensure that. the concentration of, radioactive materials released in liquid waste effluents 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 in UNRESTRICTED ' '

AREAS will result in exposures within (1) the Section II.A design objectives of Appendix I,10 CFR Part 50 to a MEMBER OF THE PUBLIC and (2) the limits of 10 CFR Part 20.106(e) to the population. The_ concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

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

Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Curile , L. ' A. , " Limits for Qualitative Detection and ~ Quantitative. Detdrmination . Application to Radiochemistry" Anal. Chem. 40, 586-93 (1968), And Eartwell,'J1 Y... " Detection Limits for Radioanalytical Counting. Techniques,": Acladtic Richfield Hanford Company Report ARH-SA-215 (June 1975). <

3/4.11.1.2 DOSE This specification is provided to implement the 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 in Section II. A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radicactive material in liquid effluents will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies that can be-potentially affected by plant 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 Part 141. The dose calculations in the ODCM 1mplement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculationel procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with

~

the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactoi Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," ' Revision 1, October 1977 and Regulatory Guide 1.113 " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

NORTH ANNA - UNIT 1 B 3/4 Il-1

~

RADIOACTIVE EFFLUENTS BASES

~

This Specification applies to the release of liquid effluents from each reactor at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned among the units sharing

. that system.

, 3 //[.11.1. 3 LIQUID RADWASTE TREATMENT The tequirement that the appropriate portions of this system be used, when specifierb provides assurance that the releases of radioactive materials in liquid i 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 the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set isrth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.

=

3/4.11.1.4 LIQUID HOLDUP TANKS The tanks listed in this Specification include all those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.

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

3/4.11.2 CASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE

, This Specification is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20. 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, either within or outside the SITE BOUNDARY to annual average concentrations exceeding the limits specified in Appendix B. Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For MEMBERS l OF THE PUBLIC, who may at times be within the SITE BOUNDARY the occupancy of

, the individual will be sufficiently low to compensate for any increase in the

_ atmospheric diffusion factor above that for the SITE BOUNDARY. The specified NORTH ANNA - UNIT 1 B 3/4 11-2 l

RADIOACTIVE EFFLUENTS BASES release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/ year to the total body or to less than or equal to 3000 mrems/ year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/ year.

This Specification applies to the release of gaseous effluents from all reactors at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.

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

Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry" Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

3/4.11.2.2 DOSE - NOBLE CASES This Specification is provided to implement the requirements of Sections II.B. III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reaconably achievable." The Surveillance Requirements implement the requirements in Section III. A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appro-priate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in 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 CO, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Caseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1. July 1977. The ODCM equations provided for determining the' air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

NORTH ANNA - UNIT 1 B 3/4 11-3 e

RADIOACTIVE EFFLUENTS BASES 3/4.11.2.3 DOSE - 10 DINE-131. TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM This Socification is provided to implement the requirements of Sections II.C. III. A and IV. A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix 1.

The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous 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 calcula-tional procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substan-tially underestimated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111 " Methods for Estimating Atmospheric Transport and Dispersion of Caseous 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 Speciffcations for iodine-131, radiciodines, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent on the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations- were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

3/4.11.2.4 GASEOUS RADWASTE TREATMENT The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable".

This Specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

6 NORTH ANNA - UNIT 1 B 3/4 11-4

RADIOACTIVE EFFLUENTS BASES 3/4.11.2.5 EXPLOSIVE GAS MIXTURE This Specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of hydrogen and oxygen. (Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen, automatic diversion to recombiners, or injection of dilutants to reduce the concentration below the flammability limits.) Maintaining the concentration of hydrogen and oxygen below their flansnability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3/4.11.2.6 GAS STORAGE TANKS The tanks included in this Specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification to a quantity that is less than the quantity which provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem in an event of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of sa uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Branch Technical Position ETSB 11-5 in NUREG-0800, July 1981.

3/4.11.3 SOLID RADIOACTIVE WASTE This Specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CVR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, vaste/ liquid / SOLIDIFICATION agent / catalyst ratios, waste oil content, waste principal chemical sonstituents, mixing and curing times.

3/4.11.4 TOTAL DOSE This Specification is provided to meet the dose limitations of 40 CFR Part 190 that have now been incorporated into 10 CFR Part 20 by 46 FR 18525.

The' Specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice NORTH ANNA - UNIT 1 B 3/4 11-5

RADIOACTIVE EFFLUENTS BASES the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of ACTION that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR Part 190.11 and 10 CFR Part 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 3.11.1 and 3.11.2. An individual is not con-sidered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

b NORTH ANNA - UNIT 1 B 3/4 11-6

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.1 MONITORING PROGRAM The radiological environmental monitoring program required by this Speci-fication provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiatien exposures of MEMBERS OF THE PUBLIC resulting from the station operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radio-active materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. The initially specified monitoring program will be effective for at least the first three years of commercial operation. Fo1Itsing this period, program changes may be initiated based on operational experience.

The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table 4.12-1 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an j( priori (before the fact) limit representing the capability of a measurement system and not as an jt posteriori (af ter the fact) limit for a particular measurement.

Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry" Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

1 NORTH ANNA - UNIT 1 B 3/4 12-1 l

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING

, BASES 3/4.12.2 LAND USE CENSUS This Specification is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the radiological environmental monitoring program are made if required by the results of this census. The best information from the door-to-door survey, from aerial survey or from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix } to 10 CTR Part 50. Restricting the census to gardens of greater than 50 m provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: 1) 20% of the garden was used for growing broad leaf vegetatign (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/m .

3/4.12/3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in crder to demonstrate that the results are reasonably valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.

k NORTH ANNA - UNIT 1 B 3/4 12-2

5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1-1.

LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1-1.

MAP DEFINING UNRESTRICTED AREAS FOR RADI0 ACTIVE CASE 0US AND LIQUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluents, which allows identification of structures and release points as well as definition of UNRESTRICTED ARE/ T within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall be as shown in Figure 5.1-1.

5.2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape with a dome roof and having the following design features:

a. Nominal inside diameter = 126 feet.

b. Nominal inside height = 190 feet, 7 inches.

c. Minimum thickness of concrete walls = 4.5 feet.

d. Minimum thickness of concrete roof = 2.5 feet.

e. Minimum thickness of concrete floor pad = 10 feet,

f. Nominal thickness of the cylindrial portion of the steel liner = 3/8 inches.

6

g. Net free volume = 1.825 x 10 cubic feet.

h. Nominal thickness of hemispherical dome portion of the 8

steel liner = 1/2 inch.

l DESIGN PRESSURE AND TEMPERATURE 5.2.2. The reactor containment building is designed and shall be maintained l

for a maximum internal pressure of 45 psig and a temperature of 280*F.

NORTH ANNA - UNIT 1 5-1

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ADMINISTRATIVE CONTROLS MEETING FREQUENCY 6.5.1.4 The SNSOC shall meet at least once per calendar month and as convened by the SNSOC Chairman or his designated alternate.

QUORUM 6.5.1.5 A quorum of the SNSOC consists of the Chairman or Vice-Chairman and two members including alternates.

RESPONSIBILITIES 6.5.1.6 The SNSOC shall be responsible for:

a. Review of 1) all procedures required by Specification 6.8.1 and changes thereto, 2) all programs required by Specification 6.8.4 and changes thereto, 3) any other proposed proce' ares or changes thereto as determined by the Station Manager to affect nuclear safety.

b. Review of all proposed tests and experiments that affect nuclear safety.

c. Review of all proposed changes to Appendix "A" Technical Specifications.

d. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.

e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Manager-Nuclear Operations and Maintenance and the Director-Safety Evaluation and Control.

f. Review of events requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months written notification to the

! Commission.

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

h. Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairman of the Station Nuclear Safety and Operating Committee.

i. Review of the Plant Security Plan and implementing procedures and F

shall submit recommended changes to the Chairman of the Station Nuclear Safety and Operating Committee.

j. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the- ~ Chairman of the Station Nuclear Safety and Operating Committee.

i NORTH ANNA - UNIT 1 6-6 1 . .

ADMINISTRATIVE CONTROLS

k. Review of every unplanned onsite release of radioactive material to the environs including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President-Nuclear Operations and to the Director-Safety Evaluation and Control.

1. Review and changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL.

AUTHORITY 6.5.1.7 The SNSOC shall:

a. Recommend to the Station Manager written approval or disapproval of items considered under 6.5.1.6(a) through (d) above.

b. Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.

c. Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to Manager-Nuclear Operations and Maintenance and the Director-Safety Evaluation and Control of disagreement between the SNSOC and the Station Manager; however, the Station Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.

RECORDS 6.5.1.8 The SNSOC shall maintain written minutes of each meeting and copies shall be provided to the Manager-Nuclear Operations and Maintenance and the Director-Safety Evaluation and Control.

6.5.2 SAFETY EVALUATION AND CONTROL (SEC)

FUNCTION 6.5.2.1 SEC shall function to provide independent review of designated activities in the areas of:

a. Nuclear power plant operations l

2

b. Nuclear engineering

c. Chemistry and radiochemistry

d. Metallurgy

.e. Instrumentation and control

f. Radiological safety NORTH ANNA - UNIT 1 67 i

- - - ~ _ _ . . . _ _

I ADMINISTRATIVE CONTROLS

g. Mechanical and electrical engineering

! h. Administrative controls and quality assurance practices

1. Other appropriate fields associated with the unique characteristics i of the nuclear power plant i

COMPOSITION i 6.5.2.2 The SEC staff shall be composed of the Director-Safety Evaluation i and Control and a minimum of three individuals who are qualified as staff

specialists. Each SEC staff specialist shall have an academic degree in an engineering or physical science field and, in addition, shall have a ninimum

of five years technical experience in one or more areas given in Specification 6.5.2.1. These staff specialists shall not be directly involved in the i licensing function.

j CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the Director-Safety

Evaluation and Control to provide expert advice to the SEC.

l l MEETING FREQUENCY 6.5.2.5 The SEC staff shall meet at least once per calendar month for the

! purpose of fostering interaction of reviews regarding safety-related opera-

] tional activities.

9 b

l REVIEW i

j 6.5.2.7 The following subjects shall be reviewed by SEC:

a. Written saf ety evaluations of changes in the stations as described i in the Safety Analysis Report, changes in procedures as described in the Safety Analysis Report and tests or experiments not described in the Safety Analysis Report which are completed without prior NRC i approval under the provisions of 10 CFR 50.59(a)(1). This review is to verify that such changes, tests or experiments did not involve a change in the Technical Specifications or an unreviewed safety

question as defined in 10 CFR 50.59(a)(2) and is accomplished by review of minutes of the Station Nuclear Safety and Operating Committee

and the design change program.

i

b. Proposed changes in procedures, proposed changes in the station, or 1 proposed tests or experiments, any of which may involve a change in

, the Technical Specifications cr an unreviewed safety question as defined in 10 CFR 50.59(a)(2). Matters of this kind shall be i referred to the Director-Safety Evaluation and Control by the Station Nuclear Safety and Operating Committee following its review prior to implementation.

NORTH ANNA - UNIT 1 6-8 1

f

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

ADMINISTRATIVE CONTROLS REVIEW (Cont'd)

c. Changes in the Technical Specifications or license amendments relating to nuclear safety prior to implementation except in those cases where the change is identical to a previously reviewed proposed change.

d. Violations and reportable occurrences such as:

1. Violations of applicable codes, regulations, orders, Technical Specifications, license requirements or internal procedures or instructions having safety significance;

2. Significant operating abnormalities or deviations from normal or expected performance of station safety-related structures, systems, or components; and

3. Reportable occurrences as defined in the station Technical Specification 6.9.1.8.

Review of events covered under this paragraph shall include the results of any investigations made and recommendations resulting from such investigations to prevent or reduce the probability of recurrence of the event.

e. The Quality Assurance Department audit program at least once per 12 months and audit reports.

f. Any other matter involving safe operation of the nuclear power stations which a duly appointed subcommittee or committee member deems appropriate for consideration, or which is referred to the Director-Safety Evaluation and Control by the Station Nuclear Safety and Operating Committee.

g. Reports and meeting minutes of the Station Nuclear Safety and Operating Committee.

AUTHORITY 6.5.2.9 The Director-Safety Evaluation and Control shall report to and advise the Manager-Nuclear Technical Services, who shall advise the Vice President-Nuclear Operations on those areas of responsibility specified in Section 5.5.2.7.

RECORDS 6.5.2.10 Records of SEC activities required by Section 6.5.2.7 shall be

, prepared and maintained in the SEC files and a summary shall be disseminated i

as indicated below each calendar month.

I

1. Vice President-Nuclear Operations

2. Nuclear Power Station Managers

3. Manager-Nuclear Operations and Maintenance NORTH ANNA - UNIT I 6- 9

ADMINISTRATIVE CONTROLS

4. Manager-Nuclear Technical Services

5. Executive Manager-Quality Assurance

6. Others that the Director-Safety Evaluation and Control may designate.

6.5.3 QUALITY ASSURANCE DEPARTMENT FUNCTION 6.5.3.1 The Quality Assurance Department shall function to audit station l activities. These audits shall encompass:

a. The conformance of facility operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months.

b. The performance, training and qualifications of the entire facility

, staff at least once per 12 months.

c. The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 ronths.

d. The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per 24 months.

e. The Station Emergency Plan and implementing procedures at least once per 24 months.

f. The Station Security Plan and implementing procedures at least once per 24 months.

g. Any other area of facility operatior considered appropriate by the Executive Manager-Quality Assurance or the Senior Vice President-Power Operations.

h. The Station Fire Protection Program and implementing procedures at

, least once per 24 months.

i. An independent fire protection and loss prevention program inspection and audit shall be performed at least once per 12 months utilizing either qualified offsite licensee personnel or an outside fire protection firm.

'j . An inspection and audit of the fire protection and loss prevention program shall be performed by a qualified outside fire consultant at least once per 36 months.

Pending NRC approval.

NORTH ANNA - UNIT 1 6-10

ADMINISTRATIVE CONTROLS

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

1. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months.

m. The PROCESS CONTROL PROGRAM and implementing procedures for pro-cessing and packaging of radioactive wastes at least once per 24 months.

n. The performance of activities required by the Quality Assurance Program to meet the provisions of Regulatory Guide 1.21, Revision 1 June 1974 and Regulatory Guide 4.1, Revision 1 April 1975 at least once per 12 months.

AUTHORITY 6.5.3.2 The Quality Assurance Department shall report to and advise the Executive Manager-Quality Assurance, who shall advise the Senior Vice President-Power Operations on those areas of responsibility specified in Section 6.5.3.1.

RECORDS 6.5.3.3 Records of the Quality Assurance Department audits shall be prepared and maintained in the department files. Audit reports shall be disseminated as indicated below:

1. Nuclear Power Station Manager l

l 2. Manager-Nuclear Operations and Maintenance

3. Manager-Nuclear Technical Services

4. Director-Safety Evaluation and Control *

5. Supervisor of area audited

6. Nuclear Power Station Manager Quality Assurance l

l l

Pending NRC approval.

1 NORTH ANNA - UNIT 1 6 -11 l

l

l ADMINISTRATIVE CONTROLS l

6.6 REPORTABLE OCCURRENCE ACTION l

6.6.1 The following actions shall be taken for REPORTABLE OCCURRENCES:

i

a. The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.

b. Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification to the Commission shall be reviewed by the SNSOC and submitted to the Director-Safety Evaluation and Control and the Manager-Nuclear Operations and Maintenance.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

1

a. The facility shall be placed in at least HOT CTANDBY within one hour,

b. The NRC Operations Center shall be notified by telephone as soon as possible and in all casas within one hour. The Manager-Nuclear Operations and Maintenance, and the Director-Safety Evaluation and Control shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the SNSOC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Commission, the Director-Safety Evaluation and Control and the Manager-Nuclear Operations and Maintenance within 14 days of the violation.

6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

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

b. Refueling operations.

O NORTH ANNA - UNIT 1 6-12

ADMINISTRATIVE CONTROLS

c. Surveillance and test activities of safety related equipment.

d. Security Plan implementation.

e. Emergency Plan implementation.

f. Fire Protection Program Implementation.

I

g. PROCESS CONTROL PROGRAM implementation.

h. OFFSITE DOSE CALCULATION MANUAL implementation.

i. Quality Assurance Program for effluent and environmental monitoring, using the guidance in Regulatory Guide 1.21, Revision 1, June 1974 and Regulatory Guide 4.1, Revision 1. April 1975.

6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be reviewed by the SNSOC and approved by the Station Manager prior to implementation and reviewed periodically as set forth in administrative procedures.

6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant supervisory staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

c. The change is documented, reviewed by the SNSOC and approved by the Station Manager within 14 days of implementation.

6.8.4 The following programs shall be established, implemented, and maintained:

a. Primary Coolant Sources Outside Containment l

A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the recirculation spray, safety injection, chemical and volume control, gas stripper, and hydrogen recombiners. The program shall include the following:

(i) Preventive maintenance and periodic visual inspection requirements and (ii) Integrated leak test requirements for each system at refueling cycle intervals or less.

NORTH ANNA - UNIT 1 6-13

8-21-80 i(

i ADMINISTRATIVE CONTROLS

b. In-Plant Radiation Monitorino A program which will ensure the capability to accurately determine i the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

(1) Training of personnel, q

] (ii) Procedures for monitoring, and (iii) Provisions for maintenance of sampling and analysis equipment.

c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit i steam generator tube degradation. This program shall include:

(1) Identification of a sampling schedule for the critical variables and control points for these variables, (ii) Identification of the procedures used to measure the values of

!( . the critical variables. -

I i ,

(iii) Identification of process sampling points, a

(iv) Procedures for the recording and management of data, h

j (v) Procedures defining corrective actions for all control point j

chemistry conditions.

l l

' (vi) A procedure identifying (a) the authority responsible for the 8 interpretation of the data, and (b) the sequence and timing of i administrative events required to initiate corrective action, and l

l (vii) Monitoring of the condensate at the discharge of the condensate pumps for evidence of condenser inleakage. When condenser in-leakage is confirmed, the leak shall be repaired, plugged, or isolated within 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months .

( .

l

' " NORTH ANNA - UNIT 1 6-13a

- ~ ~

8-21-80 ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCES 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted.

STARTUP REPORTS

6. 9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (a) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design 00 has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

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" NORTH ANNA - UNIT 1 6-14

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ADMINISTRATIVE CONTROLS l

l 6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details requested in license conditions based on other commitnents shall be included in this report.

6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of cosnercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

ANNUAL REPORTS 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.

6.9.1.5 Reports required on an annual basis shall include:

a. A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 arem/yr and their associated man-rem exposure

. according to work and job functions,2f e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.

The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20 percent of the individual total dose need not be accounted for. In the aggregate, at least 80

> percent of the total whole body dose received from external sources should be assigned to specific major work functions.

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.

8/ . This tabulation supplements the requirements of f20.407 of 10 CFR Part 20.

k NORTH ANNA - UNIT 1 6-15

ADMINISTRATIVE CONTROLS

b. The complete results of the steam generator tube inservice inspections performed during the report period (Reference Specification 4.4.5.5.b.).

MONTHLY OPERATING REPORT .

6.9.1.6 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the Director, Office of Management and Program Analysis, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Office, of Inspection and Enforcement, no later than the 15th of each month following the calendar month covered by the report.

REPORTABLE OCCURRENCES 6.9.1.7 TP.: REPORTABLE OCCURRENCES of Specifications 6.9.1.8 and 6.9.1.9 below, including corrective actions and measures to prevent recurrence, shall be reported to the NRC. Supplemental reports may be required to fully describe

final resolution of 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. }

PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP 6.9.1.8 The types of events listed below shall be reported within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Director of the Regional Office, or his designate, no later than the first working day following the event, with a written followup report within 14 days.

The written followup report shall include, as a minimum, a completed copy of 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.

a. Failure of the reactor protection system or other systems subject to limiting safety-system settings to initiate the required protective function by the time a monitored parameter reaches the setpoint specified as the limiting safety-system setting in the technical specifications or failure to complete the required protective function.

I NORTH ANNA - UNIT 1 6-16

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ADMINISTRATIVE CONTROLS

b. Operation of the unit or affected systems when any parameter or operation subject to a limiting condition for operation is less conservative than the least conservative aspect of the limiting condition for operation established in the Technical Specifications.

c. Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment.

d. Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady-state conditions during power operation greater than or equal to 1% delta k/k; a calculated reactivity balance indicating a shutdown margin less conservative than specified in the Technical Specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if suberitical, an unplanned reactivity insertion of more than 0.5% delta k/k; or occurrence of any unplanned criticality.

e. Failure or malfunction of one or more components which preventt: or could prevent, by itself, the fulfillment of the functional requirements of system (s) used to copy with accidents analyzed in the SAR.

f. Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the SAR.

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

h. Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the Technical Specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.

1. Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than that assumed in the accident analyses in the safety analysis report or Technical Specifications bases; or discovery during plant life of conditions not specifically considered a in the safety analysis report or Technical Specifications that require remedial action or corrective measures to prevent the existence rr development of an unsafe condition.

j. Offsite releases of radioactive materials in liquid and gaseous effluents that exceed the limits of Specification 3.11.1.1 or 3.11.2.1.

NORTH ANNA - UNIT 1 6-17

ADMINISTRATIVE CONTROLS

k. Exceeding the limits in Specification 3.11.1.4 or 3.11.2.6 for the storage of radioactive materials in the listed tanks. The written follow-up report shall include a schedule and a description of activities planned and/or taken to reduce the contents to within the specified limits.

THIRTY-DAY WRITTEN REPORT 6.9.1.9 The types of events listed below shall be the subject of written reports to the Director of the Regional Office within 30 days of occurrence of the event. The written report shall include, as a minimum, a completed copy of the 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.

a. Reactor protection system or engineered safety feature instru-ment 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,

b. 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.

c. 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.

d. Abnormal degradation of systems other than those specified in item 6.9.1.8(c) above designed to contain radioactive material resulting from the fission process.

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

1. A description of the event and equipment involved.

2. Cause(s) for the unplanned release.

3. Actions taken to prevent recurrence.

4. Consequences of the unplanned release.

NORTH ANNA - UNIT I 6-18 e ma

ADMINISTRATIVE CONTROLS (Continued)

CORE SURVEILLANCE REPORT 6.9.1.10 The F xy limit for Rated Thermal Power (Fxy ) in all core planes containing Bank "D" control rods and in all unrodded core planes, the surveillance power level. P , for Technical Specifications 3.2.1 and 3.2.6, and the FqfTyspeck basis shall be provided to the Director of the Regional Office of Inspection and Enforcement, with a copy to; Director, Office of Nuclear Reactor Regulation Attention: Chief of Core Performance Branch U. S. Nuclear Regulatory Commission Washington, D. C. 20555 at least 60 days prior to cycle initial criticality. In the event that the limits would be submitted at some other time during core life, they shall be submitted 60 days prior to the date the limits would become effective unless otherwise exempted by the Commission.

Any additional information needed to support the F xy AND P submittal will be by request from the NRC and need not bE included in this report.

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT

6.9.1.11 Routine Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial criticality.

The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison (as appropriate) with preoperational studies, operational controls

, 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 3.12.2.

The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations spec.ified in the Table and Figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, 1

A single submittal may be made for a multiple unit station.

i l NORTH ANNA - UNIT 1 6-19

1 ADMINISTRATIVE CONTROLS (CONTINUED)

Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the following: a summary description of the radiological environmental monitoring program; at least two legible maps

covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee in the Interlaboratory Comparison Program, required by participation Specification 3.12.3; discussion of all deviations from the sampling schedule of Table 4.12-1 and discussion of all analyses in which the LLD required by Table 4.12-3 was not achievable.

I

! SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT **

6.9.1.12 Routine Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days af ter January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality.

The Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Et platory Guide 1.21 " Measuring, Evaluating, and Reporting Radioactivity in ,olid Wastes and Releases of Radio-active Materials in Liquid and Gascous Effluents from Light-Water-Cooled Nuclear Power Plants " Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

The Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year. This report shall include an assessment of the radiation doses to the maximum exposed MEMBERS OF THE PUBLIC due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. Annual meteorological data collected over the previous year shall be in the form of joint frequency This distributions meteorological of wind speed, wind direction, and atmospheric stability.

data shall be retained in a file on site and shall be made available to the (i.e.,

l NRC upon request. All assumptions used in making these asseasments l' specific activity, exposure time and location) shall be included in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

Concurrent meteorological conditions

,or historical annual average atmospheric dispersion conditions shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the OFFSITE DOSE CALCULATION MANUAL (0pCM).

0ne map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.

- 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; however, for units with separate radwaste systems, the submittal shall specify l

the releases of radioactive material from each unit.

NORTH ANNA - UNIT 1 6-20

ADMINISTRATIVE CONTROLS If the dose to the maximum exposed MEMBER OF THE PUBLIC due to the radioactive liquid and gaseous effluents from the station during the previous calendar year exceeds twice the limits of Specification 3.11.1.2a, 3.11.1.2.b.

3.11.2.2.a. 3.11.2.2.b. 3.11.2.3.a. or 3.11.2.3.b the dose assessment shall include the contribution from direct radiation. The dose fo the maximum exposed MEMBER OF THE PUBLIC shall show conformance with 40. CFR Part 190, InvironmentalRadiationProtectionStandardsforNuclearPowerpperation.

jhe Radioactive Effluent Release Reports shall include a list of unplanned releases as required to be reported in Technical Specification 6.9.1.9.e from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.12.2.

SPECIAL REPORTS Special reports may be required covering inspections, test and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.

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NORTH ANNA - UNIT 1 6-21

ADMINISTRATIVE CONTROLS 6.9.2 Special reports shall be submitted to the Director of the NRC Regional Office listed in Appendix D, 10 CFR Part 20, with a copy to the Director.

Office of Inspection and Enforcement, U. S. Nuclear Regulatory Commission, Washington, D. c. 20555 within the time period specified for each report.

6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

l 6.10.1 The following records shall be retained for at least five years:

l

a. Records and logs of facility operation covering time interval at each power level.

b. Records and logs of principal maintenance activities, inspections,

, repair and replacement of principal items of equipment related to nuclear safety,

c. Each REPORTABLE OCCURRENCE submitted to the Cormission.

d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

e. Records of. changes made to Operating Procedures.

f. Records of radioactive shipments.

g. Records of sealed source leak tests and results.

h. Records of annual physical inventory of all sealed source material of record.

6.10.2 The following records shall be retained for the duration of the Facility Operating License:

a. Records and drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis Report.

3

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

c. Records of facility radiation and contamination surveys.

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

9 NORTH ANNA - UNIT 1 6-22 l

m ADMINISTRATIVE CONTROLS

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

f. Records of transient or operational cyclee for chose ' facility components identified in Table 5.9-1.

g. Records of reactor tests and experiments.

h. Records of training and qualification for current members of1the plant staff. '

i. Records of in-service inspections performhd pursuant to 'these Technical Specifications.

J. Records of Quality Assurance activities required by the QA Manual. _

k. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments persuant to <10 CFR 50.59.

s

1. Records of meetings of the SNSOC.

m. Records of meetings of the System Nuclear Safety and Operating '

Committee to issuance of Amendment No. 30.

n. Records of analyses required by the radiological environmental monitoring program that would permit evaluation of the accuracy of the analysis at a later date. This would include procedures effective at specified times and QA records showing that these procedures were followed.

8

o. Records of secondary water sampling and water quality,

p. Records for Environmental Qualification which are covered under the provisions of Paragraph 6.13.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained end adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREA -

6.12.1 In lieu of the " control device" or " alarm signal!' required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than'1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work NORTH ANNA - UNIT 1 6-23

v - -. - . . ~ ___ _ .- .- - .. .-

ADMINISTRATIVE CONTROLS I - Permit.* Any individual or group of individuals permitted to enter such areas l 1- shall'be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the

!- radiation dose rate in the area.

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into 'such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.

' c. An individual qualified in tht. protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for . providing positive control over the ' activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiation Work Permit.

The requirements of 6.12.1, above, shall also apply to each high t

g, 6.12.2 radiation area in which the intensity of radiation is greater than 1000 mrem /hr. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or the Plant Health j Physicist.

6.13 ENVIRONMENTAL QUALIFICATION

- 6.13.1 By no later than June 30, 1982 all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of:

Division of Operating Reactors " Guidelines for Evaluating Environmental Qualification . of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or NUREG-0588 " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment", December 1979. Copies of these documents are attached to Order for Modification of License NPF-4 dated October 24, 1980.

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6.13.2 By no later than December 1, 1980, complete and auditable records must i be available and maintained at a central location which describe the

environmental qualification method used for all safety-related electrical i equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.

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Health Physics personnel or personnel escorted by Health Physics personnel j shall be exempt from the RWP issuance requirement during the performance of

their assigned radiation protection duties, provided they comply with approved radiation protection procedures for entry in high radiation areas. ,

NORTH ANNA - UNIT 1 6-24

ADMINISTRATIVE CONTROLS 6.14 PROCESS CONTROL PROGRAM (PCP) 6.14.1 Licensee initiated changes to the PCP:

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1. Shall be submitted to the Commission in the Semiannual Radioactive l 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 rationale !

fer the change without benefit of additional or supplemental information;

b. A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and

c. Documentation of the fact that the change has ber:n reviewed and found acceptable by the SNSOC.

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

6.15 0FFSITE DOSE CALCULATION MANUAL (0DCM) 6.15.1 The ODCM shall be approved by the Commission prior to implementation.

6.15.2 Licensee initiated changes to the ODCM:

1. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain:

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change (s);

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

c. Documentation of the fact that the change has been reviewed and found acceptable by the SNSOC.

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

NORTH ANNA - UNIT 1 6-25

TABLE OF CONTENTS PAGE 1.0 DEFINITIONS .................................................. 1-1 2.0 LIMITING CONDITIONS FOR OPERATION ............................ 2-1 2.1 Non-Radiological ........................................ 2-1 3.0 ENVIRONMENTAL SURVEILLANCE ................................... 3-1 3.1 Non-Radiological Su rveillance . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 i

l 3.1.1 Abiotic - Aquatic ................................ 3-1 i

3.1.2 Biotic Aquatic ................................... 3-1 l

3.1.3 Abiotic - Terrestrial ............................ 3-1 3.1.4 Onsite Meteorology Monitoring .................... 3-1 4.0 SPECIAL SURVEILLANCE AND STUDY ACTIVITIES .................... 4-1

5.0 ADMINISTRATIVE CONTROLS ...................................... 5-1 5.1 Responsibility .......................................... 5-1 5.2 Organization ............................................ 5-1 5.3 Review and Audit ........................................ 5-1 5.3.1 Station Nuclear Safety and Operating Committee (SNSOC) .......................................... 5-1 5.3.1.1 Function ................................ 5-1 5.3.1.2 Responsibility .......................... 5-1

[ 5.3.1.3 Authority ............................... 5-3 i

5.3.1.4 Records ................................. 5-3 5.3.2 Quality Assurance Department ...................... 5-4 5.3.2.1 Function ................................. 5-4 5.3.2.2 Audits ................................... 5-4 l 5.3.2.3 Records .................................. 5-4 5.3.3 Safety Evaluation and Control (SEC) ............... 5-4 5.3.3.1 Function ................................. 5-4 j 5.3.3.2 Review ................................... 5-4 l

5.3.3.3 Responsibility ........................... 5-5 5.3.3.4 Authority ................................ 5-5 5.3.3.5 Records .................................. 5-5 5.4 State and Federal Permits and Certificates .................... 5-5 l

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l TABLE OF CONTENTS (Cant'd) i PAGE 5.5 Procedures .................................................... 5-5 5.5.1 Written Procedures ................................ 5-5 5.5.2 Operating Procedures .............................. 5-5 5.5.3 Procedures for Environmental Surveillance -

Nonradiological ................................... 5-6 5.5.4 Quality Assurance of Program Results .............. 5-6 5.5.5 Changes in Procedures, Station Design or Operation ......................................... 5-6 5.5.6 Consistency with Initially Approved Programs ...... 5-7 5.6 Station Reporting Requirements ................................ 5-7 5.6.1 Routine Reports ................................... 5-7 5.6.2 Nonroutine Reports ................................ 5-8 5.6.2.1 Nonroutine Non-Radiological Environmental Operating Report .........................5-8 5.6.3 Changes in Environmental Technical Specifications.. 5-9 5.6.4 Changes in Permits and Certifications . . . . . . . . . . . . . 5-9 5.7 Re c o rd s Re t en t ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 9 I-ii

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 NON-RADIOLOGICAL - None l

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observations in a form consistent with National Weather Service procedures. Summaries of all data and observations shall be available to the NRC upon request.

Any modification to the onsite meteorological monitoring program as described above, or planned alterations of the area in the vicinity of the meteorological tower (s) that would interfere with the measurement of meteorological conditions representative of the site, will require written approval in accordance with Section 5.6.3.

Bases The collection of meteorological data at the plant site will provide information which will be used to develop atmospheric diffusion parameters to estimate potential radiation doses to the public resulting from actual routine or abnormal releases of radioactive materials to the atmosphere, and to assess the actual impact of the plant cooling system on the atmospheric environment of the site area. A meteorological data collection program as described above is necessary to meet the requirements of subparagraph 50.36(a)(2) of 10 CFR Part 50, Appendix E to 10 CFR Part 50, and 10 CFR Part 51.

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5.5.3 Procedures for Environmental Surveillance - Nonradiological l

Not applicable.

5.5.4 Quality Assurance of Program Results The procedures document shall provide for assurance of the quality of program results, including analytical measurements. This portion of the procedures document shall document the program in policy directives, designate a responsible organization or individuals, include purchased services (e.g., contractual lab or other contract services), include audits by licensee personnel, and include proce dures for revising programs, systems to identify and correct deficiencies, investigate anomalous or suspect results, and review and evaluate program results and reports.

5.5.5 Changes in Procedures, Station Design or Operation Changes in procedures, station design or operation may be made subject to conditions described below, provided such changes are approved by the SNSOC (Review and Audit responsibility per Section 5.3).

a. The licensee may (1) make changes in the station design and operation as described in the FES, FES Addendum and the Environmental Report, (2) make changes in the procedures described in the document developed in accordance with Subsection 5.5, and (3) conduct tests and experiments not described in the document developed in acordance with Subsection 5.5, without prior Commis-sion approval, unless the proposed change, test or experiment involves a change in the objectives of the ETS, an unreviewed environmental question of substantive impact, or affects the requirements of Subsection 5.5.6 of these ETS.

b. A proposed change, test, or experiment shall be deemed to involve an unreviewed environmental question (1) if the probability of magnitude of environmental impact may be increased; or (2) if a possibility for a substantive environmental impact of a different type than any evaluated previously in the FES or FES Addendum may be created.

b I-5-6

c. Tha licances ch211 maintain racords of chrngss in proceduras cnd in facility design or operation made pursuant to this Subsection, to the extent that such changes constitute changes in procedures as described in the document developed in accordance with Section 5.5 or in the FES, FES Addendum and ER. The licensee shall also maintain records of tests and experiments carried out pursuant to paragraph "a" of this Subsection. These records shall include a written evaluation which provides the bases for the determination that the change, test, or experiment does not involve an unreviewed environmental question of substantive impact or constitute a change in the objectives of these ETS, or affect the requirements of Section 5.5.6 of these ETS. The licensee shall furnish to the Commission, annually or at such shorter intervals as may be specified in the license, a report containing descriptions, analyses, interpretations, and evaluations of such changes, tests and experiments.

d. Changes in the procedures developed in accordance with Subsection 5.5 which affect sampling frequency, location, gear, or replication shall be reported to the NRC within 30 days after their implementation. These reports shall describe the changes made, the reasons for making the changes, an evaluation of the environmental impact of these changes, and the statement required under the provisions of Subsection 5.5.6.

5.5.6 Consistency with Initially Approved Programs Any modifications or changes of the procedures developed in accordance with Subsection 5.5 must be governed by the need to maintain consistency with previously used procedures so that direct comparisons of data are technically valid. Such modifications or changes must be

~ justified and supported by adequate comparative sampling programs or studies demonstrating the comparability of results or which provide a basis for making adjustments that would permit direct comparisons.

These demonstrations of comparability shall be submitted to the NRC in accordance with the provision of Subsection 5.5.5 and 5.6.1 of these ETS.

5.6 Station Reporting Requirements 5.6.1 Routine Reports - None.

I-5-7

- t i . . *

( Nonroutine Reports f i 5.6.2

, I 5.6.2.1 Nonroutine Non-Radiological Environmental Operatine Report ,

None. -

s '

1A u l a ..

l J l 4

a .

i 4

i )

1 i 4

i i

s

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'l 4

i

(

i I

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l I

I l ,

(

158

i I

5.6.3 Changes in Environmental Technical Specifications I i

A report shall be made to the NRC prior to implementation of a change I in plant design, in plant operation, or in procedures described in Section 5.5 if the change would have a significant effect on the environment or involves an environmental matter or question not previously reviewed and evaluated by the NRC. The report shall !

include a description and evaluation of the , change and a supporting '

i benefit-cost analysis.

Request for changes in Environmental Technical Specifications shall be submitted to the Director, Office of Nuclear Reactor Regulation, for l

review and authorization. The request shall include an evaluation of the environmental impact of the proposed changes and a supporting benefit-cost analysis.

5.6.4 Changes in Permits and Certifications None

! 5.7 Records Retention 1

Records and logs relative to the following areas shall be made and retained for the life of the station:

a. Records and drawings detailing plant design changes and modifications made to systems and equipment as described in Section 5.6.3.

b. Reports from environmental monitoring, surveillance, and special surveillance and study activities required by these Environmental Technical Specifications.

I b

a i

I-5,-9

.---a-

- w - ., - , . , , , -. . - - - - - . .

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

INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS Action............................................................. 1-1 Axial Flux Difference ............................................. 1-1 Channel Calibration................................................ 1-1 Channel Check...................................................... 1-1 Channel Functional Test............................................ 1-1 Containment Intergrity............................................. 1-1 Controlled Leakage................................................. 1-2 Core Alteration.................................................... 1-2 Dose Equivalent I-131.............................................. 1-2 E-Average Disintegration Energy.................................... 1-2 Engineered Safety Feature Response T1me............................ 1-3 Frequency Notation................................................. 1-3 Gaseous Radwaste Treatment System.................................. 1-3 Identified Leakage................................................. 1-3 Member (s) of the Put11c............................................ 1-3 Offsite Dose Calculation Manual (0DCM)............................. 1-4 Operable - Operability............................................. 1-4

0perational Mode - Mode............................................ 1-4 Physics Tests...................................................... 1-4 Pre s su re Bound ry Le akage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 Process Control Program (PCP)...................................... 1-4 Purge-Purging...................................................... 1-4 Quadrant Power Tilt Rat 1o.......................................... 1-5 NORTH ANNA - UNIT 2 I

INDEX (Cont'd.)

1 PAGE

Rated Thermal Power................................................ 1-5 Rea c t o r Trip Sy s t em Re sp onse Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 Reportable Occurrence.............................................. 1-5 Shutdown Margin.................................................... 1-5 Site Boundary...................................................... 1-5 Solidification..................................................... 1-5 Source Check....................................................... 1-5 Staggered Test Basis............................................... 1-6 1

i Thermal Power.. ................................................... 1-6 Unidentified Leakage............................................... 1-6 j Unrestricted Area.................................................. 1-6 Ventilation Exhaust Treatment Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 i Venting............................................................ 1-6 4

1 4

L j

I l

t

.i i

1 NORTH ANNA - UNIT 2 II i

4 INDEX j SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i

l SECTION l Page 2.1 SAFETY LIMIT _S l 4

Reactor Core...................................................... 2-1 l Reactor Coolant System Pressure................................... 2-1

2. 2 LIMITING SAFETY SYSTEM SETTINGS l Reactor Trip Setpoints............................................ 2-5 I

i e

4 1

l BASES l

SECTION Page 2.1 SAFETY LIMITS Reactor Core...................................................... B 2-1 Reactor Coolant System Pressure................................... B 2-2 l 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip Setpoints............................................ B 2-3 6

2

}

III a

NORTH ANNA - UNIT 2 4

I i

(

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION g -

3/4.0 APPLICABILITY................................................, 3/4 0 1 l 3/4.1 REACTIVITY CONTROL SYSTEMS l

3/4.1.1 BORATION CONTROL Shutdown Margin - T,yg > 200*F............................. 3/4 1-1 Shutdown Margin - T,yg i 200*F............................. 3/4 1-3 Boron Dilution-Valve Position.............................. 3/4 1-4 Moderator Temperature Coefficient.......................... 3/4 1-5 Minimum Temperature for Criticality........................ 3/4 1-7

( 3/4.1.2 BORATION SYSTEMS Flow Paths - Shutdown...................................... 3/4 1-8 Flow Paths - Operating..................................... 3/4 1-9 Charging Pump - Shutdown................................... 3/4 1-11 Charging Pumps - Operating................................. 3/4 1-12 Borated Water Sources - Shutdown........................... 3/4 1-1?

Borated Water Sources - Operating.......................... 3/4 1-14 3/4.1.3 MOVABLE CONTROL ASSEMBLIES

, Group Height............................................... 3/4 1-16 Position Indicator Channel s-Operating. . . . . . . . . . . . . . . . . . . . . . 3/4 1-19 Position Indicator Channels-Shutdown....................... 3/4 1-20 Rod Drop Time.............................................. 3/4 1-21 Shutdown Rod Insertion Limit............................... 3/4 1-22 Control Rod Insertion Limits............................... 3/4 1-23 ,

r

{.

NORTH ANNA - UNIT 2 IV

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PACE

~3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 Axial Flur. Difference .................................. 3/4 2-1 3/4.2.2 Heat Flux Hot Channel Factor ........................... 3/4 2-5 3/4.2.3 Nuclear Enthalpy Hot Channel Factor .................... 3/4 2-9 3/4.2.4 Quadrant Power Tilt Ratio .............................. 3/4 2-12 3/4.2.5 DNB Parameters ......................................... 3/4 2-15 3/4.2.6 Axial Power Distribution ............................... 3/4 2-17 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRGHENTATION .................... 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM I N S TR UM EN TAT I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-15 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring ................................... 3/4 3-38 Movable Incore Detectors ............................... 3/4 3-42 Auxiliary Shutdown Panel Monitoring Instrumentation. . . . . 3/4 3-43 Accident Monitoring Instrumentation .................... 3/4 3-46 Fire Detection Instrumentation ......................... 3/4 3-49 Axial Power Distribution Monitoring System ............. 3/4 3-51 Radioactive Liquid Effluent Monitoring Instrumentation . 3/4 3-53 Radioactive Gaseous Effluent Monitoring Instrumentation 3/4 3-60 3/4.4 REACTOR COOLANT SYSTEM 43/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation ........................... 3/4 4-1 Hot Standby ........................................... 3/4 4-2 Shutdown .............................................. 3/4 4-3 Isolated Loop ......................................... 3/4 4-4 Isolated Loop Startup ................................. 3/4 4-5 NORTH ANNA - UNIT 2 V

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION Pace 3/4.4.2 SAFETY VALVES - SHUTD0WN.................................. 3/4 4-6 3/4.4.3 SAFETY and RELIEF VALVES - OPERATING Safety Va1ves............................................. 3/4 4-7 Relief Va1ves............................................. 3/4 4-7a 3/4.4.4 PRE 55URIZER........... ................................... 3/4 4-8 3/4.4.5 STEAM GENERATORS.......................................... 3/4 4-9 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................. 3/4 4-16 Operational Leakage....................................... 3/4 4-17

, 3/4.4.7 CHEMISTRY................................................. 3/4 4-19 3/4.4.8 SPECIFIC ACTIVITY......................................... 3/4 4-22 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.................................... 3/4 4-29 Pressurizer............................................... 3/4 4-30

>f Ove rpressure Protection Sys tems. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-31

! 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1, 2 and 3 Components..................... 3/4 4-32

} S te am Gene r ato r Supports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-33 1

1 l 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 1 3/4.5.1 ACCUMULATORS.............................................. 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,yg > 350*F............................ 3/4 5-3 i

i 3/4.5.3 ECCS SUBSYSTEMS - T < 350*F............................ 3/4 5-6 avg i 3/4.5.4 BORON INJECTION SYSTEM i

Boron Injection Tank...................................... 3/4 5-8 Heat Tracing.............................................. 3/4 5-9 3/4.5.5 REFUELING WATER STO RAGE TANK. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-10 i

i t

1./

i NORTH ANNA - UNIT 2 VI e

i

- *~ _ . - - -- __ ___*_

I 1

l INDEX !

LIMITINGCONDITIONSFOROPERATiONANDSURVEILLANCEREQUIREMENTS SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 1

3/4.6.1 CONTAINMENT Containment Integrity..................................... 3/4 6-1 Containment Leakage....................................... 3/4 6-2 Containment Air Locks..................................... 3/4 6-4 i I n te rnal P re s s u re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-6 Air Temperature........................................... 3/4 6-8

[

l Containment Structural Integrity.......................... 3/4 6-9 l 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Conta i nment Quench Sp ray System. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-10 Containment Recirculation Spray System.................... 3/4 6-11 l Chemical Addition System.................................. 3/4 6-13 1 l

3/4.6.3 CONTAINMENT ISO LATION VALVES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-14 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers........................................ 3/4 6-32 Electric Hydrogen Recombiners............................. 3/4 6-33 Waste Gas Charcoal Filter System.......................... 3/4 6-34 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM i Steam Jet Air Ejector..................................... 3/4 6-36 b

l 1

NORTN ANNA - UNIT 2 VII i

i -

1_ _ _._

i INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TUR8INE CYCLE Safety Va1ves............................................. 3/4 7-1 Auxiliary Feedwater System................................ 3/4 7-5 Emergency Condensate Storage Tank......................... 3/4 7-7 Activity.................................................. 3/4 7-8 Main Steam Trip Va1ves.................................... 3/4 7-10 Steam Turbine Assemb1y.................................... 3/4 7-11 Overspeed Protection...................................... 3/4 7-12 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION........... 3/4 7-13 3/4.7.3 COMPONENT COOLING WATER SUBSYSTEM......................... 3/4 7-14 3/4.7.4 SERVICE WATER SYSTEM...................................... 3/4 7-15 3/4.7.5 ULTIMATE HEAT SINK........................................ 3/4 7-16

['

3/4.7.6 FLOOD PROTECTION.......................................... 3/4 7-17 3/4.7.7 CONTROL ROOM EMERGENCY HABITABILITY SYSTEMS............... 3/4 7-18 3/4.7.8 SAFEGUARDS AREA VENTILATION SYSTEM........................ 3/4 7-21 3/4.7.9 RESIDUAL HEAT REMOVAL SYSTEM RHR - Operating........................................... 3/4 7-23 RHR - Shutdown............................................ 3/4 7-24 3/4.7.10 SNUBBERS.................................................. 3/4 7-25 3/4.7.11 S EALED SOURCE CONTAMINATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-51 2

3/4.7.12 SETTLEMENT OF CLASS 1 STRUCTURES. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-53

, 3/4.7.13 GROUNDWATER LEVEL-SERVICE WATER RESERVOIR................. 3/4 7-57 I

i.

1

' Vill l

NORTH ANNA - UNIT 2

INDEX I

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i

i SECTION PAGE 3/4.7.14 FIRE SUPPRESSION SYSTEMS Fi re Suppres s ion Wate r Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-59 Low Pressure CO Systems................................. 3/4 7-63 2

High Pressure CO Systems................................ 3/4 7-65 2

Halon Systems............................................ 3/4 7-67 Fire Hose Stations....................................... 3/4 7-68 3/4.7.15 PENETRATION FIRE BARRIERS................................ 3/4 7-70 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES

)

Operating................................................ 3/4 8-1 Shutdown................................................. 3/4 8-10 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. Distribution - Operating............................ 3/4 8-11 A.C. Distribution - Shutdown............................. 3/4 8-12 0.C. Distribution - Operating............................ 3/4 8-13 D.C. Distribution - Shutdown............................. 3/4 8-15 Containment Penetration Conductor Overcurrent Protective 0evices..................................... 3/4 8-16 Motor Operated Valves Thermal Overload Protection and/or Bypass 0evices.................................. 3/4 8-21 e

NORTH ANNA - UNIT 2 IX

I

.' INDEX t., e LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/A.9 REFUELING OPERATIONS 3/4.9.1

. BORON CONCENTRATION........................................ 3/4 9-1 !

3/4.9.2 INSTRUMENTATION............................................ 3/4 9-2 3/4.9.3 DECAY TIME................................................. 3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-4 3/4.9.5 COMMUNICATIONS............................................. 3/4 9-6

. 3/4.9.6 MANIPULATOR CRANE OPERA 8ILITY.............................. 3/4 9-7 3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT.............................. 3/4 9-8 3/4.9.8 RESIOUAL HEAT REMOVAL AND COOLANT CIRCULATION

. All Water Leve1s........................................... 3/4 9-9 Low Water Leve1............................................ 3/4 9-9a i

( -

3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM............. 3/4 9-10

! 3/4.9.10 WATER LEVE L- REACTO R V ES S E L. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-11

! 3/4.9.11 SPENT FUEL PIT WATER LEVEL................................. 3/4 9-12 3/4 9.12 FUEL BUILDING VENTI LATION SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-13 3/4.10 SPECIAL' TEST EXCEPTIONS ,

3/4.10.1 SHUTD0WN' MARGIN.............................. ............. 3/4 10-1 3/4.10.2 GROUP HEIGHT INSERTION AND POWER DISTRIBUTION.............. 3/4 10-2 3/4.10.3 PHYSICS TEST...................................'............ 3/4 10-3

3/4.10.4 R EACTO R COO LANT L00 P S. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 10-4 3/4.10.5 POSITION INDICATOR CHANNELS-SHUTDOWN....................... 3/4 10-5 NORTH ANNA - UNIT 2 X

- - - - - , - - - - - - - ---. _ _ _m,-.- -. _,--..4 - a - . , - - - - - , - - . y

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION FAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration............................................. 3/4 11-1 Dose...................................................... 3/4 11-5 Liquid Radwaste Treatment................................. 3/4 11-6 Liquid Holdup Tanks....................................... 3/4 11-7 3/4.11.2 GASEOUS EFFLUENTS Dose Rate................................................. 3/4 11-8 Dose-Noble Cases.......................................... 3/4 11-13 Dose-Iodine-131, Tritium, and Radionuclides in Particulate Fo rm. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-14 Gaseous Radwaste Treatment................................ 3/4 11-15 Explosive Gas Mixture..................................... 3/4 11-16 Gas Storage Tanks......................................... 3/4 11-17 3 /4.11. 3 SOLID RADIOACTIVE WASTE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-18 3/4.11.4 TOTAL D0SE................................................ 3/4 11-19 i

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING l

( 3/4.12.1 MONITORING PR0 GRAM...................................... 3/4 12-1

3/4.12.2 LAND USE CENSUS......................................... 3/4 12-13 3/4.12.3 INTERLABORATORY COMPARISON.............................. 3/4 12-14 i

I l

NORTH ANNA - UNIT 2 XI l J

INDEX ,

BASES SECTION PAGE 3/4.0 APPLICABILITY................................................ B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 80 RATION CONTR0L........................................... B 3/4 1-1 3/4.1.2 80 RATION SYSTEMS........................................... B 3/4 1-3 3/4.1.3 MOVAB LE CONTROL ASS EMB LIES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX 0IFFERENCE...................................... B 3/4 2-1 1

3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT

CHANNEL FACT 0RS............................................ B 3/4 2-4 i

3/4.2.4 QUADRANT POWER TILT RATI0.................................. B 3/4 2-5 3/4.2.5 ONB PARAMETERS............................................. B 3/4 2-6 1

l 3/4.2.6 AXIAL POWER DISTRIBUTION................................... B 3/4 2-6 l

l >

l NORTH ANNA - UNIT 2 XII

( INDEX 8ASES -

SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY -

FEATURES (ESF) INSTRUMENTATION.............. 8 3/4 3 3/4.3.3 MONITORING INSTRUMENTATION............................... B 3/4 3-1 3/4.4 REACTOR COOLANT SYSTEM t

3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION. . . . . . . . . . . . B 3/4 4-1 '

3/4.4.2 and 3/4.4.3 SAFETY AND RELIEF VALVES....... ............. B 3/4 4-2

( 3/4.4.4 PRESSURIZER.............................................. 8 3/4 4-2 3/4.4.5 STEAM GENERATORS......................................... B 3/4 4-3 3/4.4.6 REACTOR COO LANT SYSTEM LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-4 3/4.4.7 CHEMISTRY................................................ 8 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY........................................ B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS.............................. B 3/4 4-6 3/4.4.10 STRU CTU RA L INT EG R ITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-16

(

NORTH ANNA - UNIT 2 XIII

INDEX

)

BASES SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS............................................... B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS................................ 8 3/4 5-1 3/4.5.4 BORON INJECTION SYSTEM..................................... 8 3/4 5-2 3/4.5.5 REFUELING WATER STORAGE TANK (RWST)........................ 8 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS .

3/4.6.1 PRIMARY CONTAINMENT.......................................... 8 3/4 6-1 ! ;

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS......................... 8 3/4 6-3 l

i 3/4.6.3 CONTAINMENT ISOLATION VALVES................................. 8 3/4 6-3 ,

3.4/6.4 COMBUSTIBLE GAS CONTR0L...................................... 8 3/4 6-4 8 3/4 6-4 3/4.6.5 SUBSTMOSPHERIC PRESSURE CONTROL SYSTEM.......................

\ b l

NORTH ANNA - UNIT 2 XIV

INDEX

( BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE............................................ B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... B 3/4 7-4 3/4.7.3 COMPONENT E00 LING WATER SUBSYSTEM. . . . . . . . . . . . . . . . . . . . . . . . B 3/4 7-4 i

3/4 7.4 SERVICE WATER SY5 TEM..................................... B 3/4 7-4 3/4.7.5 ULTIMATE HEAT SINK....................................... B 3/4 7-5 3/4.7.6 F LOO D P ROT ECTIO N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 7-5 3/4.7.7 CONTROL ROOM EMERGENCY HABITABILITY...................... B 3/4 7-5 3/4.7.8 SAFEGUARDS AREA VENTILATION SYSTEM....................... B 3/4 7-5 3/4.7.9 RESIOUAL HEAT REMOVAL SYSTEMS............................ E 3/4 7-6 3/4.7.10 SNUBBERS................................................. B 3/4 7-6 3/4.7.11 SEALED SOURCE CONTAMINATION.............................. B 3/4 7-7 3/4.7.12 SETTLEMENT OF CLASS 1 STRUCTURES......................... B 3/4' 7-7 3/4.7.13 GROUN0 WATER' LEVEL - SERVICE WATER RESERVOIR.............. B 3/4 7-9 3/4.7.14 FIRE SUPPRESSION SYSTENS................................. B 3/4 7-9 3/4.7.15 PENETRATION FIRE BARRIERS................................ B 3/4 7-10 l 3/4.8 ELECTRICAL POWER SYSTEMS 1

l 3/4.B.1 and 3/4.B.2 A.C. AND D.C. POWER SOURCES AND DISTRIBUTION................................ B 3/4 B-1 NORTM ANNA - UNIT 2

. - ~ ~ ~

l

. s INDEX

)

BASES SECTION PAGE 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION....................................... B 3/4 9-1 l

3/4.9.2 INSTRUMENTATION........................................... B 3/4 9-1 3/4.9.3 DECAY TIME................................................ B 3/4 9-1 3/4.9.4 CONTAINMENT BUI LDING PEN ETRATIONS. . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.5 COMMUNICATIONS............................................ 8 3/4 9-1

, 3/4.9.6 MANIPULATOR CRANE OPERA 8I LITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT............................. 8 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COO LANT CIRCULATION. . . . . . . . . . . . . 8 3/4 9-2

[ 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISGLATION SYSTEM............ 8 3/4 9-2 L

l 3/4.9.10 and 3/4.9.11 WATER LEVEL-REACTOR VESSEL AND SPENT L FUEL PIT................................................ B 3/4 9-3 '

3/4.9.12 FUEL BUILDING VENTI LATICN SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-3 l

[

l 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN........................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION l LIMITS.................................................... B 3/4 10-1 l

3/4.10.3 PHYSICS TESTS............................................. B 3/4 10-1

!; 3/4.10.4 R EACTO R COO LANT L00 PS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 10-1

, 3/4.10.5 POSITION INDICATOR CHANNELS - SHUTD0WN.................... B 3/4 10-1 0

F

o 6 .

L

?

)

NORTH ANNA - UNIT 2 xy1 e a e m+ e s y .

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

^

l INDEX BASES SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS........................................ B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS....................................... B 3/4 11-2 3/4.11.3 SOLID RADI0 ACTIVE WASTE................................. B 3/4 11-5 3/4.11.4 TOTAL D0SE.............................................. B 3/4 11-5 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM...................................... B 3/4 12-1 3/4.12.2 LAND USE CENSUS......................................... B 3/4 12-2 l 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM...................... B 3/4 12-2 6

NORTH ANNA - UNIT 2 XVII

INDEX DESIGN FEATURES SECTION _ PACE 5.1 SITE E)clusion Area .............................................. 1 Li u Population Zone ......................................... 5-l!

Map Defining UNRESTRICTED AREAS For Radioactive Gaseous and Liquid Effluents ............................................ 3-2 5.2 CONTAINMENT -^

Configuration ............................................... 5-1 Design Pressure and Temperature ............................. 5-1 5.3 REACTOR CORE i I

Fuel Assemblies ............................................. 5-4 Control Rod Assemblies ...................................... 5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature ............................ 9 5-4 Volume .............................................'.......... 5-4 S.5 METEOROLOGICAL TOWER LOCATION ...................'........... 5-5 5.6 FUEL STORAGE . ,

Criticality ......................'.................,.......... 5-5.

Drainage ...........................'......................... 5-5 Capacity ..........................,............. 3........... 5-5

)

5.7 COMPONENT CYCLE OR TRANSIENT LIMIT .......................... 5-6 y

~4 b

r f

I f

NORTH ANNA - UNIT 2 XVIII I /

n l - ,'

INDEX )

ADMINISTRATIVE CONTROLS SFCTION PAGE 6.1 RESPONSIBILITY................................................ 6-1

'6. 2 ORGANIZATION 0ffsite......................................................... 6-1 Facility Staff.................................................. C-1

~.. _

Safety Engineering Staff........................................ 6-la Shift Technical Advisor......................................... 6-la 6.3 FACILITY STAFF QUALIFICATIONS................................ 6-6 6.4 TRAINING..................................................... 6-6 6.5 REVIEW AND AUDIT 6.5.1 STATION NUCLEAR SAFETY AND OPERATING COMMITTEE (SNSOC)

Function.. ..................................................... 6-6 T' Composition..................................................... 6-6 Alternates...................................................... 6-6 Meeting Frequency............................................... 6-7 Quorum......,................................................... 6-7 Responsibilities................................................ 6-7 Authcri.ty....................................................... 6-8 Records......................................................... 6-8

_ 6.5.2 SAFETY EVALUATION AND CONTROL (SEC)

Function........................................................ 6-8 Composition..................................................... 6-9 P \

- N' ORTH ANNA-UNIT 2 XIX

\

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE Consultants.................................................... 6-9 Heeting Frequency.............................................. 6-9 Review......................................................... 6-9 Authority...................................................... 6-10 Records........................................................ 6-10 6.5.3 QUALITY ASSORANCE DEPARTMENT Function....................................................... 6-11 Authority...................................................... 6-12 Records........................................................ 6-12 6.6 REPORTABLE OCCURRENCE ACTI0N................................... 6-13 6.7 SAFETY LIMIT VIOLATION......................................... 6-13 6.8 PROCEDURES..................................................... 6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS AND REPORTABLE OCCURRENCES................... 6-14 6.9.2 SPECIAL REP 0RTS.............................................. 6-21 6 .10 R ECO RD RETENTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 -2 2 6.11 RADIATION PROTECTION PR0 GRAM.................................. 6-23 6.12 H IGH RAD I ATION AR EA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 -2 3 6.13 PROCESS CONTROL PROGRAM (PCP)................................. 6-24 6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM) . . . . . . . . . . . . . . . . . . . . . . . . . 6-25 I

NORTH ANNA - UNIT 2 XX

1.0 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.

AXIAL FLUX DIFFERENCE 1.2 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals, expressed in % of RATED THERMAL POWER between the top and bottom halves of a two section excore neutron detector.

CHANNEL CALIBRATION 1.3 A CHANNEL CALIBRATION shall be the adj ustment , as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRA-TION shall encompass the entire channel including the sensor and 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.

CHANNEL CHECK 1.4 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indica-tions and/or status derived from independent instrumentation channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST 1.5 A 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.

4 k

i

' b. Bistable channels -

the injection of a simulated signal into the

sensor to verify OPERABILITY including alarm and/or trip functions.

l CONTAINMENT INTEGRITY

, 1.6 CONTAINMENT INTEGRITY shall exist when:

j 1.6.1 All penetrations required to be closed during accident j conditions are either:

i f NORTH ANNA - UNIT 2 1-1 4

6

-i-y,, - , , - - ,-,~y-,-.-- m,-. , - + , , - - - - . , . - . - - ~ ~ , - . , . - - , . -- . - - - - - . - - - - - ~ - - - . - -" * *--- - --~r-*

1.0 DEFINITIONS (Continued)

a. Capable of being closed by an OPERABLE containment auto-matic isolation valve system, or

b. Closed by manual valves, blind flanges, or deactivated auto-matic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.1, 1.6.2 All equipment hatches are closed and sealed, 1.6.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, 1.6.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and 1.6.5 The sealing mechanism associated with each penetration (e.g.

welds, bellows or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.7 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION 1.8 CORE ALTERATION shall be the movement or manipulation of any com-ponent within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a saf- conservative position.

DOSE EQUIVALENT I-131 1.9 The DOSE EQUIVALENT 1-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131 I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

t - AVERAGE DISINTEGRATION ENERGY 1.10 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor ecolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in r MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

i NORTH ANNA - UNIT 2 1-2

1.0 DEFINITIONS (Continued)

ENGINEERED SAFETY FEATURE RESPONSE TIME 1

l 1.11 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time I

interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their re-quired positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.12 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

CASE 0US RADWASTE TREATMENT SYSTEM 1.13 A GASEOUS RADWASTE TREATMENT SYSTEM is the system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment. The system is composed of the waste gas decay tanks, regenerative heat exchanger, waste gas charcoal filters, process vent blowers, vaste gas surge tanks and waste gas diaphram compressor.

IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:

a. Leakage (except. CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor coolant system leakage through a steam generator to the secondary system.

hEMBER(S) 0F THE PUBLIC 1.15 MEMBER (S) 0F 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, ocassionally enter an area that is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radio-active materials.

NORTH ANNA - UNIT 2 1-3 L

l

1.0 DEFINITIONS (Continued) 0FFSITE DOSE CALCULATION MANUAL (ODCM) 1.16 The OFFSITE DOSE CALCULATION MANUAL shall contain the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints and the specific monitoring locations of the environmental radiological monitoring program.

OPERABLE - OPERABILITY 1.17 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL MODE - MODE J.18 An OPERATIONAL MODE (i.e. , MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.1.

PHYSICS TESTS 1.19 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.20 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

PROCESS CONTROL PROGRAM (PCP) 1.21 The PROCESS CONTROL PROGRAM shall contain the current formula, sampling,

' analyses, tests and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 71 and Federal and State regulations and other requirements governing the disposal of the radioactive waste.

PURGE - PURGING 1.22 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

NORTH ANNA - UNIT 2 1-4

1.0 DEFINITIONS (Continued)

QUADRANT POWER TILT RATIO 1.23 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.24 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2775 MWt.

REACTOR TRIP LYSTEM RESPONSE TIME 1.25 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE OCCURRENCE 1.26 A REPORTABLE OCCURRENCE shall be any of those conditions specified in Specification 6.9.1.8 and 6.9.1.9.

SHUTDOWN MARGIN 1.27 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be suberitical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1.28 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased or otherwise controlled by the licensee.

SOLIDIFICATION 6

1.29 SOLIDIFICATION shall be the conversion of wet wastes into a solid form that meets shipping and burial ground requirements.

SOURCE CHECK 1.30. A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to radiation. This applies to installed

, radiation monitoring systems,

.i NORTH ANNA - UNIT 2 1-5

1.0 DEFINITIONS (Continued)

STAGGERED TEST BASIS 1.31 A STAGGERED TEST BASIS shall consist of:

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

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

THERMAL POWER 1.32 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

UNIDENTIFIED LEAKAGE 1.33 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UNRESTRICTED AREA 1.34 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY where access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM 1.35 A VENTILATION EXHAUST TREATMENT SYSTEM is the system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.36 VEhflNG is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

NORTH ANNA - UNIT 2 1-6

. )

TABLE 1.1 OPERATIONAL MODES REACTIVITY % RATED AVERAGE C001 ANT MODE gNDITION,K gg THERMAL POWER

TEMPERATURE

1. POWER OPERATION 2.0.99 > 5% 2,350*F

2. STARTUP 2.0.99 < 5%

. 2.350*F

3. HOT STANDBY < 0.99 0 2,350*F

4. HOT SHUTDOWN < 0.99 0 350*F > T

> 200*F avg

5. COLD SHUTDOWN < 0.99 0 < 200*F

6. REFUELING ** < 0.95 0 < 140*F i

l i

l l

Excluding decay heat.

- Fuel in the reactor vessel with the vessel head closure bolts less than fully

~

tensioned or with the head removed.

l NORTH ANNA - UNIT 2 1-7 i

TABLE 1.2 FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

R At least once per 18 months.

S/U Prior to each reactor startup.

P Completed prior to each release.

N.A. Not applicable.

b NORTH ANNA - UNIT 2 1-8

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding Specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time inte rvals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour ACTION shall be initiated to place the unit in a MODE in which the Specification does not apply by placing it, as applicable, in:

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual Specifications.

This Specification is not applicable in MODES 5 or 6.

3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual Specifications.

3.0.5 When a system, subsystem train, component or device is determined to be inoperable solely because its emergency power source is inoperable or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding subsystem (s),

train (s), component (s) and device (s) are OPERABLE, or likewise satisfy the requ,irements of this Specification. Unless both conditions (1) and (2) are satisfied, the unit shall be placed in at lesst HOT STANDBY within I hour, in at least HOT SHUTDOWN within the next 6 hourc, and in at least COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

This Specification is not applicable in MODES 5 or 6.

NORTH ANNA - UNIT 2 3/4 0-1

INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3. 9 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm /

trip setpoints of these channels shall be determined and adjusted in accordance with the OFFSITE DOSE CALCULATION MANUAL (0DCM).

APPLICABILITY: At all times.

ACTION:

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

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, for reasons other than a above, take the ACTION shown in Table 3.3-12. Exert best efforts to return the instruments to OPERABLE status within 30 days and, if unsuccessful, explain in the next Seminannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

c. The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1.9.b are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3. 9 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-12.

NORTH ANNA - UNIT 2 3/4 3-53

TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION O

MINIMUM CHANNELS 5 INSTRUMENT OPERABLE ACTION I

g 1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND

y AUTOMATIC TERMINATION OF RELEASE

" 26

a. Liquid Radwaste Effluent Line (1)

2. GROSS BETA OR GAMMA RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE

a. Service Water System Effluent Line (1) 26

b. Circulating Water System Effluent Line (1) 29 Y

g 3. CONTINUOUS COMPOSITE SAMPLERS AND SAMPLER FLOW MONITOR

a. Clarifier Effluent Line (1) 26

4. FLOW RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line (1) 27

TABLE 3.3-12 (Continued) h .

d

~

. RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION E

I MINIMUM CHANNELS g OPERABLE ACTION y INSTRUMENT

5. TANK LEVEL INDICATING DEVICES *

a. Refueling Water Storage Tank (1) 28

b. Casing Cooling Storage Tank (1) 28 PG Water Storage Tanks ** (1) 28 c.

M Boron Recovery Test Tanks ** (1) 28 c- d.

Y 0;

Tanks included in this Specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.

- This is a shared system with Unit 1.

TABLE 3.3-12 (Continued)

TABLE NOTATION ACTION 26 - With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and analyzed for gross radioactivity {petaorgamma)atalowerlimitofdetection of at least 10 microcurie / gram or an isotopic7 radioactivity at a lower limit of detection of at least 5x10 microcuries/ gram.

ACTION 27 - 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. Design

( capacity performance curves generated in situ may be used to estimate flow.

ACTION 28 - With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, liquid additions to this tank may continue provided the tank liquid level is estimated during all liquid additions to the tank.

ACTION 29 - With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, make repairs as soon as possible. Grab samples cannot be obtained via this pathway.

i NORTH ANNA - UNIT 2 3/4 3-56

4

$ TABLE 4.3-12 N ~

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

, CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL e

CHECK CALIBRATION TEST 4

$-D INSTRUMENT CHECK N 1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Liquid Radwaste Effluent Line D D R Q(1)

2. GROSS BETA OR CAMMMA RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE y a. Service Water System Effluent Line D M R Q(2)

U

b. Circulating Water System Effluent Line D M R Q(2)

3. CONTINUOUS COMPOSITE SAMPLERS AND SAMPLER FLOW MONITOR

a. Clarifier Effluent Line N.A. N.A. R N.A.

ll

5 TABLE 4.3-12 (Continued)

E RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

> CHANNEL I

CHANNEL SOURCE CHANNEL FUNCTIONAL g CHECK CHECK CALIBRATION TEST q INSTRUMENT

4. FLOW RATE MEASUREMENT DEVICES I l

a. Liquid Radwaste Effluent Line D(3) N.A. R Q

5. TANK LEVEL INDICATING DEVICES ***

a. Refueling Water Storage Tank D* N.A. R Q R2, s~

b. Casing Cooling Storage Tank D* N.A. R Q

c. PG Water Storage Tanks ** D* N.A. R Q

d. Boron Recovery Storage Tanks ** D* N.A. R Q

During liquid additions to the tank.

- This is a shared system with Unit 1.

f *** Tanks included in this Specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.

TABLE 4.3-12 (Continued)

TABLE NOTATION (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and concrol room alarn annunciation occur if any of the following conditions exists:

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

2. Instrument controls not set in operate mode.

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

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

2. Instrument controls not set in operate mode.

(3) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL 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 on which continuous, periodic, or batch releases are made.

b 1

s 9

0 4

NORTH ANNA - UNIT 2 3/4 3-59

INSTRUMENTATION RADIOACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive gaseous effluent monitoring instrumentatien channels shown in Table 3.3-13 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.10.2.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined and adjusted in accordance with the ODCM.

APPLICABILITY: As shown in Table 3.3-13 ACTION:

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

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, for reasons other than a above, take the ACTION shown in Table 3.3-13. Exert best efforts to return the instruments to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

c. The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1.9.b are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-13.

NORTH ANNA - UNIT 2 3/4 3-60 i

g TABLE 3.3-13 4x RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION li MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION Ci 1. PROCESS VENT SYSTEM w

a. Noble Gas Activity Monitor -

Providing Alarm and Automatic Termination of Release (1)

31

31

b. Iodine Sampler (1)

31

c. Particulate Sampler (1) w 30 d. Process Vent Flow Rate Measuring Device (1)

30 y

e

e. Sampler Flow Rate Measuring Device (1)

30

2. WASTE GAS HOLDUP SYSTEM EXPLOSIVE GAS MONITORING SYSTEM (Shared with Unit 1)

(1) ** 32

a. Hydrogen Monitor

- 32

b. Oxygen Monitor (1)

{

l

b TABLE 3.3-13-(Continued)

N RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION N

E 8 MINIMUM CHANNELS g INSTRUMENT OPERABLE APPLICABILITY ACTION n

g 3. CONDENSER AIR EJECTOR SYSTEM

31

a. Gross Activity Monitor (1)

b. Flow Rate Monitor (1)

30

4. VENTILATION VENT SYSTEM (Shared with Unit 1)

a. Noble Gas Activity Monitor (1)*

31 Y

g b. Iodine Sampler (1)*

31

c. Particulate Sampler (1)* 31

d. Flow Rate Monitor (1)*

30

e. Sampler Flow Rate Monitor (1)* 30

0ne per vent stack.

TABLE 3.3-13 (Continued)

TABLE NOTATION

At all times.

- During process vent system operation (treatment for primary system offgasee).

ACTION 30 - 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 31 - 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 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for gross activity and gamma isotopic activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 32 - With the number of channnels OPERABLE one less than required by the minimum channels OPERABLF requirement, operation of this system may continue for up to 14 days provided grab samples are taken and analyzed daily. With both channels inoperable, operation may continue provided grab samples are taken and analyzed: (1) every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during degassing operations and (2) daily during other operations.

NORTH ANNA - UNIT 2 3/4 3-63

$ TABLE 4.3-13 4

RADIOACTIVE C'ASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS g

g I

CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE g CHECK CHECK CALIBRATION TEST REQUIRED q INSTRUMENT

1. PROCESS VENT SYSTEM

a. Noble Gas Activity Monitor -

Providing Alarm and Automatic

Termination of Release D P R Q(1)

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

N.A.

a c. Particulate Sampler W N.A. N.A.

d. Process Vent Flow Rate D N.A. R Q y

p Measuring Device N.A. *

e. Sampler Flow Rate Monitor D(5) N.A. R

2. WASTE GAS HOLDUP SYSTEM EXPLOSIVE CAS MONITORING SYSTEM N.A. M **

a. Hydrogen Monitor D Q(3)

N.A. M **

b. Oxygen Monitor D Q(4)

2 TABLE 4.3-13 (Continued)

O ..

Q RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

> CHANNEL MODES IN WHICH

' CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE l @ INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED 4

u 3. CONDENSER AIR EJECTOR SYSTEM M *

a. Noble Gas Activity Monitor D R Q(2)

b. Flow Rate Monitor D N.A. R Q *

4. VENTILATION VENT SYSTEM (Shared with Unit 1)

M

R a. Noble Gas Activity Monitor D R Q(2) s~

Y b. Iodine Sampler W N.A. N.A. N.A. *

c. Particulate Sampler W N.A. N.A. N.A.. *

d. Flow Rate Monitor D N.A. R Q *

e. Sample- Flow Rate Monitor D(5) N.A. R N.A.

r TABLE 4.3-13 (Continued)

TABLE NOTATION

At all times other than when the line is valved out and/or locked.

- During process vent system operation (treatment for primary system offgases).

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

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

2. Instrument controls not set in operate mode.

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

1. Instrument indicates measured levels above the alarm setpoint.

2. Instrument controls not set in operate mode. *

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

1. One volume percent hydrogen, balance nitrogen, and

2. Four volume percent hydrogen, balance nitrogen.

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

1. One volume percent oxygen, balance nitrogen, and

2. Four volume percent oxygen, balance nitrogen.

(5) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL 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 on which continuous, periodic or batch releases are made.

l .

i

{

NORTH ANNA - UNIT 2 3/4 3-66

3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION 3.11.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 5.1-1) shall be limited to the con-centrations specified in 10 CFR Part 20 Appendix B. Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For digsolved or ~

entrained noble gases, the concentration shall be limited to 2 x 10 microcuries/ml.

APPLICABILITY: At all times.

ACTION:

With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, withcut delay restore the concentration to within the above limits.

SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.11-1.

4.11.1.1.2 The results of the radioactivity analyses shall be used in accordance with the methods in the CDCM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1.

e NORTH ANNA - UNIT 2 3/4 11-1 l

TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM l

Lower Limit Minimum of Detection Liquid Release Sampling Analysis Type of Activity (LLD)a Type Frequency Frequency Analysis (pCi/ml)

A. Batch P P Each Batch -7 Releages 'E Each Batch PrincipagGamma 5x10 Tanks Emitters

-6 I-131 1x10 P M Dissolved and 1x10 One Batch /M Entrained Gases (Gamma Emitters)E P M H-3 1x10

-5 d

Each Batch Composite Gross Alpha 1x10

-7

-8 P Q d Sr-89, Sr-90 5x10 Each Batch Composite

-6 Fe-55 1x10

-7 B. Continuous W Principa}Camma 5x10 f f l Releases Continuous Composite Emitters

-6

! I-131 1x10 Dissolved and -5 1x10 Entrained Gases (Gamma Emitters)E f

M f

H-3 1x10" l Continuous Composite

-7 l

Gross Alpha 1x10

\

i -8 f

Q f Sr-89, Sr-90 5x10 l Continuous Composite

-6 Fe-55 1x10 NORTH ANNA - UNIT 2 3/4 11-2

y .

..I TABLE 4.11-1 (Continued) 4 TABLE NOTATION D-The (LLD) is defined, for purposes of these Specifications, as;the smallest concentration of radioactive material in a sample that will yield a net count (above system background) that will be detected with 95% probability with only 5% probability of falsely concluding that a blank obser7ation represents a "real" signal.

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

4.66 s b LLD =

E.V .2.22 x 106.Y .

erp (- Aat) ; '

Where:

LLD is the "a priori" lower 11mic of detection as defined above (as microcuries per unit rass or volume),

i s is the standard deviation of the' background coun' ting rate or of tNecountingrateofablanksampleasappropriate(ascounts'per '~

minute), ,

s E is the counting efficiency (as counts per disintegration),

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

6 i 2.22 x 10 is the number of disintegrations per minute per j microcurie, l

! Y is the fractional radioch'emical yield (when applicable),

A is the radioactive decay constart for the particular radionuclide, and, At for plant' effluents is the elapsed time between the midpoint of m

sample collection and time of counting.

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

9 i

l NORTH ANNA - UNIT 2 3/4 11-3 l

l l - - -_- . --_ -. -

TART E 4.11-1 (Continued)

, ,_ TABLE NOTATION It should be recognized that the LLD is defined as an jt priori (before the fact) limit representing the capability of a measurement system and not as an 3L posteriori (after the fact) limit for a particular measurement.

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

Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mix as the situation permits, to assure representative sampling.

"The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks that are measurable and identifiable at levels exceeding the LLD, together with the above nuclides, shall also be identified and reported.

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

A continuous release is the discharge of liquid wastes of a nondiscrete n _

volume, e.g., from a volume of a system that has an input flow during the continuous release.

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

8 For certain mixtures of gamma emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the same sample in much greater concentrations.

Under these circumstances, it will be more appropriate to calculate the concentrations of such radionuclides using measured ratios with those radionuclides which are routinely identified and measured.

) >

2 NORTH ANNA - UNIT 2 3/41:-4

RADIOACTIVE EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose commitment to the maximum exposed MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each reactor unit, to UNRESTRICTED AREAS (see Figure 5.1-1) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrems to the total body and to less than or equal to 5 mrems to the critical organ,* and

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

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.2 Dose Calculations. Cumulative dose contributions from liquid effluents shall be determined in accordance with the ODCM at least once per 31 days.

4

The maximum exposed MEMBER OF THE PUBLIC and the critical organ will be addressed in the ODCM.

NORTH ANNA - UNIT 2 3/4 11-5

RADIOACTIVE EFFLUENTS 1,IQUID RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.1.3 The liquid radwaste ion exchanger system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent, from each reactor unit, to UNRESTRICTED AREAS (see Figure 5.1-1) would exceed 0.06 mrem to the total body or 0.2 mrem to the critical organ

in a 31 day period.

APPLICABILITY: At all times.

ACTION:

a. With radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report that includes the following information:

1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or sub-systems, and the reason for the inoperability,

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

3. Summary description of ACTION (s) taken to prevent a recurrence.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases shall be projected at least once per 31 days in accordance with the ODCM.

The critical organ will be addressed in the ODCM.

j NORTH ANNA - UNIT 2 3/4 11-6 i

RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each of the following unprotected outdoor tanks shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.

a. Refueling Water Storage Tank

b. Casing Cooling Storage Tank

c. PG Water Storage Tank *

d. Boron Recovery Test Tank *

e. Any Outside Temporary Tank **

APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 1

l l 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per month when radioactive materials are being added to the tank.

This is a shared system with Unit 1.

- Tanks included in this Specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste ion exchanger system.

i l NORTH ANNA - UNIT 2 3/4 11-7 l

[

RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.] The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) shall be limited to the following:

a. For noble gases: Less than or equal to 500 mrems/yr to the total body and less than or equal to 3000 mrems/yr to the skin, and

b. For iodine-131, for tritium, and for all radionuclides in particulate form with half lives greater than 8 days: Less than or equal to 1500 mrems/yr to the critical organ.*

APPLICABILITY: At all times.

ACTION:

With the dose rate (s) exceeding the above limits, without delay restore the release rate to within the above limit (s).

SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined continuously to be within the above limits in accordance with the methods and procedures of the ODCM.

4.11.2.1.2 The dose rate due to iodine-131, tritium, and all radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2.

The critical organ is defined in the ODCM.

NORTH ANNA - UNIT 2 3/4 11-8

TABLE 4.11-2 g

y ..

RADIOACTIVE CASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM

$ Lower Limit of Minimum

' Sampling Analysis Type of Detection (LLD)a E Gaseous Release Type Frequency Frequency Activity Analysis (uCi/ml)

U P P -4 A. Waste Gas Storage Each Tank Each Tank Principal Gamma Emitters 1x10

< w Tank" Grab Sample P P b -4 B. Containment PURGE Each PURGE Each PURGE Principal Gamma Emitters 1x10 Grab Sample -4 H-3 lx10 C. Process Vent M" M' Principal Gamma Emitters 1x10-

- Vent. Vent A Grab I Vent. Vent B Sample e

-6 H-3 1x10

~

Continuous W* I-131 lx10 D. All Release Types as listed in A, B, Charcoal 3

C above. Sample

-11 Continuous W" Principal Gamma Emitters 1x10 Particulate (I-131, Others)

Sample Continuous M Gross Alpha 1x10~

Composite Particulate Sample

-11 Continuous Q Sr-89, Sr-90 lx10 Composite Particulate Sample Continuous Noble Gas Noble Gases lx10" Monitor Gross Beta or Gamma

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

TABLE 4.11-2 (Continued)

'- z .

RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM

, Y l $>

I Minimum Lower Limit of i@ Sampling Analysis Type of Detection (LLD)a y Gaseous Release Type Frequency Frequency Activity Analysis (uCi/ml) w E. Condpnser Air Ejector W W Principle Gamma Emitters lx10-Vent Grab Sample -6 H-3 1x10 Steam Generatgr Blowdown Vent 4

+

S i

l r

)

i

TABLE 4.11-2 (Continued)

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

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

4.66 s b LLD =

EV 2.22 x 106 , y , ,xp (_Abt)

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 h

the counting race 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),

6 2.22 x 10 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.

NORTH ANNA - UNIT 2 3/4 11-11

TABLE 4.11-2 (Continued)

TABLE NOTATION It should be recognized that the LLD is defined as an jt priori (before the fact) limit representing the capability of a measurement system and not as an 3L posteriori (after the fact) limit for a particular measurement.

b The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, Xe-135m, 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 that are measurable and identifiable, at levels exceeding the LLD together with the above nuclides, shall also be identified and reported.

" Sampling and analysis shall also be performed following shutdown, startup, and whenever a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER occurs within a one hour period, if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant is greater than 1.0 uCi/gm; and (2) the noble gas activity monitor shows that effluent activity has increased by more than a factor of 3.

d The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2 and 3.11.2.3.

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

Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER in one hour and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for l 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLDs may be increased by a l factor of 10. This requirement applies if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant is greater than 1.0 uCi/gm and; (2) the nobic gas monitor shows that effluent activity has increased more than a factor of 3.

f For certain mixtures of gamma emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the same sample in much greater concentrations.

Under those circumstances, it will be more appropriate to calculate the concentrations of such radionuclides using measured ratios with those radionuclides which are routinely identified and measured.

NORTH ANNA - UNIT 2 3/4 11-12

l RADIOACTIVE EFFLUENTS l DOSE - NOBLE GASES l

i LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each reactor unit, from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) shall be limited to the following:

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

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

APPLICABILITY: At all times.

ACTION

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.2 Dose Calculations Cumulative dose contributions for noble gases for the current calendar quarter and current calendar year shall be determined in accordance with the ODCM at least once per 31 days.

9 NORTH ANNA - UNIT 2 3/4 11-13

RADIOACTIVE EFFLUENTS DOSE - 10 DINE-131. TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to the maximum exposed MEMBER OF THE PUBLIC from iodine-131, from tritium, and from all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, from the site to UNRESTRICTED AREAS (see Figure 5.1-1) shall be limited to the following:

i

a. During any calendar quarter: Less than or equal to 7.5 mrems to the critical organ

and,

b. During any calendar year: Less than or equal to 15 mrems to the

, critical organ *.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of iodine-131, tritium, and radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit and defines the j corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

b. The provisions of Spacifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.3 Dose Calculatior.3 Cumulative dose contributions for the current j calendar quarter and current calendar year for iodine-131, tritium and l ' radionuclides in particulate form with half lives greater than 8 days shall be determined in accordance with the ODCM at least once per 31 days.

The critical organ is addressed in the ODCM.

NORTH ANNA - UNIT 2 3/4 11-14

n - . . . . . - . .

i l

4.

. RADIOACTIVE EFFLUENTS

{ GASEOUS RADWASTE TREATMENT

, LIMITING CONDITION FOR OPERATION 4

3.11.2.4 The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST

, TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous j waste prior to their discharge when the projected gaseous effluent air doses t due to gaseous effluent releases, from each reactor unit, from the site to

! areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) would exceed 0.2 mrad for gamma radiation and 0.4. mrad for beta radiation over 31 days. The VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive i materials in gaseous waste prior to their discharge when the projected doses

' due to gaseous effluent releases, from each reactor unit, from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) would exceed 0.3 mrem to the critical organ

over 31 days.

APPLICABILITY: At all times.

ACTION:

a. With gaseous waste being discharged without treatment and in excess

of the above limits, in lieu of a Licensee Event Report, prepare

! and submit to the Commission within 30 days, pursuant to Specifica-

, tion 6.9.2, a Special Report that includes the following information:

1. Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or sub-systems, and the reason for the inoperability, j 2. ACTION (s) taken to restore the inoperable equipment to OPERABLE status, and f

[ 3. Summary description of ACTION (s) taken to prevent a recurrence.

j b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

j SURVEILLANCE REQUIREMENTS i *

, 4.11.2.4.1 Doses due to gaseous releases from the site shall be projected at j least once per 31 days in accordance with the ODCM.

i i .

1

The critical organ is addressed in the ODCM.

t NORTH ANNA - UNIT 2 3/4 11-15 i

RADI0 ACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE J

LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the waste gas decay tanks shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume or is less than 96% by volume.

APPLICABILITY: At all times.

ACTION

a. With the concentration of oxygen in the waste gas decay tanks greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ,

b. With the concentration of oxygen in the waste gas decay tanks greater than 4% volume and the hydrogen concentration greater than 2% by volume, immediately suspend all additions of waste gases to the system and reduce the concentration of oxygen to less than or equal to 2% by volume without delay.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS I

4.11.2.5 The concentrations of hydrogen and oxygen in the waste gas decay tanks shall be determined to be within the above limits by continuously monitoring the waste gases in the waste gas decay tanks with the hydrogen and oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.11.

e NORTH ANNA - UNIT 2 3/4 11-16 1

RADIOACTIVE EFFLUENTS CAS STORAGE TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to < 25,000 curies noble gases (considered as Xe-133).

APPLICABILITY: At all times.

ACTION:

. a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately 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 tank contents to within the limit.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per i 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank.

1 b

NORTH ANNA - UNIT 2 3/4 11-17 l

RADIOACTIVE EFFLUENTS 3/4.11.3 SOLID RADI0 ACTIVE WASTE LIMITING CONDITION FOR OPERATION 3.11.3 SOLIDIFICATION shall be conducted in accordance with a PROCESS CONTROL PROGRAM.

APPLICABILITY: At all times.

ACTION:

a. With the provisions of the PROCESS CONTROL PROGRAM not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.

b. The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1.9.b are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.3.1 The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive waste (e.g. , filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions).

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

b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the j i collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least 3 consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.13, to assure SOLIDIFICATION of subsequent batches of waste.

f NORTH ANNA - UNIT 2 3/4 11-18

RADIOACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.11.4 The annual (calendar year) dose or dose commitment to the maximum exposed MEMBER OF THE PUBLIC due to releases of radioactivity and radiation, from uranium fuel cycle sources shall be limited td:less than or equal to 25 mrems to the total body or the critical organ * (except the thyroid, which shall be limited to less than or equal to 75 mrems).

APPLICABILITY: At all times.

ACTION:

a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifica-tion 3.11.1.2.a. 3.11.1.2.b, 3.ll.2.2.a. 3.ll.2.2.b. 3.ll.2.3.a. or 3.11.2.3.b, calculations should be made to determine whether the above limits of Specification 3.11.4 have been exceeded. If such is the case in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that defines the corrective action to be taken to reduce sub-sequent releases to prevent recurrence of exceeding the above limits and includes the schtdule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR Part 20.405c, shall in.lude an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathwayn and direct radiation, for the calendar year that includes the release (s) covered by this report.

It shall also describe levels of radiation and concentrations of f radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose (s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request,

and a variance is granted until staf f action on the request is complete.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

l SURVEILLANCE REQUIREMENTS

\

l 4.11.4 Dose Calculations Cumulative dose contributions from liquid and gaseous i

effluents shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the ODCM.

( *The critical organ is addressed in the ODCM.

NORTH ANNA - UNIT 2 3/4 11-19 i

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 4.12-1.

APPLICABILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 4.12-1, in lieu of a Licensee Event Report, prepare and submit to the Commission, in the Annual Radio-logical Environmental Operating Report required by Specification 6.9.1.11, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 4.12-2 when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) i and defines the corrective actions to be taken to reduce radioactive i effluents so that the potential annual dose to A MEMBER OF THE PUBLIC l is less than the calendar year limits of Specifications 3.11.1.2, l 3.11.2.2, and 3.11.2.3. When more than one of the radionuclides in Table 4.12-2 are detected in the sampling medium, this report shall be submitted if:

! concentration (1) , concentration (2) . ***'> 1.0 reporting level (1) reporting level (2) l When radionuclides other than those in Table 4.12-2 are detected and l are the result of plant effluents, this report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is equal to or l

i greater than the calendar year limits of Specifications 3.11.1.2,

3.11.2.2 and 3.11.2.3. This report is not required if the measured j

l level of radioactivity was not the result of plant effluents; however, t in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

c. With milk or fresh leafy vegetable samples unavailable from one or

. more of the sample locations required by Table 4.12-1, identify locations for obtaining replacement sampler and add them to the radiological environmental monitoring program within 30 days. The l specific locations from which samples were unavailable may then be l

NORTH ANNA - UNIT 2 3/4 12-1

RADIOLOGICAL ENVIRONMENTAL MONITORING deleted from the monitoring program. In lieu of a Licensee Event Report and pursuant to Specification 6.9.1.12, identify the cause of the unavailability of samples and identify the new location (s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s).

d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.1.1 The radiological environmental monitoring samples shall be collected pursuant to Table 4.12-1 from the specific locations given in the table and figure (s) in the ODCM and shall be analyzed pursuant to the requirements of Table 4.12-1, the detection capabilities required by Table 4.12-3, and the guidance of the Radiological Assessment Branch Technical Position on Environmental Monitoring dated November, 1979, Revision No. 1.

1 I

l l

t l

l .

e NORTH ANNA - UNIT 2 3/4 12-2 5

I TABLE 4.12-1 z RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM

g Y -

> Number of Samples

$ and Sampling and Type and Frequency S Exposure Pathway a of Analysis I and/or Sample Sample Locations Collection Frequency c b 36 routine monitoring Quarterly Gamma dose

$ 1. DIRECT RADIATION quarterly.

H stations either with two or N more dosimeters or with one instrument for measuring and recording dose rate continuously to be placed as follows: 1) an .

)

inner ring of stations, one in each meteorological sector within the SITE BOUNDARY; an outer ring of u stations, one in each meteorological 57 sector within 8 km range from the site; the balance of the 3 stations to be placed in special 3 interest areas such as population centers, nearby residences, schools, and in 1 or 2 areas to serve as control stations.

2. AIRBORNE Samples from 5 locations: Continuous sampler Radioiodine Cannister:

Radiciodine and Particulates operation with sample I-131 analysis weekly.

a. 3 samples from close collection weekly or to the 3 SITE BOUNDARY more frequently if l required by dust Particulate Sampler:

I locations (in different Gross beta radioactivity sectors) of the highest loading.

calculated historical analysis folloging filter change; annual average ground-Gamma isotopic analysis level D/Q. of composite (by

The number, media, frequency, and location of samples may vary from site to site. This table presents an acceptable minimum program for a site at which each entry is applicable. Local site characteristics must be examined to determine if pathways not covered by this table may significantly contribute to an individual's dose and should be included in the sampling program.

x ,

TABLE 4.12-1 (Continued)

O y ,

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM IE z

> Number of Samples i Exposure Pathway and Sampling and Type and Frequency a Collection Frequency of Analysis Q and/or Sample Sample Locations U

N b. I sample from the vicinity location) quarterly.*

of a community having the highest calculated annual average groundlevel D/Q.

c. I sample from a control location 15-30 km

Ri distant and in the c- least prevalent wind direction.

[l.

3. WATERBORNE

a. Surface a. I sample circulating Sample off upstream, Gamma isotopic analysis" water discharge downstream and monthly. Conposite for cooling lagoon. tritium analysis quarterly.

Grab Monthly.

b. Czound Samples from 1 or 2 sources Grab Gamma isotopic" and tritium only if likely to be affected. Quarterly analysis quarterly.

! c. Sediment I sample from downstream area Semiannually Gamma isotopic analysis

with existing or potential semiannually.

recreational value.

z . TABLE 4.12-1 (Continued)

O

[ RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Number of Samples h and Sampling and Type and Frequency

, Exposure Pathway and/or Sample Sample Locations Collection Frequency of Analysis c.

3 e

to

4. INGESTION

a. Milk a. Samples from milking animals Monthly at all Gamma isotopic" and I-131 in 3 locations within 5 km times, analysis monthly .

distance having the highest dose potential. If there y

are none, then, I sample

from milking animals in each g

of 3 areas between 5 to 8 km y distant where doses are cal-v' culated to be greater than 1 mrem per yr 8.

b. I sample from milking animals at a control location (15-30 km

distant and in the least pre-I valent wind direction).

i l

b. Fish and a. I sample of commercially Annually, Gamma isotopic analysis Invertebrates and recreationally on edible portions.

important species (bass, sunfish, catfish) in vicinity of plant discharge area.

b. I sample of same species in areas not influenced by plant discharge.

2 '

TABLE 4.12-1 (Continued)

O W

sf RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM h Number of Samples

, Exposure Pathway and Sampling and Type end Frequency e and/or Sample Sample Locations ^ Collection Frequency of Analysis 3

H e y c. Food a. Samples of an ediable Monthly if available, Gamma isotopic and I-131 Products broad leaf vegetation or at harvest. analysis.

(cont'd) grown nearest each of two different offsite locations of highest predicted historical annual average ground-u, level D/Q if milk samp-3: ling is not performed, w

[ b. I sample of broad leaf vegetation grown 15-30 Monthly if available, or at harvest.

Gamma isotopic" and I-131 analysis.

km distant in the least prevalent wind direction if milk sampling is not performed.

1 z

O TABLE 4.12-1 (Continued)

E!

g t TABLE NOTATION N .

l g Iodine collection efficiency of the charcoal filters shall be determined. ~

, Y :

Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more af ter sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples

is greater than ten times the yearly mean of control samples, gamma isotopic analysis shall be performed -

! on the' individual samples.

l

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

A composite sample is one in which the quantity (aliquot) of liquid sampled is proportional to the quantity of flowing liquid and in which the method of sampling employed results in a specimen that is i

If represnetative of the liquid flow. In,this program composite sample aliquots shall be collected at time intervals that are very short (e.g., hourly) relative to the compositing period (e.g., monthly) j

{

OD in order to assure obtaining a representative sample.

8 The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM.

h If milk sampling cannot be performed, use item 4.c.

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1 E! TABLE 4.12-3 4

a DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS si .

$E LOWER LIMIT OF DETECTION (LLD) 1 Fish Milk Food Products Sediment E! Water Airborne Partieglate (pCi/kg, wet) (pCi/kg. dry)

[i Analysis (pCi/1) or Cas (pCi/m ) (pCi/kg, wet) (pCi/t) w gross beta 4 0.01 H-3 3000 Mn-54 15 130 f Fe-59 30 260 w Co-58,60 15 130 Y

- Zn-65 30 260 Y

5 Zr-Nb-95 15

60 I-131 10 0.07 1 i

Cs-134 15 0.05 130 15 60 150 l

Cs-137 18 0.06 150 18 80 180 Ba-La-140 15 15

The LLD for gamna isotopic analysis shall be used.

TABLE 4.12-3 (Continued)

TABLE NOTATION a

This list does not mean that only these nuclides are to be considered.

Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.11.

The LLD is defined, for purposes of these Specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95%

probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD =

E - V . 2.22 - Y . exp( Aft)

Where:

LLD is the "a priori" lower limit of detection as defined above, as picocuries per unit mass or volume, s is the standard deviation of the background counting rate or of .

h the counting rate of a blank sample as appropriate, as counts per minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield, when applicable.

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

At for environmental samples is the elapsed time between sample collection, or end of the sample collection period, and time of counting Typical valves of E, V, Y and at should be used in the calculation.

NORTH ANNA - UNIT 2 3/4 12-11

I 1

TABLE 4.12-2 (Continued) -j TABLE NOTATION 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. Analysis 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 pursuant to Specification 6.9.1.11.

I a

1 e

NORTH ANNA - UNIT 2 3/4 12-12

a RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 3.12.2 A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectorsofthenearestmilkgnimal,tgenearestresidenceandthenearest garden

of greater than 50 m (500 ft ) producing broad leaf vegetation.

APPLICABILITY: At all times.

ACTION:

a. With a land use census identifying a location (s) that yields a calcu-lated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, in lieu of a Licensee Event Report, identify the new location (s) in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.12.

b. With a land use census identifying a location (s) that yields a calculated dose or dose commitment (via the same exposure pathway) 25 percent greater than at a location from which samples are cur-rently being obtained in accordance with Specification 3.12.1, add the new location (s) to the radiological environmental monitoring '

program within 30 days. 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 of a Licensee Event Report and pursuant to Specification 6.9.1.12, identify the new location (s) in the next Semiannual Radioactive Effluent Release Report and also -

include in the report a revised figure (s) and table for the ODCM reflecting the new location (s).

c. The provisions of Specifications 3.0.3 and 3.0.4 are r at applicable.

SURVEILLANCE REQUIREMENTS 4.12.2 The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities. The results of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.11.

Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the site boundary in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census. Specifications for broad leaf vegetation sampling in Table 4.12-1.4c shall be followed, including analysis of control samples.

NORTH ANNA - UNIT 2 3/4 12-13

~

RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on radioactive materials (which contain nuclides produced at nuclear power stations) supplied as part of an s

Interlaboratory Comparison Program that has been approved by the Commission.

APPLICABILITY: At all times.

1 ACTION:

i

", a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrer.cc to the Commission in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.11.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.3 A summary of the results obtained as part of the above required Interlaboratory Comparison Program and in accordance with the methodology and parameters in the ODCM shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.11.

i l

l l

i l

i

(

i NORTH ANNA - UNIT 2 3/4 12-14

I l

l INSTRUMENTATION BASES 3/4.3.3.9 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION ;

1 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 offluents. The alarm /

trip setpoints for these instruments shall be calculated and adjusted in l accordance with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to'10 CFR Part 50. The purpose of tank level indicating devices is to assure the detection and control of leaks that if not controlled could potentially result in the transport of radioactive materials to UNRESTRICTED AREAS.

3/4.3.3.10 RADIOACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION 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 shall be calculated and adjusted in accordance with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the waste gas holdup system. The j 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

e NORTH ANNA - UNIT 2 B 3/4 3-4 I

- -- ---- , -= ,-w-- -

3/4.11 RADIOACTIVE EFFLUENTS BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents 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 in UNRESTRICTED AREAS will result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to a MEMBER OF THE PUBLIC and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for dissolved or entraine.I noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

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

Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry" Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

3/4.11.1.2 DOSE This specification is provided to implement the requirements of l

Sections II.A, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limitins

! Condition for Operation implements the guides set forth in Section II. A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will l be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies that 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 l of the requirements of 40 CFR Part 141. The dose calculations in the ODCM

( ' implement the requirements in Section III.A of Appendix I that conformance with i

the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with I the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses i to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating l Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and l Regulatory Guide 1.113 " Estimating Aquatic Dispersion of Effluents from l Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

NORTH ANNA - UNIT 2 B 3/4 11-1

RADIOACTIVE EFFLUENTS BASES This Specification applies to the release of liquid effluents from each reactor at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned among the units sharing that system.

3/4.11.1.3 LIQUID RADWASTE TREATMENT The requirement that the appropriate portions of this system be used, when specified, provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This Specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.

3/4.11.1.4 LIQUID HOLDUP TANKS The tanks listed in this Specification include all those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.

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

3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE

This Specification is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20. 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, either within or outside the SITE BOUNDARY to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For MEMBERS OF THE PUBLIC, who may at times be within the SITE BOUNDARY the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. The specified NORTH ANNA - UNIT 2 B 3/4 11-2

RADIOACTIVE EFFLUENTS BASES release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the SITE BOUNDARY to less than or equal to 500 arems/ year to the total body or to less than or equal to 3000 arems/ year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 arems/ year.

This Specification applies to the release of gaseous effluents from all reactors at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.

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

Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual. HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry" Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

3/4.11.2.2 DOSE - NOBLE CASES This Specification is provided to implement the requirements of Sections II.B. III. A and IV. A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable. The Surveillance Requirements implement the requirements in Section III.A o. Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appro-priate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1. July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

NORTH ANNA - UNIT 2 B 3/4 11-3

RADI0 ACTIVE EFFLUENTS BASES 3/4.11.2.3 DOSE - IODINE-131, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM This Secification is provided to implement the requirements of Sections II.C, III.A and IV. A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I.

The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous 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 calcula-tional procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substan-tially underestimated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111 " Methods for Estimating Atmospheric Transport and Dispersion of Caseous 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 iodine-131, radioiodines, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent on the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

l 3/4.11.2.4 GASEOUS RADWASTE TREATMENT The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive paterials 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 the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous effluents.

l NORTH ANNA - UNIT 2 B 3/4 11-4

RADIOACTIVE EFFLUENTS BASES 3/4.11.2.5 EXPLOSIVE GAS MIXTURE This Specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the vaste gas holdup system is maintained below the flammability limits of hydrogen and oxygen. (Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen, automatic diversion to recombiners, or injection of dilutants to reduce the concentration below the flammability limits.) Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3/4.11.2.6 GAS STORACE TANKS The tanks included in this Specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification to a quantity that is less than the quantity which provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem in an event of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Branch Technical Pocition ETSB 11-5 in NUREG-0800, July 1981.

3/4.11.3 SOLID RADIOACTIVE WASTE This Specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste / liquid / SOLIDIFICATION agent / catalyst ratios, vaste oil content, waste principal chemical ponstituents, mixing and curing times.

3/4.11.4 TOTAL DOSE This Specification is provided to meet the dose limitations of 40 CFR Part 190 that have now been incorporated into 10 CFR Part 20 by 46 FR 18525.

The' Specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice NORTH ANNA - UNIT 2 B 3/4 11-5

RADI0 ACTIVE EFFLUENTS BASES the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of ACTION that should result in the limitation of the unnual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR Part 190.11 and 10 CFR Part 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 3.11.1 and 3.11.2. An individual is not con-sidered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nucleat fuel cycle.

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NORTH ANNA - UNIT 2 B 3/4 11-6 l

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES ,

3/4.12.1 MONITORING PROGRAM The radiological environmental monitoring program required by this Speci-fication provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the station operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. The initially specified monitoring program will be effective for at least the first three years of commercial operation.

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

The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table 4.12-3 are considered optimum 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 a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

i Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry" Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

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NORTH ANNA - UNIT 2 B 3/4 12-1

l 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.2 LAND USE CENSUS This Specification is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the radiological environmental monitoring. program are made if required by the results of this census. The best infornation from the door-to-door survey, from aerial survey or from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix to 10 CFR Part 50. Restricting the census to gardens of greater than 50 m provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: 1) 20% of the garden was used for growing broad Icaf vegetatign (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/m .

3/4.12/3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monito.ing in order to demonstrate that the results are reasonably valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.

o NORTH ANNA - UNIT 2 B 3/4 12-2

5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1-1.

LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1-1.

MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluents, which allows identification of structures and release points as well as definition of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall be as shown in Figure 5.1-1.

5.2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape with a dome roof and having the following design features:

a. Nominal inside diameter = 126 feet.

b. Nominal inside height = 190 feet, 7 inches.

c. Minimum thickness of concrete walls = 4.5 feet,

d. Minimum thickness of concrete roof = 2.5 feet.

e. Minimum thickness of concrete floor pad = 10 feet.

f. Nominal thickness of the cylindrial portion of the steel liner = 3/8 inches.

0

g. Net free volume = 1.825 x 10 cubic feet.

h. Nominal thickness of hemispherical dome portion of the

, steel liner = 1/2 inch.

DESIGN PRESSURE AND TEMPERATURE 5.2.2. The reactor containment building is designed and shall be maintained for a maximum internal pressure of 45 psig and a temperature of 280*F.

l NORTH ANNA - UNIT 2 5-1 I

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ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifi-cations of ANSI N18.1 - 1971 for comparable positions and the supplemental requirements specified in the March 28, 1980 NRC letter to all licensees, except for (1) the Supervisor - Health Physics who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975 and (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents.

6.4 TRAINING 6.4.1 The Station Manager is responsible for ensuring that retraining and replacement training programs for the facility staff are maintained and that such programs meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1 -

1971 and Appendix "A" of 10 CFR Part 55 and the supplemental requirements specified in the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience identified by the SES.

6.5 REVIEW AND AUDIT 6.5.1 STATION NUCLEAR SAFETY AND OPERATING COMMITTEE (SNSOC)

FUNCTION 6.5.1.1 The SNSOC shall function to advise the Station Manager on all matters related to nuclear safety.

COMPOSITION 6.5.1.2 The SNSOC shall be composed of the:

Chairman: Station Manager Vice Chairman: Assistant Station Manager Member: Superintendent-Operations Member: Superintendent-Maintenance i Hember: Superintendent-Technical Services Member: Supervisor-Health Physics ALTERNATES f 6.5.1.3 All alternate members shall be appointed in writing by the SNSOC l Chairman to serve on a temporary basis; however, no more than one alternate l shall participate as a voting member in SNSOC activities at any one time.

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NORTH ANNA - UNIT 2 6-6

ADMINISTRATIVE CONTROLS MEETING FREQUENCY 6.5.1.4 The SNSOC shall meet at least once per calendar mon'th and as convened by the SNSOC Chairman or his designated alternate.

QUORUM 6.5.1.5 A quorum of the SNSOC consists of the Chairman or Vice-Chairman and two members including alternates.

RESPONSIBILITIES 6.5.1.6 The SNSOC shall be responsible for:

a. Review of 1) all procedures required by Specification 6.8.1 and changes thereto, 2) all programs required by Specification 6.8.4 and changes thereto, 3) any other proposed procedures or changes thereto as determined by the Station Manager to affect nuclear safety.

b. Review of all proposed test" ?nd experiments that affect nuclear safety,

c. Review of all proposed changes to Appendix "A" Technical Specifications.

d. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.

e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Manager-Nuclear Operations and Maintenance and the Director-Safety Evaluation and Control,

f. Review of events requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months written notification to the Commission.

( g. Review of f a c' '.i ty operations to detect potential nuclear safety hazards.

h. Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairman of the Station Nuclear Safety and Operating Committee.

i. Review of the Plant Security Plan and implementing procedures and

! shal.1 submit recommended changes to the Chairman of the Station l Nuclear Safety and Operating Committee.

j. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Chairman of the Station Nuclear Safety and Operating Committee.

NORTH ANNA - UNIT '2 6-7

l l ADMINISTRATIVE CONTROLS

k. Review of every unplanned onsite release of radioactive material to the environs including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President-Nuclear Operations and to the Director-Safety Evaluation and Control.

1. Review and changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL.

AUTHORITY 6.5.1.7 The SNSOC shall:

a. Recommend to the Station Manager written approval or disapproval of items considered under 6.5.1.6(a) through (d) above,

b. Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.

c. Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to Manager-Nuclear Operations and Maintenance and the Director-Safety Evaluation and Control of disagreement between the SNSOC and the Station Manager; however, the Station Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.

RECORDS 6.5.1.8 The SNSOC shall maintain written minutes of each meeting and copies shall be provided to the Manager-Nuclear Operations and Maintenance and the Director-Safety Evaluation and Control.

6,5.2 SAFETY EVALUATION AND CONTROL (SEC)

FUNCTION 6.5.2.1 SEC shall function to provide independent review of designated activities in the areas of:

a. Nuclear power plant operations 3 b. Nuclear engineering

c. Chemistry and radiochemistry

d. Metallurgy

,e . Instrumentation and control

f. Radiological safety NORTH ANNA - UNIT 2 6-8 i

ADMINISTRATIVE CONTROLS

g. Mechanical and electrical engineering

h. Administrative controls and quality assurance practices

1. Other appropriate fields associated with the unique characteristics of the nuclear power plant COMPOSITION 6.5.2.2 The SEC staff shall be composed of the Director-Safety Evaluation and Control and a minimum of three individuals who are qualified as staff specialists. Each SEC staff specialist shall have an academic degree in an engineering or physical science field and, in addition, shall have a minimum of five years technical experience in one or more areas given in Specification 6.5.2.1. These staff specialists shall not be directly involved in the licensing function.

CONSULTANTS 6.5.2.4 Consultants shall be 9tilized as determined by the Director-Safety Evaluation and Control to provide expert advice to the SEC.

MEETING FREQUENCY 6.5.2.5 The SEC staff shall meet at least once per calendar month for the purpose of fostering interaction of reviews regarding safety-related opera-tional activities.

! REVIEW 6.5.2.7 The following subjects shall be reviewed by SEC:

I

a. Written safety evaluations of changes in the stations as described in the Safety Analysis Report, changes in procedures as described in the Safety Analysis Report and tests or experiments not described in the Safety Analysis Report which are completed without prior NRC approval under the provisions of 10 CFR 50.59(a)(1). This review is to verify that such changes, tests or experiments did not involve a change in the Technical Specifications or an unreviewed safety a

question as defined in 10 CFR 50.59(a)(2) and is accomplished by review of minutes of the Station Nuclear Safety and Operating Committee and the design change program,

b. Proposed changes in procedures, proposed changes in the station, or proposed tests or experiments, any of which may involve a change in

, the Technical Specifications or ra unreviewed safety question as defined in 10 CFR 50.59(a)(2).. Matters of this kind shall be l

referred to the Directer-Safety Evaluation and Control by the Station Nuclear Safety and Operating Committee following its review prior to implementation.

NORTH ANNA - UNIT 2 6-9

1 ADMINISTRATIVE CONTROLS i

REVIEW (Cont'd)

c. Changes in the Technical Specifications or license amendments relating to nuclear safety prior to implementation except in those cases where the change is identical to a previously reviewed proposed change.

d. Violations and reportable occurrences such as:

1. Violations of applicable codes, regulatione, orders, Technical Specifications, license requirements or internal j procedures or instructions having safety significance; i

2. Significant operating abnormalities or deviations from normal or expected performance of station safety-related structures, systems, or components; and l

3. Reportable occurrences as defined in the station Technical Specification 6.9.1.8.

Review of events covered under this paragraph shall include the I

results of any investigations made and recommendations resulting from such investigations to prevent or reduce the probability of recurrence of the event.

e. The Quality. Assurance Department audit program at least once per 12 months and audit reports.

f. Any other matter involving safe operation of the nuclear power stations which a duly appointed subcommittee or committee member deems appropriate for consideration, or which is referred to the Director-Safety Evaluation and Control by the Station Nuclear Safety and Operating Committee.

g. Reports and meeting minutes of the Station Nuclear Safety and Operating Committee.

AUTHORITY 6.5.2.9 The Director-Safety Evaluation and Coatrol shall report to and advise the Manager-Nuclear Technical Services, who shall advise the Vice President-

)Ruclear Operations on those areas of responsibility specified in Section 6.5.2.7.

RECORDS 6.5.2.10 Records of SEC activities required by Section 6.5.2.7 shall be prepared and maintained in the SEC files and a summary shall be disseminated as indicated below each calendar month.

j 1. Vice President-Nuclear Operations

2. Nuclear Power Station Managers

3. Manager-Nuclear Operations and Maintenance NORTH' ANNA - UNIT 2 6-10 1

ADMINISTRATIVE CONTROLS

4. Manager-Nuclear Technical Services

5. Executive Manager-Quality Assurance

6. Others that the Director-Safety Evaluation and Control may designate.

6.5.3 QUALITY ASSURANCE DEPARTMENT FUNCTION 6.5.3.1 The Quality Assurance Department shall function to audit station activities. These audits shall encompass:

a. The conformance of facility operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months.

b. The performance, training and qualifications of the entire facility staff at least once per 12 months.

c. The results of actions taken to correct deficiencies occurring in facility equipment, st ructures , systems or method of operation that affect nuclear safety at least once per 6 months.

d. The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per 24 months,

e. The Station Emergency Plan and implementing procedures at least once per 24 months.

f. The Station Security Plan and implementing procedures at least once ,

per 24 months.

g. Any other area of facility operation considered appropriate by the Executive Manager-Quality Assurance or the Senior Vice President-Power Operations.

h. The Station Fire Protection Program and implementing procedures at a least once per 24 months.

i. An independent fire protection and loss prevention program inspection and audit shall be performed at least once per 12 months utilizing either qualified offsite licensee personnel or an outside fire protection firm.

'j . An inspection and audit of the fire protection and loss prevention program shall be performed by a qualified outside fire consultant at least once per 36 months.

Pending NRC approval.

NORTH ANNA - UNIT 2 6-11

ADMINISTRATIVE CONTROLS

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

1. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months.

m. The PROCESS CONTROL PROGRAM and implementing procedures for pro-cessing and packaging of radioactive wastes at least once per 24 months.

n. The performance of activities required by the Quality Assurance Program to meet the provisions of Regulatory Guide 1.21 Revision 1, June 1974 and Regulatory Guide 4.1, Revision 1, April 1975 at least once per 12 months.

AUTHORITY 6.5.3.2 The Quality Assurance Department shall report to and advise the Executive Manager-Quality Assurance, who shall advise the Senior Vice President-Power Operations on those areas of responsibility specified in Section 6.5.3.1.

RECORDS 6.5.3.3 Records of the Quality Assurance Department audits shall be prepared and maintained in the department files. Audit reports shall be disseminated as indicated below:

1. Nuclear Power Station Manager

2. Manager-Nuclear Operations and Maintenance

3. Manager-Nuclear Technical Services

4. Director-Safety Evaluation and Control *

5. Supervisor of area audited

6. Nuclear Power Station Manager Quality Assurance a

Pending NRC approval, ,

i NORTH ANNA - UNIT 2 6-12

ADMINISTRATIVE CONTROLS 6.6 REPORTABLE OCCURRENCE ACTION 6.6.1 The following actions shall be taken for REPORTABLE OCCURRENCES:

a. The Consnission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.

b. Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification to the Commission shall be reviewed by the SNSOC and submitted to the Director-Safety Evaluation and Control and the Manager-Nuclear Operations and Maintenance.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The facility shall be placed in at least HOT STANDBY within one hour,

b. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Manager-Nuclear Operations and Maintenance, and the Director-Safety Evaluation and Control shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the SNSOC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Commission, the Director-Safety Evaluation and Control and the Manager-Nuclear Operations and Maintenance within 14 days of the violation.

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6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

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

b. Refueling operations.

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NORTH ANNA - UNIT 2 6-13 i

ADMINISTRATIVE CONTROLS

c. Surveillance and test activities of safety related equipment.

l d. Security Plan implementation.

e. Emergency Plan implementation.

f. Fire Protection Program Implementation.

g. PROCESS CONTROL PROGRAM implementation.

h. OFFSITE DOSE CALCULATION MANUAL implementation.

i. Quality Assurance Program for effluent and environmental monitoring, using the guidance in Regulatory Guide 1.21, Revision 1. June 1974 and Regulatory Guide 4.1 Revision 1. April 1975.

6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be reviewed by the SNSOC and approved by the Station Manager prior to implementation and reviewed periodically as set forth in administrative procedures.

6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant supervisory staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

c. The change is documented, reviewed by the SNSOC and approved by the Station Manager within 14 days of implementation.

6.8.4 The following programs shall be established, implemented, and maintained:

a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the recirculation spray, safety injection, chemical and volume control, gas stripper, and hydrogen recombiners. The program shall include the following:

(i) Preventive maintenance and periodic visual inspection requirements and (ii) Integrated leak test requirements for each system at refueling cycle intervals or less.

NORTH ANNA - UNIT 2 6-14

8-31-80

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ADMINISTRATIVE CONTROLS

b. In-Plant Radiation Monitoring A program which will ansure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

(1) Training of personnel, (ii) Procedures for monitoring, and (iii) Provisions for maintenance of sampling and analysis equipment.

c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:

(i) Identification of a sampling schedule for the critical variables and control points for these variables, (ii) Identification of the procedures used to measure the values of the critical variables,

( .

(iii) Identification of process sampling points, (iv) Procedures for the recording and management of data, (v) Procedures defining corrective actions for all control point

- chemistry conditions,

(vi) A procedure identifying (a) the authority responsible for the 8 interpretation of the data, and (b) the sequence and timing of j administrative events required to initiate corrective action, and (vii) Monitoring of the condensate at the discharge of the condensate pumps for evidence of condenser inleakage. When condenser in-leakage is confirmed, the leak shall be repaired, plugged, or isolated within 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months .

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' " NORTH ANNA - UNIT 2 6-14e

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8-21-80

' ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS AND REPORTA8LE OCCURRENCES i

3 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director

of the Regional Office of Inspection and Enforcement unless otherwise noted.

i STARTUP REPORTS

6. 9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (a) receipt of an operating license, (2) amendment to l

the license involving a planned increase in power level (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, j

thermal, or hydraulic performance of the plant.

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' NORTH ANNA - UNIT 2 6-14b

8-21-80 ADMINISTRATIVE CONTROLS __

6.9.1.2 The startup report shall andress each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained daring the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details requested in license conditions based on other commitments shall be included in this report.

6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of :ommercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of comunercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

ANNUAL REPORTS 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each l year. The initial report shall be submitted prior to March 1 of the year

[ following initial criticality.

l 6.9.1.5 Reports required on an annual basis shall irclude:

a. A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 ares /yr and their associated man-rem exposure according to work and job functions,8f e.g. , reactor operations and

! surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.

The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20 percent of the individual total

, , dose need not be accounted for. In the aggregate, at least 80 l

percent of the total whole body dose received from external sources l should be assigned to specific major work functions.

'f 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.

8/ This tabulation supplements the requirements of $20.407 of 10 CFR Part 20.

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i 8-21-80 ADMINISTRATIVE CONTROLS

b. The complete results of the steam generator tube inservice inspections performed during the report period (Reference Specification 4.4.5.5.b.).

l MONTHLY OPERATING REPORT .

6.9.1.6 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the Of rector, Office of Management and Program Analysis, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Office, of Inspection and Enforcement, no later than the 15th of each month following the calendar month covered by the report.

REPORTA8LE OCCURRENCES

6. 9.1. 7 The REPORTABLE OCCURRENCES of Specifications 6.9.1.8 and 6.9.1.9 below, including corrective actions and measures to prevent recurrence, shall be reported to the NRC. Suppler. ental 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. }

t PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP l

6.9.1.8 The types of events listed below shall be reported within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telepnone and confirmed by telegraph, mailgram, or facsimila transmission to the Director of the Regional Office, or his designate, no later than the first working day following the event, with a written followup report within 14 days.

The written followup report shall include, as a minimum, a completed copy of 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.

a. Failure of the reactor protection system or other systems subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reaches the setpoint

, specified as the limiting safety-system setting in the technical specifications or failure to complete the required protective function.

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NORTH ANNA - UNIT 2 6-15

ADMINISTRATIVE CONTROLS

b. Operation of the unit or affected systems when any parameter or operation subject to a limiting condition for operation is less conservative than the least conservative aspect of the limiting condition for operation established in the Technical Specifications.

c. Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment.

d. Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady-state conditions during power l operation greater than or equal to 1% delta k/k; a calculated i reactivity balance indicating a shutdown margin less conservative than l specified in the Technical Specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if subcritical, an unplanned reactivity insertion of more than 0.5% delta k/k; or occurrence of any unplanned criticality.

1

e. Failure or malfunction of one or more components which prevents could prevent, by itself, the fulfillment of the functional requirements of system (s) used to copy with accidents analyzed i che SAR.

f. Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the SAR.

g.

Conditions arising from natural or man-made events that, as a direct result of the event, require plant shutdown, operation of safety systems, or other protective measu:es required by Technical Specifications.

l h. Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the Technical Specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.

1. Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operatior in a manner less conservative than that assumed in the accident analyses

, in the safety analysis report or Technical Specificatione 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.

j. Offsite releases of radioactive waterials in liquid and gaseous effluents that exceed the limits of Specification 3.11.1.1 or 3.11.2.1.

l NORTH ANNA - UNIT 2 6-17 i

ADMINISTRATIVE CONTROLS

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k. Exceeding the limits in Specification 3.11.1.4 or 3.11.2.6 for the l storage of radioactive materials in the listed tanks. The written follow-up report shall include a schedule and a description of activities planned and/or taken to reduce the contents to within the specified limits.

THIRTY-DAY WRITTEN REPORT 6.9.1.9 The types of events listed below shall be the subject of written I reports to the Director t,' the Regional Office within 30 days of occurrence of the event. The written eport aball include, as a minimum, a completed l copy of the 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.

4 a. Reactor protection system or engineered safety feature instru-i ment settings which are found to be less conservative than

t. hose established by the Technical Specifications but which do not prevent the fulfillment of the functional requirements of affected systems, i j

^

b. 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.

l c. 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 i d. Abnormal degradation of systems other than those specified in item 6.9.1.8(c) above designed to contain radioactive j material resulting from the fission process.

e. An unplanned offsite release of 1) more than 1 curie of radioactive

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

1. A description of the event and equipment involved.

2. Cause(s) for the unplanned release.

3. Actions taker. to prevent recurrence.

4. Consequences of the unplanned release.

i NO:-fH ANNA - UNIT 2 6-18

ADMINISTRATIVE CONTROLS (Continued) _ _ ,

CORE SURVEILLANCE REPORT 6.9.1.10 The F y limit for Rated Thermal iower..(F ) in all core planes containing Bank "D" contro'. rods and in all unrodded core planes, the surveillance power level, P , for Technical Specifications 3.2.1 and 3.2.6, and the FqfTyspeck basis shall be provided to the Director of the Regional Office of Inspection and Enforcement, with a copy to; Director Office of Nuclear Reactor Regulation Attention: Chief of Core Performance Branch U. S. Nuclear Regulatory Commission Washington, D. C. 20555 at least 60 days prior to cycle initial criticality. In the event that the limits would be submitted at some other time during core life, they shall be submitted 60 days prior to the date the limits would become effective unless otherwise exempted by the Commission.

Any additional information needed'to support the F AND P submittal will be by request from the NRC and need not bE included in this report.

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT

6.9.1.11 Routine Radiological Environmental Operating Reports covering the operation of the unit during the previc'us calendar year shall be submitted prior to May 1 of each year. The initial report shall be subm. tted prior to May 1 oftheyearfollowinginitialcriticaljty, The Annual Radiological Environmental [ Operating Reports shall include summaries.

l interpretations, and an analysis of trends'e( the_results of the radiological environmental surveillance activities for the report period, including a comparison (as appropriate) with preoperational studies, operatioral controls g

, and previous environmental surveillan'ce~ ceports, and an assessment of the observed impacts of the plant operation ,on /the environment. "Thejreforts shall ,

piso include the results of land use censuses require 6[ by

Sp ecification - 7 3.12.2. / ". i ..

The Annual Radiological Environmental Operating Reports shall include the results [,

of analysis of all radiological environmental samples and of all environmental .,5 radiation measurements taken during the period pursuant to ths locations specified in the Table and Figures in the ODCM, as well as summarfred and i tabulated results of these analyses and measurements ~ in the format of the teble in the Radiological Assessn.ent Branch Technical Position, s

A single submittal may be made for a multiple unit station.

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NORTH ANNA - UNIT 2 6-19 "

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ADMINISTRATIVE CONTROLS (CONTINUED)

Revision 1, November 1979. In the event that some individual 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.

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The reports shall also include the following: a summary description of the radiological environmental monitoring program; at least two legible maps

covering all sampling locations keyed to a table. giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program, required by Specification 3.12.3; discussion of all deviations from the sampling schedule

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of Table 4.12-1 and discussion of all analyses in which the LLD required by Table 4.12-3 was not achievable.

SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT **

6.9.1.12 Routine Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality.

The Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, " Measuring,

,, Evaluating .and Reporting Radioactivity in Solid Wastes and Releases of Radio-active Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

The Radioactive Effluent Release Report to be submitted within 60 days after January 1 of. each year. This report shall include an assessment of the radiation doses to the maximum exposed MEMBERS OF THE PUBLIC due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. Annual meteorological data collected over the previous year shall be in the form of joint frequency distributions of wind speed, wind _ direction, and atmospheric stability. This meteorological l

data shall be retained in a file on site and shall be made available to the NRC upon request. All assumptions used in making these assessments (i.e.,

specific activity, exposure time and location) shall be included in the OFFSITE DOSE CALCULATION MANUAL (ODCM). Concurrent meteorological conditions or historical annual average atmospheric dispersion conditions shall be used

,for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the OFFSITE DOSE CALCULATION MANUAL r . (0DCM).

0ne map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.

- A single submittal may be made for a multiple unit station. The submittal l} ' ,

0-should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

NORTH ANNA - UNIT 2 6-20 L-

ADMINISTRATIVE CONTROLS If the dose to the maximum exposed MEMBER OF THE PUBLIC due to the radioactive liquid and gaseous effluents from the station during the previous calendar year exceeds twice the limits of Specification 3.11.1.2a, 3.11.1.2.b, 3.11.2.2.a. 3.11.2.2.b, 3.11.2.3.a. or 3.11.2.3.b the dose assessment shall include the contribution from direct radiation. The dose to the maximum exposed MEMBER OF THE PUBLIC shall show conformance with 40 CFR Part 190,

, Environmental Radiation Protection Standards for Nuclear Power Operation.

The Radioactive Effluent Release Reports shall include a list of unplanned releases as required to be reported in Technical Specification 6.9.1.9.e from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (GLCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.12.2.

SPECIAL REPORTS Special reports may be required covering inspections, test and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.

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ADMINISTRATIVE CONTROLS 1

6.9.2 Special reports shall be submitted to the Director of the NRC Regional '

Office listed in Appendix D, 10 CFR Part 20, with a copy to the Director,

, Office of Inspection and Enforcement, U. S. Nuclear Regulatory Commission,

, Washington, D. c. 20555 within the time period specified for each report.

6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.1 The following records shall be retained for at least five years:

a. Records and logs of facility operation covering time interval at each power level.

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety,

c. Each REPORTABLE OCCURRENCE submitted to the Commission.

d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

e. Records of changes made to Operating Procedures.

f. Records of radioactive shipments.

g. Records of sealed source leak tests and results.

h. Records of annual physical inventory of all sealed source material of record.

6.10.2 The following records shall be retained for the duration of the Facility Operating License:

a. Records and drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis Report.

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

c. Records of facility radiation and contamination surveys.

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

NORTH ANNA - UNIT 2 6-22

ADMINISTRATIVE CONTROLS

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

f. Records of transient or operational cycles for those facility components identified in Table 5.7-1.

g. Records of reactor tests and experiments.

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

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

j. Records of Quality Assurance activities required by the QA Manual.

k. Records of the service life of all hydraulic and mechanical snubbers listed on Tables 3.7-4a and 3.7-4b including the date at which the service life commences and associated installation and maintenance records.

1. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

m. Records of meetings of the SNSOC.

n. Records of analyses required by the radiological environmental monitoring program that would permit evaluation of the accuracy of the analysis at a later date. This would include procedures effective at specified times and QA records showing that these procedures were followed.

o. Records of secondary water sampling and water quality. ,

p. Records for Environmental Qualification which are covered under the provisions of Paragraph 2.C(4)(e) of License No. NPF-7.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent yith the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr

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shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work NORTH ANNA - UNIT 2 6-23 k

ADMINISTRATIVE CONTROLS Permit.* Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dnse is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them,

c. An individual qualified in the protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiation Work Permit.

6.12.2 The requirements of 6.12.1, above, shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or the Plant Health Physicist.

6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 Licensee initiated changes to the PCP:

1. Shall be submitted to the Commission in the Semiannual 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 rationale for the change without benefit of additional or supplemental information;

b. A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and

c. Documentation of the fact that the change has been reviewed and found acceptable by the SNSOC.

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

Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of I

their assigned radiation protection duties, provided they comply with approved radiation protection procedures for entry in high radiation areas.

NORTH ANNA - UNIT 2 6-24

ADMINISTRATIVE CONTROLS 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation.

6.14.2 Licensee initiated changes to the ODCM:

1. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain:

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a packege of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change (s);

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

c. Documentation of the fact that the change has been reviewed and found acceptable by the SNSOC.

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

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TABLE OF CONTENTS PAGE 1.0 DEFINITIONS .................................................. 1-1 2.0 LIMITING CONDITIONS FOR OPERATION ............................ 2-1 2.1 Non-Radiological ........................................ 2-1 3.0 ENVIRONMENTAL SURVEILLANCE ................................... 3-1 3.1 Non-Radiological Surveillance ........................... 3-1 3.1.1 Abiotic - Aquatic ................................ 3-1 3.1.2 Biotic Aquatic ................................... 3-1 3.1.3 Ab io tic - Te rre s t ria l . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.1.4 Onsite Meteorology Monitoring .................... 3-1 4.0 SPECIAL SURVEILLANCE AND STUDY ACTIVITIES .................... 4-1 5.0 ADMINISTRATIVE CONTROLS ...................................... 5-1 5.1 Responsibility .......................................... 5-1 5.2 Organization ............................................ 5-1 5.3 Review and Audit ........................................ 5-1 5.3.1 Station Nuclear Safety and Operating Committee (SNSOC) .......................................... 5-1 5.3.1.1 Function ................................ 5-1 5.3.1.2 Responsibility .......................... 5-1 5.3.1.3 Authority ............................... 5-3 5.3.1.4 Records ................................. 5-3 5.3.2 Quality Assurance Department ..................... 5-4 5.3.2.1 Function ................................. 5-4 5.3.2.2 Audits ................................... 5-4 5.3.2.3 Records .................................. 5-4 5.3.3 Safety Evaluation and Control (SEC) ............... 5-4 5.3.3.1 Function ................................. 5-4 5.3.3.2 Review ................................... 5-4 5.3.3.3 Responsibility ........................... 5-5 5.3.3.4 Authority ................................ 5-5 5.3.3.5 Records .................................. 5-5 5.4 State and Federal Permits and certificates .................... 5-5 I-i

TABLE OF CONTENTS (Cent'd)

PAGE 5.5 Procedures .................................................... 5-5 5.5.1 Written Procedures ................................ 5-5 5.5.2 Operating Procedures .............................. 5-5 5.5.3 Procedures for Environmental Surveillance -

Nonradiological ................................... 5-6 5.5.4 Quality Assurance of Program Results .............. 5-6 5.5.5 Changes in Procedures, Station Design or Operation ......................................... 5-6 5.5.6 Consistency with Initially Approved Programs ...... 5-7 5.6 Station Reporting Requirements ................................ 5-7 5.6.1 Routine Reports ................................... 5-7 5.6.2 Nonroutine Reports ................................ 5-8

. 5.6.2.1 Nonroutine Non-Radiological Environmental Operating Report .........................5-8 5.6.3 Changes in Environmental Technical Specifications.. 5-9 5.6.4 Changes in Pe rmits and Certifications . . . . . . . . . . . . . 5- 9 i

5.7 Re c o rd s Re t e n t ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 9 I

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2.0 LIMITING CONDITIONS FOR OPERATION 2.1 NON-RADIOLOGICAL - None a

I-2-1

observations in a form consistent with National Weather Service procedures. Summaries of all data and observations shall be available to the NRC upon request.

Any modification to the onsite meteorological monitoring program as described above, or planned alterations of the area in the vicinity of the meteorological tower (s) that would interfere with the measurement of meteorological conditions representative of the site, will require written approval in accordance with Section 5.6.3.

Bases The collection of meteorological data at the plant site will provide information which will be used to develop atmospheric diffusion parameters to estimate potential radiation doses to the public resulting from actual routine or abnormal releases of radioactive materials to the atmosphere, and to assess the actual impact of the plant cooling system on the atmospheric environment of the site area. A meteorological data collection program as described above is necessary to meet the requirements of subparagraph 50.36(a)(2) of 10 CFR Part 50, Appendix E to 10 CFR Part 50, and 10 CFR Part 51.

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5.5.3 Procedures for Environmental Surveillance - Nonradiological Not applicable.

5.5.4 Quality Assurance of Program Results The procedures document shall provide for assurance of the quality of program results, including analytical measurements. This portion of the procedures document shall document the program in policy directives, designate a responsible organization or individuals, include purchased services (e.g., contractual lab or other contract services), include audits by licensee personnel, and include procedures for revising programs, systems to identify and correct deficiencies, investigate anomalous or suspect results, and review and evaluate program results and reports.

5.5.5 Changes in Procedures, Station Design or Operation Changes in procedures, station design or operation may be made subject to conditions described below, provided such changes are approved by the SNSOC (Review and Audit responsibility per Section 5.3),

a. The licensee may (1) make changes in the station design and operation as described in the FES, FES Addendum and the Environmental Report, (2) make changes in the procedures described in the document developed in accordance with Subsection 5.5, and (3) conduct tests and experiments not described in the document developed in acordance with Subsection 5.5, without prior Commis-sion approval, unless the proposed change, test or experiment 2

involves a change in the objectives of the ETS, an unreviewed environmental question of substantive impact, or affects the requirements of Subsection 5.5.6 of these ETS.

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b. A proposed change, test, or experiment shall be deemed to involve an unreviewed environmental question (1) if the probability of magnitude of environmental impact may be increased; or (2) if a possibility for a substantive environmental impact of a different type than any evaluated previously in the FES or FES Addendum may be created.

I-5-6

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c. The licensee shall maintain records of changes in procedures and in facility design or operation made pursuant to this Subsection, to the extent that such changes constitute changes in procedures as described in the document developed in accordance with Section 1 5.5 or in the FES, FES Addendum and ER. The licensee shall also maintain records of tests and experiments carried out pursuant to

paragraph "a" of this Subsection. These records shall include a j written evaluation which provides the bases for the determination that the chang , test, or experiment does not involve an

unreviewed environmental question of substantive impact or constitute a change in the objectives of these ETS, or affect the

requirements of Section 5.5.6 of these ETS. The licensee shall furnish to the Commission, annually or at such shorter intervals as may be specified in the license, a report containing i descriptions, analyses, interpretations, and evaluations of such j changes, tests and experiments.

d. Changes in the procedures developed in accordance with Subsection

5.5 which affect sampling frequency, location, gear, or replication shall be reported to the NRC within 30 days after

their implementation. These reports shall describe the changes

made, the reasons for making the changes, an evaluation of the environmental impact of these changes, and the statement required

, under the provisions of Subsection 5.5.6.

5.5.6 Consistency with Initially Approved Programs Any modifications or changes of the procedures developed in accordance with Subsection 5.5 must be governed by the need to maintain consistency with previously used procedures so that direct comparisons

)j of data are technically valid. Such modifications or changes must be justified and supported by adequate comparative sampling programs or i studies demonstrating the comparability of results or which provide a i basis for making adjustments that would permit direct comparisons.

These demonstrations of comparability shall be submitted to the NRC in i accordance with the provision of Subsection 5.5.5 and 5.6.1 of these ETS.

5.6 Station Reporting Requirements l

4 5.6.1 Routine Reports - None.

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( 5.6.2 Nonroutine Reports f.

5.6.2.1 Monroutine Non-Radiological Environmental Operating Report None.

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....7 5.6.3 Changes in Environmental Technical Specifications +=

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A report shall be made to the NRC prior to implementation of a change *

', in plant design, in plant operation, or in procedures described in

} Section 5.5 if the change would have a significant effect on the j environment or involves an environmental matter or question not :

j previously reviewed and evaluated by the NRC. The report shall !

} include a description and evaluation of the change and a supporting '

benefit-cost analysis.

4

Request for changes in Environmental Technical Specifications shall be submitted to the Director, Office of Nuclear Reactor Regulation, for review and authorization. The request shall include an evaluation of the environmental impact of the proposed changes and a supporting benefit-cost analysis.

! 5.6.4 Changes in Permits and Certifications j None l .

5.7 Records Retention l Records and logs relative to the following areas shall be made and retained for the life of the station:

a. Records and drawings detailing plant design changes and j modifications made to systems and equipment as described in Section 5.6.3.

b. Reports from environmental monitoring, surveillance, and special j surveillance and study activities required by these Environmental Technical Specifications.

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ML20063N338

Text

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