Text

Proposed Changes to Radiological Effluent Tech Specs
	ML20085L078
Person / Time
Site:	Brunswick
Issue date:	10/17/1983
From:	; CAROLINA POWER & LIGHT CO.
To:	;
Shared Package
ML20085L068	List: ML20085L065; ML20085L078;
References
	NUDOCS 8310210268
	Download: ML20085L078 (200)
	v • d • e

ENCLOSURE 3 2

RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS BRUNSWICK 3 TEAM ELECTRIC PLANT, UNIT NO. 1 REFERENCE No. 83TSB16 i 8310210268 831017 PDR ADDCK 05000324 PDR p

INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS ACTION............................................................1-1 AVERAGE PLANAR EXPOSURE ............................................ 1-1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE . . . . . . . . . . . . . . . . . . . . . . . . 1-1 ,

CHANNEL CALIBRATION ................................................ 1-1 CHANNEL CHECK ..................................................... 1-1 CHANNEL FUNCTIONAL TCT ............................................ 1-1 C ORE A LTE RAT ION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 CRITICAL POWER RATIO ..............................>............... 1-2 DOSE E QUIVALENT I- 131 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 E-AVERAGE DISINTEGRATION ENERGY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME . . . . . . . . . . . . . . . . 1-2 FREQUENCY NOTATION ................................................. 1-2 GASEOUS RADW ASTE TREATMENT SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 l

IDENTIFIED LEAKAGE ................................................. 1-3 ,

ISOLATION SYSTEM RESPONSE TIME . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 LIMITING CONTROL ROD PATTERN ...................................... 1-3 L INEAR HEAT GENERATION RATE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 LOGIC SYSTEM FUNCTIONAL TEST ...................................... 1-3 MAXIMUM TOTAL PEAKING FACTOR ...................................... 1-4 MEMBER (S) 0F THE PUBLIC ........................................... 1-4 MINIMUM CRITICAL POWER RATIO . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 ODYN OPTION A....................................................... 1-4 O DYN O PT ION B . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -4 0FFSITE DOSE CALCULATION MANUAL (ODCM) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 O PERABLE - O PERABILITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 OPERATIONAL CONDITION ............................................. 1-5 P HYS IC S T E STS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 PRES SURE BOUNDARY LEAKAGE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- 5 PRIMARY CONTAINMENT INTEGRITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 BRUNSWICK - UNIT 1 I Amendment No.

INDEX DEFINITIONS SECTION 1.0 DEFINITIONS (Continued) PAGE PROCESS CONTROL PROGRAM (PCP) ..................................... 1-5 PURGE - PURGING ................................................... 1-6 RATED THERMAL POWER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 REACTOR PROTECTION SYSTEM RESPONSE TIME . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 REFERENCE LEVEL ZERO . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 REPORTABLE O CCURRENCE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 ROD DENSITY ....................................................... 1-6 SECONDARY CONTAI19fENT INTEGRITY ................................... 1-6 SHUTDOWN MARGIN ................................................... 1-7 SITE BOUNDARY ..................................................... 1-7 SOLIDIFICATION .................................................... 1-7 S OURCE CHECK . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- 7 SPIRAL RELOAD ..................................................... 1-7 S P IRAL U NLOAD . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- 7 STAGGERED TEST BASIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-7 THERMAL POWER ..................................................... 1-8 i

TOTAL PEAKING FACTOR ............................................... 1-8 UNIDENTIFIED LEAKAGE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-8 UNRESTRICTED AREA ................................................. 1-8 VENTIIATION EXHAUST TREATMENT SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-8 VENTING ........................................................... 1-8 FREQUENCY NOTATION , TABLE 1.1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-9 l OPERATIONAL CONDITIONS , TABLE 1. 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-10 BRUNSWICK - UNIT 1 II Amendment No.

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION _ PAG _E_

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION . . . . . . . . . . . . . . . 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION . . . . . . . . . . . . . . . . . . . . . 3/4 3-9 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION . 3/4 3-30 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION . . . . . . . . . . . . . 3/4 3-39 3/4.3.5 MONITORING INSTRUMENTATION Seismic Monitoring Insertsmentation . . . . . . . . . . . . . . . . . . . . . . 3/4 3-44 Remote Shutdown Monitoring Instrtsnentation . . . . . . . . . . . . . . . 3/4 3-47 Po st-accident Monitoring Ins trumentation . . . . . . . . . . . . . . . . 3/4 3-50 Sour ce Range tenit o r s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 3-5 3 Chlo rine De t e ction S ys t em . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 3-5 4 Chloride Intrusion Monitors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-5 5 l

Fire Detection Ins trinnentation . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-59 Radioactive Liquid Effluent Monitoring Instrtsnentation .. 3/4 3-62 Radioactive Gaseous Ef fluent Monitoring Instrumentation . 3/4 3-68 3/4.3.6 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION . . . . . . 3/4 3-78 l

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Re circulat ion Lo o p s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/ 4 4- 1 Je t Pump s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/44-2 Idle Re circulation Loop St artup . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-3 3/4.4.2 S AFETY / RELIEF VALVE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 4-4 l 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 1

i Le akage De t e ction S ys t ems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 4-5 Ope ratio nal Isakag e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 4-6 l

BRUNSWICK - UNIT 1 V Amendment No.

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMNTS 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Co ncen t ra tion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/ 4 11-1 Dose - Liquid Ef fluents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 11-8 Liquid Radwaste Treatment System ........................ 3/4 11-9 Liquid Holdup Tanks ...................................... 3/4 11-10 3/4.11.2 GASEOUS EFFLUENTS Do s e Ra t e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-11 Do s e - No ble Ga s e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-15 Dose - Icdine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form .................................... 3/4 11-16 Gaseous Radwaste Treatment System ........................ 3/4 11-17 Ventilation Exhaust Treatment System ..................... 3/4 11-18 Explosive Gas Mixture .................................... 3/4 11-19 Main Condenaer Air Ejector Radioactivity Release Rate .... 3/4 11-20 Drywell Venting or Purging . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-21 3/4.11.3 SOLID RADI0 ACTIVE WASTE .................................. 3/4 11-22 3/4.11.4 TOTAL DOSE (40 CFR PART 190) ............................ 3/4 11-23 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONTTORING 3/4.12.1 MONITORING PROGRAM . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 12- 1 3/4.12.2 LAND USE CENS US . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 12-13 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM ....................... 3/4 12-15 BRUNSWICK - UNIT 1 IXa Amendment No.

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INDEX BASES i

l SECTION PAGE 3/4.0 AP PLICAB ILITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 0- 1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN .......................................... B 3/4 1-1 3/4.1.2 REACTIVITY ANOMALIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 1-1 3/4.1.3 CONTRO L ROD S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 1- 1 3/4.1.4 CONTROL ROD PROGRAM CONTROLS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM ............................ B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE ............... B 3/4 2-1 3/4.2.2 APRM SETPOINTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 2-3 3/4.2.3 MINIMUM CRITICAL POWER RATIO ............................. B 3/4 2-3 3/4.2.4 LINEAR HEAT GENERATION RATE .............................. B 3/4 2-5 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION ................ B 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ........................................ B 3/4 3-2 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION . . . . . . . . . . . . . B 3/4 3-2 3/4.3.5 MONITORING INSTRUMENTATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 3-2 3/4.3.6 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION . . . . . . B 3/4 3-5 l 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM ..................................... B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 4- 1 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE ........................... B 3/4 4-1 l

l l

l BRUNSWICK - UNIT 1 X Amendment No.

1 i

IMDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS (Continued) 3/4.7.3 FLOOD P ROTECTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 7- 1 3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM . . . . . . . . . . . . . . . . . . . . B 3/4 7-1 3/4.7.5 HYDRAULIC SNUB BERS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 7-2 3/4.7.6 SEALED SOURCE CONTAMINATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 7-3 3/4.7.7 FIRE SUPPRESSION SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 7-3 3/4.7.8 FIRE BARRIER PENETRATION S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 7-4 3/4.8 ELECTRICAL POWER SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 8- 1 3/4.9 REFUELING OPERATIONS 3/4.9.1 RE ACTO R MO DE SWITCH . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 9 - 1 3/4.9.2 INSTRUMENTATIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 9- 1 3/4.9.3 CONTROL ROD P O SITION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.4 D E C AY T IME . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 9 - 1 3/4.9.5 CQMMJNICATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 9- 1 3/4.9.6 CRANE AND HOIST OPERABILITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9- 1 3/4.9.7 CRANE TRAVEL-SPENT FUEL STORAGE POOL . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.8 WATER LEVEL-REACTOR VESSEL, and 3/4.9.9 WATER LEVEL-SPENT FUEL STORAGE POOL . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-2 3/4.9.10 CONTRO L RO D REMO VAL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 9 - 2 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 10-1 3/4.10.2 ROD SEQUENCE CONTROL SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 10- 1 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 10-1 3/4.10.4 RECIRCULATIO N LOOP S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 10- 1 3/4.10.5 P LANT S ERV ICE WATER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 10- 1 BRUNSWICK - UNIT 1 XII Amendment No.

INDEX BASES SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Co n c en t ra t ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E 3 /4 11-1 Dose - Liquid Effluents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 11-1 Liquid Was te Treatment System . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 11-2 Liquid Holdup Tanks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 11-3 3/4.11.2 GASEOUS EFFLUENTS Do se Ra t e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 1 1-3 Do s e-Noble Gases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 11-4 Dose-Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form . . . . . . . . . . . . . . . . . . . . . . B 3 /4 11-4 Gaseous Radwas te Treatment System . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 11-5 Ventilation Exhaust Treatment System . . . . . . . . . . . . . . . . . . . . . B 3 /4 11-5 Explo sive Ga s Mix ture . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 11-5 Main Condenser Air Ejector Radioactivity Release Rate .... B 3/4 11-6 Drywell Venting or Purging . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 11-6 3/4.11.3 SOLID RADIOACTIVE WASTE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 11-6 3/4.11-4 TOTAL DOSE (40 CFR PART 190) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 11-6 3/4.12 RADIOACTIVE ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM ....................................... B 3/4 12-1 l 3/4.12.2 LAND USE CENS US . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 12-1 i 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 12-2 l

l BRUNSWICK - UNIT 1 XIIA Amendment No.

INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE Exclusion Area ................................................ 5-1 Low Population Zone ........................................... 5-1 Site Boundary ................................................. 5-1 5.2 CONTAINMENT Configuration ................................................. 5-1 Design Temperature and Pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.3 REACTOR CORE ,

Fuel Assemblies ............................................... 5-1 Control Rod Assemblies ........................................ 5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-4 Volume ........................................................ 5-4 5.5 METEOROLOGICAL TOWER LOCATION .................................. 5-4 5.6 FUEL STORAGE Criticality ................................................... 5-5 Drainage ...................................................... 5-5 Capacity ...................................................... 5-5 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT ............................ 5-5 BRUNSWICK - UNIT 1 XIII Amendment No.

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RE S PO NS I B ILI T Y. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 1 6.2 ORGANIZATION 6.2.1 0FFSITE....................................................... 6-1 6.2.2 FAC IL ITY STAFF . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 1 6.2.3 ONSITE NUCLEAR SAFETY GROUP Function...................................................... 6-8 Re s p o n s i b i l i t i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 8 Authority..................................................... 6-8 6 . 2.4 SHIFT TECRNICAL ADVIS0R....................................... 6-8 6.3 FACILITY STAFF QUALIFICATIONS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-8 6.4 TRAINING...................................................... 6-9 6.5 REVIEW AND AUDIT 6.5.1 NUCLE AR S AFETY REVIEWERS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-9 6.5.2 SAFETY REVIEW AND CONTROL Procedures , Te s ts , and Experiment s. . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-9 Nbd i f i c a t i o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 1 1 Op e rating Li ce ns e Change s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-12 6.5.3 PLANT NUCLEAR SAFETY COMMITTEE (PNSC)

Function...................................................... 6-12 Co mp o s i t i o n. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 13 Alternates.................................................... 6-13 >

He e t ing EYeq ue n cy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 13 Qu o r um. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 1 3 l

Ac t i v i t i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 13 Authority..................................................... 6-14 Re c o rd s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 1 5 BRUNSWICK - UNIT 1 XIV Amendment No.

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.5.4 CORPORATE NUCLEAR SAFETY SECTION Function...................................................... 6-15 Organization.................................................. 6-15 Re vi e w . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . 6 - 16 Re c o r d s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 1 7 6.5.5 CORPORATE QUALITY ASSURANCE AUDIT PROGRAM Function...................................................... 6-18 Audits........................................................ 6-18 Re c o rd s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 1 9 Authority..................................................... 6-19 Pe rs o nne l. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 19 6.5.6 OUTSIDE AGENCY INSPECTION AND AUDIT P ROGRAM. . . . . . . . . . . . . . . . . . . 6-19 6.6 REPORTABL E OCCURRENCE ACTIO N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 0 6.7 S AFETY LIMIT VIO LATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-20 6.8 PROCEDURES AND PR0 GRAMS....................................... 6-20 6.9 REPORTING REQUIREMENTS I

( Routine Reports and Re portable Oc currences. . . . . . . . . . . . . . . . . . . . 6-21 l

l St a r t u p Re p o r t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 2 1 An nu a l Re p o r t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 2 2 Personnel Exposure and Monitoring Report . . . . . . . . . . . . . . . . . . . . . 6-22 Annual Radiological Environmental Operating Report . . . . . . . . . . . 6-23 l Semiannual Radioactive Effluent Releas e Repor t . . . . . . . . . . . . . . . 6-2 4 l

l Monthly Op erating Re po rt s. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-25 l

l Re po rtable Occu rrence s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 5 l

l Promp t No tification With Written Followup. . . . . . . . . . . . . . . . . . . . . 6-25 Thi r t y Day Writ t e n Re p o r t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 7 S p e ci a l Re p o r t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 2 8 BRUNSWICK - UNIT 1 XV Amendment No.

INDEX

. \DMINISTRATIVE CONTROLS -

t SECTION PAGE ,

s 6.10 RECO R D RETENTIO N. . . . . . . . . . . . . . . . . . . . . '. .l. . . . . . . . . . . . . . . . . . . . . . . 6-2 9 6.11 RADI ATION PROTECTION PR0 GRAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-31

~~

6.12 HIGH RADI ATIO N AREA. . . . . . . . . . . . . . . . . . . . . . '. . . . . . . . . . . . . . . . . . . . . 6-31 6.13 0FFSITE DOSE CALCULATION MANUAL (ODCM) ....................... 6-32 6.14 PROCES S CONTROL PRO'IRAM (PCP) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 2 6.15 MAJOR CHANGES TO LIOUID, GASEOUS ,

SOLID WASTE TRE ATMEN T SY STEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 3 w

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BRUNSWICK - UNIT 1 XVI l Amendment No.

I -

i

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

DEFINITIONS The following terms are defined so that uniform interpretation of these specifications may be achieved. The defined terms appear in capitalized type and are applicshle throughout these Technical Specifications.

ACTION ACTIONS are those additional requirements specified as corollary statements to each specification and shall be part of the specifications.

AVERAGE PLANAR EXPOSURE The AVERAGE PLANAR EXPOSURE shall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

AVERAGE PLANAR LINEAR HEAT GENERATION RATE The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of f>sel rods in the fuel bundle.

CHANNEL CALIBRATION 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 CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CtIANNEL 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 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 indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:

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

BRUNSWICK - UNIT 1 1-1 Amendment No.

DEFINITIONS CHANNEL FUNCTIONAL TEST (Continued)

b. Bistable channels - the injection of a simulated signal into the ,

channel sensor to verify OPERABILITY including alarm and/or trip

functions.

CORE ALTERATION CORE ALTERATION shall be the addition, removal, relocation, or movement of fuel, sources, incore instruments, or reactivity controls in the reactor core with the vessel head removed and fuel.in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe, conservative location.

CRITICAL POWER RATIO The CRITICAL POWER RATIO (CPR) shall be ratio of that power in the assembly which is calculated, by application of the GEKL correlation, to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be concentration of I-131, pCi/ gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of

I-131, I-132, I-133, 1-134, and I-135 actually present. The following is defined equivalent to 1 uCi of I-131 as determined from Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites": I-132, 28 pCi; I-133, 3.7 uCi; I-134, 59 uCi; I-135,12 pCf.

]

E-AVERAGE DISIlffEGRATION ENERGY 5 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 gamme energies per disintegration (in MeV) for isocopes with half lives greater than 15 minutes making up at least 95% of the total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME The FMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

. BRUNSWICK - UNIT 1 1-2 Amendment No.

i DEFINITIONS i

GASE3US RADWASTE TREATMENT SYSTEM A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary systes and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the I

environment.

IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE shall be:

l a. IAakage into Collection systems, Much as pump seal of Valve packing leaks, that is captured and conducted to a sump or collecting tank, or i

b. TAakage into the Containasnt atmosphere from sources that are both specifically located and known either not to interfere with the operation of the leskage detection systems or not be PRESSURE BOUNDARY LEAKAGE.

ISOLATION SYSTEM RESPONSE TIME The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading delays where applicable.

LIMITING CONTROL ROD PATTERN 4 A LIMITING CONTROL ROD PATIERN shall be a pattern which results in the core j being on a thermal hydraulic limit, i . e., operating on a limiting value for APLHGR, LHGR, or MCPR.

LINEAR HEAT GENERATION RATE LINEAR HEAT GENERATION RATE (LHGR) shall be the power generation in an arbitrary length of fuel rod, usually one foot. It is the integral of the heat flux over the heat transfer area associated with the unit length, usually measured in kW/ft.

LOGIC SYSTEM FUNCTIONAL TEST A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all relays and contacts of a logic circuit, from sensor output to activated device, to ensure that components are OPERABLE.

BRUNSWICK - UNIT 1 1-3 Amendment No.

J

- m---,. _ _ _w,_..w-~,_,-m,_..- -,,,, ._.___,,,__,_,_,.-,-.,,--...,_y. , . _ _ . . . _ , . , , . ..-.,y, .,--.,% ,, , ,,gy._y_.,w,r,w,%g,.y.- , _%,_y.-n

i I

DEFINITIONS MAXIMUM TOTAL PEUCING TACTOR The MAXIMUM TOTAL PEAKING FACTOR (MTPF) shall be the largest TPF which exists j in the core for w given class of fuel for a given operating condition.

MEMBER (S) 0F THE PUBLIC MEMBER (S) 0F Tile PUBLIC shall include all persons who are not occupationally l associated with the plant. This category does not include employees of the '

utility, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliverfes. This i catagory does include persons who use portions of the site for recreational, j occupational or other purposes not associate.d irtth the plant.

MINIMUM CRITICAL POWER RATIO The MIN 1 MUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists i in the core.

ODYN OPTION A ODYN OPTION A shall be analyses which refer to minimum critical power ratio limits which are determined using a transient analysis plus an analysis uncertainty penalty.

ODYN OPTION B ODYN OPTION B shall be analyses which refer to minimum critical power ratio limits determined using a transient analysis which includes a requirement for 20% scram insertion times to reduce the analysis uncertainty penalty.

OFFSITE DOSE CALCULATION MANCAL (ODCM)

The OFFSITE DOSE CALCULATION MANUAL (ODCM) is a manual which contains the current methodology and parameters to be used to calculate offsite doses resulting from the release of radioactive gaseous and liquid effluents; the methodology to calculate gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints; and, the requirements of the environmental radiological monitoring program.

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

l Implicit in this definition shall be the assumption that all necessary l attendant instrumentation, controls, normal and emergency electric 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).

BRUNSWICK - UNIT 1 1-4 Ar.eadment No.

I DEFINITIONS l

OPERATIONAL CONDITION An OPERATIONAL CONDITION shall be any one inclusive comoination of mode switch p::ition and average reactor coolant temperature as indicated in Table 1.2.

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and are 1) described in Section 13 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault in a reactor coolant system component body, pipe wall, or vessel wall.

PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE containment automatic isolation valve system, or

2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification 3.6.3.1, or

b. All equipment hatches are closed and sealed.

c. Each containment air lock is OPERABLE pursuant to Specification 3.6.1.3.

d. The containment leakage rates are within the limits of Specification 3.6.1.2.

1

e. The sesling mechanism associated with each penetration (e.g., welds, bellows, or 0-rings is) is OPERABLE.

i PROCESS CONTROL PROGRAM (PCP)

The PROCFSS CONTROL PROGRAM (PCP) 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.

BRUNSWICK - UNIT 1 1-5 Amendment No.

1

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

RATED THERMAL POWER RATED THERMAL POWER shall be total reactor core heat transfer rate to the reactor coolant of 2436 MWt.

REACTOR PROTECTION SYSTEM RESPONSE TIME REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when i the monitored parameter exceeds its trip setpoint at the channel sensor until i de-energization of the scram pilot valve solenoids.

REFERENCE LEVEL ZERO The REFERENCE LEVEL ZERO point is arbitrarily set at 367 inches above the

, vessel sero point. This REFERENCE LEVEL ZERO is approximately mid-point on i the top fuel guide and is the single reference for all specifications of vessel water level.

REPORTABLE OCCURRENCE A REPORTABLE OCCURRENCE shall be any of those conditions specified in Specification 6.9.1.13 and 6.9.1.14 ROD DENSITY ROD DENSITY shall be the number of control rod notches inserted as a fraction of the total number of notches. All rods fully inserted is equivalent to 100% ROD DENSITY.

SECONDARY CONTAINMENT INTEGRITY SECONDARY CONTAINME!ff INTEGRITY shall exist when:

a. All automatic reactor building ventilation system isolation valves or dampers are OPERABLE or secured in the isolated position.

b. The standby gas treatment system is OPERABLE pursuant to Specification 3.6.6.1.

c. At least onc door in each access to the reactor building is closed.

d. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

BRUNSWICK - INIT I 1- 5 Amendment No.

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

DEFINITIONS SHUTDOWN MARGIN SHUIDOWN MARGIN shall be the amount of reactivity by which the reactor would be suberitical assuming that all control rods capable of insertion are fully inserted except for the analytically determined highest worth rod which is assumed to be fully withdrawn, and the reactor is in the shutdown condition, cold, 68'F, and Xenon-free.

SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee, as defined by Figure 5.1.3-1.

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

SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to radiation.

SPIRAL RELOAD A SPIRAL RELOAD is the reverse of a SPIRAL UNLOAD. Except for two diagonal fuel bundles around each of the four SRMs, the fuel in the interior of the core, symmetric to the SRMs, is loaded first.

SPIRAL UNLOAD A SPIRAL UNLOAD is a core unload performed by first removing the fuel from the outermost control cells (four bundles surrounding a control blade). Unloading continues in a spiral fashion by removing fuel from the outermost periphery to the interior of the core, symmetric about the SRMs, except for two diagonal fuel bundles around each of the fotr SRMs.

STAGGERED TEST BASIS 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.

BRUNSWICK - UNIT 1 1-7 Amendment No.

DEFINITIONS THERMAL POWER THERMAL POWER shall be the total reactor core hear transfer rate to the reactor coolant.

TOTAL PEAKING FACTOR The TOTAL PEAKING FACTOR (TPF) shall be the ratio of local LHCR for any specific location on a fuel rod divided by the average LRGR associated with the fuel bundles of the same type operating at the core average bundle power.

UNIDENTIFIED LEAKAGE UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LFAKAGE.

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

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

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

BRUNSWICK - UNIT 1 1-8 Amendment No.

TABLE _1.1 FRE0fJENCY 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.

SM At least once per 16 days.

M At least once per 31 days.

Q At least once per 92 days..

SA At least once per 184 days.

A At least once per 366 days.

R At least once per 18 months (550 days).

S/U Prior to each reactor startup.

P Prior to each release.

NA Not applicable.

l i

I BRUNSWICK - UNIT 1 1-9 Amendment No.

TABLE 1.2 OPERATIONAL CONDITIONS OPERATIONAL MODE SWITCH AVERAGE COOLANT CONDITIONS POSITIONS _ TEMPERATURE

1. POWER OPERATION Run Any temperature

2. STARTUP Startup/ Hot Standby Any temperature

3. HOT SHUTDOWN Shutdown > 212*F

4. COLD SHUTDOWN Shutdown f *12 *F

5. REFUELING

Refuel ** f 212*F

Reactor vessel head unbolted or removed and fuel in the vessel.***

- See Special Test Exception 3.10.3.

- - See Special' Test Exception 3.10.1.

BRUNSWICK - UNIT 1 1-10 Amendment No.

TABLE 3.3.5.7-1 (Continued)

INSTRUMENT LOCATION MINIMUM INSTRUMENTS OPERABLE FLAME HEAT SMOKE

3. Diesel Generator Building (Cont'd)

Zone 7 23' 0 0 5 Zone 8 23' 0 0 5 "one 9 23' O 0 8 Zone 10 50' O O 9

4. Service Water Building Zone 1 4' O O 7 Zone 2 20' 0 0 6

5. A0G Building Zone 1 20' 0 0 2 Zone 2 20' O O 2 Zone 3 20' 1 5 1 Zone 4 37'-49' 1 6 0 BRUNSWICK - UNIT 1 3/4 3-61 Amendment No.

INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT t40NITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.5.8 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3.5.8-1 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 shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY: As shown in Table 3.3.5.8-1.

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, declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With less than one radioactive liquid effluent monitoring instrumen-tation channel in each release pathway OPERABLE, take the ACTION shown in Table 3.3.5.8-1. Return the instruments to OPERABLE status within 30 days or, 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 6.9.1.14.b are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.5.8 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 4.3.5.8-1.

NOTE: See Bases 3/4.3.5.8.

BRUNSWICK - UNIT 1 3/4 3-62 Amendment No.

$ TABLE 3.3.5.8-1 8!

s RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION INSTRUMENT (a) APPLICABILITY ACTION i

E! 1. Liquid Radwaste Radioactivity Effluent Monitor II (Providing alarm and automatic termination of release)

110 e

2. Liquid Radwaste Ef fluent Flow Measurement Device

111

3. Main Service Water Effluent Radioactivity Monitor

112 i

4. Stabilization Pond Ef fluent Composite Sampler ** 113

5. Stabilization Pond Effluent Flow Heasurement Device ** 114 hh 6. Condensate Storage Tank Level Indicating Device

115 4

Y 7. Service Water Effluent from Augmented Off-Gas O Precooler Radioactivity Monitor *** 112

, 8. Reactor Building Component Cooling Water (Service Water)

Radioactivity Monitors

a. Effluent from Residual Heat Removal Heat Exchanger A **** 112

b. Effluent from Residual Heat Removal Heat Exchanger B **** 112 g c. Effluent from Reactor Building Closed og Cooling Water that Exchangers **** 112 i

g g d. Effluent from Division I Residual Heat n Removal Pump Seal Coolers **** 112

= e. Effluent from Division II Residual Heat Removal Pump Seal Cooiers **** 112 e a

1 l

l TABLE 3.3.5.8-1 (Continued)

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION I

ACTIONS ACTION 110 - With less than one channel OPERABLE, effluent releases may continue provided that prior to initiating a release:

a. At least two independent samples are analyzed in accor-dance with Specification 4.11.1.1.2, and

b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving; Otherwise suspend release of radioactive effluents via this pathway.

ACTION 111 - With less than one channel OPERABLE, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump per-formance curves or tank level indicators may be used to esti-mate flow.

ACTION 112 - With less than one channel OPERABLE, effluent releases 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 sam-pies are collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 10-7 microcuries per gram.

ACTION 113 - With the stabilization pond effluent composite sampler not OPERABLE, effluent releascs may continue provided that, at least.once per day, a grab sample is collected and analyzed for principle gamma emitters as per Table 4.11.1-1. Other-l wise, suspend releases via this pathway.

j ACTION 114 - With-the stabilization pond effluent flow measuring device not OPERABLE, effluent releases via this pathway may continue provided that flow is estimated at least once per day during actual releases. The V-notch weir may be used to estimate flow. ,

l ACTION 115 - With the tank liquid level device not OPERABLE, liquid addi-

! tions may continue provided the tank liquid level is estimated once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months during all liquid additions and deletions to and from the tank.

l BRUNSWICK - UNIT 1 3/4 3-64 Amendment No.

l 1

TABLE 3.3.5.8-1 (Continued)

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION l

NOTES

At all times

- At all times other than when the line is valved out and locked. [This equipment is to be installed. Prior to installation, appropriate action statements 113 or 114 will be implemented.]

- - At all times while the A0G system precooler is in operation.

- - - At all times once these monitors are installed and become fully operational. [ NOTE: These monitors are to be installed pending completion of future plant modifications.]

(a) Refer to Appendix E of the OFFSITE DOSE CALCULATION MANUAL for specific instrumentation identification numbers.

i t

l l

BRUNSWICK - UNIT 1 3/4 3-65 Amendment No.

, TABLE 4.3.5.8-1 RADIOACTIVE LIQUID EFFLUENT NONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS s

Q CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMEN[a) CHECK CHECK CALIBRATION TEST N 1. Liquid Radwaste Radioactivity Effluent Monitor H (Providing alarm and automatic termination of release) D M R(b) q(c)

2. Liquid Radwaste Effluent Flow Measurement Device D(e) NA R Q 3 Main Service Water Effluent Radioactivity Monitor D M R(b) q(d)

4. Stabilization Pond Effluent Composite Sampler D NA R Q

5. Stabilization Pond Effluent Flow Measurement Device D NA R Q D* 6. Condensate Storage Tank level Indicating Device D(f) NA R Q Y

$ 7. Service Water Effluent from Augmented Off-Cas Precooler Radioactivity Monitor D M R(b) q

8. Reactor Building Component Cooling Water (Service Water) Radioactivity Monitors

a. Effluent from Residual Heat Removal Heat Exchanger A D M R(b) q(d)

b. Effluent from Residual Heat flemoval Heat Exchanger B D M R(b) q(d)

{

r3

c. Effluent from Reactor Building Closed Cooling Water Heat Exchangers D M R ID) Q(d)

I

$ d. Effluent from Division 1 Residual Heat Removal Pump Seal Coolers D M R(b) q(d) o

e. Effluent from Division II Residual Heat Removal Pump Seal Coolers D M R(b) q(d)

I l

TABLE 4.3.5.8-1 (Continued)

RADIOACTIVE LIOUID EFFLUENT MONITORING 1NSTRUMENTATION SURVEILLANCE REOUIREMENTS TABLE NOTATIONS (a) Refer to Appendix E of the OFFSITE DOSE CALCULATION MANUAL for specific instrumentation identification numbers.

(b) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATIOM, sources that have been related to the initial calibration shat' be used. Previously established calibration procedures may be substituted for this requirement (refer to Bases 3/4.3.5.8).

(c) 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 exist:

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

2. Circuit failure (Eigh-voltage low).

3. Instrument indicates a downscale failure.

4. Instrument controls not set in " operate" mode.

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

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

2. Circuit failure (High-voltage low).

3. Instrument indicates a downscale failure.

4. Instrument controls not set in " operate" mode.

l (e) The CHANNEL CHECK shall consist of verifying indication of flow

during periods of release. CHANNEL CHECK shall be made at least l once daily on any day on which continuous, periodic, or batch releases are made.

(f) During liquid additions to the tank.

BRUNSWICK - UNIT 1 3/4 3-67 Amendment No.

l

INSTRUMENTATION RADI0 ACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.5.9 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3.5.9-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The setpoints shall be determined in accordance with the methodology as described in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICARILITY: As shown in Table 3.3.5.9-1.

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 one radioactive gaseous effluent monitoring instrumentation channel OPERABLE, take the ACTION shown in Table 3.3.5.9-1. Return the instruments to OPERABLE status within 30 days or, 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 6.9.1.14.b are not l applicable.

SURVEILLANCE REQUIREMENTS 4.3.5.9 Each radioactive gaseous effluent monitoring instrumentatiota channel l shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHEG, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 4.3.5.9-1.

NOTE: See Bases 3/4.3.5.9.

BRUNSWICK - UNIT 1 3/4 3-68 Anendment No.

tn l TABLE 3.3.5.9-1 i! RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION INSTRUMENT (a) APPLICABILITY ACTION Bi y 1. MAIN STACK MONITORING SYSTEM

a. Noble Gas Activity Monitor 123

b. Iodine Sampler Cartridge

127

c. Particulate Sampler Filter

127

d. System Effluent Flow Rate Measurement Device

122

e. Sampler Flow Rate Measurement Device

122 w 2. REACTOR BUILDING VENTILATION MONITORING SYSTEM

a. Noble Gas Activity Monitor

123

b. Iodine Sampler Cartridge

127

c. Particulate Sampler Filter

127

d. System Effluent Flow Rate Measurement Device

722

c. Sampler Flow Rate Measurement Device

122 I

g 3. TURBINE BUILDING VENTILATION MONITORING SYSTEM 8-

a. Noble Gas Activity Monitor

123 eg re

b. Iodine Sampler Cartridge

127

c. Particulate Sampler Filter

127

tn g TABLE 3.3.5.9-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION R ACTION

, INSTRUMENT (a) APPLICABILITY H

h 3. TURBINE BUILDING VENTILATION MONITORING SYSTEM (Continued)

122

d. System Effluent Flow Rate Pleasurement Device

e. ' Sampler Flow Rate Measurement Device

122

4. MAIN CONDENSER AIR EJECTOR RADIOACTIVITY MONITOR (Prior to input to treatment system)

Noble Gas Activity Monitor ** 121 a.

(Providing alarm and automatic isolation)

M 5. MAIN CONDENSER OFF-CAS TREATMENT SYSTEM MONITOR (Downstream of AOC Treatment System)

Y 5 a. Noble Gas Activity Monitor (providing alarm)

123

6. MAIN CONDENSER OFF-GAS TREATMENT SYSTEM EXPLOSIVE GAS MONITORING SYSTEM

a. Recombiner Train A

1. Ist liydrogen Monitor *** 125

2. 2nd Ilydrogen Monitor *** 125

b. Recombiner Train B

1. Ist Hydrogen Monitor *** 125

2. 2nd Hydrogen Monitor *** 125 g

7. Il0T S110P VENTILATION MONITORING SYSTEM n

Iodine Sampler Cartridge

127 N a.

2:

127

. b. Particulate Sampler Filter

TABLE 3.3.5.9-1 (Continued _)

RADIOACTIVE GASEOUS EFFLUENT MONIT0 RING INSTRUMENTATION ACTIONS ACTION 121 - With less than one main condenser air ejector monitoring instr-umentation channel OPERABLE, gases from the main condenser off-gas system may be released to the environment for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided:

a. The GASEOUS RADWASTE TREATMENT SYSTEM is not bypassed, and

b. The main stack effluent noble gas activity monitor is OPERABLE; otherwise, be in at least HOT STANDBY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 122 - With less than one channel OPERABLE, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

ACTION 123 - With less than one channel OPERABLE, 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 noble gas activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 125 - With less than two channels OPERABLE in the operating recombiner train, operation of the train may continue provided proper function of the recombiner is assured by monitoring recombiner temperature in accordance with approved procedures.

With less than one channel OPERABLE in the operating recombiner

- train, operation of the train may continue provided grab samples from the train are collected at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and proper function of the recombiner is assured by monitoring recombiner

( temperature in accordance with approved procedures.

l ACTION 127 - With less than one channel OPERABLE. effluent releases via this pathway may continue provided samples are continuously col-lected with auxiliary sampling equipment and analyzed as i

required in Table 4.11.2-1.

l l

I i

BRUNSWICK - UNIT 1 3/4 3-71 Amendment No.

L- _

TABLE 3.3.5.9-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUKENTATION NOTES

At all times.

- During operation of the main condenser air ejector.

- - At all times during recombiner train operation.

(a) Refer to Appendix E of the OFFSITE DOSE CALCULATION MANUAL for specific instrumentation identification numbers.

l BRUNSWICK - UNIT 1 3/4 3-72 Amendment No.

m TABLE 4.3.5.9-1

=

RADIOACTIVE GASEQUS EFFLUENT HONITORING IESTRUMENTATION SURVEILLANCE REQUIREMElfrS Q CHANNEL MODES IN WHICll

CHANNEL SOURCE CHANNEL FUt;CTIONAL SURVEILLANCE INSTRUMENT (a) CHECK CHECK CALIBRATION TEST REQUIRED b

H 1. MAIN STACK MONITORING SYSTEM e

a. Noble Gas Activity Monitor D M R(b) q(d) *

b. Iodine Sampler Cartridge W NA NA NA *

c. Particulate Sampler Filter W NA NA NA

l d. System Effluent Flow Rate Measurement Device D NA R Q

D e. Sampler Flow Rate Measurement u Device D NA R Q

b

2. REACTOR BUILDING VENTILATION MONITORING SYSTEM

a. Noble Gas Activity Monitor D M R(b) q(d) .

b. Iodine Sampler Cartridge W NA NA NA *

c. Particulate Sampler Filter W NA NA NA *

d. System Effluent Flow Rate Measurement Device D NA R Q

I, n e. Sampler Flow Rate Measurement Device D NA R Q

n 4

TABLE 4.3.5.9-1 (Continued)

E g RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS M

i CHANNEL MODES IN WHICH g CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE y INSTRUMENT (a) CHECK CHECK CALIBRATION TEST REQUIRED

3. TURBINE BUILDING VENTILATION HONITORING SYSTEM

a. Noble Gas Activity Monitor D M R(b) g(d) ,

b. Iodine Sampler Cartridge W NA NA NA *

c. Particulate dampler Filter W NA NA NA

U d. System Effluent Flow Rate

[, Measurement Device D NA R Q

5 e. Sampler Flow Rate Measurt. ment Device D NA R Q *

4. MAIN CONDENSER AIR EJECTOR RADI0 ACTIVITY MONITOR (I'rior to input to treatment system)

a. Noble Gas Activity Monitor (Providing alarm and automatic isolation) D M R(b) g(c) ,.

g 5. MAIN CONDENSER OFF-CAS TREATMENT g SYSTEM MONITOR g- (Downstream of AOC Treatment System) n a. Noble Gas Activity Monitor g (Providing alarm) D M R(b) 9 ,

en TABLE 4.3.5.9-1 (Continued) a Q RADIOACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH E CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE N INSTRUMENT (a) CHECK CHECK CALIBRATION TEST REQUIRED

6. MAIN CONDENSER OFF-CAS TREATMENT SYSTEM EXPLOSIVE GAS HONITORING SYSTEM

a. Recombiner Train A

1. Ist Hydrogen Monitor D NA M ***

2. 2nd Hydrogen Monitor D NA Q(*))

I Q* M ***

b. Recombiner Train B

1. Ist Hydrogen Monitor D NA I M ***

2. 2nd Hydrogen Monitor D NA Q(*))

Qe y ,,,

y 7. IlOT SHOP VENTILATION MONITORING SYSTEM d a. Iodine Sampler Cartridge W NA NA NA *

, b. Particulate Sampler Filter W NA NA NA

4 9

9

, n O

1 1

TKILE 4.3.5.9-1 (Continued)

RADIOACTIVE CASEQUS EFFLUENT MONITORING INSTRLTENTATION SURVEILLANCE REOUIREMENTS l

TAB LE NOTATION j l

(a) Refer to Appendix E of the OFFSITE DOSE CALCULATION FMNUAL for '

specific instrumentation identification numbers.

(b) The initial CHANNEL CALB RATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIS.

These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIB RATION, sources that have been related to the initial calibration shall be used. Previously established calibration procedures may be substituted for this requirement (refer to B ases 3/4.3.5.9).

(c) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway, ac described below, and control room alarm annunciation occurs if any of the following conditions exist:

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

2. Circuit failure (High-voltage low).

3. Instrument indicates a downscale failure.

4. Instrument not set in " operate" mode.

The CHANNEL FUNCTIONAL TEST of the channel up to but not including operation of the isolation valve for this pathway shall be performed within the specified surveillance interval. Testing of the isolation valve for this pathway to demonstrate operability shall be performed during the CHAMEL CALD RATION.

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

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

2. Circuit failure (High-voltage low).

3. Instrument indicates a downscale failure.

.4 Instrument not set in " operate" mode.

BRUNSWICK - UNIT 1 3/4 3-76 Amendment No.

i

TABLE 4.3.5.9-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRIMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATION i

l (e) The CHANNEL CALIBEATION shall include the use of standard gas sam-ples containing a nominal:

1. Two volume percent hydrogen, balance nitrogen, and

2. Four volume percent hydrogen, balance nitrogen.

NOTES

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

- During operation of the main condenser air ejector.

- - During recombiner train operation.

~

BRUNSWICK - UNIT 1 3/4 3-77 Amendment No.

h k INSTRUMENTATION %s ,

3/4.3.6 ATWS RECIRCULATION PUT TRIP SYSTEM INSTRUMENTATION ( (

LIMITING CONDITION FOR OPERATION s

(%

s.

3.3.6.l s The anticipated transient without scram recirculation pump trip ,,^

(ATWS-RPT) system instrtsnentation trip systems shown in Table 3.3.6.1-1 shall be OPERABLE with their trip setpo'uts set consistent with the values shown in the Trip Setpoint column of Table '3.3.6.1-2. >

s. ..

APPLICA3ILITY: OPERATIONAL CONDITION 1.

's .

ACTION:.

.a.: With an ATWS recirculation pump trip system instrumentation trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.6.1-2, declare the trip system _ingerable until the trip system is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value. .

~

b. With the requirements for the minimum ntsaber of OPERABLE thip shtems

_ . . , per operating pump not satisfied for one Trip Function, restore the i , inoperable trip system to OPERABLE status within 14 days or be-in ac

.' least STARTUP within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . 1 SURVEILLANCE REQUIREMENTS _

ys

~

4.[.6.1.1 Each ATWS recirculation pump trip system instrumentation trip '

system shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK. - '

CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations at the f requencies

~

ihown in Table 4 . 3. 6 .1.1- 1. -

~

i.3 3 1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of i , . -

all channels shall be performed at least once per 18 months and shall include ' -

calibration of time delay relays and timers necessary for proper functioning of,the trip system.

. ~.- _

s q % i %

%1 ,,

n f,.. -

BRUNSWICK - UNIT 1 3/4 3-78 Amendment No.

e - - - - - - _ - . _ . _ _ . _ _ - - ._ - - --

as TABLE 3.3.6.1-1 m

ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION 5

R MINIhUM. K'JABER OPERABLE TRIP TRIP FUNCTION AND INSTRUMENT NUMBER SYSTEMS PER OPERATING PUMP g i

1. Reactor Vessel Water level - 1 ;

" Low, Level 2 l

(B21-LT-N024A-2.B-2 and B21-LT-N025A-2,B-2)

(B21-LTM-N024A-2,8-2 and B21-LTM-N025A-2,8-2)

2. Reactor Vessel Pressure - High I (B21-PS-N045A,B,C,D)

N i'

a s

8 8-U n

.O

=

$ TABLE 3.3.6.1-2 e

ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPCINTS ij h

M TRIP ALLOWABLE I TRIP FUNCTION AND INSTRUMENT NUMBER SETPOINT VALUE E > +112 inches *

[j 1. Reactor Vessel Water Level - > +112 inches

g low, Imvel 2 (B21-LTM-N024A-2,B-2; B21-LTM-N025A-2,B-2)

2. Reactor Vessel Pressure - High f 1120 psig < 1120 psig (B21-PS-N045A,3,C,D)

M s~

Y 8

R 9

9 n

Y

Vessel water levels refer to REFERENCE LEVEL ZERO.

, TABLE 4.3.6.1-1 pa h

ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION 'URVEILLANCE REQUIREMENTS N

p CHANNEL CHANNEL FUNCTIONAL CHANNEL

, TRIP FUtiCTION AND INSTRUMENT NUMBER CHECK TEST CALIBRATION h 1. Reactor Vessel Water Level-H low, level 2 H (B21-LT-N024A-2,B-2 and NA(a) NA R(b)

B21-LT-N025A-2,B-2)

(B21-LTM-N024A-2,B-2 and D M M B21-LTM-N025A-2, B-2)

2. Reactor Vessel Pressure - liigh NA M R (B21-PS-N045A,B,C,D)

Y v.

Y I

m

", (a) The transm'.Lter channel check is satisfied by the trip unit channel y check. A =+parate transmitter chec's is not required.

N z (b) Transmitters are exempted from the monthly channel calibration.

REACTOR COOLANT SYSTEM 3/4.4.4 CHEMISTRY LIMITING CONDITION FOR OPERATION 3.4.4 The chemistry of the reactor coolant system shall be maintained within the limits specified in Table 3.4.4-1.

APPLICABILITY: At all times.

ACTION:

a. In OPERATIONAL CONDITION 1, 2, and 3: l

1. With the conductivity or chloride concentration exceeding the limits specified in Table 3.4.4-1, but less than 10 u mho/cm at 25*C and less than 0.5 ppm, respectively, operation may continue for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and this condition need not be reported to the Commission per Specification 6.9.1.12 provided that l operation under these conditions shall not exceed 336 hours0.00389 days 0.0933 hours 5.555556e-4 weeks 1.27848e-4 months per year. The provisions of Specification 3.0.4 are not applicable.

2. With the conductivity or chloride concentration exceeding the limits specified in Table 3.4.4-1 for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during one continuous time interval or with the conductivity exceeding 10 paho/cm at 25*C or chloride exceeding 0.5 ppe, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUIDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. At all other times with the conductivity and/or chloride concentration of the reactor coolant in excess of the limit specified in Table 3.4.4-1, ' restore the conductivity and/or chloride concentration to within the limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

BRUNSWICK - UNIT 1 3/4 4-7 Anendment No.

1 - - . . - . . . - . _-. _ __-_

REACTOR COOLANT SYSTEM 3/4.4.5 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.5 The specific activity of the reactor coolant shall be limited to:

a. < 0.2 pCi/ gram DOSE EQUIVALENT I-131, and

b. $100/E pCi/ gram.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4. l ACTION:

a. In OPERATIONAL CONDITION 1, 2, and 3, with the specific activity of l the reactor coolant;

1. > 0.2 pCi/ gram DOSE ' EQUIVALENT I-131 but < 4.0 pCi/Eram, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months provided chat operation under these conditions shall not exceed 10 parcent of the unit's total yearly operating time. The provisions of Specification 3.0.4 are not applicable.

2. > 0.2 uCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or > 4.0 uCi/ gram, be in at least HOT SHUTDOWN with the main steam line isolation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3. > 100/ E pCi/ gram, be in at least HOT SHUTDOWN with the main s team line isolation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN witttia the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l b. In OPERATIONAL CONDITION 1, 2, 3, or 4, l I 1. With the specific activity of the primary coolant > 0.2 uC1/ gram DOSE EQUIVALENT I-131 or > 100/E UCi/ gram, perform the sampling '

and analysis requirements of Item 4b of Table 4.4.5-1 at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months until the specific activity of the primary coolant is restored to within its limits. A REPORTABLE OCCURRENCE report shall be prepared and submitted to the Commission pursuant to Specification 6.9.1.12. This report j shall contain the results of the specific activity analyses and the time duration when the specific activity coolant exceeded 0.2 pCi/ gram DOSE EQUIVALENT I-131 together with the below additional Information. ,

i l

BRUNSWICK - UNIT 1 3/4 4-10 knendment No.

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT STRUCTURAL INTEGRITY L EMITING COND_ITION FOR OPERATION 3.6.1.4 The structural integrity of the primary containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.4.

APPLI(%BILITY: CPERATIONAL CONDITIONS 1, 2, and 3.

I ACTION:

With the structural integrity of the primary containment not conforuing to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coclant System temperature above 212*F.

SURVEILLANCE REQUIREMENTS 4.6.1.4 The structural integrity of the primary containment shall be determined during the shutdown for each Type A containment leakage rate test by a visual inspection to the eccecsible interior and exterior surfaces of the containment and verifying no apparent changes in appearance of the surfaces or other abnormal degradation. Aay abnormal dagradation of the primary containment detected during the required inspections shall be reported to the Commission pursuant to Specification 6.9.1.12.

l i

~

I l

l l

I l

BRCNSWICK - UNIT 1 3/46-6 htandment No.

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 efflu-ents to UNRESTRICTED AREAS (see Figure 5.1.3-1) af ter dilution in the discharge canal shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved concentration shall be limited to 2 x 10~gr microcuries/ml. entrained noble gases, the 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 i concentration to within the above limits.

SURVEILLANCE REQUIREMENTS l

4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Tkble 4.11.1-1. If the stabilization pond or service water samples analyzed according to Table 4.11.1-1 indicate concentrations of any gamma-emitting radionuclides greater than 5x10~D pCi/ml (trigger level), then the liquid wastes exceeding the trigger level shall be sampled and analyzed according to the sampling and analysis program of Tkble 4.11.1-2 until such time a nuclide is less than 5x10~g the pCi/ml. sample concentration of each gamma-emitting 4.11.1.1.2 The results of 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.

NOTE: See Bases 3/4.11.1.1 BRUNSWICK - UNIT 1 3/4 11-1 Amendment No.

)

TABLE 4.11.1-1 RADIOACTIVE LIOUTn WASTE SAlfLING AND ANALYSIS PROGRAM Minimum Type of Iower Limit of Liquid Release Type Sampling Analysis Activity Detection (LLD)

Frequency Frequency Analysis (UCi/ml) (a)(e)

N A.I. Sample Tanks, P P Principal 5 x 10-7 Detergent Drain Each Each Batch Gamma Tank, and Salt Batch Emitters (8)

Water Release Tanks 1-131 1 x 10-6 (Batch Release)(h)

P Dissolved and 1 x 10-5 One M Entrained Batch /M Gases (gamma emitters)

2. Circulating P M Gross Alpha 1 x 10-7 Water Pit Each Batch Composite (c) Ib3 1 x 10-5 P Q Sr-89, Sr-90 5 x 10-8 Each Batch Composite (c) Fe-55 1 x 10-6 B. Stab P P Principal 5 x 10-7 Pondgzation Each Each c.amma Release Release Emitters (8)

D D During Du ring Periods g{ Periods g{

j Release Release i C. Service Water (d) W W Principal 5x10-7 I

(Potential During During Famma Continuous System System Emitters (8)

Releas e) Operation Operation BRUNSWICK - UNIT 1 3/4 11-2 Amendment No.

TABLE 4. I1.1-1 (Continued)

RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATION (a) The detectability limits for activity analysis are based on techni-cal feasibility limits and on the potential significance in the environment of the quantities released. For some nuclides, lower detection limits may be readily achievable; and when nuclides are measured below the stated limits, they should also be reported.

(b) When operational limitations preclude specific ganuna radionuclide analysis of each batch, gross radioactivity measurements shall be made to estimate the quantity and concentrations of radioactive material released in the batch; and a weekly sample composited from proportional aliquots from each batch released during the week shall be analyzed for principal gamma-emitting radionuclides.

(c) 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 repre-sentative of the liquids released.

(d) The stabilization pond and service water liquid release types represent potential release pathways and not actual release pathways. Surveillance of these pathways is intended to alert the -

plant to a potential problem; analysis for principal genua emitters should be sufficient to meet this intent. If analysis for principal gammaengttersindicatesaproblem(i.e.,exceedsthetriggerlevel of 5x10~ pCi/ml), then completa sampling and analyses shall be

! performed as per Table 4.11.1-2.

(e) The lower limit of detectability (LLD) is the smallest concentration of a radioactive material in an unknown sample that will be detected eith a 95% probability with a 5% probability of falsely concluding that a blank observation represents a "real" signal.

j For a particular measurement system, which may include radiochemical

! separation:

4.66 o b LLD =

E.V.2.22 x 10 .Y.exp(-lg t,)

! where:

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

o b

=

(N/t b) 2

= standard deviation of background (cpm) i BRUNSWICK - UNIT 1 3/4 11-3 Amendment No.

. . _ - . ._ .- _ __ _ _ _ _ . _ _ _ ~ _ - . . _ _ _ _ . . _ ._._____ _ -

[

TABLE 4.11.1-1 (Continued)

/ j RADI0' ACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATION N = background count rate (cpm) t b

= time background counted for (min)

E = counting efficiency, as counts per disintegration V = volume or masc of sample 2.22 x 10 6 = conversion factor (dpe/ microcurie)

Y = fractional radiochemical yield A

g

= radioactive decay constant of ith nuclide (sec-1) e, = elapsed time between sample collection and counting (sec) should be used in the

' Typical values calculation. It of E, V, be should Y, recogn and t,ized that the LLD is defined as an "a priori" (before the fact) limit representing the capability of a measurement system and not as an ",a posteriori" (af ter the fact) limit for a particular measurement.

(f) The stabilization pond is typically released over a several-day period. The pond is to be sampled and analyzed prior to consnencing release. When composite sampling instrumentation becomes available and is OPERABLE, dall'y grab sampling of the stabilization pond effluent will not be required during release and the composite l

sample will be analyzed on a weekly basis.

(g) The principal gamma emitters for which the LLD specifications apply

! exclusively are the following radionuclides: ~

Mn-54, Fe-59, Co-58, i Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list

f does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report parsuant to Specification 6.9.1.8.

(h) A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be iso-lated and then thoroughly mixed to assure representative sampling.

Once fully operational, the salt water tanks will be included as indicated in Table 4.11.1-1.

BRUNSUICK - UNIT 1 3/4 11-4 Amendment No.

l l

l TABLE 4.11.1-2 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRMi FOR POTENTIAL RELEASE PATHWAYS WHICH HAVE EXCEEDED TRIGGER LEVELS Minimum Type of lower Limit of Sampling Analysis Activity Detection (LLD)

Liquid Release Type Frequency Frequency Analysis (uCi/ml)(a)(e)

A. Stabilization P P Principal 5 x 10-7(D)

Pond Each Each Gamma Release Release Baitters(8)

D D I-131 1 x 10-6 During During Periods Periods Release ({ Release {

P Dissolved 1 x 10-5 One M and Release /M Entrained Gases (Gamma Baitters)

P Each M Gross Alpha 1 x 10-7 Release Composite (c) H-3 1 x 10-5 P

Each Q Sr-89, Sr-90 5 x 10-8 Release Composite (c) Fe-55 1 x 10-6 B. Service Water D(d) W Principal (Continupus Composite (c) Gamma 5 x 10-7(b)

Release)(hl Buitters( 8) 1-131 1 x 10-6 M M Dissolved Grab and 1 x 10-5 Sample Entrained Gases (Gamma Emitters)

D(d) M Gross Alpha 1 x 10-7 Composite (c) H-3 1 x 10-5 D(d) O Sr-89, Sr-90 5 x 10-8 l Composite (c) Fe-55 1 x 10-6 i

i BRUNSWICK - UNIT 1 3/4 11-5 Amendment No.

TABLE 4.11.1-2 (Continued)

RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM FOR POTENTIAL RELEASE PATetWAYS WHICH HAVE EXCEEDED TRIGGER LEVELS TABLE NOTATION (a) The detectability limits for activity analysis are based on technical feasibility limits and on the potential significance in the environment of the quantities released. For some nuclides, lower detection limits may be readily achievable; and when nuclides are measured below the stated limits, they should clso be reported.

(b) When operational limitations preclude specific gamma radionuclide analysis of each batch, gross radioactivity measurements shall be made to estimate the quantity and concentrations of radioactive ,

material released in the batch; and a weekly sample composited from proportional aliquots from each batch released during the week shall be analyzed for principal gamma-emitting radionuclides.

(c) 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 repre-sentative of the liquids released.

(d) Until such time as continuous proportional composite samplers are installed on the service water discharge line, daily grab sampling o.

the service water effluent will be required for use in making up the composite.

(e) The lower limit of detectability (LLD) is the smallest concentration of a radioactive material in an unknown sample that will be detected with a 95% probability with a 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

4.66a b LLD =

E+ V. 2.22 x 10 .Y. exp(-lg t,)

Where:

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

(8/'b)/2

= standard deviation of background (cpm)

N = background count rate (cpm) t b

= time background counted for (min) r i BRUNSWICK - UNIT 1 3/4 11-6 Anendment No.

t

,, - --- -..-,--,n_, -

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

TABLE 4.11.1-2 (Continued)

RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM F0k POTENTIAL RELEASE PATHWAYS WHICH HAVE EXCEEDED TRIGGER LEVELS i

TABLE NOTATION l

t E = counting efficiency, as counts per disintegration V = volume or mass of sample 2.22 x 10 6 = conversion factor (dpm/ microcurie) i Y = fractional radiochemical yield A

g

= radioactive decay constant of ich nuclide (sec-1) tg = elapsed time between sample collection and counting (sec)

Typical values of E, V, Y, and t, should be used in the calculation. It should be recognized that the LLD is defined as an "a priori" (before the fact) limit representing the capability of a i measurement system and not as an "a posteriori" (af ter the fact)

! limit for a particular measurement.

(f) The stabilization pond is typically released over a several-day period. The pond is to be sampled and analyzed prior to conumencing release. When composite sampling instrumentation becomes available and is OPERABLE, daily grab sampling of the stabilization pond effluent will not be required during release and the composite sample will be analyzed on a weekly basis.

l (g) . The principal gamma emitters for which the LLD specifications apply exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, l Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list l does not mean that only these nuclides are to be considered. Other I

gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.8.

(h) A continuous release is the discharge of liquid waste of a nondiscrete volunc, e.g., from a volume of a system that has an input flow during the continuous release.

i l

, BRUNSWICK - UNIT 1 3/4 11-7 Amendment No.

RADIOACTIVE EFFLUENTS DOSE - LIOUID EFFLUENTS LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radio-active materials in liquid effluents released to UNRESTRICTED AREAS (see Figure 5.1.3-1) shall be limited:

a. During any calendar quarter to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ, and

b. During any calendar year to less than or equal to 6 mrem to the total body and to less than or equal to 20 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated doses f rom 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 action 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 f rom liquid efflu-ents for the current calendar quarter and the current calendar year shall be determined in accordance with the ODCM at least once per 31 days.

NOTE: See Bases 3/4.11.1.2 i

i BRUNSWICK - UNIT 1 3/4 11-8 Amsndment No.

RADIOACTIVE EFFLUENTS LIQUID RADWASTE TREATMENT LIMITING CONDITION FOR OPERATIOF 3.11.1.3 The liquid radwaste treatment system shall be used to rebace the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent f rom the site to UNRESTRICTED AREAS (see Figure 5.1.3-1) would exceed 0.12 mrem to the total body or 0.4 mrem to any 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-system, and reason for the inoperability.

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

3. Summary of description of action (s) taken to prevent a recur-rence.

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

( SURVEILLANCE REQUIREMENTS 4.11.1.3 Doses due to liquiu releases f rom the site to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the ODCM.

(

NOTE: See Bases 3/4.11.1.3 i

BRUNSWIOC - UNIT 1 3/4 11-9 Amendment No.

I l

RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS Appropriate alternatives to the ACTIONS and Surveillance Requirements below can be accepted if they provide reasonable assurance that in the event of an uncontrolled release of the tanks' content, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Thble II, Colnen 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.

I LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material suspended in solution in each of the following unprotected outdoor tanks shall be limited to less than or equal to the activity indicated below, excluding tritium and dissolved or entrained gases.

OUTSIDE TANK CURIE LIMIT

a. Condensate Storage Tank 10 Ci

b. Outside Temporary Tank 10 Ci APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, without delay suspend all addition of radioactive material to the tank, uithin 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank's contents to within the limit, and describe the evente 1Sading to this condition in the next Semiannual Radioactive Ef fluent Release Report.

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

SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the tanks listed shall be determincd to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when

! radioactive materials are being added to the tank.

l NOTE: See Bases 3.4.11.1.4 F

BRUNSWICK - UNIT 1 3/4 11-10 Anendnent No.

i l

RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site to areas et an1 bejond the SITE BOUNDARY (see Fig-ure 5.1.3-1) shall be limited to the following:

a. For noble gases: Imss 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, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrems/yr to any 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 to be within the abo,ve limits in accordance with the methodology as described in the ODCM.

4.11.2.1.2 The dose rate due to iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous ef fluents shall be determined to be within the above limits in accordance with the methodology as described in the_0DCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11.2-1.

NOTE: See Bases 3/4.11.2.1 BRUNSWICK - UNIT 1 3/4 11-11 Amendment No.

TABLE'4.I1.2-1 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM n

Minimum Lower Limit of 8

Sampling Analysis Type of Detection (LLD)(a)

E Gaseous Release Type Frequency frequency Activity Analysis ( p C1/ml)

U

- A. Drywell Purge l' P Principal Gamma 1 x 10-4 Each rarge Each Purge Emmitters Grab f,ampica

-0 B. Environmental Release M(C)(d) M(c) Principal Gamma 1 x 10 Points - Main Stack, Grab Sample Emmitters (b)

Reactor Building Vent, Turbine Building Vent. H-3 1 x 10-6 w Hot Shop (h) s Continuous (*) W(f)(8) 7 Charcoal I-131 1 x 10-12 C Sample Continuous (e) g(f)(g) 1 x 10

-II Principle Gamma (b)

Emmitters Particulate Sample (I-131, others)

Continuous (e) y Composite Gross Alpha 1 x 10-11 Particulate Sample Continuous (*) Q

$ Composite Sr-89, Sr-90 1 x 10-11 E Particulat:

@ Sample Continuous (*) Noble Gas Noble Cases,

.$ Monitor Cross Beta or Gamma 1 x 10-6 f

7 a

TABLE 4.11.2-1 (Continued)

RADICACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATION (a) The lower limit of detectability (LLD) is the smallest concentration of a radioactive material in an unknown sample that will be detected with a 95% probability with a 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

4.66e b LLD =

E* V 2.22 x 10 .Y. exp(-A t,)

Where:

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

=

(N/tb )2

= standard deviation of background (cpa)

N = background count rate (cpm) t = time background counted for (min) b E = counting efficiency, as counts per disintegration V = volume or mass of sample 2.22 x 10 6 = conversion factor (dpm/ microcurie)

Y = fractional radiochemical yield A

g

= radioactive decay constant of ith nuclide (sec-I)

I t,

= elapsed time between sample collection and counting (sec)

Typical values of E, V, Y, and t, should be used in the calculation. It should be recognized that the LLD is defined as an "a priori" (before the fact) limit representing the capability of a measurement system and not as an "a posteriori" (after the fact) limit for a parilcular 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, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, 0o-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for i

BRUNSWICK - UNIT i 3/4 11-13 Amendment Nc.

1 TABLE 4.11.2-1 (Continued)

, RADIOACTIVE GASEOUS WASTE SAMPLING AND /JIALYSIS PROGRAM TABLE NOTATION particulate emissions. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to ,

Specification 6.9.1.8.

(c) With a THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within one hour, or following shutdown or start-up, sampling and analyses shall also be performed unless (1) analysis shows that the DOSE EQUIVA-LENT I-131 concentration in the primary coolant has not '.ncreased more than a factor of 3; and (2) the main condenser air ejector noble gas activity monitor shows that activity has not increased by more than a factor of 3.

(d) If during refueling, water exceeds 2 x 10 ghe tritiumtritium T Ci/m1, concentration in theshall grab samples spentbe fuel pool taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area whenever spent fuel is in the spent fuel pool. Spent fuel pool water will be sampled at least once per 7 days during refueling.

(e)' The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate enicula-tion made in accordance with Specifications 3.11.2.1, 3.11.2.2, and 3.11.2.3.

(f) Sample cartridges / filters 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 af ter changing (or af ter removal from sampler).

(g) Sampling shall be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, start-up, or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 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 does not apply if (1) analysis shows that the !

DOSE EQUIVALENT I-131 concentration in the prime.ry coolant has not increased more than a factor of 3; and (2) the main condenser air eje ctar noble gas eonitor shows that activity has not increased more than a factor of 3. This footnote doec not apply to the Hot Shop environmental release point.

(h) Monthly grab samples to be analyzed for principal gamma emitters and tritium are not applicable for the Hot Shop environmental release point. In addition, the Hot Shop release point does not have a continuous noble gas monitor and, therefore, the noble gas activity

! analysis requirements of Table 4.11.2-1 are not applicable.

BRUNSWICK - UNIT 1 3/4 11-14 Amendment No.

___._____i,____..__._.____.._._____.,_,_.__,,__,___ _ _.___

l l

RADIOACTIVE EFFLUENTS DOSE - NOBLE GASES LIMITING CC:lLITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1.3-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 10 mrad for !

gamma radiation and less than or equal to 20 mrad for beta radiation;

b. During any calendar year: Less than or equal to 20 mrad for gamma radiation and less than or equal to 40 mrad for beta radiation.

l APPLICABILITY: At all times.

ACTIONS:

a. With the calculated air dose f rom radioactive noble gases in gaseous ;

effluents exceeding any of the abcve limits, in lieu of a Licensee i Event Report, prepare and submit to the Commission within 30 days, !

pursuant _ to Specification 6.9.2, a Special Report that identifies I the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases, and the proposed corrective actions to be taken to assure that subsequent releases j 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 !

1 l

l 4.11.2.2 Dose Calculations - Cumulative dose contributions for noble gases l for the current calendar quarter and current calendar year shall be determined i

in accordance with the ODCM at least once per 31 days.

I NOTE: See Bases 3/4.11.2.2 l

l 1

1 l

BRUNSWICK - UNIT 1 3/4 11-15 Amendment No.

l l

RADIOACTIVE EFFLUENTS DOSE - 10 DINE-131, IODINE 133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1.3-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 15 mrems to any organ; and

b. During any calendar year: Less than or equal to 30 mrems to any organ.

APPLICABILITYs At all times.

ACTION:

a. With the calculated dose f rom the release of iodine-131, iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous ef fluents 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 a )plicable.

SURVEILLANCE REQUIREMENTS 4.11.2.3 Dose Calculations - Cumulative dose contributions for the current calendar quarter and current calendar year for iodine-131, iodine-133, tritium, and radionuclides in particulate form with half-lives greater titan 8 days shall be determined in accordance with the ODCM at least once per 31 days.

NOTE: See Bases 3/4.11.2.3 BRUNSWICK - UNIT 1 3/4 11-16 Amendment No.

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

RADIOACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.11.2.4 The CASE 0US RADWASTE TREATMENT SYSTEM shall be in operation.

APPLICABILITY: Whenever the main condenser air ejector (evacuation) system is in operation.

ACTION:

(

a. With ganeous radwasta f rom the main condenser air ejector system

,being discharged without treatment for more than 7 days, 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. Identification of the inoperable equipment or subsystems and the reason for 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 The readings of the relevant instruments shall be checked at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the main condenser air ejector is in use to ensure that the GASEOUS RADWASTE TREATMENT SYSTEM is functioning.

NOTE: See Bases 3/4.11.2.4 BRUNSWICK - UNIT 1 3/4 11-17 Amendment No.

RADIOACTIVE EFFLUENTS VENTILATION EXHAUST TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.11.2.5 The VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases, from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1.3-1), would exceed 0.6 mrem to any organ over 31 days.

APPLICABILITY: At all times other than when the 7ENTILATION EXHAUST TREATMENT SYSTEM is undergoiag routine maintenance.

ACTION:

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

1. Identification of any inoperable equipment or subsystems and the reason for ths 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 REOUIREMENTS f 4.11.2.5 Doses due to gaseous releases f rom the site shall be projected at

[ 1 east once per 31 days, in accordance with the ODCM, when the VENTILATION EXHAUST TREATMENT SYSTEM ie not in use.

l l

l NOTE: See Bases 3/4.11.2.5 l

BRUNSWICK - UNIT 1 3/4 11-18 Amendment No.

RADIOACTIVE EFFL_UENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.6 The concentration of hydrogen in the main condenser offgas treatment system shall be limited to less than or equal to 4% by volume.

APPLICABILITY: Whenever the main condenser air ejector system is in operation.

ACTION:

a. With the concentration of hydrogen in the main condenser offgas treatacnt system exceeding the limit, restore the concentration to within the limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ,

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

SURVEILLANCE REQUIREMENTS 4.11.2.6 The concentration of hydrogen in the main condenser offgas treatment system shall be determined to be within the above limit by continuously monitoring the waste gases in the main condenser offgas treatment system with the hydrogen monitors required OPERABLE by Table 3.3.5.9-1 of Specification 3.3.5.9.

NOTE: See Bases 3/4.11.2.6 BRUNSWICK - UNIT 1 3/4 11-19 Atendment No.

l

RADIOACTIVE EFFLUENTS MAIN CONDENSER AIR EJECTOR RADIOACTIVITY RELEASE RATE LIMITING CONDITION FOR OPERATION 3.11.2.7 The release race of the sum of the activities f rom the ncble gases measured at the main condenser air ejector shall be limited to less then or equal t'o 243,600 microcuries/second (the Kr-85a, 87, 88 and Xe-133,135,138 contribution af ter 30 minutes decay).

APPLICABILITY: During operation of the main condenser air ejector.

ACTION:

With the release rate of the sum of the activities f rom the noble gases at the main condenser air ejector exceeding the above limit, restore the gross radioactivity rate to within its limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.11,2.7.1 The radioactivity rate of noble gases at (near) the outlet of the main condenser air ejector shall be continuously monitored in accordance with Specification 3,11.2.1.

4.11.2.7.2 The release rate of the sum of the activities f rom the noble gases from the main condenser air ejector shall be determined to be within the above limit at the following frequencies by performing an isotopic analysis of a representative sample of gases taken at the discharge (prior to dilution and/or discharge) of the main condenser air ejector:

a. At least once per 31 days, or

b. Within 31 days f ollowing each refueling / maintenance outage, and

c. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following an increase, as indicated by the Condenser Air Ejector Noble Gas Activity Monitor, of greater than 50%, af ter

, factoring out increases due to changes in THERMAL POWER level, in the nominal steady state fission gas release f rom the primary coolant.

NOTE: See Bases 3/4.11.2.7 ,

BRUNSWICK - UNIT 1 3/4 11-20 Amendment No.

l RADIOACTIVE EFFLUENTS DRWELL VENTING OR PURGING LIMITING CONDITION FOR OPERATION 3.11.2.8 The drywell shall be purged to the environment at a rate in conformance with Specification 3.11.2.1.

APPLICABILITY: Whenever the drywell is vented or purged.

ACTION:

a. With the requirements of the above specification not satisfied, suspend all VENTING or PURGING of the drywell.

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

SURVEILLANCE REQUIREMENTS 4.11.2.8 A sample analysis, as defined in Table 4.11.2-1, shall be performed prior to each drywell PURGE.

NOTE: See Bases 3/4.11.2.8 BRUNSWICK - UNIT 1 3/4 11-21 Amendment No.

RADIOACTIVE EFFLUENTS 3/4.11.3 SOLID RADIOACTIVE WASTE LIMITING CONDITION FOR OPERATION 3.11.3 The solid radwaste system shall be used in accordance with a PROCESS CONTROL PROGRAM to process wet radioactive wastes to meet shipping and burial ground requirements.

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 f rom the site.

b. The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1.14.b are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.3 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, and sodium sulfate solutions).

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

SOLIDIFICATION of the batch may then be resumed using the alterna-tive SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM.

b. If the initial test specimen f rom a batch of waste f ails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection of testing of representative est specimens f rom 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 0.14, to assure SOLIDIFICATION of subsequent batches of waste.

NOTE: See Bases 3/4.11.3 BRUNSWICK - UNIT 1 3/4 11-22 Amendment No.

RADIOACT NE EFFLUENTS 3/4.11.4 TOTAL DOSE (40 CFR PART 190)

LIMITING CONDITION FOR OPERATION 3.11.4 The annual (calendar year) dose or dose commitment to any MENBER OF THE PUBLIC, due to releases of radioactivity and radiation from uranium fuel cycle sources shall be limited to leas than or equal to 25 mrems to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrems).

APPLICABILITY: At all times.

ACTION:

a. With the calculated doses f rom the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Speci-fications 3.11.1.2.a, 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, calculations should be made which, in addition to doses due to effluents, include direct radiation contributions from the reactor units and from outside storage tanks 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 subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR Part 20.405c, shall include an analysis that estimates the radiation exposure (dose) to a liEMBER OF THE PUBLIC from uranium fuel cycle sources , '.acluding all ef fluent 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 Staf f action on the request is complete.

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

BRUNSWICK - UNIT 1 3/4 11-23 Asendment No.

RADIOACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE (40 CFR PART 190)

/

SURVEILLANCE REOUREMENTS

~ .

./

4.11.4.1 Dose Calculations Cumulative dose contributions from liquid and gas-eous 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.

4.11.4.2 Cumulative dose contributions from direct radiation from the reactor units and from radwaste storage tanks shall be determined in accordance with the ODCM. This requirement is applicable only under conditions set forth in Specification 3.11.4.a.

NOTES: See Bases 3/4.11.4 BRUNSWICK - UNIT 1 3/4 11-24 Amendment No.

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM l

LIMITING CONDITION FOR OPERATION :

3.12.1 The radiological environmental monitoring program shall be conducted l as specified in Thble 3.12.1-1.

APPLICABILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12.1-1, in lieu of a Licensee Event Report, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Rep 1rt required by Specifi-cation 6.9.1.6, 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 specific location exceeding the reporting levels of Table 3.12.1-2 when averaged over any calen-dar quarter, in lieu of a Ilcensee 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 action to be taken to reduce radioactive effluents so that the potential annual dose to a MEMBER OF THE PUBLIC is less than the calendar year limits of Specifica-tions 3.11.1.2, 3.11.2.2, and 3.11.2.3. When more than one of the radionuclides in Table 3.12.1-2 are detected in the sampling medium, this report shall be submitted if:

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

reporting level (2) i 1

When radionuclides other than those in Table 3.12.1-2 are detected and are the result of plant ef fluents, this report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits of Specifications

' 3.11.1.2, 3.11.2.2, and 3.11.2.3. The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operatira Report.

c. With milk or fresh leafy vegetables unavailable from one or more of the sample locations required by Table 3.12.1-1, identify locations for obtaining replacement samples and add them to the radiological BRUNSUICK - UNIT 1 3/4 12-1 knendment No.

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RADIOLOGICAL ENVIRON" ENTAL MONITORING LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued) environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the nonitoring program and ODCM. In lieu of a Licensee Event Report and pursuant to Specification 6.9.1.8, identify the cause of unavailability of samples; and identify the new location (s) for obtaining replacement samples in the next Seniannual Radioactive Ef fluent 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 9 UIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12.1-1 from the specific locations given in the table and figure (s) in the ODCM and shall be analyzed pursuant to the requirements of Table 3.12.1-1 and the detection capabilities required by Table 4.12.1 '..

NOTE: See Bases 3/4.12.1 l

BRUNSWICK - UNIT 1 3/4 12-2 Amendment No.

I

n' TABLE 3.12.1-1 i

y RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM M

M I

@ Number Of Samples M Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations ) Collection Frequency Of Analysis g.

1. DIRECT RADIATION (b) 40 Locations. At each Q Gamma Dosh - Q location with 2 or more dosin-eters or one or more instru-ments for continuously measur-ing and recording dose rate, placed as follows:

$d An inner ring of stations,

' with at least one in each

[; meteorological sector in the

& general area of the SITE BOUNDARY as is reasonably accessible and practical; An outer ring of stations, with at least one in each meteorological sector at distances of 8 km or greater f rom the site as is reasonably accessible and practical; and y The balance of stations to be

@ placed in special interest A- areas such as population

!" centers, nearby residences, schools, and in at least one

@ or two areas to serve as control stations.

a _ _ _ _ - - -- ___ _a

en W TABLE 3.12.1-1 (Continued)

E h RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM a

Number Of Samples S Exposure Pathway and Sampling and Type and F equqncy H and/or Sample Sample locations Collection Frequency Of Ana ysis

2. AIRBORNE - 5 locations, as follows: Continuous sampler opera- Radioiodine Cannister:

Radioiodine and x tion with sample col- I-131 analysis - W Particulate 3 samples from different lection weekly or as

.. sectors as close to the required by dust loading, Particulate sampler:

SITE BOUNDARY as is whichever is more fre- Gross beta radioactivity ,.

reasonably accessible, quent. analysis 110 wing filter w one of which being at change; 3 the highest calculated Gamma isotopic analysis (*)

g annual average ground of composite (by

';3 level D/Q; location) - Q s~

l sample from the vicinity of a nearby community; and 1 sample from a control location, as for example greater than 15 km dis-tantandinalesspryc) valent wind direction

,k 3. WATERBORNE 2 locations, as follows: Composite (8) sample Gamma Is pic R. a. Surface (f). collection - M Analysis -M

,D, I sample upstream f Tritium Analysis - Q z 1 sample downstream

, TABLE 3.12.1-1 (Continued) m

$ RADIOLOGICAL ENVIRON'IEKfAL MONITORING PROGRAM a:

h Number Of Samples

' Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample locations (a) Collection Frequency of Analysis E

E 3. WATERBORNE (Continued) e

b. Sediment from I location with a sample SA shoreline taken from a downstream GammaIsgpic Analysis - SA area with existing or po-tential recreational value

4. INGESTION

a. Milk 4 locations as follows: With animals on pasture - SM Gammaisgpic analysis and I-131 Samples from milking analysis - SM (when g animals at 3 locations At other times - M animals are on pasture) within 8 km distance h

u having the highest dose At other times - M potential. gen available)

I sample from allking animals at a control location greater than '

15 km distance from the site and in a less prevalent wind direction f

g

b. Fish and Invertebrates 4 locations as follows: When in season - SA Gamma is analysis pic on edible g 3 samples of commercially portions -

SA g and recreationally impor-g tant species in the

,o vicinity of the plant dis-charge: one free swimming

E TABLE 3.12.1-1 (Continued)

E h RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM R

e E Number Of Samples O Exposure Pathway and Sarpling and Type and Frequency H and/or Sample Sample locations (a) Collection Frequency of Analysis

4. INGESTION (Continued)

b. Fish and species; one botton Invertebrates feeding species; and one (Continued) shellfish species.

I sample of a similarly w edible species from an 20 area not influenced by g plant discharge to serve 4 as a control sample.

c. Broadleaf 3 locations as follows: When available - M Csama is t pic Vegetation analysis and Samples of broadleaf vege- I-131 analysis - M tation grown in 2 sectors (when available) of historically higher D/Q values at the SITE BOUNDARY if milk sampling is not performed.

I g I sample of a similar g broadleaf vegetation grown g at a distance of greater n than 15 km from the site g: in a less prevalent wind direction if milk sampling is not performed.

t

TABLE 3.12.1-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM TABLE NOTATION (a) Specific parameters of distance and direction sector f rom the site, and additional description where pertinent, shall be provided for each and every sample location in Table 3.12.1-1 in a table and figure (s) in the ODCM. Deviations are permitted from the required sampling schedule if specimens are unobtainable due .o hazardous conditions, seasonal unavailability, malfunction o.' tutomatic 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 Radiological Environmental Operating Report pursuant to Specification 6.9.1.7. 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 programL In lieu of a License Event Report and pursuant to Specification 6.9.1.8, identify the cause of the unavailability of samples for that pathway 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).

) (b) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place l of, or in addition to, integrating dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. 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 feding.

(c) The purpose of this sample is to obtain background information. If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites that provide valid background data may be substituted.

(d) 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 l, 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.

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l BRUNSWICK - UNIT 1 3/4 12-7 Anendment No.

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TABLE 3.12.1-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM TABLE NOTATION (e) Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclidas that may be attributable to the effluents from the facility.

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

" Upstream" samples in an estuary must be taken f ar enough upstream to be beyond the plant influence. Salt water shall be sampled only when the receiving water is utilized for recreational activities.

(g) 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 representative of the liquid flow. Composite samples shall be collected with equipment that is capable of collecting an aliquot at time intervals that are short (e.g. , once per 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ) relative to compositing period (e.g. , monthly) in order to assure obtaining a representative sample.

(n) When less than three (3) milking animal locations are available for testing within an 8-km distance, sampling of broadleaf vegetation shall be performed as indicated in Table 3.12.1-1, 4.c, in lieu of milk sampling.

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BRUNSWICK - UNIT 1 3/4 12-8 Anendnent No.

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TABLE 3.12.1-2 (Continued)

E

@ REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES E

H REPORTING LEVELS S

I

@ Water Fish Milk Broadleaf Vegetation y Analysis (pC1/1) Airborne or GasesParticulg)te (pC1/m (pCi/kg. wet) (pC1/1) (pCi/kg, wet) e

~

H-3 3x 10 4 - - - -

Ma-54 1x 10 3 -

3 x 10 4 - -

Fe-59 4 x 10 2 -

1 x 10 4 - -

Co-58 4 x 10 2 -

3 x 10 4 - -

2 m Co-60 3 x 10 -

1 x 10 4 - -

s g Zn-65 3x 10 2 -

2 x 10 4 - -

w b Zr-Nb-95 4 x 10 2 - - - -

I-131 2 0.9 -

3 1 x 10 2 3

Cs-134 30 10 1 x 10 60 1 x 10 3 Cs-137 50 20 2 x 10 3 70 2 x 10 3 Ba-La-140 2 x 10 2 - -

3 x 10 2 _

3 E

n

g N s

as 5'

TABLE 4.12.1-1 E s.

x Q DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS (a) i %

h H

LOWER LIMIT OF DETECTION (LLD)(b)

Pd Airborne Broadleaf Water Particulate or Fish Milk Vegetation Sediment Analysis (pCi/1) Cases (pCi/m3 ) (pCi/kg, wet) (pCi/1) (pC1/kg, wet) (pCi/kg, dry) gross beta 4 0.01 - - - -

11 - 3 3000 - - - - -

w Mn-54 15 -

130 - - - -

U

,L Fe-59 30 -

260 - - -

o Co-58, 60 15 -

130 - - -

Zn-65 30 -

260 - - -

Zr-Nb-95 15 - - - - -

I-131 1(C) 0.07 -

1 60 -

Cs-134 15 0.05 130 15 60 150 I

$ Cs-137 18 0.06 150 18 80 180 a

$ Ba-La-140 15 - -

15 - -

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TABLE 4.12.1-1 (Continued)

DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS TABLE NOTATION (a) This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together wfth those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.7.

(b) The lower limit of detectability (ILD) is the smallest concentration of a radioactive material in an unknown sample that will be detected with a 95% probability with a 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a partienlar measurement system, which may include radiochemical j separation:

4.66a b E V 2.22 Y exp(-Ag t,)

Where:

LLD is the "a priori" lower limit of detection as defined above (as picoeuries per unit mass or volune) a b

=

(N/tb)/2

= standard deviation of background (cpm)

N = background count rate (cpm) t b

= time background counted for (min)

E = counting efficiency, as counts per disintegration V = volume or mass of sample 2.22 = conversion factor (dpm/pC1)

Y = fractional radiochemical yield A = radioactive decay constant of ith nuclide (sec~I)

= elapsed time between sample collection and counting t,

(sec)

BRUNSWICK - INIT 1 3/4 12-11 Amendment No.

TABLE 4.12.1-1 (Continued)

DETECTION CAPABILITIES FOR ENVIRONMENTAL SATLE ANALYSIS TABLE NOTATION Typical values of E, V, Y, and C, should be used in the calculation. It should be recognized that the LLD is defined as an "a priori" (before the f act) limit representing the capability of a measurement system and not as an "a posteriori" (af ter the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LT.Ds will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interferring nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Specifi ation 6.9.1.7.

(c) The LLD of gamma isotopic analysis may be used.

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BRUNSWICK - UNIT 1 3/4 12-12 Amendment No.

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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 the nearest resident, and the nearest sectorsof garden ofgreater the nearest milk than 50 m2animal, (500 f 2) t producing broadleaf vegetation. (For elevated releases as defined in Regulatory Guide 1.111, Revision 1, July 1977, the land use census shall also identify within a distance of 5 km (3 miles) the location in each of the 16 mgteorological sectors of all milk animals and all gardens of greater than 50 m producing broadleaf vegetation.)

Broadleaf vegetable sampling of at least 3 different kinds of vegetation may be perforced at the SITE BOUNDARY in each of 2 different direction cectors with the highest D/Qs in lieu of the garden census. Specifications for broad-leaf vegetation sampling in Table 3.12.1-1(4c) shall be followed, including analysis of control samples.

APPLICABILITY: At all times.

ACTION:

a. With a land use census identifying a location (s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, in lieu of a Licensee Event Report, identify the new location (s) in the next Semisnnual Radioactive Ef fluent Release Report, pursuant to Specification 6.9.1.8.

b. With a la'nd use census identifying a location (s) that yields a I calculated dose or dose commitment (via the same exposure pathway) l 20 percent greater than at a location f rom 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 central station location, having the lowest calculated dose or dose i

commitment (s) (via this same exposure pathway) may be deleted from l

this monitoring program af ter October 31 of the year in which this j land use census was conducted. In lieu of a Licensee Event Report and pursuant to Specification 6.9.1.8, identify the new location (s) in the next Semiannual Effluent Release Report; and also include in the report a revised figure (s) and table f or the ODCM reflecting the new location (s).

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

BRUNSWICK - UNIT 1 3/4 12-13 knendment No.

RADIOLOGICAL ENVIRONMENTAL MONTORING SURVEILLANCE REOUIREMENTS 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, acrial survey, or by consulting local agriculture authorities. The result of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.7.

NOTE: See Bases 3/4.12.2 l

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BRUNSWICK - UNIT 1 3/4 12-14 Amendment No.

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RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program that has been approved by the C)mmis-sion.

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 6.9.1.7.

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

SURVEILLANCE REQUIREMENTS 4.12.3 The Interlaboratory Comparison Program shall be described in the ODCM. A stamnary of the results, obtained as part of the above required Interlaboratory Comparison Program, shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.7. >

NOTE: See Bases 3/4.12.3 l

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l BRUNSWICK - UNIT 1 3/4 12-15 Amendment No.

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l INSTRUMENTATION f

RASES l I

MONITORING INSTRUMENTATION (Continued) 3/4.3.5.6 CHLORIDE INTRUSION MONITORS The chloride intrusion monitors provide adquate warning of any leakage in the condenser or hotwell so that actions can be taken to mitigate the consequences of such intrusion in the reactor coolant system. With only a minimum number

of instruments available, increased sampling frequency provides adequate

information for the same purpose.

3/4.3.5.7 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate

, warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program.

In the event that a portion of the fire detection instrumentation is .

inoperable, increasing the frequency of fire patrols in the affected areas is raquired to provide" detection capability until the inoperable instrumentacion is restored to OPERARILITY.

3/4.3.5.8 RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITT 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 l

to assure the detection and control of leaks that, if not controlled, could potentially result in the transport of radioactive materials to UNRESTRICTED AREAS. "Without delay" implies that the operator, upon determining the limiting condition for operstion is being exceeded, takes. the next appropriate l action to comply with the specification.

l The initial CHANNEL CALIBRATION for the instruments associated with footnote (b) to Table 4.3.5.8-1 shall be performed using National Bureau of Standards traceable sources which will verify that the detector operates l properly over its intended energy range and measurement range. For

! ihatruments which were operational prior to this specification being implemented, previously established calibration procedures may be substituted

! for this requirement. Subsequent CHANNEL CALIBRATIONS will be performed using sources that have been related to the initial calibration in order to ensure that the detector is still operational, but the sources need not span the full ranges used in the initial CHANNEL CALIBRATION.

B 3/4 3-4 Amendment No.

' _. _.___ BRUNSWICK - UNIT 1. _ . _ ,... _ ,__ _ _ . _ _ _ __ _ _

l BASES 3/4.3.5.9 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION

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

The alarm / trip setpoints for these instruments shall be calculated in accordance with the 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 and 64 of Appendix A to 10 CFR Part 50.

The main condenser air ejector. monitoring instrumentation, the main condenser offgas treatment system monitor, and the explosive gas monitoring instrumentation shown in Table 3.3.5.9-1 are not considered effluent monitoring instrumentation in the same sense as the other instrumentation

. listed in the Table. Therefore, their alarm / trip setpoints are not necessarily set to ensure that the limits of Specification 3.11.2.1 are not exceeded.

The main condenser air ejector monitoring instrumentation channels are provided to monitor and control gross radioactivity removed from the main condenser. The alarm / trip setpoints for the main condenser air ejector monitor are set to ensure that the limits of Specification 3.11.2.7 are not exceeded. The alarm / trip setpoint for this monitor shall be calculated in accordance with NRC approved methods to provide reasonable assurance that the potential total body accident dose will not exceed a fraction of the limits specified in 10 CFR Part 100.

This specification also includes provisions for monitoring the concentrations of potentially explosive gas mixtures in tSe offgas treatment system (hydrogen monitors).

"Without delay" implies that the operator, upon determining the limiting condition for operation is being exceeded, takes the next appropriate action to comply with the specification. _

The initial CHANNEL CALIBRATION for the instruments associated with footnote (b) to Table 4.3.5.9-1 shall be performed using National Bureau of Standards traceable sources which will verify that the detector operates properly over its intended energy range and measurement range. For instruments which were operational prior to this specification being implemented, previously established calibration procedures may be substituted

for this requirement. Subsequent CHANNEL CALIBRATIONS will be performed using

! sources that have been related to the initial calibration in order to ensure

]

that the detector is still operational, but the sources need not span the full ranges used in the initial CHANNEL CALIBRATION.

l BRUNSWICK - UNIT 1 B 3/4 3-5 Amendment No.

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

INSTRUMENTATION BASES 3/4.3.6 ATWS RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The ATWS recirculation pump trip system has been added at the suggestion of ACRS as a meanc of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the plant to this postulated event falls within the envelope of study events given in General Electric Company Topf cal Report NED0-10349, dated March, 1971.

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BRUNSWICK - UNIT 1 B 3/4 3-6 Amendment No.

F

/

3/4.11 RADIOACTIVE EFFLUENTS p

/

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 af ter dilution in the discharge canal will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Oslumn 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will not 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 methoda described in International Cbamission 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 Manuals, RASL-300 (revised annually), Currie, L. A. "Ilmits for Qualitative Detection and Quantitative Determination - Application to Radio-chemistry

Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K. , " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-3 A-215 (June 1975).

"Without delay" implies that the operator, upon determining the limiting condition for operation is being exceeded, takes the next appropriate action to comply with the specification.

l Note that for batch releases, recirculation of at least two tank volumes shall i be considered adequate for thorough mixing.

The stabilization pond and service water liquid release types represent potential release pathways and not actual releae: pachways. Surveillance of these pathways is intended to alert the plant to a potential problem; analysis for principal gamma emitters should be sufficient to meet this intent. If analysis for principal gamma emitters indicates a problem (i.e. , exceeds the trigger level of 5x10-6 uC1/ml), then complete sampling and gnalyses shall be performed as per Table 4.11.1-2. The trigger level of 5x10 pCi/ml was chosen as being sufficient to provide reasonable assurance of accountability of all nuclides released based upon lower limits of detection and expected concentrations.

3/4.11.1.2 DOSE - LIQUID EFFLUENTS 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 BRUNSWICK - UNIT 1 B 3/4 11-1 Anendment No.

i j RADIOACTIVE EFFLUENTS BASES DOSES (Continued)

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 of 10 CFR

Part 50 to assure that releases of radioactive material in liquid effluents to l UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The dose calculations in the ODCM implement the requirements in Section III. A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual

exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to j be substantially underestimated. The equations specified in the ODCM for

, calculating the doses due to the actual release rates of radioactive materials I in liquid effluents will be consistent with the methodology provided in Reg 21 story Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 1 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.

l The dose or dose commitment to a MEMBER OF THE PUBLIC is based on the 10 CFR Part 50, Appendix I, guideline of:

a. 1.5 mrem to the total body and 5.0 arem to any organ during any calendar quarter, and j b. 3 mrem to the total body and 10 mrem to any organ during any calen-dar year, f rom radioactive material in liquid effluents f rom each reactor unit to UNRE-

! STRICTED AREAS. This specification is written for a two unit site.

! 3/4.11.1.3 LIQUID RADWASTE TREATMENT SYSTEM The requirement that appropriate portions of this system be used, when speci-fied, provides assurance that the releases of radioactive materials in liquid i effluents will be kept "as low as reasonably achievable." This specification

implements the requirements of 10 CFR Part 50.36a, General Design Criteria 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 app opriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II. A i of Appendix I,10 CFR Part 50, for liquid ef fluents.

Mechanical flitration as per system design is considered to be an appropriate component of the liquid radwaste treatment system.

l The requirements of 0.12 mrem total body or 0.4 mrem to any organ in a 31-day l period is based on two reactor units having a shared liquid radwaste treatment system.

BRUNSWICK - UNIT 1 B 3/4 11-2 haendment No.

RADIOACTIVE EFFLUENTS BASES 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 do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system with the exception of the auxiliary surge tank. The auxiliary surge tank is ' excluded f rom this specification because the tank and its associated piping are all Seismic Class I.

Since the condensate storage tanks have continuous influent and effluent, stratification should not occur. Samples taken from the operating condensate transfer pump (s) vent or drain shall be deemed representative of this system.

"Without delay

implies that the operatcr, upon determining the limiting

, condition for operation is being exceeded, takes the next appropriate action to comply with the specification.

j 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE This specification is provided to ensure that the dose rate at and beyond the SITE BOUNDARY from gaseous offluents from all units on the site will be within the annual dose rate limits of 10 CFR Part 20 for UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. 1hese limits provide reasonable i assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations j

cxceeding 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 trichin the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will be sufficiently low to compensate for any increase in the atmospheric diffusion f actor above that for the SITE BOUNDARY. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC 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 mress/ 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.

With regard to footnotes (c) and (g) of Table 4.11.2-1, (1) to determine whether the DOSE EQUIVALENT I-131 concentration in the primary coolant has increased by more than a f actor of 3, the iodine-131 analysis performed af ter the transient will be compared to the most recent routine analysis for DOSE EQUIVALENT I-131 concentration performed before the transient; and (2) to determine whether the main condenser air ejector noble gas monitor has increased by more than a f actor of 3, the activity indicated on the monitor's BRUNSWICK - UNIT 1 B 3/4 11-3 Amendment No.

i

RADIOACTIVE EFFLUENTS BASES DOSE RATE (Continued) chart recorder af ter the transient will be compared to the activity indicated on the recorder just before the transient occurred.

The required detection capabilities f or radioactive materials in gaseous waste 4 samples are tabulated in terms of the Lower Limits of Detection (LLDs).

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

l "Without delay" implies that the operator, upon determining the limiting l

condition for operation is being exceeded, takes the next appropriate action to comply with the specification.

3/4.11.2.2 DOSE-NOBLE GASES 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 i ACTION statements provide the required operating flexibility and, at the same time, implement the guides set f orth in Section IV.A of Appendix I, to assure that the releases of radioactive materials in gaseous offluents to UNRESTRICTED AREAE will be kept "as low as is reasonably achievable." 1he Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through the appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Ef fluents in Routine Releases f rom Light-Water-Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY will be based upon the historical annual average atmospheric conditions. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and l

1.111. The limits of this specification are twice the 10 CFR 50 Appendix I per reactor guidelinee because they are written f or a two unit site.

l l

l BRUNSWICK - UNIT 1 B 3/4 11-4 Amendment No.

l

r t

RADIOACTIVE EFFLliENTS BASES 3/4.11.2.3 DOSE - 10DINW131,10 DINE-133, TRITIIM, AND RADIONUCLIDES IN PARTICULATE FORM This specification is provided to implement the requirements of Section II.C, III.A, and IV.A of Appendix I,10 CFR Part 50. The Limiting Conditions for Operation are the guides set f orth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and, at the same time, implements the guides set f orth in Section IV. A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED l AREAS will be kept "as low as is reasonably achievable." The ODCM

( calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substaccially underestimated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are required to be consistent with the i methodology provided in Regulatory Guide 1.109, " Calculating of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evalu-ating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and i

Dispersion of Gaseous Ef fluents in Routine Releases f rom Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for deter-mining the actual doses based upon the historical average atmospheric condi-tions. The release rate specification for iodine-131, iodine-133, tritium, and radioactive material in particulate f orm with half-lives greater than 8 days are dependent on the a isting radionuclide pathways to man in the areas at and beyond the SITE BOUNDARY. The pathways which are examined in the development of these calculations are: (1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy vegetation

! with subsequent consumption by man, (3) deposition onto grassy areas where

, nilk animals and meat producing animals graze, with consumption of the milk and meat by man, and (4) deposition on the ground with subsequent exposure of

man. The limits of this specification are twice the 10 CFR 50 Appendix I per reactor guidelines because they are written for a two rnit site.

^

3/4.11.2.4 GASE0US RADWASTE TREATMENT SYSTEM This requirement provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The GASEOUS RADWASTE TREATMENT SYSTEM refers to the 30-minute offgae holdup line, stack filter house filtration, and the Augmented Of f-Gas-Treatment System.

i BRUNSWICK - UNIT 1 B 3/4 11-5 Amendment No.

RADIOACTIVE EFFLUENTS BASES 3/4.11.2.5 VENTILATION EXHAUST TREATMENT SYSTEM This requirement provides reasonable assurance that the releases of radio-i active materials in gaseous effluents will be kept "as low as reasonably -

achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The i specified limits governing the use of the systems were specified as a suitable f raction of the dose design objectives set forth in Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous offluents. At the Brunswick Steam Electric Plant, the only VENTILATION EXHAUST TREATMENT SYSTEMS shall be those installed for the Turbine Buildings' ventilation.

3/4.11.2.6 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treatment system is main-tained below the flammability limits of hydrogen. Maintaining the concentra-tion of hydrogen below the 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.

3/4.11.2.7 MAIN CONDENSER AIR EJECTOR RADIOACTIVITY RELEASE RATE Restricting the release rate of noble gases f rom the main condenser provides reasonable assurance that the total body exposure to an individual at or beyond the exclusion boundary will not exceed a small f raction of the limits of 10 CFR Part 100 in the event this affluent is inadvertently discharged directly to the environment without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50. 243,600 microcuries/second is equal to 100 microcuries/second/MWt for a rated thermal power of 2,436 MWt.

l l

l BRUNSWICK - UNIT 1 B 3/4 11-6 Amendment No.

. . .. . .~

RADIOACTIVE EFFLUFNTS BASES 3/4.11.2.8 DRYWELL VENTING OR PURGING This specification provides reasonable assurance that releases from drywell PURGING operations will not exceed the annual dose limits of 10 CFR Part 20 for INRESTRICTED AREAS.

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 PR0 CESS CONTROL PROGRAM may include, but are not limited to waste type ,r waste pH, waste / liquid / solidification agent / catalyst ratios, waste oil content, waste principal chemical constit-uents, mixing, and curing times.

l 3/4.11.4 TOTAL DOSE (40 CFR PART 190)

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 generated radioactive effluents and direct radiation exceed 25 areas to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mress. 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 describs a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PULLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed l that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel l cycle sources is negligible, with the exception that dose contributions from l 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 Specification 3/4.11. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

I BRUNSWICK - UNIT 1 B 3/4 11-7 Amendment No.

1

1 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.1 MONITORING PROGRAM The radiological environmental monitoring program required by this specifice-tion provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest poten-tial radiation exposures of MEMBERS OF THE PUBLIC resulting from 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 measursble concentrations of radioactive materials are not higher than expected on the basis of ef fluent measurements and the modeling of the environmental exposure pathways.

The required detection capabilities for environmental sample analyses are tabulated in terms of the Lower Limits of Detection (LLDs). The LLDs required by Table 4.12.1-1 are considered optimum for routine environmental measure-ments in industrial laboratories. It should be recognized that the LLD is defined as a, priori (before the fact) limit representing the capability of a measurement system and not as jL posteriori (af ter the f act) limit for a par-ticular measurement.

Detailed discussion of the LLD and other detection limits can be found in HASL Procedure Manual, RASL-300 (revised annually), Currie, L. A. , " Limits for Qualitative Detection and Quantitative Determination Application to Radio-chensitry" Anal. Chen 40, 586-93 (1968), and Hartwell, L. K. , " Detection Limits for Radionnalytical Counting Techniques," Atlantic Richfield Hanford Company Raport ARH-S A-215 (June 1975).

Groundwater is not monitored by this specification because plant liquid efflu-ents are not tapped as a source for drinking or irrigation purposes.

3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of area 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 the census. The best information from door-to-door surveys , aerial surveys, or consulting with local agricultural authorities shall be used.

This censu's satisfies the requirements of Section IV.B.3 of Appendix I go 10 CFR Pa'rt 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/yr) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine the minimum garden size, the following assumptions were made: (1) 20% of the garden was used f or growing broadleaf vegetatiop (i.e. , similar to lettuce and cabbage; and (2) a vegetation yield of 2 kg/m BRUNSWICK - UNIT 1 ,

B 3/4 12-1 knendment No.

RADIOLOGICAL ENVIRONWgAL MONITORING BASES 3/4.12.3 INTERLABORATORY C0!PARISON PROGRAM The requirement f or parcicipation in the 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 monitor-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.

l j BRUNSWICK - UNIT 1 B 3/4 12-2 Amendment No.

l

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

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

SITE BOUNDARY 5.1.3 The SITE BOUNDARY shall be as shown in Figure 5.1.3-1. For the purpose of effluent release calculations, the boundary for atmospheric releases is the SITE BOUNDARY and the boundary for liquid releases is the SITE BOUNDARY prior to dilution in the Atlantic Ocean.

5.2 CONTAINMENT CONFIGURATION 5.2.1 The PRIMARY CONTAINMENT is a steel-lined reinforced concrete structure composed of a series of vertical right cylinders and truncated cones which form a drywell. This drywell is attached to a supprossion chamber through a series of vents. The suppression chamber is a concrete steel-lined pressure vessel in the shape of a torus. The primary containment has a minimum free air volume of (288,000) cubic feet.

DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is designed and shall be maintained for:

a. Maximum internal pressure 62 psig.

! b. Maximum internal temperature: drywell 300*F.

i suppression chamber 200*F. .

l c. Maximum external pressure 2 psig.

1 5.3 REACTOR CORE l

FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 560 fuel assemblies, with each fuel assembly containing 63 fuel rods clad with Zirealoy 2. Each fuel rod shall have a nominal active fuel length of 146 inches for 8 x 8 fuel and 150 inches for 8 x 8R fuel and contain a maximum total weight of 3,355 grams of UO 2

. The initial core loading.

BRUNSWICK - UNIT 1 5-1 Amendment No.

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

ENCLOSURE 4 RADIOLOGICAL EFFLUEtTr TECHNICAL SPECIFICATIONS BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 2 REFERENCE NO. E3TSB16 i

INDEX DEFINITIONS SECTION 1.0 DEFINITIONS PAGE ACTION ........................................................... 1-1 AVERAGE PLANAR EXPOSURE ......................................... 1-1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE ....................... 1-1 CHANNEL CALIBRATION .............................................. 1-1 CHANNEL CHECK .................................................... 1-1 CHANNEL FUNCTIONAL TEST .......................................... 1-1 l CORE ALTERATION .................................................. 1-2 CRITICAL POWER RATIO ............................................. 1-2 DOSE EQUIVALENT I-131 ........................................... 1-2 E-AVERAGE DISINTEGRATION ENERGY .................................. 1-2 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME . . . . . . . . . . . . . . 1-2 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESFONSE TIME ....... 1-3 FRE QUENCY NOTATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 GASEOUS RADWASTE TREATMENT SYSTEM ............................... 1-3

{

IDENTIFIED LEAKAGE ............................................... 1-3 ISOLATION SYSTEM RESPONSE TIME ................................... 1-3 LIMITING CONTROL ROD PATTERN ..................................... 1-3 l

LINEAR HEAT GENERATION RATE ...................................... 1-4 LOGIC SYSTEM FUNCTIONAL TEST ..................................... 1-4 MAXIMUM FRACTION OF LIMITING POWER DENSITY ...................... 1-4 l

i MAXIMUM TOTAL PEAKING FACTOR ..................................... 1-4 MEMBER (S) 0F THE PUBLIC ......................................... 1-4, MINIMUM CRITICAL POWER RATIO ..................................... 1-4 ODYN OPTION A .................................................... 1-4 ODYN OPTION B .................................................... 1-4 0FFSITE DOSE CALCULATIONAL MANUAL (ODCM) ......................... 1-5 OPERABLE - OPERABILITY ........................................... 1-5 OPERATIONAL CONDITION ............................................ 1-5 BRUNSWICK - UNIT 2 I Amendment No.

4

i INDEX l DEFINITIONS SECTION 1.0 DEFINITIONS (Continued) PAGE PHYS ICS TESTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- 5 l PRESSURE BOUNDARY LEAKAGE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 PRIMARY CONTAINMENT INTEGRITY ................................... 1-5 PROCESS C ONTROL PROGRAM (PCP) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 PURGE - PURGING ........................................... .... 1-6 RATED THERMAL P OWER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 REACTOR PROTECTION SYSTEM RESPONSE TIME ................ ......... 1-6 REFERENCE LEVE L ZERO . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- 6

~

RE PORTABLE OCCURRENCE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-7 ROD D EN S ITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-7 SECONDARY CONTAINMENT INTEGRITY ................................. 1-7 S HUTDOWN MARGIN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-7 S ITE BOUNDARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- 7 S O LID IF ICAT ION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-7 SOURCE CHECK .................................................... 1-7 S P IRAL RELOAD . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- 8 SPIRAL UNLOAD .................................................... 1-8 S TAGGERED T EST B AS IS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-8 THERMAL POWER ................................................... 1-8 TOTAL PEAKING FACTOR ............................................ 1-8 UNIDENTIFIED LEAKAGE ............................................. 1-8 U NRESTRICTED AREA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- 8 VENTILATION EXHAUST TREATMENT SYSTEM ............................ 1-9 VENTING ......................................................... 1-9 FREQUENCY NOTATION, TABLE 1.1 ................................... 1-10

. OPERATIONAL CONDITIONS , TABLE 1. 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-11 1

BRUNSWICK - UNIT 2 II Amendment No.

/ INDEX

/

y LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION . . . . . . . . . . . . . . . . . . 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-9 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION..... 3/4 3-30 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION . . . . . . . . . . . . . . . 3/4 3-39 3/4.3.5 MONITORING INSTRUMENTATION Seismic Monitoring Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-44 Remote Shutdown 2nitoring Instrunentation . . . . . . . . . . . . . . . . . . 3/4 3-47 Po s t-accident Monitoring Ins trumentation . . . . . . . . . . . . . . . . . . . 3/4 3-50 Source Range Nnit ors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 3-5 3 Chlo rine De te ction Sy s tem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 3-5 4 Chloride Intrusion Nnitors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-5 5 Fire De tection Ins trumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-59 Radioactive Liquid Ef fluent 2nitoring Instrunentation. . . . . . . 3/4 3-62 Radioactive Gaseous Ef fluent Monitoring Instrumentation. .... 3/4 3-68 3/4.3.6 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS Recirculation Pump Trip System Instrumentation . . . . . . . . 3/4 3-78 l End-of-Cycle Recirculation Pump Trip System Instrumentation .......................................... 3/4 3-82 l 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Re circula t ion Lo op s . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 4- 1 Je t P ump s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4i-2 Idle Re circulation Lo op St artup . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-3 3.4.4.2 S AFETY/RELIE F VALVE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 4-4 l

BRUNSWICK - UNIT 2 V Amendment No.

INDEX LIMITING CONDITIONS FOR OPERATION AND SERVEILLANCE REOUIREMENTS SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Co nc e n t ra t ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11- 1 Dose - Liquid Ef fluent s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 11-8 Liquid Radwas te Treatment System . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-9 Liquid Holdup Tanks ......................................... 3/4 11-10 3/4.11.2 GASEOUS EFFLUENTS Do s e Ra t e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-11 Dose - Noble Gases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 1 1 - 1 5 Dose - Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form ....................................... 3/4 11-16 Ga seous Radwas t e Tr ea tment Sys t em . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-17 Ventilation Exhaust Treatment System ........................ 3/4 11-18 Ex plo s ive Ga s Mix ture . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-19 Main Condenser Air Ejector Radioactivity Release Rate ....... 3/4 11-20 Drywell Venting or Purging . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-21 3/4.11.3 SOLID RADIOACTIVE WASTE ..................................... 3/4 11-22 3/4.11.4 TOTAL DOSE (40 CFR PART 19 0 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 1 1-2 3 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING l 3/4.12.1 MONITORING PROGRAM ......................................... 3/4 12-1 j 3/4.12.2 LAN D USE CENS US . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 12-13 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM .......................... 3/4 12-15 BRONSWICK - UNIT 2 IXa Amendment No.

l l

INDEX l BASES SECTION PAGE 3/4.0 APPLICAB ILITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 0- 1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN ........................................ B 3/4 1-1 3/4.1.2 REACTIVITY ANOMALIES ................................... B 3/4 1-1 3/4.1.3 CONTROL RODS ........................................... B 3/4 1-1 3/4.1.4 CONTROL ROD PROGRAM CONTROLS . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 1-3 3/4.1.5 STANDBY-LIQUID CONTROL SYSTEM .......................... B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2-1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE ............. B 3/4 2-1 3/4.2.2 APRM SETPOINTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 2-3 3/4.2.3 MINIMUM CRITICAL POWER RATIO ........................... B 3/4 2-3 3/4.2.4 LINEAR HEAT GENERATION RATE . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 2-5 3/4.3 INSTRUMENTATION 3/4.3.1 . REACTOR PROTECTION SYSTEM INSTRUMENTATION .............. B 3/4 3-1

3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION .................... B 3/4 3-2 l 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMElfrATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 3-2 3/4.3.4 CONTROL ROD WITHDRAWAL BLACK INSTRUMENTATION ........... B 3/4 3-2 3/4.3.5 MONITORING INSTRUMENTATION ............................. B 3/4 3-2 3/4.3.6 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 3-5 l 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM ................................... B 3/4 4-1 3/4.4.2 S AFETY/ RELIEF VALVES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 4-1 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE ......................... B 3/4 4-1 BRUNSWICK - UNIT 2 X Amendment No.

- L-- - - _ .

INDEX RASES )

SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS j Co ncen t ra t ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 1 1 -1 Dose - Liquid Ef fluents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 11-1 Liquid Radwaste Treatment Sys tem . . . . . . . . . . . . . . . . . . . . . . . B 3/4 11-2 Liquid Holdup Tanks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 11-3 3/4.11.2 GASEOUS EFFLUENTS Do s e Ra t e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 1 1-3 Dose - Noble Gases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 1 1 -4 Dose - Iodine-131, Iodine-133, Tritium, and Radionuclides in Pa rticulate Form . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 11-4 Gaseous Radwas te Treatment Sys t em . . . . . . . . . . . . . . . . . . . . . . B 3/4 11-5 Ventilation Exhaust Treatment System . . . . . . . . . . . . . . . . . . . B 3 /4 11-5 Explos ive Ga s Mixture . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 11-5 Main Condenser Air Ejector Radioactivity Release Rate . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 11-6 Drywell Venting or Purging ............................. B 3/4 11-6 3/4.11.3 SOLID RADIOACTIVE WASTE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 11-6 3/4.11.4 TOTAL DOSE (40 CFR PART 190) . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 1 1 -6 3/4.12 RADIOACTIVE ENVIRONMENTAL MONIT0 KING l

l 3/4.12.1 MONITORING PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 12-1 3/4.12.2 LAND US E CENS US . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 12- 1 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM . . . . . . . . . . . . . . . . . . . . . B 3/4 12-2 l

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1 BRUNSWICK - UNIT 2 XIIA Amendment No.

INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE Exclusion Area .................................................. 5-1

[ Iow Population Zone . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 1 Site Boundary ................................................... 5-1 5.2 CONTAINMENT Configuration ................................................... 5-1 Design Temperature and Pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.3 REACTOR CORE Fuel Assemblies ................................................. 5-1 Control Rod Assemblies .......................................... 5-4 5.4 REACTOR COOLANT SYSTEM De sign Pre s sure and Tempe rature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-4 Volume .......................................................... 5-4 3.5 METEOROLOGICAL TOWER LOCATION ................................... 5-4 5.6 FUEL STORAGE Criticality ..................................................... 5-5 Drainage ........................................................ 5-5 Capacity ........................................................ 5-5 5.7 COMPONENT CYCLIC OR TRANS IENT LIMIT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5 l

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BRUNSWICK - UNIT 2 XIII Amendant No.

INDEX

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ADMINISTRATIVE CONTROLS -

SECTION PAGE 6.1 RESPONSIBILITY.............................................. 6-1 6.2 ORGANIZATION 6.2.1 0FFSITE..................................................... 6-1 6.2.2 FACILITY STAFF.............................................. 6-1 6.2.3 ONSITE NUCLEAR SAFETY GROUP Function.................................................... 6-8 Responsibilities............................................ 6-8 Authority................................................... 6-3 6.2.4 - ' TE CHNI CAL A DVI S0 R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-8 6.3 FACILI1. '" Q UALIFI CATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-8 6.4 TRAINING.................................................... 6-9 6.5 REVIEW AND AUDIT 6.5.1 INDEPENDENT SAFETY REVIEWERS................................ 6-9 6.5.2 SAFETY EVALUATIONS AND INDEPENDENT REVIEW CONTROL Safety Evaluations.......................................... 6-9 Procedures, Tests, and Experiments.......................... 6-9 Tempora ry Pro cedure Changes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-10 Modifications............................................... 6-11 Ope ra t ing Li ce ns e Change s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-11 6.5.3 PLANT NUCLEAR SAFETY COMMITTEE (PNSC)

Function.................................................... 6-12 a

Composition................................................. 6-13 l Al t e rn a t e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 13 Meeting Frequency........................................... 6-13 Quorum...................................................... 6-13 Activities.................................................. 6-15 Authority................................................... 6-14 Records..................................................... 6-15 BRUNSWICK - UNIT 2 XIV Amendment No.

INDEK ADMINISTRATIVE CONTROLS SECTION PAGE 6.5.4 CORPORATE NUCLEAR SAFETY SECTION Function.................................................... 6-15 Organization................................................ 6-15 Review...................................................... 6-16 Re c o r d s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 1 7 6.5.5 CORPORATE QUALITY ASSURANCE AUDIT PROGRAM Function.................................................... 6-18 Audits...................................................... 6-18 Re c o r d s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 1 9 Authority................................................... 6-19 Personne1................................................... 6-19 6.5.6 OUTSIDE AGENCY INSPECTION AND AUDIT PROGRAM. . . . . . . . . . . . . . . . . 6-19 6.6 REPO RTABLE OCCURRENCE ACTIO N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 0 6.7 S AFETY LIMIT VIOLATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-20 6.8 PROCEDURE S AND PR0 GRAMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 0 6.9 REPORTING REOUIREMENTS Routine Reports and Repartable Oc currences. . . . . . . . . . . . . . . . . . 6-21 S t a r t u p Re p o r t s . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 2 1 Annual Reports.............................................. 6-22 Personnel Exposure and Monitoring Report . . . . . . . . . . . . . . . . . . . 6-22 Annual Radiological Environmental Operating Report . . . . . . . . . 6-23 Semiannual Radioactive Effluent Releas e Re por t . . . . . . . . . . . . . 6-2 4 Monthly Op e ra ting Re po rt s. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-25 ;

Re po rt able Occurrence s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 5

]

Promp t No tification With Written Followup. . . . . . . . . . . . . . . . . . . 6-25 l Thirty Day Writ t en Re po rt s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 7 S p e ci al Re p o r t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 2 8 i

1 l

BRUNSWICK - UNIT 2 xy Amendment No. ;

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.10 RE CO RD RETENTION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 9 6.11 RADIATION PROTECTION P R0 GRAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-31 6.12 HIGH RADI ATIO N ARE A. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-31 6.13 0FFSITE DOSE CALCULATIONAL MANUAL (0DCM) . . . . . . . . . . . . . . . . . . . . 6-3 2 6.14 PROCE S S CONTROL PROGRAM (PC P ). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 2 6.15 MMOR CRANGES TO LIOUID, GASEOUS, AND E.ID WASTE TREATMENT SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 3 l

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l BRUNSWICK - UNIT 2 XVI Amendment No.

l

l I

i 1.0 DEFINITIONS The following terms are defined so that uniform interpretation of these specifications may be achieved. The defined terms appear in capitalized type and are applicable throughout these Technical Specifications.

ACTION ACTIONS are those additional requirements specified as corollary statements to each specification and shall be part of the specifications.

l AVERAGE PLANAR EXPOSURE The AVERAGE PLANAR EXPOSURE shall be applicable to a specific planar height and is equal to the sum of the exposure of all of the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

AVERAGE PLANAR LINEAR HEAT GENERATION RATE The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to

a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment as necessary of the channel t

output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION 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 perfor:ned by any series of sequential, overlapping, or l total channel steps such that the entire channel is calibrated.

l l CHANNEL CHECK l

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

CHANNEL FUNCTIONAL TEST l

A CHANNEL FUNCTIONAL TEST shall be:

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

BRUNSWICK - UNIT 2 1-1 Amendment No.

I DCFINITIONS CHANNEL FUNCTIONAL TEST (Continued)

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

CORE ALTERATION ,

CORE ALTERATION shall be the addition, removal, relocation, or movement of fuel, sources, incore instruments, or reactivity controls in the reactor core with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe, conservative location.

CRITICAL POWER RATIO The CRITICAL POWER RATIO (CPR) shall be ratio of that power in the assembly which is calculated, by application of the GEXL correlation, to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be concentration of I-131, pCi/ 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 following is defined equivalent to 1 pCi of I-131 as determined from Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites": I-132, 28 pCi; E-133, 3.7 uci; I-134, 59 pCi; I-135,12 uCi.

E-AVERAGE DISINTEGRATION ENERGY 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 disintegration (in MeV) for isotopes with half lives greater than 15 minutes making up at least 95% of the total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

BRUNSWICK - UNIT 2 1-2 Amendment No.

b DEFINITIONS END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TLME The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to recirculation pump breaker trip from initial movement of the associated:

a. Turbine stop valves, and 6

b. Turbine control valves.

FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

GASEOUS RADWASTE TREATMENT SYSTEM A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system off-gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE shall be:

a. Leakage into collection systems, such as pump seal or valve packing leaks, that is 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 i operation of the leakage detection systems or not be PRESSURE BOUNDARY

! LEAKAGE.

ISOLATION SYSTEM RESPONSE TIME The ISOLATION SYSTEM RESPONSE TIME shall be chat time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading delays where applicable.

LIMITING CONTROL ROD PATTERN l A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR.

BRUNSWICK - UNIT 2 1-3 Amendment No.

DEFINITIONS LINEAR HEAT GENERATION RATE LINEAR HEAT GENERATION RATE (LHGR) shall be the power generation in en arbitrary length of fuel rod, usually one foot. It is the integral of the heat flux over the heat transfer area associated with the unit length, usually measured in kW/ft.

LOGIC SYSTEM FUNCTIONAL TEST A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all relays and contacts of a logic circuit, from sensor output to activated device, to ensure that components are OPERABLE.

MAXIMUM FRACTION OF LIMITING POWER DENSITY MAXIMUM FRACTION OF LIMITING POWER DENSITY shall be the highest value of LINEAR HEAT GENERATION RATE (LEGR) divided by the corresponding LHCR limit occurring in the reactor core.

MAXIMUM TOTAL PEAKING FACTOR The MAXIMUM TOTAL PEAKING FACTOR (MTPF) shall be the lar;;;ast TPF which exists in the core for a given class of fuel for a given operating condition.

MEMBER (S) 0F THE PUBLIC MEMBER (S) 0F THE FUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does not include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.

MINIMUM CRITICAL POWER RATIO The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.

ODYN OPTION A ODYN OPTION A shall be analyses which refer to minimun critical power ratio limits which are determined using a transient analysis plus an analysis uncertainty penalty.

ODYN OPTION B ODYN OPTION B shall be analyses which refer to minimum critical power ratio limits determined using a transient analysis which includes a requirement for 20% scram insertion times to reduc.e the analysis uncertainty penalty.

BRUNSWICK - UNIT 2 1-4 Amendment No.

DEFINITIONS / /

OFFSITE DOSE CALCULhION MANUAL (ODCM)

The OFFSITE DOSE CALCULATIONAL MANUAL (ODCM) is a manual which contains the current methodology and parameters to be used to calculate offsite doses resulting from the release of radioactive gaseous and liquid effluents; the methodology to calculate gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints; and, the requirements of the environmental radiological monitoring program.

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

Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electric 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 CONDITION An OPERATIONAL CONDITION shall be any one inclusive combination of mode switch position and average reactor coolant temperature as indicated in Table 1.2.

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related ir.strumentation and are 1) described in Section 13 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault in a reactor coolant system component body, pipe wall, or vessel wall.

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PRIMARY CONTAINMENT INTEGRITY l PRIMARY CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE containment automatic isolation valve system, or 4

2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed positic,n, except as provided in Table 3.6.3-1 of Specification 3.6.3.1, or BRUNSWICK - UNIT 2 1-5 Amendment No.

DEFINITIONS PRIMARY CONTAIMENT INTEGRITY (Continued)

b. All equipment hatches are closed and sealed.

c. Each containment air lock is OPERABLE pursuant to Specification

3. 6.1. 3.

d. The containment leakage rates are within the limits of Specification 3.6.1.2.

e. The sealing mechanism associated with each penetration (e.g. , welds ,

bellows, or 0-rings) is OPERABLE.

PROCESS CONTROL' PROGRAM (PCP)

The PROCESS CONTROL PROGRAM (PCP) shall contain the current formula, sampling, analyses, tests and determinations to be made to ensure that the processing and I 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 cogliance 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 PURGE OR PURGING is the controlled process of discharging air or gas f rom 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 containment.

RATED THERMAL POWER RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2436 MWt.

REACTOR PROTECTION SYSTEM RESPONSE TIME REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval f rom when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

REFERENCE LEVEL ZERO The REFERENCE LEVEL ZERO point is arbitrarily set at 357 inches above the vessel zero point. This REFERENCE LEVEL ZERO is approximately mid-point on the top fuel guide and is the single reference for all specifications of vessel water level.

BRUNSWICK - UNIT 2 1-6 Amendment No.

DEFINITIONS REPORTABLE OCCURRENCE A REPORTABLE OCCURRENCE shall be any of those conditions specified in Specifications 6.9.1.13 and 6.9.1.14.

ROD DENSITY ROD DENSITY shall be the number of control rc,d notches inserted as a fraction of the total number of notches. All rods fully inserted are equivalent to 100% ROD DENSITY.

SECONDARY CONIAINMENT INTEGRITY j SECONDARY CONTAINMENT INTEGERITY shall exist when:

a. All automatic Reactor Building ventilation system isolation valves or dampers are OPERABLE or secured in the isolated position.

b. The standby gas treatment system is OPERABLE pursuant to Specification 3.6.6.1.

c. At least one door in each access to the Reactor Building is closed.

d. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

SHUTDOWN MARGIN l SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor would be suberitical assuing that all control rods capable of insertion are fully inserted except f or the analytically determined highest worth rod which is assumed to be fully withdrawn, and the reactor is in the shutdown condition, cold, 68*F, and Xenon-free.

i SITE BOUNDARY I The SITE BOUNDARY shall be that line beyond which the land is neither owned,

! nor leased, nor otherwise controlled by the licensee, as defined by Figure

! 5.1.3-1.

SOLIDIFICATION l

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

SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sen.or is exposed to radiation.

BRUNSWICK - UNIT 2 1-7 Amendment No.

I

DEFINITIONS SPIRAL RELOAD A SPIRAL RELOAP is the reverse of a SPIRAL UNLOAD. Except for two diagonal fuel bundles around each of the four SRMs, the fuel in the interior of the core, symmetric to the SRMs, is loaded first.

SPIRAL UNLOAD A SPIRAL UNLOAD is a core unload performed by first removing the fuel f rom the outermost control calls (four bundles surrounding a control blade). Unloading continues in a spiral f ashion by removing fuel f rom the outermost periphery to the incarior of the core, symmetric about the SRMs, except for two diagonal f uel bundles around each of the four SRMs.

STAGGERED TEST BASIS 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.

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

TOTAL PEAKING FACTOR The TOTAL PEAKING FACTOR (TPF) shall be the ratio of local LHGR for any specifte location on a fuel rod divided by the average LHGR associated with the f uel bundles of the same type operating at the core average bundle power.

UNIDENTIFIED LEAKAGE UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.

i l UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purpose of protection of individuals f rom exposure to radiation and radioactive materials or any area within the SITE BOUNDARY used for residential quarters or industrial,

} commercial, institutional and/or recreational purposes.

BRUNSWICK - UNIT 2 1-8 Amendment No.

DEFINITIONS VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates f rom 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 componects.

VENTING VENTING is the controlled process of discharging air or gas f rom 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. Vent, used in system names, does not imply a VENTING process.

l l

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BRUNSWICK - UNIT 2 1-9 Amendment No.

l l

TABLE 1.1 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.

SM At least once per 16 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

A At least once per 366 days.

R At least once per 18 months (550 days),

S/U Prior to each reactor startup.

P Prior to each release.

NA Not applicable.

l I

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r I

r BRUNSWICK - UNIT 2 1-10 Amendment No.

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

TABLE 1.2 OPERATIONAL CONDITIONS OPERATIONAL MODE SWITCH AVERAGE COOLANT CONDITIONS POSITIONS TEMPERATURE

1. POWER OPERATION Run Any temperature

2. STARTUP Startup/ Hot Standby Any temperature Shutdown > 212 0 p

3. HOT SHUTDOWN

4. COLD E"UTDOWN Shutdown i 2120y

5. REFUELING

Refue1** f 2120y

Reactor vessel head unbolted or removed and fuel in the vessel.***

- See Special Test Exception 3.10.3.

- - See Special Test Exception 3.10.1.

1-11 Amendment No.

BRUNSWICK - UNIT 2

TABLE 3.3.5.7-1 (Continued)

INSTRUMENT LOCATION MINIMUM INSTRUMENTS OPERABLE FLAME _ HEAT SMOKE

3. Diesel Generator Building (Cont'd)

Zone 7 23' 0 0 5 Zone 8 23' 0 0 5 Zone 9 23' O O 8 Zone 10 50' O O 9

4. Service Water Buliding Zone 1 4' O O 7 Zone 2 20' 0 0 6

5. A0G Building Zone 1 20' O O 2 Zone 2 20' 0 0 2 Zone 3 20' 1 5 1 Zone 4 37' - 49' i 6 6 i

BRUNSWICK - UNIT 2 3/4 3-61 Amendment No.

I L

INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.5.8 Ihe radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3.5.8-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alare/ trip setpoints shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (CDCM).

APPLICABILITY: As shown in Table 3.3.5.8-1.

ACTION:

l' 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 sffected channel, declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With less than one radioactive liquid effluent monitoring instrumen-tation channel in each release pathway OPERABLE, take the ACTION shown in Table 3.3.5.8-1. Return the instruments to OPERABLE status within 30 days or, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not l corrected in a timely manner.

c. The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1.14.b are not applicable.

SURVEILLANCE REQUIREMENTS 1

l i

4.3.5.8 Each radioactive liquid effluent monitoring instrumentation ch nnel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOU CE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the f requencies shown in Table 4.3.5.8-1.

^

\ . ~

NOTE: See Bases 3/4.3.5.8. p

~

/

N r

l l BRUNSWICK - UNIT 2 3/h 3-62 ' '

Amendment No.

,, ,r L_-- .

m l TABLE 3.3.5.8-1

?!

s RADIOACTIVE LIQUID EFFLUENT MONIT0 KING INSTRUMENTATION O

INSTRUMENT (8) APPLICABILITY ACTION

1. Liquid Radwaste Radioactivity Effluent Monitor (Providing alarm and automatic termination of release)

110

2. Liquid Radwaste Effluent Flow Measurement Device

111

3. Main Service Water Effluent Radioactivity Monitor

112

4. Stabilization Pond Effluent Composite Sampler ** 113

5. Stabilization Pond Ef fluent Flow Measurement Device ** 114

6. Condensate Storage Tank level Indicating Device

115

7. Service Water Effluent f rom Augmented Of f-Gas Y Precooler Radioactivity Monitor *** 112 O

8. Reactor Building Component Cooling Water (Service Water)

Radioactivity Monitors

a. Effluent from Residual Heat Removal Heat Exchanger A **** 112

b. Effluent from Residual Heat Removal Heat Exchanger B **** 112 g c. Effluent from Reactor Building Closed g Cooling Water Heat Exchangeci **** 112 m d. Effluent from Division I Residual Heat S Removal Pump Seal Coolers **** 112 z

? e. Effluent from Division II Residual Heat Removal Pump Seal Coolers **** 112 L

l TABLE 3.3.5.8-1 (Continued)

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION ACTIONS ACTIOh 110 - With less than one channel OPERABLE, effluent releases may continue provided that prior to initiating a release:

a. At least two independent samples are analyzed in accor-dance with Specifics. tion 4.11.1.1.2, and l

b. At least two technically qualified members of the facility staff independently verify the release rate calculations and uischarge line valving; Otherwise suspend release of radioactive effluents via this pathway.

ACTION 111 - With less than one channel OPERABLE, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump per-formance curves or tank level indicators may be used to esti-mate flow.

ACTION 112 - With less than one channel OPERABLE, effluent releases may continue provided that, at least once peh 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab sam-pies are collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 10-7 microcuries per gram.

3 ACTION 113 - With the stabilization pond effluent composite sampler not OPERABLE, effluent releases may continue provided that, at least once per day, a grab sample is collected and analyzed for principle gamma esitters as per Table 4.11.1-1. Other-l wise, suspend releases via this pathway.

ACTION 114 - With the stabilization pond effluent flev measuring device not OPERABLE, effluent releases via this pathway may continue provided that flow is estimated at least once per day during actual releases. The V-notch weir may be used to estimate flow.

ACTION 115 - With the tank liquid level device not OPERABLE, liquid addi-tions may continue provided the tank liquid level is estimated once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months during all liquid additions and deletions to and from the tank.

BRUNSWICK - UNIT 2 3/4 3-64 Amendment No.

TABLE 3.3.5.8-1 (Continued)

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION NOTES s

At all times

- At all times other than when the line is valved out and locked. [This equipment is to be installed. Prior to installation, appropriate action scatements 113 or 114 will be implemented. ]

l *** At all times while the A0G system precooler is in operation once this monitor is installed and af ter the A0G system becomes operational; however, if the A0G system becomes operational prior to the monitor being installed, then action statement 112 will be implemented.

[ NOTE: This monitor is to be installed].

- - - At all times once these monitors are installed and become f ully operational. [ NOTE: These monitors are to be installed pending completion of future plant modifications.]

(a) Refer to Appendix E of the OFFSITE DOSE CALCULATION MANUAL for specific instrumentation identification numbers.

l BRUNSWICK - UNIT 2 3/4 3-65 Amendment No.

i i

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TABLE 4.3.5.8-1 un RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL

' INSTRUMENT (a) CHECK CHECK CALIBRATION TEST E

U l. Liquid Radwaste Radioactivity Effluent Monitor b (Providing alarm and automatic termination of release) D M R(b) q(c)

2. Liquid Radwaste Effluent Flow Measurement Device D(e) NA R Q

3. Main Service Water Effluent Radioactivity Honitor D M R(D) Q(d)

4. Stabilization Pond Effluent Composite Sampler D NA R Q

5. Stabilization Pond Effluent Flow Measurement Device D NA R Q

6. Condensate Storage Tank Level Indicating Device D(f) NA R Q w

S 7.

Service Water Effluent from Augmented Off-Gas Precooler Radioactivity Monitor D M R ID) Q

8. Reactor Building Component Cooling Water (Service Water) Radioactivity Monitors

a. Effluent from Residual Heat Removal Heat Exchanger A D M R(b) q(d)

b. Effluent from Residual lleat Removal Heat Exchanger B D M R(b) q(d)

c. Effluent from Reactor Building Closed g Cooling Water lleat Exchangers D M R(b) q(d)

. d. Ef fluent f rom Division I Residual lleat Removal Pump Seal Coolers D M R(b) q(d) z e. Effluent from Division II Residual Heat

- Removal Pump Seal Coolers D M R(b) q(d)

TABLE 4.3.5.8-1 (Continued)

RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REOUIREMENTS e

TABLE NOTATIONS (a) Refer to Appendix E of the OFFSITE DOSE CALCULATION MANUAL for specific instrumentation identification numbers.

(b) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used. Previously established calibration procedures may be substituted for this requirement (refer to Bases 3/4.3.5.8).

(c) 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 exist:

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

2. Circuit failure (High-voltage low).

3. Instrument indicates a downscale failure.

4. Instrument controls not set in " operate" mode.

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

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

2. Circuit failure (High-voltage low).

t l 3. Instrument indicates a downscale failure.

I

4. Instrument controls not set in " operate" mode.

(e) The CHANNEL CHECK sitall consist of verifying indication of flow l during periods of release. CHANNEL CHECK shall be made at least once daily on any day on which continuous, periodic, or batch releases are made.

(f) During liquid additions to the tank.

BRUNSWICK - UNIT 2 3/4 3-67 Amendment No.

t 1

1 l

INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.5.9 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3.5.9-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The setpoints shall be determined in accordance with the methodology as described I in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY: As shown in Table 3.3.5.9-1.

ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alare/ 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 one radioactive gaseous effluent monitoring instrumentation channel OPERABLE, take the ACTION shown in Table 3.3.5.9-1. Return the instruments to OPERABLE status within 30 days

or, 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 6.9.1.14.b are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.5.9 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 at the frequencies shown in Table 4.3.5.9-1.

NOTE: See Bases 3/4.3.5.9.

l l

BRUNSWICK - UNIT 2 3/4 3-68 Amendment No.

i

E TABLE 3.3.5.9-1

!i h RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION INSTRUMENT (a) APPLICABILITY ACTION h 1. MAIN STACK MONITORING SYSTEM H

I N a. Noble Gas Activity Monitor

123

b. Iodine Sampler Cartridge

127

c. Particulate Sampler Filter

127

d. System Effluent Flow Rate Measurement Device

122 l e. Sampler Flow Rate Measurement Device

122 M

2. REACTOR BUILDING VENTILATION MONITORING SYSTEM Y *

$ a. Noble Gas Activity Monitor 123

) b. Iodine Sampler Cartridge

127

c. Particulate Sampler Filter

127

d. System Effluent Flow Rate Measurement Device

122

e. Sampler Flow Rate Measurement Device

122 I

i 3. TURHINE BUILDING VENTILATION MONITORING SYSTEM

- I k a. Noble Gas Activity Monitor

123

, 0 l $ b. Iodine Sampler Cartridge

127 z

0 c. Particulate Sampler Filter

127

TABLE 3.3.5.9-1 (Continued)

{ -RADIOACf LVE CASE 00S EFFLIENT MONITORING INSTRIMENTATION INSTRIMENT

APPLICABILITY ACTION E 3. TURBINE BUILDING VENTILATION et0NITORING SYSTEM (Continued)

U u d. System Ef fluent Flow Rate Measurement Device

122

e. Sampler Flow Rate Measurement Device

122

4. MAIN CONDENSER AIR EJECTOR RADIOACTIVITY MONITOR (Prior to input to treatment system)

a. Noble Gas Activity Monitor ** 121 (Providing alarm and automatic isolation)

M 5. MAIN CONDENSER OFF-CAS TREATMENT SYSTEM MONITOR (Downstream of AOG Treatment System)

[:

5 a. Noble Gas Activity Monitor (providing alare) *** 123

6. MAIN CONDENSER OFF-GAS TREATMENT SYSTEM EXPLOSIVE CAS MONITORING SYSTEM

a. Recombiner Train A

1. Ist Hydrogen Monitor **** 125

2. 2nd Hydrogen Monitor **** 125

b. Recombiner Train B

1. **** 125 g Ist Hydrogen Monitor g 2. 2nd Hydrogen Monitor 125 s

g 7. HOT SHOP VENTILATION MONITORING SYSTEM e,

a. Iodine Sampler Cartridge

127

b. Particulate Sampler Filter

127

TABLE 3.3.5.9-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION ACTIONS ACTION 121 - With less than one main condenser air ejector monitoring instr-umentation channel OPERABLE, gases from the main condenser off-gas system may be released to the environment for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided:

a. The GASEOUS RADWASTE TREATMENT SYSTEM is not bypassed

[ Prior to the Augmented Off-Gas Treatment System becoming operational, the GASEOUS RADWASTE TREATMENT SYSTEM shall refer to the 30-minute offgas holdup line including stack filtration], and

b. The main stack effluent noble gas activity monitor is OPERAB12; otherwise, be in at least HOT STANDBY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 122 - With less than.one chsnnel OPERABLE, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

ACTION 123 - With less than one channel OPERABLE, 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 noble gas activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 125 - With less than two channels OPERABLE in the operating recombiner train, operation of the train may continue provided proper function of the recombiner is assured by monitoring recombiner temperature in accordance with approved procedures.

With less than one channel OPERABLE in the operating recombiner train, operation of the train may continue provided grab samples from the train are collected at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and proper function of the recombiner is assured by monitoring recombiner temperature in accordance with approved procedures. .

ACTION 127 - With less than one channel OPERABLE, effluent releases via this pathway may continue provided samples are continuously col-lected with auxiliary sampling equipment and analyzed as required in Table 4.11.2-1.

BRUNSWICK - UNIT 2 3/4 3-71 Amendment No.

TABLE 3.3.5.9-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION l

NOTES

At all times.

- During operation of the main condenser air ejector.

- - At all tir.as once the Augmented Off-Gas Treatment System becomes operational.

- - - At all times during recombir.c train operation.

(a) Refer to Appendix E of the OFFSITE DOSE CALCULATION MANUAL for specific instrumentation identification numbers.

t l

l BRUNSWICK - UNIT 2 3/4 3-72 Amendment No.

g TABLE 4.3.5.9-1

?!

H RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS O

I CHANNEL MODES IN WHICH c CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT (a) CHECK CHECK CALIBRATION TEST REQUIRED

1. MAIN STACK MONITORING SYSTEM

a. Noble Gas Activity Monitor D M R(D) Q(d) *

b. Iodine Sampler Cartridge W NA NA NA *

c. Particulate Sampler Filter W NA NA NA *

d. System Effluent Flow Rate Measurement Device D NA R Q *

e. Sampler Flow Rate Measurement y Ibvice D NA R Q

w

2. REACTOR BUILDING VENTILATION MONITORING SYSTEM

a. Noble Gas Activity Monitor D M R(b) q(d) ,

b. Iodine Sampler Cartridge W NA NA NA *

c. Particulate Sampler Filter W NA NA NA

k d. System Effluent Flow Rate D NA R Q

g Measurement Device B

k e. Sampler Flow Rate Measurement Device D NA R Q

tn

@ TABLE 4.3.5.9-1 (Continued) g RADI0 ACTIVE CASEOUS EFFLUENT MONITORING INSTRUNENTATION SURVEILLANCE REOUIREMENTS X

8 CHANNEL MODES IN WHICH g CHANNEL SOURCE CHANNEL FUNCTIONAL. SURVEILLANCE y INSTRUMENT (a) CHECK CHECK CALIBRATION TEST REOUIRED w

3. TURBINE BUILDING VENTILATION MONITORING SYSTEM

a. Noble Gas Activity Monitor D H R ID) O(d) *

b. Iodine Sampler Cartridge W NA NA NA *

c. Particulate Sampler Fjlter W NA NA NA *

d. System Effluent Flow Rate Measurement Device D NA R 0 *

e. Sampler Flow Rate Measurement Device D NA R 0 *

4. MAIN CONDENSER AIR FJECTOR RADI0 ACTIVITY MONITOR (Prior to input to treatment system)

a. Noble Gas Activity Monitor (Providing alarm and automatic isolation) D M R(b) g(c) aa N 5. MAIN CONDENSER OFF-GAS TREATHElfr g

g SYSTEM HONITOR g (Downstream of A0G Treatment System)(I) n Z a. Noble Gas Activity Monitor

(Providing alarm) D M R(b) n ...

to TABLE 4.3.5.9-1 (Continued)

E v3 RADI0 ACTIVE CASE 00S EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS N

O CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT (a)

@ CHECK CHECK CALIBRATION TEST REQUIRED M

u 6. MAIN CONDENSER OFF-CAS TREATME YSTEM I

EXPLOSIVE GAS MONITORING SYSTEM

a. Recombiner Train A

1. Ist Hydrogen Monitor D NA Q(*) M ****

2. 2nd Hydrogen Monitor D NA Q(e) M ****

b. Recombiner Train B

1. Ist liydrogen Monitor D NA QI ") M ****

2. 2nd liydrogen Monitor D NA QI *) M ****

h 7. HOT SHOP VENTILATION MONITORING SYSTEM

a. Iodine Sampler Cartridge W NA NA NA *

b. Particulate Sampler Filter W NA NA NA

8.

R ii

.?

TABLE 4.3.5.9-1 (Continued)

~RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REOUIREMENTS TA5 LE NOTATION (a) Refer to Appendix E of the OFFSITE DOSE CALCULATION MANUAL for specific instrumentation identification numbers.

(b) The initial CHANNEL CALIB RATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS.

These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIB RATION, sources that have been related to the initial calibr.ation shall be used. Previously established calibration procedures may be substituted for this requirement (refer to B ases 3/4.3.5.9).

(c) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway, as described below, and control room alarm annunciation occurs if any of the following conditions exist:

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

2. Circuit failure (High-voltage low).

3. Instrument indicates a downscale failure.

4. Instrument not set in " operate" mode.

The CHANNEL FUNCTIONAL TEST of the channel up to but not including operation of the isolation valve for this pathway shall be performed within the specified surveillance interval. Testing of the isolation valve for this pathway to demonstrate operability shall be performed during the CHANNEL CALIB RATION.

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

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

2. Circuit failure (High-voltage low).

3. Instrument indicates a downscale failure.

! 4. Instrument not set in " operate" mode.

4 e

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BRUNSWICK - UNIT 2 3/4 3-76 Amendment No.

l

TABLE 4.3.5.9-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRIMENTATION SURVEILLANCE REOUIREMENTS TABLE NOTATION (e) The CHANNEL CALIBRATION shall include the use of standard gas sam-ples containing a nominal:

1. Two volume percent hydrogen, balance nitrogen, and

2. Four volume perce .t hydrogen, balance nitrogen.

(f) Instrumentation for this system is only applicable once the Augu-mented Off-Gas Treatment System becomes fully operational at the Brunswick Steam Electric Plant.

NOTES

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

- During operation of the main condenser air ejector.

- - At all times other than when the line is valved out and locked (once the Augmented Off-Gas Treatment System becomes fully operational).

- - - During recombiner train operation.

l 1

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BRINSWICK - INIT 2 3/4 3-77 Amendment No.

~

INSTRUMENTATION 3/4.3.6 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUEMENTATION LIMITING CONDITION FOR OPERATION 3.3.6.1 The anticipated transient without scram recirculation pump trip (ATWS-RPT) system instrumentation trip systems shown in Thble 3.3.6.1-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Thble 3.3.6.1-2.

_ APPLICABILITY: OPERATIONAL CONDITION 1.

ACTION:

a. With an ATWS recirculation pump trip system instrumentation trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.6.1-2, declare the trip system inoperable until the trip system is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

l b. With the requirements for the Minimum Number of OPERABLE Trip Systems per Operating Pump not satisfied for one Trip Function, restore the, inoperable trip system to OPERABLE status within 14 days or be in at least STARTUP within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE REQUIREMENTS l 4.3.6.1.1 Each ATWS recirculation pump trip system instrumentation trip system shall be demonstrated OPERABLE by the performance of the CHANNEL CRECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.6.1.1-1.

< 4.3.6. 2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of l all channels shall be performed at least once per 18 months and shall include calibration of time delay relays and timers necessary for proper functioning of the trip system.

I BRUNSWICK - UNIT 2 3/4 3-78 Amendment No.

as TABLE 3.3.6.1-1 5

us ATHS RECIRCULATION PUNP TRIP SYSTEM INSTRUMENTATION N

O MININUM NUMBER OPERABLE TRIP TRIP FUNCTION AND INSTRUMENT NUMBER SiJTEMS PER OPERATING PUMP E Reactor Vesiel Water Level -

i s 1. 1

[3 Low, Level 2 (B21-LT-N024A-2,B-2 and B21-LT-N025A-2,B-2)

(B21-LTM-N024A-2,B-2 and B21-LTH-N025A-2,B-2)

2. Reactor Vessel Pressure - High 1 (B21-PS-N045A,B,C,D) i l w 22 Y

a

'l hi a

B n

.O

m TABLE 3.3.6.1-2 pa h ATWS RECIRCULATION PIMP TRIP SYSTEM INSTRUMENTATION SETPOINTS s

R TRIP ALLOWABLE TRIP FUNCTION AND INSTRUMENT NUMBER SETPOINT VALUE N 1. Reactor Vessel Water level - > + 112 inches *

" ~> + 112 inches * ~

Tow, level 2 (B21-LTM-liO24A-2,B- 2; B21-LTM-N025A-2,B-2)

2. Reacter Vessel Pressure - High f 1120 psig f 1120 psig (B21-PS-N045A,B,C,D) i U

i

. 8 F

9 B

Vessel water levels refer to REFERENCE LEVEL ZERO.

e

a, TABLE 4.3.6.1-1 so h ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS s

N CHANNEL CHANNEL FUNCTIONAL CHANNEL ,

TRIP FUNCTION AND INSTRUMENT NUMBER CHECK TEST CALIBRATION g

1. Reactor Vessel Water level -

N Low, 14 vel 2 (B21-LT-N024A-2,B-2 and NA(a) NA R(b)

B21-LT-N025A-2,B-2)

(B21-LTM-N024A-2,B-2 and D M M B21-LTM-N025A-2,B-2)

2. Reactor Vessel Pressure - liigh NA M R (B21-PS-N045A, B, C, D)

U.

v Y

? .

(a) The transmitter channel check is satisfied by the trip unit channel g check. A separate transmitter check is not required.

k (b) Transmitters are exempted from the monthly channel calibration.

U P,

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l INSTRUMENTATION l END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.6.2 The end-of-cycle recirculation pump trip (EOC-RPT) system instrumentation channels shown in Table 3.3.6.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the T-ip Setpoint column of Table 3.3.6.2-2 and with the END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME as shown in Table 3.3.6.2-3.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL PCWER. l ACTION:

a. With an end-of-cycle recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values Column of Table 3.3.6.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value.

b. With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both trip systems, place the inoperable channel (s) in the tripped condition within one hour.

c. With the number of OPERABLE channels two or more less than required by the Minimum OPERABLE Channels per-Trip System requirement for one trip system and:

4 1. If the inoperable channels consist of one turbine control valve i

channel and one turbine stop valve channel, place both inoperable channels in the tripped condition within one hour.

2. If the inoperable channels include two turbine control valve channels or two turbine stop valve channels, declare the trip system inoperable.

d. With one trip system inoperable, restore the inoperable trip system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or take the ACTION required by Specification 3.2.3.

e. With both trip systems inoperable, restore at least one trip system to OPERABLE status within one hour or take the ACTION required by Specification 3.2.3.

l l

l BRUNSWICK - UNIT 2 3/4 3-82 Amendment No.

l t _ _ _ _ - _ - _ _ _ - _ . - _

k INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.6.2.1 Each end-of-cycle recirculation pump trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.6.2.1-1.

4.3.6.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months and shall include calibration of time delay relays and timers necessary for proper functioning of the trip system.

4.3.6.2.? The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM LESPONSE TLME of

both trip systems shown in Table 3.3.6.2-3 shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least the logic of one type of channel input, turbine control valve' fast closure, or

! turbine stop valve closure, such that both types of channel inputs are tested at least once per 36 months.

4 i

1 f

l l

BRUNSWICK - UNIT 2 3/4 3-83 Amendment No.

l L

i as TABLE 3.3.6.2-1

=

E END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMElfrATION m

M A

e MINIMUM g TRIP FUNCTION OPERABLE CHANNE PER TRIP CiSTEM a S) g H

" 1. Turbine Stop Valve - Closure 2(b)

(EllC-SVOS-1X, 2X, 3X, 4X) i

2. Turbine Control Valve - Fast Closure 2(D)

(EllC-PSL-1756, 1757, 1758, 1759)

Rv w

1 w

l (a)A trip system may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance, k

provided that the other trip system in OPERABLE.

(b)These functions are bypassed when turbine first stage pressure is equivalent to THERMAL POWER

, less than 30% of RATED THERNAL POWER.

o B

n ,.

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.O 9

to l TABLE 3.3.6.2-2 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM SETPOI!ffS M

i ALLOWABLE g TRIP FUNCTION TRIP SETPOI!ff VALUE s

1. Turbine Stop Valve-Closure -< 10% closed -< 10% closed (EllC-SVOS-lX, 2X, 3X, 4X)

2. Turbine Control Valve-Fast Closure > 500 peig > 500 psig (EHC-PSL-1756,1757,1758,1759)

M.

s 3

R a

a n

O

TABLE 3.3.6.2-3 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME a

TRIP FUNCTION RESPONSE TIME (Seconds)

1. Turbine Stop Valve-Closure < 0.175 E (EHC-SVOS-1X, 2X, 3X, 4X)

M u 2. Turbine Control Valve-Fast Closure (EHC-PSL-1756, 1757, 1758, 1759) < 0.175 v.

a G

n

w TABLE 4.3.6.2.1-1

. E END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM SURVEILLANCE REQUIREMENTS cn M

R i CHANNEL FUNCTIONAL CHANNEL g TEST CALIBRATION s TRIP FUNCTION ,

H

1. Turbine Stop Valve-Closure M* R (EHC-SVOS-lX, 2X, 3X, 4X)

2. Turbine Control Valve-Fast Closure W R (EHC-PSL-1756, 1757, 1758, 1759)

M

Including trip system logic testing.

r n

E.

REACTOR COOLANT SYSTEM 3/4.4.4 CHEMISTRY LIMITING CONDITION FOR OPERATION 3.4.4. '.he chemistry of the reactor coolent system shall be maintained within the limits specified in Thble 3.4.4-1.

APPLICABILITY: At all times.

ACTION:

a. In OPERATIONAL CONDITIONS 1, 2, and 3:

l

1. With the conductivity or chloride concentration exceeding the limits specified in Table 3.4.4-1, but less than 10 u sho/cm at 25'C and less than 0.5 ppa, respectively, operation may l continue for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and this condition need not be reported to the Commission per Specification 6.9.1.12, provided l that operation under these conditions shall not exceed 336 hours0.00389 days 0.0933 hours 5.555556e-4 weeks 1.27848e-4 months per year. 'The provisions of Specification 3.0.4 are not applicable.

2. With the conductivity or chloride concentration exceeding the I limits specified in Table 3.4.4-1 for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during one continuous time interval or with the conductivity exceeding 10 paho/cm at 25'C or chloride exceeding 0.5 ppm, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l b. At all other times with the conductivity and/or chloride concentration of the reactor coolant in excess of the limit specified in Table 3.4.4-1, restore the conductivity and/or chloride concentration to within the limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

BRUNSWICK - UNIT 2 3/44-7 Amendment No.

REACTOR COOLANT SYSTEM 3/4.4.5 SPECIFIC ACTIVITY LIMITINC CONDITION FOR OPERATION 3.4.5 The specific activity of the reactor coolant shall be limited to:

a. j( 0.2 uCi/ gram DOSE EQUIVALENT I-131, and

b. < 100/E uCi/ gram.

l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4.

ACTION:

a. In OPERATIONAL CONDITIONS 1, 2, and 3, with the specific activity of l the reactor coolant:

1. > 0.2 uCi/ gram DOSE EQUIVALENT I-131 but < 4.0 pCi/ gram, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months provided that operation under these conditions shall not exceed 10 percent of the unit's total yearly operating time. The provisions of Specification 3.0.4 are not applicable.

2. > 0.2 uC1/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or > 4.0 pCi/ gram, be in at least HOT SHUTDOWN with the main steam line isolation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3. '. 100 di u Ci/ gram, be in at least HOT SHUTDOWN with the main steam line isolation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SKUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. In OPERATIONAL CONDITIONS 1, 2, 3, or 4, l

1. With the specific activity of the primary _ coolant > 0.2 pCi/ gram DOSE EQUIVALENT I-131 or > 100 /E pCi/ gram, perform the sampling and analysis requirements of Item 4b of Table 4.4.5-1 at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months until the specific activity of the primary coolant is restored to within its limits. A REPORTABLE OCCURRENCE report shall be prepared and submitted to the Commission pursuant to Specification 6.9.1.12. This report l shall contain the results of the specific activity analyses and the time duration when the specific activity coolant exceeded 0.2 pCi/ gram DOSE EQUIVALENT I-131 together with the below additional information.

BRUNSWICK - UNIT 2 3/4 4-10 Amendnent No.

P

/ ,

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CONTAINMENT SYSTEMS

/ j <

PRIMARY CONTAINMENT STRUCTURAL INTEGRITY

./ --

LIMITING CONDITION FOR OPERATION 3.6.1.4 The structural integrity of the primary containme_nt shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.4. _

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

l ACTION:

With the structural integrity of the primary containment not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 212*F.

SURVEILLANCE REQUIREMENTS 4.6.1.4 The structural integrity of the pritsary containment shall be determined during the shutdown for each Type A ' containment leakage rate test by a visual inspection of the accessible interior and exterior surfaces of the containment and verifying no apparent changes in appearance of the surfaces or other abnormal degradation. /ety abnormal degradation bf the primary containment detected during the required inspections shall be reported to the Commission pursuant to Specificatf oa 6.9.1.12. l a

r

/

I i

t BRUNSWICK - UNIT 2 3/46-6 Amendment No.

/

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 efflu-ents to UNRESTRICTED AREAS (see Figure 5.1.3-1) af ter dilution in the discharge canal shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved concentration shall be limited to 2 x 10-gr entrained noble gases, the 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 4.11.1-1. If the stabilization pondorservicewatersamplesanalyzedaccordingtoTable4.11.1-1ingicate concentrations of any gamma-emitting radionuclides greater than 5x10- pCi/ml (trigger level), then the liquid wastes exceeding the trigger level shall be sampled and analyzed according to the sampling and analysis program of Table 4.11.1-2isuntil nuclide lesssuch thantime a the sample concentration of each gamma-emitt 5x10-g i

ng uCi/ml.

4.11.1.1.2 The results of 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.

/

NOTE: See Bases 3/4.11.1.1 BRUNSWICK - UNIT 2 3/4 11-1 Amendment No.

\

TABLE 4.11.1-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Type of Lower Limit of Liquid Release Type Sampling Analysis Activity Detection (LLD)

Frequency Frequency Analysis (pCi/al) (a)(e)

A.1. Sample Tanks, P P Principal 5 x 10-7 Detergent Drain Each Each Batch Gamma Tank, and Salt Batch Emitters (8)

Water Release Tanks 1-131 1 x 10-6 (Batch Release)(h)

P Dissolved and 1 x 10-5 One M Entrained Batch /M Gases (gamma emitters)

2. Circulating P M Gross Alpha 1 x 10-7 Water Pit Each Batch Composite (c) H-3 1 x 10-5 P Q Sr-89, St-90 5 x 10-0 Each Batch Composite (c) Fe-55 1 x 10-6

^

B. Stab sation P P Principal 5 x 10-7 Pond Each Each Gamma Release Release ,

Emitters (8)

D D During During Periods Periods Release (gj Release (gj C. Service Water (d) W W Principal 5x10-7 N

(Potential During During Gamma Continuous System System Emitters (8)

Release) Operation Operation BRUNSWICK - UNIT 2 3/4 11-2 Amendment No.

TABLE 4.11.1-1 (Continued)

RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATION (a) The detectability limits for activity analysis are based on techni-cal feasibility limits and on the potential significance in the environment of the quantities released. For some nuclides, lower detection limits may be readily achievable; and when nuclides are measured below the stated limits, they should also be reported.

l (b) When operational limitations preclude specific ganuma radionuclide analysis of each batch, gross radioactivity measurements shall be made to estimate the quantity and concentrations of radioactive material released in the batch; and a weekly sample composited from proportional aliquots from each batch released during the week shall be analyzed for principal gamma-emitting radionuclides.

(c) 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 repre-sentative of the liquids released.

(d) The stabilization pond and service water liquid release types represent potential release pathways and not actual release pathways. Surveillance of these pathways is intended to alert the plant to a potential problem; analysis for principal gamma emitters should be sufficient to meet this intent. If analysis for principal gamma y tters indicates a problem (i.e., exceeds the trigger level of 5x10 uCi/ml), then complete sampling and analyses shall be performed as per Table 4.11.1-2.

(e) The lower limit of detectability (LLD) is the smallest concentration of a radioactive material in an unknown sample that will be detected with a 95% probability with a 5% probability of falsely concluding that a blank observation represents a "r_eal" signal.

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

4.66 a b E.V.2.22 x 106 .y.exp(_x e )

l where:

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

og =

(N/tb)/2

= standard deviation of background (cpm)

BRUNSWICK - LNIT 2 3/4 11-3 Amendment No.

TABLE 4.11.1-1 (Continued)

RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATION N = background count rate (cpa) t b

= time background counted for (min)

, E = counting efficiency, as counts per disintegration l

V = volume or mass of sample 2.22 x 10 6 = conversion factor (dps/ microcurie)

Y = fractional radiochemical yield A

1

= radioactive decay constant of ith nuclide (sec-I) t, = elapsed time between sample collection and counting (sec)

Typical values of E, V, Y, and t, should be used in the calculation. It should be recognized that the LLD is defined as an "a priori" (before the fact) limit representing the capability of a measurement system and not as an "a posteriori" (after the fact) limit for a particular measurement.

(f) The stabilization pond is typically released over a several-day period. The pond is to be sampled and analyzed prior to commencing release. When composite sampling instrumentation becomes available and is OPERABLE, daily grab sampling of the stabilization pond effluent will not be required during release and the composite sample will be analyzed on a weekly basis.

(g) The principal gamma emitters for which the LLD specifications apply 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 considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.8.

(h) A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be iso-laced and then thoroughly mixed to assure representative sampling.

Once fully operational, the salt water tanks will be included as indicated in Table 4.11.1-1.

l i

l l

l BRUNSWICK - UNIT 2 3/4 11-4 Amendment No.

l l

l TABLE 4.11.1-2 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM FOR PurunAL RELEASE PATHWAYS WHICH HAVE EXCEEDED TRIGGER LEVELS Minimum Type of Lower Limit of Sampling Analysis Activity Detection (LLD)

Liquid Release Type Frequency Frequency Analysis (uCi/al)(a)(e)

A. Stabilization P P Principal 5 x 10-7(b)

Pond Each Each Gamma Release Release Faitters(I)

D D I-131 1 x 10-6 During During Periods Periods Release Release P Dissolved 1 x 10-5

,One M and Release /M Entrained Gases (Gamma Faitters)

P Each M Gross Alpha 1 x 10-7 Release Composite (c) H-3 1 x 10-5 P

Each Q Se-89, Sr-90 5 x 10-8 Release Composite (c) Fe-55 1 x 10-6 B. Service Water D(d) W Principal (Continu u Composite (c) Ganuna 5 x 10-7(b)

Release)hg Faitters(8)

I-131 1 x 10-6 M M Dissolved Grab cad 1 x 10-5 Sample Entrained Gases (Ganuma Faitters )

i D(d) M Gross Alpha 1 x 10-7 Composite (c) H-3 1 x 10-5 i

l D(d) Q Sr-89, Sr-90 5 x 10-8 l Composite (c) Fe-55 1 x 10-6 g

BRUNSWICK - UNIT 2 3/4 11-5 Amendment No.

TABLE 4.11.1-2 (Continued)

RADI0 ACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM FOR POTENTIAL RELEASE PATHWAYS WHICH HAVE EXCEEDED TRIGGER LEVELS TABLE NOTATION (a) The detectability limits for activity analysis are based on technical feasibility limits and on the potential significance in the environment of the quantities released. For some nuclides, lower detection limits may be readily achievable; and when nuclides are measured below the stated limits, they sb:uld also be reported.

l (b) When operational limitations preclude specific gamma radionuclide analysis of each batch, gross radioactivity measurements shall be made to estimate the quantity and concentrations of radioactive material released in the batch; and a weekly sample composited from proportional aliquots from each batch released during the week shall be analyzed for principal gamma-emitting radionuclides.

(c) 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 repre-sentative of the liquids released.

(d) Until such time as continuous proportional composite samplers are installed on the service water discharge line, daily grab sampling of the service water effluent will be required for use in making up the composite.

l (e) The lower limit of detectability (LLD) is the smallest concentration of a radioactive material in an unknown sample that will be detected with a 95% probability with a 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

4.66a b LLD =

6 E+V.2.22 x 10 7.exp (-Ag t,)

l Where:

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

(N/tb) 2 b

= standard deviation of background (cpm)

N = back;;round count rate (cpm)

I t

b

= time background counted for (min) l BRIJNSWICK - INIT 2 3/4 11-6 Amendment No.

L_

TABLE 4.11.1-2 (Continued) /

j RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM -

_FCR POTENTIAL RELEASE PATHWAYS WHICH HAVE EXCEEDED TRIGGER LEVELS

. TABLE NOTATION E = counting efficiency, as counts per disintegration V = volume or mass of sample

2.22 x 10 6 = conversion factor (dpe/ microcurie)

Y = fractional radiochemical yield A

g

= radioactive decay constant of ith nuclide (sec-I) 1 t,

= elapsed time between sample collection and counting (sec) should be used in the Typical values calculation. Itofshould E, V, Y, be and t,ized that the LLD is defined as an recogn "a priori" (before the fact) limit representing the capability of a measurement system and not as an "a posteriori" (af ter the fact) limit for a particular measurement.

(f) The stabilization pond is typically released over a several-day I

period. The pond is to be sampled and analyzed prior to commencing release. When composite sampling instrumentation becomes available and is OPERABLE, daily grab sampling of the stabilization pond effluent will not be required during release and the composite sample will be analyzed on a weekly basis.

(g) The principal gamma emitters for which the LLD specifications apply ,

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 considered. Other gamma peaks that are identifiable, together with those of the above I nuclides, shall also be analyzed and reported in the 3emiannual

! Radioactive Effluent Release Report pursuant to Specification

! 6.9.1.8.

l (h) A continuous release is the discharge of liquid waste of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release.

BRUNSWICK - UNIT 2 3/4 11-7 Amendment No.

RADIOACTIVE EFFLUENTS DOSE - LIQUID EFFLUENTS LIMITING CONDITION FOR O_PERATION 3.11.1.2 The dose or dose commitment to a KEMBER OF THE PUBLIC from radio-active materials in liquid effluents released to UNRESTRICTED AREAS (see Figure 5.1.3-1) shall be limited:

a. During any calendar quarter to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ, and

b. During any calendar year to less than or equal to 6 mrem to the total body and to less than or equal to 20 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated doses 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 action to be taken to assure that subsequent releases will be in compliance with the above limits.

l l 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 efflu-ents for the current calendar quarter and the current calendar year shall be determined in accordance with the ODCM at least once per 31 days.

NOTE: See Bases 3/4.11.1.2 l

BRUNSWICK - UNIT 2 3/4 11-8 Amendment No.

RADIOACTIVE EFFLUENTS LIQUID RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.1.3 The liquid radwaste treat:nent 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 the site to UNRESTRICTED AREAS i (see Figure 5.1.3-1) would exceed 0.12 mrem to the total body or 0.4 mrem to any 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-system, and reason for the inoperability.

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

3. Summary of description of action (s) taken to prevent a recur-rence.

i

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

SURVEILLANCE REQUIREMENTS 4.11.1.3 Doses due to liquid releases from the site to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the ODCM.

l NOTE: See Bases 3/4.11.1.3 l

l l

BRUNSWICK - UNIT 2 3/4 11-9 Amendment No.

I a

, RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS Appropriate alternatives to the ACTIONS and Surveillance Requirements below can be accepted if they provide reasonable assurance that ic the event of an uncentrolled release of the tanks' content, the resu* ting concentrations would be less than the limits of 10 CFR Part 20, Appendix 3, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.

LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material suspended in solution in each of the following unprotected outdoor tanks shall be limited to less than or equal to the activity indicated below, excluding tritius and dissolved or entrained gases.

OUTSIDE TANK CURIE LIMIT

a. Condensate Storage Tank 10 Ci

b. Outside Temporary Tank 10 Ci APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive sster!21 in any of the above listed tanks exceeding the above limit, without delay suspend all addition of radioactive matgrial to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank's contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report.

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

l SURVEILLANCE REQUIREMENTS _

4.11.1.4 The quantity of radioactive material contained in each of the tanks listed shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank.

NOTE: See Bases 3.4.11.1.4 i

i BRUNSWICK - UNIT 2 3/4 11-10 Amendment No.

RADIOACTIVE EFFLUENTS l

3/4.11.2 GASEOUS EFFLUENTS DOSE RATE LEMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Fig-ure 5.1.3-1) 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 arems/yr to the skin, and

b. For iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrems/yr to any 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 lLait(s).

SIRVEILLANCE REOUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology as described in the ODCM.

4.11.2.1.2 The dose rate due to iodine-131, iodine-133, 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 methodology as described in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11.2-1.

l l

NOTE: See Bases 3/4.11.2.1 BRUNSWICK - UNIT 2 3/4 11-11 Amendment No.

, , TABLE 4.I1.2-1 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM a

R

' Minimum Lower Limit of Sampling Analysis Type of Detection (LLD)(a)

Gaseous Release Type Frequency Frequency Activity Analysis (p Ci/al) e

" A. Drywell Purge P P Principal Gamma 1 x 10-4 Each Purge Each Purge Emmitters )

Grab Samples

-4 B. Environmental Release M(c)(d) M(c) Principal Camma(b) 1 x 10 Points - Main Stack, Grab Sample Emmitters Reactor Building Vent, Turbine Building Vent, H-3 1 x 10-6 w Hot Shop (h) s Continuous (*) W(f)(8)

C Charcoal I-131 1 x 10-12 h Sample

-II Continuous (*) W(f)(8) Principle Gamma (b) 1 x 10 Particulate Emmitters Sample (1-131, others)

Continuous (e) M Composite Cross Alpha 1 x 10-11 Particulate Sample ,

Continuous (e) q Composite Sr-89, Sr-90 1 x 10-11 ko Particulate Sample S Continuous (*) Noble Gas Noble Gases, g Monitor Gross Beta or Gamma 1 x 10-6 ,

TABLE 4.11.2-1 (Continued)

RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATION (a) The lower limit of detectability (LLD) is the smallest concentration of a radioactive material in an unknown sample that will be detected with a 95% probability with a 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

, 4.66c b E V.2.22 x 10 6 .y.,xp(_x ge ) -

Where:

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

(N/tb) b i

= standard deviation of background (cpe) 3 N = background count rate (cpm) t b

= time background counted for (min)

E = counting efficiency, as counts per disintegration V = volume or mass of sample 2.22 x 10 6 = conversion factor (dpm/ microcurie)

Y = fractional radiochemical yield Ag = radioactive decay constant of ith nuclide (sec-1) e, = elapsed time between sample collection and counting (sec)

Tyrical values of E, V, Y, and t, should be used in the calculation. It should be recognized that the LLS it, defined as an "a priori" (before the fact) limit representing the capability of a measurement system and not as an "a posteriori" (af ter the fact) limit for a particular measurement.

, (b) The principal gamma emitters for which the LLD specification applies i exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for BRUNSWICK - INIT 2 3/4 11-13 Amendment No.

d l

TABLE 4.11.2-1 (Continued)

RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATION l particulate emissions. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.8.

(c) With a THERhAL POWER change exceeding 15 percent of RATED THERMAL POWER within one hour, or following shutdown or start-up, sampling and analyses shall also be performed unless (1) analysis shows that the DOSE EQUIVA-LENE I-131 concentration in the primary coolant has not increased more than a f actor of 3; and (2) the main condenser air ejector noble gas activity monitor shows that activity has not increased by more than a

f actor of 3.

(d) If during refueling, water exceeds 2 x 10-ghe tritium concentration in the spent fuel pool WCi/al, tritium grab samples shall be taken at least once per 7 days f rom the ventilation exhaust f rom the spent fuel pool area whenever spent fuel is in the spent fuel pool. Spent fuel pool water will be sampled at least once per 7 days during refueling.

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

(f) Sample cartridges / filters 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 af ter changing (or af ter removal f rou sampler).

(g) Sampling shall be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, start-up, or THERMAL POWER change exceeding i

15 percent of RATED THERMAL POWER in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and analyses shall be com-l pleted 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 f actor of

10. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a f actor of 3; and (2) the main condenser air ejector noble gas l

I monitor shows that activity has not increased more than a f actor of 3.

l This footnote does not apply to the Hot Shop environmental release point.

(h) Monthly grab samples to be analyzed for principal gamma emitters and

tritium are not applicable for the Hot Shop environmental release l

point. In addition, the Hot Shop release point does not have a l

continuous noble gas monitor and, therefore, the noble gas activity analysis requirements of Table 4.11.2-1 are not applicable.

BRUNSWICK - UNIT 2 3/4 11-14 Amendment No.

l

RADIOACTIVE EFFLUENTS DOSE - NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1.3-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation;

b. During any calendar year: Less than or equal to 20 mrad for gamma radiation and less than or equal to 40 mrad for beta radiation.

APPLICABILITY: At all times.

ACTIONS:

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 to be 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.

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

NOTE: See Bases 3/4.11.2.2 l

l BRUNSWICK - UNIT 2 3/4 11-15 Amendment No.

RADIOACTIVE EFFLUENTS DOSE - IODINE-131, IODINE-133. TRITIIM, AND RADION0CLIDES IN PARTICULATE FORM LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to a MEMBER OF THE PUBLIC frne iodine-131, iodine-133, tritium, and c11 radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1.3-1) shall be limited to the following:

( a. During any calendar quarter: Less than or equal to 15 mrems to any organ; and

b. During any calendar year: Less than or equal to 30 mrems to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of iodine-131, iodine-133,

, 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, iodine-133, i

tritium, and 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.

NOTE: See Bases 3/4.11.2.3 BRUNSWICK - UNIT 2 3/4 11-16 Amendment No.

/

RADIOACTIVE EFFLUENTS f i

GASEOUS RADWASTE TREATMENT SYSTEM- ,-

LIMIT ,d CONDITION FOR OPE *ATION 2 6.2.4 The GASEOUS RADi4ASTE TREATMENT SYSTEM shall be in operation.

APPLICABILITY: Whenever the main condenser air ejector (evacuation) system is in operation.,

ACTION:

a. With gaseous radwaste from the main condenser air ejector system being discharged without treatment for more than 7 days, 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. Identification of the inoperable equipment or subsystems and the reason for inoperability,

2. Action (s) taken to restore the inoperable equipment to OPERABLE statue, 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 The readings of the relevant instruments shall be checked at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the main condenser air ejector is in use to ensure that the GASEQUS RADWASTE TREATMENT SYSTEM is functioning.

l NOTE: See Bases 3/4.11.2.4 l BRUNSWICK - UNIT 2 3/4 11-17 Amendment No.

I l

RADIOACTIVE EFFLUENTS VENTILATION EXHAUST TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 8 3.11.2.5 The VENTILATION EXHAUST TREATMENT SYSTEM shall be used ta reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases. from the sita. to areas at and beyond the SITE BOUNDARY (see Figure 5.1.3-1), would exceed 0.6 mrem to any organ over 31 days. .

APPLICABILITY: At all times other than when the VENTILATION EXHAUST TRZATMENT SYSTEM ic undergoing routine mai ;enance.

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 Specification 6.9.2, a Special Report that includes the following information:

1. Identification of any inoperable equipment or subsystems and the reason for the inoperability;

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

3. Sur:: mary description of action (s) taken to prevent a recurrence.

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

l SURVEILLANCE REQUIREMENTS 4.11.2.5 Doses due to gaseous releases from the site shall be projected at least once per 31 days, in accordance with the ODCM, when the VENTILATION EXHAUST TREATMENT SYSTEM is not in use.

NOTE: See Bases 3/4.11.2.5 l

t l

l BRUNSWICK - UNIT 2 3/4 11-18 Amendment No.

l

RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE p i1 TING CONDITION FOR OPERATION 1

3.11.2.6 The concentration of hydrogen in the main condenser offgas treatment systes shall be limited to less than or eqcal to 4% by volume.

APPLICABILITY: Whenever the main condenser air ejector system is in operation.*

ACTION:

a. With the concentration of hydrogen in the main condenser offgas treatment system exceeding the limit, restore the concentration to within the limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

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

SURVEILLANCE REQUIREMENTS 4.11.2.6 The concentration of hydrogen in the main condenser offgas treatment system shall be determined to be within the above limit by continuously monitoring the waste gases in the main condenser offgas treatment system with the hydrogen monitors required OPERABLE by Table 3.3.5.9-1 of Specificatioa 3.3.5.9.

NOTE: See Bases 3/4.11.2.6

This specification shall become applicable when the offgas recombiners become operational.

l I

l l

BRUNSWICK - UNIT 2 3/4 11-19 Amendment No.

l

RADIOACTIVE EFFLUFN]

MAIN CONDENSER AIR EJECTOR RADIOACTIVITY RSLEASE RATE LIMITING CONDITION FOR OPERATION 3.11.2.7 The release rate of the sum of.the activities f rom the noble gases measured at the main condenser air ejector shall be limited to less than or equal to 243,600 microcuries/second (the Kr-85m, 87, 88 and Xe-133,135,138 contribution af ter 30 ninutes decay),

_ APPLICABILITY: During operation of the main condenser air ejector.

ACTION: ,

With the release rate of the sum of the activities f rom the noble gases at the main condenser air ejector exceeding the above limit, restore the gross radioactivity rate to within its limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REOUIREMENTS 4.11.2.7.1 The radioactivity rate of noble gases at (near) the outlet of the main condenser air ejector shall be continuously monitored in accordance with Specification 3.11.2.1.

4.11.2.7.2 The release rate of the sum of the activities from the noble gases from the main condenser air ejector shall be determined to be within the above limit at the following f requencies by performing an isotopic analysis of a representative sample of gases taken at the discharge (prior to dilution and/or discharge) of the main condenser air ejector:

'a. At least once per 31 days, or

b. Within 31 days following each refueling / maintenance outage, and

c. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following an increase, as indicated by the Condenser Air Ejector Noble Gas Activity Monitor, of greater than 50%, af ter ,

factoring out increases due to changes in TRERMAL POWER level, in the nominal steady state fission gas release f rom the primary coolant.

NOTE: See Bases 3/4.11.2.7 BRUNSWICK - UNIT 2 3/4 11-20 Anendment No.

RADIOACTIVE EFFLUENTS DRYWELL VENTING OR PURGING LIMITING CONDITION FOR OPERATION 3.11.2.8 The drywell shall be purged to the environment at a rate in conformance with Specification 3.11.2.1.

APPLICABILITY: Whenever the drywell is vented or purged.

ACTION:

a. With the requirements of the above specification not satisfied, suspend all VENTING or PURGING of the drywell.

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

SURVEILLANCE REQUIREMENTS 4.11.2.8 A sample analysis, as defined in Table 4.11.2-1, shall be performed prior to each drywell PURGE.

l NOTE: See Bases 3/4.11.2.8 l

BRUNSWICK - UNIT 2 3/4 11-21 Amendment No.

J RADIOACTIVE EFFLUENTS 3/4.11.3 SOLID RADIOACTIVE WASTE LIMITING CONDITION FOR OPERATION 3.11.3 The solid radwaste system shall be used in accordance with a PROCESS CONTROL PROGRAH to process wet radioactive wastes to meet shipping and burial ground requirements.

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 6.9.1.14.b are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.3 The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of et least one representative test sper.inen from at least every tenth batch of each type of wet radioactive waste (e.g., filter sludges, spent resins, evaporator bottoms, and sodium sulfate solutions).

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

SOLIDIFICATION of the batch may then be resumed using the alterna-tive SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM.

l

b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection of 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 provid'd in Spe.cification 6.14, to assure SOLIDIFICATION of subsequent batches of waste.

NOTE: See Bases 3/4.11.3 BRUNSWICK - UNIT 2 3/4 11-22 Amendment No.

1 i

RADIOACTIVE EFFLUENTS

, j 3/4.11.4 TOTAL DOSE (40 CFR PA2r 190) !

l LIMITING CONDITION FOR OPERATION 4

3.11,4 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity and radiation from uranium fuel cycle sources sball ha limited to less than or equal to 25 mrems to the total i body or any organ (except the thyroid, which shall be limited to less than or J equal to 75 areas).

]

l APPLICABILITY: At all times. <

! l ACTION:

a. With the calculated doses from the release of radioactive materials ,

in liquid or gaseous effluents exceeding twice the limits of Speci- l fications 3.11.1.2.a. 3.11.1. 2.b, 3.11.2.2.a, 3.11.2.2.b, 1 3.11.2.3.a, or 3.11.2.3.b, calculations should be made which, in addition to doses due to effluents, include direct radiation contributions from the reactor units and from outside storage tanks 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 subsequent releases to prevent l recurrence of e.xceeding 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 i pathways and direct radiation, for the calendar yaar 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 1 condition resulting in violation of 40 CFR Part 190 has not already I i 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 Specificatione 3.0.3 and 3.0.4 are not applicable.

l l

BRUNSWICK - UNIT 2 3/4 11-23 Amendment No.

)

l

l RADIOACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE (40 CFR FARI 190)

SURVEILLANCE,REQUREMENTS 4.11.4.'1 Dosa Calculations Cumilative dose contributions from liquid and gas-eous effluencs shall be determined in accordance with Specifiestions 4.11.1.2, I*.11.2.2, and 4.11.2.3, and in accordance with the ODCM.

4.11.4.2 Cumuistive dose contributions from direct radiation from the reactor units and from radwaste storage tanks shall be determined in accordance with the ODCM. This requirement is applicable only under conditions set forth in Specification 3.11.4.a.

NOTES: See Bases 3/4.11.4 l.

I t

l !

i l

~

BRUNSWICK - UNIT 2 3/4 11-24 Amendment No.

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 3.12.1-1.

r APPLICABILITY: At all times.

ACTION:

1

a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12.1-1, in lieu of a Licensee Event Report, prepara and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Specifi-cation 6.9.1.6, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

i

b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specific location exceeding the reporting levels of Table 3.12.1-2 when averaged over any calen-dar 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) and defines the corrective action to be taken to reduce radioactive effluents so that the potential annual dose to a MEMBER OF THE PUBLIC is less than the calendar year limits of Specifica-tions 3.11.1.2, 3.11.2.2, and 3.11.2.3. When more than one of the radionuclides in Table 3.12.1-2 are detected in the sampling medium, this report shall be submitted if:

l concentration (1) concentration (2) reporting level (1)

+

reporting level (2) + ... _> 1.0 ,

When radionuclides other than those in Table 3.12.1-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2, and 3.11.2.3. The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

c. With milk or fresh leafy vegetables unavailable from one or more of the sample locations required by Table 3.12.1-1, identify locations for obtaining replacement samples and add them to the radiological BRUNSWICK - UNIT 2 3/4 12-1 Amendment No.

RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION (Continued)

ACTION,(Continued) environmental monitoring program within 30 days. The specific locations from which samples were unavailabic ray then be deleted from the monitoring program and ODCM. In lieu of a Licensee Event l Report and pursuant to Specification 6.9.1.8, identify the cause of l 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 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12.1-1 from the specific locations given in the table and figure (s) in the ODCM and shall be at.alyzed pursuant to the requirements of Table 3.12.1-1 and the detection capabilities required by Table 4.12.1-1.

e l

NOTE: See Bases 3/4.12.1 l

BRUNSWICK - UNIT 2 3/4 12-2 Amendment No.

TABLE 3.12.1-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM h

R I

O Number Of Samples N Exposure Pathway and Sampling and Type and Frequency u and/or Sample Sample Locations (a) Collection Frequencg Of Analysis

1. DIRECT RADIATION (b) 40 Locations. At each Q Gamma Dose - Q location with 2 or more dosim-eters or one or more instru-ments for continuously measur-ing and recording dose rate, placed as follows:

O An inner ring of stations, with at least one in each U meteorological sector in the b general area of the SITE BOUNIMRY as is reasonably accessibic and practical; An outer ring of stations, with at least one in each meteorological sector at distances of 8 km or greater f rom the site as' is reasonably accessible and practical; and l

I The balance of stations to be g

g placed in special interest g areas such as population n centers, nearby residences, g schools, and in at least one

+ or two areas to serve as control stations.

l l

E TABLE 3.12.1-1 (Continued)

E l RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM n

Number Of Samples E Exposure Pathway and Sampling and Type and Frequency N and/or Sample Sample locations (a) Collection Frequency Of Analysis N

2. AIRBORNE - 5 Incations, as follows: Continuous sampler opera- Radiciodine Cannister:

Radioiodine and tion with sample col- 1-131 analysis - W Particulate 3 samples frca different lection weekly or as sectors as close to the required by dust loading, Patticciate sampler:

SITE BOUNDARY as is whichever is more fre- Gross beta radioactivity reaaonably accessible, quent. analysia 11owing filter one of which being at change; the highest calculated Gamma isotopic analysisI *)

annual average ground of composite (by g

level D/Q; location) - Q f I sample from the

- vicinity of a nearby community; and j

1 sample from a control j location, as for example l

greater than 15 km dis-tantandinalesspry-)

valent wLnd direction c

$ 3. WATERBORNE 2 locations, as follows: Composite (g) sample Surface (f) GammaIsgpic 4

ks a.

I sample upstream collection - M Analysis -M g Tritium Analysis - Q 1 sample downstream

TABLE 3.12.1-1 (Continued)

E

@ RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 2

E! Number of Samples

Exposure Pathway and Sampling and

' Type and Frequency and/or Sample Sample Locations (a) Collection Frequency of Analysis O

Lj 3. WATERBORNC (Continued) 9

b. Sediment from 1 location with a sample SA Camma Is t pic shoreline taken from a downstream Analysis * - SA area with existing or potential recreational value

4. INGESTION w a. Milk 4 locations as follows: With animals on Gamma is t 20 pasture - SM analysis {*gpic and 1-131 g Samples from milking analysis - SM (when 7

animals at 3 locations At other times - M animals are on pasture);

Within 8 km distance having the highest dose At other times - M potential {g}en available) 1 sample from milking animals at a control location greater than 15 km distance from the site and in a less prevalent wind direction

{s b. Fish and Invertebrates 4 locations as follows:

When in season - SA Gamma is analysis pic on edible B

3 samples of commercially portions - SA and recreationally impor-tant species in the vicinity of the plant dis-charge: one free swimming

m

{ TABLE 3.12.1-1 (Continued)

RADIOLOGICAL ENVIRONNENTAL MONITORING PROGRAM a

i c

% Number Of Samples M

Exposure Pathway and Sampling and Type and Frequency

" and/or Sample Sample locations (a) Collection Frequency of Analysis

4. INGESTION (Continued)

b. Fish and species; one botton Invertebrates feeding species; and one (Continued) shellfish species.

I sample of a similarly w edible species from an

]; area not influenced by g plant discharge to serve

'fm as a control sample.

c. Broadleaf 3 locations as follows: When available - M Gammaisggpic Vegetation analysis and Samples of broadleaf vege- I-131 analysis - M tation grown in 2 sectors (when available) of historically higher D/Q values at the SITE BOUNDARY if milk sampling is not performed.

I sample of a similar g broadleaf vegetation grown a at a distance of greater E. than 15 km from the site

$ in a less prevalent wind

+

E direction if milk sampling is not performed.

l l

l

TABLE 3.12.1-1 (Concinued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM TABLE NOTATION (a) Specific parameters of distance and direction sector from the site, and additional description whr.re pertinent, shall be provided for each and every sample location in Table 3.12.1-1 in a tabic and figure (s) in the ODCM. 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 Radiological Environmental Operating Report pursuant to Specification 6.9.1.7. It is recognized that, at times, it may not h* possible or practicable to continue to obtain samples of the media of choice at Cae 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 I

radiological environmental monitoring program. In lieu of a License Event Report and pursuant to Specification 6.9.1.8, identify the cause of the unavailability of samples for that pathway 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). ,

(b) 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. Film badges shall not be used as dosimeters for measuring direct radiation. 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 nimal fading.

(c) The purpose of this sample is to obtain background information. If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites that provide valid background data may be substituted.

(d) 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. ,

BRUNSWICK - UNIT 2 3/4 12-7 Amendment No.

l TABLE 3.12.1-1 (Continusd)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM TABLE NOTATION (e) Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.

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

" Upstream" samples in an estuary must be taken far enough upstream

! to be beyond the plant influence. Salt water shall be sampled only when the receiving water is utilized for recreational activities.

(g) 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 representative of the liquid flow. Composite samples shall be collected with equipment that is capable of collecting an aliquot at time intervals that are short (e.g., once per 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ) relative to compositing period (e.g., monthly) in order to assure obtaining a representative sample.

l (h) When less than three (3) milking animal locations are available for within an 8-km distance, sampling of broadleaf vegetation shall be performed as indicated in Table 3.12.1-1, 4.c, in lieu of testing with 3.0.3, 3.0.4, and 6.9.1.14.b are not applicable.

l l

BRUNSWICK - UNIT 2 3/4 12-8 Amendment No.

t l

__ _ , _ _ _ ~ _ - .

. TABLE 3.12.1-2 (Continued)

o h REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES s

Q REPORTING LEVELS i

h Water Airborne Fish Milk Broadleaf Vegetation H Analysis (pCi/1) or CasesParticug)te (pCi/m (pCi/kg, wet) (pC1/1) (pC1/kg, wet) 11 - 3 3 x 10 4 - - - -

Mn-54 1 x 10 3 -

3 x 10 4 - -

Fe-59 4 x 10 2 -

1 x 10 4 - -

Co-58 4 x 10 2 -

3 x 10 4 - - -

N u \ \

Co-60 3 x 10 2 -

1 x 10 4 - -

\

w

$ Zn-65 3 x 10 2 -

2 x 10 4 - -

Zr-Nb-95 4 x 10 2 _ _ _ _

I-131 2 0.9 -

3 1 x 10 2 3

Cs-134 30 10 1 x 10 60 1 x 10 3 \

Cs-137 50 20 2 x 10 3 70 2x 10 3 2

g Ba-La-140 2 x 10 - -

3 x 10 2 _

it u

n

$ TABLE 4.12.1-1 3

g DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS (a)

I h LOWER LIMIT OF DETECTION (LLD)(b)

H N

Airborne Broadleaf Water Particulate or Fish Milk Vegetation Sediment Analysis (pCi/1) Cases' (pCi/m ) (pCi/kg, wet) (pCi/1) (pC1/kg, wet) (pCi/kg, dry) gross beta 4 0.01 - - - -

11 - 3 3000 - - - - -

w Mn-54 15 -

130 - - -

D e Fe-59 30 -

260 - - -

5 Co-58, 60 15 -

130 - - -

Zn-65 30 -

260 - - -

Zr-Nb-95 15 - - - - -

I-131 1(c) 0.07 -

1 60 -

Cs-134 15 0.05 130 15 60 150

$ Cs-137 18 0.06 150 g 18 80 180

p. *

$ Ba-La-140 15 - -

15 - -

N P

TABLE 4.12.1-1 (Continued)

DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS 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 6.9.1.7.

(b) The lower limit of detectability (LLD) is the smallest concentration of a radioactive material in an unknown sample that will be detected with a 95% probability with a 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

4.66e b E V 2.22 Y exp(-Ag t,)

Where:

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

b (N/tb) 2

= standard deviation of background (cpm)

N = background count rate (cpm) t b

= time background counted for (min)

E = counting efficiency, as counts per disintegration i

V = volume or mass of sample l l 2.22 = conversion factor (dpm/pci)

Y = fractional radiochemical yield A

g

= radioactive decay constant of ith nuclide (sec-I) t,

= elapsed time between sample collection and counting (sec) i l

BRUNSWICK - UNIT 2 3/4 12-11 Amendment No.

l L

TABLE 4.12.1-1 (Centinund)

DETECTION CAPABILITIES POR ENVIRONMENTAL SAMPLE ANALYSIS TABLE NOTATION l

Typical values of E, V, Y, and t, should be used in the calculation. It should be recognized that the LLD is defined as an "a priori" (before the fact) limit representing the capability of a measurement system and not as an "a posteriori" (af ter the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interferring nuclides, or cther uncontrollable circumstances may render these LLDs unachievable. In such cases: tha contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report

. pursuant to Specification 6.9.1.7.

(c) The LLD of gamma isotopic analysis may be used.

l

[

BRUNSWICK - UNIT 2 3/4 12-12 Amendment No.

l

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 the nearest resident, and the nearest sectorsofofgreater garden the nearest milk than 50 m2animal, (500 f t2) producing broadleaf vegetation. (For elevated releases as defined in Regulatory Guide 1.111, Revision 1, July 1977, the land use census shall also identify within a distance of 5 km (3 miles) the location in each of the 16 mgteorological sectors of all milk animals and all gardens of greater than 50 m producing broadleaf vegetation.)

Broadleaf vegetable sampling of at least 3 different kinds of vegetation may be performed at the SITE BOUNDARY in each of 2 different direction sectors with the highest D/Qs in lieu of the garden census. Specifications for broad-leaf vegetation sampling in Tkble 3.12.1-1(4c) shall be followed, including analysis of control samples.

APPLICABILITY: At all times.

ACTION:

a. With a land use census identifying a location (s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, in lieu of a Licensee Event Report, identify the new location (s) in the next Seniannual Radioactive Ef fluent Release Report, pursuant to Specification 6.9.1.8.

b. With a land use census identifying a location (s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent greater than at a location f rom 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 central station location, having the lowest calculated dose or dose commitment (s) (via this same exposure pathway) may be deleted from this monitoring program af ter October 31 of the year in which this land use census was conducted. In lieu of a licensee Event Report and pursuant to Specification 6.9.1.8, identify the new location (s) in the next Semiannual Ef fluent 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.

I BRUNSWICK - UNIT 2 3/4 12-13 Amendment No.

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 result of the land use census shall be ,

included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.7.

NOTE: See Bases 3/4.12.2 i

l l

l BRUNSWICK - UNIT 2 3/4 12-14 Amendment No.

RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program that has been approved by the Commis-sion.

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 6.9.1.7.

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

SURVEILLANCE REQUIREMENTS 4.12.3 The Interlaboratory Comparison Program shall be described in the ODCM. A summary of the results, obtained as part of the above required Interlaboratory Comparison Program, shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.7.

I l

NOTE: See Bases 3/4.12.3 ,

BRUNSWICK - UNIT 2 3/4 12-15 Amendment No.

. - - . - - _ _ . . , . , - _ - , - .-._,___.--,..-._n . . - - - , - .

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

INSTRUMENTATION BASES MONITORING INSTRIMENTATION (Continued) 3/4.3.5.6 CHLORIDE INTRUSION MONITORS The chloride intrusion monitors provide adequate warning of any leakage in the condenser or hotwell so that actions can be taken to mitigate the consequences of such intrusion in the reactor coolant system. With only a minimum number of instruments available, increased sampling frequency provides adequate information for the same puryse.

3/4.3.5.7 FIRE DETECTION INSTRIMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overall facility fire protection program.

In the event that a portion of the fire detection instrumentation is inoperable, increasing the frequency of fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

3/4.3.5.8 RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRIMENTATION l The radioactive liquid effluent monitoring instrumentation is provided to t monitor and control, as applicable, the releases of radioactive materials in lionid effluents during actual or potential releases of liquid effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance i with 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 l Appendix A to 10 CFP Part 50. The purpose of tank level indicating devices is j to assure the detect on and control of leaks that, if not controlled, could l potentially result in the transport of radioactive materials to UNRESTRICTED AREAS. "Without delay" implies that the operator, upon determining the limiting condition for operation is being exceeded, takes the next appropriate action to comply with the specification.

The initial CHANNEL CALIBRATION for the instruments associated with footnote (b) to Table 4.3.5.8-1 shall be performed using National Bureau of Standards traceable sources which will verify that the detector operates properly over its intended energy range and measurement range. For instruments which were operational prior to this specification being implemented, previously established calibration procedures may be substituted for this requirement.

BRUNSWICK - UNIT 2 B 3/4 3-4 Amendment No.

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

)

1 INSTRUMENTATION l

2 RASES RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRIMENTATION (Continued)

Subsequent CHANNEL CALIBRATIONS will be performed using sources that have been related to the initial calibration in order to ensure that the detector is still operational, but the sources need not span the full ranges used in the initial CHANNEL CALIBRATION.

3/4.3.5.9 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRINENTATION i The radioactive gaseous effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.

The alarm / trip setpoints for these instruments shall be calculated in accordance with the 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 and 64 of Appendix A to 10 CFR Part 50.

The main condenser air ejector monitoring instrumentation, the main condenser offgas treatment system aonitor, and the explosive gas monitoring instrumen-tation shown in Table 3.3.5.9-1 are not considered effluent monitoring instrumentation in the same sense as the other instrumentation listed in the table. Therefore, their alarm / trip setpoints are not necessarily set to ensure that the limits of Specification 3.11.2.1 are not exceeded.

i The main condenser air ejector monitoring instrumentation channels are provided to monitor and control gross radioactivity removed from the main condenser. The alarm / trip setpoints for the main condenser air ejector monitor are set to ensure that the limits of Specification 3.11.2.7 are not exceeded. The alarm / trip setpoint for this monitor shall be calculated in accordance wita NRC approved methods to provide reasonable assurance that the potential total body accident dose will not exceed a fraction of the limits specified in 10 CFR Part 100.

This specification also includes provisions for monitoring the concentrations l of potentially explosive gas mixtures in the offgas treatment system (hydrogen monitors). The hydrogen monitors will become applicable when the offgcc recombiners become fully operational (prior to operation of the Augmented Off-Gas Treatment System) at the Brunswick Steam Electric Plant. There is no requirement for hydrogen monitors on the 30-minute waste gas holdup line which will serve in the interim.

"Without delay" implies that the operator, upon determining the limiting condition for operation is being exceeded, takes the next appropriate action to comply with the specification.

BRUNSWICK - UNIT 2 B 3/4 3-5 Amendment No.

I i

INSTRUMENTATION BASES RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRIMENTATION (Continued)

The initial CHANNEL CALIBRATION for the instruments associated with footnote (b) to Table 4.3.5.9-1 shall be performed using National Bureau of -

Standards traceable sources which will verify that the detector operates properly over its intended energy range and measurement range. For instruments which were operational prior to this specification being i implemented, previously established calibration procedures may be substituted for this requirement. Subsequent CHANNEL CALIBRATIONS will be performed using sources that have been related to the initial calibration in order to ensure that the detector is still operational, but the sources need not span the full ranges used in the initial CHANNEL CALIBRATION.

3/4.3.6 RECIRCULATION PINP TRIP ACTUATION INSTRIMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the plant i to this postulated event falls within an envelope of study events given in I

General Electric Company Topical Report NEDO-10349, dated March, 1971.

The end-of-cycle recirculation pump trip (EOC-RPT) system is a part of the Reactor Protection System and is a safety supplement to the reactor trip. The purpose of the EOC-RPT is to recover the loss of thermal margin which occurs at the end-of-cycle. The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity to the reactor system at a faster rate than the control rods add negative scram reactivtcy. Each EOC-RPT system trips both recirculation pumps, reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events for which the EOC-RPT protective feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.

A fast closure sensor from each of two turbine control valves provides input to one EOC-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second EOC-RPT system.

Similarly, a position switch for each of two turbine stop valves provides input to one EOC-RPT system; a position switch for each of the other two turbine stop valves provides input to the other EOC-RPT system. For each EOC-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for closure of the turbine stop valves. The operation of either logic will actuate the EOC-RPT system and trip both recirculation pumps.

Each EOC-RPT system may be manually bypassed by use of a keyswitch which is administratively controlled. The manual bypasses and the automatic operating bypass at < 30% of RATED THERMAL POWER are annunciated in the control room.

BRUNSWICK - UNIT 2 B 3/4 3-6 Amendment No.

1 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 after dilution in the discharge canal 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 not 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 (submerston) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection i (ICRP), Publication 2.

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

Detailed discussion of the LLD and other detection limits can be found in HASL Procedures Manuals, HASL-300 (revised annually), Currie, L. A. " Limits for Qualitative Detection and Quantitative Determination - Application to Radio-chemistry" 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).

, "Without delay" implies that the operator, upon determining the limiting condition for operation is being exceeded, takes the next appropriate action to comply with the specification.

Note that for batch releases, recirculation of at least two tank volumes shall be considered adequate for thorough mixing.

The stabilization pond and service water liquid release types represent potential release pathways and not actual release pathways. Surveillance of

, these pathways it intended to alert the plant to a potential problem; analysis for principal gamma emitters should be sufficient to meet this intent. If analysis for principal gamma emitters indicates a problem (i.e., exceeds the trigger level of 5x10~ UCi/ml),thencompletesamplingandgnalysesshallbe pe'r formed as per Table 4.11.1-2. The trigger level of 5x10~ pCi/mi was chosen as being sufficient to provide reasonable assurance of accountability of all nuclides released based upon lower limits of detection and expected concentrations.

3/4.11.1.2 DOSE - LIQUID EFFLUENTS 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 BRUNSWICK - UNIT 2 B 3/4 11-1 Amendscac No.

I l

RADIOACTIVE-EFFLUENTS BASES DOSES (Continued) 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 of 10 CFR Part 50 to assure that releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by 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 will be consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113,

, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine

.leactor Releases for the Purpose of Implementing Appendix I," April 1977.

The dose or dose commitment to a MEMBER OF THE PUBLIC is based on the 10 CFR Part 50, Appendix I, guideline of:

a. 1.5 area to the total body and 5.0 mrea to any organ during any calendar quarter, and

b. 3 area to the total body and 10 mrea to any organ during any calen-dar year, from radioactive material in liquid effluents from each reactor unit to UNRE-STRICTED AREAS. This specification is written for a two unit site.

3/4.11.1.3 LIQUID RADWASTE TREATMENT SYSTEM The requirement that appropriate portions of this system be used, when speci-fled, provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as reasonably achievable." This specification j implements the requirements of 10 CFR Part 50.36a, General Design Criteria 60 l of Appendix A to 10 CFR Part 50 and the design objectives given in Section I

II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment systen 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.

Mechanical filtration as per system design is considered to be an appropriate i

component of the liquid radwaste treatment system. .

The requirements of 0.12 mrem total body or 0.4 mrem to any organ in a 31-day i

period is based on two reactor units having a shared liquid radwaste treatment system.

l BRUNSWICK - UNIT 2 B 3/4 11-2 Amendment No.

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

RADIOACTIVE EFFLUENTS BASES 3/4.11.1.4 LIOUID 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 do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system with the exception of the auxiliary surge tank. The auxiliary surge tank is excluded from this specification because the tank and its associated piping are all Seismic Class I.

Since the condensate storage tanks have continuous influent and effluent, stratification should not occur. Samples taken from the operating condensate transfer pump (s) vent or drain shall be deemed representativa of this system.

"Without delay" implies that the operator, upon determining the limiting condition for operation is being exceeded, takes the next appropriate action to comply with the specification.

3/4.11.2 GASEOUS EFFLUENTS

3/4.11.2.1 DOSE RATE This specification is provided to ensure that the dose rate at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose rate limits of 10 CFR Part 20 for UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the 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 that MEMBER OF THE PUBLIC will be sufficiently low to compensate for any increase in the atmospheric diffusion i

factor above that for the SITE BOUNDARY. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC 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 mrees/ 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.

l Thic specification applies to the release of gaseous effluents from all reactors at the site.

With regard to footnotes (c) and (g) of Table 4.11.2-1, (1) to determine whether the DOSE EQUIVALENT I-131 concentration in the primary coolant has increased by more than a factor of 3, the iodine-131 analysis performed af ter the transient will be compared to the most recent routine analysis for DOSE EQUIVALENT I-131 concentration performed before the transient; and (2) to determine whether the main condenser air ejector noble gas monitor has increased by more than a factor of 3, the activity indicated on the monitor's i

l BRUNSWICK - UNIT 2 B 3/4 11-3 Amendment No.

RADIOACTIVE EFFLUENTS BASES DOSC RATE (Continued) chart recorder af ter the transient will be compared to the activity indicated I on the recorder just before the transient occurred. i The required detection capabilities for radioactive materials in gaseous waste saapies are tabulated in teras of the Lower Limits of Detection (LLDs).

Detailed disc.ussion 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 Radio-chemistry" 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).

"Without delay" implies that the operator, upon determining the limiting condition for operation is being exceeded, takes the next appropriate action to comply with the specification.

l 3/4.11.2.2 DOSE-NOBLE GASES 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 state.nents provide the required operating flexibility and, at the same time, implement the guides set forth in Section IV.A of Appendix I, to assure 4

that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The

! Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through the apprcpriate pathways is unlikely to be substantially underestimated. The dose calculations

! established in the GCCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Men from Routine Releases of Reactor Effluents for the Purpose of j Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October i 1977 and Regulatory Guide 1.111 " Methods for Estimating Atmospheric Transport l and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for l determining the air doses et and beyond the SITE BOUNDARY will be based upon the historical annual average atmospheric conditions. NUREG-0133 provides

methods for dose calculations consistent with Regulatory Guides 1.109 and

! 1.111. The limits of this specification are twice the 10 CFR 50 Appendix I per reactor guidelines because they are written for a two unit site.

BRUNSWICK - UNIT 2 B 3/4 11-4 Amendment No.

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RADIOACTIVE EFFLUENTS BASES /

3/4.11.2.3 DOSE - IODINE-131. IODINE-133. TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM This specification is provided to implement the requirements of Section II.C, III.A, and IV.A of Appendix I,10 CFR Part 50. The Limiting 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, implements the guides set forth in Section IV. A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data

such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are required to be consistent with the methodology provided in Regulatory Guide 1.109, " Calculating of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evalu-ating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for deter-mining the actual doses based upon the historical average atmospheric condi-tions. The release rate specification for iodine-131, iodine-133, tritium, and radioactive material 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 which are examined in the development of these calculations are: (1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat producing animals graze, with consumption of the milk and meat by man, and (4) deposition on the ground with subsequent exposure of l

man. The limits of this specification are twice the 10 CFR 50 Appendix I per reactor guidelines because they are written for a two unit site.

/

3/4.11.2.4 GASEOUS RADWASTE TREATMENT SYSTEM This requirement provides reasonable assurance that the releases of l

radioactive materials in gareous effluents will be kept "as low as reasonably achievable." This specification Laplements 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.

Until such time as the Augmented Off-Gas Treatment System becomes operational at the Brunswick Steam Electric Plant, the GASEOUS RADWASTE TREATMENT SYSTEM

! shall refer to the 30-minute offgas holdup line and stack filter house BRUNSWICK - UNIT 2 B 3/4 11-5 Amendment No.

RADIOACTIVE EFFLUENTS BASES 1

3/4.11.2.4 GASEOUS RADWASTE TREATMENT SYSTEM (Continued) filtration. After the Augmented Off-Gas Treatment System becomes operational, the GASEOUS RADWASTE TREATMENT SYSTEM shall refer to the 30-minute offgas holdup line, stack filter house filtration, and the Augmented Off-Gas Treatment System.

3/4.11.2.5 VENTILATION EXHAUST TREATMENT SYSTEM l

l This requirement provides reasonable assurance that the releases of radio-active materials in gaseous effluents will be kept "as low as reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of 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. At the Brunswick Steam Electric Plant, the only VENTILATION EXHAUST TREATMENT SYSTEMS shall be those installed for the Turbine Buildings' ventilation.

l 3/4.11.2.6 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treatment system is main-tained belev the flammability limits of hydrogen. Maintaining the concentra-I tion of hydrogen below the 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.

This specification will become applicable when the Off-Gas Recombiners become fully operational (prior to operation of the Augmented Off-Gas Treatment System) at the Brunswick Steam Electric Plant. There is no requirement for hydrogen monitors on the 30-minute waste gas holdup line which will serve in the interim.

l 3/4.11.2.7 MAIN CONDENSER AIR EJECTOR RADIOACTIVITY RELEASE RATE I

Restricting the release rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to an individual at or beyond the exclusion boundary will not exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50. 243,600 microcuries/second is equal to 100 microcuries/second/MWt for a rated thermal power of 2,436 MWt.

l I

BRUNSWICK - CNIT 2 B 3/4 11-6 Amendment No.

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RADIOACTIVE EFFLUENTS BASES 3/4.11.2.8 DRYWELL VENTING OR PURGING This specification provides reasonable assurance that releases from drywell PURGING operations will not exceed the annual dose limits of 10 CFR Part 20 for INRESTRICTED AREAS.

3/4.11.3 SOLID RADIOACTIVE WASTE

, his specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste / liquid / solidification

! agent / catalyst ratios, waste oil content, waste principal chemical constit-uents, mixing, and curing times.

4 3/4.11.4 TOTAL DOSE (40 CFR PART 190)

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 generated radioactive effluents and direct radiation exceed 25 areas to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mress. 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. he l 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 j other nuclear fuel cycle facilities at the same site or within a radius of 8 I km must be considered. If the dose- to any MEMBER OF THE PUBLIC is estimated I

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 Specification 3/4.11. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

BRUNSWICK - INIT 2 B 3/4 11-7 Amendment No.

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3/4.12 RADIOLOGICAL FNVIRONMENTAL MONITORING RASES 3/4.12.1 MONITORING PROGRAM The radiological environmental monitoring program required by this specifica-tion provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest poten-tial radiation exposures of MEMBERS OF THE PUBLIC resulting from 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 are not higher than expected on the basis of effluent measurements and the modeling of the environmental exposure pathways.

The required detection capabilities for environmental sample analyses are tabulated in terms of the lower Limits of Detection (LLDs). The LLDs required by Table 4.12.1-1 are considered optimum for routine environmental measure-ments in industrial laboratories. It should be recognized that the LLD is defined as a, priori (before the fact) limit representing the capability of a measurement system and not as jL posteriori (af ter the fact) limit for a par-l ticular measurement.

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

J Groundwater is not monitored by this specification because plant liquid efflu-ents are not tapped as a source for drinking or irrigation purposes.

3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of area 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 the census. The best information from door-to-door surveys, aerial

, surveys, or consulting with local agricultural authorities shall be used.

l This 10 CFRcensus satisfies the requirements of Section IV.B.3 of Appendix I go 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/yr) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a cLild. To determine the minimum garden size, the following assumptions were made: (1) 20% of the garden was used for growing broadleaf vegetatiog (i.e., similar to lettuce and cabbage; and (2) a vegetation yield of 2 kg/m BRUNSWICK - UNIT 1 B 3/4 12-1 Amendment No.

RADIOLOGICAL ENVIRONMENTAL MONITORING

.B.ASES 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in the 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 monitor-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.

i i

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l BRUNSWICK - UNIT 1 B 3/4 12-2 Amendment No.

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

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

SITE BOUNDARY 5.1.3 The SITE BOUNDARY shall be as shown in Figure 5.1.3-1. Foi the purpose of effluent release calculations, the boundary for atmospheric releases is the SITE BOUNDARY and the boundary for liquid releases is the SITE BOUNDARY prior to dilution in the Atlantic Ocean.

. 5.2 CONTAINMENT CONFIGURATION 5.2.1 The PRIMARY CONTAINMENT is a steel-liced, reinforced concrete structure composed of a series of vertical right cylinders and truncated cones which form a drywell. This drywell is attached to a suppression chamber through a series of vents. The suppression chamber is a concrete, steel-lined pressure vessel in the shape of a torus. The primary containment has a minimum free air volume of 288,000 cubic feet.

DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is designed and shall be maintained for:

a. Maximum internal pressure 62 psig.

b. Maximum internal temperature: drywell 300*F Suppression chamber 200*F

c. Maximum external pressure .2 psig.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 560 fuel assemblies with each 7 x 7 fuel assembly containing 49 fuel rods, each 8 x 8 fuel assembly containing 63 fuel

! rods; and each 8 x 8R fuel assembly containing 62 fuel rods. All fuel rods shall be clad with Zirealoy 2. The nominal active fuel length of each fuel rod shall be 144 inches for 7 x 7 fuel assemblies, 146 inches for 8 x 8 fuel assemblies, and 150 inches for 8 x 8R fuel assemblies. Each fuel rod shall contain a maximum total weight of 4430 grams of UO 2' BRUNSWICK - UNIT 2 5-1 Amendment No.

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