Text

Proposed Tech Spec Change 34A,revising Sections 3,4,5 & 6 of Apps a & B,To Incorporate Code Requirements in Response to NRC 781115 Request
	ML19347C386
Person / Time
Site:	Three Mile Island
Issue date:	10/03/1980
From:	; METROPOLITAN EDISON CO.
To:	;
Shared Package
ML19347C382	List: ML19347C381; ML19347C385; ML19347C386;
References
	NUDOCS 8010170518
	Download: ML19347C386 (100)
	v • d • e

{{#Wiki_filter:.-_ _ ( , , v TABLE OF CONTENTS Section Page TECHNICAL SPECIFIATIONS 1 DEFINITIONS 1-1 1.1 RATED POWER l-1 1.2 REACTOR OPERATING CONDTTIONS 1-1 1.2.1 COLD SHUTDOWN , 1-1 1.2.2 HOT SHUTDOWN 1-1 1.2.3 REACTOR CRITICAL l-1 1.2.4 HOT STANDBY l-1 1.2.5 POWER OPERATION 1-1 1.2.6 REFUELING SHUIDOWN 1-1 1.2.7 REFUELING OPERATION 1-2 1.2.8 REFUELING INTERVAL l-2 1.2.9 STARTUP l-2 1.2.10 TAVG l-2 1.2.11 HEATUP-COOLDOWN MODE l-2 1.2.12 STATION, UNIT, PLANT, AND FACILITY l-2 1.3 OPERABLE l-2 1.4 PROTECTIVE INSTRlDIENTATTON LOGIC 1-2 1.4.1 INSTRUMENT CHANNEL l-2 1.4.2 REACTOR PROT /CTION SYSTEM l-2 1.4.3 PROTECTION CHANNEL l-3 1.4.4 REACTOR PROTECTION SYSTDI LOGIC 1-3 1.4.5 ENGINEERED SAFETY FEATURES SYSTEM 1-3 1.4.6 DEGREE OF REDUNDANCY l-3 1.5 INSTRlDfENTATION SURVEILLANCE l-3 1.5.1 TRIP TEST l-3 1.5.2 CHANNEL TEST l-3 1.5.3 INSTRUMENT CIL\NNEL CHECK 1-3 1.5.4 INSTRUMENT CHANNEL CALIBRATION 1-4 1.5.5 HEAT BALANCE CHECK 1-4 1.5.6 HEAT BALANCE CALIBRATION 1-4 1.6 POWER DISTRIBUTION 1-5 1.6.1 QUADRANT POWER TILT l-5 l 1.6.2 REACTOR POWER IMBALANCE 1-5 1.7 CONTAINMENT INTEGRTTY l-5 1.8 FTRE SUPRESjl(y WATER SYSTFit 1-5 1.9 CHANNEL CALIBRATION 1-6 1.10 CHANNEL CHECK 1-6 1.11 CHANNEL FUNCTIONAL TEST l-6 ; 1.19 DOSE EOUTVALENT I-l31 1-6 1.27 SOURCE CHECK 1-6 1.28 SOLIDIFICATIOli 1-6 1.29 0FFSITE DOSE CALCULATION MANUAL 1-7 1 1.30 PROCESS CONTROL PROGRAM l-7 1.31 CASEOUS RADWASTE TREATMENT SYSTDJ 1-7 1.32 VENTILATION EXHAUST TREATMENT SYSTEM 1-7 l 1.33 PURCE-PURCING l-7 1.34 VE:iTING 1-7 gQ i

TABLE OF CONTENTS Section Page 2 SAFETY LIMITS AND LIMITING SAFETY SYSTDI SETTINGS 2-1 2.1 SAFETY LIMITS. REACTOR CORE 2-1 2.2 SAFETY LIMITS. REACTOR SYSTDi PRESSURE 2-4 2.3 LIMITING SAFETY SYSTDI SETTINGS, PROTECTION 2-5 INSTRUMENTATTON 3 LIMITING CONDITIONS FOR OPERATION 3-1 3.1 REACTOR COOLANT SYSTDI 3-1 3.1.1 OPERATIONAL COMPONENTS 3-1 3.1.2 PRESSURIZATION,llEATUP, AND C00LDOWN LIMITATIONS 3-3 3.1.3 MINIMUM CONDITIONS FOR CRITICALITY 3-6 3.1.4 REACTOR COOLANT SYSTDI ACTIVITY 3-8 3.1.5 CHEMISTRY 3-10 3.1.6 LEARAGE 3-12 3.1.7 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY 3-16 3.1.8 SINGLE LOOP RESTRICTIONS 3-17 3.1.9 LOW POWER PHYSICS TESTING RESTRICTIONS 3-18 3.1.10 CONTROL ROD OPERATION 3-18a 3.1.11 REACTOR INTERNAL VENT VALVES 3-18b 3.2 MAKEUP AND PURIFICATION AND CHEMICAL ADDITION 3-19 SYSTEMS 3.3 EMERGENCY CORE COOLING. REACTOR BUILDING EMERGENCY 3-21 COOLING. AND REACTOR BUILDING SPRAY SYSTDIS 3.4 TURBINE CYCLE 3-25 3.5 INSTRUMENTATION SYSTEMS 3-27 3.5.1 OPERATIONAL SAFETY INSTRUMENTATION 3-27 3.5.2 CONTROL RO.' GROUP AND POWER DISTRIBUTION LIMITS 3-33 3.5.3 ENGINEERED SAFEGUARDS PROTECTION SYSTEM ACTUATION 3-37 SETPOINTS 3.5.4 INCORE INSTRUMENTATION 3-38 3.6 REACTOR BUILDING 3-41 3.7 UNIT ELECTRICAL POWER SYSTDI 3-42 3.8 FUEL LOADING AND REFUELING 3-44 3.9 RADI0 ACTIVE MATERIALS 3-46 3.10 MISCELLANEOUS RADIOACTIVE MATERIALS SOURCES 3-46 3.11 HANDLING OF IRRADIATED FUEL 3-55 3.12 REACTOR BUILDING POLAR CRANE 3-57 3.13 SECONDARY SYSTDI ACTIVITY 3-58 3.14 FLOOD 3-59 3.14.1 PERIODIC INSPECTION OF THE DIKES AROUND TMI 3-59 3.14.2 FLOOD CONDITION FOR PLACING THE UNIT IN 110T STANDBY 3-60 3.15 AIR TREATMENT iYSTEMS' 3-61 3.15.1 DIERGENCY CONTROL ROOM AIR TREATMENT SYSTD1 3-61 3.15.2 REACTOR EUILDING PURGE AIR TREATMENT SYSTDI 3-62a 3.15.3 AUNILIARY AND FUEL HANDLING EXHAUST AIR TREATMENT SYSTDI 3-62c 3.16 SHOCK SUPPRESSORS (SNUBBERS) 3-63 . 3,17 REACTOR BUILDING AIR TD!PERATURE 3-80 3.18 FIRE PROTECTION 3-86 3.18.1 FIRE DETECTION INSTRUMENTATION 3-86 3.18.2 FIRE SUPPRESSION WATER SYSTDI 3-88 3.18.3 DELUGE / SPRINKLER SYSTEMS 3-89 11

TABLE OF CONTENTS Section Page 3.3.3.8 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION (4.3.3.8.1) 3/4 3-95 3.3.3.9 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING 3/4 3-101 INSTRUMENTATION (4.3.3.9.1) 3.11.1.1 LIQUID EFFLUENTS (4.11.1.1.1, 4.11.1.1. 2, 4.11.1.1. 3) 3/4 3-112 3.11.1.2 DOSE (4.11.1.2) 3/4 3-116 3.11.1.3 LIQUID WASTE TREATMENT (4.11.1.3.1) 3/4 3-117 3.11.1.4 LIQUID HOLDUP TANKS (4.11.1.4) 3/4 3-118 3.11.2.1 DOSE RATE (4.11.2.1.1, 4.11.2.1.2) 3/4 3-119 3.11.2.2 DOSE, NOBLE GAS (4.11.2.2.1) 3/4 3-123 3.11.2.3 DOSE, RADI0 IODINES, RADI0 ACTIVE MATERIAL IN PARTICULATE 3/4 3-124 FORM AND RADIONUCLIDES OTHER THAN NOBLE CASES (4.11.2.3.1) , 3.11.2.4 GASEOUS RADWASTE TREATMENT (4.11.2.4.1, 4.11.2.4.2) 3/4 3-125 3.11.2.5 EXPLOSIVE GAS MIXTURE (4.11.2.5) 3/4 3-126 3.11.2.7 GAS STORAGE TANKS (4.11.2.7) 3/4 3-127 3.11.3.1 SOLID RADIOACTIVE WASTE (4.11. 3.1.1, 4.11. 3.1.2) 3/4 3-128 3.11.4 TOTAL DOSE (4.11.4) 3/4 3-130 3.12.1 MONITORING PROGRAM (4.12.1.1) 3/4 3-131 3.12.2 LAND USE CENSUS (4.12.2.1) 3/4 3-139 3.12.3 INTERLABORATORY COMPARISON PROGRAM (4.12.3) 3/4 3-140 BASES 3/4.3.3.8 RADIOACTIVE LIQUID B3/4 3-142 3/4.3.3.9 RADIOACTIVE GASEOUS B3/4 3-142 3/4.11.1.1 LIQUID EFFLUENTS B3/4 3-143 3/4.11.1.2 GASEOUS EFFLUENTS B3/4 3-144 3/4.11.3 SOLID RADIOACTIVE WASTE B3/4 3-147 3/4.12.1 MONITORING PROGRAM B3/4 3-148 3/4.12.2 LAND USE CENSUS B3/4 3'-148 iii

TABLE OF CONTENTS Section . Page 3.18.4 CO2 SYSTDI 3-90 4 SURVEILLANCE STANDARDS _ 4-1 4.1 OPERATIONAL SAFETY REVIEW 4-1 / 4.2 REACTOR COOLANT SYSTDI INSERVICE INSPECTION 4-11 4.3 TESTING FOLLOWING OPENING OF SYSTEM 4-28 4.4 REACTOR BUILDING 4-29 4.4.1 CONTAINMENT LEAKAGE TESTS 4-29 4.4.2 STRUCTURAL INTEGRITY - 4-35 4.4.3 HYDROGEN PURGE SYSTDI 4-37 4.5 DfERGFNCY LOADING SEQUENCE AND POWER TRANSFER, 4-39 DIERGENCY CORE COOLING SYSTEM AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 DIERGENCY LOADING SEQUENCE 4-39

4. 5. 2. DIERGENCY CORE COOLING SYSTDI 4-41 4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTDI 4-43 4.5.4 DECAY HEAT RD10 VAL SYSTDI LEAKAGE 4-45 4.6 DfERGENCY POWER SYSTEM PERTODIC TESTS 4-46 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4-48 4.7.1 CONTROL ROD DRIVE SYSTEM FUNCTIONAL TESTS 4-48 4.7.2 CONTROL ROD PROGRAM VERIFICATION 4-50 4.8 MAIN STEAM ISOLATION VALVES 4-51 4.9 EMERGENCY FEEDWATER PD1PS PERIODIC TE5 TING 4-52 4.9.1 TEST 4-52 4.9.2 ACCEPTANCE CRITERIA 4-52 4.10 RFACTIVITY ANOMALIES 4-53 4.11 RITE ENVIRONMENTAL RADIOACTIVITY SURVEY 4-54 4.12 AIR TREATMENT SYSTDIS 4-55 4.12.1 DIERGENCY CONTROL ROOM AIR TREATMENT SYSTEM 4-55 4.12.2 REACTOR BUILDING PURGE AIR TREATMENT SYSTDI 4-55b 4.12.3 AUXILIARY AND FUEL HANDLING EXHAUST AIR TREATMENT SYSTDI 4-55d 4.13 RADI0 ACTIVE MATERIALS SOURCES SURVEILLANCE 4-56 4.14 REACTOR BUILDING PURGE EXIIAUST SYSTDI 4-57 4.15 MAIN STEM! SYSTEM INSERVIGE INSPECTION 4-58 4.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE 4-59 4.17 SHOCK SUPPRESSORS (SNUBBERS) 4-60 4.18 FIRE PROTECTION SYSTDIS 4-72 4.18.1 FIRE PROTECTION INSTRUMENTS 4-72 4.18.2 FIRE SUPPRESSION WATER SYSTDI 4-73 4.18.3 DELUGE / SPRINKLER SYSTEM 4-74 4.18.4 CO2 SYSTEM 4-74 4.18.5 HALON SYSTDIS 4-75 4.18.6 HOSE STATIONS 4-76 4.19 OTSC TUBE INSERVICE INSPECTION 4-77 4.19.1 STEMI GENERATOR SAMPLE SELECTION AND INSPECTION 4-77 METHODS 4.19.2 STEMI GENERATOR TUBE SAMPLE SELECTION AND INSPECTION 4-77 4.19.3 INSPECTION FREQUENCIES 4-79 4.19.4 ACCEPTANCE CRITERIA 4-80 4.19.5 REPORTS 4-81 4.20 REACTOR BUILDING AIR TDIPERATURE 4-86 IV

TABLE OF CONTENTS Section Page 5 DESIGN FEATURES 5-1 5.1 SITE 5-1 5.2 CONTAINMENT 5-2 5.2.1 REACTOR BUILDING 5-2 5.2.2 REACTOR BUILDING ISOLATION SYSTDI 5-3 5.3 EACTOR 5-4 5.3.1 REACTOR CORE - 5-4 5.3.2 REACTOR COOLANT SYSTDI 5-4 ' 5.4 NEW'AND SPENT FUEL STORAGE FACILITIES 5-6 5.4.1 NEW FUEL STORAGE 5-6 5.4.2 SPENT FUEL STORiGE 5-6 5.5 AIR INTAKE TUNNEL FIRE PROTECTION SYSTEMS 5-8 6 ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION 6-2 6.2.1 0FFSITE 6-2 6.2.2 FACILITY STAFF 6-2 6.3 STATION STAFF QUALIFICATIONS 6-3 6.4 TRAINING 6-3 6.5 REVIEW AND AUDIT 6-3 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) 6-3 6.5.2.A MET-ED CORPORATE TECllNICAL SUPPORP STAFF 6-5 6.5.2.B GENERAL OFFICE REVIEW BOARD (CORB) 6-7 6.6 REPORTABLE OCCURRENCE ACTION 6-10 6.7 OCCURRENCES INVOLVING A SAFETY LIMIT VIOLATION 6 10a 6.8 PROCEDURES 6-11 6.9 REPORTING REQUIRD!ENTS 6-12 6.9.1 ROUTINE REPORTS 6-12 6.9.2 REPORTABLE OCCURRENCES 6-14 6.9.3 UNIQUE REPORTING REQUIREMENTS 6-18 ' 6.10 RECORD RETENTION 6-19 l 6.11 RADIATION PROTECTION PROGRAM 6-20 l 6.12 DELETED 6-20 ' 6.13 111C11 RADIATION AREA 6-21 ! 6.14 FIRE PROTECTION INSPECTIONS 6-26 ) 6.15 PROCESS CONTROL PROGRAM 6-27 i { 6.16 O_FFSITE DOSE CALCULATION MANUAL 6-27 ! 6.17 Kt10R Cl!ANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS 6-29 ' i v

LIST OF TABLES Table Title Page 1.2 Frequency Notation 1-8 l 2.3-1 Reactor Protection System Trip Setting Limits 2-9 3.5-1 Instruments operating Conditions 3-29 3.16-1 Safety Related Shock Suppressors (Snubbers) 3-65 3.18-1 Fire Detection Instruments 3-87 3.3-11 Radioactive Liquid Ef fluent !!onitoring Instrumentation 3/4 3-96 4.3-11 Radioactive Liquid Effluent Monitoring Instrumentation 3/4 3-99 Surveillance Requirements 3.3-12 Radioactive Caseous Effluent Monitoring Instrumentation 3/4 3-102 4.3-12 Radioactive Caseous Effluent Monitoring InstrumenLation 3/4 3-107 Surveillance Requirements 4.11-1 Radioactive Liquid Waste Sampling and Analysis Program 3/4 3-113 4.11-2 Radioactive Caseous Waste Samplir and Analysis Program 3/4 3-120 l 3.12-1 Radiological Environmental Monitoring Program 3/4 3-133 l I 3.12-2 Reporting Levels for Radioactivity Concentrations in 3/4 3-140 ! EnvirJacental Sample i l 4.12-1 Maximum Valves for the Lower Limits of Detection (LLD) 3/4 3-136 4.1-1 Instrument Surveillance Requirements 4-3 4.1-2 Minimum Equipment Test Frequency 4-8 l l 4.1-3 Minimum Sampling Frequency 4-9 4.2-1 Instrument Surveillance Program 4-14 4.2-2 Surveillance Capsule Insertion & Withdrawal Schedule at 4-27a I TMI-2 4.4-1 Selected Tendons and Corresponding Inspection Periods 4-35a 4.4-2 Tendons Selected for Tendon Physical Condition Test 4-36 4.4-3 Ring Girder Surveillance 4-36g vi

LIST OF TABLES Table Title Page 4.15-1 Radioactive Liquid Waste Sampling and Analysis 4-59 ) 4.15-2 Radioactive Gaseous Waste Sampling and Analysis 4-63 4.19-1 Minimum Number of Steam Generators to be Inspected 4-84 During Inservice Inspection 4.19-2 Steam Generator Tube Inspection 4-85 6.12-1 Deleted vil

LIST OF FIGURES Figure Title 2.1-1 TMI-l Core Protection Safety Limit 2.1-2 TMI-l Core Protection Safety Limits 2.1-3 TMI-l Core Protection Safety Bases 2.3-1 TMI-l Protection System MaRimum Allowable Set Points 2.3-2 Protection System Maximum Allowable Set Points for Reactor Power Imbalance, TMI-l 3.1-1 Reactor Coolant System Heat-up/Cooldown Limitations (Applicable to 5 EFPY) 3.1-2 Reactor Coolant System, Inservice Leak and Hydrostatic Test Limitations (Applicable to 5 EFPY) 3.1-3 Limiting Pressure vs. Temperature Curve for 100 STD cc/ Liter H 0 2 3.5-1 Incore Instrumentation Specification Axial Imbalance Indication, TMI-1 3.5-2 Incore Instrumentation Specification Radial Flux Tilt Indication, TMI-l 3.5-2A Rod Position Limits for 4 Pump Operation From 0 to 125 1 5 EFPD, TMI-1 3.5-2B Rod Position Limits for 4 Pump Operation from 125 1 5 EFPD to EOC, TMI-l 1 3.5-2C Rod Position Limits for 2 and 3 Pump Operation from 0 to 125 i 1 5 EFPD, TMI-l 3.5-2D Rod Position Limite for 2 and 3 Pump Operation from 125 5 EFPD to ! EOC, TMI-l 3.5-2E Power Imbalance Envelope for Operation from 0 EFPD to EOC l I l l l 1 viii l l l l

LIST OF FIGURE-i Figure Title 3.5-2F Deleted 3.5-2G LOCA Limited Maximum Allowable Linear Heat-TMI-l

3.5-2H APSR Position Limits for Operation from 0 EFPD to EOC 3.5-2I Deleted 3.5-2J Deleted 3.5-2K Deleted 3.5-2L Deleted 3.5-2M Deleted 3.5-2N Deleted 3.5-3 Incore Instrumentation Specification, TMI-l 4.2-1 Equipment and Piping Requiring Inservice Inspection in Accordance with Section XI of the ASME Code 4.4-1 Ring Girder Surveillance, TMI-l 4.4-2 Ring Girder Surveillance Crack Pattern Chart, TMI-l 4.4-3 Ring Girder Surveillance Crack Pattern Chart, TMI-l 4.4-4 Ring Girder Surveillance Crack Pattern Chart, TMI-1 4.4-5 Ring Girder Surveillance Crack Pattern Chart, TML-1 j 5-1 Extended Plot Plan TMI 5-2 Site Topography 5 Mile Radius 5-3 Site Boundary for Gaseous Effluents 5-4 Site Boundary for Liquid Effluents ) l 6-1 Met-Ed Corporate Technical Support Staff and Station Organization Chart I ix

1.0 DEFINITIONS Dose Equivalent I-131 i 1.19 The DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, 1-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID 14844. " Calculation of Distance Factors for Power and Test Reactor Sites".

SOURCE CHECK 1.27 A SOURCE CHECK shall be the qualitative assesstaent of channel response when the channel sensor is exposed to a radioactive source. SOLTDIFICATION 1.28 SOLIDIFICATION shall be the conversion of radioactive wastes from liquid treatment systems to a uniformly distributed, ronolithic immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing). l l I TMI 1-6

OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.29 An OFFSITE DOSE CALCULATION MANUAL (ODCM) shall be a manual containing the methodology and parameters to be used in the calculation of of f-site doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints. PROCESS CONTROL PROGRAM (PCP) 1.30 The PROCESS CONTROL PROGRAM shall contain the sampling, analysis and formulation determination by which SOLIDIFICATION of radioactive wastes from 11guld systems is assured. . CASEOUS RADWASTE TREATMENT SYSTEM 1.31 A GASEOUS RADWASTE TREATFENT 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. VENTILATION EXHAUST TREATMENT SYSTEM 1.32 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in ef fluents by passing ventilation or vent exhaust gases through charcoal absorbers and/or HEPA filters for the purpose of removing lodines or particulates from the gaseous exhaust stream prior to the release to the environment. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILLATION EXHAUST TREATMENT SYSTEt15. PURGE - PURGING 1.33 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions in such a manner that replacement air or gas is required to purify the confinement. VENTING 1.34 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions in such a manner that replacement air or gas is not provided or required during VENTING. Vent used in sys&em names does not imply a VENTING process. TMI-1 1-7

                                    .                                             m   _

l TABLE 1.2 FREOUENCY NOTATION

NOTATION FREOUENCY S Shiftly (once per 12 hours)

D Daily (once per 24 .. trs) W Ifeekly (once per 7 days) M Monthly ( once per 31 days 1 i Q Quarterly (once per 92 days) SA Semi-annually (once per 184 days) l R Refueling interval (once per 18 months) S/U Prior to each reactor startup P Completed Prior to each release l N/A Not applicable l

Specified interval may be adjusted plus a minimum 25% to accommodate test schedule.

I I 1 1 \ l l

 .                                        1-8

INSTRUMENTATION RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-11 shall be OPERABLE with their alarm / trip setpoints act to ensure that the limits of Specification 3.11.1.1 are not exceeded. The ALARM / TRIP setpoints of these channels shall be determined in accordance with the Of fsite Doce Calculation Manual (ODCM). APPLICABILITY: At all times * - ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid ef fluents monitored by the af fected channel or declare the channel inoperable,

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels operable, take the ACTION shown in Table 3.3-11.

SURVEILLANCE REQUIREMENTS 4.3.3.8.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-11.

 *For RM-L6, and FT-84 operability is not required when discharges are positively controlled through the closure of WDL-V 257 and PM-L7 is operable.

i l TMI-1 3/4 3-95 t t

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TABLE 3.3-11 (Continued)

                                                                              , RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT                                                                          OPERABLE                        ACTION

3. Flow Rate Measurement Devices

a. Unit 1 Liquid Radwaste Effluent Line 1 21 (FT-84)

b. Station Effluent Discharge 1 21 (FT-146)

E e m 1 l 7

p

TABLE 3.3-11 (Continued) TABLE NOTATION ACTION 18 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, ef fluent releases may be resumed for up to 14 days, provided that prior to initiating a release:

1. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.3, and;

2. At least two technically qualified members of the Unit staff independently verify the release rate calculations and verif the discharge valve lineup.

3. Manager Unit 1 chall approve each release.

Otherwise, suspend release of rac4oactive effluents via this pathway. ACTION 20 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that, at least once per 8 hours, grab samples are collected and analyzed for gross radioactivity (beta and gamma) at a limit of detection of at least 10-7 mic rocuries/ml . ACTION 21 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, radioactive ef fluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours during actual releases. Pump curves may be used to estimate floiv. I l l 1 TMI-1 3/4 3-98

TABLE 4.3-11 RADIDACT.IVE LIQUID EFFLUENT MONITORING INSTRUf1ENTATION SURVEILLANCE REQUIREffENTS CilANNEL CilANNEL SOURCE Cl1ANNEL FUNCT10NAL INSTRUMENT CilECK CilECK CALIBRATION TEST

1. Radioactivity Monitors Providing Alarm and Automatic Isolation

a. Unit 1 Liquid Radwaste Effluents D P R(3) Q(1)

Line (RM-L6)

2. Flow Rate Monitors

a. Unit 1 Liquid Radwaste Effluent D(4) N/A R Q Line (FT-84) g

b. Station Effludnt Discharge D(4) N/A R Q (FT-146) R n

3. Cross Beta or Gamma Radioactivity Monitors Providing Alarm but not Providing Automatic Termination of Release

a. Station Ef fluent Line (RM-L7) D M R(3) Q(2) 7 E

r

TABLE 4.3-11 (Continued) TABLE NOTATICN P (1) The CHANNEL FUNCTION TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if the following condition exists:

1. Instrument indicates measured levels above the high alarm / trip setpoint. (Includes - circtut failure)

2. Instrument indicates a down scale failure. (Alarm function only.)

(Includes - circuit failure)

3. Instrument controls moved from the operate mode (Alarm function only).

(2) The CHANNEL FUNCTIONA'L 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 setpoint.

(Includes circuit failure).

2. Instrument indicctes a down scale failure (Includes - circuit failure).

3. Instrurnent controls moved from the operate mode.

(3) The initial CHANNEL CALIBRATION for radioactivity measurement instrumen-tation 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 should 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 calibra-tion should be used. (Operating plants may substitute previously established calibration pr ocedures for this' requi 'ement) (4) CHANNEL CHECK shall consist of verifying indication of flow 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, l l l 1 IMI-1 3/4 3-100 l

INSTRUMENTATION RADI0 ACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.9 The radioactive gaseous process and effluent monitoring instrumen-tation channels shown in Table 3.3-12 shall be OPERABLE with their Alarm / Trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The Alarm / Trip setpoints of these channels shall be determined in 2ccordance with the ODCM. APPLICABILITY: As shown in Table 3.3-T2: ACTION:

a. With a radioactive gaseous process or effluent monitoring instrumen-tation channel alarm trip setpoint less conservative than required by the above, immediately suspend the release of radioactive effluents monitored by affected channel or declare the channel inoperable.

b. With less than the minimum number of radioactive gaseous process or effluent monitoring instrumentation channels operabic, take the ACTION shown in Table 3.3-12.

SURVEILLANCE REQUIREMENTS 4.3.3.9.1 Each radioactive gaseous process or effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3.12. TMI-1 3/4 3-101

TABLE 3.3-12 RADIDACTIVE GASEGUS EFFLUENT tiONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION

1. Waste Gas Holdup System

a. Noble Gas Activity Monitor 1 *** 25 (RM-A7)

b. Effluent System Flow Rate 1 *** 26 Measuring Device (FT-46)

2. Waste Gas lloldup System Explosive Gas Monitoring System

a. Hydrogen Monitor 2 *

30 E

b. Oxygen Monitof 2 *

30 4

n 4 4 4 E e

TABLE 3.3-12 (Continued) RADI0 ACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION

3. Containment Purge Monitoring System

a. Noble Gas Activity Monitor (RM-A9) 1

27

b. Iodine Sampler (RM-A9) 1 31

c. Particulate Sampler (RM-A9) 1

31

d. Effluent System Flow Rate 1

26 Measuring Device (FR-148)
26

e. Sampler Flow Rate Monitor 1 8

7 n n E

TABLE 3.3-12 (Continued) RADIDACTIVE GASEOUS EFFLUENT 110NITGn

G INSTRUMENTATION MINIMutt CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACIIGN

4. Condenser Vent System

a. Noble Gas Activity Monitor (RM-AS) 1

27 8

7 m n M A iE

                                                                                                                                                                           =.       -                      _ _ _ ..

TABLE 3.3-12 (Continued) R@l0 ACTIVE GASEQUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION

5. Auxiliary and Fuel Handling Building Ventilation System

27

a. Noble Gas Activity 1 Monitor (RM-A8) or (RM-A4 and RM-A6)

b. Iodine Sampler (RM-AO) or (RM-A4 and 1

31 RM-A6.)
31

c. Particulate Sampler 1 (RM-A8) or (RM-A4 and RM-A6)

d. Ef fluent System Flow 1

26 Rate Measuring Device 'o]

(FT-151) or (FT-149 and FT-150), A

e. Sanpler Flow Rate Monitor 1

26 g E

TABLE 3.3-12 (Continued) TABLE NOTATION

         *At all times.
        **During waste gas isoldup system operation.
       ***0perability is not required when discharges are positively controlled through the enclosure of WO6-V47 and RM-A8 and F-151 are operable.

ACTION 25 With the number of channels OPERABLE less than required by the Minimura Channels OPERABLE requirement, the contents of the tank may be released to the environment for up to 14 days provided that prior to initiating the release: i

1. At least two independent samples of the tank's contents are analyzed, and

2. At least two technically qualified members of the unit staff independently verify the release rate calculations and verify the discharge valve lineup.

3. The Manager Unit 1 shall approve each release.

Otherwise, suspend release of radioactive effluents via this pathway. ACTION 26 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 28 days provided the flow rate is estimated at least once per 4 hours. ACTION 27 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 28 days nrovided grab samples are taken at least once per 8 hours and these samples are analyzed for gross activity within 24 hours. ACTION 30 With ..he number of channels OPERABLE one less than raquired by the Minimum Channels OPERABLE requirement, operation of this system may continue for up to 14 days. With both channels inoperable, be in at least HOT STANDBY within 6 hours. ACTION 31 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 28 days, provided samples are contin-uously collected with auxiliary sampling equipment. TMI-l 3/4 3-106

TABLE 4.3-12 RADIDAl'TIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Sl'RVEILLANCE REQUIREf TENTS CHANNEL CHANNELS SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST APPLICABILITY

1. Waste Gas Holdup System (RM-A7)

a. Noble Gas Activity Monitor P P R(3) Q(1) * * *

b. System Effluent Flow Rate * *

P N/A R Q teasuring Device

2. Waste Gas Holdup System Explosive Gas Monitoring System

a. Hydrogen Monitor D N/A Q(4) M

b. Oxygen Monitor **

D N/A Q(5) M S 7 I 5

TABLE 4.3-12 (Continued) RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CilANNEL CHANNELS SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST APPLICABILITY

3. Containment Purge Vent System

a. Noble Gas Activity Monitor (RM-A9) D P R(3) M(1) *

b. Iodine Sampler (RM-A9) W N/A N/A N/A *

c. Particulate Sampler (RM-A9) W N/A N/A N/A *

d. System Effluent flow Rate D N/A R Q Measuring Device,

e. Sampler Flow Rate Monitor D N/A R N/A

m e-t

                                                                                                                                             ,A iE

TABLE 4.3-12 (Continued) RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNELS GURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK , CHECK CALIBRATION TEST APPLICABILITY

4. Condenser Vent System

a. Noble Gas Activity Monitor (RM-AS: D M R(3) Q(2)

O eA A

n 5

                                                                                                                  .E e

TABLE 4.3-12 (Continued) RADIDACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CilANNEL CilANNELS SOURCE CHANNEL FUNCIIONAL INSTRUMENT CHECK CilECK CALIBRATION TEST _ APPLICABILITY S. Auxiliary and fuel Handling Building Ventilatten System

a. Noble Gas Activity Monitor (RM-A8) D M R(3) Q(1) or (RM-A4 or RM- A6)

b. Iodine Sampler (RM-A8) or (RM-A4 and W N/A N/A N/A RM-A6)

c. Particulate Sampler (RM-A8)or (RM-A4 W N/A N/A N/A and RM-A6)

d. System Effluent Flow Rate D N/A R Q Measurement Devices

e. Sampler flow Rate D N/A R Q Measurement Device P"4 n

7 n L I e-

TABLE 4.3-12 (Continued) TABLE NOTATION

          *At all times.
        **During waste gas holdup system operation.
       ***0perability is not required when discharges are positively controlled through the closure of WDC-V47 and RM-A8 and FT-151 are operable.

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway (for DLA8 or RM*A4 and FM-A6 only supply ventalation is isolated) and control room alarm annunciation occurs if the following condirion exists:

1. Instrument indicates measured levels above the high alarm setpoint (Includes circuit failure).

2. Instrument indicates a downscale failure. (Alarm function only)

(Includes circuit failure).

3. Instrument controls moved from the operate mode. (Alarm function only)

(2) 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 setpoint.

(Includes circuit failure)

2. Instrement indicates a downscale failure. (Includes circuit failure)

3. Instrument controls moved from the operate mode.

(3) The initial CHANNEL CALIBRATION for radioactivity measurement instrumentation 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 should 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 should be used. (Operating plants may substitute previously established calibration procedures for this requirement) (4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

1. One volume percent hydrogen, balance nitrogen, and

2. Four volume percent hydrogen, balance nitrogen.

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

1. One volume perce* oxygen, balance nitrogen, and

2. Four volume percent oxygen, balance nitrogen.

TMI-l 3/4 3- 111

3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIOUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION 3.11.1.1 The concentration of radioactive material released at anytime from the site to unrestricted areas (see Figure 5-4) shall be limited to the concen-trations specified in 10 CFR Part 20, Appendix B, Table II Column 2 for radio-nuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 3 x 10-3 uCi/ml - total activity. APPLICABILITY: At all times. ACTION: 1 i

a. With the concentration of radioactive material released from the site to unrestricted areas exceeding the above limits, immediately restore concen-tration within the above limits.

b. If action "a" cannot be met, then be in:

1. At 1 cast HOT STANDBY within 1 hour,

2. At least HOT SHUTDOWN within the next 6 hours, and

3. At least COLD SHUTDOWN within the following 30 hours.

SURVEILLANCE REQUIREMENTS 4.11.1.1.1 The radioactivity content of each batch of radioactivity liquid waste shall be determined prior to release by sampling and analysis in accordance with Table 4.11-1. The results of pre-release analyses shall be used with the calculational methods in the ODCM to assure that the concentration at the point of release is maintained within the limits of Specification 3.11.1.1. 4.11.1.1.2 Post-release analysis of samples composited from batch releases shall be performed in accordance with Table 4.11.1. The results of the previous post-release analysis shall be used with the calculational methods in the ODCM to assure that the concentrations at the point of release were maintained within the limits of Specification 3.11.1.1. 4.11.1.1.3 The radioactivity concentration of liquids discharged from continuous release points shall be determined by collection and analysis of samples in accordance with Table 4.11-1. The results of the analyses shall be used with the calculational methods of the ODCM to assure that the concentration at the point of release are maintained within the limits of Specification 3.11.1.1. TMI-1 3/4 3-112

TABLE 4.11-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Sampling Minimum Lower Limit Liquid Release Type Frequency Analysis Type of Activity of Detection Frequency Analysis (LLD) (uCi/ml)a P P H-3 1 x 10-> A.1 Batch Waste Each batch Each batch Principal Gamma 5 x 10-7 Release Tankse , d Emitters 9' f ,

                                             .                   I-131          1 x 10-6 One Batch /M          M        Dissolved and      1 x 10-4 Entrained Gases (Gamma Emitters)

P Q Gross Beta E 5 x 10-8 Each Batch Composite C Gross alpha 1 x 10-7 P-32 1 x 10-6 Sr-89, Sr-90 5 x 10-8 Fe-55 1 x 10-6 W A.2 Continuous ContinuousC Composite C Principal Gamma 5 x 10-7 Releases Emitters 9 f (RML-7) I-131 1 x 10-6 l M GRAB Sample M Dissolved an' 1 x 10-5 1 Entrained Cases I (gamma emitters) M H-3 1x 10-3 ) Continuouse Composite C l Gross alpha 1 x 10-7 i C l Continuouse Composite c Sr-89, Sr-90 5 x 10-8 l I Fe-55 1 x 10-6 P-32 1 x 10-6 i TMI-1 3/4 3-113

TABLE 4.11-1 (Continued) TABLE NOTATION

a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95?4 probability with S?e concluding that a blank observation represents 1 "(probability of falselyreal" For a particular measurement system (which may include radiochemical separation):

4.66 sb LLD = E x V x 2.22 x 106 x Y. x exp (- A AC) Where: LLD is the "a priori" lower limit of detection as defined above (as microcurie per unit mass or volume), sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute), E is the counting ef ficiency (as counts per transfor nation), V 3s the sample size (in units of mass or volume), 2.22 x 106 is the number of transformations per minute per micro-curie, Y is the fractional radiochemical yield (when applicable), X is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples). The value of ob used in the calculation of LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appro-priate) rather than on an unverified theoretically predicted variance. Typical values of E, V, Y, and at shall be used in the calculation,

b. A camposite 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 campling employed results in a specimen which is repre-sentative of the liquids released.

c. To be representative of th ' quantities and concentrations of radioactive materials in liquid ef fluents, samples shall be collected continuously in proportion to the rate of flow of the ef fluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sarrpie to be representative of the effluent release.

TMI-1 3/4 3- 114

TABLE 4.11-1 (Continued) TABLE NOTATION

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

Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed, by a method described in the ODCM, to assure represen-tative sampling,

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

f. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60,-

Zn-65, Mo-99, Cs-134, Cs-137, Cc-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and indentifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses shall be reported as " lens than" the nuclide's LLD, and shall not be reported as being present at the LLD level for that nuclide. The "less than" values shall not be used in the required dose calculations.

g. If the calculated gross beta concentration exceeds 1 x 10-7 microcuries/ml, an isotopic analysis shall be performed to determir;c either the concentra-

 ,        tion of Sr-89 and of Sr-90, or the concentration of gross strontium assuming that all strontium present is Sr-90.

l 1 TMI-1 3/4 3- 115 f

                                                                   . . ~ _     , - -

, RADI0 ACTIVE EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION e 3.11.1.2 The dose or dose commitment to an individual from radioactive materials i in liquid effluents released from the unit to the site boundary (see Figure 5-4) shall be limited:

a. During any calendar quarter t,o 11.5 mrem to the total body and to 15 mrem to any organ,

b. During any calendar year to 13 mrem to the total body and to $10 mr m to any organ.

APPLICABILITY: At all times. ACTION: I

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding

any of the above limits, in lieu of any other report required by Specification 6.9.1 prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) j and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and during the subsequent 3 calendar i quarters so that the cumulative dose or dose committment to any l individual from such releases during these four calerdar quarters is .

within 3 meem to the total body and 10 mrem to any organ. This i Special Report shall also include (1) the result of radiological ) analyses of the drinking water source, and (2) the radiological impact on finished drinking water supplies with regards to the requirements , of 40 CFR 141, Safe Drinking Water Act. I SURVEILLANCE REQUIREMENTS 4.11.1.2. Doce Calculations. Cumulative dose contributions from liquid effluents shall be determined in accordance with the Offsite Dose Calculation Manual (00CM) at least once a conth. TH1-1 3/4 3- 116 l

RADI0 ACTIVE EFFLUENTS LIQUID WASTE TREATMENT LIMITING CONDITION FOR OPERATION 1 3.11.1.3 The liquid radwaste treatment system shall be OPERABLE. The appro-priate portions of the system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected dose- 'tna to the liquid effluent from tha site (see figure 5-4) when averaged over 31 days, would exceed 0.06 mrem to the total body or 0.2 mrem to any organ. APPLICABILITY: At all times. . AETION:

a. With the liquid radwaste treatment system inoperable for more than 31 days or with radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days a Special P.epor t which 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 ' l

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

l l SURVEILLANCE REQL'IREMENTS 4.11.1.3.1 Doses due to liquid releases shall be projected at least once per 31 days, in accordance with the ODCM. 4.11.1.3.2 The liquid radwaste treatment system shall be demonstrated OPERABLE by operating the liquid radwaste treatment system equipment for at least 60 minutes quarterly unless the liquid radwaste system has been utilized to process radioactive liquid effluents durint the previous 92 days. 1 l 1 1 TMI-1 3/4 3- 117

RADIOACTIVE EFFLUENTS t LIOUID HOLDUP TANKS LIMITING CONDITION FOR OPERATIONS 1 3.11.1.4 The quantity of radioactive material contained in each of the following i tanks shall be limited to less than or equal to 10 curies, excluding tritium j and dissolved or entrained noble gases.

a. 'Outside temporary tank APPLICABILITY: At all times.

ACTION: 3 l

a. With the quantity of radioactive material in any of the above listed ;

tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents to within the limit. l SURVEILLANCE REQUIREMENTS i l 4.11.1.4 The quantity of radioactive material contained in each of the above l listed tanka shall be determined to be within the above limit by analyaing l a representative sample of the tank's contents weekly when radioactive l materials are being added to the tank. I I i 3/4 3-118

P DI0 ACTIVE 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 (see Figure 5-3) shall be limited to the following:

a. For nobic gases: < 500 mrem /yr to the total body and 1 3000 mrem /yr to the skin, and

b. For all radiciodines and for all radioactive materials in particulate form and radicauelides (other than noble gases) with half lives greater than 8 days: $ 1500 mrem /yr to any ocgan.

APPLICABILITY: At all times. ACTION: .

a. With the release rate (s) exceeding the above linits, immediately decrease the release rate to comply with the above limit (s)

b. If action "a" cannot be met, then be in:

1. At least HOT STANDBY within 1 hour

2. At least HOT SHUTDOWN within the next 6 hours, and

3. At least COLD SHUTDOWN within the following 30 hours.

SURVEILLANCE REOUIREMENTS 4.11.2.1.1 The dose rate due to nobic gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods and , I procedures of the ODCM. . 1 4.11.2.1.2 The dose rate of radioactive materials, other than noble gases,

in gaseous ef fluents shall be determined to be within the above limits in I accordance with methods and procedures of the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program, specified in Tabic 4.11-2.

l l l TMI-1 3/4 3- 119 l

TABLE 4.11-2 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Sampling Minimum Lower Limit Gaseous Release Type Frequency Analynis Type of Activity of Detection Frequency Analysis (LLD) (uCi/ml)a P P A. Waste Gas Each Tank Each Tank Principal Gamma 1 x 10-4 Storage Tank Grab Emitters 8 Sample P P B. Containment Each Purge b Each Purge b Principal Gamma 1 x 10-4 Purge Grab Emitters 9 Sample H-3 1 x 10-6 C. -Auxiliary and Mb ,c,e gb Principal Camma 1 x 10-4 Fuel Handling Grab Emitters 9 Building Sample Ventilation H-3 1 x 10-6 Wd D. All Release Continuousf Charcoal I-131 1 x 10-12 Type as Sample listed in A, B, I-133 1 x 10-10 C above. Wd Continuousf Particulate Principal Gamma 1 x 10-11 Emitters 9 (I-131, Others) Q Continuousf Composite Gross alpha 1 x 10-11 Particulate Sample O Continuousf Composite Sr-89h, Sr-90 1 x 10-11 Particulate Sample Continuouuf Noble Gns Noble Cases 1 x 10-6 Monite Gross Beta and Gamma E. Condenser vacuun Mb ,h Principal Gamma 1x 10-4 Pumps Exhaust h Grab sample Emitters

                                                   ,j-3 1                     1 x 10-6 TMI-1                               3/4 3-120

P TABLE 4.11-2 (Continued)

TABLE NOTATION

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

For a particular measurement system (which may include radiochemical separation): 4.66 sb LLD = . E x V x 2.22 x 106 x Y x exp (- Aat) . Where: LLD is the "a priori" lower limit of detection as defined above (as microcurie per unit mass or volume), ob is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute), E is the counting efficiency (as counts per transformation), V is the sample size (in units of mass or volume), 2.22 x 106 is the number of transformations per minute per micro-curie, Y is the fractional radiochemical yield (when applicable), A is the radioactive decay constant for the particular radionuclide, and l 6 t is the elapsed time between midpoint of sam-le collection and time of counting ( for plant effluents, not environmental samples). The value of sb used in the calculation of LLD for a detection system shall be based on the actual observed variance of the background counting rara or of the counting rate of tha blank samples (as appro-priate) ratner than on on unverified theoretically predicted variance. Typical values of E, V, Y, and a t shall be used in the calculation, i

b. Analyses shall also be performed following shutdown, startup or a thermal power level change exceeding 15% of RATED THERMAL POWER in a one hour period.

c. Tritium grab samples shall be taken at least on'3 per 24 hours wnen the ,

refueling canal is flooded. THI-1 3/4 3- 121

TABLE 4.11-2 (Continued) TABLE NOTATION

d. Samples shall be changed weekly and analyses shall be completed within 48 hours after changing (or after removal from sampler).

c. Tritium grab sampics shall be taken weekly from the ventilation exhaust from the spent fuel pool area whenever spent fuel is in the spent fuel pool,

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

g. The principal gamma emitters for which the LLD specification applies exclu-sively 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, 2n-65, Mo-99, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported.

Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses shall be reported as "less than" the nuclide's LLD and shall not be reported as being present at the LLD level for that nuclide. The "less than" values shall not be used in the required dose calculations,

h. Applicable only when condenser vacuum is established.

TMI-l 3/4 3- 122

RADI0 ACTIVE 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 unit (see -gure 5-4) shall be limited to the following:

a. During any calendar quarter: f, 5 mrad for gamma radiation and

           < 10 mrad for beta radiation.

b. During any calendar year: j,10 mrad for gamma radiation and < 20 mrad for beta radiation.

APPLICABILITY: At all times. ACTION: With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lie J of any other report required by Specification 6.9.2 prepare and submit to the Commission within 30 days, a Special Report which identifies the cause(s) for exceeding che limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive noble gases in gasecus effluents during the remainder of the current calendar quarter and during the subsequent 3 calendar quarters, so that the cumulative dose during these four calendar quarters is within 10 mrad for gamma radiation and 20 mrad for beta radiation. SURVEILLANCE REQUIREMENTS 4.11.2.2.1 Dose calculations Cumulative dose contributions for the current calendar quarter and current calendar year shall bc determined in accordance with the Offsite Dose Calculation Manual (ODCM) c,onthly. IMI-1 3/4 3- 123

RADI0 ACTIVE EFFLUENTS DOSE. RADI0 IODINES. RADI0 ACTIVE MATERIAL IN PARTICULATE FORM. AND RADIONUCLIDES OTHER THAN NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to an individual from radiciodines, radioactive materials in particulate form, and radionuclides (other than noble gases) with half-lives greater than 8 days in gaseous effluents released from the unit (See Figure 5-3) shall be limited to the following:

a. During any calendar quarter to f,7.5 mrem to any organ.

b. During any calendar year f15 mrem to any organ.

APPLICABILITY: At all times. ACTION: With the calculated dose from the release or radiciodines, radioactive materials in particulate form, or sadionuclides (other than noble gases) with half-lives greater than 8 days in gaseous effluents exceeding any of the above limits, in lieu of ar.y other report required by Specification 6.9.2 prepare and submit to the Commission within 30 days, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to reduce the releases of radiciodines, radioactive materials in particulate form, and radio-nuclides other than noble gases with half-lives greater than 8 days in gaseous effluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters so that the cumulative i dose or dose committment to an individual from such releases during i these four calendar quarters is within 15 mrem to any organ. SURVEILLANCE REQUIREMENTS l l 4.11.2.3.1 Dose Calculations Cumulative dose contributions for the current l calendar quarter and current calendar year shall be determined in accordance l with the ODCM monthly. i I 1 l TMI-1 3/4 3- 124

RADI0 ACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.2.4 The GASEGUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the GASEOUS RADWASTE TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous ef fluent air doses due to gaseous ef fluent releases from the site (see Figure 5-3), when averaged over 31 days, would exceed 0.2 mead for gamma radiation and 0.4 mrad for beta radiation. The appropriate portions of the VENTILATION EXHAUST TREATUENT 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 (see Figure S-3) when averaged over 31 days would exceed 0.3 mrem to any organ. APPLICABILITY: At all times. ACTION:

a. With the GASEOUS RACWASTE TREATMENT SYSTEM and/or the VENTILATION EXHAUST TREATMENT SYSTEM inoperable for more than 31 days or with gaseous waste being discharged without treatment and in excess of the above limits, in lieu of any other report required by Specifica-tion 6.9.2, prepare and submit to the Commission within 30 days, ,

a Special Report which 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.

I 1 SURVEILLANCE REQUIREMENTS l 4.11.2.4.1 Doses due to gaseous releases from the site shall be projected at least once per 31 days, in accordance with the ODCM. 4.11.2.4.2 The GASEGUS RADWASTE TREATMENT SYSTEM and VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated OPERABLE by operating the GASEOUS RADWASTE IREATMENT SYSTEM equipment and VENTILATION EXHAUST TREATMENT SYSTEM equipment for at least 60 minutes, at least Quarterly unless the appro-priate system has been utilized to process radioactive gaseous ef fluents during the previous 92 days. l TMI-1 3/4 3-125 1

RADIOACTIVE EFFLUENTS EXPLOSIVE baC 3' r0RE LIMITING CONI.s . ION FOR OPERATION 3.11.2.5 The concentration of oxygen in the waste gas holdup system shall be limited to less than or equal to 2% by volume whenever the hydrogen concen-tration exceeds 4% by volume. APPLICABILITY: At all times. . ACTION:

a. With the concentration of oxygen in the waste gas holdup systen greater than 2% by volume but less than or equal to 4% and the h'drogan concen-tration greater than 4% by volume, reduce the oxygen con.entration to the above limits within 48 hours,

b. With the concentration of oxygen in the waste gr3 holdup system greater than 4% by volume and the hydrogen concentration greater th,an 4 % by volume , immediately suspend all additions of waste gases to the system and reduce the, concentration of oxygen to less than or equal to 4% by volume within I hour and 2% by volume within 48 hours after initially exceeding 2% by volume.

l i SURVEILLANCE REQUIREMENTS - 4.11.2.5 The concentrations of hydrogen and oxygen in the waste gas holdup system shall be determined to be within the above limits by continuously monitoring the waste gases in the waste gas holdup system with the hydrogen and oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.10. l l l l TMI-l 3/4 3-126 1

RADI0 ACTIVE EFFLUENTS GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION l 3.11.2.7 The quantity of radioactivity contained in each gas storage tank shal) be limited to <8800 curies noble gases (considered as Xe-133). APPLICABILITY: At all times. . ACTION: ! With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents to within the limit. l SURVEILLANCE REQUIREMENTS 4.11.2.7 The concentration of radioactivity contained in each vent header shall be determined weekly. If the concentration of the vent header exceeds 10.7 uCi/cc, daily samples shall be taken of each tank being added to, to determine if the tank (s) is within the above limit. l l i l THI-1 3/4 3-127 e

RADI0 ACTIVE EFFLUENTS 3/4.11.3 SOLID RADI0 ACTIVE WASTE LIMITING CONDITION FOR OPERATION ' 3.11.3.1 The solid radwaste system shall be operable and used as applicable in the process control program used for the SOLIDIFICATION and packaging of radioactive wastes, and to ensure the meeting of the requirements of 10 CFR Part 20 and/or 10 CFR Part 71 prior to shipment of containers of radioactive wastes from the site. , APPLICABILITY: At all times. ACTION:

a. With the packaging requirements of 10 CFR Part 20 and/or,10 CFR Part 71 not satisfied, suspend shipments of defectively packaged solid radioactive wastes from the site.

b. With the solid radwaste system inoperable for more than 31 days, in lieu of ar, other report required by Specification 6.9.2, prepare and submit to the Commission within 30 days, a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystems and the reasons for inoperability,

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

3. A description of alternative used for SOLIDIFICATION and packaging of wastes, and

4. Summary description of action (s) taken to prevent a recurrence.

SURVEILLANCE 4.11.3.1.1 The solid radwaste system shall be demonstrated OPERABLE Quarterly. TMI-1 3/4 3- 128

SURVEILLANCE REQUIREFENTS (Continued) 0- Operating the solid radwaste system at least once in the previous 92 days in accordance with the PROCESS CONTROL PROGRAM or;

b. Verification of the existence of a valid contract for SOLIDIFICATION to be performed by a Contractor in accordance with a PROCESS CONTROL PROGRAM.

4.11.3.1.2 The Process Control Program shall be used to verify the SOLIDIFI-CATION of at least one representative fest specimen from at least every tenth batch of each type of radioactive waste required to be solidified by the Process Control Program,

a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFI-CATION of the batch under test shall be susperJed until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the Process Control Program, and a subsequent test verifies solidification. Solidi fi-cation of the batch may then be resumed using the alternative SOLID-IFICATION parameters determined by the Process Control Program.

b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the Process Control Program shell provide for the collection and testing of representative test specimens from each l consecutive batch of the same type of wet waste until 3 consecutive l initial test specimens demonstrate SOLIDIFICATION. The Proccas l Can' trol Program shall be modified as required, to assure SOLIDIFI-l CATION of subsequent batches of waste.

I l l l l IMI-1 3/4 3-129

RADI0 ACTIVE EFFLUENTS 3/4.11.4 TOTAL D0gr LIMITING CONDITION FOR OPERATION 3.11.4 The dose or dose commitment to any member of the public, due to releases of radioactivity and radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrem) over 12 consecutive months. . APPLICABILITY: At n11 times. ACTION: With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifications 3.11.1. 2. a , 3.11.1. 2. b , 3.11. 2. 2. 0, 3.11.2.2.b, 3.11.2.3.a, or 3.11.2.3.b, in lieu of any other report regt. ired by Specification 6.9. 2, prepare and submit a Special Report to the Director, Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, within 30 days, which defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of Specification 3.11.4. This Special Report shall include an analysis which estimates the radiation exposure (dose) to a member of the public from uranium fuel cycle sources (including all ef fluent pathways and direct radiation) for a 12 consecutive month period that includes the release (s) covered by this report. If the estimated dose (s) exceeds the limits of Specificatien 3.11.4, and if the release condition rasulting in violation of 40 CFR 190 has not already been corrected, the Spe-ial Report shall include a request for a variance in accordance with the provisions of 40 CFR 190 and including the specified information of sec.190.11(b). Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the requirements for dose limita-tion of 10 CFR Part 20, as addressed in other sections of this technical specification. SURVEILLANCE REQUIREMENTS 4.11.4 Dose Calculations Cumulative dose contributions from liquid and gaseous ef fluents 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. TMI 3/4 3- 130

3/4.12 RADICLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1 The radiolcgical environmental monitoring program shall be conc'ucted as specified in Table 3.12-1. APPLICABILITY: At all times. , ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12-1, prepare and submit to the Com-mission, in the Annual Radiological Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

b. With the level of radioactivity in an environmental sampling medium exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter, a report pursuant to 6.9.1.13. When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:

concentration (1) + concentration (2) + > 1.0 limit level (1) limit level (2) When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specification 3.11.1.2, 3.11.2.2 and 3.11.3.3. This reoort 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 j Radiological Environmental Operating Report. i 1 1 l l l l l TMI-1 3/4 3- 131 i

c. With milk or fresh leafy vegc& ables unavailable from one or more of the sample locations required by ra ble 3.12-1 in lieu of any other report required by Specfication 6.9.1 prepared and submit to the commission within 30 days. A Special Report which identifies the !

cause of the unavailability of samples and iocntifies locations for obtaining replacement samples. The locations fru.7 which samples were unavailable may then be deleted from those required by Table 3.12-1, provided the locations from which the replacemes:t samples were obtained are added to the environmental monitoring program as replacement locations. SURVEILLANCE REQUIREMENTS 4.12.1.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the locations given in the table and figure in the ODCM and shall be analyzed pursuant to the requirements of Table 3.12-1 and 4.12-1. l l TMI-1 3/4 3- 132 j

TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Number of Samples Exposure Pathway and Sampling and Type and frequency and/or Sample Sample Locations ** Collection Frequency of Analysis

1. AIRBORNE Radiciodine and A minimum of 5 locations Continuous operation of Radiciodine canister.

Particulates from Table 1 of the ODCM. sampler with sample col- Analyze at least once lection as required by per 7 days for I-131. dust loading but at least once per 7 days. Particulate sampler. Analyze for gross beta radioactivity > 24 hours following filter change. Perform gamma isotopic analysis on each sample m when gross beta activity C is > 10 times the calendar A yearly mean of control y samples. Perform gamma ;z isotopic analysis on composite (by location) sample at least once per 92 days.

2. DIRECT RADIATION A minimum of 38 locations At least once per 92 days. Gamma case. At least from Table 2 of the ODCM once per 92 days.

(using either 2 dosimeters or at least 1 instrument for continuously measuring and recording dose rate at each location). 7

    ** Sample locations are given on the figure and table in the ODCM.

TABLE 3.12-1 (Continued) RADIOLOGICAL ENVIRONMENTAL MONITORING PR0rRAli Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations ** Collection Frequency of Analysis

3. WATERBORNE

a. Surface A minimum of 2 locations Composite

cample collected Gamma isotopic analysis from Table 3 of the OnCM. over a period of < 31 days.

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

b. Drinking A minimum of 2 locations Composite

sample collected Gross beta and gamm-from Table 3 of the ODCM. Over a period of j:,31 days, isotopic analysis o.

each composite sample. sr Tritium analysis of C composite sample at least A once per 92 days. e

c. Sediment from A minimum of 2 locations At least once per 184 days. Gamma isotopic analysis Shoreline (1 Control and 1 Indicator) of each comple.

from Tabic 4 of the ODCM. o Composite samples shall be collected by collecting an aliquot at intervals not exceeding 24 hours. o* Sample locations are shown on the figure in the ODCM. 7 E

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TABLE 4.12-1 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)a,e Airborne Particulate I Water or Gas Fish Milk Food Products Sediment Analysis (pCi/1) (pCi/m3) (pCi/kg, wet) (pCi/1) (pCi/Kg, wet) (pCi/Kg, dry) gross beta 4 1 x 10-2 3H 2000 54Mn 15 130 59Fe 30 260 58,60Co 15 130 e C 652n 30 260 4 95 Zr 30 95-Nb 15 131 1 1c 7 x 10-2 1 60 134Cs 15 5 x 10-2 130 15 60 150 137Cs 18 6 x 10-2 150 14 80 180 140Ba 60 60 140La 15 15 7 E

TABLE 4.12-1 (Continued) TABLE NOTATION

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

For a particular measurement system (wt._en may include radiochemical separation): 4.66sb LLD = E . V . 2.22 . Y exp(- A o t) w are LLD is the lower "a priori" limit cf detection as defined above (as pCi per unit mass or volume), sb is the standard deviation of the background operating rate or of the counting rate of a blank sample as appropriate (in counts per minute). E is the counting ef ficiency (as counts per transformation). V is the sampic size (in units of mass or volume). 2.22 is the number of transformation per minute per picoeurie. Y is the fractional radiochemica] yield (when applicable). Is the radioactive decay constant for the particular radionuclide. Is the elapsed time between sample collection (or end of the sample col-lection period) and time of counting. The value of sb used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the l LLD for a radionuclide determined by gamma-ray spectrometry, the background { shall include the typical contributions of other radionuclides normally present in the samples (e.g., potassium-40 in milk samplem' . Typical values l of E, V, Y and A t shall be used in the calculations. l TMI-1 3/4 3- 137 l l

TABLE 4.12-1 (Continued) TABLE NOTATION

b. LLD for drinking water.

c. Other peaks which are measured and identifiable, together with the radioactivity in Table 4.12-1, shall be identified and reported.

e I t If11-1 3/4 3- 138 l l

RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDIT10N FOR OPERATION 3.12.2 A land use census shall be conducted during the grazing season to determine the location of the nearest milk animal in each of the 16 meteoro-logical sectors within a distance of 5 miles. Broad leaf vegetation sampling at the site boundary or closest landsite location in a sector with the highest annual average D/Q shall be conducted during the harvest season. APPLICABILITY: At all times. ACTION:

a. With a land use census identifying a location (s) which yields a calculated dose or dose committment greater than the values currently being calculated in Specification 4.11.2.3, in lieu of any other report required by Specification 6.9.2, prepare and submit to the Commission within 30 days, a Special Report which identifies the new locations.

b. With a land use census identifying a location which yields a cal-culated dose or dose committment (via the same exposure pathway) greater than at a location from which samples are currently being

, obtained in accordance with Specification 3.12.1, in lieu of any other report required by Specification 6.9.2, prepare and submit to the Commission within 30 days, a Special Report which identifies the new locations. The new location shall be added to the radiological environmental monitoring program within 30 days. The sampling location, excluding the control station location, having the lowest calculated dose or dose committments (via the same exposure pathway) may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted. SURVEILLANCE REQUIREMENTS 4.12.2.1 The land use census shall most likely be conducted at least once per 12 months between the dates of June 1 and October 1, using that information which will provide the best results such as, door-to-door survey, aerial survey, or by consulting local agriculture authorities. TMI-1 3/4 3-139

TABLE 3.12-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Reporting Levels Water Airborne Particulate Fish Milk Food Products Analysis (pCi/1) or Gases (pCi/m3) (pCi/kg, wet) (pCi/1) (pCi/Kg, wet) H-3 2 x 104 ( ) Mn-54 1 x 103 3 x 104 Fe-59 4 x 102 1 x 104 Co-58 1 x 103 3 x 10 4 Co-60 3 x 102 1 x 104 g e Zn-65 3 x 102 2 x 104 A Zr-Nb-95 4 x 102 h I-131 2 0.9 3 1 x 102 Cs-134 30 10 1 x 103 60 1 x 103 Cs-137 50 20 2 x 103 70 2 x 103 Ba-La-140 2 x 102 3x 102 (a) For drinking water samples. This is 40 CFR Part 141 value. 7 5

RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 3.12.3 Analysis shall be performed on radioactive materials supplies as part of an Interlaboratory Comparison Program which has been approved by NRC. APPLICABILITY: At all times. - ACTION: With analyses not being performed as required above, report the correction actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report. SURVEILLANCE REQUIREMENTS 4.12.3 A summary of the results obtained as part of the above required Inter-laboratory Comparison Program and in accordance with the ODCM shall be included in the Annual Radiological Environmental Operating Report. TMI-l 3/4 3 141

INSTRUMENTATION BASES 3/4.3.3.8 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases. The alarm / trip setpoints for these instru-ments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. . 3/4.3.3.9 RADI0 ACTIVE GASEOUS EFFLUENT INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases. The alarm / trip setpoints for these instru-ments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. I i i

TMI-1 83/43-142

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3/4.11 RADIDACTlVE EFFLUENTS BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION This s,9cificatien is provided to ensure that the concentration of radio-active materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II. This limitation provides additional assurance that the levels of radioactive materials in bodi~es of water outside the site will not result in exposures within )1) the Section II.A design objectives of Appendix. I,10 CFR Part 50, to an individual and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for noble gases is based upon the assumption that Xe-135 a the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in '.nternational Commission on Radiological Protection (ICRP) Publication 2. 3/4.11.1.2 DOSE This specification is provided to implement the requirements of Sections II. A, III . A and IV. A of Appendix I,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II. A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV. A of Appendix I to assure that the releases of radioactive material in liquid ef fluents will be kept "as low as is reasonably achievable". Also, for fresh water sites with drinking water supplies which can be potentially af fected by plant operations, there is reason-able assurance that the operation of the facility will not result in radionuclide concentrations' in the finished drinking water that are in excess of the require-ments of 10 CFR 20. The dose calculations in the ODCM 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 an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid ef fluents 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 Ef fluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.113. l TMI-1 B 3/4 3- 14 3 1

RADI0 ACTIVE EFFLUENTS BASES This specification applies to the release nf liquid effluents from each reactor at the site. > 3/4.11.1.3 LIQUID WfjTE TREATMENT The use of the liquid radwaste tre'atment system ensures that this system will be available for use whenever liquid effluents need treatment prior to release to the environment. The appropriate portions of this system provides assurance that the releases of radioactive materials in liquid effluen'.s will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and design objective Section II.D of Appendix A to 10 CFR Part

50. The specified limits governing the use of appropriate portions of the hquid radwaste treatment system were specified as a suitable fraction of the guide set forth in Section II. A of Appendix I,10 CFR Part 50, for liquid ef fluents. .

3/4.11.2 GASEOUS EFFLUENTS 3/4.1,1. 2.1 DOSE RATE The specification is provided to ensure that the release rate at anytime at the exclusion area boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an Individual in an unrestricted area, either within or outside the exclusion area boundary, to annaul average concentrations exceeding the limits - specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For individuals who may at times be t'ithin the exclusion area boundary, the oc upancy of the individual will be s ifficiently low to compensate for any incaase in the atmospheric diffusion factor above that for the exclusion area boundary. The specified release rate limits restrict, at all times, the cor-responding gamma and beta dose rates above background to an individual at or beyond the' exclusion area boundary to i 500 mrem / year to the total body or to

< 3000 mrem / year to the skin. These release rate limits also "estrict, at all times, the corresponding thyroid dose rate above background to an infant via the cow-milk-infant pathway to i 1500 mrem / year for the nearest cow to the plant.

TMI-1 03/43 144

RADI0 ACTIVE EFFLUENTS BASES This specification applies to the release of gaseous effluents from all reactors at the site. 3/4.11.2.2 DOSE, NOBLE GASES This specification is provided to implement the requirements of Sections II.8, III. A and IV. A of Appendix I,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same , time implement the guides set forth in Section IV. A of Appendix I assure that the releases of radioactive material in gaseous ef'luents will be kept "as low as is reasonably achievable." The Surveillaner u uirements implement the requirements in Section III.A of Appendix I th c conform with the guides of Appendix I to be shown by calculational prociouces based on models and data such that the actual exposure of an individe:1 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 radio-active noble gases in gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man # rom 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 from Light-Water-Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for determining the air doses at the exclusion area boundary will be based uponm the historical average atmospheric condi tions . NUREG-0133 provides methods for dose calculations consistent wiht Regulatory Guides 1.109 and 1.111. 3/4.11.2.5 DOSE. RADIDIODINES, RADI0 ACTIVE MATERIAL IN PARTICULATE FORM AND RADIONUCLIDES OTHER THAN NOBLE GASES This specification is provided to impleme'.i the requirements of Sections II .C, III . A and IV. A of Appendix I,10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time , implement the guides set forth in Section IV. A of Appendix I to assure that the l releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III. A of Appen-dix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an indi-vidual through appropriate pathways is unlikely to be substantially under-estimated. The ODCM calculational methods approved by NRC 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, "Calcu-lating of Annual Doses to Man fromRoutine Releases or Reactor Effluents for the Purpose of Evaluating Compliance with 10 CfR Part 50, Appendix I," Revision I, i l I TMI-1 83/43-145 1 L

RADI0 ACTIVE EFFLUENTS BASES 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 determing the actual doses based upon the historical average atmospheric conditions. The release rate specifications for radiciodines, radioactive material in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the unrestricted area. The pathways which are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy arecs where milk animals and meat producing animals graze with consumption of the railk and meat by man, and 4) deposition on the ground with subsequent exposure of man. 3/4.11.2.4 GASEOUS WASTE TREATMENT The use of gaseous radwaste treatment system ensure that the system will be available f or whenever gaseous effluents require treatment prior to release to the environment. The appropriate portions of this system provide reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and design objective Section IID of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suiteble fraction of the guide set forth in Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous effluents. l l 1 J TMI-1 83/43-146

RADIDACTIVE EFFLUENTS BASES 3/4.11.2.7 GAS STORAGE TANKS Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tanks contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Standard Review Plan 15.7.1, " Waste Gas System Failure." , 3/4.11.3 SOLID RADI0 ACTIVE WASTE The use of the solid radwaste system ensures that the system will be available for use whenever solid radwastes need processing and packaging prior to being shipped offsite. This specification implements the requirements of 10 CFR Part 50.36a. 1 11 I - 1 B 3/4 3-147

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.1 MONITORING PROGRAM The radiological monitoring program required by this specification provides trasurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of individuals resulting from the station operation. This monitoring program thereby supplements the radiological ef fluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measure-ments and modeling of the environmentaf exposure pathways. The initially specified monitoring program will be effective for at least the first three years of commericial operation. Following this period, program changes may be initiated bastd on operational experience. 3/4 12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of unre-stricted areas are identified and that modifications to the monitoring program are made if required by the results of this census. The best survey information from the door-to-door, aerial or consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were used,

1) that 20% of the garden war used for growning broad leaf vegetation (i.e. ,

similar to lettuce and cabbage), and 2) a vegetation yeild of 2 kg/ square meter. 3/4 12.3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an Inter laboratory 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 a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid. TMI-1 B 3/4 3- 148 L

5.0 DESIGN FEATURES 5.1 SITE Applicability Applies to the location and extent of the exclusion boundary, restricted area, and low population zone. Objective To define the above by location and distance description. Specification 5.1.1 The Three Mile Island Nuclear Station Unit 1 is located in an area of low population density about ten miles southeast of Harrisburg, Pa. It is in Londonderry Township of Dauphin County, Pennsylvania, about two and one-half miles north of the southern tip of Dauphin County, where Dauphin is coterminal with York and Lancaster Counties. The station is located on an island approximately three miles in length situated in the Susquehanna River upstream from York Haven Dam. Figure 2-3 of the TMI Unit 1 FSAR is an aerial photo of the site showing the plant orientation and immediate surroundings. The exlusion area as defined in 10 CFR 100.3, is a 2,000 ft radius, including portions of Three Nile Island, the river surface around it, and a portion of Shelley Island, which is owned by Met-Ed. The minimum distance of 2,000 f t occurs on the shore of the mainland in a due easterly direction from the plant as shown on Figure 2-3 of the FSAR. Figure 1-1 of the FSAR is a plot plan showing the physical location of the fence which defines the "destricted Area" surrounding the plant. The minimum distance of the

        " Restricted Area" is approximately 560 feet and is from the centerline of the TMI Unit 2 Reactor Building to a point on the westerly shoreline of Three Mile Island. Figure 5-1 is the Extended Plot Plan for Three Mile Island and includes the Exclusion Area and the meteorological tower locations. The minimum distance to the outer boundary of the low pop-ulation zone is two miles as shown on Figure 5-2. For discharge points for gaseous ef fluents, see figure 5-3 and for liquid ef fluents, see Figure 5-4.

THI-1 5-1

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l LIQUID' EFFLUENT OUTFALL DESCRIPTIONS 001 - Represents the main discharge for liquid offluents discharged from the island. 002 003 )

             ) Are redundantemergency   outfalls to 001 004 I

1 1 5-11

2) review of violations of applicable federal statute 9 codes, 9

regulations and internal station procedures and instructions having nuclear safety significance,

f. Evaluating plant operations for and providing assistance in planning future activities to the Unit Superintendent.

g. Perform special reviews and investigations and submit reports thereon as directed by the Manager-Generation Division, the Manager-Generation Operations-Nuclear or Unit Superintendent.

h. Review of the Plant Security Plan and implementing procedures as they relate to nuclear safety and shall submit recommended changes to the Unit Superintendent.

i. Review of the Emergency Plan'and implementing procedures and shall submit recommended changes to the Unit Superintendent.

j. Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence to the Superintendent and to the Met-Ed Technical Support Staff,

k. Review of major changes to radwaste systems.

AUTHORITY 6.5.1.7 The Plant Operations Review Committee shall:

a. Recommend to the Unit Superintendent in writing approval or disapproval of items considered under 6.5.1.6(a) through (d) above.

b. If requested by the Unit Superintendent for 6.5.1.6(a) through (d) and at all times for 6.5.1.6(3), render determinations with regard to whether or not each item considered constitutes an unreviewed safety question,

c. Provide immediate written notification to the Manager-Generation Operations Nuclear of any unresolvable disagreements between PORC and the Unit Superintendent as they may relate to nuclear safety; however, the Unit Superintendent shall have" responsibility for resolution of such disagreements pursuant to 6.1.1 above.

Note: The Plant Operations Review Committee shall be advisory to the Unit Superintendent. Nothing herein shall relieve the Unit Super-intendent of his responsibility for overall safety of plant opera-tions including taking immediate emergency actions. RECORDS 6.5.1.8 The Plant Operations Review Committee shall maintain at the station written minutes of each meeting and copies shall be provided to the Unit Superintendent , Manager-Generation Operations-Nuclear, Manager-Generation Engineering, and the General Of fice Review Board Secretary. 6.5.2.A MET-ED CORPORATE TECHNICAL SUPPORT STAFF ORGANIZATION 6.5.2.A.1 The organization of the Met-Ed Corporate Technical Support Staff is as shown on Figure 6-1 and consists of the Manager-Generation Operations Nuclear, Manager-Generation TMI-l 6-5

i

k. Periodically audit the areas listed below to verify compliance with the Three Mlle Island Operating Quality Assurance Plan, Fire Protection Program Plan, internal rules and procedures, federal regulations, and operating license provisions:

1) The 18 Criteria of 10CFR50, Appendix B

2) Normal Unit Operation

3) Inservice Insoection

4) Refueling .

5) Radiological Controls

6) Station Maintenance

7) Technical Specifications

8) Training and Qualifications of Station Staff

9) Emergency Plan

10) Industrial Security Program

11) Fire Protection Program and implementing procedures In performing these audits, written procedures and/or check-lists shall be used. As a minimum, each area shall be audited at least once every two years.

1. Review the radiological environmental monitoring program and the results thereof at least once per 24 months per the 00A Plan.

m. Review the OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months.

n. Review the PROCESS CONTROL PROGRAM and implementing procedures for solidification of radioactive wastes at least once per 24 months.

o. Review the performance of activities required by the Quality Assurance Program to meet the criteria of Regulatory Guide 4.15, December 1977 at least once per 24 months.

TMI-1 6-6 a

6.8 PROCEDURES 6.8.1 Written procedures and administrative policies shall be established, implemented and maintained that meet or exceed,the requirements and recommendations of Sections 5.1 and 5.3 of ANSI N18.7-1972 and Appendix "A" of USNRC Regulatory Guide 1.33 November 1972 except as l provided in 6.8.2 and 6.8.3 below. Implementation of the Fire Protection Program shall be by means of written procedures. 6.8.2 Each nuclear safety related procedure and administrative policy of 6.8.1 above, and changes theteto, shall be reviewcd by the Plant Operations Review Committee and approved by the Unit Superintendent prior to implementation and periodically as may be set forth in each document. 6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:

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

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

c. The change is documented', reviewed by the Plant Operations Review Committee and approved by the Unit Superintendent within 7 days of implementation.

6.8.4 Jritten procedures shall be established, implemented and maintained covering the activities referenced below:

a. PROCESS CONTROL PROGRAM implementation.

b. OFFSITE DOSE CALCULATION MANUAL implementation.

c. Quality Assurance Program for effluent and environmental monitoring, using the guidance in Regulatory Guide 4.15.

TMI-1 6-11

6.9 REPORTING REQUIREMENTS (cont'd) (4) Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady state conditions during power operation greater than or equal to I?e Ak/k: a calculated reactivity balance indicating a shutdown margin less conservative than specified in the technical specifications; short-term reactivity increases that correspond to a rer etor period of less than 5 seconds or, if sub-critical, an unplanned reactivity insertion of more than 0.5?f A k/k; or occurrence of any unplanned criticality. (5) Failure or malfunction of one or more components which prevents or could prevent, by itselt, the fulfillment of the functional requirements of system (s) used to cope with accidents analyzed in the FSAR. (6) Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional require-ments of systems required to cope with accidents analyzed in the FSAR. Note: For itema 6.9.2A(5) and 6.9.2A(6) reduced redundancy that does not result in a loss of system function. need not be reported under this section but may be reportable under items 6.9.2.B(2) and 6.9.2.B(3). (7) Conditions arising from natural or man-made events that, as a direct result of the event require plant shutdown, uperation of safety systems, or other protective measures required by tech-nical specifications. (8) Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the FSAR or in the bases for the Technical Specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the safety analyses. (9) Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the FSAR or Technical Specifications bases; or discovery during plant life of conditions not specifically considered in the FSAR or Technical Specifications that require remedial action or

                 ' corrective measures to prevent the existence of development of an unsafe condition. *

(10) Of fsite releases of radioactive materials in liquid and gaseous l ef fluents which exceed the limits of Specification 3.11.1.1 or I 3.11.2.1. (11) Exceeding the limits in Specification 3.11.2.7 for the storage of radioactive materials in the listed tanks. I

                            *This item is intended to provide for reporting of         j potentially generic problems.                             i TMI-1                                    6-16 l

6.9 REPORTING REQUIREMENTS (cont'd) B. Thirty Day Written Reports. 1/ The reportable occurrences discussed below shall be the subject of written reports to the Director of the appropriate Regional Office within thrity days of occurrence of the event. The written report shall include narrative material to provide complete explanation of the cause of the event, circum-stances surrounding the event, any corrective action, and component failure data. (1) Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillment of the functional requirements of af fected systems. . (2) Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown reqaired by a limiting condition for operation. Note: Routine surveillance testing, instrument calibration, or preventative maintenance which require system con-figurations as described in items 6.9.2.B(1) and 6.9.2.B(2) need not be reported except where test results themselves reveal a degraded mode as described above. (3) Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems. (4) Abnormal degradation of systems other than those specified in item 6.9.2.A(3) above designed to contain radioactive material resulting from the fission process. Note: Sealed sources or calibration sources are not included under this item. Leakage of valve packing or gaskets within the limits for identified leakage set forth in technical specifications need not be reported under this item. (S) An unplanned offsite release of 1) more than 1 curie of radio-active material in liquid ef fluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radiciodine in gaseous effluents. (6) Measured levels of radioactivity in an environmental campling medium determined to exceed the reporting level values of Table 3.12.2 when averaged over any calender quarter sampling period. TMI-1 6-17 i

        -                                         -                   ,   -m      w

ADMINISTRATIVE CONTROLS ANNUALRADIOLOGICALENVIRONMENTALOPERATINGREPORT2/ 6.9.4.1 Routine radiological environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. 6.9.4.2 The annual radiological environmental operating reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous environnental surveillance reports and an assess-ment of the observed impacts of the plant operaticn on the environment. The reports shall also include the results of the land use censuses required by Specification 3.12.2. If harmful effects or evidence of irreversible damage are detected by the monitorirg, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem. The annual radiological environmental operating reports shall include summarized and tabulated results in the format of Regulatory Guide 4.8, December 1975 of all radiological environmental samples taken during the report period. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supple-mentary report. The reports shall also include the following: a summary description of the radiological environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions from one reactor; and the results of licensee participation in the Interlsboratory Comparison Program, required by Specification 3.12.3. SEMANNUALRADI0ACTIVEEFFLUENTRELEASEREPORT2/ 6.9.5.1 Routine radioactive effluent release reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days af ter January 1 and July 1 of each year. 2/A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station: however, for units with separnte radwaste systems, the submittal shall specify the releases of radioactive material from each unit. TMI-1 6-18a l

7 ADMINTSTRATlVE CONTROLS 6.9.5.2 The radioactive effluent release reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, "'ieasuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radio-acitve Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof. The radioactive effluent release report to be submitted 60 days after January 1 of each year and shall include an annual summary of hourly meteorological data collected over the previous year. Thisr annual summary may be either in the form of an hour-by-hour listing of wind speed, wind direction, atmospheric stabilit.y, and precipitation (if measured) on magnetic tape, or in the form of joint frequency distribut i.ons of wind speed, wind direction, and atmospheric stability. This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to individuals due to their activities inside the site boundary (Figures 5.1-3 and 5.1-4) during the report period. All assumptions used in making these assessments (i.e., specific activity, exposure time and location) shall be included in these reporte. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frt',Jency and measurement) stall be used ror determining the gaseous pathway deses. The assessment of radiation doses shall be performed in accor-dance wiih the Offsite Dose Calculation Manual (ODCM). The radioactive effluent release report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed real individual from reactor releases and other nearby uranium 1el cycle sources (including doses from prirrary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev.1. The radioactive ef fluent release reports shall include the following information for each type of solid waste shipped offsite during the report period:

a. container volume, b, total curie quantity (specify whether determined by measurement or estimate),

c. principal radionuclides (specify whether determined by measurement ,

or estimate), j i

d. type of waste (e.g. spent resin, compacted dry waste, evaporator l bottoms), '

e. type of container (e.g. , LSA, Type A, Type 8, Large Quantity) and

f. solidification agent (e.g. , cement , urea formaldehyde) . ,

i 1 TMI-1 6-18b i 1

ADMINISTRATIVE CONTROLS The radioactive effluent release reports shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis. The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP) made during the reporting period. Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the next Semi Annual Radioacitve Effluent Release Report. l 1 l l l l l TMI-I 6-18c

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

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

i. Records of quality assurance activities required by the 00A i Plan.

j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59. .

k. Plant Operations Review Committee and General Office Review Based Minutes.

1. Records of analyses required by the radiological environmental monitoring program.

6.11 RADIATION PROTECTIONS PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFP Part 20 and shall be approved, maintained and adhered to for all operations invol'ving personnel radiation exposure. 6.12 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous and solid) , 6.12.1 Licensee initiated major changes to the radioactive waste systema ! (Liquid, gaseous and solid):

1. ahall be reported to the Commission in the Annual Report for the period in which the evaluation was reviewed by PORC. The discus-sion of each change shall contain:

a. A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59;

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

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

d. An evaluation of the change which shows tha predicted releases of radioact ive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those p eviously predicted in the licence application and amendments thereto; TMI 6-20

c. An evaluation of the change which shows the expected maximum exposures to individual in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto;

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

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

h. Documentation of the fact that the change was reviewed and found acceptable by the PORC.

2. Shall become ef fective upon review and acceptance by the PORC.

1 l i l TMI-1 6-20a l l

ADMINISTRATIVE CONTROLS 4 6.16 PROCESS CONTROL PROGRAM (PCP) 6.16.1 The PCP shall be approved byn the Commission prior to implementation 4 6.16.2 Licensee initiated changes to the PCP:

1. Shall be submitted to the Commission in the semi-annual Radio-l active Effluent Release Report for the period in which the

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

a. sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;

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

c. documentation of the fact that the change has been reviewed and found acceptable by the PORC.

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

i . 6.17 0FFSITE DOSE CALCULATION MANUAL (ODCM) I 6.17.1 The ODCM shall be approved by the Commission prior to implementation. 6.17.2 Licensee initiated changes to the ODCM:

1. Shall be submitted to the Commission in the semiannual Radiation Ef fluent Release Report. This submittal shall contain:

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

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

c. documentation of the fact that the change has been reviewed and found acceptable by the PORC.

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

I TMI-1 6-27 t

i ENVIRONMENTAL TECHNICAL SPECIFICATIONS

TABLE OF CONTENTS i

Page 1.0 DEFINITIONS ............................................... 1 4 j 2.0 LIMITING CONDITIONS FOR OPEPATION . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.1 Thermal .............................................. 2 i 2.2 Chemical ............. ............................... 7 2.2.1 Chlorine ...................................... 7 2.2.2 Suspended and Dissolved Solids ................ 10 2.2.3 pH ............................................ 13 3.0 DESIGN FEATURES AND OPERATING PRACTICES ................... 15 3.1 Operation of Mechanical Draft Cooling Tower .......... 15 3.2 Chemical Usage ....................................... 16 3.2.1 Water Treatment ............................... 16 3.2.2 Sulfuric Acid for Cooling Tower Circuits ...... 17 g 3.2.3 Concentration of Naturally Occurring Salts .... 17 l 3.2.4 Chlorination .................................. 17 ! 3.2.5 Sanitary Wastes ............................... 18 I 3.2.6 Solid Wastes .................................. 18 l 1 5.0 ADMINISTRATIVE CONTROLS ................................... 19 5.1 Responsibility ....................................... 19 i 5.2 Organization ......................................... 20 l 5.3 Audit and Review ..................................... 20 i 1 5.4 Action to be Taken if a Limiting Condition for Operation ! is Exceeded .......................................... 21 i 5.5 Procedures ........................................... 21 5.5.1 Written Procedures for Activities ............. 21 1 5.5.2 Plant operating Procedures .................... 22 5.5.3 Review of Procedures .......................... 22 i

TABLE OF CONTENTS (Cont'd) Page 5.6 Plant Reporting Requirements ........................... 22 5.6.1 Routine Reports ................................. 22 5.6.2 Non Routine Reports ............................. 23 5.7 Records Retention ...................................... 24 t 11

1.0 DEFIETJONS t The succeeding frequently used terms are explicitl'/ defined so that a uniform interpretation of the specifications cay be achieved. Startup - The reactor shall be con'sidered in the startup code when the shutdown margin is reduced with the intent

of going. critical. e k I, 3 I, D

4. ,

9 9

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

2.0 LIMITING CONDITIONS FOR OEERATION 2.1 Thermal Monitoring Requirement Objective Objective To define operating limits for plant thermal To monitor and record temperatures during normal discharge under normal plant operation and plant operation and cooldown;in order to' provide cooldown such that the following conditions the control room operator with' data necessary to are met. appraine, take corrective action and review the results of that action in order to maintain (1) The maximum width of the zone delincated operation within limits and at the same time by the 5*F isotherm shall not exceed 25% attempt to match unit discharge temperature to of the width of the channel into which river temperatures. ' i the effluent is discharged. (2) The maximum mixed river temperature shall not exceed 87*F. (3) The maximum change in mi'.ced river temper-ature shall not exceed 2*F/hr. Specification Specification l ~

a. During the period between April 1 and a. The discharge temperature recorder located in September 30, the following effluent the control room shall be used for monitoring

, temperature limits will apply: the plant discharge temperature. Should this temperature recorder be out of service, the (1) During normal operation discharge discharge temperature recorder located in the temperature shall be no greater than mechanical draft cooling tower pumphouse shall i 7'F above inlet temperatures or-3*F be used for monitoring the plant discharge t below inlet temperature. temperature. . 4 ? i

__ . _ _ m.. _ . .. . _ - . _ _ _ _ - . - _ . _ _ _ _ _ _ _ _ - ~.. .

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                                                                                                                                                                                   ~

1 2.0 LIMITING CONDITIONS FOR OPERATION Spo'cifica tion (Con t ' d)- Specification (Cont'd) p_y_g (2) During reactor cooldoun 'cond!tions .' discharge temperature shall not exceed 4 -

b 3 12*F above inlet temperature and this b 9 temperature dif ferential shall not- .

i

    $ KEN                   be changed by more than 20P during i                           any one-hour period.            .

6~~l) (3) If intake water temperature is 87"F M or higher discharge temperature during normal operation shall be maintained ' 1 02fb) at or below tiver ambient. i . During the period between October 1 and b. ihe delta ten >perature recorder located in I b. 3 ebEE3 March 3L the following effluent temper- thu control roum shall by used for monitoring b ature limits uill apply:

the dif ference between river water inlet ~

temperature and the dis, charge temperature. (1) During normal operation discharge Should this delta temperature recorder be out temperature chall be no greater than of service, the dif ference between the river 12*F above inlet temperature or 3*F vater inlet temperature and the discharge below inlet temperature, temperature shall be obtained from recorders

                                                                                        - located in            the mechanical draft cooling tower
                       .(2) During reactor cooldown conditions                               pumphounc.

discharge temperature shall not execed . ' 20*F above inlet temperature and this . temperature differential shall not. be changed by more than 20F during . ' any one-hour period. .

                        .                               .                                                                                                                            l 9

4

_ . . _ . - __. -_. - . . .. . - . - . .- ~ [ .D ,-

                                                                                                                              ; Q..      l 4-2.0       LIMITING CONDITIONS FOR OPERATION                                                                         ,
                ' Bases    ,

Bases _ t Natural draft cooling towers are utilized to Instrumentation is requir,ed.for twoLdifferent' . cool the large heat load of the condenser, purposes and is located in two places., One The mechanical draft cooling tower coola a group of instruments is located in the control mixture of service water and the natural room to provide operator control intelligence.. . draft cooling tower blowdown. The effluent The second group of instruments is located in from the mechanical draft cooling tower the mechanical draft cooling tower pumphouse - discharges to the river. and serves in connection with the automation of . the towers but also serves to provide additional-For normal operation, one pump and up to three time-history recorded data and backup information fans will be operated to affect maximum cooling for operator control intelligence in event without intentionally discharging below river control room instrumentation is out of service. ambient. The tower will be operated manually by the operator from the control room to affect In the control room .the following enables the maximum cooling without intentionally dis- operator to monitor and tontrol discharge charging below river ambient. The tower was . temperatures: designed to limit discharges to 870F on the 7 !: hottest dcy, a. Delta temperatura recorder -- discharge temperature minus river water inlet temperature. As an operator aid, the MDCT can be operated , , in the automatic mode which shifts fans to b. River water inlet temperature on computer. half speed, reduces the number of fans operating - and shifts fan operation from cell to cell. The c. Heated water temperature'to tower indicated, automatic mode is used to help prevent icing of a the MDCT while maintaining discharge temperature d. Discharge temperature (of. tower is recorded as close as possible to river water inlet and indicated. temperature. The automatic control system, l however, does not assure compliance with environ . In the cooling tower pumphouse, the following . mental Technical Specifications. The operator instrumentation is available on multipoint and ! will take manual control when necessary to pre- continuous pen' dragging recorders: - vent icing or to improve cooling tower operation ,

              . with regard to discharge temperature. It is        a. River water -inlet temperature.

expected that during sustained cold periods, the . discharge will average 3 c above river ambient, b. Heated water temperature to tower.- '

e

                                                                                                .,    ,         ;                                             .I

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

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                                                                                                                       ~        ~
                                                                                                                                                                                                                                    '     ~

u; -: 2.0 LI:lITING CONDITIONS FOR OPERATION -

                                                                                                                                                                                                  .i '                                         .
                                                                                                                                                                               .                       ,Ati                s i.

Bases (Cont'd) Bases (Cont'd)

llowever, since the tower performance is a c. Discharge L temperature of tower. - i function of air uct bulb temperature and since 4. . the wet bulb can increase many degrees in hours

d. Air dry bulb temperature.g-l v.

while the river temperature tracks much more

                                                                                                                                              ,   .(                                                 ;

slowly, the tower's performance can become in- e. ' Cooling tower basin water', temperature ne'ar - . effective. The worst example of this mismatch louvers. . - . L is a sudden warm day in winter with a frozen

                                                                                                                                                                                                ~ y' river. At such times the tower will be shut                                                               '                               '

L.

                                                                                                                                                                                                         ..                               x down, since continued operation would result in.                                                 ,
                                                                                                                                                                                                                                                          ~
                                                                                                                                          ,       ,9

, higher temperatures. - - - t .. tv As an example of how TMI might perform with ' .;' , these restrictions with the tower shut down *

                                                                                                                                                                                                  ,1                                                      i due to air / river temperature mismatch, the mixed river temperature, assuming a 33 F                                                                     .                 .

river and a winter river flow of 10,000 cfs

would be 0.1 F above river ambient based on '

normal plant operation. , For cooldown operation, two pumps will be

operated to pump over the mechanical draft '
* ~

cooling tower fill. The tower is designed

to cool the ef fluent flow on the hot test day to 87 F. If the tower should be in the T automatic mode of operation, the operator -

will shift to manual operation to achieve ' '

,                  maximum cooling at the beginnning of cool-                                                                                                   .

down. Note that an incrpased heat load is - present at the beginning of cooldown which , reduces the probability of freezing. i - l , , i Near the end of the cooldown, the operator ,

                  .may shift to automatic. control to preclude                  __
                                                                                                                                                       ,                                                                           .                      I

) freeze-up. ,- .,

                                                                                                                                       ~ .~ '                                                                                                              :

If cooldown should occur at a time of air / 1 f i e . ! E

A

                                                                                                                                                                                                                    .+                   s.              ].
                                                                                                                                                    ..s 3

 . _ . - - _ - _ - - .                . -      -    -.   . _ _ . - . -- _.           -  ._       -.,         . . . -         -          . . -                _ . - - -                            . -. .                  . - _ - ~ . .
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        .                 2.0             LIMITING CONDITIONS FOR OPERATION                                                                                                                                                   x.."
                                                                                                                                                                                                                                             -r
                                                                                                                                                                                                                                                            5

t.

Bases (Cont'd) , si i rifer temperature mismatch (as described under

normal operation earlier) and should the tower operation add heat, the tower will be bypassed.
6, ,

                                                                                                                                                                                                                           .! 6 t

If the unit were to be 'copled down with the .,. , mechanical draft cooling tower not operating.

the mixed river temperature at beginning of *
cooldown would bc <+ 3%' above river ambient based on a 33 F river with 10,000 cfs flow. ,

The above operating practices and the . I i effluent temperature limits in this speci- ~ j fication will insure compliance with the

objectives. ,

o '} e t t 0

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                                                                    - t-2.0                1,lMITING CONDITICUS FOR OPERATION 2.2                Chemien1

?.2.1 Chlorine Monitoring Requirement O!>J ee t i ve Objective The purpose of this specification is to , The purpose of this specification is to ensure l i:ni t t.he discharge of chlorine to levels that the chlorine concentration at the plant which are not harmful to the biota in the river water discharge is monitored in such a way Cusquehannu River. as t.o assure compliance with Specification 2.2.1. Specification geeification ,

a. The total Chlorine concentration, as The total chlorine concentration shall be monitored ceasured at the pInnt river discharge, and recorded continunily at the plant river water shall. not exceed 0.2 ppm and the free discharge and in the cooling tower blevdown.

chlorine component shall be less than lustrument.s employing the amperometric principle 0.1 ppa except as discussed in b. below. or instruments employing another method of equiva-

                                                                             . lent, accuracy and standardized against the umpero-

h. For one consecutive 90 day period during metric method shall be used. If the automated the first tuo years of plant operation, the

monitoring equipment at the plant river water tot.al residual chlorine concentration discharge is out of service, an analysis for measured at t.he plant river discharge chlorine concentrat. ion vill be made daily during

                ,       shall not exceed 0.5 ppm. The starting               a chlorinat.lon period. The analysis shall be date for this 90 day period will be                   performed on samples taken at 10, 30, and 50 relected by the Metropolitan Edison                   minutes following the start of the chlorinat. ion Ctepany and t.he Director of the Regional             period.

0.erations 1 Office notified before it.s c urune nc emen t. . For a temporary period not exceeding two years, a reading of 3 ppm or Jess total residual chlorine in the cooling tower blowdown shall be t.aken as

f *D

  .0      LIMITING CONDITIONS FOR OPEltATION Specificatiori (cont'd)              .

Specification (cont'd)

e. The tot.al-duration of chlorine discharge evidence that the concentration'of chlorine

_to the river at levels greater than originating in the blowdown is 0.01 ppm or less , 0.01 ppm shall not e<ceed 2 hours per in the discharge to the river. During the 90-day, day period identified in part b of Specification 2.2.1, the plant operating staff.or their agents shall determine the adequacy of the 1 ppm limit to assure the upper limit of 0.01 ppm in the l discharge to the river. If the.1 ppm limit is found inadequatt, ur if the plant operators so elect, the operators shall submit a new limit on residual chlorine in the cooling tower blowdown

to the AEC Office of lleculation for approval.

If the 1 ppm limit is found inadequate, a nev

                                                                        . limit must be approved by the AEC Office of

, llegulation within two years of the start of operation of the plant. If the 1 ppm limit is found to be adequate, the plant.. operators shall submit evidence of t.his adequacy to the AEC 3 Office of llegulat. ion for concurn,ence within two years of the start of operation of the plant, i Eases Bases , Ituned on the AEC staff review, including as Monitoring residual chlorine at the discharge backd round an evaluation by William A. Brungs,* to the river will assure compliance with the

   .'illiam A Brungs, " Effects of Residual Chlorine on Aquatic Life," J. Water Pollution Control Federation, M, 1

2180-93-(1973).

                 .- .                                            n,                                                       ^

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                      ' A             eht i eeea eo        fl ser R             c      ofh vuu                      ib su E             non           t il q                h           qe                aoaace u P             ow        f        rf                wl gi nemt                                  xv O             ct eos                f e               l nnit rotd oei r      i ees                  giih raeiiit                                     r R             eras              h        o         nwt coutb            ms                      n

_ O no et t th i ael q i e nie F if yiI nt rmrt hegcl rw h r oa upe cd ni oet _ S omse l c tlf d ppg aot endb N l pup . po one l ah ow o O h posrt d5 ihh ut ot t I c u eaes o t t t tf q tl o . _ T neveh s i0 na i oaf ib nd I ) d 2.ihi rt e r art w ee om irl g e D d e0t tRh c eol ea _ N ' n n t nx pt pph ech yl el r O t i oooa i e ot t nt c wia C n b t ct no ype aa auowh o m opmga nnengi auhherd ouet t sr uit qn es c G C d N ( cup nh er - s i uso e gni I pi esat 0 eel s nved anid _ T s l r guoh o 9 glb neaga ir I e ae1 aqpc n rb aem a el on M s t v0 ms s eaat md menol e _ I L B a oi t r0dS t di a uois hh naoonahi ohh T ccct cad t mcc w _. 0 _ 2 ~ . 4 i ' ,;ii I 1

m . __ . _ . . _ _ . _ . _. 2.0 LItilTING. CONDITIONS FOR OPERATION

     . 2.2.2 Suspended and Dissolved Solids                                   Monitoring Requirement Obicetive                                                        Objective                  ,,

The purpose of this specification is to limit The purpose of this specification is to ensure the addition of suspended and dissolved solids compliance with Specification 2.2.2. to the Susquehanna River. , ,, , Specification Specification

a. The dissolved solids concentration as An analysis for suspended and' dissolved solids measured at the plant river water shall be performed on samples taken at the plant discharge, shall nor exceed 500 ppm river water discharge during the discharge of as a monthly average and shall not cach tank of neutralized regenerant wastes or execed 700 ppm at any time. Suspended at'wcekly intervals, as a minimum frequency.

solids shall not exceed 560 ppm at any . time. Inventory and log records shall be kept in such a way as to show the quantity of sulfuric acio

b. The quantity of sulfate ion released to used in domineralizer regen'eration and the the river from regeneration of the makeup quantity used in treatment of the circulating 4

water demineralizers and from the addition vator and/or its makeup. of sulfuric acid to the circulating water (and/or its makeup) shall not exceed , , 4,620,000 pounds / year. Bases Bases . , The dissolved and suspended solids values Any significant changes in dis's'olved solids are based on che Water Quality Management should be observed during tlje discharge of 1 f

        ,*                                                         f

_. - . . _ ._. _ _ . _ _ . . . _ _ _ ._m - _ _ _ _ . _ _ _ _ . . _ _ . _ _. _ _ _ . _ _ _ _ -. _. __

              ,.,,.,                                                    fr?                                                                        -

2.0 LIMITING CONDITIONS FOR OPERATION Bases (Cont'd) Bases (Cont'd) Permit No. 2270204,' approved by the Depart- regenerant wastes or during cooling tower blow-ment of Environmental Resources,.Commonucalth, down. The specified monitoring frequencies of Pennsylvania, on August 17, 1971. The only provide itssurance that neither source'is outside significant additions of dissolved solids and/ the specitication. The addiEion of a suspended or sulfates to the plant discharge are neutral . solids analysis provides monitoring of the dia-ized regenerant wastes from the cycle makeup tomaccous carth pressure filters. Inventory and demineralizers and blowdown from the cooling log records will make it possible to calculate towers. The spent regenerants consist.of the quantity of sulfate use,d in the form of salta removed from the river water prior to sulfuric acid in domineralizer regeneration and its use as makeup to the reactor or turbine in treatment of the circulating water system. plant systems plus spent sodium hydroxide This allows calculation of the meximum quantitics - and sulfuric acid regenerants. The cooling of sulfate discharged from these activitics. touer blowdown contains only salts from'the * , river which have been concentrated by evaporatio'n, with their composition cha'nged due to the addition of sulfuric acid and the l resultant loss of bicarbonate-carbonate (in the form of carbon dioxide discharged to the atuosphere). No other significant soerces of 3 increased dissolved solids exist in the plant. The limit on the quantity of sulfate released is baued on the expected normal quantities ,' of sodium sulfate released from the makeup water system and sulfuric acid added to the circulating water system, as given in Sections 3.2.1 and 3.2.2 of these Specifications. In ] view of the minimal expected impact on.the 1

                                                                     ^ f . s*

f 2.0 1.IMITING CONDITIONS FOR OPERATION i Bases (Cont'd) I river of the discharges allowed in these specifications, the limit on the. quantity discharged is set at twice the quantity , e expected to be discharged. Because there ' is no background of experience for this , particular type of plant, and therefore ' there is substantial uncertainty in the requirement for plant process water makeup, - an additional 50% allouance was provided to the limit on sulfate discharged from the makeup water demineralizer regeneration or.ly. This additional allowance is about-111,000 lb/yr, or about 2% of the total permitted discharge. All significant sources of suspended solids within the plant are filtered through dia-tomaccous carth pressure filters before discharge. Uhile suspended solids will be concentrated due to evaporation in tha cooling towers, settling will occur in the o cooling tower basins. The cooling tower , blowdown is thus not expected to raake a significant increase in the suspended . solids concentration at the plant discharge. P 1 m o n.

  .                                                 e                            i:

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

p% i' q% f. i t

2.0 LIMITING CONDITIONS FOR OPERATION 1

Monitoring Requirement _ 2.2.3 pH Objective Objective The purpose of this specification is to limit The purpose of this specification is to ensure the pH of plant discharges to values which compliance with Specification 2.2.3 vill produce no harmful effects to the Susquehanna River. .

Spec i fication Soccification The pli, as measured at the plant discharge and A determination of the pH'of the contents of each at the vaste neutralizing tank prior to release, tank of neutralized regenerant vastes vill be shall have a value of nct less than 6.0, nor ande prior to release using installed instru-more than 9 0, except that during those periods ' mentation. All necessary adjustments to meet the when the intake pli is creater than 9 0, the plant specification vill be made prior to initiation of discharge pH shall not exceed the intche pH, and the release. If the installed instrumentation is

that during those periods when the intake pH is out of service, the necessary analyses vill be less than 6.0, the plar.t discharge pH shall not perrormed prior to initiating the discharge using be less than the intake pH. laboratory instrumentation. An analysis for p!! vill be performed on a sample taken from the plant river water discharge during the release of each tank of regenerant vastes, or at weekly intervals as a minimum frequency. Bases Bases i The pH of the Susquehanna River as measured in Discharge of neutralized regenerant vastes is the

               -     the vicinity of Three Mile Island is variable               only normal plant operation which could cause a and values spanning almost the entire range                 change in the pH of th.e discharge since all sumps 1                                                                                                           .

. _ . . . __.      .    . _ . . ____._.-.____.._._....__-..m.            . . _ _ _ _ _ . _ . _ . _ . _ _ . _ . _ . . . _ _ _ ._       . _ . . -     _ _ _ _ _ _. . m_. .. _      __ . ..
                  ?                                                                   g pn.                 ,

l

                                                                           - 14   1   -                                                                                                    ,

2.0 LIMITING CONDITIONS FOR OPERATION , T , - 6 4 ' i Bases (Cont'd) Bases (Cont'd) G {

         ,           from 5.0 to 10.0 have been recorded.             I.imiting and drains which are potential receivers of N

the pli of discharge to,the normal range of chemicals are collected in this tank. . values insures that no pli related damage to river ecosystems or bicta vill result. M

                     'Ihe limits on the pli of the waste neutralizing                                                                                                    ,
                    , tank discharge vill preclude sizable changes                                                              ,
                                                                                                                                                                              , gg in the pl! of the discharge to the river. For example, adding 300 gym of pli 9 0 neutralizing g

tank discharge to a pli 8.0 stream at 17,250 . cpn would raise its pli a calculated 0.06 unit , assumind no buffering action. e 0 l 4 V

        .                                                       3.0      DESIGN FEATURES AND OPERATING PRACTICES Objective This section contains a description of design features and operating practices which, if changed, might have a significant' environmental impact.

Specification - - If operating practices or design features are planned .which deviate from those described in the bases below, an analysis of their potential environmental impact will be made and a course of action taken to alleviate potential adverse impacts. In addition, if the ecology of the river significantly changes at a future date as, for exa=ple, by major changes in water che=istry or reintroduction of shad, an analysis

               - of expected impacts and a course of' action to mini =1:e the impacts                                ,

will be provided. Bases 3.1 Oce' ration of Mechanical Draft Cooling Tower Natural ~ draft cooling towers are utilized to cool the large heat load of dhe condenser. The techanical draf t cooling tower cools a mixture (~ of service cooling water and a small a=ount of natural draft cooling : I h- tower blowdown, which represents a much reduced heat load. The ef fluent from the mechanical draf t cooling tower discharges to the river. For nor=al operation, one pu=p will be operated with up to three f ans to affect maximum cooling without intentionally discharging below river a=bient. The tower will be operated manually by the operator f rom the control roo= to affect maximum cooling without intentionally discharging below river ambient. As an operator aid, the MDCT can be operated in the automatic mode which shif ts fans to half speed, reduces the number of fans operating and shifts fans operation from cell to cell. The automatic code is used to help prevent icing of the MDCT while -maintaining discharge temperature as close as possible to river water inlet temperature. The automatic control system, however, does not assure compliance with environmental Technical Specifications. The operator will take manual control when necessary to prevent icing or to improve. cooling tower operation with regard to discharge te=perature. During sustained cold periods, the discharge will average 3 F above river ambient. However, since the tower performance is a function of air wet bulb temperature which can increase cuch more rapidly than the river te=perature, the tower's per-formance can become ineffective. An example would be a sudden warm day while the river is still f rozen. At such time the tower is shut down since its operation would result in increasing the discharge ( ~ te mpera tu re . During such periods, the discharge te=perature is (- approximately 10 F above river ambient. For cooldown operation two pu=ps are operated to pump over the mechanical draft cooling tower fill. If the tower is in the automatic code of - operation, it is shif ted to manual operation to achieve maximum

l t r

! cooling at the beginning of cooldown. With average winter weather l p' conditions, the tower discharge is appr'oximately 12 F above river (<~ ambient at the beginning of cooldown and reduces to approximately 3 F some 12 hours later. Near the end of cooldown the tower may be l . shifted back to automatic control to pr'eclude freeze-up. l 3.2 Chemical Usace This section describes the chemicals used in the plant which are discharged.t.o the environment. The equipment in which the chemicals . are used along the quantities per batch or rate of continuous discharge ~ l

l. ~and expected discharge freq6ency are included. -

3.2.1 Water Treatment The clarifier continually receives approxi=ately 0.05 lb. of cationic

                            - polyelectrolyte and 0.6 lb. of anionic clay per 1000 gallons of water treated to recove suspended solids from the river water.

l Assuming an avert,e flow of 100 gpo. through the clarifier, sludge containing approxi=ately 60 lb. of clay and 5 lb. of polyelectrolyte plus a highly variable amount of suspended solids recoved f rom the river water is blown down fro = the clarifier each day. The sludge is processed n diatomaceous earth pressure filters and the filtrate is released to the plant river water discharge. The solids ccmponent is pressed into dewatered blocks. Their disposal is described under solid waters. ( A cation - anion string in the cycle =akeup de=ineralizer system uses

    \-                         2260 lb. of sulfuric acid and 1340 lb. of sodiu= hydrode for each regeneration. An additional 2350 lb. of sodica hydroxide is required to neutralize the spent regenerants prior to discharge, resulting in 3270 lb. of neutralized sodium sulfate contained in approximately 70,000 gallons of water. Based upon a decineralized water use of       40,000 gallons per day and a production of 300,000 gallons between regenerations , this quantity would be released each 7.5 days. Release rates are based upon flow through the mechanical draf t cooling tower.

l

J i A mixed bed unit in the cycle cakeup demineralizer uses 320 lb. of a sulfuric acid and 800 lb. of sodium hydroxide for each regeneration. i An additional 72 lb. of sulfuric acid is required to neutralize the spent regenerants prior to discharge, resulting in 568 lb. of neutralized sodium sulfate contained in approximately 50,000 gallons - - of water. Based upon a demineralized water use of 40,000 gallons 4 per. day and a production of.2,000,000 gallans between regeneratiods,

                                                          ~                             -

this, quantity. iould be:re' leased each 50 days. Release rates are based l upon flow thro 6gh~the m'echanical draft cooling tower. , 1- . , The six Powdex condensate polishing units are of the wound element < filter type precoated with pewdered resin. The spent resin is back- ] washed off the ele =ents and treated in the same canner as clarifier.

sludge. With five of the six units in service during normal i operation, and a 25 day service cycle for each u' nit, one unit would i be backwashed each 5 days. This would result in ISO lb. of powdered

. resin to be pressed into dewatered blocks and the discharge of - j 14,000 - 17,000 gallons of filtered, demineralized water to the cechanical draf t cooling- tower. l 3.2.2 Sulfuric Acid for Cooling Tower Circuits l Sulfuric acid is added to t'he circulating water in the condenser j cooling water circuits, for pH control, at an average rate of j f- 6,000 lb. per day. This acid forms sulfates with various cations i .1, in the cooling water and is eventually released with the 2,000 gpa. j N blowdown from the cooling towers. This will result in an increcental increase of approxi=ately 2S ppa. sulfates in the effluent returning ] to the river, assuming 18,000 gpa. flow through the mechanical draf t l cooling towers. 3.2.3 Concentration of Naturally Occurring Salts i

!                              In addition to the acid added to the cooling tower circuits there is i                              a concentration of naturally occur' ring salts in the river water by

! abour r factor of 5 due to evaporation in the cooling towers. Assuming i an ave; age concentration of 238 pga. for dissolved salts in the river j- water, the concentration in the blowdown from the cooling towers would be approximately 1200 ppm. Assuming a cooling water flow of 18,000 gpm. through the techanical draft cooling tower this would result in , an incrc= ental increase of 120 ppa. dissolved salts in the effluent l to the river. 3.2.4 Chlorination . 1 4 ~ The water taken from the river is treated with approximately 100 lb.

pcr day of chlorine to control the growth of biological slines in

+ 1 ( ~ ' '.

  .f-t 0
             , - - - - - - . - - . - . ~ , . - - - -   , . , _ , - ,   ----.a_.,.     -e.n. w.n ..w ,  ,e,c,,,, n . . , e-.   .n . , . . - ,,-   ..w .,n   - m.   , -r

s. . .

i i 1 river water: piping and heat exchangers. The chlorine will be added (~' I in 15 cin. periods as dictated by biological growth. Up to an

additional- 1000 lb. of chlorine will be injected into the condenser cooling circuits. Addition will be in 15 ninute periods as dictated 1

by biological growth. Chlorination periods for the condenser cooling f circuits:and river water circuits will be staggered so that blowdown i from the condenser cooling circuit will make a minimus contribu. tion ,_ to' chlorine concentration at the riv'er water discharge.. This' ' ... contribution to the river water discharge chlorine concentration will , i be even lower due'to the fact that'any volatile chlorine components vill be lost during passage through the natural draft cooling towers. I 1 3.2.5 Sanitarv Wastes i

The sanitary waste treatment plant produces up to 10,000 gallons of

^ treated sanitary wastes. Biological oxygen de=and is reduced by i approxi=ataly. 93% in the aeration tanks and aerobic digester and is further reduced by the addition of sodium hypochlorite. 1

Phosphate reduction is accomplished by the addition of li=e which

{ recoves approxi=ately 80% of the input phosphate. The final effluent , of the treat =ent plant is mixed with effluent from the cachanical draf t cooling towed prior to discharge to the river. 3 2- 3.2.6 Solid Wastes , ,e Suspended solids from the water treatment facilities are separated ( frc= the carrier water by filtration through diatomaceous earth pressure-filters. The resulting slurry is further dewatered by pressing until the cake has a toisture content of approxi=ately 50% and the blocks are trucked offsite to an approved sanitary i landfill. .

!~

Approximately once each can'th an additional 1500-1700 lb. of cake

!                  will be produced as a result of processing sludge from the sanitary waste systen. This waste is also trucked offsite for disposal in y

an approved landfill. I Trash collected from the river by the plant intake screens as well as colid wastes from the oil fired incinerator are also hauled offsite for disposal in a landfill. ) i i e < d

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5.0 ADMINISTRATIVE CONTROLS Objective To describe the administrative and management controls established to provide continuing protection to the environment and to implement the environmental technical specifications. Specifications 5.1 Responsibility Corporate responsibility for implementation of the Environmental Technical Specifications and for assuring that plant operations are controlled in such a manner as to provide continuing protection to j the environment have been assigned by the President or Metropolitan Edison Company to the Chief Executive Officer. This responsibility is carried out by the Generation Division through the organization set forth in Figure 6-1. Responsibility for compliance with these Environmental Technical Specifications rests with the Manager Unit 1. The procedures I and controls necessary to ensure compliance are implemented through the statf of the Director TMI-1. The Manager Unit 1 is responsible g for the environmental compatibility of plant operations, and he shall ensure that: A. All proposed changes to the procedures delineated in Section 5.5 of these Environmental Technical Specifications and design changes to such equipment or syitems as is the subject of these procedures are reviewed by the station staff to determine whether or not they might involve a significant environmental impact. B. All proposed changes considered under 5.1.A above which were determined thereunder to possibly involve a significant environmental impact are analyzed to determine the extent of the impact. 1 C. All pronosed changes to the procedures delineated in Section ) 5.5 of these Environmental Technical Specifications and design . changes to such equipment or systems as is the subject of these l procedures that would have a significant adverse effect on the l environment or which involve a significant environmental matter i or question not previously revicued and evaluated by the NRC are l reported to the NRC prior to implementation. Proposed changes I which the analysis shows would have a favorable environmental impact or which involve a significant environmental matter or question previously reviewed and approved by the NRC are forwarded to the Director Technical Functions for independent l review. l l

D. Reports are submitted and records are kept in accordance with 5.6 and 5.7 of the Environmental Technical Specifications. Violations of these Environmental Technical Specifications are investigated and appropriate corrective action taken to prevent recurrence. Responsibility for the independent review functions concerning environmental matters as defined in section 5.3 of these Environmental Technical Specifications has been assigned by the Chief Executive Officer to the Director Technical Functions. When the review function is performed by the Radiological and Environmental Controls Section, the Chief Operating Executive shall ensure that necessary audits of those review fufetions are performed independently under the j direction of the Dirdctor Nuclear Assurance. When organizations other than Metropolitan Edison Company are utilized to establish and execute portions of these Environmental Technical Specifications, compliance with the Environmental Technical Specifications in such instances shall remain the responsibility of Metropolitan Edison Company. 5.2 Organization Organization of the personnel responsible for implementation, audit and review of these Environmental Technical Specifications including the Corporate level is as shown on Figure 6-lof these Technical l , Specifications. In all matters pertaining to compliance with these Environmental Technical Specifications, the Manager Unit 1 shall report g , to and be directly responsible to the Director. Unit 1. l l 1 5.3 Audit and Review Independent review functions for environmental matters will be performed under the direction and control of the Director Technical Functions. l Independent review of environmental matters relating to these Environmental Technica1' Specifications will be conducted by the Radiological and Environmental Controls Section, reporting to the Director Technical Functions. Their review will be audited by the Director Nuclear Assurance. These audits and reviews will encompass: I 9

A. Coordination of Environmental Technical Specifications development with the Safety Technical Specifications to avoid conflicts and maintain consistency. B. Proposed changes to the Environmental Technical Specifications i and the evaluated impact of the change. ~ C. Evaluation of proposed changes conducted in compliance with 5.lB and 5.lC. . . D. Results of the Environmental Monitoring Programs prior to their submittal in-each Annual Environmental Monitoring Report. i E. Reports of investig'ations of reported instances of violation of Environmental Technical Specifications and associated corrective action. 5.4 Action to be Taken if a Limiting Condition for Operation is Exceeded I Follow any remedial action permitted by the Technical Specification until the limiting condition can be met. i All instances of exceeding a Limiting Condition for Operation will be promptly investigated. A report of each occurrence of a violation of the provisions in specifications of the Limiting Conditions for Operation of these Environmental Technical Specifications will be prepared as specified in Section 5.6.2. 5.5 Procedures 5.5.1 The following written procedures will be prepared to ensure compliance with various activities involved in carrying out the Environmental Technical Specifications. Procedures will include applicable check lists and instructions, sampling, instrument calibration, analysis, and actions to be taken when limits are approached or exceeded. Testing frequency of any alarms will be included. These frequencies will be determined from experience with similar instruments in similar environments and from manufacturers' technical manuals. A. Operation of mechanical draft cooling touers. B. Operation of chlorination system. C. Discharge of neutralized regenerant wastes. D. Analysis for chlorine concentration, dissolved solids concentration, suspended solids concentration and pil at , the plant discharge. E. Calibration of chlorine moniter.

5.5.2 The following plant operating procedures shall include provisions to ensure the related systems and components are operated in compliance with the Limiting Conditions for Operation established as part of the Environmental Technical Specifications. A. Circulating water and natural draft cooling tower operating procedure. B. Mechanical draft cooling tower operating procedure.

C. Circulating water chlorination system operating procedure. D. River water chlorination system operating procedure. E. Discharge of neutralized regenerant westes operating procedure. F. Industrial waste treatment plant operating procedure. G. Sump pump and drainage system operating procedure. 5.5.3 All procedures described aLove and all changes thereto will be reviewed periodically under the cognizance of the Director Technical Function. however, temporary changes to these procedures which do not change the intent of the original procedure may be made providing such changes are approved by two members of the Plant Management Staff. i Such procedure change approval will be documented. 5.6 Plant Reporting Requirements - 5.6.1 Routine Reports A. Annual Environmental Operating Report 1 Nonradiological Report. A report on the environmental surveillance programs for the previous 12 months of operation shall be submitted to the Director of the NRC Regional Office (with a copy to the Director, Of fice of Nuclear Reactor Regulation) as a separate document within 90 days after. January 1 of each year. The report shall include summaries, interpretations, and statistical evaluation of the results of the nonradiological environmental surveillance activities as deemed appropriate j by the licensee and the environmental monitoring programs ; required by limiting conditions for operation for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment. If harmful effects or evidence of irreversible damage are detected by the monitoring, the licensee shall provide an analysis of the problem and a proposed course of action to alleviate the problem.

5.6.2 Non Routine Reports Nonradiological In the event a Limiting Condition for Operation is exceeded, a report will be made within 24 hours by telephone and telegraph to the Office of Inspection and Enforcement-Region 1 followed by a written report within two weeks (cc to the Director of Nuclear Reactor Regulation) . The written report and, to the extent possible, the preliminary telephone and telegraph report, will:

1. Describe, analyze and evaluate the occurrence including extent and magnitude of the impact;

2. Describe the cause of the eccurrence; and

3. Indicate the corrective action taken (including any significant changes made in procedures) to preclude repetition of the occurrence and to prevent similar occurrences involving similar components or systems.

5.6.2.4 Changes A. When a change to the plant design, to the plant operation or to the procedures described in Section 5.5 is planned which would have a significant adverse effect on the environment or which involves an environmental matter or question not previously reviewed and evaluated by the NRC, a report on the change will be made to the NRC prior to implementation. The report will include a description and evaluation of the change including a supporting benefit-cost analysis. B. Changes or additions to permits and certificates required by Federal, State, local and regional authorities for the protection of the environment will be reported. When the required changes are submitted to the concerned agency for approval, they will also be submitted to USNRC for information. The submittal will include an evaluation of the environmental impact of the change. C. Requests for changes in Environnental Technical Specifications will be submitted to the USSRC for prior review and authorization. The request will include an evaluation of the i impact of the change, including a supporting benefit-cost i i analysis. ; l l 5.6.2.5 other If harmful ef fects or evidence of irreversible damage are detected by the monitoring programs, the licensee shall provide an analysis l of the problem and shall develop a course of action to be taken to i i l _ ._ _ _ . , .

alleviate the problems. If the ecology of the river significantly changes at a future date as, for example, by major changes in water chemistry or reintroduction of shad, the licensee shall provide an analysis of expected impacts and a course of action to minimize the impacts. 5.7 Records Retention 5.7.1 Records and logs. relative to the following areas will be retained for the life of the plant.

1. Records and drawing changes reflecting plant design changes made to systems and equipment as described in Section 5.6.2.4.

2. Records of environmental surveillance data.

3. Records to demcastrate compliance with the Limiting Conditions for Operation in Section 2.

5.7.2 All other records and logs relating to the Environmental Technical Specifications shall be retained for 5 years. t l

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ML19347C386

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