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Tech Spec Change Request 34A for Tech Spec Sections 1.1 - 3.12.3 Re Definitions,Safety Limits & Limiting Safety Sys Settings & Limiting Conditions for Operation
	ML20008E290
Person / Time
Site:	Three Mile Island
Issue date:	10/03/1980
From:	; METROPOLITAN EDISON CO.
To:	;
Shared Package
ML20008E278	List: ML20008E277; ML20008E282; ML20008E290; ML20008E292;
References
	NUDOCS 8010240508
	Download: ML20008E290 (78)
	v • d • e

{{#Wiki_filter:,* , y TABLE OF CONTENTS Section Page l TECHNICAL SPECIFICATIONS 1 DEFINITIONS- 1-1 1.1 RATED POWER l-1 1.2 REACTOR OPERATING CONDITIONS 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 SHUTDOWN 1-1 1.2.7 REFUELING OPERATION 1-2 I 1.2.8 REFUELING INTERVAL l-2 1.2.9 STARTUP l-2 1.2.10 TAVG 1-2 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 INSTRUMENTATION LOGIC 1-2 1.4.1 INSTRUMENT CHANNEL l-2 1.4.2 REACTOR PROTECTION SYSTEM l-2 1.4.3 PROTECTION CHANNEL 1-3 1.4.4 REACTOR PROTECTION SYSTEM LOGIC 1-3 1.4.5 ENGINEERED SAFETY FEATURES SYSTEM l-3 1.4.6 DEGREE OF REDUNDANCY l-3 1.5 INSTRUMENTATION SURVEILLANCE l-3 1.5.1 TRIP TEST 1-3

        -,      l.5.2         CHANNEL TEST                          l-3 1-3 l.5.3         INSTRUMENT CHANNEL CHECK 1.5.4          INSTRUMENT CHANNEL CALIBRATION       1-4 1.5.5         HEAT BALANCE CHECK                    1-4 1.5.6         HEAT BALANCE CALIBRATION              14 1.6        VOWER DISTRIBUTION                       1-5 1.6.1         QUADRANT POWER TILT                   l-5 1.6.2          REACTOR POWER IMBALANCE              l-5 1.7        CONTAINMENT INTEGRITY                    l-5 1.8        FIRE SUPRESSION_ WATER SYSTEM            l-5 1.9        CHANNEL CALIBRATION                      1-6 1.10      CHANNEL CHECK                            1-6 1.11       CHANNEL FUNCTIONAL TEST                  l-6 1.19       DOSE EOUIVALENT I-131                   1-6 1.27.      SOURCE CHECK                            1-6 1.28       SOLIDIFICATION                          1-6 1.29       0FFSITE DOSE CALCULATION MANUAL         l-7 1.30       PROCESS CONTROL PROGRAM                 l-7 1.31       GASEOUS RADWASTE TREATMENT SYSTQ1       1-7 1.32       VENTILATION EXHAUST TREATMENT SYSTEM    l-7 1.33       PURCE-PURCING                           l-7
               'l.34        VENTING                                 l-7 i

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 . o TABLE OF CONTENTS Section    ,                                                   Page 2       SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS      2-1 2.1       SAFETY LIMITS. REACTOR CORE                          2-1 2.2       SAFETY LIMITS, REACTOR SYSTEM PRESSURE               2-4 2.3       LIMITING SAFETY SYSTEM SETTINGS PROTECTION           2-5 INSTRUMENTATION 3       LIMITING CONDITIONS FOR OPERATION                      3-1
                                                    ~

3.1 REACTOR COOLANT SYSTEM 3-1 3.1.1 OPERATIONAL COMPONENTS 3-1 3.1.2 PRESSURIZATION, HEATUP, AND COOLDOWN LIMITATIONS 3-3 3.1.3 MINIMUM CONDITIONS FOR CRITICALITY 3-6 3.1.4 REACTOR COOLANT SYSTEM ACTIVITY 3-8 3.1.5 CHD1ISTRY 3-10 3.1.6 LEAKAGE 3-12 3.1.7 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY 3-16 3.1.8 SINGL'i 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 SYSTD!S 3.3 EMERGENCY CORE COOLING. REACTOR BUILDING EMERGENCY 3-21 COOLING AND REACTOR BUILDING SPRAY SYSTEMS 3.4 TURBINE CYCLE 3-25 3.5 INSTRUMENTATION SYSTDIS 3-27 3.5.1 OPERATIONAL SAFETY INSTRUMENTATION 3-27 3.5.2 CONTROL R0D GROUP AND POWER DISTRIBUTION LIMITS 3-33 3.5.3 ENGINEERED SAFEGUARDS PROTECTION SYSTD1 ACTUATION 3-37 SETPOINTS 3.5.4 INCORE INSTRUMENTATION 3-38 3.6 REACTOR BUILDING 3-41 3.7 UNIT ELECTRICAL POWER SYSTEM 3-42 3.8 FUEL LOADING AND REFUELING 3-44 3.9 RADIOACTIVE MATERIALS 3-46 3.10 MISCELLANEOUS RADIOACTIVE MATERIALS SOURCES 3-46 3.11 @NDLING OF IRRADIATED FUEL 3-55 3.12 FE. ACTOR BUILDING POLAR CRANE 3-57 3.13 SECONDARY SYSTEM 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 HOT STANDBY 3-60 3.15 AIR TREATMENT SYSTDIS 3-61 3.15.1 EMERGENCY CONTROL ROOM AIR TREATMENT SYSTDI 3-61 3.15.2 REACTOR BUILDING PURCE AIR TREATMENT SYSTEM 3-62a 3.15.3 AUXILIARY AND FUEL HANDLING EXHAUST AIR TREATMENT SYSTEM 3-62c 3.16 SHOCK SUPPRESSORS (SNUBBERS) 3-63 3.17 REACTOR BUILDING AIR TEMPERATURE 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 EFFLl?DTS (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, RADIOACTIVE 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 C AS 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 CENSU3 (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 E 10 GRAM B3/4 3-148 3/4.12.2 LAND USE CEN'US B3/4 3-148 G iii

l . 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 SYSTEM 4-37

4.5 DfERGENCY LOADING SEQUENCE AND POWER TRANSFER, 4-39 EMERGENCY CORE COOLING SYSTEM AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 DIERGENCY LOADING SEQUENCE 4-39

4. 5. 2. EMERGENCY CORE COOLING SYSTEM 4-41 4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTEM 4-43 4.5.4 DECAY HEAT REMOVAL SYSTEM LEAKAGE 4-45 4.6 DfERGENCY POWER SYSTEM PERIODIC TESTS 4-46 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4-48 4.7.1 CONTROL ROD DRIVE SYSTDi 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 PUMPS PERIODIC TESTING 4-52 4.9.1 TEST 4-52 4.9.2 ACCEPTANCE CRITERIA 4-52 4.10 REACTIVITY ANOMALIES 4-53 4.11 SITE ENVIRONMENTAL RADIOACTIVITY SURVEY 4-54 4.12 AIR TREATMENT SYSTEMS 4-55 4.12.1 DIERGENCY CONTROL ROOM AIR TREATMENT SYSTEM 4-55 4.12.2 REACTOR BUILDING PURGE AIR TREATMENT SYSTEM 4-55b 4.12.3 AUXILIARY AND FUEL HANDLING EXHAUST AIR TREATMENT SYSTDI 4-55d 4.13 RADIOACTIVE MATERIALS SOURCES SURVEILLANCE 4-56 4.14 REACTOR BUILDING PURGE EXHAUST SYSTai 4-57 4.15 MAIN STEAM SYSTEM INSERVICE INSPECTION 4-58 4.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE 4-59 4.17 S110CK 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 SYSTD1 4-73 4.18.3 DELUGE / SPRINKLED SYSTEM 4-74 4.18.4 CO2 SYSTDI 4-74 4.18.5 HALON SYSTDtS 4-75 4.18.6 HOSE STATIONS 4-76 4.19 OTSG TUBE INSERVICE INSPECTION 4-77 4.19.1 STEM! GENERATOR SAMPLE SELECTION AND INSPECTt0N 4-77 METHODS 4.19.2 STEMI GENERATOR TUBE SM1PLE 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 SIT.E 5-1 5.2 CONTAINMENT 5-2 5.2.1 REACTOR BUILDING 5-2 5.2.2 REACTOR BUILDING ISOLATION SYSTEM 5-3 5.3 REACTOR 5-4 5.3.1 REACTOR CORE - 5-4 5.3.2 REACTOR COOLANT SYSTEM 5-4 ' 5.4 NEW'AND SPENT FUEL STORAGE FACILITIES 5-6 4 5.4.1 NEW FUEL STORAGE 5-6 5.4.2 SPENT FLEL STORAGE 5-6 5.5 AIR INTAKE TUNNEL FIRE PROTECTION SYSTEMS 5-8 e 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 0.3 STATION STAFF QUALIFICATIONS 6-3 6.4 TRAINING 6-3 6.5 FEVIEU AND AUDIT 6-3 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) 6-3 6.5.2.A MET-ED CORPORATE TECHNICAL SUPPORT STAFF 6-5 6.5.2.B GENERAL OFFICE REVIEW BOARD (GORB) 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 REQUIREMENTS 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 6.11 RADIATION PROTECTION PROGRAM 6-20 6.12 DELETED 6-20 6.13 HICil RADIATION AREA 6-21 6.14 FIRE PROTECTION INSPECTIONS 6-26 6.15 PROCESS CONTROL PROGRAM 6-27 i 6.16 0FFSITE DOSE CALCULATION MANUAL 6-27 I 6.17 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS 6-29 l v

LIST OF TABLES Table Title Pjygg

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 Effluent Monitoring Instrumentation 3/4 3-96 4.3-11 Radioactive Liquid Effluent Monitoring Instrumentation 3/4 3-99 Surveillance Requirements 3.3-12 Radioactive Gaseous Effluent Monitoring Instrumentation 3/4 3-102 4.3-12 Radioactive Gaseous Effluent Monitsring Instrumentation 3/4 3-107 Surveillance Requirements 4.11-1 Radioactive Liquid Waste Sampling and Analysis Program 3/4 3-113 4.11-2 Radioactive Gaseous Waste Sampling and Analysis Program 3/4 3-120 3.12-1 Radiological Environmental Monitoring Program 3/4 3-133 3.12-2 Reporting Levels for Radioactivity Concentrations in 3/4 3-140 Environmental Sample 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 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 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 I Table Title Page 4.15-1 Radio _ active 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 l l l G vii

1 y .. .e. LIST OF FIGURES Figure Title 2.1-1 TMI-1 Core Protection Safety Limit 2.1-2 TMI-1 Core Protection Safety Limits 2.1-3 TMI-1 Core Protection Safety Bases 1 2.3-1 TMI-1 Protection System MaRimum Allowable Set Points I 2.3-2 Protection System Maximum Allowable Set Points for Reactor

                         . Power Imbalance, TMI-1 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)                                                l 3.1-3     Limiting Pressure vs. Temperature Curve for 100 STD cc/ Liter H 2O l

3.5-1 Incore Instrumentation Specification Axial Imbalance Indication, TMI-1 1 3.5-2 Incore Instrumentation Specification Radial Flux Tilt Indication, TMI-1 3.5-2A Rod Position Limits for 4 Pump Operation From 0 to 125 i 5 EFPD, TMI-1 3.5-2B Rod Position Limits for 4 Pump Operation from 125 1 5 EFPD to EOC, TMI-1 3.5-2C Rod Position Limits for 2 and 3 Pump Operation from 0 to 1251 - 5 EFPD, TMI-1 3.5-2D Rod Position Limits for 2 and 3 Pump Operation from 125 i 5 EFPD to EOC, TMI-1 3.5-2E Power Imbalance Envelope for Operation from 0 EFPD to EOC viii i

.~.

LIST OF FICURES Figure Title-

4. ,

3.5-2F Deleted-3.5-2G LOCA Limited Maximum Allowable Linear Heat-TMI-l

\-

3.5-2H APSR Position Limits for Oparation from 0 EFPD to EOC 3.5-2I Deleted 3.5-2J Deleted . 3.5-2K Deleted s 3.5-2L Deleted 3.5-2M Deleted . 3.5-2N Deleted 3.5-3 Incore Instrumentation Specification, TMI-1 4.2-1 Equipment and Piping Requiring Inservice Inspection ira 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, THI-1 4.4-3 Ring Girder Surveillance Crack Pattern Chart, TMI-l 4.4-4 Ring Girder Surveillance Crack Pattern Chart, TMI-l 4.4-5 Ring Girder Surveillance Crack Pattern Chart, TMI-l 5-1 Extended Plot Plan TMI 5-2 Site Topography 5 Mile Radius 5-3 Site Boun/ary for Gaseous Ef fluents 5-4 Site doundary for Liquid Effluents 6-1 Met-Ed Corporate Technical Support Staff and Station Organization Chart ix

i i 1.0 DEFINITIONS - Dose Equivalent I-131 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, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Iable 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 aseessment of channel response when the channel sensor is exposed to a radioactive source. SOLIDIFICATION 1.28 SOLIDIFICATION shall be the conversion of radioactive wastea from liquid treatment systems to a uniformly distributed, e.onolithic immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing). - TMI-1 1-6

OFFSITE DOSE CALCULATION MANUAL ((X)CM) 1.29 An OFFSr"E DOSE CALCULATION MANUAL (ODCM) shall be a manual containing the. methodology and par uneters to be used in the calculation of off-site doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous ' and liquid effluent monitoring instrumentation alare/ trip setpoints. . r PROLESS CONTROL PROGRAM (PCP), 1.30 The PROCESS CONTROL PROGRAM shall contain the sampling, analysis and formulation determination by which SOLIDIFICATION of radioactive wastes from

                'lig.id systems is assured.                         .

GASEOUS RADWASTE TREATMENT SYSTEM

1. .T 1 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant systt:n off graes from the primary system and providing for delay or holdup for the purpose c.f redo ing 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 effluents by passing ventilation or vent exhaust gases through charcoal absorbers end/or HEPA filters for the purpose of removing lodines or pat';iculates from the q aeous exhaust stream prior to the release to the environment. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILLATION EXHAUST TREATMENT SYSTEMS. PURGE - PURGING 1.33 PURGE or PURGING is the controlled process of discharging air or gas from a 'confinemer.t to maintaia temperature, pressure, humidity, concentration or other operating conditions in such a manner that replacement' air or gas is required to , purify the confinement. l VENTING

. 1.34 . VENTING is the controlled process of discharging air or gas from a t

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 system names does not imply a VENTING process. l . THI-1 1-7

                     - - - -                        , , ,, . . -       ,,,w,- , , , , v - + - - ~ - -* w- w-~e----*-ear-e +   ww '""'m-~T-   th*m'r

TABLE 1.2 FREQUENCY NOTATION

NOTATION FREQUENCY S Shiftly (once per 12 hours)

D Daily (once per 24 hours) W Weekly (once per 7 days) M Monthly ( once per 31 days - Q Quarterly (once per 92 days) SA Semi-annually (once per 184 days)

             .                                   Refueling interval (once per 18 months)

S/U Prior to each reactor startup P Completed Prior to each release N/A Not applicable

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

i l l l l

        ,                                       1-8

INjiTRUAENTATION 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 set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The ALARH/ TRIP setpoints of these channels shall be determined in accordance with the Off site Dose Calculation Manual (00CM). 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 effluents monitored by the effected channel or declare the channel inoperable.

b. With less Ulan 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 MCSES c... 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 RM-L7 is operable.

THI-1 3/4 3-95

                                                                            %  1

TABLE 3.3-11

                                                                               . RADI0 ACTIVE LIQUID EiFLUENT MONITORING INSTRUMENIATION                                 .

MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION ,

1. Gross Radioactivity Monitors Providing Automatic Termination '

of Release L

a. Unit 1 Liquid Radweste Effluent 1 18 Line (RM-L6)

2. Gross Radioactivity Monitors Not Providing Automatic Termination of Release

a. Station Effluent Line (f:4-t i) 1 20 $

A R n e t V E

e

.g

I L TABLE 3.3-11 (Continued)

                                     ,  RADIDACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT                                                 OPERABLE                                      ACTION J. Flow Rate Measurement Devices                                             .

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

b. Station Ef fluent Discharge 1 21 (FT-146)

E

                                                                                                                          .A P

5 9 4 g

                                                  . . _ . , .         _                . _ . . . _ _ _ _. -. _ ,    -- __    _____._m -

TABLE 3.3-11 (Continued) . TABLE NOTATION ACTION 18 With the number of channels DPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent 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 technicakly qualified members of the Unit staff independently verify the release rate calculations and verif the discharge valve lineup.

3. Manager Unit 1 shall approve each release.

Otherwise, suspend release of radioactive effluents via this pathway. ACTION 20 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releasee 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 microcuries/ml. ACTION 21 With the number of. channels OPERABLE .less than required by the Minimum Channels OPERABLE requirement, radioactive effluent 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 flo'w. 'l l l l l l

THI-1 3/4 3-98

i TABLE 4.3-11 RADIDACT.IVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIRDtENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST

1. Radioactivity Monitors Providing Alarm and Automatic Isolation ,

a. Unit 1 Liquid Radweste Effluents D P R13) Q(1) '

Line (RM-L6)

2. Flow Rate Monitors

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

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

( n

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

a. Station Effluent Line (RM-L7) D M R(3) Q(2)

E e #

TABLE 4.3-11 (Continued) TABLE NOTATION f (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 - circuit failure)

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

4 (Includes - circuit failure) f 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). I 2. Instrument indicates a down scale failure (Includes - circuit .L failure) .

3. Instrument controls moved from the operate mode.

4 (3) The initial CH/iNEL 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-l tion should be used. (Operating plants may substitute previously established calibration procedures for this' requirement) (4) CHANNEL CHECK shall consist of verifying indication of flow during periods l of release. CHANNEL CHECK shall be made at least once daily on any day on l which -continuous, periodic, or batch releases are made. l THI-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 3.11.2s1 are not exceeded. The Alarm / Trip setpoints of these channels shall be determined in accordance with the ODCH. APPLICABILITY: As shown in Table 3.3-72. 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 operable, take the ACTION shown in Tab 13 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 l

TABLE 3.3-12 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION

1. Weste Gas Heldup System

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

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

2. Waste Gas Holdup System Explosive Gas Monitoring System

a. Hydrogen Monitor 2 *

30 y

30

b. Oxygen donitot 2 4

n e

TABLE 3.3-12 (Continu:d) RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM *' e CHANNELS ' OPERABLE APPLICABILITY ACTION INSTRUMENT i r

3. Containment Purge Monitoring System Noble Gas Activity Monitor (RM-A9)

27

a. 1 Iodine Sampler (RM-A9)

31

b. 1

31

c. Particulate Sampler (RM-A9) 1 Effluent System Fion Rate

26

d. 1 Measuring Device N.5-1t.3)

Sampler flow Rate Monitor

26

e. 1

    '                                                                                                                                                               E 7

m n W O 9 -e

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

TABLE 3.3-12 (Continued) RADI0 ACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT ' OPERABLE APPLICABILITY ACTION i 1 4." Condenser Vent System

a. Noble Gas Activity Monitor (RM-AS) 1 *' 27 7

n - n k E. e e __ ww- -

                                                            - - - - , - -        ,              e < +,    , - ,

TABLE 3.3-12 (Continued) 1 RADI0 ACTIVE CASE 005 EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS OPERABLE APPLICABILITY ACTION INSTRUMENT

5. Auxiliary and Fuel Handling Building Ventilation System

27

a. Noble Gas Activity 1 Mon!. tor (RM-A8) or (RM-A4 and RM-A6)

31

b. Iodine Sampler (RM-A8) or (RM-A4 and 1 RM-A6.)

Particulate Sampler 1 a 31 c. (RM-A8) or (RM-A4 and RM-A6) Erfluent System Flow

26

d. 1 Rate Measuring Device 8 (PT-151) or (FT-149 and FT-150) T n

e. Sampler Flow Rate Monitor 1

26 -g e

                                                                                                        >4 4

e

TABLE 3.3-12 (Continued) TABLE NOTATION

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

ACTION 25 With the number of qhannels OPERABLE less than required by the Minimum 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:

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 channe'Is 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 orovided grab samples are taken at least once per 8 hours and these samples are analyzed for gross activity within 24 hours.

ACTION 30 With the number of channels OPERABLE one less than required 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. 1 TMI-l 3/4 3-106 1 l l 1

i ! . TABLE 4.3-12

f RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNELS SOURCE CHANNEL FUNCTIONAL C. 2Cl( CHECK CALIBRATION TEST APPLICABILITY INSTRUMENT

1. Waste Gas Holdup System (RM-A7)

Q(1) k**

a. Noble Gas Activity Monitor P P R(3)

b. System Effluent Flow Rate P N/A R Q Measuring Device

2. Waste Gas Holdup System Explosive Gas Monitoring System

a. Hydrogen Monitor D WA Q(4) M

b. Oxygen Monitor D N/A Q(5) H E

t n r

 .                                                                                                                          7

TABLE 4.3-12 (Continu d) RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTA' ION SURVEILLANCE REQUIREMENTS CHANNEL CHANNELS SOURCE CHANNEL FUNCT10NAL 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 Efflueni Flow Rate D WA R Q Measuring Device

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

                                                                                                                                                                                            ,A   >

R n

                      ..e -                                                                                                                .

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

4. Condenser Vent System -

a. Noble Gas Activity Monitor (RM-A5) D M R(3) Q(2) -

r

t

                                                                                                                               ,A R

n 4 7 5

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

5. Auxiliary and Fuel Handling ,

Building Ventilation 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 o

N

                                                                                                                       ,A n

w J.

5 _ ._ _ _ l

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 clasure of WDG-V47 and RM-A8 and FT-151 are operable.

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway (for RM-A8 or RM*A4 and RM-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) 9 (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. Instrument 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, sourcer 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 percent oxygen, balance nitrogen, and

2. Four volume percent oxygen, balance nitrogen.

TMI-1 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 radiocctive 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/mi - total activity. APPLICABILITY: At all times. ACTION:

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 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 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 sampics 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 SAWLING 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 10-2 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       i x 10-4 Entrained Gases (Gamma Emitters)

P Q Gross Beta 8 5 x 10-8 Each Batch Compositec c P-32 , 1 x 10-6 Sr-89, Sr-90 5 x 10-8 e Fe-55 1 x ".0-6 W A.2 Continuous Continuouse Composite c Principal Gamma 5 x 10-7 Releases Emitters 9 f (RML-7) I-131 1 x 10-6 M GRAB Sample M Dissalved and 1 x 10-5 Entrained Gases (gamma emitters) M H-3 1 x 10-3 Continuouse Composite C Gross alpha 1 x 10-7 Continuouse Com sitec Sr-89, Sr-90 5 x 10-8 Fe-55 1 x 10-6 P-32 1 x 10-6 i TMI-1 3/4 3-113

 .  *                              )JP_E 4.11-1 (Continued _),

TABLE NOTATION

a. Ihe LLD is the smallest concentration of radioactive material in a sample

that will be detected with 95:5 probability with 5%(probability of falsely concluding that a blank observation represents a "real" signal) uocating. For a particular measurement system (which may include radiochemical separation): 4.66 sb LLD = E x V x 2.22 x 100 x Y.x exp (- L.t) 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 efficiency (as counts per transformation), , V is the sample size (in units of mass or volume), 2.27 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 At is the elapsed time between midpoint of sample 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 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 a t shall be used in the calculation.

b. A composite sample is one in which the quantity of liquid sampled is proportion.a1 to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is repre-sentative of the liquids released.

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

l THI-1 3/43114

TABLE 4.11-1 (Continued) TABLE NOTATION

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

Prior to sampling fr- . alyses, each batch shall be isolated, and then thoroughly mixed, b, s method described in the 00CM, to assure represen-tative sampling.

e. A continuous ral ne is the discharga of liquid wastes of a nondiscrete volume; e.g., f:s.n 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, Ce-141, and Ce-144. This list does not'mean that only these nuclides are to be detected and reported. Other peaks which are measurable and indentifiable, tcgether with the above nuclides, shall also be l'entified and reported. Nucildes 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.

g. If the calculated gross beta concentration exceeds 1 x 10-7 microcus as/ml, an isotopic analysis shall be performed to determine either the concentra-tion of Sr-89 and of Sr-90, or the concentration of gross strontium assumint
that all strontium prcsent is Sr-90.

1 1 THI-1 3/4 3- 115

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

a. During any calendar (parter (o j).5 mrem to the total bcdy and to f,5

mrem to any organ,

b. During any calendar year to f,3 mrem to the total body and to jJO mrem to any organ.

APPLICABILITY: At all times. I ACTION:

a. With the calculated dose from the release of radioactive materials in 11guld 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) 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 quarters so that the cumulative dose or dose committment to any individual from such releases during these four calendar quarters is within 3 mrem to the total body and 10 mrem to any organ. This 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.

SURVEILLANCE REQUIREMENTS 4.11.1.2. Dose Calculations. Cumulative dose contributions from liquid effluents shall be determined in accordance with the Offsite Dose Calculation Manual (00CH) at least once a conth. l THI-1 3/4 3- 116 l

RADI0 ACTIVE EFFLUENTS LIQUID WASTE TREATMENT

LIMITING CONDITION FOR OPERATION 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 doses due to the liquid effluent from the 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. . ACTION:

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 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.

SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses d 2 to liquid releases shall be pr6jected at least once per 31 days, in acco 9nce with the ODCM. 4.11.1.3.2 The liqu.'d radwaste treatment system shall be demonstrated OPERABLE by operating 'he liquid radwaste treatment system equipment for at least 60 minutes qu -terly unless the liquid radwaste system has been utilized to process radioactive liquid eff1 tents during the previous 92 days. THI-1 3/4 3- 117 , i I i l l 1

RADIOACTIVE EFFLUENTS LIOUID IOLDUP TANKS LIMITING CONDITION FOR OPERATIONS 3.11.1.4 The quantity of radioactive mater 3a1 contained in each of the following tanks shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.

a. Outside temporary tank ,

APPLICABILITY: At all times. ACTION:

a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents to within the limit.

4 SURVEILLANCE REQUIREMEh7S 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents weekly when radioactive materials are being added to the tank. L L 3/4 3-118

RADIOACTIVE EFFLUENTS 3/4.11.2 CASEOUS EFFLUENTS DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate, due to radioactive materials released in gaseous

                'ffluents from the site (see Figure 5-3) shall be limited to the following:

a. For noble gases: 1 500 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 radionuclides (other than noble gases) with half lives greater than 8 days: 1 1500 mrem /yr to any organ.

APPLICABILITY: At all times. ACTION:

a. With the release rate (s) exceeding the above limits, immediately decrease the release rate to comply with the above limit (s) 1

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, r.

SURVEILLANCE REQUIRDENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be deterniined to be within the above limits in accordance with the methods and procedures of the ODO!. 4.11.2.1.2 The dose rate of radioactive :naterials, other than noble gases, in gaseous effluents shall be determined to be within the above limits in 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 Table 4.11-2. t THI 3/4 3- 119 l

TABLE _4.11-2 RADI0 ACTIVE CASE 005 MASTE SAMPLING AND ANALYSIS PROGRAM Sampling Minimum Lower Limit Gaseous Release Type Frequency Analysis 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 Crab Emitters 8 t Sample h' i P P

8. Containment Each Purge b Each Purge b Principal Camma 1 x 10-4 Purge Grab Emitters 9

            .<                         Sample H-3                     1 x 10-6 C. Auxiliary and          Mb ,c,e       Mb           Principal Gamma         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 Camma 1 x 10-11 Emitters 9 (I-131, Others) Q Continuousf Composite Gross alpha 1 x 10-11 Particulate Sample Q Continuousf Composite Sr-89h, Sr-90 1 x 10-11 Particulate l Sample Continuousf Noble Gas Noble Cases 1 x 10-6 Monitor Gross Beta and Gamma E. Condenser vacuum Mb ,h M Principal Gamma 1 x 10-4 Pumps' Exhaust h Grab sample Emitters H-3 1 x 10.6 i THI-1 3/4 3-120 l n ww* - r

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 55 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 100 x Y x exp (- A4c) , 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 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 a t is the elapsed time between midpoint of sample 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 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 A t shall be used in the calculation. l l b. Analyses shall also be performed following shutdown, startup or a thermal l power level change exceeding 15% of RATED THERMAL POWER in a one hour l period.

c. Tritium grab samples shall be taken at least once 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).

e. Tritium grab samples 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 exc 's-sively are the following radionuclides: Kr-8*/, Kr-88, Xe-133, Xe-133m, Xe-135 and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-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 analyres chall 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.

THI-1 3/4 3- 122

c

 . o RADI0 ACTIVE EFFLUENTS DOSE, NOBLE CASES LIMITING CONDITION FOR OPERATION 3.11.2.2     The air dose due to noble gases released in gaseous effluents from the unit (see Figure 5-4) shall be limited to the following:

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

                      < 10 mrad for beta radiation.

b. During any calendar year: < 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 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 cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive noble gases in gaseous 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 be determined in accordance with the Offsite Dose Calculation Manual (ODCM) monthly. 1 I l 1 TMI-1 3/4 3- 123

i 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 radiolodines, 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. 4 ,

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

APPLICABILIYi At all times. ACTION: With the calculated dose from the release or radiciodines, f.adioactive materials in particulate form, or radionuclides (other thar noble gases) with half-lives greater than 8 days in gaseous effluents exceeding any of the above limits, 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 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 dose or dose committment to an individual from such releases during these four calendar quarters is within 15 mrem to any organ. SURVEILLANCE REQUIREFENTS 4.11.2.3.1 Dose Calculations Cumulative dose contributions for the current ~ calendar quarter and current calendar year shall be determined in accordance with the ODCM monthly. - l l l THI-1 3/4 3- 124

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

a. With the GASEQUS RADWASTE 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.

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

RADIOACTIVE EFFLUEns'S EXPLOSIVE CAS MIXTURE LIMITING CONDITION 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 system greater than 2% by volume but less than or equal to 4% and the hydrogen concen-tration greater than 4% by volume, reduce the oxygen concentration to

b. the Withabove limits within 48 the concentration hourt,in the waste gas holdup system of oxygen greater than 4% by volume and the hydrogen concentration greater than 4% by volume, imecdiately 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.

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 Specificat. ion 3.3.3.10. TMI-1 3/4 3-126

RADI0 ACTIVE EFFLUENTS GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION 3.11.2.7 The quantity of radioactivity contained in each gas storage tank shall 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. 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. i l l 1 THI-1 3/4 3-127 l

RADIDACTIVE EFFLUENTS 3/4.11.3 SOLID RADI0 ACTIVE WASTE ] LIM' TING 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, end to ensure the meeting of the requirements of 10 CFR Part 20 and/or 10 CFR Part 71 rzior to shipment of containers of radioactive wastes from the site. , f 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 any 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:

l

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 j of wastes, and

4. Summary der _1ption of action (s) taken to prevent a recurrence.

SURVEILLANCE 4.11.3.1.1. The solid radwaste system shall be cemonstrated OPERABLE Quarterly. I l TMI-1 3/4 3- 128 - l ; l . I

   . o SURVEILLANCE REQUIREMENTS (Continued)

e. 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 suspended 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 shall provide for the collection and testing of representative test specimens from each consecutive batch of the scme type of net waste until 3 consecutive initial test specimens demonstrate SOLIDIFICATION. The Process Control Program shall be modified as required, to assure SOLIDIFI-CATION of subsequent batches of waste.

THI-1 3/4 3-129

RADI0 ACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE 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) uver 12 consecutive months. .

        '  APPLICABILITY: At all 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.a, 3.11.2.2.b, 3.11.2.3.a, or 3.11.2.3.b,

                  ;n lieu of any other report required by Specification 6.9. 2, prepare and 3ubmit 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 effluent 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 Specification 3.11.4, and if the release condition resulting in violation of 40 CFR 190 has not alreadj been corrected, the Special 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

i. - 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 g specification.

1 SURVEILLANCE REQUIREMENTS 4.11.4 Dose Calculations Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, - and in accordance with the ODCM. h THI-1 - - 3/4 3- 130 t

    . o 3/4.12       RADIOLOGICAL--ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1      The radiological environmental monitoring program shall be conducted i ._          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-i mission, in the Annual Radiological Operating Report, a description of l 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) f When radionuclides other than those in Table 3.12-2 are detected and i 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 report is not required if the measured level of ' radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual 1 Radiological Environmental Operating Report. l I THI-1 3/4 3- 131 4 e ,- -

                                                                       +c   , ,   - -.,-.- .   ,-e ,x- , --   m 1 -- - .- , - - ,

c. With milk or fresh leafy vegetables unavailable from one or more of the sample locations required by Table 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 l cause of the unavailability o# samples and identifies locations for i obtaining replacement samples. The locations from which samples were unavailable may then be deleted from those required by Table 3.12-1, provided the locatiota from which the replacement 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 t able and figure in the DOCH and shall be analyzed pursuant to the requirements ai Table 3.12-1 and 4.12-1. e THI-1 3/4 3- 132 i

J ' i 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 Radioiodine 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 n when gross beta activity C is > 10 times the calendar /s yearly mean of control y samples. Perform gamma ); 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 dose. At least from Table 2 of the ODCN once per 92 days.

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

   .  .. ,.mpie soc.tions .re ,1ven on th. f1,ur. .ne <asie 1n the ee<M.

gi e

                                                        ,,v..       ,        -.                  - . .

TABLE 3.12-1 (Continued) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM I 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

sample collected Gamma isotopic analysis from Table 3 of the ODCM. Over a period of j:,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 Cross beta and gamma from Table 3 of the ODCM. over a period of j:,31 days. isotopic analysis of' each composite sample. e Tritium analysis of C composite sample at least -A
once per 92 days. , i

                                                                                                                       ?s     ,

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 sample.

from Table 4 of the ODCH.

Composite samples shall be collected by collecting an aliquot at intervals not exceeding 24 hours.

    ** Sample locations are shown on the figure in the ODCM.

. J.

 -             -                                                                                                       5 G
                                                                                                .- v                      __

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

4. INGESTION A minimum of 4 locations At least once per 15 days Gamma isotopic and

a. Milk I-131 analysis from Table 5 of the ODCM. when animals are on pasture; at least once per 31 days of each sample.

at other times. Fish-an ATminimumof2 locations One sample in season, or at Gamma isotopic analysis

b. on edible portions.

Invertebrates from Table 6 of the ODC21. least once per 184 days if not seasonal. One sample of each 'of the following species: y

1. Preditor (channel catfish or Bluegill or Pumpkinseed). di

2. Prey

c. Food Products A minimum of 4 locations At time of harvest. One Gamma isotopic analysis from Table 7 of the ODCM sample of each of the fol- on edible portion.

(when available), lowing classes of food products:

1. Fruits

2. Vegetables Indicator Location and At time of harvest. One I-131 analysis.

Control Location sample of broad leaf vegetation

    ** Sample locations are shown on the figure in the ODCH.                                                             1 5

O e

TABLE 4.12-1 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)a,e Airborne Particulate Water or Gas Fish Milk Food Products Sediment Analysis (r?i/1) (pCi/m3) .(pC1/kg, wet) (pCi/1) (pCi/Kg, wet) (pCi/Kg, dry) gross beta 4 1 x 10-2 3M 2000 54Mn 15 130 59Fe 30 260 58,60Co 130 e; 15 C 65Zn 30 260 y 95 Zr . 30 {

     >]-Nb              15 131 1                jc                                 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 5 O O -

                   ,                 . . . _ . _ _ . _ _ . -                         _         _         -4          ,      ,                          ,

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 (which may include radiochemical separation):

4.66sb I LLD = i E . V . 2.22 . Y . exp(- A a t) where-LLD is the lower "a priori" limit of detection as defined above (as pCi per unit mass or volume). sb is the standard deviation of thp background operating rate or of the counting rate of.a blank sample as appropriate (in 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 is the number of transformation per minute per picoeurie. Y is the fractional radiochemical yield (when applicable). Is the radioactiv.e 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 i rate er of the counting rate of the blank samples (as appropriate) rather i than on an unverified theoretically predicted variance. In calculating the s 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 samples). Typical values of E, V, Y and A t shall be used in the calculations. THI,1 L3/4 3- 137

                                                         - ,r       -      - -     ._-p-   , - -

TABLE 4.12-1 (Continued) TABLE NOTATION

b. LLD for drinking water.

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

e I i 1 l l I THI-1 3/4 3- 138 s - , - - r, , use-- --- - - - - g p-, - -r-

RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDITION 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 REQUIEEMENTS 4.12.2.1 The land use census shall mout 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 consult ing local agriculture authorities. THI 3/4 3-139 D** j] ;3 __ m . -

TABLE 3.12-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Reporting Levels Water Airborne Particulate Fish Hilk Food Products Analysis (pCi/1) or Cases (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 104 ; Co-60 3 x 102 1 x 104 Zn-65 3 x 102 2 x 104 A Zr-Nb-95 4 x 102 { 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 3 x 102 m (a) For drinking water samples. This is 40 CFR Part 141 value. E e'

.m

RADIOLOGICAL ENVIRONMENTAL MONITORING

           - 3/4.12.3 INTERLABORATORY COMPARIS0N 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 app.oved 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. f TMI-l 3/4 3 141 _. . _ , - . - . ,..e-, w - ,.

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 00CM to ensure that the alerm/ trip will occur prior to exceeding the limits of 10 CFR Part 20. . 3/4.3.3.9 RADI0 ACTIVE CASE 0US EFFLUENT INSTRUMENTATION The radioacti e 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 w.ith NRC approved methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. 1 THI-1 B .3/4 3- 142 l I

3/4.11 RADI0 ACTIVE EFFLUENTS

           ' BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1          CONCENTRATION This specification is provided to ensure that the concentration of radio-active materials released in liquid waste effluents f rom the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix 8, 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 is the controlling radioirotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in Internaticnal Canmission on Radiological Protection (ICRP)

Publication 2. 3/4.11.1.2 DOSE Thir, specification is provided to implement the requirements of Sections II.A, 171.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 opera ing flexibility and at the same time implement the guides set forth in Section_ IV.A of Appendix I to assure thrt the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". Also, for fresh water sites with drinking water supplies which can be potentially af fected by plant operation; , there is reason-able assurance that the operation of the facility will not res J t in radionuclide concentrations in the finished drinking water that are in exce is of the require-ments of 10 CFR 20. The dose calculations in the ODCM impleme it the requirements in Section III.A of Appendix I that conformance with the guide of Appendix I is to be shown by calculational procedures based on models and da a such that the actual exposure of an individual through appropriate pathways it unlikely to be substantially underestimated. _ The equations sp cified in the OL 7M for calculating the doses due to the actual release rates of radioactive materia.s in liquid effluents will be consistent with the methodology provided in Regalatory Guide 1.109,. " Calculation of Annual Doces to Man from Routine Releases oJ 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.

              .TMI-1 B 3/4 3- 143 5

              ~ RADI0 ACTIVE EFFLUENTS BASES-This specification applies to the release of liquid effluents from each reactor at the site.                                                                  r 3/4.11.1.3        LIQUID WASTE TREATMENT
                       ' The use of the liquid radwaste treatment 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 effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50' and design objective Section II.D of Appendix A to 10 CFR Part a

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

3/4.11.2 GASEOUS EFFLUENTS 1 3/4.11.2.1 DOSE RATE I 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 - spec'ified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For individuals who may at times be within the exclusion area boundary, the occupancy of' the individual will be sufficiently low ~ to compensate for any L increase in' the al.caspheric diffusion factor above that for the exclusion area boundary. .The specified release rate limits restrict, at all times, the cor-responding gamme and beta dose rates above backgrnund to an individual at or beyond the exclusion area boundary.to < 500 mrem / year to the total body or to

                  < 3000 mrem / year.to the skin. These release rate limits also restrict, at all Times, t.he corresponding thyroid daue rate above background to an infant via the cow-milk-infant pathway to < 1500 mrem / year for. the' nearest cow to the plant.

1 oo o TMI-1 B 3/4 3 144 J es ev Ju ev Ju ebu .\ W as 9

RADI0 ACTIVE EFFlufNTS 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 ConditionThefor Operation implements the guides set forth in Section II.B of Appendix I. ACTION statements provide the requir ed 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 effluents will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conform with the guides of Appendix I to be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the 00CM for calculating the doses due to the actual release rates of radio-active no51e gases in gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1. iO9, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for determining the air doses at 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. 3/4.11.2.3 DOSE. RADIOI0 DINES, RADI0 ACTIVE MATERIAL IN PARTICULATE FORM AND RADIONUCLIDES OTHER THAN NOBLE GASLS This specification is provided to implement the requirements of Eections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guiden 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 Sect ion IV. A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methnds specified in the surveillance requirements implement the requirements in Section III. A of Appen-dix I that conformance with the guiden 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 in 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 of Reactor Ef fluents for the Revision I, I," Purpose of Evaluating Compliance with 10 CfR Part 50, Appendix THI-1 03/43-145

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 pr rticulate form and radionuclides other than noble gases are dependent on t,te existing radionuclide pathways to man, in the unrestricted area. The pa'. sways which are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposition of radionucliceu onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man. 3/4.11.2.4 GASEOUS WASTE TREATMENT The use of gaseous radwaste treatment system ensure that the system will be available for 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 suitable fraction of the guide set forth in Sections II.B and II.C of Appendix 1,10 CFR Part 50, for gaseous effluents. TMI-1 03/43-146

1 s '. RADI0 ACTIVE EFFLUENTS BASES 3/4.11.2.7 -CAS STORACE TANKS Restricting the quantity of radioactivity contained in each gas storagt 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. J t TIII-1 B 3/4 3-147 1

a. .

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.1 MONITORING PROGRAM The radiological monitoring program required by this specification provides maasurements of radiation and of radioactive materials in those exposure pathways

          . and for those radionuclides which lead to the highest potential radiation exposures of individuals resulting from the station operation.      This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the 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 based on operational experience. 3/4 12.2 LAND USE CENSUS This specification is provided to ensure that changes F +he use of unre-stricted areas are identified and that modifications to the .nonitoring 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 was 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 ihe requirement for participation in an Inter laboratory Comparisen 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 onet of a quality assurance program for environmental monitoring in' order to demonstrate that the results are reasonably valid. l THI-1 H 5/4 3- 148 1

5.0 DESIGN FEATURES 5.1 SITE: Applicability Applies to the location and. extent of the exclusion bour.dary, restricted area, and low population zone. Objective To define the above by location and iistance-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, inci 4 ing portions of , Three Mile Island, the. river surface around it, and e portion'of Shelley Island, which is owned by Met-Ed. The minimum distance of 2,000 ft occurs on the chorn of the mainland -in a due casterly direction from the

                  - plant 'as nhown on Figure 2-3 of. the FSAR. Figure 1-1 of the FSAR is a plot plan showinq the physical location of the fence which defines the
                   " Restricted 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.

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1 . LIQUID' EFFLUENT OUTFALL DESCRIPTIONS

.001 - Represents the main discharge for liquid cf fluents discharged from the island. .

002 } Are redundant emergency outfalls to 001

;.                                       003  )

004 i.4-1 } i ?- 4 I k. 4 5 4 r i f )' t / 4

                                            ~
    ',I                                                    '5-11' V                             -
                               . i.: .

2)[ review of violationa of applicable federal statutes, codes,

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.

9 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 c,hanges to radwaste systemt.

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.

Provide immediate written notification to the Manager-Generation ~ c. 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 'bove. a 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 includinq taking immediate emergency actions. RECORDS The Plant Operations . Review Committee shall maintain at the station 6.5.1.8 written minutes of each meeting and copies shall be provided to the Unit Superintendent, Manager-Generation Operations-Nuclear, Manager-Sneration Engineering, and the General Office Review Board Secretary.

6.5.2.A MET-ED CORP 0 RATE TECHNICAL SUPPORT STAFF l ORGANIZATION l The organization of the Met-Ed Corporate Technical Support Staff is as

6.5.2.A.1 shown on Figure 6-1 and consists of the Manager-Generation Operations Nuclear, Manager-Generation * " D TMI-l' ** * " 9

                                                . 6-5_

k. Periodically audit the areas listed below to verify compliance with the Three Mile 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 i; 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
'. cast once every two years.

                 ~1. Review the radiological environmental monitoring program and the recults 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 foi 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.

4 1 THI-1~ 6-6 a e e s.. -v v +

6.8 PROCEDURES 6.8.1 Writi3n 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 of USNRC Regulatory Guide 1.33 November 1972 except as Appendix "A" 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 proceoure and administrative policy of ' 6.8.1 above, and changes theceto, shall L; reviewed by the Plant Operations Review Committee and approved by the document. 6.8.3 Temporary changes to procedures of 6.8.1 dbove may be made provided:

s. 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 af fected.

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

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

n. 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.

i 4 I i l - 6-11 THI-1 l l - _ _ _ .

e - 4- ' 'S 4 6.9 LREPORTING-REQUIREMENTS'(cont'd) (4). Reactivity anomalies involving disagreement.with the predicted

                                       .value of reactivity balance'under steady state conditions during pow'er operation greater than or equal to 1% Ak/k; a calculated
                                                             ~

reactivity balar.ce indicating a shutdown margin less conservative than specified -in the technical l specifications; short-term

                                       - reactivity increases that correspond to a reactor period of less than 5 seconds or, if sub-critical, an unplanned reactivity insertion of more than 0.5% A k/k; or occurrence of any unplanned criticality.

(5) Failure or malfunction of one or more components which prevents or could prevent, by .itself, the- fulfillment -of the functional requirements of system (s) used to cope with accidents analyzed in , the FSAR. (6) . Personnel errortor procedural inadequacy which prevents or coulJ ! ' prevent, by itself, the fulfillment of the functional require-ments of systeme required to cope with accidents analyzed in the FSAR. Note: .For items 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 i items 6.9.2.B(2), and 6.9.2.B(3) . (7)- Conditions arising from natural or man-made events that, as a I direct result of the event require plant shutdown, operation of safety systems, or other protective messares required by tech-F nical. specifications. L. (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. i (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 th.e accident analyses in U the FSAR or Technical Specifications bases; or discovery during plant life of conditions not specifically considered in the FSAR g' or Technical Specificalions that require remedial action or correct ive measures to prevent the existence of development of an unsafe condition. * - (10) .0ffsite releases of radioactive materials in liquid and gaseobs effluents 'which exceed the limits of Specification 3.11.1.1 or 3.11.2.1. (11) Exceeding the' limits in Specification 3.11.2.7 for the storage

           "              u               :of radioactive materials in the listed tanks.

F -*Tliisi item'is. intended to provide !ar reporting of l potentially generic problems. THI-1. 6-16 , r s 77.. > z

6.9 REPORTING REQUIREMENTS-(cont'd)

JB. Thirt'y Day Written Reports. l/ ' 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 fr.r operation or plant shutdown required 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. u (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. l 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. (5) <An unplanned of fsite release of 1) more than 1 curie of radio- ^ active material 'in ' liquid effluents, 2) more than 150 curies of

                                                                                                   ~
                                      - noble gas in gaseous ef fluents, or 3) more than 0.05 curies of radiciodine in gaseous effluents.

(6) Measured. levels of radioactivity in an environmental sampling

                                      . medium determined to exceed the reporting level values of Table 3.12.2 when: averaged over any calender quarter sampling period.
                                                                            ~

i 4

                    .THI                                       6-17
                                                     <            w     -     v - v-  , - ,    .~, -,   -     -v m. ,

ADMINISTRAllVE 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 l'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 environmental surveillance reports and an assess-ment of the obsersed impacts of the plant operation 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 monitoring, 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 fracn one reactor; and the results of licensee participation in the Interlaboratory Comparison Program, required by Specification 3.12.3. SEMIANNUALRADI0ACTIVEEFFLUENTRELEASEREPORT2[ 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. 3/A single submittal m:iy be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station: however, for units with separate radwante systema, the submittal shall specify the releases of radioactive material from each unit. THI-1 6-18a sf "

ADMINISTRATIVE 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, " Measuring, 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. This 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 distributions 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 reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accor-dance with 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 Trom reactor releases and other nearby uranium fuel cycle sources (including doses from primary 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 effluent 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 quant ity (specify whether determined by measurement or estimate),

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

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

e. type of container (e.g. , LSA, Type A, Type B, Larqe Quantity) and

f. solidificat ion asjent (e.g. , cement , uren fnrmaldehyde) .

THI-1 6-10b

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. . i Any changes to the 0FFSITE DOSE CALCULATION MANUAL shall be submitted with the next Semi Annual Radioacitve Effluent Release Report. i l l l

4 TMI-l 6-18c 1

.i

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.

1. Records of quality assurance activities required by the 00A 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 Board Minutes.

1. Records of analyses required by the radiological environmental monitoring program. i 6.11 RADIATION PROTECTIONS PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR 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 systems (Liquid, gaseous and solid):

1. Shall 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 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 the predicted releases of radioactive materials in liquid and gaseous ef fluents and/or quantity of solid waste that differ from those previously predicted in the license applicatinn and amendments thereto; A

THI 6-20

e. 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 compar ison of the predicted releases of radioactive materials, in liqui t 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.

l TM1-1 6-20a

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

1. Shall be submitted to the Commission in the semi-annual Radio-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.

6.17 0FF5ITE DOSE CALCULATION MANilAL (ODCM) 6.17.1 The ODCM shall be approved by the Canmission prior to implementation. G.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 supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each paqe numbered and provided with an approval and date box, together with appropriate analyses or evaluationn juntifying the change (n):

b, a determination that the change will not reduce the accuracy or reliability of done* eniculationn or setpnint determinations: and

c. document at ion of the fact that the change has been reviewed asal found acceptable by Ihe PORC.

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

l THI-1 6-27 0 5" 'MD~ 'g' J }D dau.1.}k dJu /ftla w_}}

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