ML20236F197

Proposed Tech Specs Providing Administrative Changes

ML20236F197
Person / Time
Site:	Three Mile Island
Issue date:	07/28/1987
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20236F178	List: ML20236F175 ML20236F184 ML20236F197
References
NUDOCS 8708030248
Download: ML20236F197 (55)

Text

_ - _

i TABLE OF CONTENTS Section Page TECHNICAL SPECIFICATIONS 1 DEFINITIONS 1 -1 1.1 RATED POWER 1 -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 1 -1 1.2.4 Hot Standby 1 -1 1.2.5 Power Operation 1 -1 1.2.6 Refueling Shutdown 1 -1 1.2.7 Refueling Operation 1-2 1.2.8 Refueling Interval 1-2 1.2.9 Startup 1-2 1.2.10 TAvg 1-2 1.2.11 Heatup-Cooldown Mode 1-2 1.2.12 Station, Unit, Plant, and Facility 1-2 1.3 OPERABLE l-2 1.4 PROTECTIVE INSTRUMENTATION LOGIC 1-2 ,

1.4.1 Instrument Channel 1-2 l 1.4.2 Reactor Protection System 1-2 1.4.3 Protection Channel 1-3 1.4.4 Reactor Protection System Logic 1-3 1.4.5 Engineered Safety Features System 1-3 1.4.6 Degree of Redundancy 1-3 1.5 INSTRUMENTATION SURVEILLANCE 1-3 1.5.1 Trip Test 1-3 1.5.2 Channel Test 1-3 1.5.3 Channel Check 1 -4 1.5.4 Channel Calibration 1 -4 1.5.5 Heat Balance Check 1-4 1.5.6 Heat Balance Calibration 1-4 1.6 POWER DISTRIBUTION 1-5

1. 6.1 Quadrant Power Tilt 1-5 1.6.2 Reactor Power Imbalance 1-5 1.7 CONTAINMENT INTEGRITY 1-5 1.8 FIRE SUPRESSION WATER SYSTEM 1-5 1.12 DOSE EQUIVALENT I-131 1-6 1.13 SOURCE CHECK l-6 1.14 SOLIDIFICATION 1-6 1.15 0FFSITE DOSE CALCULATION MANUAL 1-6 1.16 PROCESS CONTROL PROGRAM l-6 1.17 GASEOUS RADWASTE TREATRENT SYSTEM 1-6 1.18 VENTILATION EXHAUST TREATMENT SYSTEM 1-6 1.19 PURGE-PURGING l-7 1.20 VENTING l-7 1.21 REPORTABLE EVENT 1-7 1.22 MEMBER (S) 0F THE PUBLIC 1 -7 i

Amendnent No. 11, 72. 129.

8708030240 070728 PDR ADOCK 05000289 "

-. .2

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l TABLE OF CONTENTS Section Page 3.16 SH0CK SUPPRESSORS (SNUBBERS) 3-63 j 3.17 REACTOR BUILDING AIR TEMPERATURE 3-80 ,

3.18 FIRE PROTtCTION 3-86 3.18.1 Fire Detection Instrumentation 3-86 3.18.2 Fire Suppression Water System 3-88 3.18.3 Deluge / Sprinkler Systems 3-89 3.18.4 C02 System 3-90 3.18.5 Halon Systems 3-91 3.18.6 Fire Hose Stations 3-92 3.18.7 Fire Barrier Penetration Seals 3-94 3.19 CONTAINMENT SYSTEMS 3-95 3.20 SPECIAL TEST EXCEPTIONS 3-95a 3.20.1 Low Power Natural Circulation Test 3-95a '

3.21 RADI0 ACTIVE EFFLUENT INSTRUMENTATION 3-96 l 3.21 .1 Radioactive Liquid Effluent Instrumentation 3-96 3.21 .2 Radioactive Gaseous Process and Effluent Monitoring 3-100 Instrumentation 3.22 RADI0 ACTIVE EFFLUENTS 3-106 3.22.1 Liquid Effluents 3-106 ,

3.22.2 Gaseous Effluents 3-111 3.22.3 Solid Radioactive Waste 3-118 3.22.4 Total Dose 3-119 3.23 RADIOLOGICAL ENVIRONMENTAL MONITORING 3-120 3.23.1 Monitoring Program 3-120 3.23.2 Land Use Census 3-125 3.23.3 Interlaboratory Comparison Program 3-127 4 SURVEILLANCE STANDARDS 4 -1 4.1 OPERATIONAL SAFETY REVIEW 4-1 4.2 EACTOR COOLANT SYSTEM 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 Deleted 4-37 4.5 EMERGENCY LOADING SEQUENCE AND POWER TRANSFER, EMERGENCY 4-39 CORE COOLING SYSTEM AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 Emergency 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 EMERGENCY POWER SYSTEM PERIODIC TESTS 4-46 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4-48 4.7.1 Control Rod Drive System Functional Tests 4-48 4.7.2 Control Rod Program Verification 4-50 111 Amendment No. 72, 81, 108, 129

I TABLE OF CONTENTS Section Page 4.8 MAIN STEAM ISOLATION VALVES -

4-51 4.9' DECAY HEAT REMOVAL CAPABILTTY - PERIODIC TESTING 4-52 1 4.9.1 Emergency Feedwater System - Periodic Testfeg 4-52 (Reactor Coolant Temperature Greater Than 250*F) 4.9.2 Decay Heat Removal Capability - Periodic Testing 4-52a (Reactor Coolant Temperature 250*F or Less) 4.10 REACTIVITY AN0MALIES 4-53 4.11 _ REACTOR COOLANT SYSTEM YENTS 4-54 4.12 AIR TREATMENT SYSTEMS 4-55 i 4.12.1 Emergency Control Room Air Treatment System 4-55 i 4 .12.2 Reactor Building Purge Air Treatment System 4-55b i

-4.12.3 Auxiliary and Fuel Handling Building Air Treatment System 4-55d 4.12.4- Fuel Handling Building ESF Air Treatment System 4-55f 4.13 RADI0 ACTIVE MATERIALS SOURCES SURVEILLANCE 4-56 4.14 DELLIED 4-56 4.15 FD31r3 TEAM SYSTEM INSERVICE INSPECTION 4-58 4.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE 4-59 4.17- SHDCK SUPPRESSORS (SNUBBERS) 4 4.18. FIRE PROTECTION SYSTEMS 4-72 4.18.1 Fire Protection Instruments 4-72 4.18.2 . Fire Suppression Water System 4-73 4.18.3 Deluge / Sprinkler System 4-74 4.18.4 CO2 System 4-74 4.18.5 Halon Systems 4-75

-4.18.6 Hose Stations 4-76 4.18.7 Fire Barrier Penetration Seals 4-76a

'4.19 OTSG TUBE INSERVICE INSPECTION 4-77 4.19.1 Steam Generator Sample Selection and Inspection Methods 4-77 4.19.2 Steam Generator Tube Sample Selection and Inspection 4-77 4.19.3 Inspection Frequencies 4-79 j 4,19.4 Acceptance Criteria 4-80 4.19.5 Reports 4-81 4.20 REACTOR BUILDING AIR TEMPERATURE 4-86 4.21 RADI0 ACTIVE EFFLUENT INSTRUMENTATION 4-87 l 4.21.1 Radioactive Liquid Ef fluent Instrumentation 4-87 4.21.2 . Radioactive Gaseous Process and Effluent Monitoring 4-90 Instrumentation 4.22 RADI0 ACTIVE EFFLUENTS 4-95 4.22.1 Liquid Effluents 4-95 4.22.2 Gaseous Effluents 4-1 01 4.22.3 Solid Radioactive Waste 4-107 4.22.4. Total Dose 4-108 4.23.1 Monitoring Program 4-117 4.23.2 Land Use Census 4-1 21 4.23.3 Interlaboratory Comparison Program 4-122 iv Amendment No. 11, 28, 30, 41, 47, 55, 72, 78, 95, 97, 119, 122, 129

LIST OF TABLES TABLE TITLE PAGE 1.2 Frequency Notation 1-8

2. 3-1 Reactor Protection System Trip Setting Limits 2-9 3.1. 6.1 Pressure Isolation Check Valves Between the Primary 3-15a ,

Coolant System and LPIS

3. 5-1 Instruments Operating Conditions 3-29 3.5-2 Accident Monitoring Instruments 3-40c 3.5-3 Post Accident Monitoring Instrumentation 3-40d 3.18-1 Fire Detection Instruments 3-87 3.21 -1 Radioactive Liquid Effluent Monitoring Instrumentation 3-97 3.21 -2 Radioactive Gaseous Process and Effluent 3-1 01 Monitoring Instrumentation 3.23-1 Radiological Environmental Monitoring Program 3-122 3.2.3-2 Reporting Levels for Radioactivity-Concentration 3-126 in Environmental Samples 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 4.1 -4 Post Accident Monitoring Instritmentation 4-10a 4.19-1 Minimum Number of Steam Generators to be 4-84 Inspected During Inservice Inspection 4.19-2 Steam Generator Tube Inspection 4-85 4.21 -1 Radioactive Liquid Effluent Monitoring 4-88 Instrumentation Surveillance Requirements 4 .21 -2 Radioactive Gaseous Effluent Monitoring 4 -91 Instrumentation Surveillance Requirements 4.22-1 Radioactive Liquid Waste Sampling & Analysis Program 4-96 4.22-2 Radioactive Gaseous Waste Sampling & Analysis Program 4-102 4.23-1 Maximum Values for the Lower Limits of Detection (LLD) 4-118 .

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vi Amendment No. 59, 72, 100, 106, 118

. forms the automatic system that protects the reactor by control rod trip. It includes the four protection channels, their associated instrument channel inputs, manual trip switch, all rod drive control protection trip breakers, and activating relays or coils.

1.4.3 PROTECTION CHANNEL A PROTECTION CHANNEL as'shown in Figure 7.1-1 of the updated FSAR (one of three'or l

.one of four. independent channels, complete with sensors, sensor power supply units, amplifiers, and bistable modules provided for every reactor protection safety parameter) is a combination of instrument channels forming a single digital output to the protection system's coincidence logic. It includes a shutdown bypass circuit, a protection channel bypass circuit and a reactor trip module.

1.4.4 REACTOR PROTECTION SYSTEM LOGIC This system utilizes reactor trip module relays (cotis and contacts) in all four of the protection channels as shown in Figure 7.1-1 of the' updated FSAR, to provide {

reactor trip signals .for de-energizing the six control rod drive trip breakers.

The control rod drive trip breakers are arranged to provide a one-out-of-two-times-two logic. Each element of the one-out-of-two-times-two logic is controlled by a separate set of two-out-of-four logic contacts from the four reactor protection channels.

1.4.5 ENGINEERED SAFETY FEATURES SYSTEM

'This system utilizes relay contact output from individual channels arranged in three analog sub-systems and two two-out-of-three logic sub-systems .as shown in Figure 7.1-4 of the updated FSAR. The logic sub-system is wired to provide l appropriate signals for the actuation of redundant engineered safety features -

equipment on a two-of-three basis for any given parameter.

1.4.6 DEGREE OF REDUNDANCY The difference between the number of operable channels and the number of channels which, when tripped, will cause an automatic system trip.

1.5 INSTRUMENTATION SURVEILLANCE l

-1.5.1 TRIP TEST

\

A TRIP TEST is a test of logic elements in a protection channel to verify their j associated trip action. '

l.5.2 CHANNEL TEST A CHANNEL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practical to verify OPERABILITY, including alarm and/or trip functions.

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1.5.3 CHANNEL CHECK

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

1.5.4 CHANNEL CALIBRATION An instrument CHANNEL CALIBRATION is a test, and adjustment (if necessary), to establish that the channel output responds with acceptable range and accuracy to known values of the parameter which the channel measures or an accurate simulation ol' these values. Calibration shall encompa:s the entire channel, including equipment actuation, alarm, or trip and shall be deemed to include the channel test.

1.5.5 HEAT BALANCE CHECK A HEAT BALANCE CHECK is a comparison of the indicated neutron power and core I thermal power, i

1. 5. 6_ HEAT BALANCE CALIBRATION A HEAT BALANCE CALIBRATION is an adjustment of the power range channel amplifiers output to agree with the core thermal power as defined by a weighted primary and l secondary heat balance considering heat losses. The weighting factor, a is shown

. in the figure below as a function of power level. The equations below define the value of a as a function of power level and the use of- a in determining the core thermal power.

1.0 ,

o BW' o

0 I

power

• 0 1-4

Core Thermal Power = a Qsec + (1-a ) Qprim for, POWER less than or equal to 15%, a=0 POWER greater than 15% AND less than 50%

a = POWER - 15 85 WHERE:

POWER = Oprim 100 Qmax POWER greater than 50% AND less than 100%

a =

P_0WER - 15 WHERE: POWER = Osec 100 85 Qmax POWER greater than or equal to 100%, a=1 1-4a

(Definitions 1.9 - 1.11 have been deleted).

1.12 DOSE EQUIVALENT I-131 The DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same _ thyroid aose as the quantity and isotopic

- mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID 14844, " Calculation of Distance Factors for Power and Test Reactor Sites". [0r in Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977.]

1.13 SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

1.14 SOLIDIFICATION SOLIDIFICATION shall be the conversion of radioactive wastes from liquid treatment systems to a uniformly distributed, monolithic immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (f ree-sta ndi ng) .

1.15 0FFSITE DOSE CALCULATION MANUAL (00CM)

An 0FFSITE DOSE CALCULATION MANUAL (0DCM) shall be a manual containing the methodology and parameters to be used in the calculation of off-site doses due to l radioactive gaseous and liquid effluents and in the calculation of gaseous and i liquid effluent monitoring instrumentation alarm / trip setpoints, and in the conduct of the environmental radiological monitoring program.

1.16 PROCESS CONTROL PROGRAM (PCP)

The PROCESS CONTROL PROGRAM shall contain the sampling, analysis, and formulation determination by which SOLIDIFICATION of radioactive wastes from liquid systems is assure d.

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

1.18 VENTILATION EXHAUST TREATMENT SYSTEM 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 and/or HEPt.

filters for the purpose of removing iodines or particulate from the gaseous exhaust system prior to the release to the environment. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEMS.

1-6 Amendment No. 72

-1.19 PURGE - PURGING PURGE or PURGING is the controlled process of _ discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions in such a manner that replacement air or gas is required to purify the confinement.

1.20 VENTING VENTING is the controlled process of dischargir sir as gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions in such a manner that replacement air or gas is not provided. Vent used in system name does not imply a VENTING process.

1 .21 REPORTABLE EVENT

' A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

1.22 MEMBER (S) 0F THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the GPU System, GPU contractors or' vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. ,

i 1 -7 Amendment No. 72

o TABLE 1.2 FREQUENCY NOTATION'

-NOTATION FREQUENCY' S Shif tly (once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />)

!' D- Daily (once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)  !

W .~ Weekly (once per 7 days)

M Monthly (once per 31 days)

'( Quarterly (once:per 92 days)

S/A Semi-Annually (once per 184 days)

R Refueling Interval P S/U Prior to each reactor startup, if not.

done during the previous 7 days P Completed prior to each release N/A (NA) Not applicable

-E Once per 18 months

.The' Surveillance Requirements shall be performed within the specified time interval with:

A. A maximum allowable extension not to exceed 25% of the surveillance interval, and B. A total- maximum combined interval time for any 3 consecutive tests, not to exceed 3.25 times the specified surveillance i nterval . ,

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I 18 Amendment No. 72 1

____-.___....__._-.-_--_-A

3.1. 6.9 Loss of reactor coolant through reactor coolant pump seals and system valves to connecting systems which vent to the gas vent header and from which coolant can be returned to the reactor coolent system shall not be considered as reactor coolant leakage and shall not be subject to the consideration of Specifications 3.1. 6.1, 3.1. 6. 2, 3.1. 6. 3, 3.1. 6. 4, 3.1.6.5, 3.1.6.6 or 3.1.6.7, cxcept that such losses when added to leakage shall not exceed 30 gpm. If leakage plus losses exceeds 30 gpm, the reactor shall be placed in HOT SHUT 00WN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection. l 3.1.6.10 Operating conditions of POWER OPERATION, STARTUP and HOT SHljTDOWN apply to the operational status of the high pressure isolation valves between the primary coolant system and the low pressure injection system.

a. During all operating conditions in this specification, all pressure isolation valves listed in Table 3.1.6.1 that are located between the primary coolant system and the LPIS shall function as pressure isolation devices except as specified in 3.1.6.10.b. Valve leakage shall not exceed the amount indicated in Table 3.1.6.1.(a)

b. In the event that integrity of any high pressure isolation check valves specified in Table 3.1.6.1 cannot be demonstrated, reactor operation may continue provided that at least two valves in each high pressure line having a non-functional valve are in and remain in, the mode corresponding to the isolated condition. (b)

c. If Specification 3.1.6.10.a or 3.1.6.10.b cannot be met, an orderly shutdown shall be accomplished by achieving HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Bases Any leak of radioactive fluid, whether from the reactor coolant system primary boundary or not, can be a serious problem with respect to in-plant radioactive contamination and required cleanup or, in the case of reactor coolant, it could develop into a still more serious problem and, therefore, the first indications of such leakage will be followed up as soon as practical. The unit's makeup system has the capability to makeup considerably more than 30 gpm of reactor coolant leakage.

Water inventory balances, monitoring equipment, radioactive tracing, boric acid i crystalline deposits, and physical inspections can disclose reactor coolant leaks.

(a) For the purpose of this specification integrity is considered to have been demonstrated by meeting Specification 4.2.7.

(b) Motor operated valves shall be placed in the closed position and power supplies deenergized.

3-13 Order dtd. 4/20/81 L ------- _ _ -- - - - - -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

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e. Core flood tank (CFT) vent valves CF-V3A and CF-V3B shall be closed and the breakers to the CFT vent valve motor operators shall be tagged open, except when at asting core flood tank level and/or pressure. Specification 3.0.1 applies.

3.3.1.3 Reactor Building Spray System and Reactor Building Emergency Cooling System i The following components must be OPERABLE:

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a. Two reactor building spray pumps and their associated spray j nozzles headers and two. reactor building emergency cooling I fans and associated cooling units (one in each train).

Specification 3.0.1 applies.

b. The sodium hydroxide (Na0H) tank shall be maintained at 8 ft.

6 inches lower than the BWST level as measured by the BWST/Na0H tank differential pressure indicator. The Na0H tank concentration shall be 10.0 .5 weight percent (%).

c. All manual valves in the discharge lines of the sodium hydroxide tank shall be locked open.

3.3.1.4 Cooling Water Systems - Specification 3.0.1 applies.

a. Two nuclear service closed cycle cooling water pumps must be OPERABLE. l

b. Two nuclear service river water pumps must be OPERABLE. l l

c. Two decay heat closed cycle cooling water pumps must be OPERABLE. l 1 1

d. Two decay heat river water pumps must be OPERABLE.

e. Two reactor building emergency cooling river water pumps must l be OPERABLE.  !

l I I 3.3.1.5 Engineered Safeguards Valves and Interlocks Associated with the I Systems in Specifications 3.3.1.1, 3.3.1.2, 3.3.1.3, 3.3.1.4 are OPERABLE. Specification 3.0.1 applies.

3.3.2 Maintenance shall be allowed during power operation on any i component (s) in the makeup and purification, decay heat, RB ]

emergency cooling water, RB spray, CFT pressure instrumentation, J CFT level instrumentation, BWST level instrumentation, or cooling water systems which will not remove more than one train of each system from service. Components shall not be removed from service so that the affected system train is inoperable for more than 72 consecutive hours. If the system is not restored to, meet the requirements of Specification 3.3.1 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, th? reactor shall be placed ia a COLD SHUTDOWN condition within twelve hours.

3-22 Amendment No. 33, 80, 98 l

3.21 RADI0 ACTIVE EFFLUENT INSTRUMENTATION 3.21.1 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.21.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.21-1 shall be OPERABLE with their alarm / trip setpoints set to ensure j that the limits of Specification 3.22.1.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the 0FFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY: At all times

• ACTION:

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

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Tabl e 3.21 -1. Exert best efforts to return the instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semi-Annual Effluent Release Report why the inoperability was not corrected in a timely manner.

• For FT-84, and RM-L6, operability is not required when discharges are positively controlled through the closure of WDL-V257.

• For RM-L12 and associated IWTS/IWFS flow interlocks, operability is not required when discharges are positively controlled through the closure of IW-V72, 75 and IW-Y280, 281.

• For FT-146, operability is not required when discharges are positively controlled through the closure of WDL-V257, IW-V72, 75 and IW-Y280, 281.

BASES 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 instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.

3-96 Amendment No. 72, 88

l l

l TABLE 3.21-1 (continued)

TABLE NOTATION ACTION 18 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue, provided that prior to initiating a release: r

1. At least two independent samples are er.alyzed in accordance with Specifications 4422.1.1A & B 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. Operations and Maintenance Director 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 releases via this pathway may commence or continue provided that grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least lx10-/ microcuries/m1, prior to initiating a release and ,

at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during release.

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, provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump curves may bc used to estimate fl ow.

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3-99 Amendment No. 72, 88

i 3.21.2 RADI0 ACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.21.2 The radioactive gaseous process and effluent monitoring instrumentation channels shown in Table 3.21-2 shall be OPERABLE with their alarm / trip setpoints set to enst re that the limits of Specification 3.22.2.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the OFFSITE D03E CALCULATION MANUAL (0DCM).

APPLICABII.ITY: As shown in Table 3.21-2.

ACTION:

a. With a radioactive gaseous process or effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the abcve specification, immediately suspend the release of radioactive effluents monitored by the affected channel or declare the channel inoperable.

h. With less than the minimum number of radioactive gaseous process or effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.21 -2. Exert best efforts to return the instrumentation l to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semi-Annual Effluent Release Report why the inoperability was not corrected in a timely manner.

BASES The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases. The alarm / trip setpoints for these l instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.

The low range condenser offgas noble gas activity monitors also provide data for determination of steam generator primary to secondary leakage rate. Channel operability requirements are based on an ASLB Order dated October 31, 1984.

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3-100 Amendment No. 72,103 l

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4 TABLE 3.21-2.

(Continued)-

TABLE ~ NOTATION .

• At all ' times.

• • During waste gas holdup system operation.

• • • Operability is not required when discharges are positively controlled

.through the closure of WDG-Y47, and RM-A8 and FT-151 are operable.

• • • • 'During Fuel Handling Building ESF Air Treatment System Operation.

1. At all times during containment purging.

-## At all times when condenser vacuum is established.

l ACTION 26 'With' the number of-channels OPERABLE less than required by the' Minimum l Channels OPERABLE requirement, the contents of the tank may be released to the environment provided that prior to initiating the release: j

1. At least two independent samples of the tank's contents are analyzed in accordance with Table 4.22-2, Item A, and l l

2. At least twol technically qualified members of the Unit staff .

independently verify the release rate calculations and verify the discharge valve lineup.

1

3. The Operations & Maintenance Director, Unit 1, shall approve each release.

Otherwise, suspend release of radioactive effluent via this pathway.

ACTION 26 With the number of channels OPERABLE less than required by the Minimum '

Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

. ACTION 27 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken.at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the initial samples are analyzed for gross _ activity (gamma scan) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the channel has been declared inoperable. If RM-A9 is declared inoperable, see also Specification 3.5.1, Table 3-5.1, Item C.3.f.

ACTION 30 1. With the number of channels OPERABLE less than required by the Miriimum Channels OPERABLE requirement, a grab sample shall be collected and analyzed for the inoperable gas channel (s) at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With both channels inoperable, a grab sample shall be collected and analyzed for the inoperable gas channel (s):

(a) at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during degassing operations.

(b) at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during other operations (e.g. Feed and Bleed).

3-105 Amendment No. . 72, 78, 104, 122

I l

l l

TABLE 3.21-2 (Continued) 1ABLE NOTATION ACTION 30 2. If the inoperable gas channel (s) is not restored to service within (CONT'D.) 14 days, a special report shall be submitted to the Regional Administrator of the NRC Region I Office and a copy to the Director, Office of Inspection and Enforcement within 30 days of declaring the channe1(s) inoperable. The report shall describe (a) the cause of the monitor inoperability, (b) action being taken to restore the instrument to service, and (c) action to be taken to prevent recurrence.

ACTION 31 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that within four hours after the channel has been declared inoperable, samples are continuously collected with auxiliary l sampling equipment.

ACTION 32 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 that one OPERABLE channel remains in service or is placed in service within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. After 28 days, or if one OPERABLE channel does not remain in service or is not placed in service within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the provisions of 3.0.1 apply.

ACTION 33 With the number of channels OPERABL E less than required by the Minimum Channels OPERAP,LE requirement, either restore the inoperable channel to OPERABLE status within 7 days, or prepare and submit a special report within 30 days outlining the action (s) taken, the cause of the inoperability, and plans and schedule for restoring the system to OPERABLE status.

l l 3-105a Amendment No. 72, 103, 104, 122

[

-3.22 RADI0 ACTIVE EFFLUENTS 3.22.1: LIQUID EFFLUENTS-3.22.1.1 CONCENTRATION E LIMITING ~ CONDITION FOR OPERATION 3.22.1.1 ~.The concentration of radioactive material released at anytime from the unit to. unrestricted areas (see-Figure 5-4) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix = B, Table II Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 3 x 10-3 uCi/cc total activity.

' APPLICABILITY: At all times.

ACTION:

a. With the concentration of radioactive material released from the unit to

. unrestricted areas exceeding the above limits, immediately restore concentration within the above limits,

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

1. At least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />,

2. At least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

3. At least COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

BASES This. specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the unit to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II. LThis limitation provides additional assurance that the levels of radioactive materials .in bodies of water outside the site will not result in exposures with .(1)~ the Section II. A design objectives of Appendix I,10 CFR Part 50, to a MEMBER OF THE PUBLIC and (2) the limits of 10 CFR Part 20.106 (e) to i the population. The concentration limit for noble gases is based upon the  ;

assumption the Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP)

Publication 2.

I l

3-106 Amendment No. 72 i

a_.-___.-_. ___ _ _ I

e:

RADIOACTIVE EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION 3.22.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC 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 quarter to < 1.5 mrem to the total body and to s 5 mrem to aliy crgan.

b. During any calendar year to < 3 mrem to the total body and to 1 10 mrem to any organ.

APPLICABILITY: At all times ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the NRC Region I Administrator 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 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 commitment 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 regard to the requirements of 40 CFR 141, Safe Drinking Water Act.

BASES This specification is provided to implement the requirements of Sections II.A, III. A, and IV. A of Appendix I,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclides concentrations in the finished drinking water that are in excess of the requirements of 10 CFR 20. The dose calculations in the ODCM implement 1

l 3-107 l Amendment No. 72,129

the requirements in Section III.A. of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. l The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents will be consistent with the methodology provided in Regulatory Guide 1.109,

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

Revision 1, October,1977, and Regulatory Guide 1.113, " Estimating Aquatic Dispersion- of Effluents from Accidental and Routine 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.

i l

l 3-108 Amendment No. 72,129

a RADI0 ACTIVE EFFLUENTS 3.22.2 GASE0US EFFLUENTS 3.22.2.1 DOSE RATE LIMITING CONDITION FOR OPERATIONS 3.22.2.1 The dose rate due to radioactive materials released in gaseous effluencs from the site (see Figure 5-3) shall be limited to the following:

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

b. For I-131, I-133, tritium and all radionuclides in particulate form with half lives greater than 8 days: less than or equal to 1500 mrem /yr to any organ.

APPLICABILITY: At all times.

ACTION:

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

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

1. At least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

2. At least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

3. At least COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

BASES The specification is provided to ensure that the release rate at anytime at the site 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 a MEMBER OF THE PUBLIC in an unrestricted area, either within or outside the site boundary, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For MEMBERS OF THE PUBLIC who may at times be within the site boundary, the occupancy of the MEMBER OF THE PUBLIC will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the exclusion area boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the site boundary to less than or equal to 500 mrem / year to the total body or to less than or equal to 3000 l mrem / year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to an infant via the cow-milk-infant pathway to less than or equal to 1500 mrem / year for the nearest cow to the plant. '

3-111 Amendment No. 72

-o ~

7r 4

.RADI0 ACTIVE EFFLUENTSh ]

' GASEOUS EFFLUENTS. ]

' DOSE-NOBLE ~ GASES 1

<y LIMITING CONDITION FOR OPERATION r

3.22.2.2~ The' air. dose due to ' noble. gases released in gaseous effluents from the unit to areas at and beyond the site boundary (see Figure 5-3) ,

L shall be limited to the following: '

1 .a.- During any . calendar quarter: < 5 mrad for gamma radiation and < 10 mrad for beta radiation and,

b. During any calendar-year: < 10 mrad for gamma radiation and

< 20 mrad for beta radiation.  ;

APPLICABILITY: At all times.

' ACTION:

a.. With the calculated air dose from radioactive noble gases in gaseous' effluents exceeding any of the above limits, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report.which identifies the cause(s)- for. exceeding the .

limit (s) and defines the corrective actions that have been taken to ' reduce the releases and the proposed corrective

~

actions to be..taken to assure that subsequent releases will be.

in compliance with. the above limits.

BASES-

' This. specification applies to the release of radioactive' materials .in gaseous effluents from TMI-1.

This specification-is provided to implement the requirements of Section II.8, III. A and .IV. A of Appendix I,10.CFR Part 50. The limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low.as is reasonably achievable." The Survei' lance Requirements implement the requirements in Section III.A ~of Appendix I.that conformance with the guides of Appendix I be shown by calculational procedures based on models and data. such that the actual exposure of a MEMBER OF THE PUBLIC 'through the l

?? appropriate pathways--is unlikely to be substantially underestimated. The dose-

. calculation methodology and parameters established in the ODCM for calculating

.the' doses due to~ the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Release of Reactor 3-112 Amendment No. 72,129

RADI0 ACTIVE EFFLUENTS DOSE - 10 DINE-131,10 DINE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM LIMITING CONDITION FOR OPERATION 3.22.2.3 The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, l tritium, and all radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents released from the unit to areas at and beyond the site boundary (See Figure 5-3) shall be limited to the following:

a. During any calendar quarter: 5 7 5 mrem to any organ, and

b. During any calendar year: $ 15 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

With the calculated dose from the release of iodine-131, iodine-133, tritium, and radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

BASES This specification applies to the release of radioactive materials in gaseous effluents from TMI-1.

i This specification is provided to implement the requirements of Section II.C, III. A and IV. A of Appendix I,10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I. The ACTION statement provides the required operating flexibility and at the same time implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III. A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methodology and l parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of 3-113 Amendment 72,129

L RADI0 ACTIVE EFFLUENTS 3.22.2.4 GASEOUS RADWASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION ,

3.22.2.4 'The GASE0US RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the GASEOUS RADWASTE. TREATMENT SYSTEM shall be used to reduce radioactive materials in the gaseous waste prior. to their discharge when the monthly projected gaseous l effluent air doses due to untreated gaseous effluent releases from the unit (see Figure 5-3) would exceed 0.2 mrad for gamma radiation and 0.4 mrad for

-beta radiation. The appropriate portions.of the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the monthly projected doses due to gaseous effluent l releases from the site (see Figure 5-3) would exceed 0.3 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the GASEOUS 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, prepare and submit to the NRC Region I Administrator 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. A summary description of action (s) taken to prevent a recurrence.

BASES The use of the GASE0US RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that gaseous effluents are treated as appropriate 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 Critorion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous effluents.

3-115 Amendment No. 72

t RADI0 ACTIVE EFFLUENTS 3.22.2.6 WASTE GAS DECAY TANXS LIMITING CONDITION FOR OPERATION 3.22.2.6 The quantity of radioactivity contained in each waste gas decay tank shall be limited < 8800 curies noble gases (considered as Xe-133). l APPLICABILITY: At all times.

ACTION:

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

BASES Restricting the quantity of radioactivity contained in each waste gas decay l tank provides assurance that in the event of an uncontrolled release of the tanks contents, the resulting total body exposure to a MEMBER OF THE PUBLIC 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."

3-117 Amendment No. 72

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Q QN Q N 4

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T C 1 1 E ( A AAA .. A A (

H S N NNNW N WW N M C

s m p e e m r t y u e u s a P r s y l p

u s S r s e c s e s r l i i t se P o g D m apr r o e wmP w t L e r t d u o n l o w s eP g L os n p t o y nse n Cp o u i l S ol FCil G o i o n F i e Rde S eo t c

'n o tnh l v T vL a o M r o antli e O l s u m N e i iaol uL al r t r O n t t thBAB n Vee c e I i a- c iC w o vl A h T g r

wn e n ffro i l el n

T P d o t ItoooL t oL o I a ei o n t a r r i e R M et r oesscG l tGt a r C Fa P tmssaS o nSn r o S n t uuooeT s oTo T c E o yn k ArLLRO I COC n D i ce n t S I t

a nm i Ws W W P L eu S Fn . . . . F . . S p E

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4.21 RADI0 ACTIVE EFFLUENT INSTRUMENTATION

~4.21.1 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION.

. SURVEILLANCE REQUIREMENTS

'4.21.1. Each radioactive liquid effluent monitoring instruantation channel-shall be demonstrated OPERABLE by performance of.the CHANNEL CHECK, SOURCE

' CHECK; CHANNEL CALIBRATION, and CHANNEL TEST operations during the MODES and at- the frequencies shown in'. Table 4.21-1.

p 4-87 Amndnent No. 72

- _ _ _ _ _ _ - _ - _ _ _ _-______-_____________a

p-

?

a'

- '4.' 21.2 RADIOACTIVE GASEOUS. PROCESS AND EFFLUENT MONITORING INSTRUMENTATION

. SURVEILLANCE REQUIREMENTS

- 4. 21 .2 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 TEST operations at' the frequencies shown in Table' 4.21-2.

4 i

i 4-90  ;

Amendnent No. 72  !

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U AB a b c d e FT a b c R

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TABLE _4.21-2 (Continued) ~

TABLE NOTATION

• At all times.

~** During waste gas holdup system operation.

Operability is not required when discharges are positively controDed through the closure of WDG-V47, and RM-A8 and FT-151 are operable.

• • • • During Fuel Handling Building ESF Air Treatment System Operation.

1. - At all times during containment purging.

1. 1. At all times when condenser vacuum is established.

(1 ) The CHANNEL TEST shall also demonstrate that automatic isolation of this pathway for the Auxiliary and Fuel Handling Building Ventilation System, the supply ventilation is isolated 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)

(Includes circuit failure)

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

(2) The CHANNEL TEST shall also demonstrate that control room alann 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 down scale 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 standard.; certified by the National Bureau of Standards or using standards that have been obtained' from suppliers that participate in measurement assurance activities with NBS. These standards should permit calibrating the system over its intended '

range of energy and' measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration should be used.

(Operating plants may substitute previously established calibration procedures for this requirement.)

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

1. One volume percent nydrogen, balance nitrogen, and

2. Four volume percent hydrogen, balance nitrogen.

l 4-93 Amendment No. 72, .122

{

l TABLE 4.21-2 (Continued)

TABLE NOTATION l (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. 4 4-94 Amendment No. 72

[

4.22 RADI0 ACTIVE EFFLUENTS 4.22.1 LIQUID EFFLUENTS SURVEILLANCE REQUIREMENTS 4.22.1.1 CONCENTRATION

4. 22.1.1. A The radioactivity content of each batch of radioactive liquid waste shall be determined prior to release by sampling and analysis in accordance with Table 4.22-1. 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 ifmits of specification 3.22.1.1.

4. 22.1.1. B Post-release analysis of samples composited from batch releases shall be performed in accordance with Table 4.22-1. 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 3.22.1.1.

4.22.1.1.C The radioactivity concentration of liquids discharged from continuous release points shall be determined by collection and analysis of samples in accordance with Table 4.22-1. The results of the analysis shall be used with the calculational methods of the ODCM to assure that the concentration at the point of release is maintained within the limits of Specification 3.22.1.1.

t 4-95 l 1

Amendment No. 72

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Table 4.22-1 (Continued)

TABLE NOTATION

a. The LLD is defined, for purposes of this specification, as the smallest concentration of radioactive material in a sample that will yield a net count above system background that will be detected with 95% probability with 5%

probability of falsely concluding that a blank observation represents a "real" signal .

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

LLD =

E x V x 2.22 x 106 x Y x exp (- AAt)

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 disintegration),

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

2.22 x 106 is the number of disintegrations per minute per microcurie, Y is the fra:'.ional radiochemical yield (when applicable),

A is the radioactive decay constant for the particular radionuclides, and At is the elapsed time between midpoint of sample collection and time of counting. l Typical values of E, V, Y and At shall be used in the calcula'. ion.

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

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

4-98 Amendment No. 72

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 coaposite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.

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

Prior to sampling for analyses, each batch shall be isolated, and the thoroughly mixed, by a method described in the ODCM, to assure representative sampling.

e. A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g., from a volume or 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, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semi-Annual Radioactive Effluent Release Report pursuant to TS 6.9.4.

d l 4-99 Amendment No. 72 4 L________...__

'4.22.1.2 ' DOSE CALCULATIONS-Cumulative dose contributions from liquid effluents shall be determined in accordance with the Offsite Dose Calculation Manual (ODCM) at least once a. month.

4.22.1.3 LIQUID WASTE TREATMENT 4.22.1.3.1 Doses due to liquid releases shall be projected at least once a month, l in accordance with the ODCM. I 4.22.1.3.2 The liquid radwaste treatment system shall be demonstrated OPERABLE by operating the liquid radwaste treatment system equipment for at least 60 minutes quarterly unless the liquid radwaste system has been utilized to. process radioactive liquid effluents during the previous 92 days. j 4.22.1.4 LIQUID HOLDUP TANKS l

The quantity of radioactive material contained in each of the tanks specified in Specification 3.22.1.4 shall be determined to be within the limit by analyzing a representative sample of the tank's content weekly when radioactive materials are being added to.the' tank.  !

i 1

i l

l 4-100 Amendment No. 72 L______=_-___-- - _

I 1

)

l 4.22.2 GASEOUS EFFLUENTS i SURVEILLANCE REQUIREMENTS j 4.22.2.1 DOSE RATES I 1

4.22.E.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the limits of Specification 3.22.2.1.a in accordance with the methods and procedures of the ODCM.

1- 4.22.2.1.2 The dose rate of radioactive materials, other than noble gases, in l gaseous effluents shall be determined to be within the limits-of Specification 3.22.2.1.b in accordance with methods and procedures of

-f the ODCM by obtaining representative samples and performing analyses in l accordance with the sampling and analysis program, specified in Table 4.?2-2.

3 l

4 -101 i

Amendment No. 72

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Table 4.22-2 (Continued)

TABLE NOTATION

a. The LLD is defined, for purposes of this specification, as the smallest concentration of radioactive material in a sample that will yield a net count above system background that will be detected with 95% probability with 5%

probability of falsely corcluding that a blank observation represents a "real" signal .

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

4* S b

LLD =

E x V x 2.22 x 106 x Y x exp (- AAt)

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

I 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 disintegration),

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

2.22 x 106 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),

A is the radioactive decay constant for the particular radionuclides, and A t is the elapsed time between midpoint of sample collection and time of counting.

Typical values of E, V, Y, and A t shall be used in the calculation.

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

b. Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within one hour unless (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent  ;

activity has not increased by more than a factor of 3. '

c. Tritium grab samples from the spent fuel pool area shall be taken at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the refueling canal is flooded.

4-104 Amendment No. 72

l-Table 4.22-2 (Continued) l TABLE NOTATION

d. Charcoal cartridges and pardculate filters shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing (or after removal from sampler).

e. Tritium grab samples shall be taken weekly 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 3.22.2.1, 3.22.2,.2, and 3.22.2.3.-

g. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135 and Xe-138 for gaseous emmissions and Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semi-Annual Radioactive Effluent Release Report pursuant to TS 6.9.4.

h. Applicable only when condenser vacuum is established. Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within one hour unless (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3.

i. Gross Alpha, Sr-89, and Sr-90 analyses do not apply to the Fuel Handling Building ESF Air Treatment System.

j. If the Condenser Vent Stack Continuous Iodine Sampler is unavailable, then alternate sampling equipment will be placed in service within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, k.- Applicable only when condenser vacuum is established.

4-105 Amendment No.- 72, 122, 130

'4.22.2 GASEOUS EFFLUENTS SURVEILLANCE REQUIREMENTS 4.22.2.2 ' DOSE, NOBLE GAS Cumulative dose contributions from noble gas effluents for the current

. calendar quarter and current calendar year shall be determined in

.accordance with the OFFSITE DOSE CALCULATION MANUAL (0DCM) monthly.

4.22.2.3: . DOSE,10 DINE-131,10 DINE-133, TRITIUM, AND RADIONUCLIDES IN

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.Cu'nulative dose contributions from Iodine-131, Iodine-133, Tritium, and ' 't radionuclides in particulate fortn with half lives greater than 8 days l for the current calendar quarter and current calendar year shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM) monthly.

4.22.2.4 GASEOUS WASTE TREATMENT Doses due,to gaseous releases from the unit shall be projected monthly in accordance with the ODCM.

4.22.2.5 EXPLOSIVE GAS MIXTURE The concentrations of hydrogen and oxygen in the waste gas holdup system shall be determined to be within the limits of TS 3.22.2.5 by monitoring the waste gases in the Waste Gas Holdup System with the hydrogen and oxygen monitors covered in-Table 3.21-2 of Specification i 3.21 . 2. a 4.22.2.6 WASTE GAS ~ DECAY TANK The concentration of radioactivity contained in the vent header shall be determined weekly. If the concentration of the vent header exceeds 10.7 pCi/cc, daily samples shall be taken of each waste gas decay tank l being added to, to determine if the tank (s) is ! 8800 Ci/ tank.  ;

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4-106 l i

Amendment No. 72  ;

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( ~ l E4.22.3E SOLID RADIOACTIVE-WASTE -

a SURVEILLANCE REQUIREMENTS 4.22.3.1 SOLID RADWASTE SYSTEM-The l solid radwaste system shall be demonstrated OPERABLE quarterly by.

a. ' Operating.the solid radwaste system at least once in the previous  !

'92' days in accordance with the PROCESS CONTROL PROGRAM or,

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,b.: Verification of.the existence of a valid contract for .

SOLIDIFICATION to be performed by a Contractor in accordance with a PROCESS CONTROL. PROGRAM.

4.22.3.2. PROCESS CONTROL PROGRAM The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION

.of at'least one representative test specimen from at least every tenth

. batch of. each type of radioactive waste required.to be solidified by

r. the Process Control Program.

1 a.: If any ' test specimen fails ~ to verify. SOLIDIFICATION, the SOLIDIFICATION of the batch under test:shall be suspended until such time as.

. additional test- specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance-with the Process Control:

Program,' and.a subsequent test verif fes . solidification. Solidification.

of the batch-may then be resumed:using the alternative SOLIDIFICATION parameters ' determined by the PROCESS CONTROL PROGRAM..

b. -If the initial test specimen from a batch of waste-fails.to verify-SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative i.est specimens from each

' consecutive batch of the same type of wet waste until 3 cor.secutive

. initial test specimens demonstrate SOLIDIFICATION, The PROCESS CONTROL PROGRAM shall be, mndified'as required, to assure SOLIDIFICATION of subsequent batches of waste.

j f .y-4-107 Amendment No. 72

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i 4.22.'4 , TOTAL DOSE

SURVEILLANCE REQUIREMENT 4.22.4.1 DOSE CALCULATION'-

i Cumulative dose contributions from liquid and~ gaseous effluents shall be

' determined in.accordance with TS 4.22.1.2, 4.22.2.2 and 4.22.2.3 and in i

accordance with the ODCM. i p

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4-103 (Pages 4-109 to 4-116 intentionally blank)

' Amendment No. 72

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5.0 DESIGN FEATURES 5.1 SITE Applicability Applies to the location and extent of the exclusion boundary, restricted area, and low population zone.

Objective To define the above by location and distance description.

Specification 5.1.1 The Three Mile Island Nuclear Station Unit 1 is located in an area of low population density about ten miles southeast of Harrisburg, PA. It is in Londonderry Township of Dauphin County, Pennsylvania, about two and one-half miles north of the southern tip of Dauphin County, where Dauphin is coterminal with York and Lancaster Counties. The station is located on an island approximately three miles in length situated in the Susquehanna River upstream from York Haven Dam. Figure 2.1-3 of the updated TMI Unit 1 l FSAR is an extended plot plan of the site showing the plant orientation and immediate surroundings. The exclusion area as defined in 10 CFR 100.3, is a 2,000 ft. radius, including portions of Three Mile Island, the river surface around it, and a portion of Shelley Island, which is owned by Met Ed. The minimum distance of 2,000 ft, occurs on the shore of the mafiiland in a due easterly direction from the plant as shown on Figure 2.1-3 of the FSAR. Figure 1.1-1 of the FSAR is a plot plan showing the physical location of the fence which defines the " Restricted Area" l 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. T.S. l 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 population zone is two riles as shown on T.S. Figure 5-2. For discharge points for gaseous effluents, see T.S. Figure 5-3 and for liquid effluents, see T.S. Figure 5-4.

5-1 Amendment No. 72

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6.9.4 SEMIANNUAL ~RADI0 ACTIVE EFFLUENT RELEASE REPORT Note: A single submittal may be made for the station. The submittal should combine those sections that are common to both units at the station however, for units with separate radwaste systems, the submittal shall specify the release of radioactive material from each unit.

6.9.4.1 Routine ~ Radioactive Effluent Release Reports covering the operations of the unit during the prwfous 6 months c,f operation shall be l submitted within 60 days after January 1 and July 1 of each year.

6.9.4.2 The following information shall be included in both Radioactive Effluent Release Reports to be submitted each year:

6.9.4.2.1 The R'adioactive Effluent Release Reports shall include a sumary of l 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 Radioactive 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.

6.9.4.2.2 The Radioactive Effluent Release Reports shall include the following l information for each type of solid waste shipped offsite during the report period:

a. . container volume,  !

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

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

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

e. type of shipment (e.g., LSA, Type A, Type B) and

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

6.9.4.2.3 The Radioactive Effluent Release Reports shall include a summary of l unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents made during the reporting 1

period.

4 6.9.4.2.4 The Radioactive Effluent Release Reports shall include any changes  !

made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) '

and to the 0FFSITE DOSE CALCULATION MANUAL (0DCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.23.2.

6-18 -!

I Amendment No. 72, 77, 129 1 i

6.9.4.2.5 The Radioactive Effluent Release Reports shall include the instrumentation not returned to OPERABLE status within 30 days per TS 3.21.1.b and TS 3.22.2.b. ,

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6.9.4.3 The following information shall be included in the Radioactive i Effluent Release Report to be submitted 60 days after January 1 of each year.

6.9.4.3.1 The Radioactive Effluant Release Report to be submitted 60 days after l January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual l summary may be either in the form of an hour-by-hour listing of wind j speed, wind direction, atmosphere stability, and precipitation (if i measured) on magnetic tape, or in the form of joint frequency ,

distribution of wind speed, Wind direction, and atmospheric stability. I 6.9.4.3.2 The Radioactive Effluent Release Report to be submitted 60 days after l January 1 of each year shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released i from the unit or station during the previous calendar year. l l

6.9.4.3.3 The Radioactive Effluent Release Report to be submitted 60 days after l January 1 of each year shall include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the site boundary (Figures 5-3 and 5-4) 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 eccordance with the 0FFSITE DOSE CALCULATION MANUAL (0DCM).

l 6.9.4.3.4 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 likcly most exposed real individual from 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 contributions from Liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev.1.

6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:

a. Records of normal station operation including power levels and periods of operation at each power level.

6-19 Amendment No. 72, 77, 129

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I b.' Records-of principal maintenance activities, including

-inspection, repairs, substitution, or replacement of principal

. items of equipment important to safety. ,

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c. All REPORTABLE EVENTS.

d..  : Records of periodic' checks, tests and calibrations,

e. Records of reactor physics tests and other special tests important to safety.

f. Changes _ to operating' procedures important to safety.

g. Records of solid radioactive shipmentr.

e.

l 6-19a Amendment No. 72, 77, 129 1

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