ML20081D156

Proposed Change 7 to Radiological Effluent Tech Specs

ML20081D156
Person / Time
Site:	Cooper
Issue date:	03/07/1984
From:	NEBRASKA PUBLIC POWER DISTRICT
To:
Shared Package
ML20081D153	List: ML20081D156 ML20081D161 NLS8400083, Application for Amend to License DPR-46,consisting of Proposed Change 7 to Radiological Effluent Tech Specs.W/ Offsite Dose Assessment Manual for Assessment of Gaseous & Liquid Effluents. No Fee Required
References
TAC-08140, TAC-8140, NUDOCS 8403150171
Download: ML20081D156 (72)

Text

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i m Revised Technical Specifications for Radiological Effluents

, Revised Pages: 1 5 87 216u 216a3 j 111 Sa 216n 216v 216a4 i iv Sb 216o 216w 216a5 v 48 216p 216x 216a6 1 63 216q 216y 216a7 2 63a 216r 216z 216a8 3 78 216s 216a1 216a9 4 81 216t 216a2 216a10 i'

216all 216a19 216a26 231a 216a12 216a20 216a27 231b 216a13 216a21 216a28 231c 216a14 216a22 216a29 235 216a15 216a23 225a 235a 216a16 216a24 226 235b 216a17 216a25 231 235c 216a18

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8403150171 840307 PDR ADOCK 05000296 P. PDR _

1 RADIOLOGICAL TECHNICAL SPECIFICATIONS l e

TABLE OF CONTENTS Page No.

1.0 DEFINITIONS 1 - Sa LIMITING SAFETY SAFETY LIMITS SYSTEM SETTINGS 1.1 FUEL CLADDING INTEGRITY 2.1 6 - 22 4

1.2 REACTOR COOLANT SYSTEM INTEGRITY 2.2 23 - 26 SURVEILLANCE

[IMITINGCONDITIONSFOROPERATION REQUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 27 - 46 3.2 PROTECTIVE INSTRUMENTATION 4.2 47 - 92 A. Primary Containment Isolation Functions 47 B. Core and Containment Cooling Systems Initiation 47 and Control (CS, LPCI, HPCI, RCIC, ADS)

C. Control Rod Block Actuation 47 D. Radiation Monitoring Systems - Isolation and 48 Initiation Functions

1. Steam Jet Air Ejector Of f-Gas System 48

2. Reactor Building Isolation and Standby Gas 48 Treatment Initiation

3. Liquid Radwaste Discharge Isolation 48

4. Main Control Room Ventilation 48

5. Mechanical Vacuum Pump Isolation 49 E. Drywell Leak Detection 49 F. Primary Containment Surveillance Information 49 Readouts G. Recirculation Pump Trip 49 3.3 REACTIVITY CONTROL 4.3 93 - 106 A. Reactivity Limitations A 93 B. Control Rods B 94 C. Scram Insertion Times C 97 D. Reactivity Anomalies D 98 E. Recirculation Pumps E 98 F. Restrictions F 98 G. Scram Discharge Volume G 98 3.4 STANDBY LIQUID CONTROL SYSTEM 4.4 107 - 113 A. Normal Operation A 107 B. Operation with Inoperable Components B 108 C. Sodium Pentaborate Solution C 108 1

TABLE OF CONTENTS (Cont'd) e Page No.

SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REOUIREMENTS 3.12 ADDITIONAL SAFETY RELATED PLANT CAPABILITIES 4.12 215 - 215f A. Main Control Room Ventilation A 215 B. Reactor Building Closed Cooling Wate stem B 215b C. Service Water System C 215c D. Battery Room Vent D 215c 1

3.13 RIVER LEVEL 4.13 216 3.14 FIRE DETECTION SYSTEM 4.14 216b 3.15 FIRE SUPPRESSION WATER SYSTEM 4.15 216b b

3.16 SPRAY AND/OR SPRINKLER SYSTEM i (FIRE PROTECTION) 4.16 216e 3.17 CARBON DIOXIDE SYSTEM 4.17 216f 3.18 FIRE HOSE STATIONS 4.18 216g 3.19 FIRE BARRIER PENETRATION FIRE SEALS 4.19 216h 3.20 YARD FIRE HYDRANT AND HYDRANT HOSE HOUSE 4.20 2161 f

3 3.21 ENVIRONMENTAL / RADIOLOGICAL EFFLUENTS 4.21 216n A. Inst rumentation 216n

, B. Liquid Effluents 216x

C. Gaseous Effluents 216a4 D. Effluent Dose Liquid / Gaseous 216all E. Solid Radioactive Waste 216a12 F. Monitoring Program 216a13 C. Interlaboratory Comparison Program 216a20 5.0 MAJOR DESIGN FEATURES 5.1 Site Features 217 5.2 Reactor 217 l' 5.3 Reactor Vessel 5.4 Containment 217 217 5.5 Fuel Storage 218

, 5.6 Seismic Design- 218

! 7.7 Barge Traffic 218 6.0 ADMINISTRATIVE CONTROLS 6.1 Organization 219

, 6.1.1 Responsibility 219 6.1.2 Offsite 219 6.1.3 Plant Staff - Shift Complement 219 6.1.4 Plant Staff - Qualifications- 219a i.

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TABLE OF CONTENTS (Cont'd) 1 Page No.

SURVEILLANCE LIMITING CONDITIONS FOR OPERATICN REQUIREMENTS 6.2 Review and Audit 220 6.2.1.A Station Operations Review Committee (SORC) 220 A.1 Membership 220 A.2 Meeting Frequency 220 A.3 Quorum 220 A.4 Responsibilities 220 A.5 Authority 221 A.6 Records 221 A.7 Procedures 222 6.2.1.B NPPD Safety Review and Audit Board (SRAB) 222 B.1 Membership 223 B.2 Meeting Frequency 223 B.3 Quorum 223 B.4 Review 223 B.5 Authority 224 B.6 Records 225 B.7 Procedures 225 B.8 Audits 225 6.3 Procedures and Programs 226 6.3.1 Introduction 226 6.3.2 Procedures 226 6.3.3 Maintenance and Test Procedures 226 6.3.4 Radiation Control Procedures 226

.A High Radiation Areas 226a 6.3.5 Temporary Changes 226a 6.3.6 Exercise of Procedures 226a 6.3.7 Programs 226a

.A Systems Integrity Monitoring Program 226a

.B Iodine Monitoring Program 226a

.C Environmental Qualification Program 226a 6.4 Record Retention 228 6.4.1 5 year retention 228 6.4.2 Life retention 228 i

6.4.3 2 year retention 229 6.5 Station Reporting Requirements 230 l 6.5.1 Routine Reports 230

.A Introduction 230

.B Startup Report 230 l .C Annual Reports 230

.D Monthly Operating Report 231

.E Annual Radiological Environmental Report 231

.F Semiannual Radioactive Material Release Report 231a l

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Page No.

SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS 6.5.2 Reportable Occurrences 231c l

.A Prompt Notification with Written Followup 232

.B Thirty Day Written Reports 234 i

6.5.3 Unique Reporting Requirements 235 i 6.6 Environmental Qualification - 235b

! 6.7 Systems Integrity Monitoring Program 235b i 6.8 Iodine Mcnitoring Program 235b

.; 6.9 Process Control Program 235b 4

6.10 Offsite Dose Assessment Manual (ODAM) 235c 6.11 Major Changes to Radioactive Waste Treatment Systems 235c 1

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II0 DEFINITIONS i

The succeeding frequently used terms are explicitly defined so that a uniform l interpretation of the specifications may be achieved. l A. Thermal Parameters

1. Critical Power Ratio (CPR) - The critical power ratio is the ratic of 1 that assembly power which causes some point in the assembly to experience transition boiling to the assembly power at the reactor condition of interest as calculated by application of the GEXL correlation. (Reference NED0-10958)

2. Maximum Fraction of Limiting Power Density - The Maximum Fraction

, of Limiting Power Density (MFLPD) is the highest value existing in the core of the Fraction of Limiting Power Density (FLPD).

3. Minimum Critical Power Ratio (MCPR) - The minimum critical power ratio corresponding to the most limiting fuel assembly in the core.

4. Fraction of Limiting Power Density - The ratio of the linear heat generation rate (LHGR) existing at a given location to the design LHGR for that bundle type. Design LHCR's are 18.5 KW/ft for 7x7 bundles and 13.4 KW/ft for 8x8 bundles.

5. Transition Boiling - Transition boiling means the boiling regime between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.

B. Alteration of the Reactor Core - The act of moving any component in the region above the core support plate, below the upper grid and within the shroud. Normal control rod movement with the control rod drive hydraulic system is not defined as a core alteration. Normal movement of in-core instrumentation is not defined as a core alteration.

C. Cold Condition - Peactor coolant temperature equal to or less than 212 F.

D. Design Power - Design power means a steady-state power level of 2486 thermal megawatts. This is 104.4% of Rated Power (105% of rated steam flow) and the poner to which the safety analysis applies.

E. Engineered Safeguard - An engineered safeguard is a safety system the actions of which are essential to a safety action required to maintain the consequences of postulated accidents within acceptable limits.

E.A 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 dose if inhaled by an adult as the quantity and isotopic mixture of I-131, I-132, 1-133, I-134, ana I-135 actually present. The dose equivalent I-131 concentration is calculated by: equiv. I-131 = (I-131) + 0.0096 (I-132) + 0.18 (I-133) +

0.0025 (I-134) + 0.037 (I-135).

E.B Exhaust Ventilation Treatment System - An EXHAUST VENTILATION TREATMENT SYSTEM (EVTS) is a system intended to remove radioiodine or radioactive material in particulate form from gaseous effluent by passing exhaust ventilation air through charcoci absorbers and/or HEPA filters before exhausting the air to the environment. An EVTS is not intended to affect noble gas in gaseous effluent. Engineered Safety Feature (ESF) gaseous treatment systems are not considered to be EVTS. The Standby Gas Treatment System is an ESF and not an EVTS. EVTS are specifically identified la ODAM Figure 3-1.

F'. Functional Test - A functional test is the manual operation or initiation of

, a system, subsystem or component to verify that it functions within design tolerances (e.g. the manual start of a core spray pump to verify that it runs and that it pumps the required volume of water).

F.A Caseous Radwaste Treatnent System - A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by col-lecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior-to release to the environment.

C. Mot Standby Condition - Hot standby condition means operation with coolant temperature greater than 212*F, system pressure less than 1000 psig, and the mode switch in "Startup/ Hot Standby".

H. Immediate - Immediate means that the required action will be initiated as soon as practicable conside.ing the safe operation of the unit and the importance of the required action.

I. Instrumentation

1. Instrument Functional Test - Analog instrument functional test means the injection of a siculated signal into the instrument as close to the sen-sor as practical to verify the proper instrument channel response, alarm and/or initiating action. Bistable channels - the injection of a simu-lated signal into the sensor to verify OPERABILITY including alarm and/

or trip functions.

2. Instrument Calibration - An instrument calibration means the adjustment, as necessary, of an instrument signal output so that it corresponds, within acceptable range, and accuracy, to a known value(s) of the parameter which the instrument monitors. Calibration shall encompass the entire instru-ment including sensor, alarm /or trip functions and shall include the func-tional test. The calibration may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is cali-brated.

3. Instrument Channel - An instrument channel means an arrangement of a sen-sor and auxiliary equipment required to generate and transmit a signal related to the plant parameter monitored by that instrument channel.

4. Instrument Check - An instrument check is the qualitative determination of acceptable operability by observation of instrument behavior during operation. This determination shall include, where possible, comparison of the instrument with other independent instruments measuring the same variable.

5. Logic System Functional Test - A logic system functional test means a test of relays and contacts of a logic circuit from sensor to activated device to ensu're components are operable per design intent. Where practicable, action will go to completion; i.e., pumps will be started and valves operated.

6. Protective Action - An action initiated by the protection system when a limiting safety system setting is reached. A protective action can be at a channel or system level.

7. Protective Function - A system protective action which results from the protective action of the channels monitoring a particular plant condition.

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• 8. Simulated Automatic Actuation - Simulated automatic actuation means applying a simulated signal to the sensor to actuate the circuit in question.

8.A Source Check - A SOURCE CHECK shall be the qualitative assessment of chan-nel response when the channel sensor is exposed to a source of radioactivity.

9. Trip System - A trip system means an arrangement of instrument channel trip signals and auxiliary equipmant required to initiate action to accomplish a protective function. A trip system may require one or more inr:rument channel trip signals related to one or more plant parameters in order to initiate trip system action. Initiation of protective action may require the tripping of a single trip system-or the coincident tripping of two trip systems.

J. Limitine Conditions for Operation (LCO) - The limiting conditions for operation specify the minimum acceptable levels of system phrformance necessary to assure safe startup and operation of the facility. When these conditions are met, the plant can be operated safely and abnormal situations can be safely controlled.

Limiting Conditions for Operation (LCO) shall be applicable during the operational conditions specified for each spacifi-cation.

. Adherence to the requirements of the LCO within the specified time interval shall constitute compliance with the specification.

In the event the LCO is restored prior to expiration of the specified time interval, completion of the LCO action is not required.

In the event an LCO cannot be satisfied because of circumstances in excess of those addressed in the specification, the facility shall be placed in HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in 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 /> unless corrective measures are completed that permit operation under the LCO for the specified time interval as measured from initial discovery. Exception to these requirements shall be stated in the individual specifications.

Entry into an operational condition shall not be made unless the conditions of the LCO are met without reliance on the actions specified in the LCO unless otherwise excepted. This provision shall not prevent passage through operational conditions required to comply with the specified actions of an LCO.

K. Limiting Safety System Setting (LSSS) - The limiting safety system settings are settings on instrumentation which initiate the automatic protective action at a level such that the safety limits will not be exceeded. The region between the safety limit and these settings represent a' margin with normal operation lying below these settings. The margin has been established so that with proper operation of the instrumentation the safety limits will never be exceeded.

L'. Moda - The reactor modc is established by the mode selector-switch. The

, modes include refuel, run, shutdown and startup/ hot standby which are defined as follows:

1. Refuel Mode - The reactor is in the refuel mode when the mode switch is in the refuel mode position. When the mode switch is in the refuel position, the refueling interlocks are in service.

2. Run Mode - In this mode the reactor system pressure is at or above 825 psig and the reactor protection system is energized with APRM protection (excluding the 15% high flux trip) and RSM interlocks in service.

3. Shutdown Mode - The reactor is in the shutdown mode when the reactor mode switch is in the shutdown mode position.

4. Startup/ Hot Standby - In this mode the reactor protection scram trips initiated by the main steam line isolation valve closure are bypassed when reactor pressure is less than 1000 psig, the low pressure main steam line isolation valve closure trip is bypassed, the reactor protection system is energized with APRM (15% SCRAM) and IRM neutron monitoring system trips and control rod withdrawal interlocks in servica L.A Normal Ventilation - Normal ventilation is tha controlled process of discharging and replacing air from/to a confinement to maintain temperature, humidity, or other conditions necescary for p resonnel safety and entry. The contents of the atmosphere being discharged from the confinement will have been established prior to establishing normal ventilation following a purging / venting operation.

L.B Offsite Dose Assessment Manual (ODAM) - An OFFSITE DOSE ASSESSMENT MANUAL (ODAM) shall be a manual containing the methodology and parameters to be used ir the calculation of offsite doses due to radioactive gaseous and liquid effluents, calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints, and describes the Environmental Radiation Monitoring Program.

M. Operable - A system or component shall be considered operable when it is capable of performing its intended function in its required manner.

N. Operating -. Operating means that a system or component is performing its intended functions in its required manner.

O. Operating Cycle - Interval between the end of one refueling outage and the end of the next subsequent refueling outage.

P. Primary Containment Integrity - Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:

1. All manual containment isolation valves on lines connected to the reactor coolant system or containment which are not required to be open during accident conditions are closed.

2. At least one door in each airlock is closed and sealed.,

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3. All eutomatic containm:nt isolation valves are operable or de-activated )

in the isolated position.

4. All blind flanges and manways are closed.

P.A Purge - Purging - Purge or Purging is the controlled process of discharging air or gas from a confinement to establish temperature, pressure, humidity, concentra-tion or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

Q. Rated Power - Rated power refers to operation at a reactor power of 2381 megawatts thermal. This is also termed 100% power and is the maximum power level authorized by the operating license. Rated steam flow, rated coolant flow, rated neutron flux, and rated nuclear system pressure refer to the values of these parameters when the reactor is at rated power. Design power, the power to which the safety analysis applies, is 104.4% of rated power (105% of rated steam flow), which corresponds to 2486 megawatts thermal.

R. Reactor Power Operation - Reactor power operation is any operation with the mode switch in the "Startup/ Hot Standby" or "Run" position with the reactor critical and above 1% rated power.

S. Reactor Vessel Pressure - Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detectors.

T. Refueling Outage - Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the plant after that refueling.

U. Safety Limits - The safety limits are limits within which the reasonable maintenance of the fuel cladding integrity and the reactor coolant system integrity are assured. Violation of such a limit is cause for unit shut-down and review by the Nuclear Regulatory Commission before resumption of unit operation. Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory review.

V. Secondary Containment Integrity - Secondary containment integrity means that the reactor building is intact and the following conditions are met:

1. At least one door in each access opening is closed.

2. The standby gas treatment system is operable.

3. All automatic ventilation system isolation valves are operable or secured in the isolated position.

W. Shutdown - The reactor is in a shutdown condition when the mode switch is in the " Shutdown" or " Refuel" position.

1. Hot Shutdcwn means conditions as above with reactor coolant temperature greater than 212*F.

2. Cold Shutdown means conditions as above with reactor coolant temperature equal to or less than 212*F and the reactor vessel vented, t

l V. A Solidification - SOLIDIFICATION shall be the conversion of radioactive vastes from liquid systems to a solid which is as uniformally distributed as reasonably achievable with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).

X. Spiral Reload - Pertains to the spiral reloading of the core with fuel, at least 50% of which has previously accunulated a minimum exposure of 1000 MWD /T.

Y. Surveillance Frequency - Surveillance requirements shall be applicable during the operational conditions associated with individual LCO's unless otherwise stated in an individual Surveillance Requirement.

Each Surveillance Requirement shall be performed within the specified time interval with:

a. A maximum allowable extension not to exceed 25% of the surveillance interval.

b. A total maximum combined interval time for any 3 consecutive l surveillance intervals not to exceed 3.25 times the specified interval.

1 Performance of a Surveillance Requirement within the specified time i interval shall constitute compliance with operability requirements for an l LCO unless otherwise required by the specification.

Z. Surveillance Interval - The surveillance interval is the calendar time between surveillance tests, checks, calibrations and examinations to be performed upon an instrument or component when it is required to be operable. These tests may be waived when the instrument, component or system is not required to be operable, but the instrument, component or system shall be tested prior to being declared operable or as practicable following its return to service.

Z.A Venting - Venting is the controlled process of discharging air or gas from a confinement to establish temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not pro-vided or required during venting. Vent, used in system names, does not imply a venting process.

Z.B Offsite - Offsite means outside of the exclusion area as defined in 10CFR Part 100.3. The exclusion area boundary around Cooper Station is defined in Figure 1.1 and may also be referred to as the Site Boundary.

Z.C Member of the Public - A Member of the Public is a person who is not occupationally associated with NPPD and who does not normally frequent the Cooper Station. The category does not include contractors, contractor employees, vendors, or persons who enter the site to make deliveries, to service equipment, or work on the site.

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LIMITING CONDfTf0N FOR OPERATION SURVEILLANCE REQUIREMENTS 3.2 (cont'd.) 4.2 (cont'd.)

D. Radiation Monitoring Systems - D. Radiation Monitoring Systems -

Isolation & Initiation Functions Isolation & Initiation Functions

1. Steam Jet Air Ejector Of f-Gas System 1. Steam Jet Air Ejector Of f-Gas System

a. Operability of the Steam Jet Instrumentution surveillance require-Air Ejector Of f-Gas System ments are given on Table 4.2.D.

monitor is defined in Table 3.21.A.2.

b. The time delay setting for closure of the steam jet air ejector isolation valves shall not exceed 15 minutes,

c. Other limiting conditions for operation are given on Table 3.2.D and Specifications l 3.21.A.2 and 3.21.C.6.

2. Reactor Building Isolation and 2. Reactor Building Isolation and Standby Gas Treatment Initiation Standby Gas Treatment Initiation The limiting conditions for opera- Instrumentation surveillance require-tion are given on Table 3.2.D and ments are given on Table 4.2.D. -

Specification 3.21.A.2.

3. Liquid Radwaste Discharge , 3. Liquid Radwaste Discharge Isolation Isolation Instrumentation surveillance re-The limiting conditions for opera- quirements are given on Table l tion are given on Tabic 3.2.D and 4.2.D. l l Specification 3.21.B.

4. Main Control Room Ventilation 4. Main Control Room Ventilation Isolation Isolation The limiting conditions for opera- The instrument surveillance require-tion are given on Table 3.2.D and ments are given on Table 4.2.D.

l the Section entitled " Additional Safety Related Plant Capabilities."

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COOPER NUCLEAR STATION TABLE 3.2.D RADIATION MONITORING SYSTEMS THAT INITIATE AND/OR ISOLATE SYSTEMS Number of Sensor Instrument Setting Channels Provided Action System I. D. No. Limit by Design (1)

Sicam Jet Air Ejector Of f-Cas RMP-RM-150 A & B (3) 2 A System

-Reactor Building Isolation RMP-RM-452 A & B j:,100 mr/hr 2 B and Standby Cas. Treatment  :

. Initiation Liquid Radwaste Discharge RMP-RM-1 (2) 1 C Isolation Main Control Room Ventilation (RMV-RM-1) 4x10 CPM 1 D '

Isolation >

k Y Machanical Vacuum Pump Isolation (4) RMP-RM-251 A-D 3 times normal full power 4 E background. Alarm at 1.5 times normal full ,

pcwer background 4

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NOTES'FOR TABLE 3.2.D

1. Action required when component operability is not assured.

A. (1) If radiation level exceeds 1.0 ci/sec (prior to 30 min. delay line) for a period greater than 15 consecutive minutes, the off-gas iso-

lation valve shall close and reactor shutdown shall be initiated immediately and the reactor placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

A. (2) Refer to Specification 3.21.A.2. l B. Cease refueling operations, isolate secondary containment and start SBGT.

C. During release of radioactive wastes, the effluent control monitor shall be set to alarm and automatically close the waste discharge valve prior to exceeding the limits of Specification 3.21.B.1.

D. Refer to Section entitled " Additional Safety Related Plant Capa-bilities".

E. Refer to Section 3.2.d.5 and the requirements for Primary Contain-i ment Isolation on high main steam line radiation. Table 3.2.A.

2. Trip settings to correspond to Specification 3.21.B.1.

3. Trip settings to correspond to Specification 3.21.C.6.a. ,

4. Minimum number of channels operable shall be one during mechanical vacuum pump operation.

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I COOPER NUCLEAR STATION TABLE 4.2.D MINIMUM TEST AND CALIBRATION FREQUENCIES FOR RADIATION MONITORING SYSTEMS Instrument Instrument System I.D. No. Functional Test Freq. Calibration Freq. Check Instrument Channels Steam Jet Air Ejector Off-Cas System RMP-RM-150 A & B (i2) (12) (12)

Reactor Building Isolation and RMP-RM-452 A & B (12) (12) (12)

Standby Cas Treatment Initiation Liquid Radwaste Discharge Isolation RMP-RM-1 (11) (11) (11)

Main Control Room Ventilation RMV-RM-1 Once/ Month (1) Once/3 Months Once/ Day Isolation Mechanical Vacuum Eump Isolation RMP-RM-251, A-D See Tables y 4.1.1 & 4.1.2 Logic Systems 1

SJAE Off-Gas Isolation Once/ Year Standby Gas Treatment Initiation Once/6 Months Reactor Building Isolation Once/6 Months  !

Liquid Radwaste Disch. Isolation once/6 Months a

Main Control Room Vent Isolation Once/6 Montha Mechanical. Vacuum Pump Isolation Once/ Operating Cycle

2~ l SOTES FOR TABLES 4.2.A THROUGH 4.2.F l

l. Initiallyonceeverymonthuntilexpoeure (M as defined on Figure 4.1.1) is 2.0 X 10 ; thereafter, according to Figure 4.1.1 (after NRC approval). The compilation of instrument failure' rate data may include data obtained from other boiling water reactors for which the same design instrument

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operates in an environment similar to that of CNS.

2. Functional tests shall be performed before each startup with a required frequency not to exceed once per week.

3. This instrumentation is excepted from the functional test definition. The functional te'st will consist of applying simulated inputs. Local alarm lights representing upscale and downscale trips will be verified but no rod bluck will be produced at this time. Tho' inoperative trip will be initiated to produce a rod block (SRM and IRM inoperative also bypassed with the mode switch in RUN). The functions that cannot be verified to produce a rod block directly will be verified during the ope'r ating cycle.

4. Simulated automatic actuat3.on shall be performed once each operating cycle.

Where possible, all logic system functional tests will be performed using the test Jacks.

5. Reactor low. vater level, high drywell pressure and high radiation main steam line tunnel are not included on Table 4.2.A since they are tested on Table 4.1.2.

6. The logic system functional tests shall. include an actuation of time delay relays and timers necessary for proper functioning of the trip sy. stems.

7. These units are tested as part of the Core Spray System tests.

8. The flow bias comparator will be tested by putting one flow unit in " Test" (producing 1/2 scram) and adjusting the test input to obtain comparator rod block. The. flow bias upscale will be verified by observing a local upscale trip light during operation and verifying that it will produce a rod block during the operating cycle.

9. Performed during operating cycle. P.ortions of the logic is checked more frequently during functional tests'of the i functions that produce a rod block.

10. The detector will be inset 4 during each operating cycle and the proper amount of travel into the cot arified.

11. Surveillance requirements for this s m are defined in Table 4.21.A.I. l

12. Surveillance requirenents' for this spacem are defined in- Table 4.21. A.2. l l

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3'. 2 BASES (Cont'd)

Both instruments are required for trip but the instruments are so designed that any instrument failure gives a downscale trip. The trip setting of 1.0 ci/sec (prior to 30 min, delay) provides an improved capability to detect fuel pin cladding failures to allow prevention of serious degradation of fuel pin cladding integrity which might result from plant opsration with a misoriented or misloaded fuel assembly. This limit is more restrictise than 0.39 ci/sec noble gas release rate at the air ejectors (after 30 min, delay) which was used as the source term for an accident analysis of the augmented off-gas system. Using the .39 ci/sec source term, the maximum off-site to*.al body dose would be less than the .5 rem limit.

2. Reactor Building Isolation and Standby Cas Treatment Initiation Two radiation monitors are provided which initiate the Reactor Building Isolation function and operation of the standby gas treatment system. The trip is actuated by one hi-hi or two downscale indications.

Trip settings of <100 mr/hr for the monitors in the ventilation exhaust ducts are based upon initiating normal ventilation isolation and standby gas treat-ment system operation so that none of the activity released during the re-fueling accident lesves the Reactor Building via the normal ventilation path but rather all the activity is processed by the standby gas treatment system.

3. Liquid Radwaste Discharge Isolation The liquid radwaste monitor assures that cl1 liquid discharged to the discharge canal does not exceed the limits of Specification 3.21.B. Upon sensing a high l discharge level, an isolation signal is generated which closes the radwaste discharge valve. The set point is adjustable to compensate for variable isotopic discharges and dilution flow rates.

4. Main Control Room Ventilation The main control room ventilation isolation is provided by a detector monitoring the intake of the control room ventilation system. Automatic isolation of the normal supply and exhaust and the activation of the emergency filter system is provided by the radiation detector trip function at the predetermined trip level.

5. Mechanical Vacuum Pump The mechanical. vacuum pump isolation prevents the exhausting of radioactive gas -

thru the 1 minute holdup line upon receipt of a main steam line high radiation signal.

E. Drywell Leak Detection Flow transmitters are used to record the flow of liquid from the drywell sumps. An air sampling system is also provided to detect leakage inside the primary containment.

LIMITING CONDITYON FOR OPERATION SURVEILLANCE REQUIREMENTS 3.21 ENVIRONMENTAL / RADIOLOGICAL EFFLUENTS 4.21 ENVIRONMENTAL / RADIOLOGICAL EFFLUENTS A. Instrumentation A. Instrumentation

1. Liquid Effluent Monitoring 1. Liquid Effluent Monitoring Applicability: As shown in Table a. Each radioactive liquid effluent 3.21.A.I. monitoring instrumentation chan-nel shall be demonstrated OPER-Specification: ABLE by performance of the CHAN-NEL CHECK, SOURCE CHECK, CHANNEL

a. The radioactive liquid effluent CALIBRATION and CHANNEL FUNCTIONAL monitoring instrumentation chan- TEST operations during the modes nels shown in Table 3.21.A.1 shall and at the frequencies shown in be OPERABLE with their alarm and Table 4~.21.A.1.

trip setpoints set to ensure that the limits of 3.21.B.1 are not b. Radioactive liquid effluent moni-exceeded, tor alarm and trip setpoints shall be determined in the manner

b. With a radioactive liquid effluent described in the ODAM.

monitoring instrumentation channel alarm and trip setpoint less con-servative than required, reset without delay to meet Specifica-tion 3.21.A.1.a. suspend the release of radioactive liquid effluents monitored by the affected channel, declare the channel inoperable, or change the setpoint so it is acceptably conservative,

c. With less than the minimum required number of radioactive liquid effluent monitoring instrumentation channels operable, take the ACTION shown in Table 3.21.A.1.

d. If the minimum number of instrument channels is not returned to OPERABLE status within 31 days, in lieu of any other report, explain in the next Semiannual Radioactive Effluent Report why the instrument was not repaired in a timely manner.

e. The provisions of Definition J are not applicable. The reporting provisions of Specification 6.5.2 are not applicable.

l l

l

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o TABLE 3.21.A.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CilANNELS INSTRUMENT OPERABLE APPLICABILITY

• ACTION-

1. Gross Beta or Gamma Radioactivity Menitor Providing Automatic Isolation

a. Liquid Radwaste Effluent Line 1

• 18

2. Gross Beta or Camma Radioactivity Monitors Not Providing Automatic Isolation

a. Service Water Effluent Line 1

• 20 Z

3. Flow Rate Measurement Devices

a. Liquid Radwaste Effluent Line 1

• 21 i

NOTES FOR TABLE 3.21.A.1

• During releases via this pathway.

+ Channel (s) shall be OPERABLE and in service as indicated except that outages for maintenance and required tests, checks, or calibrations are permitted.

ACTION 18 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may be resumed, providing that prior to initiating a release:

1. At least two independent samples are analyzed in accordance with Specification 4.21.B.1.a and;

2. At least one technically qualified member of the Facility Staff independently verifies the release rate calculations and discharge valving which were determined by another qualified member.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 20 With the numbers of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that at least once every day a grab sample is collected and analyzed for gross radioacgivity (beta or gamma) at a lower limit of detection not greater than 10 mci /ml.

ACTION 21 With the number of channel.2 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 /> during actual releaser.

i

-216p-

~

TABLE 4.21.A.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT -

CHECK CHECK CALIBRATION TEST

1. Gross Beta or Camma Radioactivity Monitors Providing Alarm and Automatic Isolation

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

2. Cross Beta or Gamma Radioactivity Monitors Providing Alarm but not Providing Auto-matic Isolation

a. Service Water System Effluent Line D* M R(3) Q(2)

'h i

3. Flow-Rate Measurement Devices

a. Liquid Radwaste Effluent Line D (4 )

• NA R SA 4

. i NOTES FOR TABLE 4.21.A.1 s l

• During releases via this pathway.  !

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway occurs for Conditions 1 and 2 below and control room alarm annunciation occurs for Conditions 1, 2, and 3 below.

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

2. Circuit failure.

3. Instrument indicates a downscale failure.

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

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

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

(3) The CHANNEL CALIBRATION shall be performed according to established station calibration procedures.

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

FREQUENCY NOTATION:

S =

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

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

W =

At least once per 7 days.

M =

At least once per 31 days.

Q =

At least once per 92 days.

SA =

At least once per 184 days.

A = At least once per year.

R = At least once per 18 months.

S/U = Prior to each reactor startup.

P = Completed prior to each release.

NA = Not applicable.

-216r-

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.21.A (Cont'd) 4.21.A (Cont'd)

2. Gaseous Effluent Monitoring 2. Gaseous Effluent Monitoring Applicability: As shown in Table a. The setpoints shall be deter-3.21.A.2. mined in accordance with the method described in the ODAM.

Specification:

b. Each radioactive gaseous efflu-

a. The radioactive gaseous effluent ent monitoring instrumentation monitoring instrumentation chan- channel shall be demonstrated nels shown in Table 3.21.A.2 OPERABLE by performance of the shall be OPERABLE with their CHANNEL CHECK, SOURCE CHECK, alarm setpoints set to ensure CHANNEL CALIBRATION, and CHAN-that the limits of Specification NEL FUNCTIONAL TEST operations 3.21.C.1 are not exceeded. during the modes and at the frequencies shown in Table

b. With a radioactive gaseous ef- 4.21.A.2.

fluent monitoring instrumentation channel alarm setpoint less con ~

servative than a value which will ensure that the limits of 3.21.C.1 are met, reset without delay to comply with Specification 3.21.A.2.a. declare the channel inoperable; immediately suspend release; or change the setpoint l so it is acceptably conservative.

c. With less than the minimum re-quired number of radioactive gaseous effluent monitoring instrumentation channels operable, take the ACTION shown in Table 3.21.A.2.

d. If the minimum number of instru-ment channels are not returned to OPERABLE status within 31 days, in lieu of any other report, expla!n in the next Semiannual Radioactive Effluent Report shy the instrument was not repaired in a timely manner.

e. The provisions of Definition J are not applicable. The reporting provisions of Specification 6.5.2 are not applicable.

I w

-216s-I

TABLE 3.21.A.2 ~

RADIOACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CilANNELS INSTRUMENT OPERABLE APPLICABILITY # PARAMETER ACTION

1. Steam Jet Air Ejector

a. Noble Gas Activity Monitor 1 *** Noble Gas 25 Radioactivity Rate Measurement d

b. Effluent System Flow Rate Meaeuring Device 1

• System Flow Rate Measurement 26

2. Augmented Offgas Treatment System Explosive Gas Monitoring System

a. Hydrogen Monitor (2% monitor) 2 **  % llydrogen 28

3. Reactor Building Ventilation Monitor System i

a. Noble Gas Activity Monitor 1

• Radioactivity Rate Measurement 27

• 29

b. Iodine Sampler Cartridge 1 Verify Presence of Cartidge Particulate Sampler Filter 1

• Verify Presence of Filter 29 c.

d. Effluent System Flow Rate Measuring Device 1

• System Flow Rate Measurement 26 i

e. Sampler Flow Rate Measurement Device 1

• Sampler Flow Rate Measurement 26

4. (****)

a. Noble Gas Activity Monitor 1 Radioactivity Rate Measurement 27

b. Iodine Sampler Cartridge 1

• Verify Presence of Cartridge 29 Particulate Sampler Filter 1

• Verify Presence of Filter 29 c.

d. Effluent System Flow Rate Measuring Device 1

• System Flow Rate Measurement 26

e. Sampler Flow Rate Measuring Device 1

• Sampler Flow Rate Measurement 26

NOTES FOR TABLE 3.21.A.2

1. Channels shall be operable and in service as indicated except that outages are permitted for the purpose of required tests, checks, calibrations, or for maintenance.

• During releases via this pathway.

• • During Augmenteo Offgas Treatment System Operation.

• • • During operation of the Steam Jet Air Ejector.

• • • • Elevated Release Point (ERP) Monitoring Syrtem, Radwaste Building Ventilation Monitoring System, and Turbine Building Vantilation Monitoring System.

ACTION 25 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, gases from the main condenser offgas treat-ment system may be released to the environment for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided:

1. The offgas delay system is not bypassed; and

2. The main stack system noble gas activity monitor is OPERABLE:

Otherwise, be in at least HOT STANDBY within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-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 day and these a

samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 28 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of the augmented offgas treatment system may continue with one channel operable provided that the recombiner exhaust temperature is monitored. With only one of the preceeding methods operable, operation of the augmented offgas treatment system may continue provided gas samples are collected at least once per day and analyzed with-in the ensuing 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

ACTION 29 With the number of samplers OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided samples are continuously collected with auxiliary sampling equipment as required in Table 4.21.C.1.

-216u-

TABLE 4.21.A.2 '

R ADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST

1. Steam Jet Air Ejector

a. Noble Gas Activity Monitor D*** M R(3) 'Q(2) R(1)

b. Effluent System Flow Rate Measuring Device D* NA R Q

2. Augmented 0ffgas Treatment System Explosive Gas Monitoring System

a. Hydrogen Monitor (2% Monitor) D** NA Q(4) M

3. Reactor Building Ventilation Monitoring System

a. Noble Gas Activity Monitor (KAMAN) D* M R(3) Q(5)

'b. Iodine Sampler Cartridge W* NA NA NA

c. Particulate Sampler Filter W* NA NA NA f

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

e. Sampler Flow Rate Measuring Device D* NA R Q

f. Isolation Monitor (CE) D*- Q a(3) R(1)

4. (****)

a.. Noble Gas Activity Monitor (KAMAN) D* M R(3) Q(5)

b. Iodine Sampler W* NA NA NA

c. Particulate Sampler W* NA NA NA

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

e. Sampler Flow Rate Monitor D* NA R Q u

. NOTES FOR TABLE 4.21.A.2

• During releases via this pathway.

• • During augmented offgas treatment system opera.lon.

• • • During operation of the Steam Jet Air Ejector.

• • • • Elevated Release Point (ERP) Monitoring System, Radwaste Ventilation Monitoring System, and Turbine Building Ventilation Monitoring System (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciaticn occurs if any of the following conditions exists:

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

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

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

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

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

(3) The CHANNEL CALIBRATION shall be established in accordance with established station calibration procedures.

(4) The CHANNEL CALIBRATION shall include the use of a standard gas sample containing a percentage of hydrogen to verify accuracy of the monitoring channel in its operating range.

(5) Same as (2) except Parts 3 and 4 are deleted.

FREQUENCY NOTATION:

S = At least'once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

=

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

W =

At.least once per 7 days.

M =

At least once per 31 days.

Q = At least once per 92 days.

SA = At least once per 184 days.

A = At least once per year.

R = At least once per 18 months.

S/U = Prior to each reactor startup.

P = Completed prior to each release.

NA = Not applicable.

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LIMITING CONDITXON FOR OPERATION SURVEILLAMCE REOUIREMENTS 3.21 (Cont'd) 4.21 (Cont'd)

B. Liquid Effluents B. Liquid Effluents Applicability: At all times. 1. Concentration Specification: a. Radioactive liquid wastes shall be sampled and analyzed accord-

1. Concentration ing to Table 4.21.B.1.

a. The concentration of radioactive b. The analytical results shall material in water Offsite be used with methods in the (Figure 1.1) due to radioactive ODAM to verify that the average liquid effluent shall not exceed concentration beyond the site the concentration specified in boundary does not exceed 10 CFR Part 20.106 for radio- Specification 3.21.B.1.a.

nuclides other than dissolved when Sr-89, Sr-90 and Fe-55 or entrained noble gases. For concentrations are averaged dissolved or entrained noble over no more than 3 months gases, the concentgation shall and other radionuclide not exceed 2 x 10 pCi/ml concentrations are averaged total activity. over no more than 31 days,

b. With the concentration of radio-active material released Offsite exceeding the limit, attend to the cause without delay and restore the concentration within

, the above limit.

c. The provisions of Specification 6.5.2 do not apply.

I l

-216x-

TABLE 4.21.B.1 .

RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRA!!

Lower Limit Minimum of Detection' Sampling Analysis Type of Activity (LLD)

Liquid Release Type Frequency Frequency Analysis (pCi/ml)(1)

- (2)

1. Batch Waste Release Tanks (5) P P Principal Gamma 5 x 10 Each Batch Each Batch Emitters (7)(8) -6 I-131 1 x 10

-5 P M(9) Dissolved and I x 10 One Batch /M Entrained Cases (gamma emitters)

P M H-3 1 x 10-Each Batch Composite (3)(9) Gross Alpha 1 x 10-

-8 P Q(9) Sr-89, Sr-90 5 x 10 4 Each Batch Composite (3)(9) Fe-55 1 x 10 6 5

7 2.A. Plant Service Water W W(9) Principal Gamma 5 x 10- (2)

Effluent (6) Grab Sample Erait t ers (7) (8) 2.B. Plant Continuous Grab Sample W(9) Principal Gamma 5 x 10 -7(2)

Discharge (10) Composite (4) Emitters (7)(8)

I-131 1 x 10 -6

-5 M M(9) Dissolved and 1 x 10 Grab Sample Entrained Cases (gamma emitters)

Proportional (4) M(9) H-3 1 x 10 Composite (4) Gross Alpha 1 x 10 Proportional (4) Q(9) Sr-89, Sr-90 5 x 10--6 Composite (4) Fe-55 1 x 10

NOTES FOR TABLE 4.21.B.1 (1) The LLD is the smallest concentration of the radioactive material in a sample that will be detected with 95% probability (5% probability of falsely concluding that a i blank observation represents a "real" signal). ,

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

.66 s LLD

• b i E *.V.* 2.22

• Y

• exp (- Aat)

, Where:

1-4 LLD is the "a priori" lower limit of detection as defined above (as picocurie per unit mass or volume),

s is the standard deviation of the background counting rate or of the b

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 is.the number of transformations per minute per picocurie, 4

Y is the fractional radiochemical yield (when applicable),

A is the radioactive decay constant for the particular radionuclide, and i

At is the elapsed time between mfdpoint of sample collection and time of

); counting (for plant effluents, not environmental samples).

The value of s3used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance.. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the back-l ground shall include the typical contributions of other radionuclides normal?.y l present in the samples. Typical values of E,'V, Y, and at shall be used in 4

the calculation.

(2) For certain radionuclides with low gamma yield or low energies, or for certain radio-j nuclide mixtures, it'may not be possible to measure radionuclides-in concentrations near the LLD. Under these circumstances, the LLD may be ig reased inversely propor-tionally to the magnitude of the' gamma yield (i.e., 5 x 10 /I, where I is the photon abundance expressed ~as a decimal fraction), but in no case shall the LLD, as calcu-n lated in this manner for a specific radionuclide, be greater than 10% of'the MFC value specified.in 10 CFR 20, Appendix B. Table'II, Column 2.

(3) 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) To be representative of the quantities and concentrations of-radioactive materials in.

liquid ~ effluents, daily grab samples shall be collected in proportion-to the rate of m flow'of.the effluent stream. Prior.to analyses, all samples taken for the composite shall be' thoroughly mixed in order lfor the composite sample to be representative of :

.the effluent release.

[.

-216z-

l l

NOTES FPy TABLE 4.21.B.1 (Continued) l ~

l (5) A batch release is the discharge of liquid wastes of a discrete volume. Prior to l

sampling for analyses, each batch shall be isolated and then thoroughly mixed.

! (6) A grab sample of plant service water effluent shall be analyzed at least once each week in accordance with Table Itgm 2.A. In the event the radioactivity concentra-tion in a sample exceeds 3 x 10 uCi/ml, or in the event the plant service water effluengmonitorindicatesthepresenceofanactivityconcentrationgreaterthan 3 x 10 uCi/ml, sampling and analysis according to Table Item 2.B. shall commence and shall be performed as long as the condition persists.

(7) The principal gamma emitters for which the LLD specification will apply are exclu-sively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99 Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level. When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the Semiannual Radioactive Effluent Release Report.

(8) If an isotopic analysis is unavailable, batch releases may be made for up to 14 d ys provided the gross beta / gamma concentration to the unrestricted area is < 1 x 10 9 pc/ml and the sample is analyzed when the instrumentation is once again available.

(9) Analysis may be performed after release.

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

FREQUENCY NOTATION:

S = At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. A = At least once per year.

D = At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. R = At least once per 18 months.

W = At least once per 7 days. S/U = Prior to each reactor startup.

M = At least once per 31 days. P = Completed prior to each release.

Q = At least once per 92 days. NA = Not applicable.

SA = At least once per 184 days.

-216al-

L'IMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS

+

3.21.B (Cont'd) 4.21.B (Cont'd)

2. Liquid Dose 2. Liquid Dose

-a. The dose to a Member of the Public a. Dose Assessment - An assessment due to radioactive material in of compliance with Specification liquid effluents offsite (see 3.21.B.2.a shall be made in Figure 1.1) shall not exceed accordance with the Offsite Dose Assessment Manual (0 DAM) at least

~

1.5 mrem to the total body or 5 mrem to any body organ during once per 31 days.

any calendar quarter and not more than 3 mrem to the total body or 10 mrem to any body organ during any calendar year.

b. In the event Specification 3.21.B.2.a

b. In any quarter in which radio-is exceeded, prepare and submit to active liquid releases are made the Commission within 31 days after and the radwaste system is not the end of the quarter in which the operated, a projection of the limit was exceeded, pursuant to prospect of compliance with Specification 6.5.3, a Special Specification 3.21.B.2.a shall Report in lieu of any other report be made in accordance with the which identifies the cause(s) for ODAM.

exceeding the limit (s) and defines the corrective actions to be taken.

c. Appropriate parts of the system shall be used to reduce the concen-tration of radioactive materials in liquid wastes prior to their dis-charge when the pre-release analy-sis indicates a radioactivity con-centration, excluding tritium and a

noble gases, in excess of 0.01

' pCi/ml. -

j d. With radioactive liquid waste being discharged without treatment in ex-cess of the limit in Specification 3.21.B.2.c, prepare and submit to the Commission within 31 days after the end of the quarter in which the limit was exceeded, pursuant to Specification 6.5.3, a Special-Report in lieu of any other report which includes the following information:

s

1) Indentification of equipment or subsystems not OPERABLE and

~

the reason for nonoperability.

~

2) Action (s) taken to restore the-nonoperable equipment to OPER-ABLE status.

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

e. The provisions of' Definition J are not applicable.

-216a2-

UIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.21.B (Cont'd) 4.21.B (Cont'd)

3. Temporary, outside Storage 3. Temporary, outside Storage Tanks of Radioactive Liquid Tanks of Radioactive Liquid

a. In the event temporary, a. When radioactive liquid unprotected tanks are is being added to a used outside to store temporary, unprotected radioactive liquid, outside storage tank, the the contents of each liquid shall be sampled tank shall not exceed and analyzed for radio-10 curies, excluding activity at least once H-3 and dissolved per 7 days, noble gas,

b. If the quantity of radioactive material in a temporary, unprotected storage tank outside exceeds 10 curies, excluding H-3 and dissolved noble gas, immediately suspend addition of radioactive '

material and begin measures to reduce the content to 10 curies or less without delay and describe the events leading to the condition i in t:1e next Semiannual Radioactive Materials _

Release Report.

c. The provisions of Definition J are not applicable.

l l

l -216a3-L , . _ _ . . _ . _ _

LIMITZNG CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.21 (Cont'd) 4.21 (Cont'd)

C. Gaseous Effluents C. Gaseous Effluents Applicability: At all times.

Specification:

1. Concentration 1. Concentration

a. The dose rate Offsite due to a. The release rate of radioactive radioactive noble gases shall noble gas shall be monitored not exceed 500 mrem /yr to the according to Specification total body or 3000 mrem /yr to 3.21.A.2.

skin.

b. The dose rate Offsite due to H-3, b. A radioactive noble gas effluent I-131, I-133, and radioactive monitor shall be set to cause material in particulate form automatic alarm when the monitor having half-lives of 8 days or alarm setpoint, determined as more in gaseous effluent shall specified in the ODAM, is not exceed 1500 mrem /yr to any exceeded.

body organ when the dase rate due to H-3, Sr-89, Sr-90, and alpha emitting radionuclides is averaged over no more than 3 months and the dose rate due to other radionuclides is averaged over no more than 31 days.

c. In the event a limit in Speci-

c. An assessment of compliance with fication 3.21.C.1.a or b is Specification 3.21.C.1.b shall exceeded, decrease the release be made in accordance with the rate to comply with the limit. ODAM at least once every 31 days,

d. The provisions of Definition J are not applicable.

2. Noble Cases Dose 2. Noble Cases Dose

a. The air dose Offsite (see Figure a. Dose Assessment - An assessment 1.1) due to noble gases released of compliance with Specification in gaseous effluents shall not 3.21.C.2.a shall be made in exceed 5 mrad from gamma radi- accordance with the ODAM at ation and 10 mrad from beta least once every 31 days.

radiation during any calendar quarter. The air dose Offsite due to noble gases released in gaseous effluents shall not ex-ceed 10 mrad from gamma radi-ation and 20 mrad from beta radiation during any calendar year.

I

-216a4--

1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENTS 3.21.C (Cont'd) 4.21.C (Cont'd)

b. With the calculated air dose from l radioactive noble gases in gaseous effluents exceeding Specification 3.21.C.2.a. prepare and submit to the Commission within 31 days after i the end of the quarter in which '

the limit was exceeded, pursuant 3 to Specification 6.5.3 a Special Report in lieu of any other report which identifies the cause(s) and defines the corrective actions taken,

c. The provisions of Definition J are not applicable.

3. Iodine and Particulate 3. Iodine and Particulate

a. The dose to a Member of the Public a. Radioactive gaseous effluent due to I-131, I-133, and radio- other than noble gases shall active material in particulate be sampled and analyzed as form having a half-life greater specified in Table 4.21.C.1.

than 8 days in gaseous effluents Offsite (see Figure 1.1) shall not exceed 7.5 mrem to any organ during any calendar quarter and 15 mrem to any organ during any calendar year.

b. In the event Specification b. Dose Assessment - An assessment 3.21.C.3.a is exceeded, prepare of compliance with Specification and submit a Special Report to 3.21.C.3.a shall be performed the NRC within 31 days after the in accordance with the ODA!!

end of the quarter in which the at least once every 31 days.

specification was exceeded, pursuant to Specification 6.5.3.B and in lieu of any other report, which identifies the cause(s) and describes the corrective action taken.

c. The provisions of Definition J are not applicable.

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l LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENTS l

' l 3.21.C (Cont'd) 4.21.C (Cont'd)

4. Gaseous Releases 4. Gaseous Releases

a. Every reasonable effort shall be a. In any month in which radioactive made to operate at least one material in gaseous effluent is train of charcoal adsorbers in being released without treatment, the Offgas Treatment System a projection of the prospect of whenever the main condenser air compliance with Specification ejector is in operation except 3.21.C.4.b shall be made at during startup or shutdown with least once every 31 days in reactor power less than 10% of accordance with the ODAM.

rated or when system cannot function due to low offgas flow,

b. The Exhaust Ventilation Treatment b. Operation of the Offgas Treatment System (EVTS) shall be operated System charcoal adsorbers shall ,

to treat radioactive materials in be verified by using the gaseous effluent air when the projected effluent monitoring program in dose to a Member of the Public Specification 3.21.A.2.

due to the activity in air ef-fluent via the EVTS would exceed 0.3 mrem to any body organ,

c. In the event radioactive gas from the main condenser air ejector is discharged in effluent air for mere than 7 days without treat-ment by charcoal adsorbers or in the event air is discharged via an exhaust ventilation treatment system for more than 31 days without treatment and the limit of Specification 3.21.C.4.b is exceeded, prepare and submit a Special Report to the NRC, pursuant to Specification 6.5.3 and in lieu of any other report, which identifies the inoperable equipment and describes the corrective action taken.

d. The provisions of Definition J are not applicable.

-216a6-

TABLE 4.21.C.1 ,

RADIOACTIVE CASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum of Detection Sampling Analysis Type of Activity (LLD)

Gaseous Release Type Frequency Frequency Analysis (uC1/mi) (1)

A. Elevated Release Point M(3) M(3) Principal Gamma 1 x 10~ (2)

(ERP), Reactor-Bldg Vent, Grab Emitters (6)

Augmented Radwaste Sample Bldg Vent,

-6 Turbine Bldg Vent Q(9) Q H-3 1 x 10 (Gaseous) Grab Sample B. All Release Iypes as Continuous (5) W(4) 1131 1x10-f2

Listed in A Above Charcoal I-133 1 x 10 3 Sample Y Continuous (5) W(4) Principal Gamma 1 x 10~ (2)

Particulate Emitters (6)

Sample (I-131, Others)

~

Continuous (5) Q Sr89, Sr-90 1x10_ff Composite Gross Alpha 1 x 10 Particulate Sample (7)

-6 Continuous (5) Noble Gas Cross Noble Cases 1 x 10 Monitor Beta and Gamma (8) i

NOTES FOR TABLE 4.21.C.1 (1) The LLD is the smallest concentration of radioactive material in a sample that will

, be detected with 95% probability (5% probability of falsely concluding that a blank l

} observation represents a "real" signal).

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

g, 4.66 s b

E . V. 2.22 . Y. exp (-Aat)

Where:

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

s is the standard deviation of the background counting rate or of the count-b ing 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 is the number of transformations per minute per picocurie,  ;

i 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 s use in e ca cu a i n e ra ee n system b

shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples. Typical values of E, V, Y, and At shall be used in the calculation.

(2) For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possible to measure radionuclides in con-centrations near the LLD. Under these circumstances, the LLD may be incrgased inversely proportional to the magnitude of the gamma yield (i.e., 1 x 10 /I, where I is the photon abundance expressed as a decimal fraction), but in no case shall the LLD, as calculated in this manner for a specific radionuclide, be greater than 10% of the MPC value specified in 10 CFR 20, Appendix B, Table II, Column 1.

(3) Analyses shall also be performed following an increase as indicated by the gaseous release monitor of greater than 50% in the steady state release, after factoring out increases due to power changes or other operational occurrences, which could alter the mixture er radionuclides.

-216a8-

(.4) Analyses shall also be parformsd following an increase as indicated by the gaseous release monitor of greater than 50% in the steady state release, after factoring out increases due to power changes or other operational occurrences, which could alter the mixture of radionuclides. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less are analyzed, the corresponding LLD's may be increased by a factor of 10.

(5) 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 accor-

dance with Specifications 3.21.C.1, 3.21.C.2 and 3.21.C.3.

(6) The principal gamma emitters for which the LLD specification will apply are exclusitely the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99 Cs-134, Cs-137, Ce-141, and Ce-144 for particulate' emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide. When unusual circumstances cause LLD's higher than required for more than 31 days, the reasons shall be docume:ted in the Semiannual Radioactive Effluent Release Report.

(7) A quarterly composite particulate cample shall include a portion of each weeks particulate samples collected during the quarter.

i (8) The noble gas continuous monitor shall be calibrated using laboratory aralysis of the grab samples from A and B on Table 4.21.C.1 or using reference sources.

(9) A H-3 grab sample will also be taken when the reactor vessel head is removed.

This sample will be taken at the ERP or Reactor Building vent whichever will be representative dependent upon the head removal vaccum procedure.

FREQUENCY NOTATION:

S = At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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

W = At least once per 7 days.

M = At least once per 31 days.

= At least once per 92 days.

Q SA = At least once per 184 days.

A = At least once per year.

R = At least once per 18 months.

S/U = Prior to each reactor startup.

P = Completed prior to each release.

NA = Not applicable.

l I

1 I -216a9-A

i i

i LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS _

l 3.21.C (Cont'd) 4.21.C (Cont'd)

5. Hydrogen Concentration 5. Hydrogen Concentration

a. The concentration of hydrogen in a. The concentration of hydrogen the augmented offgas treatment sys- in the augmented offgas treat-tem downstream of the recombiners ment system downstream of the shall be limited to j:, 2% by volume, recombiners shall be determined

b. With the concentration of hydrogen ~by continuously monitoring the in the augmented offgas. treatment waste gases in the main condenser system downstream of thecrecombiners offgas treatment system with the exceeding the limit, restore the hydrogen monitors required OPER-concentration to within'the limit ABLE by Table 3.21.A.2.

within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

c. The provisions of Definition J are not applicable. The reporting provisions of Specification 6.5.2 are not applicable.

6. Air Ejector 6. Air Ejector

a. The gross radioactivity (beta and/ a. The gross radioactivity (beta and/

or gamma) rate of noble gases meas- or gamma) rate of noble gases ured at the main condenser air ejec- from the main condenser air ejec-tor shall be limited to < 1 Ci/see tor shall be determined at the at the air ejector. following frequencies by perform-

b. With the gross radioactivity (beta ing an isotopic analysis of a and/or gamma) rate of noble gases representative sample of gases at the main condenser air ejector taken at the discharge (prior to exceeding Specification 3.21.C.6.a. dilution and/or discharge) of the restore the gross radioactivity rate main condenser air ejector:

to within its limit within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY with-in the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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1 LIMITING CONDITION FOR OPERATION SURVETLLANCE REQUIREMENTS 3.21.C (Cont'd) 4.21.C.6 (Cont'd)  ;

7. Containment At least once per 31 days 1)

a. Whenever the primary containment is during normal operation.

vented / purged, it shall be vented / Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following an 2) purged through the Standby Gas Treat- increase, as indicated by the ment System, unless governed by Condenser Air Ejector Noble another Technical Specification, Gas Activity Monitor, of NRC directive, or District commitment greater than 50%, after fac-to the NRC. This specification does toring out increases due to not apply to Normal Ventilation.

changes in THERMAL POWER

b. The provisions of Definition J are level, in the nominal steady not applicable. The reporting pro- state fission gas release visions of Specification 6.5.2 from the primary coolant.

are not applicable.

D. Effluent Dose Liquid / Gaseous b. The radioactivity rate of noble Applicability: At all times. E"* * * "

the main condenser air ejector Specification: shall be monitored in accordance

1. The dose or dose commitment to a (ac- with Table 3.21.A.2.

tual) member of the public due to radiation and radioactive releases D. Effluent Dose Liquid / Gaseous from Cooper Station shall not exceed 75 mrem to his thyroid or 25 mrem 1. Dose Calculations - The cumu-to his total body or any other lative dose to a Member of the body organ during a calendar year. Public contributed by radioactive In the event the calculated dose material in' gaseous and liquid from radioactive material in liquid effluents shall be calculated at or gaseous effluents exceeds two times least once per year in accordance the limit of Specification 3.21.B.2.a, with the ODAM in order to verify 3.21.C.2.a. or 3.21.C.3.a. prepare and compliance with Specifi-submit a Special Report, in lieu of any cation 3.21.D.

other report, to the Commission pursuant to Specification 6.5.3 within 31 days which 1) defines actions to be taken to reduce releases and prevent recurrence and 2) results of an exposure analysis including effluent pathways and direct radiation to determine whether the dose or dose commitment to a member of the public due to radiation and radioactive releases from Cooper Station during the calendar year through the period covered by the calculation was less than limits stated in this Specification. If the estimated dose exceeds the limits stated herein, ar.d if the condition resulting in doses exceeding these limits has not already been corrected, submission of the Special Report shall be deemed a timely request for a variance in accord with provisions of 40 CFR Part 190, provided information specified in 40 CFR Part 190.11(b) is included. In that event, a variance 10 granted until NRC Staff action on the item is complete.

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENTS l 3.21.D (Cont'd) 4.21 (Cont'd)

2. The provisions of Definition J are j not applicable.

E. Solid Radioactive Waste E. Solid Radioactive Waste Applicability: During solid radwaste 1. Operating parameters and limits processing. for the solidification of radio-active waste were established dur-Specification: ing preparational testing of the system. Radioactive waste solid-

1. The appropriate equipment of the ification shall be performed in solid radwaste system shall be oper- accordance with established para-ated to process radioactive waste meters and limits. In addition, containing liquid and liquid des- every 10th batch of dewatered tined for disposal subject to 10 waste will be sampled prior to CFR Part 61 to a form that meets solidification and analyzed for applicable requirements of 10 CFR pH.

Part 61.56 before the waste is shipped from the site. 2. Each drum of solidified or dewatered radioactive waste will

2. Suspend delivery to a carrier for be inspected, prior to capping, co transport of any container of insure that there is no free vaste subject to Specification standing liquid on top of the 3.21.E.1 which does not comply solid waste.

with 10 CFR Part 61.56.

3. The Semiannual Radioactive Mate-rial Release Report in Specifi-cation 6.5.1.F shall include the following information for radio-active solid waste shipped off-site during the report peried:

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LTMITING CONDITION FOR OPERATION SURVEZLLANCE REOUIREMENTS 3.21 (Cont'd) 4.21 (Cont'd)

3. The provisions of Definition J are a. Container burial volume, not applicable,

b. Total curie quantity (determined

.by measurement or estimate),

c. Principal gamma radionuclides (determined by measurement or estimate),

d. Type of waste,

e. Solidification agent.

F. Monitoring Program F. Monitoring Program Applicability: At all times, 1. Radiological environmental samples 1 shall be collected and analyzed Specificacion: as specified in Table 3.21.F.1.

1. As a minimum the radiological envi- 2. A land use censue shall be con-ronmental monitoring program shall ducted annually and shall iden-be conducted as specified in Table tify the location of the nearest 3.21.F.1. Analytical techniques garden that is greater than 500 used shcIl be such that the detec- square feet in area and that tion capabilities in Table 3.21.F.2 yields edible leafy vegetables, are achieved. the location of the nearest milk

2. In the event the radiological en- animal, and the location of the vironmental monitoring program is nearest resident in each of the not conducted as specified in 16 meteorological sectors within Table 3.21.F.1, prepare and submit three miles of the Station. The to the Commission in the Annual land use census shall be conduc-Operating Report the reasons for ted at least once per 12 months.

not conducting the program in ac-cordance with Table 3.21.F.1 and 3. The results of sample analyses the plans for preventing a recur- performed shall be summarized rence. in the Annual Radiological

3. When the radioactivity in a sampled environmental medium, averaged over 4. The results of the land use cen-a calendar quarter, exceeds an ap- sus shall be included in the propriate value stated in Table Annual Radiological Environmental 6.5-2, prepare and submit to the Report.

Commission within 31 days from the end of the affected calendar quar-ter a Special Report in accordance with 6.5.3 which includes an evaluation of any release con-ditions, environmental factors or-other conditions which caused the value(s) of Table 6.5-2 to be ex-ceeded. If the radioactivity in  ;

environmental sample (s) is not at-tributable to release from the l

I

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.21.F (Cont'd) 4.21 (Cont'd)

Station, the Special Report is not required; instead the sample (s) result (s) shall be reported and explained in the Annual Radiologi-cal Environmental Report.

4. When environmental sampling medium is not available from a sampling location designated in Table 3.21.F.1, the cause and the loca-tion where replacement samples were obtained shall be reported ,

in the Annual Radiological Envi-ronmental Report.

5. In the event a location is identi-fled at which the calculated.per-sonal dose associated with one or more exposure pathways exceed 120%

of the calculated dose at the max-imum dose location associated with like pathways at a location where sampling is conducted as specified in Table 3.21.F.1, then the path-ways having maximum exposure poten-tial at the newly identified loca-tion will be added to the radiol-ogical monitoring program and to Table 3.21.F.1 at the next SRAB meeting if samples ~are reasonably attainable at the new location.

Like pathways monitored (sampled) at a location, excluding the control station location (s), having the lowest associated calculated per-sonal dose may be deleted fro,m Table 3.21.F.1 at the time the new pathway (3) and location are added.

6. A change in Table 3.21.F.1 shall be described in the Annual

, Radiological Environmental Report.

7. The provisions of Definition J are not applicable. -

~

l l

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TABLE 3.21.F.1 ~

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Sampling and Type and Frequency and/or Sample Sample Stations Collection Frequency of Analysis

1. Airborne

a. Radioiodine At least 5 locations Continuous operation of sampler with Radioidine canister: Ana-and Partic- sample collection as required by dust lyze at least once per 7 days ulate loading but at least once per 7 days. for I-131.

Particulate sample: Analyze for gross beta radioactivity

> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following filter changg. Perform gamma iso-topic analysis on each sample in which gross beta activity is >10 times the yearly mean of control samgles. Perform

, gamma isotopic analysis on y composite (by location) sam-p ple at least once per 92 days.

2. Direct Radi- At least 32 locations Thermoluminescent Dosimeters (TLD)U Camma dose: At least once per ation exchange and read-out at least once 92 days.

per 92 days.

3. Waterborne.

a River Water At least 2 locations Collect a one (1) gallon grab sample Gamma isotopic analysis of at least once per 31 days. each sample. Composite grab sample for tritium analysis at least once per 92 days.

b. Ground Water At least 2 locations Collect a one (1) gallon grab sample Gamma isotopic and tritium at least once per 92 days. analysis of each sample.

c. Sediment At least I location Two (2) times a year, once in the Gamma isotopic analysis of from Shore- spring and once in the fall. each sample.

line

TABLE 3.21.F.1 (CONTINUED)

• RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Sampling and Type and Frequency and/or Sample Sample Stations Collection Frequency of Analysis

4. Ingestion b

a. Milk At least one location At least once ger 15 days during Peak Gamma isotopic and I-131 (Nearest Pasture Period ; at least. once pcr analysis of each sample.

Producer) 31 days at other times.

b. Milk At least 2 locations At least once per 92 days. Gamma isotopic and I-131 (Other analysis of each sample.

Producers)

c. Fish At least'2 locations Two times per year (once in the Gamma isotopic analysis on summer and once in the fall), edible portions.

Attempt to include the following:

1. Bottom feeding species j

t 2. Middle-Top feeding species

d. Food Prod- At least 3 locations At time of harvest. Sample one of Gamma isotopic analysis on i ucts (Vege- the following classes of food products edible portion.

tables) at each location.

1. Flowers & f ruits"

2. Tubers"

3. Roots At least I location At time of harvest. One sample of I-131 analysis.

I broad-leaf

• vegetation, b

NOTES FOR TABLE 3.21.F.1

a. DELETED

b. Ge(Li) gamma isotopic analysis refers to high resolution Ge(Li) gamma spectrum analysis as follows: the sample is scanp2d for gamma-ray activity.

If no activity is found for a selected nuclide, the detection sensitivity for that nuclide will be calculated using the counting time, detector efficiency, gamma energy, geometry, and detector background appropriate to the particular sample ir. question. The following nineteen (19) nuclides shall be analyzed for routinely:

Be-7 Ru-103 Ce-144 K-40 Ru-106 Ra-226 Mn-54 I-131 Th-228 Fe-59 Cs-134 Co-58 Cs-137 Co-60 Bata-140 Zn-65 Ce-141 Zr-95 Nb-95 i

Any radionuclide detected, i.e., having a measured concentration greater than the LLD, vnether or not it is one of the 19 nuclides listed above, shall be regarded

as present in the sample,

c. Thermoluminescent Dosimeters (TLD) is a single phosphore. Two or more phosphores in one package are considered to be two or more dosimeters,

d. Peak Pasture Period is June 1 through September 30 of each year.

e. Vegetables are classified as follows:

- Flowers and fruits: Artichoke, broccoli, cauliflower, corn, cucumber; egg-plant, okra, pepper, pumpkin, squash, and tomato.

- Tubers: Potato.

l. - Roots: Beet, carrot, parsnip, radish, rutabaga, sweet potato, and turnip.

- Leaves (broad leaf): Cabbage, lettuce, spinach.

4 4

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TABLE 3.21.F.2

• DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS Lower Limit of Detection (LLD)*

Airborne Particulate Water or Gaa Fish Milk Food Products Sedfment Anaysis (pC1/1) (pCi/m3) (pci/kg, wat) (pci/l) (pci/kg, wet) (pci/kg, dry)

-2 gross beta 4 1 x 10 3 2000 H

54 15 130 Mn 597 , 30 260 58,60 15 130 Co h 65 Zn 30 260 g

U g 95 30-Zr e

95 15 Nb

-2 131 9 7 x 10 1 60 7

-2 134 15 5 x 10 130 15 60 150 Cs

-2 137 18 6 x 10 150 18 80 180 Cs 140 Ba 140g 15 15 Note: 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.

e

l NOTES FOR TABLE 3.21.F.2

a. The LLD is the "a priori" smallest concentration of radioactive material in a ,

sample that will be detected with 95% probability (5% probability of falsely concluding that a blank observation represents a "real" signal).

For a particular measurement system (which may include radio-chemical separation):

LLD = 4.66 S b E.V. 2.22 . Y . exp(-AAt)

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

S is the standard deviation of the background counting rate or of the b

counting rate of a blank sample as appropriate (as counts per minute)

I 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 traneformation per minute per picoeurie Y is the fractional radiochemical yield (when applicable)

A is the radioactive decay constant for the particular radionuclide At is the elapsed time between sample collection (or midpoint of the sample collection period) and time of counting The value of Sb used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unver-ified theoretically predicted variance. In calculating the LLD for a radionu-clide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radio-nuclides normally present in the samples (e.g., potassium-40 in milk samples).

Analyses shall be performed in such a manner that the stated LLD's will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLD's unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

b. LLD for drinking water.

-216a19-

EIMXTING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.21 (Cont'd) 4.21 (Cont'd)

G. Interlaboratory Comparison Program G. Interlaboratory Comparison Program Applicability: Applicable at all times 1. A brief summary of results ob-to Radiological Environmental Monitoring tained as part of the Interlab-Program, oratory Comparison Program shall be included in the Annual Specification: Radiological Environmer.tal Report, pursuant to Specification

1. Analyses shall be performed on 6.5.1.E.

radioactive materials supplied as part of an Interlaboratory Com-parison Program which has been ap-roved by the NRC.

2. With analyses not being performed ,

as required in Specification 3.21.G.1, report the corrective ac-tions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Report.

3. The provisions of Definition J are not applicable, i

i l

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3'.14-3.19/4.14-4.19 BASES 3.14/4.14 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This cap-ability is required in order to detect and locate fires in their early stages.

Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overall facility fire protection program.

In the event that a portion of the fire detection instrumentation is in-operable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is returned to service.

3.15-3.18/4.15-4.18 FIRE SUPPRESSION SYSTEMS THE OPERABILITY of the fire suppression systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in any portion of the facility where safety related equipment is located. The fire suppression system consists of the water system, spray and/or sprinklers, CO 2 and fire hose stations. The collective capability of the fire suppression systems is adequate to minimize potential damage to safety related equipment and is a major element in the facility fire protection program.

In the event that portions of the fire suppression systems are inoperable, alternate backup fire fighting equipment is required to be made available in the affected areas until the affected equipment can be restored to service.

In the event the fire suppression water system becomes inoperable, im-mediate corrective measures must be taken since this sytem provides the major fire suppression capability of the plant. The requirement for twenty-four hour report to the Commission provides for prompt evaluation of the acceptability of the corrective measures to provide adequate fire suppression capability for the continued protection of the nuclear plant.

3.19/4.19 FIRE BARRIER PENETRATION SEALS The functional integrity of the fire barrier penetration seals ensures that fires will be confined or adequately retarded from spreading to adjacent portions of the facility. This design feature minimizes the possibility of a single fire rapidly involving several areas of the facility prior to detection and extinguish-ment. The fire barrier penetration seals are a passive element in the facility fire protection program and are subject to periodic inspections, During periods of time when the seals are not functional, a continuous fire watch is required to be maintained in the vicinity of the affected seal until the seal is restored to functional status.

Fire barrier penetration seals include cable penetration barriers, fire doors, and fire dampers.

-216a21- )

3'.21 & 4.21 BASES 3.21.A & 4.21.A INSTRUMENTATION 3.21.A.1 & 4.21.A.1 Liquid Effluent Manitoring The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the release of radioactive material in liquid effluents. The OPERABILITY and use of these instruments implements the requirements of 10 CFR Part 50, Appendix A, General Design Criteria 60, 63, and 64. The alarm and/or trip setpoints for these instruments are calculated in the manner described in the ODAM to assure that the alarm and/or trip will occur before the limit specified in 10 CFR Part 20.106 is exceeded.

Control of the normal liquid discharge pathway is assured by station procedures governing locked discharge valves and valve line-up verification.

3.21.A.2 & 4.21.A.2 Gaseous Effluent Monitoring The radioactive gaseous eff: :ent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The location of this instrumentation is indicated by a Figure in the ODAM, a simplified flow diagram showing gaseous effluent treatment and monitoring equipment. The alarm / trip setpoints for these instruments shall l be calculated in accordance with methods in the ODAM, which have been reviewed by NRC, to ensure that the alarm will occur prior to exceeding the limits of 10 CFR Part 20.

The process monitoring instrumentation includes provisions for monitoring the concentra-tions of potentially explosive gas mixtures in the augmented offgas treatment system. The OPERABILITY and use of this instrumentation is consistent with the requirements of i General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

In the event no flow rate measurement device is operable on a gaseous stream, alternative 24-hour estimates are adequate since the system design is constant flow and loss of flow is alarmed in the control room.

3.21.B & 4.21.B LIOUID EFFLUENTS 3.21.B.1 & 4.21.B.1 Concentration This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20.106. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures within (1) the Section IV. A guides on technical specifications in Appendix I, 10 CFR Part 50, for 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 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.

Since service water is not a normal or expected source of significant radioactive release, routine sampling and_gonitoring for radioactivity is precautionary. An activity con-centration of 3 x 10 pCi/ml in service water effluent is diluted in the discharge canal to about 1.5% of the 10 CFR 20 Appendix B Table 2 Column 2 concentration with only one circulating water pump operating. During normal Station operation the dilution would be even greater. By monitoring service water effluent continuously for radio-activity and by confirmatory sampling weekly, reasonable assurance that its activity concentration can be kept to a small fraction of the 10 CFR Part 20.106 limit and within the Specification 3.21.B.2.a limit is provided.

By monitoring service water continuously and liquid radwaste continuously during dis-

! charge with the monitor set to alarm or trip before the limit specified in 10 CFR 20.106 is exceeded, reasonable assurance of compliance with Specification 3.21.B.1.2 is provided.

i Verification that radioactivity in liquid effluent averaged only a small fraction of the concentration limit is provided by calculations demonstrating compliance with Specifica-tion 3.21.B.2.a.

l

! -216a22-

3,21 & 4.21 BASES (Cont'd) 3.21.B & 4.21.B LIQUID EFFLUENTS (Cont'd) 3.21.B.1 & 4.21.B.1 Concentration (Cont'd)

Compliance with 10CFR Part 20.106 implies that the concentration limit represented by 10CFR Part 20, Appendix B, Table 2 will be met within a suitable and reasonable averaging time for assessing compliance. That averaging time is dependent upon the resolving time of the measurements or estimates which are used to evaluate compliance. Assessment of compliance is done by sampling and analysis according to Specification 4.21.B.1.2, by estimating or measuring the maximum release flow and the minimum dilution flow coincident during the period of release represented by tas sample, and by computing the concentration as a fraction of the limit in the unrestricted area periodically on the basis of these data.

3.21.B.2 & 4.21.B.2 Liquid Dose Specifications 3.21.B.2, 3.21.C.2 and 3.21.C.3 implement the requirements of 10 CFR Part 50.36a and of 10 CFR Part 50, Appendix I, Section IV. These specifications state limiting conditions for operation (LCO) to keep levels of radioactive materials in LWR effluents as low as is reasonably achievable. Compliance with these specifications will also keep average releases of radioactive material in effluents at small per-centages of the limits specified in 10 CFR Part 20.106. Surveillance Requirements provide for the measurement of releases and calculation of doses to verify compliance with the Specifications. Action statements in these Specifications implement the requirements of 10 CFR Part 50.36(c)(2) and 10 CFR Part 50, Appendix I, Section IV. A in the event an LCO is not met. Annual dose limitations stated in Specifications 3.21.B.2, 3.21.C.2, and 3.21.C.3 are not strict limits as used elsewhere in the Technical Specifications (are not aa immediate safety concern) but do obligate NPPD to take the applicable reporting action required in Specifications 3.21.B.2.b, 3.21.C.2.b. or 3.21.C.3.b.

10 CFR Part 50 contains two distinctly separate statements of requirements pertaining to effluents from nuclear power reactors. The first concerns a description of equipment to maintain control over radioactive materials in effluents, determination of design objectives, and means to be employed to keep radioactivity in effluents ALARA. This requirement is stated in Part 50, Section 34a and Appendix I,Section II. Appendix I,Section III stipulates that conformance with the guidance on design objectives be demonstrated by calculations (since demonstration is expected to be prospective). The other is a requirement for developing limiting conditions for operation in technical specifications. It is stated in 10 CFR Part 50, Section 36a and Appendix I,Section IV.

Both the intent of the Commission and the requirement are clearly stated in the Opinion of the Commission;7 relevant paragraphs from that document follow:

Section 50.36a(b) of 10 CFR Part 50 provides that licensees shall be guided by certain considerations in establishing and implementing operating procedures speci-fied in technical specifications which take into account the need for operating flexibility and at the same time ensure that the licensee will exert his best efforts to keep levels of radioactive materials in effluents as low as practicable.

The Appendix I that we adopt provides more specific guicance to licensees in this respect.

l

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

3.21 & 4.21 BASES (Cont'd) 3.21.B & 4.21.B LIQUID EFFLUENTS (Cont'd) l 3.21.B.2 & 4.21.B.2 Liquid Dose (Cont'd)

A. The Rule Section IV of Appendix I specifies action levels for the licensee. If, for any individual light water cooled nuclear power reactor, the quantity of radioactive material actually released in effluents to unrestricted areas during any calendar  !

quarter is such as to cause radiation exposure, calculated on the same basis as the design objective exposure, which would exceed one-half the annual design objective exposure, the licensee shall make an investigation to identify the causes of these high release rates, define and initiate a program of action to correct the situation, and report these actions to the Commission within 30 days of the end of the calendar quarter.  !

The conclusion of the NRC Staff in the Appendix I Rulemaking Hearing agrees with that of the Commission. The Staff recommended, "...that the limiting conditions for oper-ation described in Appendix I,Section IV be applicable upon publication to technical specifications included in any license authorizing operation of a light water cooled nuclear power reactor. . ." (p. 73) . (Cont'd)

The action to be taken by a licensee in the event a limiting condition is e3ceeded, i is stated in Appendix I,Section IV.A and in the Oninion of the Commission. Techni-cal Specifications 3.21.B.2, 4.21.B.2, 3.21.C.2, 4.21.C.2, 3.21.C.3 and 4.21.C.3 for Cooper Station conform to this requirement.

Guidance for developing technical specifications for surveillance and monitoring is included in Appendix I Section IV.B.

Although "it is expected that the annual releases of radioactive material in effluents

.from light water cooled nuclear power reactors can generally be maintained within the

! levels set forth as numerical guides for design objectives in Section II" (Appendix I, SectionI{),norecommendationwasmadebyeitgertheStaffinitsConcluding Statement or by the Commission in its Opinion that design objective values should

' appear as technical specification limits. The Opinicn of the Commission and the i

statement of Appandix-I are clear. Limiting conditions of operation (LCO) related to the quantity of radioactive material in effluents released to an unrestricted area stated in technical specifications-shall conform to Appendix-I,Section IV.A.

Licensee action in the event an LCO is exceeded should be in accord with Section IV.A.

Finally, surveillance and monitoring of. effluents and the environment should conform 4

to Section IV.B..

With the implementation of Specification 3.21.B.2 and 4.21.B.2 there is reasonable assurance that Station operation will not cause a,radionuclide concentration in public.

drinking water taken from the River that exceeds the. standard for anthropogenic-radioactivity in community drinking water.-

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3.21 & 4.21 BASES (Cont'd) 3.21.B & 4.21.B LIOUID EFFLUENTS (Cont'd) 3.21.B.2 & 4.21.B.2 Liquid Dose (Cont'd)

Specification 3.21.B.2.c implements the requirements of 10 CFR Part 50.36a(< t) that operating procedures be established and followed and that equipment be maintali i and used to keep releases to the environment as low as is reasonably achievable. 4 .e OPERABILITY of the liquid radwaste treatment system ensures that the appropriate portions will be available for use whenever liquid effluents require treatment prior to release to the environment. The specification that the portions of the system which were used to establish compliance with the design obj ectives in 10 CFR Part 50, Appendix I, Section II be used when specified provides reasonable assurance that releases of radioactive material in liquid effluent will be kept as low as is reason-ably achievable. The activity concentration, 0.01 pC1/ml, below which liquid rad-waste treatment would not be cost beneficial, and therefore not required, is demonstrated below:

Thequantityofradioactivematerialingiquideffluent released annually from Cooper Station has been calculated to be total iodines 3.65 curies total others (less H ) 0.7 total 4.35 curies (Coat'd)

The population dose commitment resulting from the radioactive material in liquid effluent released annually has been calculated to be thyroid 1.95 manrem total body _0_. 5 6 total 2.5 manrem Therefore, population doses are about 0.5 manrem per curie of 1 dine released and about 0.8 manrem per curie of other radionuclides (less H3 ) released in liquids. It would be conservative to assume one manrem committed per curie released in liquid effluent.

The volume of liquid waste processed and intended for discharge is estimated to be:

6 Low Purity Waste 5700 gal / day 1.8 x 10 gal /yr Chemical Waste +

Demin Regenerant Waste 6 4000 gal / day 1.2 x 10 gal /yr The annual costs to operate the radwaste processing equipment, neglecting credit for capital recovery, are estimated according to Regulatory Guide 1.110 to be:

Dirty Waste Ionex $ 88,000/yr Evaporator $114,000/yr Unit volume operating costs are about:

Cost to ion exchanger = $ 88,000 = $0.05/ gal 1.8E+6 gal Cost to evaporate = $114,000 = $0.10/ gal 1.2E+6 gal

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3.21 & 4.21 BASES (Cont'd) 3.21.B & 4.21.B LIQUID EFFLUENTS (Cont'd) 3.21.B.2 & 4.21.B.2 Liquid Dose (Cont'd)

Assuming the costbenefit balance is $1,000 expenditure per manrem reduction and assuming teatment removes all radioactivity from the liquid, then (1) the activity concentration in a batch below which treatment is not cost-beneficial is 6

$ 88,000 1 curie 10 uCi 1 manrem C = 1.8E+6 gal x 3785 ml manrem curie $1,000 gal C = 0.013 uCi/ml (Cont'd)

(2) the activity concentration below which evaporation is not cost-beneficial is 0

$114,000 x 1 curie x 10 uCi x 1 manrem C = 1.2E+6 gal x 3785 m1 manrem curie $1,000 gal C = 0.025 pCi/ml Therefore, to one significant digit, radwaste treatment of liquids containing less than 0.01 uCi/ml is not justified.

NRC Commissioners, " Opinion of the Commission," in the Appendix I Rulemaking Hearing, Docket Rm502, p. 101102, April 30, 1975.

NRC Staff, " Concluding Statement of the Regulatory Staff," in the Appendix I Rule-making Hearing, Docket RM502, pp. 17, 69, 73, 115, February, 1974.

NRC Commissioners, p. 101.

4 NRC Staff, op. cit.

5 NRC Commissioners, op. cit. l

~1 6 1 Demonstration of Compliance with 10 CFR 50 Appendix I, Revision 1 and Supplement 2, I Nebraska Public Power District, Cooper Nuclear Station, January 9, 1978.

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3.21 & 4.21 BASES (Cont'd) 3.21.C & 4.21.C CASE 0US EFFLUENTS 3.21.C.1 & 4.21.C.1 Concentration j Specification 3.21.C.1.a is included to assure that a measure of control is provided 1

over the concentration of radionuclides in air leaving the exclusion area. Radio-l active noble gases are monitored by instruments that provide a measure of release rate and cause automatic alarm when the noble gas concentration offsite is expected to exceed the dose rate specified in 3.21.C.1.a. With prompt action to reduce the radioactive noble gas concentration in effluent following alarm initiation, it can be maintained at a small fraction of the annual limit. The specified release rate limits

restrict the corresponding gamma and beta dose rates above background 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.

l Radiaiodines and radionuclides in particulate form are sampled with integrating samplers that permit assessment of the average release rate during each sample col-lection period. By complying with Specifications 3.21.C.2 and 3.21.C.3 the average offsite concentration will be maintained at a small fraction of the 10 CFR Part 20.106 concentration limit.

3.21.C.2 & 4.21.C.2 Noble Cases Assessments of dose required by Specifications 4.21.C.2 and 4.21.C.3 to verify com-pliance with Appendix I,Section IV is based on measured radioactivity in gaseous effluent and on calculational methods stated in the ODAM. Pathways of exposure and location of individuals are selected such that the dose to a nearby resident is un-likely to be underestimated. Dose assessment methodology described in the ODAM for gaseous effluent will be consistent with the methodology in Regulatory Guides 1.109 and 1.111. Cumulative and projected assessments of dose made during a quarter are based on historical average, or reference (the same period of record used in the design objective Appendix I evaluation) atmospheric conditions. Assessments made

, for the annual radiological environmental report will be based on quarterly and annual averages of atmospheric conditions during the period of release.

The bases for Specifications 3.21.C.2 and 4.21.C.2 are also discussed in the bases for Specifications 3.21.B.2 and 4.21.B.2.

3.21.C.3 & 4.21.C.3 Iodine and Particulates The bases for Specifications 3.21.C.3 and 4.21.C.3 are discussed in the bases for Specifications 3.21.B.2 and 4.21.B.2.

, -216a27-

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

3.21 & 4.21 BASES (Cont'd) 3.21.C & 4.21.C CASE 0US EFFLUENTS (Cont'd) 3.21.C.4 & 4.21.C.4 Gaseous Radwaste System The OPERABILITY of the gaseous radwaste treatment system and the ventilation exhaust treatment systems ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and design objective Section IID of Appendix I to 10 CFR Part 50.

The specified limits governing the use of appropriate portions of the systems are specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

3.21.C.5 & 4.21.C.5 Hydrogen Concentration This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treatment system is maintained below the flammability limits of hydrogen and oxygen. While the Augmented Treatment System is in service the hydrogen and oxygen concentrations are prevented from reaching the flammability limits. Maintaining the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3.21.C.6 & 4.21.C.6 Air Ejector Restricting the gross radioactivity rate of noble gases froc, the main condenser pro-vides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.

3,21.C.7 & 4.21.C.7 Containment This specification provides reasonable assurance that releases from drywell purging operations will not exceed the annual dose limits of 10 CFR Part 20 for unrestricted areas.

L 3.21.D & 4.21.D EFFLUENT DOSE LIOUID/ CASEOUS

This specification is provided to meet the reporting requirements of 40 CFR Part 190.

I

! -216a28-l

7.21 & 4.21 BASES (Cont'd) 3.21.D & 4.21.D EFFLUENT DOSE LIQUID / CASE 00S (Cont'd)

In the event an analysis is required to determine compliance with 40 CFR 190, the dose to a member of the public due to radiation direct from the station will be estimated with the aid of environmental TLD, PIC, or similar environmental radiation dosimetry.

A contribution from another fuel cycle facility is not added since there is no licensed fuel cycle facility within 50 miles of Cooper Station.

3.21.E & 4.21.E SOLID RADIOACTIVE WASTE The OPERABILITY of the solid radwaste system ensures that the system will be avail-able for use whenever solid radwastes require materials processing and packaging prior to being shipped offsite. This specification implements the requirements of 10 CFR Part 50.36a and General Design Criteria 60 of Appendix A to 10 CFR Part 50.

3.21.F & 4.21.F MONITORING PROGRAM The radiological envit;nmental monitoring program, including the land use census, is conducted to satisfy the requirements of 10 CFR Part 50, Appendix I, Section IV.B.2 and 3. The radiological monitoring program required by this specification provides measurements 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 measure-able concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environ-mental exposure pathways.

The environmental monitoring program described in Table 3.21.F.1 is the minimum pro-gram which will be maintained. The Offsite Dose Assessment Manual (ODAM) describes in detail the actual monitoring program which is performed to ensure compliance with the specified minimum program. Control of the radiological environmental monitoring program, including the ODAM, rests with the Environmental Affairs Division of Operations and not the Cooper Nuclear Station organization.

The land use census is conducted annually to identify changes in use of the unre-stricted area in order to recommend modifications in monitoring programs for evalu-ating individual doses from principcl exposure pathways.

The need to adjust the program to current conditions and to assure that the integrity of the program is maintained are thereby provided. 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 determir. this minimum garden size, the following assumptions were used, 1) that 20% of the garden was used for growing broad leaf vegetation (i.e. , similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/ square meter.

3.21.G & 4.21.G INTERLABORATORY COMPARISON PROGRAM The requirement for participation in a Interlaboratory Comparison Program is pro-vided to ensure that independent checks on the precision and accuracy of the meas-urements of radioactive material in environmental sample matrices are performed as

.part of a quality assurance program for environmental monitoring in order to demon-strate that the results are reasonably valid. Participation in an Interlaboratory Comparison Progran is contingent upon availability of samples supplied by the NRC or samples approved by the NRCr

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6.2 (cont'd)

a. Verification of compliance with internal rules, procedures (for example: normal, off-normal, emergency, operating, maintenance, surveillance, test, and radiation control procedures) and applicable license conditions at least once per 24 months.

b. The training, qualification, and performance of the operating staff at least once per 24 months,

c. The Emergency Plan and implementing procedures at least once per 12 months,

d. The Security Plan and implementing procedures at least once per 12 months.

e. The facility fire protection and its implementing procedures at least once per 24 months.

A fire protection and loss prevention inspection will be performed f.

utilizing either qualified off-site licensee personnel or an out-side fire protection consultant at least once per 12 months.

g. An inspection and audit by an outside qualified fire protection consultant shall be performed at least once per 36 months.

h. The Radiological Environmental Monitoring Program and the Offsite Dose Assessment Manual with their implementing procedures at least once every 24 months.

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l I

r 6.3 PROCEDURES AND PROGRAMS l .

l 6.3.1 Introduction

! Station personnel shall be provided detailed written procedures to be l used for operation and maintenance of system components and systems that could have an effect on nuclear safety.

6.3.2 Procedures Written procedures and instructions including applicable check off lists shall be provided and adhered to for the following:

A. Normal startup, operation, shutdown and fuel handling operations of the station including all systems and components involving nuclear safety.

B. Actions to be taken to correct specific and forseen potential or actual malfunctions of safety related systems or components including responses to alarms, primary system leaks and abnormal reactivity changes.

C. Emergency conditions involving possible or actual releases of radio-active materials.

D. Implementing procedures of the Security Plan and the Emergency Plan.

E. Implementing procedures for the fire protection program.

F. Administrative procedures for shift overtime.

G. Implementing procedures for the Offsite Dose Assessment Manual.

l 6.3.3 Maintenance and Test Procedures The following maintenance and test procedures will be provided to satisfy routine inspection, preventive maintenance programs, and operating license requirements.

A. Routine testing of Engineered Safeguards and equipment as required by the facility License and the Technical Specifications.

B. Routine testing of standby and redundant equipment.

C. Preventive or corrective maintenance of plant equipment and systems that could have an effect on nuclear safety.

D. Calibration and preventive maintenance of instrumentation that could affect the nuclear safety of the plant.

E. Special testing of equipment for proposed changes to operational procedures or proposed system design changes, o.3.4 Radiation Control Procedures Radiation control procedures shall be maintained.and made available to all station personnel. These procedures shall show permissible radiation exposure, and chall be consistent with the requirements of 10 CFR 20.

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l 6.7.1.C (Cont'd)

1. A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) re-ceiving exposures greater than 100 mrem /yr and their associated man rem exposure according to work and job functions, 1/ e.g. , reactor operations and surveillance, inservice inspection, routine maintenance, special main-tenance (describe maintenance), waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20%

of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

2. A summary description of facility changes, tests or experi-ments in accordance with the requirements of 10CFR50.59(b).

3. Pursuant to 3.8. A. a report of radioactive source leak testing. This report is requ'. red only if the tests reveal the presence of 0.005 microc 2 ries or more of removable contamination.

D. Monthly Operating Report Routine reports of operating statistics, shutdown experience, and a narrative summary of operating experience relating to safe operation of the facility, shall be submitted on a monthly basis to the individual designated in the current revision of Reg, i Guide 10.1 no later than the tenth of each month following the l calendar month covered by the report.

E. Annual Radiological Environmental Report

1. Routine radiological environmental reports covering the surveillance activities related to the Station operation during the previous calendar year shall be submitted to the NRC before May 1 of each year.

2. The Annual Radiological Environmental Report shall include the following:

a. A summary of doses to a Member of the Public Offsite due to Cooper Station aqueous and airborne radioactive effluents, calculated in accordance with methods compatible with the ODAM.

b. A summary of the results of the land use census required in Specification 4.21.F.2.

1/ This tabulation supplements the requirements of $20.407 of 10CFR Part 20.

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

6.5.1.E (Cont'd)

c. Summarized and tabulated results in the format of Table 6.5-1 of analyses of samples required by the radiological

- environmental monitoring program, and taken during the 2 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 a results. The missing data shall be submitted as soon as possible in a supplementary report.

d. A summary description of the radiological environmental monitoring program including any changes; a map of all sampl-ing locations keyed to a table giving distances and directions from the reactor; and the results of participation in the Inter-laboratory Comparison Program, required by Specification 3.21.G.

e. A summary of meteorological data collected'during the year

! shall either be included in the Annual Radiological Environmental -

Report or retained by NPPD and made available to the NRC upon

i. request.

F. Semiannual Radioactive Material Release Report

1. A report of radioactive materials released from the Station during

the preceding six months shall be submitted to the NRC within 60 days after January 1 and July 1 of each year *.

2. A Semiannual Radioactive Material Release Report shall include the following:

a

a. A summary by calendar quarter of the quantities of radioactive liquid and gaseous effluents released from the Station. The data should be reported in the format recommended in Regulatory Guide 1.21, Appendix B, Tables 1 and 2.

b. A summary of radioactive solid waste shipped from the Station,

including information named in Specification 4.21.E.3.

l

c. A summary of meteorlogical data collected during the year shall be included in the Semiannual Report submitted within 60 days after January 1 of each year.

d. A list and brief description of each unplanned release of.

, gaseous or liquid radioactive effluent that causes a limit in Specification 3.21.B.1.a. 3.21.B.2.a 3.21.C.1.a. 3.21.C.2.a, or 3.21.C.3.a to be exceeded.

• It should be noted that this data has.not normally been available to the District within-60 days and a verbal extension has typically been required from the NRC CNS Project

. Manager.

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c -- m g .-e+- h q mm. y  % v-?

TABLE 6.5-1 ,

ENVIRONMENTAL RADIOLOGICAL MONITORING PROGRAM

SUMMARY

Name of Facility Cooper Nuclear Station Docket No. 50-298 Location of Facility Nemaha, Nebraska Reporting Period (County, State)

Type & Lower Limit All Indicator Control Medium of Pathways Total No. of Locations Location with Highest Annual Mean Locations No. of Sampled of Analyses Detection (l) Mean[](2) Name Mean[](2) Mean[](2) Reportable (Unit of Measurement) Performed (LLD) Range (2) Distance & Direction Range (2) Range (2) Occurrences rk M

.q-Tabic Notes:

(1) Nominal Lower Limit of Detection (LLD) as defined in Definition K.A.

(2) Mean and Range based upon detectable measurements only. Fraction of detectable measurements at specified locations indicated in brackets [].

. .. - - - - -.=-- - . . . - , -.

i

. 6.5.2 Rsportable Occurrances Reportable occurrences, including corrective actions and measures to prevent reoccurrence, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a licensee ~

event report shall be completed and reference shall be made to the original report date.

d i

1 .

1 f

i 1

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6.5.3 Uniqua Rsporting Requiremtnts Reports shall be submitted to the Director, Nuclear Reactor Regulation, USNRC, Washington, D.C. 20555, as follows:

A. Reports on the following area shall be submitted as noted:

Area Reference Submittal Date

1. Secondary Containment 4.7.C.1 90 Days After Leak Rate Testing (1) Completion of Each Test.

I Note: (1) Each integrated leak rate test of the secondary containment shall be the subject of a summary technical report. This report should include data on the wind speed, wind direc-tion, outside and inside temperatures during the test, concurrent reactor building pressure, and emergency venti-lation flow rate. The report shall also include analyses and interpretations of those data which demonstrate com-pliance with the specified leak rate limits.

B. Special Reports Special reports (in lieu of Licensee Event Reports) may be required covering inspections, test and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.

Special reports shall be submitted to the NRC Regional Administrator within the time period specified for each report.

1. Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table 6.5-2 when averaged over any calendar quarter sampling period.

When more than one of the radionuclides in Table 6.5-2 are detected in the sampling medium, this report shall be submitted if:

Concentration (1) . Concentration (1)

(2) + * * * > 1.0 Limit Level Limit Level (2) -

When radionuclides other than these in Table 6.5-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications 3.21.B.2.a and 3.21.C.3.a. 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 RadiologictT Environmental Report.

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

TABLE 6.5-2 REPS,.lTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Reporting Levels Water Airborne Particulate Fish Milk Food Products Analysis (pCi/l) or Cases (pC1/m3) (pCi/Kg, Wet) (pC1/1) (pC1/Kg, Wet)

H-3 2E + 4(a) 3E + 4(c) i Mn-54 IE+3 3E + 4 Fe-59 4E + 2 IE + 4 Co-58 IE + 3 3E + 4 Co-60 3E + 2 IE + 4 Zn-65 3E + 2 2E + 4 i Zr-Nb-95 4E + 2(b)

- U$

E 'I-131 2 0.9 3 lE+2 Cs-134 30 10 lE + 3 60 IE + 3 Cs-137 50 20 2E + 3 70 2E + 3 Ba-La-140 2E + 2(b) 3E + 2(b)

T l (a) For drinking water samples. This is the 40 CFR 141 value.

(b) Concentration of parent or daughter.

(c) For. samples of water not used as a source of drinking water.

6.6 Envircnm;ntal Qualification A. By no later than June 30, 1982 all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of:

Division of Operating Reactors " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588 " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment", December 1979.

Copies of these documents are attached to Order for Modification of License DPR-46 dated October 24, 1980.

B. By no later than December 1, 1980, complete and auditible records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the D0R Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further quelified.

6.7 Systems Integrity Monitoring Program A program shall be established to reduce leakage from systems outside the primary containment that wo 21d or could contain highly radioactive fluids during a serious accident to as low as practical levels. This program shall include provisions establishing preventive maintenance and periodic visual inspection requirements, and leak testing requirements for each system at a frequency not to exceed refueling cycle intervals.

6.8 Iodine Monitoring Program A program shall be established to ensure that capability to accurately determine the airborne iodine concentration in vital creas under accident conditions. This program shall include training of personnel, procedures for monitoring and provisions for maintenance of sampling and analysis equipment.

6.9 Process Control Program (PCP)

The PCP shall be 3 manual detailing the program of sampling, analysis and formulation determination by which SOLIDIFICATION of radioactive waste from liquid systems is assured consistent with Specification 3.21.E and the surveillance requirements of these Technical Specifications.

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j 6.10 Offsite Dona Assessmtnt Manual (ODAM)

6.10.1 The ODAM shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid efflu-j ents and in the calculation of gaseous and liquid effluent monitoring l instrumentation alarm / trip setpoints consistent with the applicable LCO's contained in these Technical Specifications. The ODAM also describes the

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Environments 1 Radiation Monitoring Program.

I 6.10.2 District Initiated Changes A. Shall be submitted to the Commission by inclusion in the Semi-annual ,

Radioactive Material Release Report for the period in which the change (s) was made effective and shall contain:

! 1. Sufficiently detailed information to totally support the ration-1 ale for the change without benefit of additional or supplemental i

information. Information submitted should consist of a package j of those pages of the ODAM to be changed with each page numbered and provided with a signed approval and date box, together with appropriate analyses of evaluations justifying the change (s).

2. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations.

3. Documentation of the fact that the change has been reviewed and found acceptable by the SORC.

B. Shall become effective upon review and acceptance by the SORC.

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} 6.11 Major Changes To Radioactive Waste Treatment Systems (Liquid, Gaseous, and Solid) 6.11.1 The radioactive vaste treatment systems (liquid, gaseous, and_ solid) are those systems described in the facility Safety Analysis Report and amend-ments thereto, which are used to maintain that control over radioactive materials in gaseous and liquid effluents and in solid waste packaged for-offsite shipment required to meet the LCO's set forth in Specifications 3.21.B, 3.21.C. 3.21.D, and 3.21.E. The NRC is notified of major changes to these systems under the provisions of 10 CFR Part 50.59 and Part 50.71 (USAR revisions).

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ENVIRONMENTAL TECHNICAL SPECIFICATIONS APPENDIX B 10 OPERATING LICENSE NO. DPR-46 FOR THE COOPER NUCLEAR STATION NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 (All 84 pages of these Appendix B Technical Specifications have been deleted in their entirety by the generation of Radiological Environmental Technical Specifications (RETS) in Appendix A.)

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