Text

Proposed Radiological Effluent Tech Specs.
	ML19270G328
Person / Time
Site:	Cooper
Issue date:	06/06/1979
From:	; NEBRASKA PUBLIC POWER DISTRICT
To:	;
Shared Package
ML19270G327	List: ML19270G326; ML19270G328; ML19270G330; ML19270G333;
References
	TAC-08140, TAC-11959, TAC-8140, NUDOCS 7906060151
	Download: ML19270G328 (63)
	v • d • e

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4 . i 4 ATTACHMENT I PROPOSED RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS For Cooper Nuclear Station (Note: This cover sheet is not part of the proposed Technical Specifications) 2273 105 7 900 060 /[/ <-i

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RADIOLOGICAL TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Page No.

1.0 DEFINITI0!iS 1 - Sa l LIM 1 RING SAFETY SAFETY LIMITS SYSTEM SETTINGS 1.1 FUEL CLADDING INTEGRITY 2.1 6 - 22 1.2 REACTOR COOLANT SYSTEM INTEGRITY 2.2 23 - 26 SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 27 - 46 3.2 PROTECTIVE INSTRUMENTATION 4.2 47 - 92 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 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 3.5 CORE AND CONTAINMENT COOLING SYSTEMS 4.5 114 - 131 A. Core Spray and LPCI Subsystems A 114 B. Containment Cooling Subsystem (RHR Service Water) B 116 C. HPCI Subsystem C 117 D. RCIC Subsystem D 118 E. Autsmatic Depressurization Systen E 119 F. Minimum Low Pressure Cooling System Diesel F 120 Generator Availability G. Maintenance of Filled Discharge Pipe G 122 1:. Engineered Safeguards Compartments Cooling H 123 3.6 PRIMARY SYSTEM BOUNDARY 4.6 132 - 158 A. Thermal and Pressurization Limitations A 132 22L73 106

.4 Jg b TABLE OF CONTENTS (Cont'd)

SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS Page No.

3.14 Fire Detection System 4.14 216b 3.15 Fire Suppression Water System 4.15 216b 3.16 Spray and/or Sprinkler System (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 Environmental / Radiological Effluents 4.20 216m A. Instrumentation 216m B. Liquid Effluents 216w C. Gaseous Effluents 216a3 D. Ef fluent Dose Liquid / Gaseous 216a8 E. Solid Radioactive Waste 216a9 F. Monitoring Program 216a10 G. Interlaboratory Comparison Program 216all 5.0 MAJOR DESIGN FEATURES 217 - 218 5.1 Site Features 217 5.2 Reactor 217 5.3 Reactor Vessel 217 5.4 Containment 217 5.5 Fuel Storage 2.o 5.6 Seismic Design 218 5.7 Barge Traffic 218 6.0 ADMINISTRATIVE CONTROLS 219 - 237 6.1 Organization 219 6.2 Review and Audit 220 6.2.1.A Station Operations Review Committee 220

1. Membership 220

2. Meeting Frequency 220

3. Quorum 220

4. Responsibilities 220

5. Authority 2273 07 221

6. Records 221

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TABLE OF CONTENTS (Cont'd)

Page No.

6.2.1.B NPPD Safety Review and Audit Board 222

1. Membership 223

2. Meeting Frequency 223

3. Quorum 223

4. Responsibilities 223

5. Authority 224

6. Records 225

7. Procedures 225

8. Fire Inspection 225 6.3 Station Operating Procedures 226 6.4 Actions to be taken in the Event of Occurrences 227 Specified in Section 6.7.2.A 6.5 Actions to be taken if a Safety Limit is Exceeded 227 6.6 Station Operating Records 228 6.7 Station Reporting Requirements 230

1. Routine Reports 230 A. Requirements 230 B. Startup Report 230 C. Annual Reports 230 D. Monthly Operating Report 231 E. Annual Radiological Environmental Report 231a F. Semiannual Rad'oactive Material Release 231c Report

2. Reportable Occurrences 231c A. Prompt Notification with Written Followup 232 B. Thirty Day Written Reports 234

3. Unique Reporting Requirements 235 A. Testing Reports 235 B. Special Reports 235 6.8 Radiological Environmental Monitoring Manual (REMM) 235b 6.9 Offsite Dose Assessment Manual (ODAM) 235b 6.10 Major Changes to Radioactive Waste Treatment Systems 235c 2273 108

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. s 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 toler-ances (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 Gaseous Radwaste Treatment 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.

G. Hot Standby Condition - Hot standby condition means operation with coolant tem-perature 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 considering 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 simulated 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 the 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 oper-ation. This determination shall include, where possible, comparison of the instrument with other independent instruments measuring the same para-meter.

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 ensure components are operable per design intent. Where practicable, action will go to completion; i.e., pumps will be started and valves operated.

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

8. Simulated Automatic Actuation - Simulated automatic actuation means apply-ing 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 radioactive source.

9. Trip System - A trip system means an arrangement of instrument channel trip siguais and auxiliary equipment required to initiate action to accomplish a protective function. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in ordce 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. Limiting Conditions for Operation (LCO) - Th0 11miting conditions for operation specify the minimum acceptable levels of system performances 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 opers-tional conditions specified for each specification.

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 feeflity shall be placed in HOT SHUT-DOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months 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 otFatvise excepted. This provision shall not prevent passage through opera-tional 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 eyeeeded. 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.

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K.A Lower Limit of Detection (LLD) - The LLD is the smallest concentration of radio-active material in a sample that will be detected with 95% probability with 5%

probability of falsely concluding that a blank observation represents a "real" signal. Methods for determining the LLD are contained in the Of fsite Dose Assessment Manual (ODAM).

L. Mode - The reactor mode 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 850 psig and the reactor protection system is energized with APRM protection (exclud-ing 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 if 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 service.

L.A Offsite Dose Assessment Manual (ODAM) - An OFFSITE DOSE ASSESSMENT MANUAL (ODAM) shall be a manual containing the methodology and parameters to be used in the calculation of of f site doses due to radioactive gaseous and liquid ef fluents and in the calculation of gaseous and liquid ef fluent monitoring instrumentation alarm / trip setpoints.

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

l 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 - Primacy 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 automa t ic containment Isolation valves are operable or deactivated in the isolated position.

4. All blind flanges and manways are closed.

P.A Radiological Environmental Monitoring Manual (RutM) - A RADIOLOGICAL ENVIRON-

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MENTAL MONITORING MANUAL (REMM) shall be a manual describing the Radiological Environmen al Monitoring Program. Control of the REMM shall rest with the Environmental Af fairs Division of the Power Operations Group and not the Cooper Nuclear Station organization.

Q. Rated Power - Rated power refers to operation at a reactor power of 2381 mega-wa tts thermal. This is also termed 100% power and is the maximun 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 105% of rated power, which corresponds to 2500 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 af ter that refueling.

U. Safety Limits - The safety limits are limits within which the reasonable main-tenance of the fuel cladding integrity and the reactor coolant system integrity are assured. Violation of such a limit is cause for unit shutdown 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" position.

1. Hot Shutdown means conditions as above with reactor coolant temperature greater than 2120F.

2. Cold Shutdown means conditions as above with reactor coolant temperature equal to or less than 2120F and the reactor vessel vented.

2273 112 W.A Solidification - SOLIDIFICATION shall be the conversion of radioactive wastes from liquid systems to a solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).

X. 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:

1. A maximum allowable extension not to exceed 25% of the surveillance inter-val.

2. A total maximum combined interval time for any 3 consecutive surveillance intervals not to exceed 3,25 times the specified interval.

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

Y. 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 oper-able, but the instrument, component or system shall be tested prior to being declared operable or as practicable following its return to service.

Z. Ventilation Exhaust Treatment System - A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal absorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.2.C (Cont'd) 4.2.C D. Radiation Monitoring Systems - D. Radiation Monitoring Systems -

Isolation & Initiation Functions Isolation & Initiation Functions l 1. Main Condenser Air Ejector 1. Main Condenser Air Ejector l Off-Gas System Off-Gas System

a. Operability of the Main Condenser Instrumentation surveillance re-Air Ejector Off-Gas System is de- quirements are given on Table fined in Table 3.20.A.2. 4.20.A.2.

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

c. Other limiting conditions for operation are given on Table 3.2.D and Specifications 3.20.A.2 and 3.20.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 re-tion are given on Table 3.2.D and quirements are given on Table Specification 3.20.A.2. 4.2.D.

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3. Liquid Radwaste Discharge 3. Liquid Radwaste Discharge Isolation Isolation The limiting conditions for opera- Instrumentation surveillance re-tion are given on Table 3.2.D and quirements are given on Table Specification 3.20.B. 4.2.D and Table 4.20.A.1. l l

4. Main Control Room Ventilation 4. Main Control Room Ventilation Isolation Isolation The limiting conditions for opera- The instrument surveillance re-tion are given on Table 3.2.D and quirements are given on Table the Section entitled "Additiona! 4.2.D.

Safety Related Plant Capabilities."

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

Main Condenser Air Ejector Off-Gas RMP-RM-150 A & B (3) 2 A System Reactor Building Isolation RMP-RM-452 A & B < 100 mr/hr 2 B and Standby Gas Treatment Initiation Liquid Radwaste Discharge RMP-RM-2 (2) 1 C Isolation Main Control Room Ventilation (RMV-RM-1) 4 x 103 CPM 1 D

& Isolation Y

Mechanical Vacuum Pump Isolation RMP-RM-251 A-D 3 times normal 4 E full power back-ground. Alarm at 1.5 times normal full power back-ground.

N N

N U

M

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

A. (2) Refer to Specification 3.20.A.2.

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 3.20.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-ment Isolation on high main steam line radiation. Table 3.2.A.

2. Trip settings to correspond to Specification 3.20.B.1.

3. Trip settings to correspond to Specification 3.20.C.6.a.

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COOPER NUCLEAR STATION TABLE 4.2.D MINIML'M TEST AND CALIBRATION FREQUENCIES FOR RADIATION MONITORING SYSTEMS Instrument Functional Calibration Instrument System I.D. No. Test Freq. Freq. Check Instrument Channels Steam Jet Air Ejector Off-Gas System RMP-RM-150 A & B (13) (13) (13)

Reactor Building Isolation and RMP-RM-452 A & B (13) (13) (13)

Standby Gas Treatment Initiation Liquid Radwaste Discharge Isolation (RMP-RM-2) (12) (12) (12)

Main Control Room Ventilation RMV-RM-1 Once/ Month (1) Once/3 Months once/ Day Isolation 1 Mechanical Vacuum Pump Isolation RMP-RM-251, A-D See Tables j$ 4.1.1 & 4.1.2 Logic Systems SJAE Off-Gas Isolation Once/ Year Standby Gas Treatment Initiation Once/6 Months N

()3 Reactor Building Isolation once/6 Months Liquid Radwaste Disch. Isolation once/6 Months Main Control Room Vent Isolation Once/6 Months N

Mechanical Vacuum Pump Isolation Once/ Operating Cycle

NOTES FOR TABLES 4.2. A THROUGH 4.2.F

1. Initially once every month until exposure (M as defined on Figure 4.1.1) is 2.0 X 105 ; thereafter, according to Figure 4.1.l(after AEC approval). The compilation of instrument failure rate data may include data obtained from other boiling water reactors for which the same design instrument operates in an environment similar to that of CNS.

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

3. This instrumentation is excepted from the functional test definition. The functional test will consist of applying simulated inputs. Local alarm lights representing upscale and downscale trips will be verified but no rod block will be produced at this time. The 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 operating cycle.

4. Simulated automatic actuation shall be performed once each operating cycle.

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

5. Reactor low water 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 systems.

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. Portions of the logic is checked more fre-quently during functional tests of the functions that produce a rod block.

10. The detector will be inserted during each operating cycle and the proper amount of travel into the core verified.

11. The RSCS Rod Group C Bypass function is required for the first 6500 FND/T of the initial core loading. This function is provided by two pressure transducers which sense turbine first stage pressure which is then correlated with core thermal power. This bypass function assures that control rod worths are con-trolled as described in the Basis for Specification 3.3.b.3.

12. Surveillance requirements for this system are defined in Table 4.20.A.1.

13. Surveillance requirements for this system are defined in Table 4.20.A.2.

2273 118 3.2 BASES (Cont'd)

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 leaves the Reactor Building via the normal ventilation path but rather all the activity is processed by the standby gas treatment system.

Flow transmitters are used to record the flow of liquid from the drywell sumps.

An air sampling system is also provided to detect leakage '_nside the primary containment.

For each parameter monitored, as listed in Table 3.2.F, there are two (2) channels of instrumentation. By comparing readings between the two (2) chan-nels, a near continuous surveillance of instrument performance is available.

Any deviation in readings will initiate an early recalibration, thereby main-taining the quality of the instrument readings.

The recirculation pump trip has been added as a means of limiting the con-sequences of the unlikely occurrence of a failure to scram during an antici-pated transient. The response of the plant to this postulated event falls within the envelope of study events given in General Electric Company Topical Report, NED010349, dated March, 1971.

The liquid radwaste monitor assures that all liquid discharged to the discharge canal does not exceed the limits of Specification 3.20.B. Upon sensing a high discharge level, an isolation signal is generated which closes of radwaste dis-l charge valve. The set point is adjustable to compensate for variable isotopic discharges and dilution flow rates.

The main control room ventilation isolation is provided by a detector moni-toring 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.

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.

2273 119 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.20 ENVIRONMENTAL / RADIOLOGICAL EFFLUENTS 4.20 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 3.20.A.1. monitoring instrumentation chan-nel shall be demonstrated OPER-Specification: ABLE by performance of the CHAN-NEL CHECK, SOURCE CHECK, CRANNEL

a. The radioactive liquid effluent CALIBRATION and CHANNEL FUNCTIONAL monitoring instrumentation chan- TEST operations during the MODES nels shown in Table 3.20.A.1 shall and at the frequencies shown in be OPERABLE with their alarm and Table 4.20.A.l.

trip setpoints set to insure that the limits of 3.20 B.1 are not b. Radioactive liquid effluent noni-exceeded. tor alarm and trip setpoints shall be determined in the manner

b. With a radioactive liquid effluent described in the ODAM. Auditable monitoring instrumentation channel records of the setpoints and set-alarm and trip setpoint less con- point calculations shall be main-servative than required, reset tained, immediately to meet Specification 3.20.A.1.a, suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable,

c. With a radioactive liquid effluent monitoring instrumentation channel 9 L

39g icU inoperable, take the ACTION shown in Table 3.20.A.l.

d. The provisions of Definition J are not applicable. The reporting provisions of Specification 6.7.2.B.2 are not applicable.

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TABLE 3.20.A.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY + ACTION

1. Gross Beta or Gamma Radioactivity Monitor Providing Automatic Isolation

a. Liquid Radwaste Effluent Line 1

18

2. Gross Beta or Gamma Radioactivity Monitors Not Providing Automatic Isolation i

b! a. Service Water Effluent Line 1

20

$i

3. Flow Rate Measurement Devices

a. Liquid Radwaste Ef fluent Line** 1

21 N

()j 4. Tank Level Monitor u

a. Condensate Storage Tank 1 *** 22

NOTUS FOR TABLE 3.20.A.1

During releases via this pathway.

- Pump curves may be utilized to estimate flow; in such cases, Action Statement 21 is not required.

- - During liquid addition to the tank.

+ Channel (s) shall be OPERABLE and in service 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 for up to 14 days, providing that prior to initiating a release:

1. At least two independent samples are analyzed in accordance with Specification 4.20.B.1.c 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 charnels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 45 days provided that at least once every day a grab sample is collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection not greater than 10-6 pCi/ml.

ACTION 21 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, ef fluent releases via this pathway may continue for up to 14 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

ACTION 22 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, liquid additions to this tank may continue for up to 28 days provided the tank liquid level is estimated during all liquid additions to the tank.

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TABLE 4.20.A.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST

1. Gross Beta or Gamma 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 Provinding Alarm but not Providing Auto-matic Isolation (4)

a. Service Water System Effluent Line D* M R(3) Q(2) 4 E

i

3. Finw Rate Measurement Devices

a. Liquid Radwaste Effluent Line D(4)* NA R SA N

N wl

4. Tank Level Monitor N e. Condensate Storage D** NA R Q trJ

NOTES FOR TABLE 4.20. A.1

During releases via this pathway.

- During liquid additions to the tank.

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

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

2. Circuit failure.

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 exist:

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 include the use of a known (traceable to the National Bureau of Standards radiation measurement system) radioactive source positioned in a reproducible geometry with respect to the sensor and emitting beta and gamma radiation in the ranges measured by the channel during normal operation.

(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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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

W = At least once per 7 days.

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 = Prior to each release.

NA = Not applicable.

2273 124

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.20.A (Co.nt'd) 4.20.A (Cont'd)

2. Gaseous Effluent Monitoring 2. Gaseous Effluent Monitoring Applicability: As shown in Table a. The setpoints shall be deter-3.20.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 channels channel shall be demonstrated shown in Table 3.20.A.2 shall be OPERABLE by performance of the OPERABLE with their alarm setpoints CHANNEL CHECK, SOURCE CHECK, set to ensure that the limits of CHANNEL CALIBRATION, and CHAN-Specification 3.20.C.1 are not NEL FUNCTIONAL TEST operations exceeded, during the MODES and at the frequencies shown in Table

b. With a radioactive gaseous effluent 4.20.A.2.

monitoring instrumentation channel alarm setpoint less conservation c. Auditable records of the set-than a value which will ensure that points and setpoint calcula-the limits of 3.20.C.1 are met, tions shall be maintained.

reset immediately to comply with Specification 3.20.A.2.a or declare the channel inoperable,

c. With one or more radioactive gase-ous effluent monitoring instrumen-tation channels inoperable, take the ACTION shown in Table 3.20.A.2.

d. The provisions of Definition J are not applicable. The reporting provisions of Specification 6.7.2.B.2 are not applicable.

2273 125

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TABLE 3.20.A.2 .

RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY # PARAMETER ACTION

1. Main Condenser Air Ej ector

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

b. Effluent System Flow Rate Measuring Device 1

System Flow Rate Measurement 26

2. Augmented Offgas Treatment System Explosive Gas Monitoring System

a. Hydrogen Monitor 1 ** % Hydrogen 28

, 3. Reactor Building Ventilation Monitor System U

g' a. Noble Gas Activity Monitor 1

Radioactivity Rate Measurement 27 i

b. Iodine Sampler Cartridge 1

Verify Presence of Cartidge 29

c. Particulate Sampler Filter 1

Verify Presence of Filter 29

d. Effluent System Flow Rate Measuring Device 1

System Flow Rate Measurement 26

e. Sampler Flow Rate Measurement Device 1

Sampler Flow Rate Measurement 26

4. (****)

a. Noble Cas Activity Monitor 1

Radioactivity Rate Measurement 27 py b. Iodine Sampler Cartridge 1

Verify Presence of Cartridge 29 rs)

'sJ c. Particulate Sampler Filter 1

Verify Presence of Filter 29 U

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 3.20. A.2

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

During releases via this pathway.

- During Augmented Offgas Treatment System Operation.

- - During operation of the Main Condenser Air Ejector.

- - - Main Stack Monitoring System, Augmented Radwaste Building Ventilation Monitoring System, Radwaste Area (Building) Ventilation Monitoring System (b, c, and e only),

Turbine Building Vent 1'ation 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 envitonment for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 houra.

ACTION 26 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 28 days provided the flow rate is estimated at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

AC1!.ON 27 With the number of channels OPEE4BLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 28 days provided grab samples are taken at least once per day and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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 for up to 28 days provided gas samples are collected at least once per day and analyzed within the ensuing 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 29 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 28 days, provided samples are continuously collected with auxiliary sampling equipment for periods on the order of seven (7) days and analyzed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after the end of the sampling period.

2273 127

-216t-

TABLE 4.20.A.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST

1. Main Condenser Air Ejector

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

b. Noble Gas Activity Recorder D** NA R Q

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

2. Augmented Offgas Treatment System Explosive Gas Monitoring System

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

, 3. Reactor Building Ventilation Monitoring System U

a. Noble Gas Activity Monitor D* R(3)

Q Q(2) R(1)

b. Iodine Sampler Cartridge D* NA NA NA

c. Particulate Sampler Filter D* NA NA NA

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

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

4. (****)

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

b. Noble Gas Activity Recorder D* NA R Q

c. Iodine Sampler D* NA NA NA

d. Particulate Sampler D* NA NA NA N

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

f. Sampler Flow Rate Monitor D* NA R Q

NOTES FOR TABLE 4.20.A.2

During releases via this pathway.

- During augmented offgas treatment system operation.

- - During operation of the main condenser air ejector.

- - - Fbin Stack Monitoring System, Augmented Radwaste Ventilation Monitoring System, Radwaste Area (Building) Ventilation Monitoring System (c, d, and f only), Tur-bine Building Ventilation Monitoring System (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exist:

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

2. Circuit failure.

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 exist:

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

2. Circuit fa ilu re.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

(3) The CHANNEL CALIBRATION shall include the use of known (traceable to the National Bureau of Standards radiation measurement system) radioactive source positioned in a reproducible geometry with respect to the sensor and emitting beta and/or gamma radiation in the range measured by the channel during normal operation.

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

FREQUENCY NOTATION:

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

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

W = At least once per 7 days.

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 = Prior to each release.

NA = Not applicable.

2273 129

-216v-

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.20 (Cont'd) 4.20 (Cont'd)

B. Liquid Effluents B. Liquid Effluents Applicability: At all times. 1. Concentration Specification: a. The concentration of radioactive material in liquid effluents

1. Concentration released from the site 'shall be monitored in accordance with

a. The concentration of radioactive Table 3.20.A.l.

material released from the site to the unrestricted area (Figure b. The liquid effluent monitors 4.20.B.2) shall not exceed the having provisions for automatic concentration specified in 10 CFR termination of liquid releases, Part 20.106 for radionuclides as listed in Table 3.20.A.1, other than dissolved or entrained shall be used to limit the con-noble gases. For dissolved or centration of radioactive mate-entrained noble gases, the concen- rial released from the site to tration shall not exceed 2 x 10-4 not more than the value given in pCi/ml total activity. Specification 3.20.B.I.a.

b. With the concentration of radio- c. The radioactivity content of each active material released from the batch of liquid waste to be dis-site to the unrestricted area charged shall be determined prior exceeding the limit, restore the to release by sampling and analy-concentration within the limit and ses in accordance with Table notify the Commission pursuant to 4.20.B.l. The results of pre-Specification 6.7.2.A. release analyses shall be used with the calculational methods in the ODAM to establish alarm and trip points to assure that the concentration at the unrestricted area boundary does not exceed the value in Specification 3.20.B.l.a.

d. The radioactivity concentration in liquid effluents shall be deter-mined by collection and post-release analysis of samples in accord with Table 4.20.B.l. Cal-culational methods presented in the ODAM shall be applied to these concentration measurements at least once per month to calculate the average concentration at the unrestricted area boundary.

22173 130

-216w-

n t o ) ) ) )

ii 1 2 2 2 mt ( ( ( (

ic) ) 7 6 5 5 7 8 7 7 6 5 57 8 L eDl - - - - - - - - - - - - -

t L m 0 0 0 00 0 0 0 0 0 00 0 reL/ 1 1 1 1 1 1 1 1 1 1 1 1 1 eD( i w C p

x x x xx x x x x x xx x of Lo ( 5 1 1 1 1 5 5 5 1 1 1 1 5 y

t a s a a s i

v m) e m) m) e m8 d s m8 m8 d s is a( na 0 a( a( na 0 ti G) aC a 9 G) G) aG a 9 cs 7 h - / 7 h -

A y l( dd p r l( l( dd p r l as ee l S as as ee l S M f a pr vn A pr pr vn A A on i e li ,

i e i e li ,

R A ct1 oa s 9 ct ct1 oa s 9 G e nt3 sr s 8 nt nt3 sr s 8 O p ii1 st 3 o - ii i i1 st 3 o -

R P

y T

rm- i n - r r rm rm- in - r r PEI DE HG S PE PEI DE HG S S

I S

Y L ) )

A 9 9 N ( (

A ) ) ) ) )

3 3 4 4 4 D h ( ( ( ( (

N y c e e e A sc t ) t ) t )

e e

) t ) ) t ) t 1 min a 9 i 9i 9 9i 9 9i 9i

. G use myu B ( s ( s ( ( s ( ( s ( s B. IN il q P

h M Mo p Qo p W Wop M Mo p Qo p 0L nae c m m m m m

2. P i nr a o o o o o M MAF E C C C C C 4A S

E LE BT AS TA W ) ) )

D 4 4 4 M e ( e ( (

I h / h h l s l s s U y c h c c p u Q gc c p u u nn t

a t t m o m o o I t a a a u a u u L i e B a B B S n S n n l u P PB P P W i M i i E pq h h h b t V me c e c c a n b

a t

n t

n I ar a n a a r o r o o T SF E O E E G C G C C C

A )

O ) 6 I 5 (

D ( e A s g R k r n a a h T c e e s p r i s e D y a t T e a s e

l e W u o

s R e u a c n e e l

e t

s i) v6 r(

i t NNNU _U. .

n R a et o W S n C d

i h e

t u t u c nl n q t af a i a lf l L B PE P

. A. B.

1 2 2 ,

R',

NOTES FOR TABLE 4.20.B.1 (1) The lower limit of detection (LLD) is defined in Definition K.A.

(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 increased inversely proportionally to the magnitude of the gamma yield (i.e., 5 x 10-7/1, 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 2.

(3) A composite sample is one in which the quantity cf 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 concentratious of radioactive materials in liquid effluents, samples shall be collected in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.

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

(6) A grab sample of plant service water effluent shall be analyzed at least once each week in accordance with Table Item 2.A. In the event the radioactivity concen-tration in a sample exceeds 3 x 10-6 pCi/ml, or in the event the plant service water effluent monitor indicates the presence of an activity concentration greater than 3 x 10-6 pCi/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 exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measur-able and identifiable, together with the above nuclides, shall also be identi-fled and reported. Nuclides which are belou 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 semi-annual Radioactive Effluent Release Report.

(8) If an isotopic analysis is unavailable, batch releases may be made for up to 14 days provided the gross beta / gamma concentration to the unrestricted area is

_< 1 x 10-7 pc/ml.

(9) Analysis may be performed after release.

FREQUENCY NOTATION:

S = At least once per 1: hours.

9977 11 D = At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . CL'J IJ 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 =

Prior to each release.

NA = Not applicable.

-216y-

k htsa A$a A

_s) a

\

l s _ ..., _

==

\ -- -% -- . . . _ ,

_ . . . ~ ..s ~-

0._._ _.

\

'y

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j

== = = = =-

J . O .*:0".:".*"*

. . - I -h_ .

\ Sp euene W .08 m Set. .r a. .g O ~

. , .,, %, \

.. *..e e.....- / '

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g x

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e ._.

~a g i I i

, f e =- E5 ' -

i l

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-m,*.

,e ii 4

_ m. -

0 P3-A*44e en Figure 4.20.B.2 Exclusion Area Boundary 273 133 For Gascous and Liquid Effluents

-216z-

LIMITING CONDITION FOR OPERATION SURVE1LLANCE REQUIREMENTS 3.20.B (Cont'd) 4.20.B (Cont'd)

2. Liquid Dose

a. The dose or dose commitment to an a.

individual from radioactive mate- Dose Calculation - Cumulative dose rials in liquid effluents released contributions from liquid efflu-to unrestricted areas (see Figure ents during each calendar quarter 4.20.B.2) shall not exceed 1.5 shall be determined in accordance mrem to the total body or 5 mrem with the Offsite Dose Assessment to any organ during any calendar Manual (ODAM) at least once per quarter, 31 days.

b. b. Doses due to liquid releases to With the calculated dose from the release of radioactive materials unrestricted areas shall be pro-in liquid effluents exceeding the jected at least once per 31 days.

above limit, prepare and submit to c.

the Commission within 30 days, pur- The liquid radwaste system shall suant to Specification 6.7.2.B a be demonstrated OPERABLE at least Special Report which identifies the once per 92 days unless the 11-cause(s) for exceeding the limit (s) quid radwaste system has been utilized to process radioactive and defines the corrective actions to be taken. liquid effluents during the pre-vious 92 days.

c. Every reasonable effort shall be made to maintain the liquid rad-waste treatment system OPERABLE.

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 noble gases, in excess of 0.01 pCi/ml.

d. With radioactive liquid waste being discharged without treatment in ex-[} j}4 cess of the above limit, prepare and submit to the Commission within 30 days, pursuant to Specification 6.7.2.B. a Special Report which in-cludes the following information:

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.

                                                -216cl-

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS .

3. 20.B (Cont 'd) 4.20.B (Cont'd)

3. Condensate Storage Tank 3. Condensate Storage Tank

a. The quantity of radioactive mate- a. The quantity of radioactive mate-rial contained in the Condensate rial contained in the Condensate Storage Tank shall be limited to Storage Tank shall be determined

_< 25 curies, excluding tritium and by analyzing a representative dissolved or entrained nobel gases. sample of the tank's contents at least once per 7 days when radio-

b. With the quantity of radioactive active materials are being added material in the Condensate Storage to the tank.

Tank exceeding 25 curies, immedi-ately suspend all additions of radioactive material to the tank and within 48 hours either reduce the tank contents to within the limit or provide prompt notifica-tion to the Commission pursuant to Specification 6.7.2.A. The writ-ten followup report shall include a schedule and description of acti-vities planned and/or taken to re-duce the tank contents to within the limit.

c. The provisions of Definition J are not applicable.

2273 135

                                               -216a2-

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.20 (Cont'd) 4.20 (Cont'd) C. Gaseous Effluents C. Gaseous Effluents Applicability: At all times. Specification:

1. Total Dose 1. Total Dose

a. The concentration of radioactive a. The release rate of radioactive noble gas in air offsite due to noble gas shall be monitored gaseous effluents shall not exceed according to Specification the concentration specified in 10 3.20.A.2.

CFR Part 20.106.

b. A radioactive noble gas effluent

b. With the concentration exceeding monitor shall cause automatic the limit in 3.20.C.I.a, decrease alarm when the concentration ex-the release rate to comply with the ceeds the monitor alarm setpoint, limit and provide prompt notifica- determined as specified in the tion to the Commission pursuant to ODAM.

Specification 6.7.2.A.

c. The provisions of Definition J are not applicable.

2. Noble Gases 2. Noble Gases

a. The air dose in unrestricted areas a. Dose Calculations - Cumulative (see Figure 4.20.B.2) due to noble dose contributions during each gases released in gaseous effluents calendar quarter shall be deter-shall not exceed 5 mrad from gamma mined in accordance with the radiation and 10 mrad from beta method in the ODAM at least once radiation during any calendar quar- every 31 days.

ter.

b. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.7.2.B, a Special Report which identifies the cause(s) for exceeding the lim-

                                                                         !       b it(s) and defines the corrective actions to be taken.

c. The provisions of Definition J are not applicable.

                                              -216a3-

TABLE 4.20.C.1

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

Gaseous Release Type Frequency Frequency Analysis (pCi/ml)(1) A. 1. Main Stack M(3) M(3) Principal Gamma 1 x 10-4(2)

2. Reactor Bldg Vent Grab Emitters (7)

3. Augmented Radwaste Sample Bldg Vent

4. Turbine Bldg Vent Q Q H-3 1 x 10-6 (Gaseous) Grab Sample B. All Release Types as Continuous (6) W(4) 1-131 1 x 10-12 b Listed in A Above, Charcoal I-133 1 x 10-10 5 & Radwaste Bldg Vent Sample

' (Iodine & Particulate) Continuous (6) W(4) Principal Gamma 1 x 10-II(2) Particulate Emitters (7) Sample (I-131, Others) Continuous (6) M Gross Alpha 1 x 10-II Particulate Sample (5) Continuous (6) Q Sr-89, Sr-90 1 x 10' N Composite N Particulate N Sample (8) U4 _ Continuous (6) Noble Gas Gross Noble Gases 5 x 10-6 u Monitor Beta and Gamma (9) N

NOTES FOR TABLE 4.20.C.1 (1) The lower limit of d etection (LLD) is defined in Definition K. A. (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 increased inversely proportionally to the magnitude of the gamma yield (i.e., 1 x 10-4/1, 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 of radionuclides. (4) 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 of radionuclides. When samples collected for 24 hours are analyzed, the corresponding LLD's may be increased by a factor of 10. (5) One week's filter will be analyzed for alpha cach month. (6) 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 3.20.C.1, 3.20.C.2 and 3.20.C.3. (7) The principal gamma emitters for which the LLD specification will apply are exclusively 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, tobather 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 result in LLD's higher than required, the reasons shall be documented in the semi-annual effluent report. (8) A quarterly composite particulate sample shall include one of every three successive particulate samples collected during the quarter. (9) The noble gas continuous monitor shall be calibrated using laboratory analysis of the grab samples from A and B on Table 4.20.C.1 or using reference sources. FREQUENCY NOTATION: S = At least once per 12 hours. W D =

At least once per 24 hours. 2273 138 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 = Prior to each release. NA = Not applicable.

                                               -216a5-

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.20.C (Cont'd) 4.20.C (Cont'd)

3. Iodine and Particulate 3. Iodine and Particulate

a. The dose to an individual from ra- a. Dose Calculations - Cumulative dioiodines, radioactive materials dose contributions during each in particulate form, and radionu- quarter shall be determined in clides other than noble gases with accordance with the ODAM at least half-lives greater than 8 days in once every 31 days.

gaseous effluents released to un-restricted areas (see Figure 4.20.B.2) shall not exceed 7.5 mrem during any calendar quarter.

b. With the calculated dose from the release of radionuclides, radio-active materials in particulate f o rm , or radionuclides other than noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commis-sion within 30 days following the end of the calendar quarter in which the release occurred pur-suant to Specification 6.7.3.B a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken.

c. The provisions of Definition J are not applicable.

4. Gaseous Radwaste System 4. Gaseous Radwaste System

a. The gaseous radwaste treatment a. Doses due to gaseous releases to system and the ventilation exhaust unrestricted areas shall be pro-treatment system shall be OPERABLE. jected at least once per 31 days The gaseous radwaste treatment using calculational methods in system shall be operated to reduce the ODAM. The appropriate sys-radioactive materials in gaseous tems shall be demonstrated OPER-wastes prior to their discharge ABLE at least once per 92 days when the projected gaseous ef flu- unless the appropriate system ent air doses due to gaseous efflu- has been utilized to process ra-eut releases to unrestricted areas dioactive gaseous effluents dur-(see Figure 4.20.B.2) when averaged ing the previous 92 days.

over 31 days would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation. The ventila-tion exhaust treatment system shall be operated to reduce the radio- 2273 139 active materials in gaseous waste prior to their discharge when the projected gaseous effluent doses due to gaseous effluent releases to unrestricted areas (see Figure 4.20.B.2) when averaged over 31 days would exceed 0.3 mrem to any organ. -216a6-

LIMITlNG CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 1 3.20.C (Cont'd) 4.20.C (Cont'd)

b. With gaseous wast - being dis-charged for more ta n 31 days with-out treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.7.2.B, a Special Report which includes the following information:

1) Identification of equipment of subsystems not OPERABLE and the reason for nonoperability.

2) Action (s) taken to restore the non-operable equipment to OPERABLE STATUS.

3) Sewary description of ac-clon(s) taken to prevent a recurrence.

c. The provisions of Definition J are not applicable.

5. Hydrogen Concentration 5. Hydrogen Concentration

a. The concentration of hydrogen in a. The concentration of hydrogen or the augmented offgas treatment sys- oxygen in the augmented offgas tem downstream of the recombiners treatment system downstream of the shall be limited to 1 4% by volume, recombiners shall be determined by continuously monitoring the

b. With the concentration of hydrogen waste gases in the main condenser oc oxygen in the augmented offgas offgas treatment system with the treatment system downstream of the (hydrogen) monitors required OPER-recombiners exceeding the limit, ABLE by Table 3.20.A.2.

restore the ( oncentration to within the limit within 48 hours.

c. The provisions of Definition J are not applicable. The reporting provisions of Specification 6.7.2.B.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 5 (100 pCi/ tor shall be determined at the sec/MWt). 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 (100 pCi/sec/MWt), re- dilution and/or discharge) of the store the gross radioactivity rate main condenser air ejector:

to within its limit within 72 hours or be in at least HOT STANDBY with-in the next 12 hours.

                                                                   }}[}      j

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.20.C (Cont'd) 4.20.C.6 (Cont'd)

1) At least once per 31 days.

2) Within 4 hours following an increase, as indicated by the Condenser Air Ejector Noble Gas Activity Monitor, of greater than 50%, af ter fac-toring out increases due to changes in THERMAL POWER level, in the nominal steady state fission gas release f rom the primary coolant.

7. Containment

a. Whenever the primary containment is vented / purged, ic shall be vented / purged through the Standby Gas Treatment System. With this specification not satisfied, sus-pend all venting / purging of the containment.

b. The provisions of Definition J are not applicable. The reporting provisions of Specification 6.7.2.B.2 are not applicable.

D. Ef fluent Dose Liquid / Gaseous D. Ef fluent Dose Liquid / Gaseous Applicability: At all times. 1. Dose Calculations - The cumu-lative dose to an individual Specification: contributed by radioactive ma-terial in gaseous and liquid

1. The dose or dose commitment to a ef fluents shall be calculated at real individual from all uranium least once per year in order to fuel cycle sources is limited to verify compliance with Specifi-

             < 25 mrem to the total body or any               cation 3.20.D.

organ (except the thyroid, which is limited to 5 75 mrem) over a period of 12 consecutive months. With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifications 3.20.B.2, 3.20.C.2, or 3.20.C.3, prepare and submit a 22'73 141 Special Report to the Commission pursuant to Specification 6.7.3.B and limit the subsequent releases such that the dose or dose commit-ment to a real individual from all uranium fuel cycle sources is lim-ited to < 25 mrem to the total body or any organ (except thyroid, which is limited to < 75 mrem) over 12 consecutive months. This Special 6a8-(Cont d)

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.20.D (Cont'd) 4.20 (Cont'd) Report shall include an analysis which demonstrates that radiation exposures to all real individuals from all uranium fuel cycle sources (including all effluent pathways and direct radiation) are less than the 40 CFR Part 190 Standard. Otherwise, obtain a variance from the Commission to permit releases which exceeds the 40 CFR Part 190 Standard.

2. The provisions of Definition J are 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 solidify and package radio- meters and limits. In addition, active waste and meet the require- every 10th batch of dewatered ments of 10 CFR Part 20 and 10 CFR waste will be sampled prior to so-Part 71 prior to shipment of radio- lidification and analyzed for pH.

active wastes from the site.

2. When the requirements of 10 CFR 2. Each drum of solidified radio-Part 20 and 10 CFR Part 71 are not active waste will be visually satisfied, prepare and submit to inspected, prior to capping, to the Commission, pursuant to Speci- insure that there is no free fication 6.7.3.B, a Special Report standing liquid on top of the which includes the following infor- solidified waste.

mation:

3. The Semiannual Radioactive Mate-

a. Indentification of the inoperable r!al Release Report in Specifi-equipment, cation 6.7.1.F shall include the

b. Cause of inoperability, following information for each

c. Action (s) taken to restore the inop- type of solid waste shipped off-site during the report period:

erable equipment to operable status,

d. Length of time the above require-ments were not satisfied, b. Total curie quantity (determined by measurement or estimate),

e. Identification and description of alternate equipment, if any, that c. Principal gamma r-dionuclides may be used to satisfy the above (determined by nuasurement or requirement, estimate),

f. Action taken to prevent recurrence. d. Type of waste,

3. The provisions of Definition J are e. Type of container, not applicable.

f. Solidification agent.

                                             -216a9-

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.20.F (Cont'd) 4.20 (Cont'd)

3. In the event a location is identi-fled at which the calculated per-sonal dose associated with one or more exposure pathways exceeds the calculated dose associated with like pathways at a location where sampling is conducted as specified by the REMM, then the pathways hav-ing maximum exposure potential at the newly identified location will be added to the radiological moni-toring program and to the REKM at the next SRAB meeting if samples are reasonably attainable at the new location. Like pathways moni-tored (sampled) at a location, excluding the control station lo-cation (s), having the lowest asso-ciated calculated personal dose may be deleted from the REMM at the time the new pathway (s) and location are added.

6. A change in the REMM shall be des-cribed in the Monthly Operating Report within 90 days after the change was made effective.

7. The provisions of Definition J are not applicable.

G. Interlaboratory Comparis n 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 and in accordance with the Radi-Specification: ological Environmental Monitoring Manual (REMM) shall be included

1. Analyses shall be performed on in the Annual Radiological En-radioactive materials supplied as vironmental Report, pursuant to part of an Interlaboratory Com- Specification 6.7.1.A.E.

parison Program.

2. With analyses not being performed as required in Specification 3.20.G.1, report the corrective ac-tions taken to prevent a recurrence to the Cornission in the Annual Radiological Environmental Report.

22[3 lf}

3. The provisions of Definition J are not applicable.

                                             -216all-

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.20 (Cont'd) 4.20 (Cont'd) F. Monitoring Program F. Monitoring Program Applicability: At all times. 1. Radiological environmental samples shall be collected and analyzed Specification: as specified in the REMM.

1. A radiological environmental moni- 2. A land use census shall be con-toring program shall be conducted ducted annually and shall iden-in accordance with the Radiologi- tify the location of the nearest cal Environmental Monitoring Manual garden that is greater than 500 (REMM). square feet in area and that yields edible leafy vegetables,

2. In the event the radiological en- the location of the nearest milk vironmental monitoring program is animal, and the location of the not conducted as specified in the nearest resident in each of the REMM, prepare and submit to the 16 meteorological sectors within Commission in the Annual Operating three miles of the Station. The Report the reasons for not conduct- land use census shall be conduc-ing the program in accordance with ted at least once per 12 months, the REMM and the plans for prevent-ing a recurrence. 3. The results of sample analyses performed in accordance with the

3. When the radioactivity in a sampled REMM shall be summarized in the environmental medium, averaged over Annual Radiological Environmental a calendar quarter, exceeds an ap- Report.

propriate value stated in Table 6.7-2, prepare and submit to the 4. The results of the land use cen-Commission within 30 days from the sus shall be included in the end of the affected calendar quar- Annual Radiological Environmental ter a Special Report which includes Report. an evaluation of any release con-ditions, environmental factors or other conditions which caused the value(s) of Table 6.7-2 to be ex-ceeded. If the radioactivity in environmental sample (s) is not at-tributable to release from the 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.

                                                                       }}[}     )ff

4. When environmental sampling medium is not available from a sampling location designated in the REMM, the cause and the location where replacement samples were obtained shall be reported in the Annual Radiological Environmental Report.

                                               -216a10-

3.20 & 4.20 BASES 3.20.A & 4.20.A INSTRUMENTATION 3.20.A.1 & 4.20.A.1 Liquid Effluent Monitoring 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. 3.20.A.2 & 4.20.A.2 _ Gaseous Effluent Monitoring The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseaus 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 be calculated in accordance with methods in the ODAM, which have beon 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 General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. 3.20.B & 4.20.B LIQUID EFFLUENTS 3.20.B.1 & 4.20.B.1 Concentration This specifcation is provided to ensure that the conceatration 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 radicisotope and its MPC in air (submersion) was converted to an equivalent concentr.ition 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 mci /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 3.20.B.2.a limit is provided. 2273 145

                                              -216a12-

3.20 & 4.20 BASES (Coat'd) 3.20.B & 4.20.B LIQUID EFFLUENTS (Cont'd) 3.20.B.2 & 4.20.B.2 Liquid Dose Specifications 3.20.B.2, 3.20.C.2 and 3.20.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 tha 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. 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. of the Commission;Ithe Commission and the requirement are clearly stated in the opinion Both the intent of 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-fled 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 guidance to licensees in this respect. 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. 2 The concludion 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) 2273 146

                                                       -216a13-

3.20 & 4.20 BASES (Cont'd) 3.20.B & 4.20.B LIQUID EFFLUENTS (Cont'd) 3.20.B.2 & 4.20.B.2 Liquid Dose (Cont'd) The action to be taken by a licensee in the event a limiting condition is e3ceeded, is stated in Appendix I, Section IV.A and in the opinion of the Commission. Techni-cal Specifications 3.20.B.2, 4.20.B.2, 3.20.C.2, 4.20.C.2, 3.20.C.3 and 4.20.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, Section IV), no recommendation was made by either the Staff in its Concluding Statement 4 or by the Commission in its Opinion5 that design objective values should appear as technical specification limits. The opinion of the Commission and the statement of Appendix 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 to Section IV.B. With the implementation of Specificat .20.B.2 and 4.20.B.2 there is reasonable assurcnce that Station operation will noc cause a radionuclide concentration in public drinking water taken from the River that exceeds the standard for anthropogenic radioactivity in community drinking water. The equations in the ODAM for calculating doses due to measured releases of radioactive material in liquid effluent will be consistent with the methodology in Regulatory Guides 1.109 and 1.113. The assessment of personal doses will examine potential exposure pathways including consumption of fish and water taken irom the River downstream of the discharge canal. Specification 3.20.B.2.c implements the requirements of 10 CFR Part 50.36a(a)(1) that operating procedures be established and followed and that equipment be maintained and used to keep releases to the environment as low as is reasonably achievable. The 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 objectives 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 pCi/ml, below which liquid rad-waste treatment would not be cost-beneficial, and therefore not required, is demonstrated below: The quantity of radioactive material in liquid effluent released annually from Cooper Station has been calculated to be 6 total iodines 3.65 curies total others (less H ) 0.7 total 4.35 curies (Cont'd) 2273 l47

                                               -216a14-

3.20 & 4.20 BASES (Cont'd) 3.'3.8 6 4.20.B LIQUID EFFLUENTS (Cont'd) 0.20.B.2 & 4,20.B.2 Liquid Dose (Cont'd) The population dose commitment resulting from the radioactive material in liquid effluent released annually has been calculated to be thyroid 1.95 man-rem total body 0.56 total 2.5 man-rem Therefore, population doses are about 0.5 man-rem per curie of iodine released and about 0.8 man-rem per curie of other radionuclides (less H3 ) released in liquids. It would be conservative to assume one man-rem committed per curie released in liquid effluent. Tha volume of liquid waste processed and intended for dir charge is estimated to be: Low Purity Waste 5700 gal / day - 1.8 x 106 gal /yr Chemical Waste + Demin Regenerant Waste 4000 gal / day - 1.2 x 106 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 Assuming the cost-benefit balance is $1,000 expenditure per man-rem 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

                             ,000 C=$1.8E+6     gal x 3785 m1 x 1 curie x 106 pCi x 1 man-rem man-rem    curie       $1,000 gal C = 0.013 pCi/ml (Cont'd) 2273 148
                                                -216al5-

3.20 & 4.20 BASES (Cont'd) 3.20.B & 4.20.B LIQUID EFFLUENTS (Cont'd) 3.20.B.2 & 4.20.B.2 Liquid Dose (Cont'd) (2) the activity concentration below which evaporation is not cost-beneficial is

                      , $114,000                x 1 curie x 106 pCi x_1 man-rem 1.2E+6 gal x 3785 ml     man-rem    curie      $1,000 gal C = 0.025 pCi/ml Therefore, to one significant digit, radwaste treatment of liquids containing less than 0.01 pCi/ml is not justified.

I NRC Commissioners, " Opinion of the Commission," in the Appendix I Rulemaking Hearing, Docket Rm-50-2, p. 101-102, April 30, 1975. 2 NRC Staff, " Concluding Statement of the Regulatory Staff," in the Appendix I Rule-making Hearing, Docket RM-50-2, pp. 17, 69, 73, 115, February, 1974. 3 NRC Commissioners, p. 101. 4 NRC Staff, op. cit. S NRC Commissioners, op. cit. 6 Demonstration of Compliance with 10 CFR 50 Appendix I, Revision 1 and Supplement 2, Nebraska Public Power District, Cooper Nuclear Station, January 9, 1978. 2273 149

                                              -216a16-

3.20 & 4.20 BASES (Cont'd) 3.20.B & 4.20.B LIQUID EFFLUENTS (Cont 'd)

3. 20.B. 3 & 4. 20.B. 3 _ Condensate Storage Tank Restricting the quantity of radioactive material contained in the Condensate Storage Tank provides assurance that in the event of an uncontrolled release of the tanks' con-tents, the resulting dose commitment to an individual in an unrestricted area will not exceed 0.5 rem.

3.20.C & 4.20.C GASEOUS EFFLUENTS 3.20,C.1 & 4.20.C.1 Total Dose Specification 3.20.C.I.a is included to assure that a measure of control is provided over the concentration of radionuclides in air entering the unrestricted area. Radio-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 unrestricted area limit specified in 10 CFR Part 20, Appendix B. With prompt action to reduce the radioactive noble gas concentration in effluent following alarm initiation, it can be maintained at a samil fraction of the technical specifi-cation 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 _5 (3000) mrem / year to the skin. Radiciodines 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 3.20.C.2 and 3.20.C.3 the average offsite concentration will be maintained at a small fraction of the 10 CFR Part 20.106 concentration limit. 3.20.C.2 & 4.20.C.2 Noble Cases Assessments of dose required by Specifications 4.20.C.2 and 4.20.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 3.20.C.2 and 4.20.C.2 are also discussed in the bases for Specifications 3.20.B.2 and 4.20.B.2. 3.20.C.3 & 4.20.C.3 Iodine and Particulates The bases for Specifications 3.20.C.3 and 4.20.C.3 are discussed in the bases for Specifications 3.20.B.2 and 4.20.E.2. 2273 150

                                                  -216a17-

7.20 & 4.20 BASES (Cont'd) 3.20.C & 4.20.C GASEOUS EFFLUENTS (Cont'd) 3.20.C.4 6 4.20.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. 3.20.C. 5 & 4.20.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. 3.20.C.6 6 4.20.C.6 Air Ejector Restricting the gross radioactivity rate of noble gases from 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. 3.20.C.7 6 4.20.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. 3.20.D & 4.20.D EFFLUENT DOSE LIQUID / GASEOUS This specification is provided to meet the reporting requirements of 40 CFR Part 190. A contribution from another fuel cycle facility is not added since there is no licensed fuel cycle facility within 50 miles of Cooper Station. 3.20.E & 4.20.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.

                                               -216a18-              2273 151

3.20 & 4.20 BASES (Cont'd) 3.20.F & 4.20.F MONITORING PROGRAM The radiological environmental 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 land use census is conducted annaully to identify changes in use of the unre-stricted area in order to recommend modifications in monitoring programs for evalu-ating individual doses from principal exposure pathways. In order that the monitoring program may be adjusted on the bases of operational experience and the land use census, it is described in a separate Radiological Environmental Monitoring Manual (REKM). The REMM describes required sampling locations, sampling and collection frequency, type and frequency of analyses, lower limits of analytical detection, the land use census, and the Interlaboratory Comparison Program. Changes to the REMM, and therefore to the conduct of the program may be made by the licensee only in accord with conditions stated in Specification 6.8.2 and 3.20.F. 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 determine this minimum garden size, the following assumptions were used, 1) that 207. of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vege-tation yield of 2 kg/ square meter. 3.20.G & 4.20.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. 22!73 152!

                                            -216a19-

6.2 (Cont'd) tary material reviewed; copies of the minutes shall be forwarded to the Chairman of the NPPD Safety Review and Audit Board and the Director of Power Supply within one month.

7. Procedures:

Written administrative procedures for Committee operation shall be prepared and maintained describing the method for submission and content of presentations to the committee, provisions for use of subcommittees, review and approval by members of written Committee evaluations and recommendations, dissemination of minutes, and such other matters as may be appropriate. B. NPPD Safety Review and Audit Board. The board must: verify that operation of the plant is consistent with company policy and rules, approved operating procedures and operating license provisions; review safety related plant changes, proposed tests and procedures; verify that unusual events are prompt-ly investigated and corrected in a manner which reduces the proba-bility of recurrence of such events; and detect trends which may not be apparent to a day-to-day observer. Audits of selected aspects of plant operation shall be performed with a frequency commensurate with their safety significance and in such a manner as to assure that an audit of all nuclear safety related activities is completed within a period of two years. Periodic review of the audit programs should be performed by the Board at least twice a year to assure that such audits are being accomplished in accordance with requirements of Technical Specifications. The audits shall be performed in accordance with appropriate written instructions or procedures and should include verification of com-pliance with internal rules, procedures (for example, normal, off-normal, emergency, operating, maintenance, su rveillance, test and radiation control procedures and the energency and security plans), regulations involving nuclear safety and operating license provisions; training, qualification and performance of operating staff; and corrective actions following abnormal occurrences or unusual events. A representative portion of procedures and records of the activities perforced during the audit period shall be audited and, in addition, observations of performance of operating and maintenance activities shall be included. Written reports of such audits shall be reviewed at a scheduled meeting of the Board and by appropriate members of management including those having responsibility in the area audited. Follow-up action, including reaudit of deficient areas, shall be taken when indicated. In addition to the above, the Safety Review and Audit Board will audit the facility Fire Protection Program, Radiological Environ-mental Monitoring Manual, 0?fsite Dose Calculation Manual and their implementing procedures at least once every 24 months. 2273 153

                                           -222-

6.3 Station Operating Procedures 6.3.1 Station personnel shall be provided detailed written procedures to be used for operation and maintenance of system components and sysrems that could have an effect on nuclear safety. 6.3.2 Written integrated and system 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 abnor-mal reactivity changes. C. Emergency conditions involving possible or actual releases of radioactive materials. D. Implementing procedures of the Security Plan and the Emergency Plan. E. Implementing procedures for the fire protection program. F. Implementing procedures for the Offsite Dose Assessment Manual. 6.3.3 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 Specifi-cations. 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. 6.3.4 Radiation control procedures shall be maintained and made available to all station personnel. These procedures shall show permissible radiation exposure, and shall be consistent with the requirements of 10 CFR 20. 2273 154

                                             -226-

6.7 Station Reporting Requirements 6.7.1 Routine Reports A. Requirements In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Director of the appropriate NRC Regional Of fice of Inspection and Enforcement unless otherwise noted. B. Startup Report

1. A summary report of plant startup and power escalation testing shall be submitted following:

a. Receipt of an operating license.

b. Amendment to the license involving a planned increase in power level,

c. Installation of fuel that has a dif ferent design or has been manufactured by a different fuel supplier.

d. Modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

The report shall address each of the tests identified in the FSAR and shall include a description of the measured valucs of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfac-tory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

2. Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criti-cality, whichever is earlicct. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

C. Annual Reports Routine reports covering the subj ects noted in 6.7.1.C.1 6.7.1.C.2, 6.7.1.C.3 and 6.7.1.C.4 for the previous calendar year shall be submitted prior to March 1 of each year. 2273 155

                                          -230-

6.7.1.C (Cont'd)

1. A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving 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 maintenance (describe mainte-nance), waste processing, and refueling. The dose assignment to various duty functions may be estimaten based on pocket dosimeter, TLD, or film badge measurements. Small exposurcs 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 as-signed to specific major work functions.

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

3. Pursuant to 6.6.2G, Design Eatigue Usage, a listing of the number of events identified in 6.6.2.G.2.b will be tabulated and compared to the design or allowed quantity of comparabic or more severe events. In those cases where recalculation of fatigue usage is required per 6.6.2.G.2.c and the calculated usage exceeds two times the design usage limit of the Code, the report will define the inservice inspections that will be performed on that portion of the RCPB to monitor for crack init ia tion.

4. Pursuant to 3.8. A, a report of radioactive source leak testing.

This report is required only if the tests reveal the presence of 0.005 microcuries 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 opera-tion of the facility, shall be submitted on a monthly basis to the Director, Office of Management and Program Analysis, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the appropriate Regional Office, no later than the 15th of each month following the calendar month covered by the report. In addition, any changes to the Radiological Environmental Monitoring Manual of Specification 6.8 or changes to the Offsite Dose Assessment Manual of Specification 6.9 shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made effective. A major change to a radioactive waste treatment system, as defined in Specification 6.10.3, shall be reported to the Commission pur-suant to Specification 6.10 by description in the Monthly Operating Report for the period in which the change was made. 2273 156

                                             -231-

6.7.1 (Cont'd) E. Annual Radiological Environmental Report

1. A report of radiological environmental surveillance activi-ties related to the Station operation during each calendar year shall be submitted to the NRC before May 1 of the fol-lowing year. Each report shall include the information specified in Specification 6.7.1.E.2.

2. The Annual Radiological Environmental Report shall include a summary description of the radiological environmental monitoring program, including a map of all sampling locations.

3. The report shall include a summary of results of the land use census required in Specification 4.20.F.2.

4. The report shall include a summary of results of analyses obtained by participation in the Interlaboratory Compari-son Program required by Specification 3.20.G.

5. The report shall include summarized and tabulated results in the format of Table 6.7-1 of analyses of samples required by the radiological environmental monitoring program. In the event that some results are not available, the reasons shall be explained in the report. In the event the missing results are obtained, they shall be submitted in a supple-mentary report.

6. The Annual Radiological Environmental Report shall include:

a. Summaries, evaluation, and interpretation of the results of radiological environmental monitoring during the year.

b. Comparison of results with operational controls, pre-operational studies, and previous environmental sur-veillance reports as appropriate.

c. An assessment of the observed radiological environ-mental impact of plant operation.

d. A summary of meteorological data collected during the year in the form of joint frequency distributions of wind speed, wind direction and atmospheric stability.

e. An assessment of off-site radiation doses due to radio-active liquid and gaseous ef fluents released from the Station during each calendar quarter of the year and during the year. The dose assessment shall be performed in accordance with the ODAM.

2273 157

                                            -231a-

TABLE 6.7-1 ENVIRONMENTAL RADIOLOGICAL MONITORING PROGRAM SL4L%RY 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 h 0 i N N N u Ln CC Table 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 [].

6.7.1 (Cont'd) F. Semiannual Radioactive Material Release Report

1. A report of radioactive materials released from the Station shall be submitted to the NRC within 60 days after January I and July 1 of each year. Each report shall include the information specified in Specification 6.7.1.F.2 covering the preceeding six months.

2. A Semiannual Radioactive Material Release Report shall include a summary by calendar quarter of the quantities of radioactive liquid and gaseous effluents and radioactive solid waste released from the Station. The data should be reported in the format recommended in Regulatory Guide 1.21, Appendix B, Tables 1, 2, and 3.

3. A Semiannual Radioactive Material Release Report shall include the following information related to each unplanned release radioactive material in gaseous or liquid effluent to offsite environs:

a. A description of the event and equipment involved.

b. Cause(s) of the unplanned release.

c. Actions taken to prevent recurrence.

d. Consequences of the unplanned release.

6.7.2 Reportable Occurrences 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. 22'73 159

                                           -231c-

6.7.2.A (Cont'd)

4. Reactivity anomalies, involving disagreement with the predicted value or reactivity balance under steady state conditions during power operation, greater than or equal to 1% Ak/k; a calculated reactivity balance indicating a shutdown margin less conserva-tive than specified in the technical specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if suberitical, an unplanned reactivity insertion of r, ore than 0.5% Ak/k or occurrence of any unplanned criticality.

5. Failure or malfunction of one or more components which prevents or could prevent, by itself, the fulfillment of the functional requirements of system (s) used to cope with accidents analyzed in the SAR.

6. Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional require-ments of systems required to cope with accidents analyzed in the SAR.

Note: For items 6.7.2.A.5 and 6.7.2.A.6 reduced redundancy that does not result in a loss of system function need not be reported under this section but may be reportable under items 6.7.2.B.2 and 6.7.2.B.3 below.

7. Conditions arising from natural or man-made events that, as a direct result of the event require plant shutdown, operation of safety systems, or other protective measures required by tech-nical specifications.

8. Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the technical specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.

9. Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the safety analysis report or technical specifications bases; or discovery during plant life of conditions not specifically considered in the safety analysis report or technical specifi-cations that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.

Note: This item is intended to provide for reporting of poten-tially generic problems.

10. Occurrence of an unusual or important event that causes a significant environmental impact, that affects potential environ-mental impact from unit operation, or that has high public or potential public interest concerning environmental impact from unit operation.

22'73 160

                                              -233-

6.7.2 (Cont'd) B. Thirty Day Written Reports The reportable occurrences discussed below shall be the subject of written reports to the Director of the appropriate Regional Office within thirty days of occurrence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as naeded, by additional narrative material to provide complete explanation of the circumstances surrounding the event.

1. Reactor protection system or engineered safety feature instru-ment settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillment of the functional requirements of af-fccted systems.

2. Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation.

Note: Routine surveillance testing, instrument calibration, or preventative maintenance which require system configura-tions as described in items 6.7.2.B.1 and 6.7.2.B.2 need not be reported except where test results themselves reveal a degraded mode as described above.

3. Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engi-neered safety feature systems.

4. Abnormal degradation of systems other than those specified in item 6.7.2. A.3 above designed to contain radioactive material resulting from the fission process.

Note: Sealed sources or calibration sources are not included under this item. Leakage of valve packing or gaskets within the limits for identified leakage set forth in technical specifications need not be reported under this item.

5. An unplanned offsite release of 1) more than 1 curie of radio-active material in liquid effluents, 2) more than 150 curies of noble gas in gaseous ef fluents, or 3) more than 0.05 curies of radiolodine in gaseous effluents. The report of an unplanned offsite release of radioactive material shall include the fol-lowing information:

a. A description of the event and equipment involved.

b. Cause(s) for the unplanned release.

c. Actions taken to prevent recurrence.

d. Consequences of the unplanned release.

_234- 2273 161

6.7.2.B (Cont'd)

6. Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table 6.7-2 when averaged over any calendar quarter sampling period.

When more than one of the radionuclides in Table 6.7-2 are d6tected in the sampling medium, this report shall be submitted if: Concentration (1) Concentration (2) Limit Level (1) + Limit Level (2) + ...> 1.0 When radionuclides other than those in Table 6.7-2 are detected and are the result of plant effluents, thia report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limita of Specifications 3.20.B.2.a 3.20.C.2.a, and 3.20.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 condi-tion shall be reported and described in the Annual Radiological Environmental Report. 6.7.3 Unique Reporting Requirements A. Testing Reports Reports shall be submitted to the Director, Nuclear Reactor Regula-tion, USNRC, Washington, D. C. 20555, as follows: 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.

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 may be required covering inspections, test and main-tenance 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 Director of the Office of Inspection and Enforcement Regional Office within the time period specified for each report. 2273 162

                                               -235-

TABLE 6.7-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Reporting Levels Water Airborne Particulate Fieh Milk Vegetables Analysis pCi/l or Gases (pCi/m 3) (pCi/Kg, Wet) (pci/1) (pci/Kg, Wet) H-3 3E + 4 Mn-54 1E + 3 3E + 4 Fe-59 4E + 2 1E+4 Co-58 1E + 3 3E + 4 Co-60 3E + 2 1E + 4 Zn-65 3E + 2 2E + 4 Zr-Nb-95 4E + 2 m I-131 2 0.9 3 1E + 2 Cs-134 30 10 1E + 3 60 IE+3 Cs-137 50 20 2E + 3 70 2E + 3 Ba-La-140 2E + 2 3E + 2 N N N U LN

6.8 RADIOLOGICAL ENVIRONMENTAL MONITORING MANUAL (REMM) 6.8.1 The REMM shall describe the Radiological Environmental Monitoring Program. The description shall identify: A. Sampling locations. B. Sampling and collection frequency. C. Type and frequency of analyses. D. Maximum values for lower limits of detection. E. Land use census. F. Interlaboratory Comparison Program. 6.8.2 District Initiated Changes A. Shall be submitted to the Commission by inclusion in the Monthly Oper-ating Report pursuant to Specification 6.7.1.D within 90 days of the date the change (s) was made effective and shall contain:

1. Sufficiently detailed information to totally support the ration-ale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the REMM to be changed with each page numbered and provided with a signed approval and date box, together with appropriate analyses or evaluations justifying the change (s).

2. A determination that the change will not substantially reduce the information collected on measurable levels of radiation and radioactive materials in the environment.

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

B. Shall become effective upon review and acceptance by the SRAB. 6.9 0FFSITE DOSE ASSESSMENT MANUAL (ODAM) 6.9.1 The ODAM shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid efflu-ents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints consistent with the applicable LCO's contained in these Technical Specifications. 6.9.2 District Initiated Changes A. Shall be submitted to the Commission by inclusion in the Monthly Operating Report pursuant to Specification 6.7.1.D within 90 days of the date the change (s) was made ef fective and shall contain:

1. Sufficiently detailed information to totally support the ration-ale for the change without benefit of additional or supplemental information. Information submitted should consist of a package 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 justifyirig the change (s).

                                           -235b-                       2273 164

6.9.2.A (Cont'd)

2. A determination that the change will not reduce the accuracy or reliability of dose calculations or sc*. point 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. 6.10 MAJOR CHANCES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (LIQUID, GASEOUS, AND SOLID) 6.10.1 The radioactive waste 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 ef fluents and in solid waste packaged for offsite shipment required to meet the LCO's set forth in Specifications 3.20.B. 3.20.C. 3.20.D. and 3.20.E. 6.10.2 Major changes to the radioactive waste systems (liquid, gaseous, and solid) shall be made by either of the following methods. For the purpose of this specification ' major changes' is defined in Specification 6.10.3 below. A. District Initiated Changes

1. The Commission shall be informed of all changes by the inclusion of a suitable discussion of each change in the Semiannual Radioactive Material Release Report for the period in which the changes were made. The discussion of each change shall contain:

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

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

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

d. An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste from those pre-viously predicted in the license application and amend-ments thereto.

e. An evaluation of the change which shows the expected max-imum exposures to individual in the unrestricted area and to the general population from those previously estimated in the license application and amendments thereto.

f. A comparison of the predicted releases of radioactive materials in liquid and gaseous effluents and in solid waste to the actual releases for the period in which the changes were made.

2273 165

                                           -235c-

6.10.2.A (Cont'd)

g. An estimate of the exposure to plant operating personnel as a result of th9 change.

h. Documc~.tation of the fact that the change was reviewed and found acceptable by SORC.

2. The change shall become effective upon review and acceptance by SORC.

B. Commission Initiated Changes

1. The applicability of the change to the facility shall be deter-mined by the SORC after consideration of the facility design.

2. The licensee shall provide the Commission with written notifica-tion of its determination of applicability including any neces-sary revisions to reflect facility design.

3. The change shall become effective on a date specified by the Commission.

6.10.3 Background and definition of what constitutes ' major changes' to radioactive waste systems (liquid, gaseous, and solid). A. Background

1. 10 CFR Part 50, Section 50.34a(a) requires that each application to construct a nuclear power reactor provide a description of the equipment installed to maintain control over radioactive material in gaseous and liquid effluents produced during normal reactor operations including operational occurrences.

2. 10 CFR Part 50, Section 50.34a(b)(2) requires that each applica-tion to construct a nuclear power reactor provide an estimate of the quantity of radionuclides expected to be released annually to unrestricted areas in liquid and gaseous effluents produced during normal reactor operation.

3. .10 CFR Part 50, Section 50.34a(3) requires that each application to construct a nuclear power reactor provide a description of the provisions for packaging, storage and shipment offsite of solid waste containing radioactive materials resulting from treatment of gaseous and liquid effluents and from other sources.

4. 10 CFR Part 50, Section 50.34a(3)(c) requires that each applica-tion to operate a nuclear power reactor shall include (1) a des-cription of the equipment and procedures for the control of gas-cous and liquid effluents and for the maintenance and use of equipment installed in radioactive waste systems and (2) a revised estimate of the information required in (b)(2) if the expected releases and exposures dif fer significantly from the estimate submitted in the application for a construction permit.

2273 166

                                             -235d-

6.10.3.A (Cont'd)

5. The Regulatory staf f's Safety Evaluation Report and amendments thereto issued prior to the issuance of an operating license contains a description of the radioactive waste systems installed in the nuclear power reactor and a detailed evaluation (including estimated releases of radioactive materials in liquid and gaseous waste and quantities of solid waste produced from normal opera-tion, estimated annual aximum exposures to an individual in the unrestricted ar.e and estimated exposures to the general pop-ulation) which shows the capability of these systems to meet the appropriate regulations.

6. The Regulatory staff's Final Environmental Statement issued prior to the issuance of an operating license contains a detailed evalu-ation as to the expected environmental impact from the estimated releases of radioactive material in liquid and gaseous ef fluents.

B. Definition

                " Major Changes" to radioactive waste systems (liquid, gaseous, and solid) shall include the following:

1. Major changes in process equipment, components, structures and effluent monitoring instrumentation from those described in the Safety Analysis Report (SAR) and evaluated in the staff's Safety Evaluation Report (SER) (e.g., deletion of evaporators and installation of demineralizers; use of fluidized bed calciner/

incineration in place of cement solidification systems).

2. Major changes in the design of radwaste treatment systems (liquid, gaseous, and solid) that could significantly alter the charac-teristics ard/or quantities of effluents released or volumes of solid waste otored or shipped offsite from those previously con-sidered in the SAR and SER (e.g., use of asphalt system in place of cement).

3. Changes in system design which may inve.11date the accident analy-sis as described in the SER (e.g. , changes in tank capacity that would alter the curies released).

4. Changes in system design that could potentially result in a signi-ficant increase in occupational exposure of operating personnel (e.g., use of skid mounted equipment, use of mobile processing equipment).

2273 167

                                           -235e-

                                                                                   .'?.                                                   .

s NPPD MANAGEMENT ORGANIZATION CHART Ceneral Manager KPPD Safety Review and Audit Board

  /,            Assistant General Manager                                    Assistant General Manager La              Power Projects Group                                         Power Operations Group cd 1

Director of Power Supply Director of LicensinE Director of

                                                                                           & Quality Assurance        Environmental Affairs Cooper Nuclear Station              Cooper Nuclear Station Engineering Support               Station Superintendent             Quality Assurance Pbnager  Environmental lanager
          )

N N U

     ,,,,                                                                                 Cooper Nuclear Station

Responsible for the Fire Protection Program Quality Assurance Supervisor Q

Figure 6.1.1 NPPD Management Organization Chart}}

ML19270G328

Text

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