ML19262C378
| ML19262C378 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 01/21/1980 |
| From: | NEBRASKA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML19262C377 | List: |
| References | |
| RTR-NUREG-0473, RTR-NUREG-473 TAC-08140, TAC-8140, NUDOCS 8002110458 | |
| Download: ML19262C378 (70) | |
Text
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7 ATTACHMENT ~
PROPOSED' RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS For Cooper Nuclear Station (Note: This cover sheet is not part of the proposed Technical Specifications)
Revised' January 21, 1980 1947 308 8002110 Igg
.' RADIOLOGICAL TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Page N:.
1.0 DEFINITIONS 1 - 5a SAFETY LIMITS LIMITING SAZETY 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 - 1C6 A. Reactivity Limitations A 93 B. Control Rods B 94 C. Scram Insertion Times D. C 97 Reactivity Anomalies E. D 98 Recirculation Pumps E 98 3.4 STANDBY LIQUID CONTROL SYSTEM 4.4 107 - 113 A. Normal Operation B. A 107 Operation with Inoperable Components B C. 108 Sodium Pentaborate Solution C 108 3.5 CJRE AND CONTAINMENT COOLING SYSTEMS 4.5 114 - 131 A. Core Spray and LPCI Subsystems B. A 114 Containment Cooling Subsystem (RHR Service Water) B 116 C. HPCI Subsystem C 117 D. RCIC Subsystem E. D 118 Automatic Depressurization System E F. 119 Minimum Low Pressure Cooling System Diesel F 120 Generator Availability G. Maintenance of Filled Discharge Pipe H. G 122 Engineered Safeguards Compartments Cooling H 123 3.6 PRIMARY SYSTEM BOUNDARY 4.6 132 - 158 A. Thermal and Pressurization Limitations A 132 1947 309 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. Effluent Dose Liquid / Caseous 216a8 E. Solid Radioactive Waste 216a9 F. Monitoring Program 216a10 G. Interlaboratory Comparison Program 216a17 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 218 5.6 Seismic Design 218 5.7 Barge Traffic 218 6.0 ADMINISTRATIVE CONTROLS 211 - 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
5.
Responsibilities Authority 1947 310 220 221
- 6. Records 221
-lii-
TABLE _g CO,ITENTS (Cont'd)
Page No.
6.2.1.B NPPD Safety Review and Audit Board 222
- 1. Membership 223
- 2. Meeting Frequcacy 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 ta' en 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 Repu,t .
231a F. Se=iannual Radioactive Material Release 231c Report -
- 2. Reportable Occurrer.ces 231c A. Procpt 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 Process Control Program 235b 6.9 Offsite Dose Assessment Manual (ODAM) 235b 6.10 Major Changes to Radioactive Waste Treatment Systems 235c i947 3II
-iv-
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 Caseous Radwaste Treatment System - A GASEOUS RAD'JASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous ef fluents 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 prio:
to release to the environment.
C. Hot Standby Condition - Hot standby condition means operation with coolant tem-perature greater than 2120 F, system pressure less than 1000 psig, and the mode switch in "S tartup/Ho t Standby".
H. Immediate - Immedicte 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 kn,wn value(s) of the patameter 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 st:h 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 Cht:k - An instrument check is the qualitative determination of acceptable >perability by observation of instrument behavior during oper-ation. This determinatic; 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 ecmponents are operable per design intent. Where practicable, action will go to completion; i.e. , pumps will be started and valves operated.
1947 312
- 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 cf chan-nel respanse when the channel sensor is exposed to a radioactive source.
- 9. Trip System - A trip system means an arrangement of instrument channel trip signals 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 cne or more plant parameters in order to initiate trip system action. Initiation of protective action may require the tripping of a single trip system or the. coincident tripping of two trip systems.
~~
J. Limiting Conditions for Operation (LCO) - The limiting 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 opera-tional conditions specified for each specification.
Adherence to the requirements of .the LCO within the specified time interval shall constitute compliance.with the speci.fication. In the event the LCO is restored prior to expiration of the specified time interval, completion of the LCO action is not required.
In the event an LCO cannot be satisfied because of circumstances in excess of those addressed in the specification, the facility shall be placed in HOT SHUT-DOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> unless corrective measures are completed that permit operation under the LCO for the specified time interval as measured from initial discovery. Exception to these requirements shall be stated in the individual specifications.
Entry into an operational condition shall not be made unless the conditions of the LCO are met without reliance on the actions specified in the LCO unless otherwise excepted. This provision shall not prevent passage through opera-tional conditions required to comply with the specified actions of an LCO.
K. Limiting Safety System Setting (LSSS) - Tha limiting safety system settings are settings on instrumentation which initiate the automatic protective action at a level such that the safety limits will not be exceeded. The region between the safety limit and these settings represent a margin with normal operation lying below these rettings. The margin has been established so that with p, roper operation of the instrumentation the safety limits will never be exceeded.
l'947 313 K .'A Lower Limit of Detection (LLD) - The LLD is the smallest concentration of radic-active material in a sample that will be detected with 95% probability with 5%
probability of falsely concluding that a blank observation renra m u "real" sJen,1. *
- hods for determining the LLD are contained in th notes able 3.20.F.2. '-
L. Mode - The reactor mode is established by the mode s= lector-switch. The taodes 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 APMi protection (exclud-ing the 15% high flux trip) and RSM interlocks in service.
- 3. Shutdo*m Mode - The reactor is in the shutdo'wn mode when the reactor mode switch if in the shutdown mode position.
- 4. S t artup /Ho t Standby - In this mode the reactor protection scram trips initiated by the main steam line isolation valve closure are bypassed when reactoi 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 d*hdrawal incarlo
( L.A Normal Ventilation - Nomal ventilation is the controlled process of discharging I and replacing air f rom /to a confinement to maintain temperature, humidity, or other conditions necessary for personnel safety and entry. The contents of the atmosphere being discharged from the confinement will have been established prior to establishing normal ventilation followine a purgine/ venting operation.
L.B Offsite Dose Assessment Manual (ODAM) - An 0FFSITE DOSF ASSESSMENT MANUAL (ODAM) shall be a manual containing the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints.
M. Operable - A system or component shall be considered operable when it is capable of performing its intended function.
N. Operating - Operating means that a system or component is performing its intended functions in its required manner.
O. Operating Cycle - Interval between the end of one refueling outage and the end of the next subsequent refueling outage.
P. Primary Containment Integrity - Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:
- 1. All manual containment isolation valves on lines connected to the reactor coolant system or containment which are not required to be open during accident conditions are closed. ~
1947 314
- 2. At least one door in each airlock is closed and sealed.
-4_
- 3. All automatic containment isolation valves are operable or deactivated in the isolated position.
- 4. All blind flanges and manways are closed, wN V P.A Process Control Program - The Process Control Program outlines the solidification of radioactive waste from liquid systems. It does not substitute for station I
operating procedures, but provides a general description of equipment, controls, and practices to be considered during waste solidification to assure solid wastes.
P.B Purge - Purging - Purge or Purging is the controlled process of discharging air l or gas f rom a confinement to establish temperature, pressure, humidity, concentra-tion or other operating condition, in such a manner that replacement air or gas is required to purify the confinement" A A Q. Rated Power - Rated power refers to operation at a reactor power of 2381 mega-watts thermal. This is also termed 100% power and is the maximum power level authorized by the operating license. Rated steam flow, rated coolant flow, rated neutron flux, and rated nuclear system pressure refer to the values of these parameters when the reactor is at rated power. Design power, the power to which the safety analysis applies, is 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 :he "Startup/ Hot Standby" or "Run" position with the reactor critical and above 1% rated power.
S. Reactor Vessel Pressure - Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detectors.
T. Refueling Outage - Refueling Outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the plant after that refueling.
U. Safety Limits - The safety limits are limits within which the reasonable 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 "egulatory 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
" Shutdown"- The reactor is in a shutdown condition when the mode switch is in the position.
- 1. Hot Shutdown means conditions as above with reactor coolant tempe ra tu re greater than 212 F.
1947 315 2.
Cold Shutdown means conditions as above with reactor coolant temperature equal to or less than 212 F and the reactor vessel vented.
'J . 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 Frecuency - 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 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 it.terval.
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 compenent when it is required to be acerable. These tests may be waived when the instrument, component or system is . t 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 EKHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal absorbers and/or 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.
nor Engineered Safety Feature (ESF) atmospheric cleanup systems are nsidmd to be "r';TILATIO" rVHAUST TREATMENT SYSTEM components.
Z.A Venting confinement
- Venting is the controlled process of discharging air or gas from a to establish temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not pro-j vided or required during venting. Vent, used in system names, does not imply a venting process.
Y ^
l l
i I
1947 316
-Sa-
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-Cas System Off-Gas Systen
- a. Operability of the Main Condenser Instrumentation surveillance re-Air Ejector Of f-Gas System is de- quirements are given on Table fined in Table 3.20.A.2. 4.20.A.2.
l
- 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.
l
- 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.l.
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 " Additional 4.2.D.
Safety Related Plant Capabilities."
1947 317 n
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NOTES FOR TABLE 3.2.D
- 1. Action required when component operability is not assured.
A. (1) If radiation level exceeds 1.0 ci/sec (prior to 30 min. delay line) for a period greater than 15 consecutive minutes, the off-gas iso-lation valve shall close and reactor shutdown shall be initiated immediately and the reactor placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
A. (2) Refer to Specification 3.20.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.l.
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.
~1947 319
-63a-
COOPER NUCLEAR STATION TABLE 4.2.D MINIMUM 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 Of f-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 i Mechanical Vacuum Pump Isolation RMP-RM-251, A-D See Tables j$ 4.1.1 & 4.1.2 Logic Systems i SJAE Off-Gas Isolation Once/ Year Standby Gas Treatment Initiation Once/6 Months Reactor Building Isolation once/6 Months Liquid Radwaste Disch. Isolation ,
once/6 Months Main Control Room Vent Isolation Once/6 Months Mechanical Vacuum Pump Isolation (($ Once/ Operating ga Cycle N
W N
CD
NOTL ' rOR 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 <o;trumentation 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 b ecrified 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 func*ioning 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 conparator 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 MWD /T of the initial core loading. This function is provided by two pressure transducers which sense turbine first stage pressure which is then correJated 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.l.
- 13. Surveillance requirements for this system are defined in Table 4.20.A.2.
1947 321 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 inside 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-nelg, i 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 event.s 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.
5947322 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.20 ENVIRONMENTAL / RADIOLOGICAL EFFLUENTS 4.20 ENVIRON TNTAL/ RADIOLOGICAL EFFLLw _S A. Instrumentation A. Instrumentation
- 1. Liquid Effluent Monitoring 1. Lii tid Efflucnt Monitoring Applicability: As shown in Table a. Eac!. radioactive liquid ef fl; ant 3.20.A.l. monitoring ins trumentation clan-nel shall be demonstrated OPER-Suecification: ABLE by performance of the CEC;--
3EL CHECK, SOURCE CHECK, CIU2. L
- a. The radioactive liquid effluent CALIBRATION and CHANNEL FUNC:205AL monitoring instrumentation chan- TEST operations duding the M: ES nels shown in Table 3.20.A.1 shall and at the frequencies shown is 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 m:ni-exceeded. tor alarm and trip setpoints shall be determined in the manner
- b. With a radioactive liquid effluent described in the ODAM. Audi:1b le monitoring instruuentation channel records of the setpoints anc set-alarm and trip setuoint less con- point calculations shall be rain-servative than required, reset tained, immediately to meet Specification 3.20.A.1.a, suspend the release of radioactive liquid effluents monitored by the af fected channel, or declare the channel inoperable.
- c. With a radioactive liquid effluent monitoring instrumentation channel inoperable, take the TIO' -' ~en in Table 1 'O.A ' Times specified in the action statements may be exceeded upon notification to the Commission pursuant to the Special j Reporting requirement of Specif1- /
cation 6.7.3.B.
- d. The provisions of Definition J are not applicable. The reporting provisions of Specification 6.7.2.B.2 are not applicable.
1947 323
-216c-
N O
I 8 0 1 2 T 1 2 2 2 C
A
+
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I L
I B
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- N O
C I
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f I L T P A P T A N
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T S
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I 1
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EE LU BL AF TF E
s D r r I o o U t t Q i i I n n }
L o o M M (e E e V y n yn n I t i t o e i T i L ii n L C v vt i A i n t i a L s t O t o n t l e n k I ci e co t c e n D at u as n i u a A oa l oI e v l T R il f i u e f d o f d c l D f e as E ai f E g RI e
R t f t a a E n e r ac t am e t o mi s a
mo r m s r t mt mt e e a o S aa w au t r w t T Gm d CA a u d i e N o a W s a n t E rt R r g a R o a M ou on e e M s U A d i c M d n R a i ad i i l e T t g u t i v e u e d S en q ev r t q v n N Bi i B o e a i e o I d L r S R L L C si sP sv s w k oo ot o n r r .
r o . l .
a .
GP a GN a F a T a 1 2 3 4
- 4N u)r\ c=
HEY r
NOTES FOR TABLE 3.20.A.1 O
During releases via this pathway.
- n.rinc 11 uid ,a ition e the en k.
+ Channel (s) shall be OPERABLE and in service except that outages for maintenance and required tests, checks, or calibrations are permitted.
ACTION 18 With tue 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.l.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.
O t he rwi se , suspend release of radioactive effluents via this pathway.
ACTION 20 With the numbers of channels OPERABLE less than required by the Minimum Channels OPERABLE r- tirement, effluent releases via this pathway may continue for up t 30 .ays provided that at least once every day a grab sample is collectec and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection not greater than 10-6 pci/ni, ACTION 21 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE v ~ irement, effluent releases via this pathway may continue for up tc 30 1ays provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> o-. ng actual releases. -
ACTION 22 With the number of channels OPERABLE less than required by the Minimum Channels "cRABLE requirement, liquid additions to this tank may continue for up t 30 days provided the tank liquid level is estimated during all liquid acu tions to the tank.
i947 525
-216o-
TABLE 4.20.A.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CilANNEL CilANNEL SOURCE CllANNEL FUNCTIONAL INSTRUMENT CllECK CllECK CALIBRATION TEST
- 1. Cross Beta or Gamma Radioactivity Monitors Providing Alarm and Automatic Isolation
- a. Liquid Radwaste Effluents Line D* P R(3) Q(1)
- 2. Gross Beta or Gamma Radioactivity Monitors Provinding Alarm rut not Providing Auto-matic Isolation j
r a. Service Water System Effluent Line D* M R(3) Q(2)
$ 1
- 3. Flow Rate Measurement Devices
- a. Liquid Radwaste Effluent Line D(4)* NA R SA
- 4. Tank Level Monitor sg) a. Condensate Storage D** NA R Q 4
N cf N-C7s "
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 exists:
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instrument indicates a downscale failure.
- 4. Instrument controls not set in operate mode.
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instrument indicates a downscale failure.
- 4. Instrument controls not set in opeta c 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 t e made at least once daily on any day on which continuous, periodic, or batch releases are made.
FREQUENCY NOTATION:
=
S At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D = At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
W =
At least once per 7 days.
M =
At least once per 31 days.
=
Q At least once per 92 days.
SA =
At least once per 184 days.
A =
At least once per year.
R = At least once per 18 months.
S/U = Prior t, each reactor startup.
P =
kmpleteDriortoeachrelease.
NA = Not-applicable.
f, 4I sci
-216q-
LIMITING CONDITION FOR OPERATION ' SURVEILLANCE REQUIRDIENTS 3.20.A (Cont'd) 4. 20. A (Cont 'd)
- 2. Caseous 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 setpints 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. declare the cbw 4 i_noper or imme-release,
- c. With one or more radioactive gase-ous effluent monitoring instrumen-tation channels inoperable, take
- he ACTION sho in Table 3. 20_. A . 2.
Times specitled in -ne ACTI6N statements maybe exceeded upon notification to the Commission pursuant to the Special Reporting (
requirement of Specification 6.7.3.B. )
- d. The provisions of Definition J are not applicable. The reporting provisions of Specification 6.7.2.B.2 are not applicable.
)94T7 328
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TABLE 3.20.A.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CllANNELS INSTRUMENT OPERABLE APPLICABILITY # PARAMETER ACTION
- 1. Main Condenser Air Ejector
- 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 M' itoring System
- a. Ilydrogen Monitot (4% monitor) 1 ** % llydrogen 28
- 3. Reactor Building Ventilation Monitor System b a. Noble Gas Activity Monitor 1
- Radioactivity Rate Measurement 27 5
Y b. Iodine Sampler Cartridge 1
- Verify Presence of Cartidge 29
- c. Particulate Sampler Filter
- Verify Presence of Filter 1 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 Gas Activity Monitor 1
- Radioactivity Rate Measurement 27
- b. Iodine Sampler Cartridge
- Verify Presence of Cartridge 1 29
- c. Particulate Sampler Filter 1
- Verify Presence of Filter 29 23, d. Effluent Syste' Flow Rate Measuring Device 1
- System Flow Rate Measurement 26 N
- e. Sampler Flow Rate Ileasuring Device 1
- Sampler Flow Rate Measurement 26 u
N 4
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 Monitorin; System, Radwaste Area (Building) Ventilation Monitoring System (b, c, and e only ,
Turbine Building Ventilation Monitoring System.
ACTION 25 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, gases from the main condenser offgas treat-ment system may be released to the environment for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided:
'l . The offgas delay system is not bypassed; and
- 2. The main stack system noble gas activity monitor is OPERABLE:
Otherwise, be in at least HOT STANDBY within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
ACTION 26 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE *aauirement, effluent releases via this pathway may continue for up t- 30 days provided the flow rate is estimated at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 27 With the number of channels OPERA 3LE less than required by the Minimum Channels OPERABLE rn uirement, effluent releases via this pathvay may continue for up tc e 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 28 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement -
eration of the augmented offgas treatment system may continue for up to days provided gas samples are collected at least once per day and ana yzed within the ensuing 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
ACTION 29 With the number c OPERABLE less than required by the Minimum Channels OPERABLE r> uirement, effluent releases via this pathway may continue f or up tc 30 days, provided camn'as 're continuous 1v aollected wi*h auxiliary sampling equioment.
1947 330
-216t-
TABLE 4.20.A.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CllANNEL CilANNEL SOURCE CIIANNEL FUNCTIONAL INSTRUMENT CllECK CllECK C\LIBRATION TEST
- 1. Main Condenser Air Ejector
- a. Nob e Gas Activity !!onitor D*** M R(3) Q(2) R(1)
- b. Effluent System Flow Rate Measuring Device D* NA R Q
- 2. Augmented Offgas Treatment System Explosive Gas Monitoring System
- a. Ilydrogen Monitor (4% Monitor) D** NA Q(4) M
- 3. Reactor Building Ventilation Monitoring System
- a. Noble Gas Activity Monitor D* Q R(3) Q(2) R(1) 5 b. Iodine Sampler Cartridge We NA NA NA 5 ,
[
- c. Particulate Sampler Filter W* 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* !! R(3) Q(2)
- b. Iodine Sampler W* NA NA NA
_u -u
- c. Particulate Sampler W* 'NA NA NA
- d. Effluent System Flow Rate Measuring Device D* NA R Q N
- e. Sampler Flow Rate Monitor D* NA R Q u
LeJ
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.
- Main 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 exists:
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instru=ent indicates a downscale failure.
- 4. Instrument controls not set in operate mode.
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instrument indicates a downscale failure.
- 4. Instrument controls not set in operate mode.
(3) The CHANNEL CALIBRATION shall 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 durinc normal operation.
(4) The CH/a'NEL CALIBRATION shall include the use oht least t h tandard gas samples containing a percentage of hydrogen to verity accuracy of the monitoring channel over its operating range.
FREQUENCY NOTATION:
S =
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D =
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
W =
At least once per 7 days.
M =
At least once per 31 days.
=
Q At least once per 92 days.
SA =
At least once per 184 days.
A =
At least once per year.
R =
At least once per 18 months.
S/U =
P-4or *, each reactor startup.
P =
h ete'dJ rior to each release. ,
NA = Not applicable./17 332
-216v-
LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENTS 3.20 (Cont'd) 4.20 (Cont'd) B. Liquid Effluents B. Liouid 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 cate-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.1.a.
- b. With the concentration of radio- c. The radioactivity content of each active material released from th~e 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-concentra on within the above ses in accordance with Table limit. 4.20.B.1. The results of pre-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.1.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.1. 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.
I947 333
-216w-
TABLE 4.20.B.1 RADIOACTIVE LIQUID WASTE SAMPLI;;G Al;D A'JALYSIS PROGRA51 Lower Limit Minimum of Detection Sampling Analysis Type of Activity (LLD) Liquid Release Type Freques.cy Frequency Analysis (pCi/ml)(1)
- 1. Batch Waste Release Tanks (5) P P Principal Gamma 5 x 10-7(2)
Each Batch Each Batch Emi t te rs ( 7) (8) 1-131 1 x 10-6 P 11(9) Dissolved and 1 x 10-5 One Batch /M Entrained Cases P M 11 - 3 1 x 10-5 Each Batch Composi t e (3) (9) Gross Alpha - 1 x 10-6 P Q(9) Sr-89 Sr-90 5 1 - Each Batch' Compos i t e (3) (9) 9 x 10-6 s T 2.A. Plant Service Water W W(9) Principal Gamma Effluent (6) Grab Sample 5 x 10-7(2) Emit te ra (7) (8) 2.B. Plant Continuous Discharge (6 Proportional 4) W(9) Principal Gamma Composite (4) 5 x 10-7(2) Emi t te rs (7) (8) I-131 1 x 10-6 M M(9) Dissolved and 1 x 10-5 Grab Sample Entrained Gases Proportiona :4) M(9) 11 - 3 1 x 10-5 Compo:ii t e (4) Cross Alpha iv 'n-7 N Proportiona 4) Q(9) Sr=Bo -90 .
-R
. Compon i t e (4 ) Fe-55 (l0) 8 71 x 10-6 U -
C=
NOTES FOR TA3LE 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 radie-nuclide mixtures, it may not be possible to measure radionuclides in concentrati:as near the LLD. Under these circumstances, the LLD may be increased inversely pre;:r-tionally to the magnitude of the gamma yield (i.e., 5 x 10-7/I, where I is the pt: ten abundance expressed as a decimal f raction), but in no case shall the LLD, as cal:u-lated in this manner for a specific radionuclide, be greater than 10% of the MFC value specified in 10 CFR 20, Appendix B, Table II, Column 2. (3) A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liqu'd waste discharged and in which the method of sampling emplcyed results in a specimen which is representative of the liquids released. (4) To be representative of the quantities and concentrations of radioactive materials in liquid effluents, 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 efflu-ent release. (5) A ba*ch relo7so is Pho A4sc', ee of lian4d vastos of a dim- ete "^'"-^ Prior .tc batch sha h isol en tnoroughly mixad. (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 concentra-tion in a sample exceeds 3 x 106 uCi/ml, or in the event the plant service water ef fluent monitor indicates the presence of an activity concentration greater than 3 x 106 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 exclu-sively the folicwing radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides. shall also be identified and reported. Nuclides which are belcw the LLD for the a..alyses should not be reported as being present at the LLD level. When unusual circumstances result in LLD's higher than required, the reasens shall be documented in the semiannual Radioactive Effluent Release Report. (8) If an isotopic analysis is unavailable, batch releases may be made for up to 14 days provided the gross beta /ganna concentration to the unres_tricted area i s_r 1 x 10-/ pc/mlw- Nsamples are anal %d wiien the instruGE"Etio'n is once agalii available.
-_m ~ _w (9) Analysis may be performed after release.
(10) P-J2 and Fe-55 sampling requirements may be deleted upon completion of a comprehen-sive sampling program which shows that thcse isotopes do not constitute a significant, dose at the unrestricted area boundary. The Commission shall be notified of such a j deletion bv inclusion in the Monthly Operation Report.
%-- d -
)
A FREQUENCY NOTATION: S = At least once per 12 hours. A = At least once per year. D = At least once per 24 hours. R = At least once per 18 months. W = At least once per 7 days. S/U = Prior to each reactor startup. M = At least once per 31 days. P = prior to each release. Q = At least once per 92 days. NA = Not applicable. SA = At least once per 184 days. 1947 335
-216y-
ass $ $CLAs
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n- i a o> I
~*
t ., . . - . , ,
\ -
-. .. ., , ~= ~ ~'" ~ ~'
- . \-.
4 *q
\ l \ .
I .= 1 f
\e. %,
\.
i l; . . ; ,,, / .:.1:= ;~.:; - 0 I* 1 J *:.U.".'.***.'.".'.., g
- !, .l y . ,.
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,, I .,
;x i* "
s/c +,
\ "
i\ O. %- i ,, v . I
' 4 % #y f :=',=' :::--
)
- -'= :-~:: 1.
' - - - - x -
, , c ,
t \ ( ; N. .*- 'g
*%, \
4 - , ~ , . - - .
*'\
; j 1
-Q .; y n , e -
e
\
\ *.,%. -
g .i .. I .l N.% -
) ..
a
\
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- q
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' . N -
=~-. : ;l ke .i i
.I :s i
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( m l1 . . . . .
-l- . _ _ . . _..
j - ... - l 0 P 3 A 46 e e. Figure 4.20.B.2 Exclusion Area Eoundary Caseous and L id Effluents }hk7 33b
- 216z-
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS
- 3. 20.B (Cont 'd) 4.20.B (Cont'd)
- 2. Liquid Dose
- 2. Liquid Dose
- a. The dose or dose commitment to an a.
individual from radioactive mate- Dose Calculation - Cumulative dose co g ,tions from liquid efflu-rials in liquid effluents released to unrestricted areas (see Figure ents shal' be determined in accord-
- 4. 20.B . 2) shall not exceed 1.5 ance witn the Offsite Dose Asses-mrem to the total body or 5 mrem sment Manual (ODAM) at least once per 31 days, to any organ during any calendar quarter,
- b. Doses due to liquid releases to unrestricted areas shall be pro-
- b. With the calculated dose from the jected at least once per 31 days.
release of radioactive materials in liquid effluents exceeding the
- c. The appropriate portions of the above limit, prepare and submit to the Commission within 30 days, pur- liquid radwaste system sM 11 he suant to Specification 6.7.2.B, a domonsr* ted OPERABLE by operating Special Report which identifies the he equioment at least five min-cause(s) for exceeding the limit (s)
,itos (E leas t-unce per 97. cay.
unless the liquid radwaste system and defines the corrective actions has been utiliaed to process radic-to be taken. active liquid effluents during
- 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-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 OPEPMLE 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.
)g4[ }}[
-216al-
LIMITI!;G CO::DITIO:; FOR OPERATIO!; SURVEILLA!;CE REQUIRE!'E!iTS 3.20.B (Cont'd) 4.20.B (Cont'd)
- 3. Co A nn -1r a - :w a Ta . an 3. Cond -nara crora-, Tank anc
("U"utside Tempora ry Tanks outside Temporary Tanks
- a. The quantity of radioactive mate- a. The quantity of radioactive c2te-rial contained in the Condensate rial co tainad in each o. the Storage Tank shall ba limited to 'above listed tanks s ia11 e cc-
< 25 curies, lid 2a the vatsice te m
~'
aea by analyzing a repre-ren.gorary tanks shall l lamlimited sentative sample of the tank's to 10 cu g excluding tritium contents at least once per 7 and cissolved or entrained nobel days when radioactive materials gases, are being added to the tank.
- b. With the qua tit of radioactive m- , ri ,1 i- any or ne above listed tanks exceeding the re-spective limit Ammediately sus-pend all additions of radioact!.ve material to the tank and within 48 hours reduce the tank contents to vi hin the limit.
- c. The provisfons of Definition J are not applicable.
1947 338
-216a2-
LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENTS 3.20 (Cont'd) 4.20 (Cont'd) C. Gaseous Effluents C. Caseous Effluents Applicability: At all times. Specificatiou:
- 1. Total Dose 1. Tots' Dose
- a. The concentration of radicactive a. The release rate of radioactice 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 Pa rt 20.106.
- b. A radioactive noble gas ef fluant
- b. With the concentration exceeding monitor shall cause automatic the limit in 3.20.C.l.a, decrease alarm when the concentration ex-the release rate to comply with the ceeds the monitor alarm setpcint, 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()
applicable.
- 2. Noble Cases 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 detar-shall not exceed 5 mrad f rom gamma mined in accordance with the radiation and 10 mrad from beta method in the ODAM at least c:ce 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 limite, 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-it(s) and defines the corrective actions to be taken.
- c. The provisions of Definition J are not applicable.
-216a3-I947 339
TABLE 4.20.C.1 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum of Detection Samp18.ng 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. t.ugmented Radwaste Sample Bldg Vent
- 4. Turbine Bldg Vent Q Q 11 - 3 1 x 10-6 (Gaseous) Grab Sample B. All Release Types as Continuous (6) W(4) I-131 1 x 10-12 tb Listed in A Abc,ve, Charcoal I-133 1 x 10-10 5 6 Radwaste Bldg Vent Sample (Iodine 6 Particulate)
Continuous (6) W(4) Principal Gamma 1 x 10-11(2) Particulate Emitters (7) Sample (I 131, Others) Continuous (6) M Gross Alpha 1 x 10-II Particulate Sample (S) Continuous (6) Q Sr-89, Sr-90 1 x 10" Composite Particulate _ Sample (8) 4 43" Continuous (6) Noble Gas Gross Noble Gases S x 10-6
-J Monitor Beta and Gamma (9) w CD
NOTES FOR TABLE 4.20.C.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., 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 vill be analyzed for alpha each 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, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141, and Cc-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reportec. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the , analyses should not be repor:ed as being present at the LLD level for that nuclide. When unusaal circumstances result in LLD's higher than required, the reasons shall be documented in the semi-annual ef fluent 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 Tcble 4.20.C.1 or using reference sources. FREQUENCY NOTATION: S = At least once per 12 hours. D = At least once per 24 hours. W = At least cnce 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 = Pri
- each reactor startup.
P = omolet rior to each release. NA = Not app 1 cable. _
-216 a 5- )hk7 b41
LIMITING CONDITION FOR OPERATION SURV2ILLWCE 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.
diciodines, radioactive materials Dose Calculations - Cumulative 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 mcem during any calendar quarter.
- b. Uith the calculated dose from the release of radionuclides, radio-active materials in particulate form, 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 Spea.ification 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. Gareous 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. Theacoropriate([(({}
radioactive materials in gaseous shall be wastes prior to their discharge ~=f y operat-when th2 proj ected gaseous ef flu-five_ minutes at ent air doses due to gaseous efflu- teast unce per 92 cays unless ent releases to unrestricted areas the appropriate system has been (see Figure 4.20.B.2) when averaged utilized to process radioactive over 31 days vauld exceed 0.2 mrad gaseous effluents during the for gamma radiation and 0.4 mrad previous 92 days. for beta radiation. The ventila-tion exhaust treatment system shall be operated to reduce the radio-active meterials in gaseous waste prior to their discharge when the projected gaseous ef fluent doses due to gaseous effluent releases to unrestricted areas (see Figure , 4.20.B.2) when averaged over 31 'l ' days would exceed 0.3 mrem to any 8 1OA/ I /7 J40 organ.
-216a6-
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 1 3.20.C (Cont'd) 4.20.C (Cont'd)
- b. With gaseous wastes being dis-charged for more than 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) Summary description of ac-tion (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 < 4% by volume. recombiners shall be determined by continuously monitoring the
- b. With the concentration of hydrogen _
waste gases in the main condenser or 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 concentration 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 Ej ector 6. Air Ejector
- a. The gross radioactivity (beta and/ a.
or gamma) rate of noble gases meas-The gross radioactivity (beta and/ or gamma) rate of noble gases ured at the main condenser air ejec- from the main condenser air ejec-tor shall be limited to < (100 pCi/ tor shall be determined at the sec/MWt). following frequencies by perform-
- b. With the gross radioactivity (beta in8 an is t pic analysis of a and/or gamma) rate of neble gases r presentative sample of gases at the main condenser air ejector t ken 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.
)hk7 -)43
-216a7-
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
, from the primary coolant.
- b. The radioactivity rate of noble gases at or near the outlet of I
the main condenser air ejector shall be continuously monitored
- 7. Containment in accordance with Table 3.20.A.2)
- a. Whenever the primary containment is vented / purged, it shall be vented / purged through the Standby Gas Treatment System. With this specification not satisfied, sus-pend all venting /purgine of the
'ne innene uis specificati ces not ap ly to Normal Ventila-tion.
- b. The provisions of Definition J are not applicable. The reporting provisions of Specification 6.7.2.B.2 are not applicable.
D. Effluent Dose Liould/ Gaseous D. Effluent Dose Liquid / Gaseous Applicability: At all times. l. Dose Calculations - The cumu-Specification: lative dose to an individual contributed by radioactive ma-1. terial in gaseous and liquid The dose or dose commitment to a effluents shall be calculated at real individual f rom 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 < 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 iOA7 i/7I 3A74 Special Report to the Commission pursuant to Specification 6. 7. 3.B -216a8-(Cont'd)
LIMITING CO'iDITION FOR OPERATION SURVEILLANCE REOUIREMENTS 3.20.D (Cont 'd) 4.20 (Cont' d) 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 Report shall include an analysis which demonstra*es that radiation ev,osures t any raal individual w1*hin,a atte cadius including all ef fluentMaafi'and direct radiation) are less than the 40 CFR Part 190 Standard. Other-wise, 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 Naste E. Solid Radioactive Waste Applicability: During solid radwaste 1. Operating parameters and limits processing. for the solidification of radio-Specification: active waste were established dur-ing preparational testing of the system. Radioactive waste solid-
- 1. The appropriate equipment of the ification shall be performed in solid r'dwo te system shall be oper- accordance with establ shed para-sen in accorcance w th te Process me*o- and limits and in accord-Control Program o so ial&y ano ance with the rocess Contro' pacsage raatoacmive waste and meet Program in acoltion, every 10th the requirements of 10 CFR Part 20 oaten of dewatered waste will be and 10 CFR Part 71 prior to shipment sampled prior to solidification of radioactive wastes from the site. and analyzed for pH.
- 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 6.7.3.3, 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 rial Release Report in Specifi-equipment, cation 6.7.1.F shall include the following information for each
- b. Cause of insperability, type of solid waste' shipped off-site during the report period:
1947 345
-216a9-
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.20 (Cont'd) 4.20 (Cont'd)
- c. Action (s) taken to restore the inop- ]
erable equipment to operable status, j a. Container burial volume,
- d. Identification and description of b. Total curie quantity (determined alternate equipment, if any, that by measurement or estimate),
may be used to satisfy the above requitement y . Principal gamma radionuclides (determined by measurement or
- e. Action taken to prevent recurrence. es tima t e) ,
- 3. The provisions of Definition J are d. Type of waste, not applicable.
- e. Type of container,
- f. Solidification agent.
F. Monitoring Program F. Monitoring Program Applicability: At all times.
- 1. Radiological environmental samples Specification:
shall be collected and -nalv ed as specified i Table J.20.F.1.
- 1. As 7 minimum the radiological envi-ronmental monitor ng progran shall 2. A land use census shall be con-be comiucteA as specifiec in Tablo^
ducted annually and shall iden-3.20.F.1. .inalytical techniques tify the location of the nearest um anall be such that the detec- g rden that is greater than 500 t tion capabilities in Table 3.20.F.2 square feet in area and that are achieved. yields edible leafy vegetables,
%- the location of the nearest milk.
- 2. In the event the radiological en- animal, nd the location of the vironmental monitoring program is nearest resident in each of the e dnctad as specified in 16 mete r 1 gical sectors within y T20TF prepare and submit three miles of the Station. The to tne Lommission in the Annual land use census shall be conduc-0; 2 rating Report the reasons for ted at least once per 12 months.
not conducting *he croPram in ac-cordance with rable 3.20.F.1 and 3. Tho results or_ sample analyses the plans for preventing a recur- perf rmed shall be summarized
- c. In tne Annual Radiological Environmental Report.
- 3. When the radioactivity in a sampled environmental medium, averaged over 4. The results of the land use cen-a calendar quarter, exceeds an ap- sus shall be included in the propriate value stated in Table Annual Radiological Environmental 6.7-2, prepare and submit to the ep rt.
Commission within 30 days from the end of the affected calendar quar-ter a Special Report which includes 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 saaple(s) is not at-t ributable to release f rom the gJ ~Kkh
-216a10-
LIMITING CONDITION FOR GPERATION SURVEILLANCE REOUIRE>1ENTS 3.20.F (Cont'd) 4.20 (Cont'd) Station, the Special Report is not required; instead the sample (s) result (s) shall be reported and explained in the Annual Radiologi-cal Environmental Report.
- 4. When environmental sampling medium is not available from a sam ling
'ne
- on designated ir table 3.20.F.1 the cause and ne loca-tion waere replacement samples were obtained shall be reported in the Annual Radiological Envi-ronmental Report.
- 5. In the event a location is identi-fled at which the calculated per-sonal dose associated with one or more exposure pathways exceeds the calculated dose associated with like pathways at a location where compline is cr" ducted as specified -
n Table _ 3. 20.F_. lb then the pa th- - ways having maximum exposure poten-tial at the newly identified loca-tion will be added to the radiol-alcal anni oring program and to Table 3.20.F. at the next SRAB meeting if samples are reasonably attainable at tae new location. - Like pathways monitored (sampled) - at a location, excluding the control station location's), having the lowest associated calculated per-sana doce nav be deleted from able 3.90.F. at the time the new pathway (s) and location are added.
- 6. A change in shall be described in tne Monthly Operat-ing Report within 90 days after the change was made effective.
- 7. The provisions of Definition J are not applicable.
1947 347
-216all-
TABLE 3.20.F.1 RADIOLOGICAL E!1VIR0!PctE!1TAL F1011ITORI!1G PROGRAF 1 Exposure Pathway !1 umber of Sampling and Type and Frequency Sample Stations" Collection Frequency of Analysis and/or Sample {
- l. Airborne At least 5 locations Continuous operation of sampler with Radioidine canister: Ana-
- a. Radiciodine and Partic- in accordance with the sample collection as required by dust lyze at least once per 7 days ulate Radiological Environ- loading but at least once per 7 days. for I-131.
mental lionitoring }!an- } I ual (REFBI). Particulate sample: Analyze for gross beta radioactivity
> 24 hours following filter
\ change. Perform gamma iso-l topicb analysis on each sample '
in which gross beta activity is >10 times the yearly mean of control sam les. Perform gamma isotopic analysis on
' composite (by location) sam-d ple at least once per 92 days.
T
)
C'
- p. Direct Radi- At least 8 locations Thermoluminescent Dosimeters (TLD) Gamma dose: At least once per ation in accordance with the exchange and read-out at Icast once 92 days.
RDDI, with 2 dosime- per 92 days. ters at each location.
- 3. Waterborne
- a. River Water At least 2 locations Collect a four (4) liter grab sample Gamma isotopicb analysis of I in accordance with at least once per 31 days. eacn sample. Composite grab h the RDDI. sample for tritium analysis at I
least once per 92 days.
- b. Ground Water At least I location Collect a four (4) liter grab sample Gamma isotopic b and trititm at least once per 92 days. analysis of each sample.
"j . in accordance with the RDDi.
- c. Sediment At least I location Two (2) times a year, once in the Gamma isotopic b analysis of from Shore- in accordance with spring and once in the fall. each sample, u line the RDIM.
CD
TABLE 3.20.F.1 (CONTINUED) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Sampling and Type and Frequency and/or Sample Sample Stations a Collection Frequency of Analysis
- 4. Ingestion
- a. Milk At least 4 locations At least once per 15 days during Peak Gamma isotopicb and I-131 l in accordance with Pasture Periodc; at least once per analysis of each sample, the REMM. 31 days at other times.
- b. Fish At least 2 locations Two times per year (once in the Gamma isotopicb analysis on in accordance with summer and once in the fall). edible portions.
I the REMM. Attempt to include the following:
- 1. Bottom feedir.g species
- 2. Middle-Top feeding species
, c. Food Prod- At least 3 locations At time of harvest. Sample one of Gamma isotopicb analysis on l Z ucts (Vege- in accordance with the following classes of food products edible portion. p { tables) the REMM. at each location. C 8
- 1. Flowers & fruitsd
- 2. Tubers d I 3. Roots d At least I location At time of harvest. One sample of I-131 analysis, in accordance with broad-leafd vegetation.
the REMM.
.c=-
N u
.5:n-
NOTES FOR TABLE 3.20.F.1
- a. Sample station locations are shown on Figure 1.F.1 of the Radiological Envi-ronmental Monitoring Manual (RDD1) maintained by the Environmental Af f airs Division of the Power Operations Group.
- b. Ge(L1) gamma isotopic analysis refers to high resolution Ge(L1) gamma spectrum analysis as follows: the sample is scanned for gamma-ray activity. If no activity is found for a selected nuclide, the detection sensitivity for that nuclide will be calculated using the counting time, detector efficiency, gamme energy, geometry, and detector background appropriate to the particular sample in question. The following nineteen (19) nuclides shall be analyzed for and routinely reported:
Be-7 Ru-103 Ce-144 K-40 Ru-106 Ra-226 Mn-54 I-131 Th-228 Fe-59 Cs-134 Co-58 Cs-137 ) Co-60 Zn-65 BaLa-140 f Ce-141 Zr-95 Nb-95 Any nuclide detected, having a concentration greater than the LLD shall be reported quantitatively whether or not it is one of the above 19 nuclides.
- c. Peak Pasture Period is June 1 through September 30 of each year.
- d. Vegetables are classified as follows:
- Flowers and fruits: Artichoke, broccoli, caulificwer, corn, cucumber; egg-plant, okra, pepper, pumpkin, squash, and tomato.
N - Tubers: Potato. I
- Roots: Beet, carrot, parsnip, radish, rutabaga, sweet potato, and turnip.
- Leaves (broad leaf): Cabbage, lettuce, spinach.
[947 350
-216a14-
TABLE 3.20.F.2 DETECTION CAPABILITIES FOR ENVIRONMENTet SAMPLE ANALYSIS Lower Limit of Detection (LLD)a Airborne Particulate Water or Gas Fish Milk Food Products Sediment Anaysis (pCi/1) (pCi/m3) (pCi/kg, wet) (pCi/1) (pC1/kg, wet) (pCi/kg, dry) gross beta 4 1 x 10-2 3 1I 2000 54 Mn 15 130 59 Fe 30 260 58,60 Co 15 130 65 2n 30 260 a, 95 Zr 30 I 95 Nb 15 131 7 ib 7 x 10-2 1 60 l134 Cs 15 5x 10-2 130 15 60 150 137 Cs 18 6x 10-2 150 18 80 180 s 140 Ba 60 60 4
- l l4O La y 15 15 l
u Note: This list does not mean that only these nuclides are to be detected and reported. Other peaks which are { measurable and identifiable, together with the above nuclides, shall also be identified and reported.
NOTES FOR TABLE 3.20.F.2
- a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falselv concluding that a blank observation represents a "real" signal.
For a particular measurement system (which may include radio-chemical separation): i g, 4.66 s b E.V . 2.22 . Y . exp(-AAt) Where I LLD is the lower limit of detection as defined above (as pCi per unit mass or volume) f sb is the standard deviation of the background counting rate or of the j counting rate of a blank sample as appropriate (as counts per minute) ) l E is the counting ef ficiency (as counts per transformation) V is the sample size (in units of mass or volume) 2.22 is the number of transformation per minute per picoc".ie Y is the fractional radiochemical yield (when applicable) l A is the radioactive decay constant for the particular radionuclide At is the eltpsed time between sample coll _ction (or end of the sample collection period) and time of counting The value of sb used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unver-ified theoretically predicted variance. In calculating the LLD for a radionu-clide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radio-nuclides normally present in the samples (e.g., potassium-40 in milk samples). Analyses shall be performed in such a manner that the stated LLD's will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLD's unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.
- b. LLD for drinking water.
F947 352
-216a16-
LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENTS G. Interlaboratory Comparison Program C. Interlaboratory Comparison Program Applicability: Applicable at all times 1. A brief summary of results ob-to Radiological Environmental Monitoring tained as part of th Inta-lab-Program. oratory Compariso rogram saalg be included in the Annual Specification: Radiological Environmental Report, pursuant to Specification
- 1. Analyses shall be performed on 6.7.1.A.E.
radioactive materials supplied as part of an Interlaboratory Ccm-parison Program.
- 2. With analyses not being performed as required in Specification 3.20.G.1, :eport the corrective ac-tions taken to prevent a recurrence to the Commission in the Annual Radiological Enviro.. ental Report.
- 3. The provisions of Definitior J are not applicable.
{h b
~216a17-
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 afora the 14,it e- ified n 10 CFR Po-* 20.106 ie xceeded. vontrol v. the normal liquid discharge pathway is assured by station procedu es governing locked discharge valves and valve line-up verification. - ' 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 gaseous effluents. The location of this instrumentation is indicated by a Figure in the ODAM, a simplified flow diagram showing gaseous ef fluent treatment and monitoring equipment. The alarm / trip setpoints for these instruments shall be calculated in accordance with methods in the ODAM, which have been reviewed by NRC, to ensure that the alarm will occur prior to exceeding the limits of 10 CFR Part 20. The process monitoring instrumentation includes provisions for monitoring the concentra-tions of potentially explosive gas mixtures in the augmented of fgas 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 concentration of radioactive materials released in liquid waste ef fluents f rom 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 Xc-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2. Since service water is not a normal or expected source of significant radioactive release, routine sampling and monitoring for radioactivity is precautionary. An activity con-centration of 3 x 10-6 pCi/ml in service water effluent is diluted in the discharge canal to about 1.5% of the 10 CFR 20 Appendix B Table 2 Column 2 concentration with only one circulating water pump operating. During normal Station operation the dilution would be even greater. By monitoring service water effluent continuously for radio-activity and by confirmatory sampling weekly, reasonable assurance that its activity concentration can be kept to a small fraction of the 10 CFR Part 20.106 limit and within the Specification 3.20.B.2.a limit is provided. 1 1347 354
-216a18-
3.20 & 4.20 BASES (Cont'd) 3.20.B & 4.20.B LIOUID EFFLUENTS (Cont'd) 3.20.B.2 & 4.20.B.2 Licuid 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 specificaticas will also keep average releases of radioactive material in effluents at small per-centages of the limits specified in 10 CFR Part 22.106. Surveillance Requirements provide for the measurement of releases and calculation of doses to verify compliance with the Specifications. Action statements in these Specifications implement the requirements of 10 CFR Part 50.36(c)(2) and 10 CFR Part 50, Appendix I, Section IV.A in the event an LCO is not met. 10 CFR Part 50 contains two distinctly separate statements of requirements pertaining to effluents from nuclear power reactors. The first concerns a description of equipment to maintain control over radioactive materials in effluents, determination of design objectives, and means to be employed to keep radioactivity in effluents ALARA. This requirement is stated in Part 50, Section 34a and Appendix I, Section II. Appendix I, Section III stipulates that conformance with the guidance on design objectives be demonstrated by calculations (since demonstration is expected to be prospective). The other is a requirement for developing limiting conditions for operation in technical specifications. It is stated in 10 CFR Part 50, Section 36a and Appendix I, Section IV. Both the intent of the Commission and the requirement are clearly stated in the opinion of the Commission;l relevant paragraphs from that document follow: Section 50.36a(b) of 10 CFR Part 50 provides that licensees shall be guided by certain considerations in establishing and implementing operating procedures speci-fied in technical specifications which take into account the need for operating flexibility and at the same time ensure that the licensee will exert his best efforts to keep levels of radioactive materials in effluents as low as practicable. The Appendix I that we adopt provides more specific guidance to licensees in this respect. A. The Rule Section IV of Appendix I specifies act!'n levels for the licensee. If, for any individual light-water-cooled nuclear powet reactor, the quantity of radioactive material actually released in effluents to unrestricted areas during any calendar quarter is such as to cause radiation exposure, calculated on the same basis as the design-objective exposure, which would exceed one-half the annual design-objective exposure, the licensee shall make an investigation to identify the causes of these high release rates, define and initiate a program of action to correct the situation, and report these actions to the Commission within 30 days of the end of the calendar quarter. The conclusion of the NRC Staff in the Appendix I Rulemaking Hearing 2 agrees with that of the Commission. The Staf f 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) 1947 355
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3.20 & 4.20 BASES (Cont'd) 3.20.B & 4.20.B LIOUID 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 exceeded, is stated in Appendix I, Section IV.A and in the Opinion of the Commission.3 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 f rom 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' or by the Commission in its Opinion 5 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 Specification 3.20.B.2 and 4.20.B.2 there is reasonable assurance that Station operation will not cause a radionuclide concentration in public drinking water taken from the River that exceeds the standard for anthropogenic radioactivity in community drinking water. 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 from 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 1 a 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 H3) 0.7 total 4.35 curies (Cont'd) i947 356
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3.20 & 4.20 BASES (Cont'd) 3.20.B & 4.20.B LIOUID EFFLUENTS (Cont'd) 3.20.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. The volume of liquid waste processed and intended for discharge 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 Cuide 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 = S114,000 = $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 S 88.000 1 curie x 106 uCi C = 1.dE+6 gal x 3785 m1 x x 1 man-rem man- rem curie $1,000 gal C = 0.013 pCi/ml (Cont'd) 1947 357
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3.20 & 4.20 BASES (Cont'd) 3.20.B & 4.20.B LIQUID EFFLUENTS (Cont'd) 3.20.B.2 & 4.20.3.2 Liquid Dose (Cont'd) (2) the activity concentration below which evcporation is not cost-beneficial is C = $114,000 x 1 curie x 106 uCi x 1 man-rem 1.2E+6 gal x 3785 m1 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. 5NRC Commissioners, op. cit. 6 Demonstration of Cocpliance with 10 CFR 50 Appendix I, Revision 1 and Supplement 2, Nebraska Public Power District, Cooper Nuclear Station, January 9, 1978.
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3.20 & 4.20 BASES (Cont'd) 3.20.3 & 4.20.B LIOUID EFFLUENTS (Cont'd) 3.20.B.3 & 4.20.B.3 Condensate Storage Tank and outside Temporary Tanks WW 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' cen-tents, the resultino dose commitment to an individual in an unrestricted area will not axe ad m N incAuded N specirica M those outd5or - anks tnat are not surrounded by liners, dikes, or walls capable of holding the tanks contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system. 3.20.C & 4.20.C GASEOUS EFFLUENTS 3.20.C.1 & 4.20.C.1 Total Dose Specification 3.20.C.1.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. _W ith prompt action to reduce the radioactive noble gas concentration in effluent following alarm initiation, it can be maintained at a small 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) arem/ year to the total body or to _< (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 Gases 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 ef fluent 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 en 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 Specificationc 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.B.2. . j947 359
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3.20 & 4.20 BASES (Cont'd) 3.20.C & 4.20.C GASEOUS EFFLUENTS (Cont'd) 3.20.C.4 & 4.20.C.4 Gaseous Radwaste System The OPERABILITY of the gaseous radwaste treatment system and the ventilation exhaust t reatment systems ensures that the systems will be available for use whenever gaseous ef fluents 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 flamnability 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 & 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 & 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 / CASE 0US 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 o 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. 1947 360
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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 measura-able concentrations of radioactive materials and levels of radiatian are not higher than expected on the basis of the effluent measurements and modeling of the environ-mental exposure pathways.
?
The environmental monitoring program described in Table 3.20.F.1 is the minimum pro-gram which will be maintaincd. The Radiological Environmental Monitoring Manual (REMM) is an internal control document which describes in detail the actual mon-itoring program which is performed to ensure compliance with the specified minimum p rog ram. Control of the radiological environmental monitoring progrm m including the R EFS!, rests with the Environmental Affiars Division of the Power Operattsns and not th rganizatio - 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-atingwindividual doses from principal exposure pathways.- M - The need to adjust the program to , current conditions and tc assure that tde 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 20% of the garden was usec for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vegetation 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.
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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 a'nd 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, surveillance, test and radiation control procedures and the emergency and security plans), regulations involving nuclear safety and operating license provisiens; 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 performed 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 responsibilit" 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 fra Protection Program, Radiological Environ-mental Monitorin rogram Offsite P:se Calculation Manual and their implementing procecutes at least once every 24 months. 1948 002
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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 systems that could have an effect on nuclear safety. 6.3.2 Written integrated and system procedures and instructions including applicable check of f 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 saf ety. B. Actions to be taken to correct specific and forseen potential or actual malfunctions of saf ety 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, prevencive maintenance programs, and operating license requirements. A. Routine testing of Engineered Saf eguards and equipment as required by the f acility 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 pr^ventive maintenance of instrumentation that could affect the nuclear safety of the plant. E. Special testing of equipment for proposed charses 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. 1948 003
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6.7 Station Reporting Reouirements 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 Office 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 different 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 values 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
,ther co=mitments shall be included in this report.
- 2. Stattup 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 earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion af 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. T948 004L
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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 expcsures 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 pr'ocessing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be as-signed to specific major work functions.
- 2. A summary description of facility changes, tests or experi-ments in accordance with the requirements of 10CFR50.59(b) .
- 3. Pursuant to 6.6.2G, Design Fatigue Usage, a listing of the nutber of events identified in 6.6.2.G.2.b will be tabulated and compared to the design or allowed quantity of comparable 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 crach initiation.
- 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 f acility, 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 Monitorin Program or changes to the Offsite Dose Assessment Manual 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 shall be reported to the Commission pursuant to Specification 6.10 by descrip-tion in the Monthly Operating, Report for the period in which the change was made. 1948 005
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6.7.1 (Cont'd) E. Annual Radiological Environmental Report
- 1. Routine radiological environmental reports covertag the sur-veillance activities related to the Station operation during the previous calendar year shall be submitted to the NRC before May I l of each year.
- 2. The Annual Radiological Environmental Report sitall include the I following: )
- a. Summaries, interpretations, and an analysis of trends of results of the radiological environmental surveillance activities for the report period, including a conparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment.
- b. A sucmary of the results of the land use census required in Specification 4.20.F.2.
- c. Suc=arized and tabulated results in the format of Table 6.7-1 of analyses of samples required by the rsdiological environmental monitoring program, and taken during the report period. In the event that some results are not
\ available for inclusion with the rcport, tl.a report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as l
possible in a supplementary report.
- d. A summary description of the radiological envirormental monitoring program; a map of all sampling locations keyed to a table giving distances and directions from the reactor; and the results of participation in the Interlaboratory Comparison Program, required by Specification 3.20.G.
- e. A suc=ary of meteorological data collected during the year in the form of joint frequency distributions of wind speed, wind direction and atmospheric stability,
- f. An assessment of off-site radiation doses due to radioactive liquid and gaseous effluents released from the Station f during each calendar quarter of the year and during the I
year. The dose assessment shall be performed in accordance with methods compatible with NRC approved ccues and procedures. A A Y 1948 006
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TABLE 6.7-1 ENVIRONMENTAL RADIOLOGICAL MONITORING PROGRAM
SUMMARY
Name of Facility Cooper Nuclear Station Docket No. 50-298 Location of Facility Nemaha, Nebraska Reporting Period (County, State) Type & Lower Limit Ali indicator Control Medium of Pathways Total No. of Locations Location with liighest Annual Mean Locations No. of Sampled of Analyses Detection (l) Mean[](2) Name Mean[](2) Mean[](2) Reportable (Unit of Measurement) Perfo rmed (LLD) Range (2) Distance & Direction Ra nge (2) Range (2) Occurrences d> M T b CD C O N Table Notes: (1) Nominal Lower Limit of Detection (LLD) as defined in Definition K.A. (?) Menn and Ranp,e honed upon detectable mennuremenen only. Frnction of elet ectnhin monnu remain e n n t npucirteil inentlunn 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 1 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 Semiannua.' 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 Semiar.nual 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. 1948 008
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6.7.2.A (Con t ' d)
- 4. Reactivity anomalies, involving disagreement with the predicted value of reactivity balance under steady state conditions during power operation, greater than or equal to 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 more 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 procedu..! inadequacy which prevents or could prevent, by itself, the fult_llment of the functional require-ments of systems required to cope with accidents snalyzed 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 f rom natural or man-made events that, as a di"ect result of the event require plant shutdown, operation of sa',ety 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 '.n 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 tha t 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.
l948 009
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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 repos t form shall be supplemented, as needed, 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 fou- 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.
No +_ e : Sealed sources or calibration sources are not included under this item. Leakage of valve packing or gaskets within the limits for identified leakage set forth in technical specifications need not be reported under this item.
- 5. An unplanned of fsite release of 1) more than 1 curie of radio-active material in liquid ef fluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radiciodine in gaseous effluents. The report of an unplanned offsite release of radioactive ma terial 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 relea. .
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6.7.2.B (Cont'd)
- 6. Measured levels cadioactivity in an environmental sampling medium determine ;o 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 detected 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 ef fluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications 3.20.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 ef fluents; 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. l' '90 Days After Leak Rate Testing (1) Completion of Each Test.
Note: (1) Each integra ed leak rate test of the secondary containment shall be the sabject 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. I948 011
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TABLE 6. 7-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Reporting Levels c Water Fish Milk d Leaf Vegetation AirborneParticul<gte or Cases (pCi/m ) (pC1/Kg, Wet) (pCi/1) (pbi Analysis pCi/1 , H-3 :+ Mn-54 lE + 3 3E + 4 Fe-59 4E + 2 lE + 4 Co-58 lE + 3 3E + 4 Co-60 3E + 2 lE + 4 Zn-65 3E + 2 2E + 4 4 Zr-Nb-95 4E + U y- I-131 2 0.9 3 lE + 2 Cs-134 30 10 lE + 3 60 lE + 3 Cs-137 50 20 2E + 3 70 2E + 3 Ba-La-140 2E + 3E + 2
") For drinking water samples. This is the 40 CFR 141 valve. b-(b) Total for parent and daughter. A CD CD 4
6.8 PROCESS CONTROL PROGRAM (PCP) 1 6.8.1 The PCP shall be a manual detailing the program of sampling, analysis and formulation determination by which SOLIDIFICATION of radioactive waste froa liquid systems is assured consistent with Specification 3.20.E and the surveillance requirements of these Technical Specifications. The PCP shall be submitted to the Commission at the time of proposed Radio-logical Effluent Technical Specifications and shall be subject to review and approval by the Commission prior to implementation. 6.8.2 District Initiated Changes A. Shall be submitted to the Commission by inclusion in the Semiannual Radioactive Ef fluent Release Report for the period in which the chaage(s) was made effective and shall contain: I
- 1. Sufficiently detailed inforaation to totally support the rationale for the change without benefit of additional or supplemental l information;
- 2. A determination that the change did not reduce the overall confor:ance of the solidified waste product to existing criteria for solid wastes; and
- 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, 6.9 0FFSITE DOSE ASSESSMENT MANUAL (ODAM) 6.9.1 The ODAM shall describe the methodolor.v and parameters tc 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 LC0's contained in these Technical Specifications. 6.9.2 District Initiated Changes A. Shall be submitted to the Commission by inclusion in the honthly Operating Report pursuant to Specification 6.7.1.D within 90 days of the o.'.te 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 ODAM to be changed with each page numbered and provided with a signed approval and date box, together with appropriate analyses of evaluations justifying the change (s) .
- 2. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint dete rmina tions .
- 3. Documentation of the fact that the change has been reviewed and found acceptable by the SORC.
B. Shall become effective upon review and acceptance by the SORC.
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1948 013
6.10 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS , (LIQUI", CASEOUS, AND SOLID) 6.10.1 The radioactive waste treatment systems (liquid, gaseous, and solid) are thote systems described in the facility Safety Analysis Report and amend-ments thereto, which are used to mainta4.n that control over radioactive materials in gaseous and liquid effluents and in solid waste packaged for offsite shipmenc 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 solii) 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 that the change could be made (in accordance with 10 CFR 50.59).
- b. Suf fiulent detailed information to totally support the reason for the change without benefit of additional or supplemental information.
- c. A detailed description of the equipcent, 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.
1948 014
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6.lO.2.A (Cont'd)
- g. An estimate of the exposure to plant operating personnel as a result of the change.
- h. Documentation of the fact that the change was reviewed and found acceptable by SORC.
- 2. The change shall become ef f ective upon review and acceptance by SORC.
B. Commission Initiated Changes
- 1. The applicability of the change to the f acility shall be deter-mined by the SORC after consideration of the facility design.
- 2. The licensee shall provide the Cormission 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. .
to radioactive 6.10.3 Background and definition of what constitutes ' major changes' waste systems (liquid, gaseous, and solid). A. Background
- 1. 10 CFR Part 50, Section 50.34a(a) requires that each applicatioc 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 eachrapplicae 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 'l) a des-cription of the equipment and procedures for the cor. trol of gas-eous and liquid effluents and for the maintenance addfuse of equipment installed in radioactive waste systems and (2) a revised estimate of the information required in (b)(2) if the expectei releases and exposures differ significantly frc. the estimate submitted in the application for a constructior, permit.
I948 015
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6.10.3.A (Cont'd)
- 5. The Regulatory staff'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 maximum exposures .to an individual in the unrestricted area and estimated exposures to the general pop-ulation) which shows the capability of these systems to meet the appropria te regula tions.
- 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 effluer.ts.
B. Definition
" Major Changes" to radioactive waste systems (liquid, gaseous, and solid) shall include the following:
- 1. }hjor changes in process equipment, comnonents, 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 and/or quantities of effluents released or volumes of solid waste stored 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 ir system design which may invalidate 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 equipoent, use of mobile processing e quipment) .
1948 016
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