ML20027B502
| ML20027B502 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/17/1982 |
| From: | INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | |
| Shared Package | |
| ML17319B516 | List: |
| References | |
| NUDOCS 8209210140 | |
| Download: ML20027B502 (168) | |
Text
{{#Wiki_filter:: ATTACHMENT 1 TO AEP:NRC:0055F D. C. COOK - UNIT 1 8200210/%
a ATTACHMENT 1 TO AEP:NRC:0055F i LIST OF CHANGES TO APPENDIX A TECHNICAL SPECIFICATION D. C. COOK - UNIT 1 DPR-58 REMOVE PAGES INSERT PAGES I-A IV IV IX-A XV-A XVI XVI XVII XVII XVIII XVIII 1-4 1-4 1-6 to 1-7 1-6 to 1-9 3/4 3-57 to 3/4 3-68 3/4 11-1 to 3/4 11-17 3/4 12-1 to 3/4 12-10 B 3/4 3-5 B 3/4 11-1 to B 3/4 11-5 B 3/4 12-1 5-1 5-1 5-9 6-5 to 6-9 6-5 to 6-9 6-11 to 6-13 6-11 to 6-13 6-16 to 6-21 6-16 to 6-27
INDEX DEFINITION 3 SECTION PAGE 50URCE CHECK .................................................. 1 - 6 PROCESS CONTROL PROGRAM (PCP) ................................. 1 - 6 SOLIDIFICATION ................................................ 1 - 6 0FFSITE DOSE CALCULATION MANUAL (ODCM) ........................ 1 - 6 GASEOUS RADWASTE TREATMENT SYSTEM .............................. 1-6 VENTILATION EXHAUST TREATMENT SYSTEM .......................... 1 - 6 PURGE-PURGING ................................................. 1 - 6 VENTING ........................................................ 1-6 FEMB ER ( S ) 0F THE P UBLIC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 7 SITE BOUNDARY ................................................. 1 - 7 UNRESTRICTED AREA ............................................. 1 - 7 , D. C. COOK - UNIT 1 I-A i
? - INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 1 3/4.2 POWER DISTRIBUTION LIMITS l 3.4.2.1 Axial Flux Diff erence . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 2-1 3/4.2.2 Heat Flux Ho t Channel Factor . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-5 3/4.2.3 Nuclear Enthalpy Hot Channel Factor . . . . . . . . . . . . . . . . 3/4 2-9 ] 1 3/4.2.4 Quadrant Power Tilt Ratio . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-11 3/4.2.5 DNB P ar mae t er s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 2-13 3/4.2.6 Axial P ower Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-15 3/4.3 INSTRUMENTATION { 3/4.3.1 REACIOR TRIP SYSTEM INSTRUMENTATION . . . . . . . . . . . . . . . . 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM
- INSTRUMENTATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 3-15
! 3/4.3.3 MONITORING INSTRUMENTATION
- Radiation Monitoring Instrumentation . . . . . . . . . . . . . . . 3/4 3-35 Movable Incore De tectors . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 3-39 Seismic Instrumentation ............................ 3/4 3-40 Meteorological Instrumentation ..................... 3/4 3-43 Remote Shutdown Instrumentation .................... 3/4 3-46 Fire Detection Instrumentation ..................... 3/4 3-51 Radioactive Liquid Effluent Instrumentation ........ 3/4 3-57 Radioactive Gaseous Process and Effluent Monitoring Instrumentation ......................... 3/4 3-62 3/4.4 REACTOR COOLANT SYSTEM ,
3/4.4.1 REACTOR COOLANT LOOPS Normal Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-1 3/4.4.2 SAFETY VALVES - SHUTDOW.i . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 4-4 3/4.4.3 S AFETY VALVES. - OPERATING . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-5 , 3/4.4.4 PRESSURIZER ........................................ 3/4 4-6 3/4.4.5 ST EAM GENERATORS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 4-7 D. C. COOK - UNIT 1 IV
-s
v . LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration ......................................... 3/4 11-1 Dose .................................................. 3/4 11-4
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Liquid Waste Treatment ................ ............... 3/4 11-5 Liquiu Holdup Tanks ................................... 3/4 11-6 3/4.11.2 GASEQUS EFFLUENTS Dose Rate ............................................. 3/4 11-7 - Dose - Noble Gases ..................................... 3/4 11-10 Dose - Radiciodines, Radioactive Material in Particulate Form, and Radionuclides other than N ob l e G a s e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-11 Gaseous Radwaste Treatment ......... 4.................. 3/4 11-12 Explo s iv e G as .M ix tur e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-13 Gas Storage Tanks ...................................... 3/4 11-14 3/4.11.3 SOLID RADI0 ACTIVE WASTE ............................ 3/4 11-15 3/4.11.4 TOTAL DOSE ......................................... 3/4 11-17 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3 /4.12.1 MONITORING PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 12-1 3/4.12.2 LAND USE CENSUS ................................... . 3/4 12-9 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM ................. 3/4 12-10 D. C. CC0K - UNIT 1 IX-A
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INDEX BASES SECTION 3/4.11 RADIOACTIVE EFFLUENTS 3 /4 .11.1 LIQUID EFFLUENT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 / 4 11-1 3/4.11.2 GASEOUS EFFLUENTS .................................. B 3/4 11-2 3/4.11.3 SOLID RADI0 ACTIVE WASTE ............................ B 3/4 11-5 3/4.11.4 TOTAL DOSE ......................................... B 3/4 11-5 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3 /4.12.1 MONITORING PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 12-1 3 / 4.12 . 2 LAND USE CENSUS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 12-1 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM ................. B 3/4 12-1 D. C. COOK - UNIT 1 XV-A s
INDEX , DESIGN FEATURES SECTION PAGE 5.1 SITE
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Exclusio'n Area............................................... 5-1 Low Po p ul a ti o n Zon e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 Si te Boundary for Gaseous and Liquid Effluents. . . . . . . . . . . . . . 5-1 5.2 CONTAINMENT Co n fi gu r a t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 Design Pressure and Temperature.............................. 5-4 Pene'trations'................................................. 5-4 5.3 REACTOR CORE Fuel Assemblies.............................................. 5-4 Co n t ro l Ro d As s emb l i es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-4 1 5.4 REACTOR COOLANT SYSTEM
- Des i gn P re ssure and Tempe rature. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-4 Volume....................................................... 5-5 5.5 EMERGENCY CORE CCOLING SYSTEMS......'......................... 5-5
- 5.6. FUEL STORAGE Criticality................................................. 5-5 l 0rainage........................,........................... 5-6 i Capacity.................................................... 5-6 5.7 SEISMIC CLASSIFICATION...................................... 5-6
- c. . o. .w. e.ei . n .0' nm G 't C
- l *. e*.4.
L v" C n' *i I u" 'I . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .-6 4 1 5.9 COMPONENT CYCLE OR TRANSIENT LIMIT.......................... 5-6 D. C. COOK - UNIT 1 XVI S
*W
-IflDEX .
ADMIT!ISTRATIVE C0?iTROLS SECTION-- . PAGE 6.1 RESPONSIBILITY.............................................. 6-1
- ~
6.2 ORGAttIZATI0tt 0ffsite...........'.......................................... 6-1 Facility Staff.................. . ........................... 6-1
'6.3 FACILITY STAFF 00ALIFICATIONS............................... 6-5 6.4 TRAINING.................. ................................. 6-5 6.5 REVIEW AND AUDIT 6.5.1 PLANT NUCLEAR SAFETY REVIEW t0MMITTEE 6-6 Functi,on.................................................. l l
Composition............................................... 6-6 Alternates................................................ 6-6
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Meeting Frequency.......................................... 6-6 Quorum.................................................... 6-6 Re s p o n s i b i l i ti e s . . . . . . . . . . . . . . . . . . . . . . . . . .'. . . . . . . . . . . . . . . . 6-7 Authority................................................. 6-8 Records................................................... . 6-8
' 6.5.2 NUCLEAR SAFETY AND DESIGN REVIEW COMMITTEE'
' Function.................................................. 6-8 g
Composition............................................... 6-9 A l t e r n a t e s . . . . . . . . . . . . . . . . . . . . . . . . . 3. . . . . . . . . . . . . . . . . . . . . 6-9
- i. >
i D. C. COOK - UNIT 1 XVII J a W c- -- - - , - - -- - - . --- - - - m.-
ADMINISTRATIVE CONTROLS SECTION PAGE , { Consultants .............................................. 6-9 l Meeting Frequency ........................................ 6-9 QuorulR ................................................... 6-9 Review ................................................... 6-10 Audits ................................................... 6-11 Authority ................................................ 6-12 Records .................................................. 6-12 6.6 REPORTABLE OCCURRENCE ACTION ............................. 6-12 6.7 SAFETY LIMIT VIOLATION ................................... 6-13 6.8 PROCEDURES ............................................... 6-13 6.9 REPORTING REQUIREMENTS l 6.9.1 ROUTINE REPORTS AND REPORTABLE OCCURRENCES . . . . . . . . 6-14 6.9.2 SPECIAL REPORTS ................................... 6-22 l 6.10 RECORDS RETENTION ........................................ 6-23 6.11 RADIATION PROTECTION PROGRAM ............................. 6-24 6.12 HIGH RADIATION AREA ...................................... 6-24 6.13 ENVIRONMENTAL QUALIFICATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-24 6.14 PROCESS CONTROL PROGRAM .................................. 6-25 6.15 0FFSITE DOSE CALCULATION MANUAL .......................... 6-25 6.16 MAJOR CEANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS ........................................ 6-26 1 D. C. COOK - UNIT 1 XVIII
{. ~ [' OEFINITIONS PRESSURE SOUNOARY LEAXAGE 1.16 PRESSURE BOUNDARY LEAXAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall. CONTROLLED LEAXAGE 1.17 i CONTRCLLED LEAXAGE shall be that seal water flow supplied to the ' reactor coolant pump seals. j OUADRANT POWER TILT RATIO 1.18 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper-excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one execre detector inoperable, the remaining three detectors shall be used for computing the average. DOSE EQUIVALEIT I-131 1.19 00SE EQUIVALENT I-131 shall be that concentration of I-131 (uCi/ gram) which alone would produce the same thyroid dose as the quantity and iso-topic mixture of I-131, I-132. I- U3, 1-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distanca Factors for Fower and Test Reactor Sitas, or in NRC Regulatory Guide 1.109 Rev.1, October 1977. STAGGERED TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist of:
- a. A test schedule for n systems, subsystems, trains or other designated ccmponents obtained by dividing the specified test interval into n equal subintervals,
- b. The testing of one system, subsystem, trtin or other designated ccmponent at the beginning of each subinterval.
O. C. CC0K - UNIT 1 1-4
1.0 DEFINITIONS . SCURCE CHECK 1.27 A SOURCE CHECK shall be the qualitative assessment of Channel response when the Channel sensor is exposed to a radioactive scurce. PROCESS CONTRCL PRCGRAM (PCP) L.23 The PRCCESS CCNTROL PROGRAM shall contain the current formula, sampling, analysis, tests and determinations to be made to ensure that the precessing and packaging of solid radicactive wastes will be acccmplished in such a way as to assure ccmoliance with 10 CFR 20,10 CFR 71, Federal and State regulations and other requirements governing the shipment and disposal of radicactive waste. SOLIDIFICATION 1.29 SOLIDIFICATION shall be the conversion of radioactive liquid, resin and sludge wastes-frem liquid systems into a form that meets shipping and burial site requirements. OFFSITE 00SE CALCULATICN MANUAL (OCCM) 1.30 The OFFSITE 00SE CALCULATION MANUAL shall contain the methodology and para-meters used in the calculation of offsite doses due to radiactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints and the conduct of environmental radiological monitoring program. CASEOUS RACWASTE TREATMENT SYSTEM 1.31 A GASEGUS RA0 WASTE TREATMENT SYSTEM is any system designed and installed to reduce radicactive gaseous effluents by collecting primary - coolant system off-gases frem the primary system and providing for delay or holdup for the purpose of reducing the total radicactivity prior to ~~ release to the environment. VENTILATION EXHAUST TREATMENT SYSTEM 1.32 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gasecus radioicdine 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 icdines or carticulates frca the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect en ncble gas' effluents. Engineered Safety Feature (ESF) 2 atmoscheric cleanuo systems are not censidered to be VENTILATICN EXHAUST TREATMENT SYSTEM ccmconents.
PURGi-PURGING 1.33 ? URGE cr PURGING is the controlled process of discharging air er gas frem a confinement to maintain temcerature, pressure, humidity, concentratien or other operating condition, in such a manner that re-l placement air er gas is required to purify the cenfinement. ! VENTING I 1.34 VENTING is the centrolled pr: cess of discharging air or gas fr:m a confinement to maintain temcerature, orassure, humidity, c:ncentratien or other ccerating c nditien, in such a manner that reolacement air or l gas is not provided er recuired during VENTING. Vent, used in system
- names, does not imely a VENTING process.
j
- D. C. COOK - UNIT 1 1-6 i
OEFINITIONS MEMBER (S) 0F THE PUBLIC 1.35 MEMBER (S) 0F THE PUBLIC shall include all perso6s who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or its vendors. Also excluded from this category are persons who enter tha site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant. SITE BOUNDARY 1.36 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased or otherwise controlled by the licensee. UNRESTRICTED AREA 1.37 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY to which access is not controlled by the licensee for purposes of protection of individuals frem exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial, commercial, institutional and/or recreational purposes. e l l l l l l l D. C. COOK - UNIT 1 1-7 l
TABLE 1.1
'OPERATICNAL MCOES 4
REACTIVITY ". RATED AVERAGE COOLANT , MCDE CONDITION, X.ff THEEfAL PCWER* TEMPERATURE
- 1. PCWER OPERATION ,1 0.99 > 5% > 350*F
- 2. STARTUP 1 0.99 1:5 1 350*F
- 3. HOT STANDBY < 0.99 0 1 350*F
- 4. HOT SHUTDOWN < 0.99 0 350*F >Tavg ~
> 200*F
- 5. COLD SHUTDOWN < 0.99 0 < 200*F
- 6. REFUEL.ING** < 0.95 0 < 140*F
= Excluding decay heat.
** Reactor vessel head unbolted or removed and fuel in the vessel.
t 1 D. C. COOK - UNIT 1 1-8
TABLE 1.2 FREQUENCY NOTATION NOTATION FREQUENCY t S AT LEAST ONCE PER 12 HOLTS D AT LEAST ONCE PER 24 HOURS 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 R AT LEAST ONCE PER 549 DAYS S/U PRIOR TO EACH REACTOR START-UP P COMPLETED PRIOR TO EACH RELEASE N.A . NOT APPLICABLE D. C. COOK - UNIT 1 1-9 l
INSTRUMENTATION RADIOACTIVE LIOUID EFFLUENT INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.9 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are i not exceeded. APPLICABILITY: As shown in Table 3.3-12. ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel
. alarm / trip setpoint less conservative than a value which will ensure that the limits of 3.11.1.1 are met, without delay suspend the release .of radioactive liquid effluents monitored by the affected channel, reset, ) or declare the channel inoperable.
- b. With one or more radioactive liquid effluent monitoring instrumentation channels inoperable, take the applicable ACTION shown in Table 3.3-12.
- c. The provisions of Specifications 3.0.3, 3.0.4 and 6.9.1.13 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.9.1 The setpoints shall be determined in accordance with methodology 1 as described in the ODCM and shall be recorded. 4.3.3.9.2 Each radioactive liquid effluent monitoring instrumentation ' channel shall be demonstrated OPERABLE by performance of the CHANNEL l CHECK, SOURCE CHECX, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequenc'es shown in Table 4.3-8. D. C. COOK - UNIT 1 3/4 3-57 -
TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum i Channels ' Instrument Operable Apolicability Action
- 1. Gross Radioactivity Monitors Providing !
Automatic Release Termination
- a. Liquid Radwaste (1)
Effluent Line At times of release 23
- b. Steam Generator (1) At times of release Blowdown Line 24
- c. Steam Generator (1) At times of release 24 Blowdown Treatment Effluent
- 2. Gross Radioactivity -
Monitors Not Providing Automatic Release Termination
- a. Service Water System (1) At all t'imes 25 Effluent Line
- 3. Continuous Composite Sampler Flow Monitor
- a. Turbine Building Sump (1) At all times Effluent Line 25 1
l
- 4. Flow Rate Measurement Devices
- a. Liquid Radwaste Line (1)
- b. Discharge Pipes
- At times of release 26 (1) At all times NA
- c. Steam Generator Blowdown Treatment Effluent (1) At times of release 26
- Pump curves and valve settings may be utilized to estimate flow; in such cases, Action Statement 26 is not applicible.
D. C. COOK - UNIT 1 3/4 3-58
TABLE 3.3-12 (Continued) TABLE NOTATION Action 23 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may be resumed for up to 30 days, provided that prior to initiating a release:
- 1. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1 and;
- 2. At least two technically qualified members of the Facility Staff independently verify the discharge valving. Otherwise, suspend release of radioactive effluents via this pathway.
Action 24 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are analyzed for gross radioactiv gamma) at a limit of detection of at least 10-}tyaci/ (beta gram:or
- 1. At least once once per 8 hours when the specific activity of the secondary coolant is >0.01 pei/ gram DOSE EQUIVALENT I-131. .
- 2. At least once per 24 hours when the specific activity of the secondary coolant is <0.01 uci/ gram. DOSE EQUIVALENT _
I-1 31 . - Action 25 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that at least once per 8 hours, grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 10 -'uci/ml. Action 26 With the number of Channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours during actual releases. D. C. COOK - UNIT 1 3/4 3-59
TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL INSTRUMENT FUNCTIONAL CHECK CHECK CALIBRATION TEST
- 1. Gross Beta or Ganna Radioactivity Monitors providing alarm and automatic isolation
- a. Liquid Radwaste D* P R(3) Q(1)
Effluent Line S. Steam Generator D* M R(3) Q(1) Blowdown Effluent Line
- c. Steam Generator D* M R(3) Q(1)
Blowdown Treatment Effluent Line
- 2. Gross Beta or Gamma Radioactivity Monitors Providing Alarm But Not Providing Automatic Isolation
- a. Service Water D M System Effluent R(3) Q(2)
Line
- 3. Continuous Composite Samplers
- a. Turbine Building D N/A N/A N/A Sump Effluent Line
- 4. Flow Rate Monitors
- a. Liquid Radwaste D(4)* N/A R Effluent Q
- b. Steam Generator 0(4)* N/A N/A N/A Blowdown Treatment Line -
- During Releases. Via This Pathway D. C. COOK - UNIT 1 3/4 3-60
A TABLE 4.3-8 (Cont) TABLE NOTATION (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 control not set in operating 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 setpoint.
** 2. Circuit failure.
** 3 . Instrument indicates a downscale failure. -
** 4. Instrument controls not set in operating mode.
(3) The initial CHANNEL CALIBRATION shall be performed using one or -- more sources with traceability back to the National Bureau of Standards. These sources shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used. (4) CHANNEL CHECX shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days on which continuous, periodic or batch releases are made.
- Instrument indicates, but does not provide for automatic isolation.
** As equipment becomes operational.
D. C. COOK - UNIT 1 3/4 3-61 ,
^
Instrumentation l Radioactive Gaseous Process and Effluent Monitorino Instrumentation Limiting Condition for Operation __ i 3.3.3.10 The radioactive gaseous process and effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of 3.11.2.1 are not exceeded. Applicability: As shown in Table 3.3-13. . Action:
- a. With a radioactive gaseous process or effluent monitoring instrumentation channel alarm / trip setpoint less conservative than a value which will ensure that the limits of 3.11.2.1 are met, without delay suspend the release of radioactive gasecus effluents monitored by.the affected channel, reset, or declare the channel inoperable.
- b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13.
c. The provisions of Specification 3.0.3, 3.0.4 and G.9.1.13 are not applicable. .. Surveillance Requirements 4.3.3.10.1 The setpoints shall be determined in accordance with methodology as described in the ODCM and shall be recorded.* 2 4.3.3.10.2 Each radioactive gaseous process or effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-9. This surveillance requirement does not apply to the Waste Gas Holdup System Hydrogen and Oxygen Monitors, as their setpoints are not addressed in the 00CM. O. C. COOK - UNIT 1 3/4 3-62 ~
TABLE 3.3-13 Radioactive Gaseous Effluent Monitorino Instrumentation Minimu., Channels Instrument Operable Apolicability Action
- 1. Waste Gas Holdup System Explosive Gas Monitoring System
- a. Hydrogen Monitor (1) **
30
- b. Oxygen Monitor (2) **
29
- 2. Condenser Evacuation System
- a. Noble Gas Activity (1) **** 28 Monitor
- b. Flow Rate Monitor (1) **** 27
- 3. Auxiliary Building Ventilation System
- a. Noble Gas Activity (1) * ~
28 Monitor
- b. Iodine Sampler (1)
- 32 Cartridge
- c. Particulate (1)
- 32 Sampler Filter
- d. Effluent System (1)
- 27 Flow Rate Measuring Device ,
- e. Sampler Flow Rate (1)
- 27 Measuiing Device
- 4. Containment Purge System ***
- a. Noble Gas Activity (1) ****) 31 Monitor
- b. Particulate Sampler (1) ****I 32
- 5. Waste Gas Holdup System
- a. Noble Gas Activity (1) * ** *2 33 Monitor Providing Alarm and Tennination of Gas Decay Tank Releases
- 6. Gland Seal Exhaust
- a. Noble Gas Activity (1) ****
Monitor 28
- b. Flow Rate Monitor (1) ****
27 D. C. COOK - UNIT 1 3/4 3- 63
TABLE 3.3-13 (Cont) ~
** At all times Monitors sample containment atmosphere Automa tic not es). containm termination of purge on high containment activity. .
**** Ouring releases via this pathway !
I For 3.9.9 and purgeforpurposes other requirements. only, see Technical Specifications 3.3 3 1
. . , 3.4.6.1, 2
For gas decay tank releases only, see 3. for additional requirements . l l
-o D. C. COOK - UNIT 1 3/4 3-64
o' TABLE 3.3-13 (Cont) TABLE NOTATICN Action 27 With the number of channels OPEMBLE less than required by tne Minimum Channels OPEMBLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours. Acticn 2B With the number of channels OPEMBLE less than required by the i Minimum Channels OPEMBLE requirement, effluent releases via i this pathway may continue for up to 30 days provided grab samples are taken at least once per 8 hours and these samples are analy:ed
- for gross activity within 24 hours.
Action 29 With the number of channels OPEMBLE one less than required by the Minimum Channels OPEMBLE requirement, operation of this system may continue for up to 14 days. With 2 channels in-operable, operation of this system may continue for up to 14 days, provided grab samples are taken and analyzed every 12 hours. i Action 30 With the number of channels OPEMBLE less than required by the Minimum Channels OPEMBLE requirement, operation of this system may continue for up to 14 days, provided grab samples are taken and analyzed every 12 hours. Action 31 With the number of channels OPEMBLE less than recuired by the Minimum Channels OPEMBLE requirements, immediately suspend PURGING of radioactive effluents via this pathway. Action 32 With ttle number of channals OPEMBLE less than required by the . j Minimum Channels CPEMBLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samoles are continucusly collected with auxiliary sampling equipment as required in Table 4.11-2. Action 33 With the numcer of channels OPEMELE less than required by the Minimum Channels OPEMBLE requirement, the contents of the tank (s) may be released to the environment for uo to 14 days provided that prior to initiating the release:
- a. At least two independent samples of the .ank's contants are analy:ed and,
- b. At least two tachnically qualified memcers of the Facility Staff independently verify the release rate calculaticns and discharge valve lineups; otherwise, suspend release of radicactive effluents via this
;athway.
D. C. COOK - UNIT 1 3/4 3-65
TABLE 4.3-9 Radioactive Gaseous Effluent Monitorina Instrumentation Surveillance Recuirements Channel Channel Source Channel Functional Instrument Check Check Calibration Test
- 1. Waste Gas Holdup System Explosive Gas Monitoring System
- a. Hydrogen D* *
- NA Monitor Q(3) M
- b. Oxygen D **
- NA Monitor Q(4) M
- c. Oxygen D**
- NA Q(4) M Monitor (Alt.)
- 2. Condenser Evacuation System
- a. Noble Gas Activity D*
- M R(2) Q(1)
Monitor _.
- b. System Effluent D*
- NA R Flow Rate Q
- 3. Auxiliary Building Ventilation System
- a. Noble Gas Activity D* M Monitor R(2) Q(1)
- b. Iodine Samoler W* NA NA
- c. Particulate Sampler W* NA NA NA NA
- d. System Effluent D* NA R Q Flow Rate Measure-ment Device
- e. Sampler Flow Rate D* NA R Q Measurement Device
- 4. Containment Purge System
- a. Noble Gas Activity D** P Monitor R(2) Q(5)
- b. Particulate Sampler W** NA NA NA
- 5. Waste Gas Holdup System
- a. Noble Gas Activity P** P R(2) Q(5)
Monitor Providing Alarm & Termination of Gas Decay Tank Re, leases D. C. COOK - UNIT 1 3/4 3-66 m -
TABLE 4.3-9(Cont)
- 6. Gland Seal Exhaust
- a. Noble Gas-Activity D** M R(2)
Monitor Q(1)
- b. System Effluent D** NA R Q Flow Rate At all times During Release Via This Pathway
*** During Waste Gas Holdup System Operation (Treatment for Primary System Offgases)
I D. C. COOK - UNIT 1 3/4 3-67
TABLE 4.3-9 (Cont) TABLE NOTATION 1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists: 1.
**2. Instrument indicates measured levels above the alarm setpoint.
Circuit failure.
**3. Instrument indicates a downscale failure.
**4. Instrument controls not set in operate mode.
- 2) The initial CHANNEL CALIBRATION shall be performed using one or more sources with traceability back to the National Bureau of Standards. These sources shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used.
3) The CHANNEL CALIBRATION shall include the use of standard gas samples - containing a nominal:
- 1. One volume percent hydrogen, balance nitrogen, and
- 2. Four volume percent hydr. ogen, balance nitrogen.
- 4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
- 1. One volume percent oxygen, balance nitrogen, and o
- 2. Four volume percent oxygen, balance nitrogen.
5) 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.
**2. Instrument indicates measured levels above the alarm / trip setpoint.
Circuit failure.*
**3. Instrument indicates a downscale failure.*
**4. Instrument controls not set in operate mode.*
- Instrument indicates, but does not provide automatic isolation.
** As equipment becomes operational .
D. C. COOK - UNIT 1 3/4 3-68
B 3/4.11 RADI0 ACTIVE EFFLUENT 5 . 3/4.11.1 Liquid Effluents Concentra tion Limiting Condition for Operation 3.11.1.1 The concentration of radioactive material released at any time from the site to unrestricted areas (see Figure 5.1-3) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B. Table II,
- Column 2, for radionuclides other than dissolved or entrained noble gases.
For dissolged or entrained noble gases, the concentration shall be limited to 2 X 10- uci/ml total activity. Appl icability: At all times.
- Action:
With the concentration of radioactive materi,al released from the site ex-ceeding the above limits, without delay restore the concentration to within the above limits. , SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.11-1. j 4.11.1.1.2 The result of radioac'tive analysis shall be used in accordance with the methods of the ODCM to assure that all concentrations at the point of release are maintained within the limits of Specification 3.11.1.1. 1 f D. C. COOK - UNIT 1 3/4 11-1
l l TABLE 4.11-1 l Radioactive Liquid Waste Sampling and Analysis Program Lower Limit a Minimum Type Of of Detection Liquid Release Sampling Analysis Activity (LLD) Type Frequency Frequency Analysis uci/ml A. Batch Waste -
-7 Release Tanks c P P. Principal SX10 Gamma Emitters' c Each Batch Each Batch I-131 IX10 "
-5 P Dissolved 1X10 and Entrain-ed Gases (Gamma One Batch /M M Emitters)
P M H-3 1X10-5 Each Batch Composite b Gross Alpha IX10-/ P Sr-89,Sr-90 -8 Q b SX10 Each Batch Composite ' re-oo TXTO O B. Plant Continuous , Daily W- Principal 5X10-7 Releases d Gamma b Emitters' , Composite I-131 1X10 " n . Dissolved 1X10-5 Grab Sample and Entrain-ed Gases (Gamma Emitters) Daily M b H-3 1X10i Comoosite Gross Alpha IX10 '
~
Daily Q b Composite Sr-89.Sr-90SX163 - Fe-55 l1X10-6 ._ . D. C. COOK - UNIT 1 3/4 11-2
.- - .---,, - - - - - - - - - - - - + - _
1 TABLE 4.11-1 (Cont)
. TABLE NOTATION
- a. The lower limit of detection (LLD) is defined in Table Notation
- a. of Table 4.12-1 of Specification 4.12.1.1.
- b. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is re-presentative of the liquids released.
- c. A batc5 release is the discharge of liquid wastes of a discrete volume.
Prior '.a sampling for analysis, each batch shall be isolated and re-circulated to ensure thorough mixing.
- d. A continuous release is the discharge of liquid waste of a non-discrete volume; e.g. from a volume of system that has an input flow during the continuous release.
- e. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144. This list does not mean that only these nuclides are to be detected and reported.
Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. 5 4 D. C. COOK - UNIT 1 3/4 11-3 } s 1
RADIOACTIVE EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose commitment to an individual from radioactive material in liquid effluents released to unrestricted areas (see Figure 5.1-3) shall be limited:
- a. During any calendar quarter to 5,1.5 mrem to the total body and to g,5 mrem to any organ, and
- b. During any calendar year to 5,3 mrem to the total body and to 5,10 mrem to any organ.
APPLICABILITY: At all times. ACTION: a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and __ submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause (s) for exceed-ing the limit (s) and defines the corrective actions taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be within the above limits. This Special Report shall also include (1) the results of radiological analyses of the drinking water source, and (2) the radiological impacts on finished drinking water supplies with regard to the
, requirements of 40 CFR 141, Safe Drinking Water Act. (Applicable only if drinking water supply is taken from the receiving water body.)
- b. The provisions of Specification 3.0.3, 3.0.4 and 6.9.1.13 are not applicable.
SURVEILLANCE REOUIREMENTS 4.11.1.2 Dose Calculations. Cumulative dose contributions from liquid - effluents snall be determined in accordance with the Offsite Dose Calcu-lation Manual (00CM) at least once per 31 days. D. C. COOK - UNIT 1 3/4 11-4
Radioactive Effluents Licuid waste Treatment ' Limiting Condition For Ooeration 3.11.1.3 The liquid radwaste treatment system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the pro-jected doses due to the liquid effluent from the site (see Figure 5.1-3) when averaged over 31 days, would exceed 0.06 mrem to the t'otal body or 0.2 mrem to any organ. Acolicability: At all times. Action:
- a. With radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special '
Report which includes the following information: i
- 1. Identification of the inoperable' equip. ment or subsystems and the reason for inoperability,
- 2. Action (s) taken to restore the inoperable equipment to operable
. s ta tus , and
- 3. Summary description. of action (s) taken to prevent recurrence.
- b. The provisions of Specification 3.0.3, 3; 0.4 and 6. 9. l'.13 a're not. .
appl ica,ble. Surveillance Recuirements
! 4.11.1.3 Doses due to liquid releases to UNRESTRICTED AREAS shal.1 be pro-jected at least once per 31 days, in accordance with the 00CM, whenever liquid releases are being made without being processed by the liquid rad-waste treatment system. '
D. C. COOK - UNIT 1 3/4 11-5
Radioactive Effluents Liquid Holdup Tanks * . Limiting Condition For Operation 3.11.1.4 The quantity of radioactive material contained in each of the following tanks shall be limited to less than or equal to 10 curies, excluding tritium and ' dissolved or entrained noble gases,
- a. Outside temporary tanks.
Applicability: At all times. Action:
- a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, without delay suspend all addit.fons of
~
radioactive material to the tank and within 48 hours reduce the tank contents to within the limit.
- b. The provisions of Specification 3.0.3, 3.0.4 and 6.9J .-13 a re. not applicable. .
Surveillance Reauirements " 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents _at least once - per 7 days when radioactive materials are being added to the tank. , Tanks included in this Specification are those outdoor tanks that a are not surrounded by liners, dikes, or walls capable of holding '
^
the tank contents and that 'do not have tank overflows and sur-rounding area drains connected to the liquid radwaste treatment system.
~
t D. C. COOK - UNIT 1 3/4 11-6 H=
Radioactive Effluents
-3/4.11.2 Gaseous Effluents ,
Dose Rate Limiting Condition For Operation 3.11.2.1 The dose rate due to radioactive materials released in gaseous ! effluents from the site (See Figure 5.1-3) shall be limited to the - following: . a. For noble gases: 5_500 mrem /yr to the total body and < 3000 mrem /yr to the skin, and
'b. For all radiciodines and for all radioactive materials in particulate form and radionuclides (other than noble gases) with half-lives greater than 8 days: < 1500 mrem /yr to any organ.
4 ! Applicability: At all -times. Action: With the dose rate (s") exceeding the above limits, without delay decrease tne release rate to within the above limit (s). Surveillance Reauirements
~
4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the 0,DCM. 4
~
4.11.2.1.2 The dose rate due to radioactive materials, other than noble e gases, in gaseous effluents shall be determined to be within the above ' limits,in accordance with the methods and procedures of the ODCM by ob-
. taining representative samples and performing analyses in accordance with the sampl'ing and analysis program specified in Table 4.11-2.
e ,,9 = l i_ I it D. C. COOK - UNIT 1 3/4 11-7 T , l . . - -
TABLE 4.11-2 RADICACTIVE GASECUS WASTE SAMFLING AND ANALYS:S ?RCGRAft Minimum Ty;e of Loner Limi t Analysis of Cetecticn Gaseous Release Type Frequency Frequency Activity Analysis ( uci/ml )a l' P P Principal Gamma ' Each Tank Each
! a. 'Jaste Gas Storage Tank Grab Sample Tank Sitters e 1 X 10 #
I l
' P P Principal Gamma ,
Sitters e 1 X 10 -4 Each Purge Each
- b. Contaiment Purge Grab Sample b ourge b g,3
;)x79 %
W Mb
- c. Condenser EvaCJation Grab PFincipal Gat".ma System and Glana Seal Sample b Particulate .
l Samole Si tters e ' 1 X 10 t Exhaust
- i
, Mb l H-2 ' " O ~5 l Mb i Charecal I-131 Sampie l i X 10-12 Continuous" I,
l Noble Gas v-r'-e- l Noble Gases l1XION'
! e
' W , Charcoal I-131 1 X 10 -i2 .
- d. Auxiliary Building Ven;' Continucus d Sample C
W 1 Continuous d Particula te Principal Gamma I samri e Si tters e 1 Y 10-I,' fr M l Continucus d Canposite i Particulate G...s Alcha 1 I 10,;) famel e Continucus d " lC cs b H-3 1 X 10-5 l
~! I q !
d Centir.ucus Cemccsite Sr-89, Sr-9C f1X10-II Pam:f culata i Sample f 0 l Centinucus ! 1Dd.i's iNeoleGases l 1 X 10
- As ecui: ment :ec:mes :ceracicnal
- D. C. COOK - UNIT I 3/4 11-3 l , _ _ _ _ _ - - - - - - -
l TABLE 4.11-2 (cont) TABLE NOTATION
- a. The lower limit of detection (LLD) is defined in Table Notation a.
of Table 4.12-1 of Specification 4.12.1.1. i
- b. Analyses shall also be performed following any operational occurrence which has altered the mixture of radionuclides as indicated by RCS analysis. (ie., start-up.)
i
- c. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing. Analyses shall l also be performed at least once per 24 hours for 7 days following each shutdown, startup or similar operational occurrence which lead to significant increases or decreases in radiciodine in the Reactor Coolant System. When samples collected for 24 hours are analyzed, the corresponding LLD's may be increased by a factor of 10.
- d. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specification 3.11.2.1, 3.11.2.2, and 3.11.2.3.
- e. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133M, Xe-135 and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58 Co-60, Zn-65, Mo-99, Cs-134. Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these i' nuclides are to be detected and reported. Other peaks which are measureable and identifiable, together with the above nuclides, shall also be iden.tified and reported.
1
- 0. C. COOK - UNIT 1 3/4 11-9
==
- - + - 9 my m - ,-rmw- , -- - - -- --.y . --a. -
--9------~ g g
a . RADI0 ACTIVE EFFLUENTS DOSE, NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose in UNRESTRICTED AREAS due to noble gases released in gaseous effluents shall be limited to the following:
- a. During any calendar quarter, to 1 S mrad f'or gamma radiation and 5 01 mrad for beta radiation;
- b. During any calendar year, to < 10 mrad for gamma radiation and 5 02 mrad for beta radiation.
APPLICABILITY: At all times. ACTION:
- a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to '
Specification 6.9.2, a Special Report which identifies the - __ cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent re-leases will be within the above limits.
- b. The provisions of Specification 3.0.3, 3.0.4 and 6.9.1.13 are not applicable.
SURVEILLANCE REOUIREMENTS l 4.11.2.2 Dose Calculations Cumulative dose contributions for the total time period snali be determined in accordance with the Offsite Dose Calculation Manual (CDCM) at least once every 31 days. ? l I I D. C. COOK - UNIT 1 3/4 11-10 e s w se
- . - , . , - _ . - - . - , , ,.7 ,
.7-
. 1 i
RADIOACTIVE EFFLUENTS DOSE, RADIOIODINES, RADIOACTIVE MATERIAL IN PARTICULATE FORM, AND RADIONUCLIDES OTHER THAN NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to A MEMBER OF THE PUBLIC f rom radioiodines, radioactive materials in particulate form, and radionuclides other than noble gases with half-lives greater than 8 days in gaseous effluents released to unrestricted areas shall be limited to the following:
- a. During any calendar quarter to n 7.5 mrem te any organ;
- b. During any calendar year to 4 15 mrem to any organ.
APPLICABILITY: At all times. ACTION:
- a. With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides other than noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions taken to reduce the releases and the proposed corrective action to be taken to assure that subeequent releases will be within the above limits.
- b. The provisions of Specification 3.0.3, 3.0.4, and 6.9.1.13 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.3 DOSE CALCULATIONS Cumulative dose contributions for the total time period shall be determined in accordance with the ODCM at least once every 31 days. gD. C. COOK - UNIT 1 3/4 11-11
Radioactive Effluents Gaseous Radwaste Treatment Limitinc Condition For Ooeration 3.11.2.4 The gaseous radwaste treatment system ano the ventilation exhaust treatment system shall be used to reduce the radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases to unrestricted areas (See Figure 5.1.3) when averaged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation. The ventilation exhaust treatment system shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases to unrestricted areas (See Figure 5.1-3) when averaged over 31 days would exceed 0.3 mrem to any organ. Acolicability: At all times. Action: a, 'dith gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Ccmmission -witnin 30 days, pursuant to Specification 6.9.2, a Special Report wnien in-cludes the following information:
- 1. Identification of the inoperable equipment or subsystems and the reason for inoperability.
- 2. Action (s) taken to restore the inoperable equipment to operable status.
! b. i The provisions of Specification 3.0.3, 3.0.4 and 6.g.l.13 are not applicable. Surveillance Recuirements 4.11.2.4 Coses due to gaseous releases to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with tne COCM, whenever the gaseous waste treatment system or ventilation exhaust treatment system is not operational. O. C. COOK - UNIT 1 3/4 11-12
RADI0 ACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE
~
LIMITING CONDITION FOR 0PERATION 3.11.2.5 The concentration of oxygen in the waste gas holdup system shall be limited to < 2% by volume if the hydrogen in the system is > 4% by volume. APPLICABILITY: At all times. ACTION:
- a. With the concentration of oxygen in the waste gas holdup system
> 25 by vo.lume but 5 4% by volume and containing > 4% hydrogen, restore the concentration of oxygen to < 2% or reduce the hydrogen concentration to < 4% within 48 hours.
- b. With the concentration of oxygen in the waste gas holdup system or tank > 45 by volume and > 4% hydrogen by volume without delay suspend all additicns of waste gases to the system or tank and reduce the concentration of oxygen to < 2% or the concentration of hydrogen to
< 4% within 48 hours in the system or tank.
- c. The provisions of Specification 3.0.3, 3.0.4 and 6.9.1.13 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentration of oxygen in the waste gas holdup system shall be determined to within the above limits by continuously monitoring the waste gases in the waste gas holdup system with the oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.10. D. C. COOK - UNIT 1 3/4 11-13 m.
i RADIOACTIVE EFFLUENTS GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to 43,800 curies noble gas (considered as Xe-123). APPLICABILITY: At all times. ACTION:
- a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, without delay suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents to within the limit.
- b. The provisions of Specification 3.0.3, 3.0.4 and 6.9.1.13 are not applicable.
SURVEILLANCE REOUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas -- storage tank shall be determined to be within the above limit at least once per 4 days by analysis of the Reactor Coolant System noble gases. D. C. COOK - UNIT 1 3/4 11-14
. t RADI0 ACTIVE EFFLUENTS 3/4.11.3 SOLID RADI0 ACTIVE WASTE LIMITING CONDITION FOR OPERATION 3.11.3 The solid radwaste system shall be used as applicable in accordance with a PROCESS CONTROL PROGRAM for the SOLIDIFICATION and packaging of radioactive wastes to ensure meeting the requirements of 10 CFR Part 20 and of 10 CFR Part 71 prior to shipment of radioactive wastes from the site.
APPLICABILITY: At all times. ACTION:
- a. With the packaging requirements of 10 CFR Part 20 and/or 10 CFR Part 71 not satisfied, suspend shipments of defectively packaged solid radioactive wastes from the site.
- b. With the solid radwaste system inoperable for more than 31 days, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days
~ __
pursuant to Specification 6.9.2 a Special Report which includes the following information:
- 1. Identification of the inoperable equipment or subsystems and the reason for inoperability,
- 2. Action (s) taken to restore the inoperable equipment to operable status,
- 3. A description of the alternative used for SOLIDIFICATION and packaging of radioactive wastes, and
- 4. Summary description of action (s) taken to prevent a recurrence.
- c. The provisions of Specification 3.0.3, 3.0.4 and 6.9.1.13 are not applicable..
D. C. COOK - UNIT 1 3/4 11-15
SOLID RADIOACTIVE WASTE SURVEILLANCE REOUIREMENTS 4.11.3.1 The solid radwaste system shall be demonstrated operable at least once per 92 days by:
- a. Operating the solid radwaste system at least once in the previous 92 days in accordance with the PROCESS CONTROL PROGRAM, or b.
Verification of the existence of a valid contract for SOLIDIFICATION to be performed by a contractor in accordance with a PROCESS CONTROL PROGRAM. 4.11.3.2 THE PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive waste (e.g. filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions). a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and -- a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch . may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM. b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least 3 consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.14 to assure SOLIDIFICATION of subsequent batches of waste. D. C. COOK - UNIT 1 3/4 11-16
- f RADIOACTIVE ErFLUENTS 3/411.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.11.4 The dose or dose commitment to a real individual from all uranium fuel cycle sources is limited to < 25 mrem to the total body or any organ (except the thyroid, which is limited to < 75 mrem) over a period of 12 consecutive months.
APPLICABILITY: At all times. ACTION: a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifi-cation 3.11.1.2.a , 3.11.1.2.b, 3.11.2.2.a , 3.11.2.2.b, 3.11.2.3.a or 3.11.2.3.b, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Director, Nuclear Reactor Regulation, U.S. Nuclear Regulatory Comission, Washington, D. C. 20555, within 30 days, which defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of Specification 3.11.4. This Special Report shall include - an analysis which estimates the radiation exposure (dose) to a member of the public from uranium fuel cycle sources (including all effluent pathways and direct radiation) for a 12 consecutive month period that includes the release (s) covered by this report. If the estimated dose s exceeds the limits of Specification 3.11.4, and if the release conditio(n) resulting in violation of 40 CFR 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190 and including the specified infor-mation of 5190.11(b). Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the requirements for dose limitation of 10 CFR Part 20, as addressed in other sections of this Technical Specification.
- b. The provisions of Specification 3.0.3, 3.0.4 and 6.9.1.13 are not applicable.
SURVEILLANCE REOUIREMENTS 4.11.4 Dose Calculations Cumulative dose contributions from liquid and gaseous effluents snall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3 and with the ODCM. D. C. COOK - UNIT 1 3/4 11-17
i 3/4.12 RADIOLOGICAL Et1VIR0ftMEtlTAL tt0tlITORIflG LIMITING CONDITION FOR OPERATI0t1 3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12-1. APPLICABILITY: At all times. ACTI0ft: a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12-1, prepare and submit to the Commission in the Annual Radiological Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a reccurence. (Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, or to malfunction of automatic sampling equipment. If the latter, every effort shall be made to complete the corrective action prior to the end of the next sampling period.)
- b. __
With the level of radioactivity in an environmental sampling medium - at one or more of the locations specified in Table 3.12-1. exceeding the limits of Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Comission within 30 days from the end of the affected calendar quarter, a Special Report which includes an evaluation of any release conditions, environmental factors or other aspects which caused the limits of Table 3.12-2 to be exceeded. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event,
':he condition shall be reported and described in the Annual Radiological Environmental Operating Report. When more than one of the radio-nuclides in Table 3.12-P. are detected in the sampling medium, this report shall be submitted ,if:
concentration (1) concentration (2)
+
+...>l limit level (1) limit level (2)
When radionuclides other than those in Table 3.12-2 are detec'ted and are the result of plant effluents, this report shall be sub-mitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specification 3.11.1.2, 3.11.2.2 and 3.11. 2.3.
RADIOLOGICAL ENVIR0!!MEtlTAL MONITORING LIMITING CONDITION FdR OPERATION (C0flTIflVED}
- c. With milk or fresh leafy vegetable samples unavailable from any of the sample -locations required by Table 3.12-1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause of the unavaila-bility of samples and identifies locations for obtaining replace-I ment samples. The locations from which samples were unavailable may then be deleted from Table 3.12-1 provided the locations from which the replacement simples were obtained are added to the environmental monitoring program as replacement locations, if available.
- d. The provisions of Specification 3.0.3, 3.0.4 and 6.9.1.13 are not applicable.
SURVEILLANCE REOUIREMENTS 4.12.1 ' The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the locations given in the table 7.nd figures in the ODCM and shall be analyzed pursuant to the requirements of Tables 3.12-1 and 4.12-1. e e 9 D. C. COOK - UNIT 1 3/4 12-2
? TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sampling and Type of Frequenc and/or Samoles Samole Locations Collection Freauency of Anal is
- 1. Airborne
- a. Radiciodine Al-A6 (Site) Continuous operation Radiciodine canis and Particulates of sampler with Analyze NewrBuffalo, Sample Collection Weekly for I-131 South Bend, as required by i Dowagiac, and Dust Loading But At Particulate sampi l Coloma are Least once Per 7 Days Gross Beta Rad-l Background ioactivity follow Filter Changea, posite (by loca-tion) for gamma isotopic quarterl
- 2. Direct Radiation a) T1-T9 (Site) At least once per Gamma Dose. At 92 days Least Once Per b) New Buffalo 92 days .
South Bend ._- Dowagiac Coloma c) 10 TLD tionitor Locations in the Five Mile Radius
- 3. Waterborne ,
- a. Surface L1 , L2 , L3 Composite Sample Gamma Isotopic Over One-Month Period Analysis monthly .
Composite For tritium analysis Quarterly,
- b. Ground W1-W 7 Quarterly Gamma Isotopic and Tritium analy Quarterly .
- c. Drinking St. Joseph Compositt Sample Gross Beta and Lake Township Collected over a Gamma Isotopic New Buffalo Period of < 31 days Analysis of each Composite
- Sample composite sample.
Over a 2 week Tritium Analysis Period if I-131 of composite Analysis is Performed. Quarterly. I-131 analysis on each composite whe
" Ccmposite samples shall be collected by collecting an aliquot the dose calculate at intervals not exceeding 24 hours. for the consumptio of the water is greater than 1 mre D. C. COOK - UNIT 1 3/4 12-3 per year.
TABLE 3.12-1 (Cont)
- d. Sediment from L2, L3 2/ year Gamna Isotopic Shoreline Analyses Semi-Annually.
- 4. Ingestion
- a. Milk Stevensville At least once per Gamma Isotopic Bridgman 15 days when animals and I-131 Analysi Galien are on Pasture. At of Each Sample.
Dowagiac Least Once Per 31 South Bend Days at Other Times.
- b. Fish Plant Site 2/ year Gamna Isotopic Off-Si te Analysis on Edibl.
Portion.
- c. Food Products Plant Site At time of Harvest Gamma Isotopic l Off-Site (approx. One Sample of Each Analysis on_Edibit 20 mi) of the Following Portion. _ --
Classes of Food Products l
- 1. Grapes Plant Site At time of Harvest Gamma Isotopic One sample of Broad Analysis Leaf Vegetation r
aParticulate sample filters should be analyzed for gross beta 24 hours or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air or water is greater than 10 times the yearly mean of control samples for any medium, gamna isotopic analysis should be performed on the individual samples. 1 1 D. C. COOK - UflIT 1 3/4 12-4 , I
TABLE 3.12-2 Repcrting Levels For Radicactivity Concentrations In Environmental Samples Water ' Food Analysis (pci/1) Airborne Partieglate Fis Milk Products or Gases (pci/m ) (pcf r<hg, wet) (pci/1) (pci/k
! wet) g, l 4 1 H-3 2 X 10 i 3 1 Mn-54 4 1 X 10 3 X 10 2 4 Fe-59 4 X 10 1 X 10 3 4 Co-58 1 X 10 3 X 10 2 4 CO-60 3 X 10 1 X 10 2 4 Zn-65 3 X 10 2 X 10 Zr-Nb-95 4 X 10 I-131 2 0.9 3 1 X 10 Cs-134 30 10 1 X 10 60 1 X 10
.Cs-137 50 20 2 X 10 70 2 X 10
, Sa-La-140 2 X 10 2 3 X 10 D. C. COOK - UNIT 1 3/4 12-5
TABLE 4.12-1 Maximum Values For The Lower Limits of Detection (LLD)a,c Analysis Airborne Particulate Water or Fish Milk (pci/1) Food Products Sediments Gas 3 (pci/kg) (pci/1) (pci/kg, wet) (pci/kg,dr (pci/m ) wet b -2 Gross Beta 4 1 X 10 H-3 2000 . Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 Zn-65 30 260 Zr-95 30 . . - Hb-95 15 Z-131 1 7 X 10 1 60 Cs-134 15 6 X 10 130 15 60 150 Cs-137 18 6 X 10 150 18 60 180 Ba-140 60 60 La-140 15 15 D. C. COOK - UNIT 1 3/4 12-6
TABLE 4.12-1 (Cont) TABLE NOTATION a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely cor.cluding that a blank observation represents a "real" signal. For a particular measurement system (which may include radiochemical separation): LLD = 4.66 sb E*V 2.22 Y exp (-Aat) where LLD is the "a priori" (as pci per unit mass orlower limit of detection as defined above volume), s h is the standard deviation of the background counting rate or of tne counting rate of a blank sample as appropriate (as counts per minute), __ E is the counting efficiency (as counts per transformation), V is the sample size (in units of mass or volume), 2.22 is the number of transfomation per minute per picocurie, Y is the fractional radiochemical yield (when applicable), A is the radioactive decay constant for the particular radionuclide and at is the elapsed time between sample collection (or end of the sample collection period) and time of counting. , The volume 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 appro-priate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide detemined by gama-ray spec-trometry, the background shall include the typical contributions of other radionuclides normally present in the samples (e.g. , potassium - 40 in milk samples). D. C. COOK - UNIT 1 3/4 12-7
Table 4.12-1 (Cont) Table Notation Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes,the presence of inter-I ferring nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contribution factors will be identified and described in the Annual Radiological Environmental Operating Report.
- b. LLD for drinking water,
- c. Other peaks which are measurable and identifiable, together with the radionuclides in Table 4.12-1, shall be identifi,ed and reported.
- 0. C. COOK - UNIT 1 3/4 12-8
~. . . -
L E I l RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION - 3.12.2 A land use census shall be conducted and shall identify the loca-tion of the nearest milk animal, the nearest residence and the nearest garden
- of greater than 500 square feet producing fresh leafy vegetables in each of the 9 land covering meterolog.ical sectors withi.n a distance of five miles. '
APPLICABILITY: At all times. ACTION:
- a. With a land use census identifying a location (s) which yields "
a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, pre-pare and submit to the Comission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the new location (s). -
- b. With a land use census identifying a location (s) which yields I a calculated dose or dose commitment (via the same exoosure patnway) 20 percent creater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1, i prepare and submit to the Comission within 30 days, pursuant to -
Specification 6.9.2, a Special Report which identifies the new location. The new location shall be added to the radiological environmental monitoring program within 30 days, if possible. - The sampling location having the lowest calculated dose or cose commitment (via the same exposure pathway) may be deleted from ; this monitoring program after (October 31) of the year in which -_ this land use census was conducted. -
- c. The orovisions of Specification 3.0.3, 3.0.4 and 6.9.1.13 are not applicable.
i SURVEILLANCE REQUIREMENTS 4.12.2. The land use census shall be conducted at least once per 12 months between the dates of June 1 and October 1, by door-to-door survey, - aerial survey, or by consulting local agriculture authorities. Broao leaf vegetation samoling may be performed at the site boundary in the direction sector with the highest D/Q in lieu of the garden census. D. C. COOK - UNIT 1 3/4 12-9
~
Radiological Environmental Monitoring 3/4 12.3 Interlaboratory Comoarison Program Limiting Condition For Operation 3.12.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the Commission. Apolicability: At all times. Action:
- a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report.
- b. The provisions of Specification 3.0.3, 3.0.4 ana 6.9.1.13 arc not applicable. -
Surveillance Reouirements 4.12.3 A summary of the results obtained as part of the above required Interlaboratory Comparison Program and in accordance with the ODCM (or participants in the EPA crosscheck program shall provide the EPA program code designation for the unit) shall be included in the Annual Radiological Environmental Operating Report. < D. C. COOK - UNIT 1 3/4 12-10
INSTRUMENTATION BASES 3/4.3.3.9 RADI0 ACTIVE LIOUID EFFLUENT INSTRUMENTATION 3/4.3.3.9 The radioactive liquid effluent instrumentation is provided to
- monitoe and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases. The alarm / trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm / trio will occur 4 prior to exceeding the limits of 10 CFR Part 20.
4 The OPERASILITY and use - of this instrumentation is consistent with the requirements. of General Dssign Criteria specified in Section 11.3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant. 4 3/4.3.3.10 RADI0 ACTIVE GASEOUS EFFLVENT INSTR!) MENTATION . 3/4.3.3.10 The radiaoctive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materi.als in gaseous effluents during actual or ' potential releases. The alarm / trip setpoints for these instruments shall be calculated in accordance with NRC -- approved methods in the 0D04 to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation l also includes provisions for monitoring the concentrations of potentially explosive gas mixtures in the waste gas holdup system. The OPERABILITY and use of this instrumentation is consistent with the requirements of . General Design Criteria specified in Section 11.3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant. e 1 1
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D. C. COOK - UNIT 1 B 3/a 3- 5
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4 3/4.11 RADI0 ACTIVE EFFLUENTS BASES 3/4.11.1 LIOUID EFFLUENTS 3/4.11.1.1 CONCENTRATION. This specification is oroviced to ensure that the concentration of radioactive materials released in liquid waste effluents frcm the site to unrestricted areas will be less than :ne concentration levels spect-fied in 10 CFR Part 20, Appendix 3, Table II. This limitation provices additional assurance that the levels of radioactive materials in bodies of water cutside the site will not result in exposures within (1) the Section II.A design objectives of Appendix I,10 CFR Part 50, to an individual and (2) the limits of 10 CFR Part 20.106(e) to the poculation. The c ncentration limit for noble gases is basrd upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in the International Commission on Radiological Protec-tion (ICRP) Publication 2. 3/4.11.1.2 COSE. This specification is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix I,10 CFR Part 50. The Limiting Ccndition for Operaton implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility anc at the same time implement the guides set forth in Section IV.A of Accendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide c:ncentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculations in the 00CM imolement the requirements in Section III.A of Appendix I that conformance with the guides of Acpendix ! be shown by calculational procecurs; based on models anc data such tnat the actual exposure of an individual thrcugh appPopriate pathways is unlikely to be substantially underestimated. The equations specified in the CCCM for calculating the doses due to the actual release rates of radioactive materials in licuid effluents will,be censistent with the methodology provided in Regulatory Guide 1.109,
" Calculation of Annual Cases to Man frem Routine Releases of Reacter Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Accendix I,"
Revisicn 1, October 1977, anc Regulatory Guide 1.113, " Estimating Acuatic Dispersicn of Effluents frem Accidental and Rcutine Reactor Releases for the Purpose of Implementing Acpendix I," April 1977. NUREG-0133 crevices methocs for dose calculations consistent with Regulatory Guice 1.109 and 1.113. D. C. COOK - UNIT 1 3 3/4 11-1
RADI0 ACTIVE 5FFLUESTS ' BASES This specification applies to the release of liquid effluents frem each reactor at the site. The liquid effluents from the shared system are proportioned among the units sharing the system. 3/4.11.1.3 LIQUID WASTE TREATMENT. The operability of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The i requirements that the appropriate portions of this system be used when speci-fied provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This speci-fication implements the requirements of 10 CFR Part 50.36a, General Design Criteria Section 11.1of the Final Safety Analysis Report for the Conald C." Cook Nuclear Plant, and design objective Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents. 3/4.11.1.4 LIQUID HOLCUP TANXS Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the, event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potabic water supply and the nearest surface water supply in an unrestricted area. 3/4.11.2 GASECUS EFFLUENTS 3/4.11.2.1 DOSE RATE. This specification is provided to ensure that the dose rate any any time at the SITE BOUNDARY frem gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Aapendix 3, Table II. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exocsure of an individual in an unrestricted area, Oc annual average concentrations exceeding the limits specified in Appendix 3, Table II of 10 CFR Part 20 (10 CFR Part 20.106 (b)). For individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently low to comoensate for any increase in the atmosoneric diffusien factor above tnat for the site boundary. The specified release rate limits restrict, at all times, the corre-spending gamma and beta dose rates above background to an individual at or beyond the site boundary to < or to <(3000) mrem / year to the skE.(500) mrem / year to tne total body These releasa rate limits also restrict, at all times, the corres;:ending thyroid dose rate acove back-ground to an infant via the cow-milk-infant pathway to < 1500 mrem / year for the nearest cow to the plant. This soecification apolies to the release of gasecus effluents from i all reactors at tne site. The gaseous effluents frem the shared system are ;:rocortioned amcng the units snaring that system. D. C. COOK - UNIT 1 B 3/4 11-2
ql RADI0 ACTIVE EFFLUENTS BASES ! 3/4.11.2.2 DOSE, NOBLE GASES This specification is provided to implement i the requirements of Sections II.8, III. A and IV. A of Appendix I,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in section IV.A of Appendix I to assure that the releases of radio-4 active material in gaseous effluents will be kept "as low as is reasonable i achievable". The Surveillance Requirements implement the requirements in Section III.A of Appendix I that confonn with the guides of Appendix I to 4 be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated. The dose calculations 3 established in the ODCM for calculating the doses due to the actual release i rates of radioactive noble gases in gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1.109, " Calculation.of ' Annual Doses to Man from Routine Releases of Reactor Effluents for the - ' Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision -- 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmos-i pheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. The ODCM i equations orovided for determining the air doses at the site boundary will 4 be based upon the historical average atmospherical conditions. NUREG-0133 i provides methods for dose calculations consistent with Regulatory Guides j 1.109 and 1.111. 3/4.11.2.3 DOSE, RADI0 IODINES, RADI0 ACTIVE MATERIAL IN PARTICULATE FORM AND RADIONUCLIDES OTHER THAN NOBLE GASES. This specification is provided to implement the requirements of Secti.ons II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides
- set forth in Section II.C of Appendix I.
The ACTION statements provide the required operating flexibility and at i the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A cf Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to i be substantially underestimated. The ODCM calculational methods approved by the i NRC for calculating the doses due to the actual release rates of the subject materials are required to be consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Comoliance with i 10 CFR Part 50, Appendix- I, " Revision 1, October 1977 and Regulatory Guide l 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of t D. C. COOK - UNIT 1 B 3/4 11-3 1
! RADIOACTXVE EFFLUENTS
- BASES Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,"
Revision 1, July 1977. These equations also provide for detennining the 4.ctual doses based up the hictorical average atmospheric conditions. The release rate specifications for radiciodines, radioactive material in particulate form and radionuclides other than noble gases are dependant on the existing radionuclide pathways to man, in the unrestricted area. The pathways which are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat pro-ducing animals graze with consumption of the milk and meat by man, and
- 4) deposition on the ground with subsequent exposure of man.
3/4.11.2.4 GASEOUS WASTE TREATMENT The operability of the gaseous radwaste treatment system and the ventila-tion exhaust treatment systems ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion Section 11.1 of the Final Safety Analysis Report for The Donald C. Cook Nuclear Plant and design objective Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous effluents. 3/4 11.2.5 Exolosive Gas Mixture This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treat- - ment system is maintained below the flammability limits of hydrogen and oxygen mixtures. Maintaining the' concentration of hydrogen or oxygen below their flammability limits provides that the releases of radioactive materials will be controlled ir, conformance with the requirements of the General Design Criterion specified in Section 11.1 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant. 3/4 11.2.6 Gas Storage Tanks Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an in-dividual at the nearest site boundary will not exceed 0.5 rem. This is consistant with Standard Review Plan 15.7.1. " Waste Gas System Failure." D. C. COOK - UNIT 1 3 3/4 11 4 -
RADI0 ACTIVE EFFLUENTS BASES 3/4.11.3 SOLID RADI0 ACTIVE WASTE The operability of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require processing and packaging prior to shipment offsite. This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion specified in Section 11.1 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant. The process parameters included in establishing the PROCESS CONTROL PROGRAll may include, but are not limit-ed to waste type, waste pH, waste / liquid / solidification agent / catalyst ratios, and waste curing oil content, waste principal chemical constituents, mixing time. 3/4.11.4 TOTAL DOSE The specification is provided to meet the dose limitations of 40 CFR 190. The specification requi~r es the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed hvice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action which should result.in the limitations of dose to a member of the public for 12 consecutive months to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commit-ment to the member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle sidered.facilities at the same site or within a radius of 5 miles must be con-If the dose to any member of the public is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provi'sion of 40 CFR 190.11 is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation which is part of the nuclear fuel cycle. l D. C. COOK - UNIT 1 8 3/4 11-5 L .
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3/4.12 RADX0LOGXCAL ENVfRONMENTAL MONITORING ;, ;, BASES 1 3/4.12.1 f10NITORING PROGRAM 1 The radiological monitoring program required by this specification provides measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of individuals resultingifrom tne station operation. This monitoring program thereby supplements the' radiological effluent monitoring program by verifying that the measurable concentration of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and mcdeling of the envirornnental exposure pathways. The initially specified monitoring program will be effective for at least the first three years of commercial operation. Following this period, program changes may be initiated based on operational experience. The detection capabilities required by Table 4.12-1 are the state-of-the art for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined an an a priori (before _ the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement. _ Analyses shall be performed in such a manner that the. stated LLDs will ~ be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interferring nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report. 3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in-the use of unrestricted areas are identified and that modifications to the moattoring , program are made if required by the results of tr.u. census. This census , i satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (25 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, tN following assumptions were used,1) that 20% of the garden was usc ir growing broad leaf vegetation (i.e. similiar to lettuce and cabbW) 9d 2) a vegetation field of 2 kg/ square meter. 3/4.12.3 INTERLABORATORY COMPAP.1:0N PRGGFJM The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental l sample matrices are perforned as part of the quality assurance progtam l for environmental monitoring in order to demonstrate that he results are reasonably valid. i D. C. COOK - UNIT 1 B 3/4 12-1
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5.0 DESIGN FEATURES l l 1 5.1 SITE i
' Exclusion Area-5.1' .1 The exclusion area shall be shown in Figure 5.1-1.
4 Low Population Zone 5.1.2 The low population zone shall be shown in Figure 5.1-2. Site Boundary For Gaseous and Liquid Effluents 5.1.3 The site boundary for gaseous and liquid effluents shall be shown in Figure 5.1-3. l
; 5.2 CONTAINMENT CONFIGURATION 5.2.1 'he reactor containment building is a steel lined, reinforced T
- concrete building of cylindrical shape, with a dome roof and having the following design features:
- a. Nominal inside diameter = 115 feet.
- b. Nominal inside height = 160 feet.*
s c. Minimum thickness of concrete walls = 3' 6".
- d. Minimum thickness of concrete roof = 2' 6".
e.' Minimum thickness fo' conrete floor pad = 10 feet.
- f. Nominal thickness of steel liner, side and dome = 3/8 inches.
- g. Nominal thickness of steel liner, bottom = 1/4 inch.
- h. ' Net free volume = 1.24 x 100 e.ubic feet.
- From grade (Elev.-608') to inside of dome.
D. C. COOK - UNIT 1 5-1
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ADMINISTRATIVE CONTROLS 6.3 FliCILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positiens, except for (1) the Radiation Protection Supervisor who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and (2) the Shif t Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in Plant design, and response and analysis of the Plant for transients and accidents. 6.3.2 Until the newly appointed Operations Superintendent obtains a Senior Reactor Operator's License, all of his licensed functions will be performed by a full time assistant who holds a current Senior Reactor Operator's License. 6.4 TRAINING 6.4.1 A retraining and replacement training program for the f acility staff shall be maintained under the direccion of the Training Coordinator and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55. 6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Plant Manager and shall meet or exceed the requirements of Section 27 of the NFPA Code-1976. 6 i i i D. C. COOK - UNIT 1 6-5
I ADMINISTRATIVE CCNTROLS 6.5 REVIEW AND AUDIT 6.5.1 PLANT NUCLEAR SAFETY REVIEW COMMITTEE (PNSRC) FUNCTION 6.5.1.l' The PNSRC shall function to advise the Plant Manager on all matters related to nuclear safety. COMPOSITION 6.5.1.2 The PNSRC shall be composed of the: Chairman: Plant Manager or designated alternate Member: Assistant Plant Managers Member: Operations Superintendent Member: Technical Superintendent . Member: Maintenance Superintendent Member: Control and Instrument Supervisor Member: Nuclear / Computer Engineering Supervisor Member: Plant Chemical Supervisor Member: Performance Supervising Engineer - Member: Plant Radiation Protection Supervisor Member: Shift Supervisor Member: Environmental Coordinator ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the PNSRC Chairman to serve on a temporary basis; however, no more than two alternates snall participate as voting members in PNSRC activities ! at any one time. MEETING FRE0VENCY I 6.5.1.4 The PNSRC shall meet at least once per calendar month and as convened by the PNSRC Chairman or his designated alternate. QUORUM 6.5.1.5 A quorum of the PNSRC shall consist of the Chairman or his designated alternate and four memoers including alternates. l D. C. COOK - UNIT 1 6-6
ADMINISTRATI'/E CONTROLS RESPONSIBILITIES 6.S.I.6 The PNSRC shall be responsible for:
- a. Review of 1) all procedures required by Specification 6.8 and changes thereto, 2) any other proposed procedures or changes thereto as determined by the Plant Manager to affect nuclear safety,
- b. Review of all proposed tests and experiments that affect nuclear safety.
- c. Review of all proposed changes to Appendix "A" Technical Specifications,
- d. Review of all propose'd changes or modifications to plant systems or equipment that affect nuclear safety.
- e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Chairman of the NSDRC. . _ __
- f. Review of those REPORTABLE OCCURENCES requiring 24 hour notification to the Commission.
9 Review of facility operations to detect potential safety hazards.
- h. Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairman of the NSDRC.
- i. Review of the Plant Security Plan and implementing procedures and shall submit recommended changes to the Chairman of the NSDRC.
- j. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Chairman of the NSDRC.
- k. Review of every uncianned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the NSDRC.
- 1. Review of changes to the PROCESS CONTROL PROGRAM, OFFSITE DOSE CALCULATION MANUAL, and radwaste treatment systems.
D. C. COOK - UNIT 1 6-7
ADMINISTRATIVE CONTROLS ' AUTHORITY 6.5.1.7 The PNSRC shall:
- a. Recommend to the Plant Manager written approval or disapproval of items considered under 6.5.1.6 (a) through (d) above.
- b. Render determinations in writing with regard to whether or not each item considered under 6.5.1.6 (a) through (e) above constitutes an unreviewed safety question.
- c. Provide written notification within 24 hours to the NSDRC of disagreement between the PNSRC and the Plant Manager; however, the Plant Manager shall have responsibility- for resolution of such disagreements pursuant to 6.1.1 above.
RECORDS 6.5.1.8 The PNSRC shall maintain written minutes of each meeting and copies shall be provided to the Chairman of the NSDRC. 6.5.2 NUCLEAR SAFETY AND DESIGN REVIEW COMMITTEE (NSDRC) FUNCTION 6.5.2.1 The NSDRC shall function to provide independent review and audit of designated activities in the areas of; -
- a. nuclear power plant operations
- b. nuclear engineering
- c. chemistry and radiochemistry
- d. metallurgy
- e. instrumentation and control
- f. radiological safety
- g. mechanical and electrical r.igineering
- h. quality assurance practices D. C. COOK - UNIT 1 6-8 -
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COMPOSITION 6.5.2.2 The NSDRC shall be composed of the: Chairman: Assistant Vice President, Nuclear Engineering Member: Vice Chairman, Engineering and Ccustruction Member: President and Chief Operating Officer of I&MECo Member: Executive Vice President, Construction & New York Engineering Member: Vice President, Mechanical Engineering Member: Vice President, Electrical Engineering Member: Vice President, Engineering Administration Member: Assistant Vice President, Design Division , Member: Assistant Vice President, Environmental Engineering Division j Member: Plant Manager, D. C. Cook Plant Member: Manager, Nuclear Safety and Licensing Section Alternate: Assistant Chief Mechanical Engineer ! Alternate: Assistant Plant Manager, D. C. Cook Plant Alternate: Executive Assistant to the President of I&MECo Alternate: Assistant Division Manager, Nuclear Engineering ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the NSDRC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in NSDRC activities at any one time. CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the NSDRC Chairman to provide expert advice to the NSDRC. MEETING FREQUENCY 6.5.2.5 The NSDRC shall meet St least once per calendar quarter during the initial year of facility operation following fuel loading and at least once per six months thereaf ter. QUORUM 6.5.2.6 A quorum of NSDRC shall consist of the Chairman or his designated alternate and at least six NSDRC members including alternates. No more than a minority of the quorum shall have line responsibility f or operation of the f acility. D. C. COOK - UNIT 1 6-9
O 1, l ACMINISTRATI'/E CCNTRCLS i AUDITS 6.5.2.8 Audits of facility activities shall be performed uncer the cognizance of the NSDRC. These audits shall enccmpass:
- a. The confermance of facility cperatien to provisiens contained within the Technical Soecification and applicable license conditiens at least once per 12 months.
- b. The performance, training and cualifications of the entire facility staff at least once per 12 months.
- c. The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least once por 6 months.
- d. The performance cf activities recuired by the Guality Assurance '
Frccram to met: the c:itaria of Acpencix "3",10 CFR 50, a: leas cnce per 24 mentas,
- e. The Facility Emergency Plan and implementing procedures at least once per 24 months.
- f. The Facility Security Plan and imolementing precedures at least once per 24 months.
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- g. Any other area of facility operatien censidered appropriate by the NSCRC.
- h. The Facility Fire Protec icn Program and implementing prececures at least cnce per 24 months.
i. An independent fire protection and loss preventien program inspecticn and audit shall be performed a: least once per 12 months utili:ing either qualified offsite licensee cersennel er an cu: side fire protecticn firm.
- j. An inscection and audit of the fire prc:acticn and loss pre-vention program shall be performed cy a qualified cu: side fire censultant at least once per 36 men:ns.
- k. The radioicgical environmental mcnitoring program and the results nereof at least once per 12 mon:hs.
i. The CFFSITE OCSE CALC'JLATION MANUAL anc imolementing procacures at leas: ence per 24 menths. D. C. COOK - UNIT 1 6-11
ADMINISTRATIVE CONTROLS
- m. The PROCESS CONTROL PROGRAM and implementing procedures for solidification of radioactive wastes at least once per 24 months.
- n. The performance of activities required by the Quality Assurance >
Program to meet the criteria of Regulatory Guide 1.21, Rev. 1, June 1974 and Regulatory Guide 4.1, Rev. 1, April 1975 at least once per 12 months. AUTHORITY 6.5.2.9 The NSDRC shall report to and advise the Vice Chairman, Engineering and Construction, AEPSC, on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8. RECORDS 6.5.2.10 Records of the NSDRC activities shall be prepared, approved and distributed as indicated below
- a. Minutes of each NSDRC meeting shal be prepared, approved and forwarded to the Vice Chair =an, Engineering and Construction, AEPSC, within 14 days following each. meeting.
- b. Reports of reviews encompassed by Section 6.5.2.7 above, shall
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be prepared, approved and forwarded to the Vice Chairman, Engineering and Construction, AEPSC, within 14 days following completion of the review. .
- c. Audit reports encempassed by Section 6.5.2.8 above, shall be forwarded to the Vice Chairman, Engineering and Construction, AEPSC, and to the management positions responsible for the areas audited within 30 days after completion of the audit.
6.6 REPORTABLE OCCURRENCE ACTION
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6.6.1 The fallowing actions shall be taken for REPORTABLE OCCURRENCES:
- a. The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.
- b. Each REPORTA3LE OCCURRENCE requiring 24 hour notification to the Commission, shall be reviewed by PNSRC and submitted to the NSDRC Chairman.
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D. C. C00K - UNIT 1 6-12
ADMINISTRATIVE CONTROLS 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
- a. The facility shall be placed in at least HOT STANDBY within once hour.
- b. The Safety Limit Violation shall be reported to the Commission and to the Chairman of the NSDRC within 24 hours.
- c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PNSRC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence.
- d. The Safety Limit Violation Reporr shall be submitted to the Commission, the Chairman of the NSDRC, and the Executive Vice President-Construction and New York Engineering, AEPSC, within 14 days of the violation.
6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and maintained covering the activities ref erenced below:
- a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, November 1972.
- b. Refueling operations.
- c. Surveillance and test activities of safety related equipment.
- d. Security Plan implementation.
- e. Emergency Plan implementation.
- f. Fire Protection Program implementation.
- g. PROCESS CONTROL PROGRAM implementation.
I
- h. OFFSITE DOSE CALCULATION MANUAL implementation.
- 1. Quality Assurance Program for affluent and environmental monitoring using the guidance in Regulatory Guide 1.21, Rev.1, June 1974 and Regulatory Guide 4.1, Rev.1, April 1975.
6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the PNSRC and approved by the Plant Manager prior to implementation and reviewed periodica31y as set forth in administrative procedures. 4 D. C. COOK - UNIT 1 6-13 1 - . - . - - . - . _ - _ . - - , _ - _ - - - . . . - - - ,- - - .--- - --- - -. - -- - - - - -
( . *. ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT I 6.9.1.6 Routine radiological environmental operating reports covering i the operation of the unit during the previous calendar year shall be j subnitted prior to May 1 of each year. , 6.9.1.7 The annual radiological environmental operating reports shall include summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveil-j lance reports and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of the land use censuses required by Specification 3.12.2. If harmful effects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course 4 of action to alleviate the problem. The annual radiological enviconmental operating reports shall include summarized and tabulated results in the format of 3.12-2 of all -- - radiological environmental samples taken during the report period. In the event that some results are not available for inclusion with the report, the report shall be subnitted noting and explaining tne reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.
' The reports shall also include the following: a summary description of the radiological environmental monitoring program including sampling methods i for each sample type, size and physical characteristics of each sample type, sample preparation methods, analytical methods, and measuring equip-ment used; a map of all sample locrtions keyed to a table giving distances and directions from one reactor; the result of land use census required by the Specification 3.12.2; and the results of participation in the Interlaboratory Comparison Program required by Specification 3.12.3.
j 3/ A single submittal may be made for a multiple unit station. The subnittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the i submittal shall specifty the releases of radioactive material for each i unit. I i i . 5
- 0. C. COOK - UNIT 1 6-16 i
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ADMINISTRATIVE CONTROLS-SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 3/ 6.9.1.8 Routine radioactive effluent release reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. 6.9.1.9 The radioactive effluent . release reports shall include a ' summary of the quantities of radioactive liquid and gaseous effluents and solid i waste released from the units as outlined in Regulatory Guide 1.21',
" Measuring, Evaluating and Reporting in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water Cooled Nuclear Power Plants," with data sunnarized on a quarterly basis following the format of Appendix B, thereof.
The radioactive effluent release report to be submitted 60 days after January 1 and July 1 of each year shall include a quarterly summary of hourly meteoregical data collected during the reporting period. This summary may be either in the form of an hour-by-hour listing of wind speed, wind direction, atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of joint frequency distributions of wind speed, wind direction and atmospheric stability. The report submitted 60 - -- days after January 1 shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. These reports shall include as assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary (Figure 5.1-3) during the reporting period. All assumptions used in making these assessments (i.e., specific activity, exposure time and location) shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in effluents (as determined by sampling frequency and measurement) gaseousshall be used for detennining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the Offsite Dose Calculation Manual (ODCM). 3/ A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. D. C. COOK - UNIT 1 6-17
1 ACMINSTRATIVE CONTROLS The radioactive effluent release report to be si.bmitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev.1. The radioactive effluent release report shall include the following information for each type of solid waste shipped offsite during the report period:
- a. Volume (cubic meters),
- b. Total curie quantity (specify whether determined by measurement or estimate),
- c. Principal radionuclides (specify whether determined by measurement or estimate),
- d. Type of waste (e.g., spent resin, compacted dry waste, evaporator bo ttoms) , -
- e. Type of container (e.g., LSA, Type A, Type B, large Quantity),
and
- f. Solidification agent (e.g. , cement).
The radioactive effluent release report shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous , and liquid effluent on a quarterly basis. The radioactive effluent release reports shall include any change to the PROCESS CONTROL PROGRAM (PCP) and the OFFSITE DOSE CALCULATION MANUAL (ODCM) made during the reporting period. MONTHLY REACTOR OPERATING REPORT i 6.9.1.10 Routine reports of operating statistics and shutdcwn experience shall be submitted on a monthly basis to the Director, Office Of Management and Program Analysis, U.S. Nuclear Regulatory Comnission, Washington, D.C. 20555, with a copy to the Regional Office of Inspection and Enforcement, no later than the 15th of each month following the calendar month covered by the report. REPORTABLE OCCURRENCES 6.9.l.11 The REPORTABLE OCCURRENCES of Specification 6.9.1.12 and 6.9.1.13 l below, including corrective actions and measures to prevent recurrence, 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.
ADMIMISTRATIVE CONTROLS PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP 6.9.l.12 The types of events listed below shall be reported within 24 hours i by telephone and confirmed by telegraph, mailgram or facsimile transmission l to the Director of the Regional Office, or his designate no later than the first_ 14 days. working day following the event, with a written followup report within The written followup report shall include, as a minimum a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material event. to provide complete explanation of the circumstances surrounding the a. Failure of the reactor protection system or other systems, subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reaches the setpoint specified as the limiting safety system setting in the technical specifications or failure to complete the required protective function. b. Operation of the unit or affected systems when any parameter or operation subject to a limiting condition for operation is less conservative than the least conservative aspect of the limiting - condition for operation established in the technical specifications. c. Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment. d. Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady state conditions during power operation greater than or equal to l%ak/k; a calculated reactivity balance indicating a SHUTDOWN MARGIN less conservative than specified in the technical specification; short-term reactivity increases that correspond to a reactor period of-less than 5 sec*,nds or, if subcritical, an unplanned reactivity insertton of more tnen 0.5%ak/k; or occurrence of any unplanned criticality. e. Failure or malfunction of one or more comconents which prevents or could prevent, by itself, the fulfillment of the functional requirements in the SAR. of system (s) used to cope with accidents analyzed f. Personnel error or procedural inadequancy which prevents or could prevent, by itself, the fulfillment of the functional recuirements of systems required to cope with accidents analyzed in the SAR. D. C. COOK - UNIT 1 6-19
ADMINISTRATIVE CONTROLS
- g. Conditions arising from natural or man-made events that, as a 1
direct result of the event require unit shutdown, operation of safety systems, or other protective measures required by technical specifications.
- h. Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis
- report or in the bases for the technical specifications that have '
or could have permitted reactor operation in a manner less conservative than assumed in the analyses.
- i. 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 analysis in the safety analysis report or technical specification bases; or discovery during unit life of conditions not specifically considered in the safety analysis report or technical specifications that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.
J. Offsite releases of radioactive material in liquid and gaseous - effluents which exceed the limits of Specification 3.11.1.1 or - 3.11.2.1.
- k. Exceeding the limits in Specification 3.11.1.4 or 3.11.2.6 for the storage of radioactive materials in the listed tanks. The written follow-up report shall include a schedule and a description of activities planned and/or taken to reduce the contents to with-in the specified limits.
D. C. COOK - UNIT 1 6-20
ADMINISTRATIVE CONTROLS THIRTY DAY WRITTEN rep 0RTS 6.9.l.13 The types of events listed below shall be the subject of written reports to the Director of the Regional Office within thirty days of occurence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report fonn. Infornation provided on the licensee event report fann shall be supplemented, as needed, by additional narrative material to provide complete explana-tion of the circumstances surrounding the event.
- a. Reactor protection system or engineered safety feature in-strument settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems.
- b. Conditions leading to operation in a degraded mode pennitted by a limiting condition for operation or plant shutdown re-quired by a limiting condition for operation.
- c. Observed inadequacies in the implementation of administrative .
or procedural controls which threaten to cause reduction of .- degree of redundancy provided in reactor protection systems or engineered safety feature systems.
- d. Abnormal degradation cf systems other than those specified in 6.9.1.12.c above designed to contain radioactive' material resulting from the fission process.
- e. An unplanned offsite release of 1) more than 1 curie of radio-active material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radioiodine in gaseous effluents. The report of an unplanned offsite release of radioactive material shall include the following information:
- 1. A description of the event and equipment involved.
- 2. Causes(s) for the unplanned release.
- 3. Actions taken to prevent recurrence.
4 Consequences of the unplanned release, f. Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table 3.12-2 when averaged over any calendar quarter sampling period. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the co'ndition shall be repcrted and described in the Annual Radiological Environmental Operating Report. D. C. COOK - UNIT 1 6-21
ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional 0Ffice within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference Specifications:
- a. Inservice Inspection Program Review, Specification 4.4.10.
- b. ECCS Actuation, Specifications 3.5.2 and 3.5.3.
- c. Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3.
- d. Inoperable Meteorological Monitoring Instrumentation, Specification 3.3.3.4.
- e. Seismic event analysis, Specification 4.3.3.3.2.
- f. Sealed Source leakage on excess of limits, Specification
- t. 7.7.1.3.
- g. Fire Detection Instrumentation, Specification 3.3.3.7.
- h. Fire Suppression Systems, Specifications 3.7.9.1, 3.7.9.2, 3.7.9.3 and 3.7.9.4 D. C. COOK - UNIT 1 6-22
ADMINISTRATIVE CONTROLS 6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:
- a. Records and logs of unit operation covering time interval at each power level.
- b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
- c. ALL REPORTABLE OCCURRENCES submitted to the Commission.
- d. Records of surveillance activities, inspections and !
calibrations required by these Technical Specifications.
- e. Records of changes made to the procedures required by Specification 6.8.1.
- f. Records of sealed source leak tests and results,
- g. Records of annual physical inventory of all sealed source material on record.
6.10.2 The following records shall be retained for the duration of the Facility Operating License:
- a. Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report.
- b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories,
- c. Records of radiation exposure for all individuals entering radiation control areas.
- d. Records of gaseous and liquid radioactive material released to the environs.
- e. Records of radioactive shipments.
- f. Records of transient or operational cycles for those facility components identified in Table 5.9-1.
- g. Records of training and qualification for current members of the Plant staff.
- h. Records of in-service inspections performed pursuant to these Technical Specifications.
- 1. Records of Quality Assurance activities required by the QA Manual,
- j. Records of reviews performed f or changes made to procedures or equipment or review of tests and experiments pursuant to 10 CFR 50.59.
- k. Records of meetings of the PNSRC and the NSDRC.
- 1. Records for Environmental Qualification which are covered under the provisions of paragraph 6.13.
- m. Records of reactor tests and experiments.
D. C. COOK - UNIT 1 6-23
' ADMINISTRATIVE CONTROLS 6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consiMent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.
6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20:
- a. A High Radiation Area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a High Radiation Area and entrance thereto shall be controlled by issuance of a Radiation Work Permit and any individual or group of individuals permitted to enter such areas shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area.
- b. A High Radiation Area in which the intensity of radiation is greater than 1000 mrem /hr shall be subject to the provisions of 6.12.1.a above, and in addition, locked doors shall be l prnvided to prevent unauthorized entry into such areas and the keys shall be maintained under the adminstrative control of the Shift Operating Engineer on duty.
6.13 ENVIRONMENTAL QUALIFICATION 6.13.1 By no later than June 30, 1982, all safety-related electrical equipment in the facility shall be qualified in accordance with the providions of: Division of Operating Reactors " Guidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors" (DOR Guidelines), or, NUREG-0588 " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment", December 1979. Copies of these documents are attached to Order for Modification of License No. DPR-74, dated October 24, 1980. 6.13.2 By no later than December 1,1980, complete and auditable l records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the D0R Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified. D. C. COOK - UNIT 1 6-24
l ADMINISTRATIVE CONTROLS l 6.14 PROCESS CONTROL PROGRAM (PCP) i
, 6.14.1 The PCP snall be approved by the Ccmmission prior to implementation.
6.14.2 Licensee initiated changes to the PCP: 1. Shall be submitted to the Commission in the semi-annual Radioactive Effluent Release Report for the period in which the change (s) was made. This subnittal shall contain:
- a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;
- b. A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and
- c. Documentation of the fact that the change has been reviewed and found acceptable by the PNSRC.
- 2. Shall become effective upon review and acceptance by the D'.'SnC. - -
5.15 0FFSITE 00SE CALCULATION MANUAL (ODCM) 5.15.1 The ODC4 shall be approved by the Commission prior to implementation. 5.14.2 Licensee initiated changes to the ODCM: 1. Shall be submitted to the Ccmmission in the Semi-Annual Radioactive Effluent Release Report in the next report after the report period the change (s) was made effective. This submittal shall contain:
- a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and orovided with an approval and date box, together with appropriate analyses or evaluations justifying the change (s);
b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and
- c. Documentation of the fact that the change has been reviewed and found acceptable by the PNSRC.
D. C. COOK - UNIT 1 6-25
ADMINISTRATIVE CONTROLS
- 2. Shall become effective upon review and acceptance by the PNSRC.
6.15.3 Commission initiated changes:
- 1. Shall be determined by the PNSRC to be applicable to the facility after consideration of facility design.
- 2. The licensee shall provide the Commission with written noti-fication of-their determination of applicability including any necessary revisions to reflect facility design.
6.16 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS (Licuid, Gaseous and Solid) 6.16.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid): 1. Shall be reported to the Comission in the Annual Operating Report for the period in which the evaluation was reviewed by the (PNSRC) The discussions of each change shall contain:
- a. A summary of the evaluation that led to the determination that ~
the change could be made in accordance with 10 CFR 50.59;
- b. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;
- c. A detailed description of the equipment, components and processes involved and the interfaces with other plant systems;
- d. An evaluation of the change which shcws the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted,in the license application and amendments thereto;
- e. An evaluation of the change which shows the expected maximum exposure to individuals in the unrestricted area and to the general population that differ from those previously estirrated 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 prior to when the changes are to be made; D. C. COOK - UNIT 1 6-26 .
l . ADMINISTRATIVE CONTROLS
- g. An estimate of the exposure to plant operating personnel as a result of the change; and I
- h. Documentation of the fact that the change was reviewed and found acceptable by the PNSRC.
- 2. Shall become effective upon review and acceptance by the PNSRC.
6.16.2 Commission initiated changes:
- 1. The applicability of the change to the facility shall be determined by the (PNSRC) after consideration of the facility design.
- 2. The licensee shall provide the Commission with written notification of its determination of applicability including any necessary revisions to reflect facility design.
l
- 0. C. COOK - UNIT 1 6-27
l 1 l ATTACHMENT 2 5 TO AEP:NRC:0055F D. C. COOK - UNIT 2 l l l
l ATTACHMENT 2 TO AEP:NRC:0055F LIST OF CHANGES TO APPENDIX A TECHNICAL SPECIFICATION D. C. COOK - UNIT 2 DPR-74 REMOVE PAGES INSERT PAGES I-A IV IV IX-A XIV-A XV XV XVI XVI XVII XVII 1-4 1-4 1-6 to 1-8 1-6 to 1-10 3/4 3-53 to 3/4 3-54 3/4 3-53 to 3/4 3-66 3/4 11-1 to 3/4 11-17 3/4 12-1 to 3/4 12-10 B 3/4 3-3 8 3/4 3-3 B 3/4 11-1 to B 3/4 11-5 B 3/4 12-1 5-1 5-1 5-9 6-5 to 6-9 6-5 to 6-9 6-11 to 6-13 6-11 to 6-13 6-16 to 6-21 6-16 t- i-28
INDEX DEFINITIONS SECTION PAGE SOURCE CHECK .................................................. 1 - 6 i' PROCESS CONTROL PROGRAM (PCP) . . . . ............................. 1 - 6 SOLIDIFICATION ................................................ 1 - 7 0FFSITE DOSE CALCULATION MANUAL (ODCM) ........................ 1 - 7 GASEOUS RADWASTE TREATMENT SYSTEM ............................. 1 - 7 VENTILATION EXHAUST TREATMENT SYSTEM .......................... 1 - 7 i PURGE-FURGING ................................................. 1 - 7 VENTING ....................................................... 1 - 7 MEMBER ( S ) 0 F THE P UBLIC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 8 SITE BOUNDARY ................................................. 1 - 8 UNRESTRICTED AREA ............................................. 1 - 8 D. C. COOK - UNIT 2 I-A 4 4 r - - - - -,,- -- - ~- -n- - , - - , ,
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 Axial Flux Dif f erence . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 2-1 3/4.2.2 Heat Flux Hot Channel Factor ...................... 3/4 2-5 3/4.2.3 RCS F low Rat e and R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 2-9 3/4.2.4 Quadrant Power Tilt Ratio . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-13 3.4.2.5 DNB Paraseters .................................... 3/4 2-15 3/4.2.6 Axial Power Dis tribution . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 2-17 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION . . . . . . . . . . . . . . . 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTDi INSTRUMENTATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 3-14 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation .............. 3/4 3-34 Movable Inco re Detector s . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 3-3 8 Meteorological Instrumentation . . . . . . . . . . . . . . . . . . . . 3/4 3-39 Remote Shutdown Instrumentation ................... 3/4 3-42 Post-Accident Instrumentation . . . . . . . . . . . . . . . . . . . . . 3 /4 3-45 Axial Power Distribution Monitoring System ........ 3/4 3-48 Fire Detection Instrumentation .................... 3/4 3-50
, Radioactive Liquid Effluent Instrumentation ....... 3/4 3-53 Radioactive Gaseous Process'and Effluent Monitoring Instrumentation ........................ 3/4 3-58 3/4.3.4 TURBINE OVERSPEED PROTECTION ...................... 3/4 3-65 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS No rmal Op eration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 4-1 3/4.4.2 SAFETY VALVES - SHUTDOWN .......................... 3/4 4-4 3/4.4.3 SAFETY VALVES - OPERATING . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-5 3/4.4.4 P RE S SURIZ ER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/44-6 3/4.4.5 STEAM GENERATORS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 4-7 D. C. COOK - UNIT 2 IV
/
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration ......................................... 3/4 11-1 Dose .................................................. 3/4 11-4 Liquid Waste Treatment ...............................;. 3/4 11-5 Liquid Ho ldup Tanks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 11-6 3/4.11.2 GASEOUS EFFLUENTS Dose Rate ............................................. 3/4 11-7 Dose - Noble Gases ....................,................. 3/4 11-10 Dose - Radiciodines, Radioactive Material in Particulate Form, and Radionuclides other than Noble Cases ............................................. 3/4 11-11 G a s e ou s Rad wa s t e Tre a tm en t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-12 Explo s ive G as Mix tur e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-13 Gas Storage Tanks ...................................... 3/4 11-14 3/ 4.11. 3 SOLID RADIOACTIVE WASTE . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-15 3/4.11.4 TOTAL DOSE ......................................... 3/4 11-17 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3 /4.12.1 MONITORING PRCGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 12-1 3/4.12.2 LAND USE CENSUS ................................... 3/4 12-9 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM ................. 3/4 12-10 D. C. COOK - UNIT 2 IX-A t N
INDEX BASES SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS ................................. B 3/4 11-2 3/4.11.3 SOLID RADIOACTIVE WASTE ........................... B 3/4 11-5 3/4.11.4 TOTAL DOSE ........................................ B 3/4 11-5 3/4.12 RADIOLOGICAL ENVIR0hWAL MONITORING 3/4.12.1 MONITORING PROGRAM ................................ B 3/4 12-1 3/4.12.2 LAND USE CENSUS ................................... B 3/4 12-1 3/4.12.3 INIERLABORATORY COMPARISON PROGRAM ................ B 3/4 12-1 D. C. COOK - UNIT 2 XIV - A
-- +- --
m , -
\ i l INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE Exclusion Area ........................................... 5-1 Low P op ul a tion Z on e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 Site Boundary f or Gaseous and Liquid Ef fluents . . . . . . . . . . . 5-1 5.2 CONTAINMENT Co nf ig ur a t io n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 1 Design Pressure and Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.3 REACTOR CORE Fuel Assemblies .............................. ........... 5-4 Control Rod Assemblies ................................... 5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . 5-4 Volume ................................................... 5-5 5.5 METEOROLOGICAL TOWER LOCATION ............................ 5-5 5.6 FUEL STORAGE Criticality - Spent Fuel ................................. 5-5 Criticality - New Fuel ................................... 5-5 Drainage ................................................. 5-6 Capacity ................................................. 5-6 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT . . . . . . . . . . . . . . . . . . . . . . 5-6 D. C. COOK - UNIT 2 XV
INDEX . ADMINISTRATIVE CONTROLS SECTION , PAGE 6.1 RESPONSIBILITY.............................................. 6-1 6.2 ORGANIZATION . 0ffsite..................................................... 6-1 Facility Staff.............................................. 6-1
'6.3 FACILITY STAFF 00ALIFICATIONS............................... ,
6-5 6.4 TRAINING.................. ................................. 6-5 6.5 REVIEW AND AUDIT 6.5.1 PLANT NUCLEAR SAFETY REVIEW ' COMMITTEE F u n c t i,o n . . . . . . . . '. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-6 Composition................................................ 6-6
. A l t e rn a t e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-6 -
~
Meeting Frequency.......................................... 6-6 l Quorum.................................................... 5-6 Responsibilities.......~................................... 6-7 Authority................................................. 6-8 Records................................................... . 6-8 6.5.2 NUCLEARSAFETYANDDESIGNREVIEWCCMMITTEE
' Function.................................................. . 6-8 Comcosition............................................... 6-9 Al t e r n a t e s . . . . . . . . . . . . . . . . . . . < . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-9
- l. >
I D. C. COOK - UNIT 2 XVI 4 e
#4P
ADMINISTRATIVE CONTROLS SECTION PAGE Consultants .............................................. 6-9 Meeting Frequency ........................................ 6-9 Quorum ................................................... 6-10 l Review ................................................... 6-10 Audits ................................................... 6-11 Authority ................................................ 6-12 l Records .................................................. 6-12 6.6 REPORTABLE OCCURRENCE ACTION ............................. 6-12 6.7 SAFETY LIMIT VIOLATION ................................... 6-13 6.8 PROCEDURES ............................................... 6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS AND REPORTABLE OCCURRENCES . . . . . . . . . . 6-14 6.9.2 S P EC IAL REPO RT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 2 6.10 RECORD RETENTION ......................................... 6-23 6.11 RADIATION PROTECTION PROGRAM ............................. 6-24 6.12 HIGH RADIATION AREA ...................................... 6-24 6.13 ENVIRONMENTAL QUALIFICATION .............................. 6-25 6.14 PROCESS CONTROL PROGRAM (PCP) ............................ 6-26 6.15 0FFSITE DOSE CALCULATION MANUAL (ODCM) ................... 6-26 6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS . . . . . 6-27 D. C. C0CK - UNIT 2 XVII
~
OEFINITIONS - UNIDENTIFIED LEAKAGE 1.15 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED 1.EAXAGE. PRESSURE BOUNDARY LEAKAGE 1.16 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall. CONTROLLED LEAKAGE 1.17 CONTROLLED LEAKAGE shall b.e that seal water flow supplied to the reactor coolant pump seals. CUADRANT POWER TILT RATIO 1.18 -QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper , excore detector calibrated output to the average of .th.e upper excore detector calibrated octputs, or the ratic of the maximdm lower execre detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.- With one excore detector inoperable, the remaining three detectors shall be used for computing the average. . DOSE EQUIVALENT I-13T 1.19 COSE EQUIVALENT I-131 shall be that concentration of I-131 (uci/ gram) which alone would produce the same thyroid dose as the quantity and iso-topic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors ! for Pcwer and Test Reactor Sites, or in NRC Regulatory Guide 1.109 Rev.1, l October 1977. i D. C. COOK - UNIT 2 l-4 r
DEFINITIONS PHYSICS TESTS 1.25 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 13.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission. E'- AVERAGE DISINTEGRATION ENERGY 1.26 f shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energfes per disnitegration (in MeV) for isotopes, other than todines, with half lives greater than 15 minutes, making up at least 95% of the total non-fodine activity in the coolant. SOURCE CHECX 1.27 A SCURCE CHECX shall be the qualitative assessment of Channel response when the Channel sensor is exposed to a radioactive source. PRCCESS CCNTROL PRCGRAM (.CCP) 1.28 The PRCCESS CCNTRCL PRCGRAM shall contain the current formula, samoling, analysis, tests and determinations to be made to ensure that the precessing and packaging of solid radioactive wastes will be accomplished in such a way as to assure ccmoliance with 10 CFR 20,10 CFR 71, Federal and State regulations and other requirements governing the shipment and disposal of radioactive wasta. D. C. COOK - UNIT 2 1-6 ,
DEFINITIONS SOLIDIFICATION 1.29 SOLIDIFICATION shall be the conversion of radioactive liquid, resine and sludge wastes from liquid systems into a form that meets ; l shipping and burial site requirements. OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.30 The OFFSITE DOSE CALCULATION MANUAL shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints and the conduct of environmental radiological monitoring program. GASEOUS RADWASTE TREATMENT SYSTEM 1.31 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary l coolant system off-gases from the primary system and providing for delay l or holdup for the purpose of reducing the total radioactivity prior to release to the environment. VENTILATION EXHAUST TREATMENT SYSTEM 1.32 A VENTILATION EXHAUST TREATMENT SYSTEM is any systam designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust I gases through charcoal absorbers and/or HEPA filters for the purpose of removing iodines or particulate from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. PURGE-PURGING 1.33 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. VENIING 1.34 VENTING is the controlled process of discharging air or gas f rom a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process. D. C. COOK - UNIT 2 1-7
DEFINITIONS MEMBER (S) 0F THE PUBLIC 1.35 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the Plant. This category does not include employees of the utility, its contractors or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the Plant. SITE BOUNDARY 1.36 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased or otherwise controlled by the licensee. UNRESTRICTED AREA 1.37 An UNRESTRICIED AREA shall be any area at or beyond the SITE BOUNDARY to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial, commercial, institutional and/or recreational purposes . D. C. COOK - UNIT 2 1-8
f TABLE 1.1
/
OPERATICNAL ."CDES REACTIVITY ". RATED AVERAGE CCCLANT MCDE CONDITION, X.f< THER4AL POWER
- TDtPERATURE
- 1. POWER OPERATION ,
> 0.99 > 5% > 350*F
- 2. STARTUP 1 0.99 < 5% 3 350*F
- 3. HOT STANCBY < 0.99 0 1 350*F 4 HOT SHUTDCWN < 0.99 0 350*F >Tavg
> 200*F
- 5. COLD SHUTDOWN < 0.99 0 < 200*F
- 6. REFUELING ** < 0.95 0 < 140*F Excluding cecay neat.
Reactor vessel head unbolted or removed and fuel in the vessel. 1
- 0. C, COOK - UNIT 2 1-9
i l . '. l* 8 i TABLE 1.2 FREQUENCY NOTATION NOTATION FREQUENCY S AT LEAST ONCE PER 12 E0URS D AT LEAST ONCE PER 24 HOURS 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 R AT LEAST ONCE PER 549 DAYS S/U PRIOR TO EACH REACTOR START-UP P COMPLETED PRIOR TO EACH RELEASE N.A. NOT APPLICABLE D. C. COOK - UNIT 2 1-10
INSTRUMENTATION RADI0 ACTIVE LIOUID EFFLUENT INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.9 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. APPLICABILITY: As shown in Table 3.3-12. ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than a value which will ensure that the limits of 3.11.1.1 are met, without delay suspend the release .of radioactive liquid effluents monitored by the affected channel, reset, or declare the channel inoperable.
- b. With one or more radioactive liquid effluent monitoring instrumentation channels inoperable, take the applicable ACTION shown in Table 3.3-12.
- c. The provisions of Specifications 3.0.3, 3.0.4 and 6.9.1.13 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.9.1 The setpoints shall be determined in accordance with methodology as described in the ODCM and shall be recorded. 4.3.3.9.2 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECX, SOURCE CHECX, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-8. D. C. COOK - UNIT 2 3/4 3- 53 i
l . TABLE 3.3-12 RADI0 ACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Instrument Operable Acolicabilftv Action
- 1. Gross Radioactivity Monitors Providing Automatic Release Termination
- a. Liquid Radwaste (1)
Effluent Line At times of release 23
- b. Steam Generator (1) At times of release Blowdown Line 24
- c. Steam Generator (1) At times of release 24 Blowdown Treatment Effluent
- 2. Gross Radioactivity --
Monitors Not Providing Automatic Release Termination
- a. Service Water System (1) At all times 25 Effluent Line
- 3. Continuous Composite Sampler Flow Monitor
- a. Turbine Building Sump (1) At all times Effluent Line 25
- 4. Flow Rate Measurement Devices
- a. Liquid Radwaste Line (1)
- b. Discharge Pipes" At times of release 26 (1) At all times NA
- c. Steam Generator Blowdown Treatment Effluent (1) At times of release 26
- Pump curves and valve settings may be utilized to estimate flow; in such cases, Action Statement 26 is not applicable.
D. C. COOK - UNIT 2 3/4 3-54
O TABLE 3.3-12 (Continued) TABLE NOTATION Action 23 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may be resumed for up to 30 days, provided that prior to initiating a release:
- 1. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1 and;
- 2. At least two technically qualified members of the Facility Staff independently verify the discharge valving. Otherwise, suspend release of radioactive effluents via this pathway.
Action 24 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are analyzed for gross radioactivjty (beta or gamma) at a limit of detection of at least 10~ uci/ gram:
- 1. At least once once per 8 hours when the specific activity of the secondary coolant is >0.01 uci/ gram DOSE EQUIVALENT I-131. .
- 2. At least once per 24 hours when the specific activity of the secondary coolant is --0.01
< uci/ gram DOSE EQUIVALENT I-131.
Action 25 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that at least once per 3 hours, grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 10 -'uci/ml. Action 26 With the number of Channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours during actual releases. D. C. COOK - UNIT 2 3/4 3- 55
- , , - - - -- - - , e.- , , - -- -
l . TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK ~ CHECK CALIBRATION TEST
- 1. Gross Beta or Gamma Radioactivity Monitors providing alarm and automatic isolation
- a. Liquid Radwaste D* P R(3) Q(1)
Effluent Line
- 5. Steam Generator D* M R(3) Q(1)
Blowdown Effluent Line
- c. Steam Generator D* M R(3) Q(1)
Blowdown Treatment Effluent Line 2. Gross Beta or Gamma Radioactivity Monitors Providing Alarm But Not Providing Automatic Isolation
- a. Service Water D M System Effluent R(3) Q(2)
Line e
- 3. Continuous Composite Samplers
- a. Turbine Building D N/A N/A N/A Sump Effluent Line
- 4. Flow Rate Monitors
- a. Liquid Radwaste D(4)* N/A R Q Effluent
- b. Steam Generator D(4)* N/A N/A N/A Blowdown Treatment Line -
- During Releases Via This Pathway D. C. COOK - UNIT 2 3/4 3 56
TABLE 4.3-8 (Cont) FABLE NOTATION (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 control not set in operating 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 setpoint.
** 2. Circuit failure.
** 3. Instrument indicates a downscale failure. -
** 4. Instrument controls not set in operating mode.
(3) The initial CHANNEL CALIBRATION shall be performed using one or -- more sources with traceability back to the National Bureau of Standards. These sources shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used. (4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days on which continuous, periodic or batch releases are made.
- Instrument indicates, but does not provide for automatic isolation.
** As equipment becomes operat' 7nal .
D. C. COOK - UNIT 2 3/4 3- 57 ,
Instrumenta tion Radioactive Gaseous Process and Effluent Monitoring Instrumentation Limiting Condition for Operation ,3.3.3.10 The radioactive gaseous process and effluent monitoring instrumentation
- channels shown in Table 3.3-13 shall be OPERABLE with their alarm / trip
,setpoints set to ensure that the limits of 3.11.2.1 are not exceeded. ,Apolicability: As shown in Table 3.3-13. Action:
- a. With a radioactive gaseous process or effluent monitoring instrumentation channel alarm / trip setpoint less conservative than a value which will ensure that the limits of 3.11.2.1 are met, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, reset, or declare the channel inoperable.
- b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13. .
- c. The provisions of Specification 3.0.3, 3.0.4 and 6. 9.1.13 are not applicable. _.
Surveillance Reouirements 4.3.3.10.1 The setpoints shall be determined in accordance with methodology as described in the ODCM and shall be recorded.* 4.3.3.10.2 Each radioactive gaseous process or effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-9. This surveillance requirement does rot apply to the Waste Gas Holdup System Hydrogen and Oxygen Monitors, as their setpoints are not addressed in the 00CM. D. C. COOK - UNIT 2 3/4 3-58 '
)
TABLE 3.3-13 Radioactive Gaseous Effluent Monitorina Instrumentation Minimum Channels Instrument Operable Acolicability Action
- 1. Waste Gas Holdup System Explosive Gas Monitoring System
- a. Hydrogen Monitor (1) **
30
- b. Oxygen Monitor (2) **
29
- 2. Condenser Evacuation System
- a. Noble Gas Activity (1) ****
28 Monitor
- b. Flow Rate Monitor (1) **** 27
- 3. Auxiliary Building Ventilation System
- a. Noble Gas Activity (1) * ~
28 Monitor
- b. Iodine Sampler (1)
- 32 Cartridge
- c. Particulate (1)
- Sampler Filter 32
- d. Effluent System (1)
- 27 Flow Rate Measuring Device
- e. Sampler Flow Rate (1)
- Measubing Device 27
- 4. Containment Purge System ***
- a. Noble Gas Activity (1) ****3 31 Monitor
- 6. Particulate Sampler (1) ****) 32
- 5. Waste Gas Holdup System
- a. Noble Gas Activity (1) ****2 33 Monitor Providing Alarm and Termination of Gas Decay Tank Releases
- 6. Gland Seal Exhaust
- a. Noble Gas Activity (1)
Monitor **** 28
- 5. Flow Rate Monitor (1) ****
27 D. C. COOK - UNIT 2 3/4 3- 59
TABLE 3.3-13 (C n:) At all times
*** During waste gas ncidu: system c:eratica (treat:ent for ricacy system gases).
Moni: Ors sa= le c:ntainment at cs:nere no: c:n sinmen : urge. Au:::a.ic termination of purge on high centainment activity.
"** Curing releases via this pa:hway I
For purge pur oses only, see Tecnnical icecifications 3.3.3.1, 3.2.6.1, and 3.9.9 for other recuirements. 2 For gas cecay tank releases only, see 3. for accitional recuirements. l O D. C. COOK - UNIT 2 3/4 3-60
b TABLE 3.3-13 (Cont) TABLE NOTATICN Action 27 With the number of channels OPEMBLE less than required by the Minimum Channels OPEMBLE requirement, effluent releases via this pathway may continue for up to 30 days provided the ficw rate is estimated at least cnce per 4 hours. Action 28 With the number of channels OPEMBLE less than required by tne Minimum Channels OPEUSLE requirect t, effluent releases via this pathway may continue for up t; 30 days provided grab samples are taken at least once per 8 hours and these samples are analy:ed for gross activity within 24 hours. Action 29 With the number of channels OPEMBLE one less than required by the Minimum Channels OPEMBLE requirement, operaticn of this system may centinue for up to 14 days. With 2 channels in-operable, operatien of this system may continue for up to 14 days, provided grab samples are taken and analyzed every 12 hours. Action 30 With the number of channels OPEMBLE less than required by the Minimum Channels OPERABLE requirement, operation of this system may continue for up to 14 days, provided grab samoles are taken and analy:ed every 12 hours. Action 31 With the number of channels OPEMBLE less than recuired by the Minimum Channels OPERABLE requirements, immediately suspend PURGING of radioactive effluents via this pathway. Action 32 With the numoer of channels OPEMBLE less than required by the Minimum Channels CPEMBLE requirement, effluent releases via the affected pathway may centinue for up to 30 days provided samoles are continuously collected with auxiliary sampling equipment as recuired in Table 4.11-2. Action 33 Wit the numcer of channels CPEMBLE less than required by the Minimum Channels OPERABLE requirement, the centants of the t:nk(s) may be released to the environ:.ent for uo to 14 days provided that prior to initiating the release:
- a. At least two independent samoles of the tank's contents are analy:ed and,
- b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineues; otherwise, sus:end release of radicactive effluents via this pa taway.
D. C. COOK - UNIT 2 3/s 3 -61
l . TABLE 4.3-9 Radioactive Gaseous Effluent Monitorino Instrumentation l Surveillance Recu1rements Channel Channel Source Channel Functional Instrument Check Check Calibration Test !
- 1. Waste Gas Holdup System Explosive Gas Monitoring System
- a. Hydrogen D* *
- NA Q(3) M Monitor
- b. Oxygen D **
- NA Q(4) M Monitor
- c. Oxygen D**
- NA Q(4) M Monitor (Al t. )
- 2. Condenser Evacuation System
- a. Noble Gas Activity D*
- M R(2) Q(1)
Monitor _
- b. System Effluent D*
- NA R Q Flow Rate
- 3. Auxiliary Building Ventilation System
- a. Noble Gas Activity D* M R(2) Q(1)
Monitor
- b. Iodine Sampler W* NA NA
- c. NA Particulate Sampler W* NA NA NA
- d. System Effluent D* NA R Q Flow Rate Measure-ment Device
- e. Sampler Flow Rate D* NA R Q Measurement Device
- 4. Containment Purge System
- a. Noble Gas Activity D** P Monitor R(2) Q(5)
- b. Particulate Sampler W** NA NA NA
- 5. Waste Gas Holdup System
- a. Noble Gas Activity P** P R(2) Q(5)
Monitor Providing Alarm & Termination of Gas Decay Tank Re, leases D. C. COOK - UNIT 2 3/4 3 62
TABLE a.3-9 Cen: 2s1cc zea. .:,Anaus:
- o. s
- 3. NOI! '3as AC*iVi*y O" *i E ( 2,' [(I) l-!cn i ~c t
- 3. System Effluent 0" tlA R Q Flow Rate At all times During Release 'lia This Pathway
'" Curing Waste Gas Holdup System Operation (Treatment f:r Primary System Offgases)
D. C. COOK - UNIT 2 2/a 3-63
l . TABLE 4.3-9 (Cont) TABLE NOTATION 1) 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 setpoint.
**2. Circuit failure.
** 3 . Instrument indicates a downscale failure.
**4. Instrument controls not set in operate mode.
- 2) The initial CHANNEL CALIBRATION shall be performed using one or more sources with traceability back to the National Bureau of Standards. These sources shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used.
- 3) The CHANNEL CALIBRATION shall include the use of standard gas samples --
containing a nominal:
- 1. One volume percent hydrogen, balance nitrogen, and
- 2. Four volume percent hydrogen, balance nitrogen.
- 4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
- 1. One volume percent oxygen, balance nitrogen, and
- 2. Four volume percent oxygen, balance nitrogen.
5) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control rocm alarm annunciation occurs if any of the following conditions exists: 1.
**2. Instrument indicates measured levels above the alarm / trip setpoint.
Circuit failure.*
**3. Instrument indicates a downscale failure.*
**4. Instrument controls not set in operate mode.*
- Instrument indicates, but does not provide automatic isolation.
** As equipment becomes operational.
D. C. COOK - UNIT 2 3/4 3-64
1* . INSTRUMENTATION 3/4.3.4 TURBINE OVERSPEEb PROTECTION LIMITING CONDITION FOR OPERATION 3.3.4.1 At least one turbine overspeed protection system shall be OPERABLE. , APPLICABILITY: MODES 1, 2 and 3. ACTION:
- a. With one stop valve or one control valve per high pressure turbine steam lead inoperable or with one reheat stop valve or one reheat intercept valve per low pressure turbine steam lead inoperable, operation may continue for up to 72 hours provided the inoperable valve (s) is restored to OPERABLE status or at least one valve in the affected steam lead is closed; otherwise, isolate the turbine from the steam supply within the next 6 hours.
- b. With the above required turbine overspeed protection system otherwise inoperable, within 6 hours either restore the system to OPERABLE status or isolate the turbine from the steam supply.
SURVEILLANCE RE0VIREMENTS 4.3.4.1.1 The provisions of Specification 4.0.4 are not applicable. 4.3.4.1.2 The above required turbine overspeed protection system shall be demonstrated OPERABLE:
- a. At least once per 7 days by cycling each of the following valves through at least one complete cycle from the running position.
- 1. Four high pressure turbine stop valves.
- 2. Four high pressure turbine control valves.
- 3. Six low pressure turbine reheat stop valves.
- 4. Six low pressure turbine reheat intercept valves.
D. C. COOK - UNIT 2 3/4 3- 65
l .
. o INSTRUMENTATION LIMITING CONDITION FOR OPERATION
- b. At least once per 31 days by direct observation of the move-ment of each of the above valves through one complete cycle from the running position.
- c. At least once per 18 months by performance of a CHANNEL CALIBRATION on the turbine overspeed protection systems.
- d. At least once per 40 months by disassembling at least one of each of the above valves and performing a visual and surface inspection of valve seats, disks and stems and verifying no unacceptable flaws or corrosion.
e e e D. C. COOK - UNIT 2 3/4 3- 66
3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 Liouid Effluents Concentra tion . Limiting Condition for Operation 3.11.1.1 The concentration of radioactive material released at any time from the site to unrestricted areas (see Figere 5.1-3) shall be limited to the concentrations specified in .10 CFR Part 20, Appendix 8 Table II, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolged or ' entrained noble gases, the concentration shall be limited to 2 X 10- uci/ml total activity. Applicability: At all times, j Action: With the concentration of radioactive material released from the site ex-ceeding the above limits, without delay restore the concentration to within the above limits. SURVEILLANCE REQUIREMENTS l 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.11-1. 4.11.1.1.2 The result of radioactive analysis shall be used in accordance with the methods of the ODCM to assure that all c.:ncentrations at the point of release are maintained within the limits of Specification 3.11.1.1. i S D. C. COOK - UNIT 2 3/4 11-1 e 4 e ,c-, -+ - - , -
-, - - r ,
. l .
TABLE .4.11-1 Radioactive Liquid Waste Sampling and Analysis Program Lower Limit a Minimum Type Of of Detection L1 quid Release Sampling Analysis Activity (LLD) Type Frequency Frecuency Analysis uci/ml A. Batch Waste -7 Release Tanks c P P Princi SX10 Gamma ' pal Emitters * , J Each Batch Each Batch I-131 IX10'"
~
P Dissolved IX10-5 and Entrain-ed Gases
. (Gamma One Batch /M M Emitters)
P M H-3 1X10-5 Each Batch Composite b Gross Alpha ; 1X10-/ P Sr-89,Sr-90 8 Q b Each Batch Composite .I re-co SX10 lX10- 6 B. Plant'Continu'ous Daily , W Principal SX10 7 Releases d Gamma b Emitters' , Composite I-131 ::1X10
- M M Dissolved IX10-5 Grab Sample and Entrain-ed Gases (Gamma Emitters)
Daily M
. b H-3' 1X10i Comoosite Gross Alpha llX10 /
Daily Q b Composite Sr-89, Sr-90'5X10_3 Fe-55 l1X10-6 D. C. COOK - UNIT 2 3/4 11-2 .
. .-,r,,.r -
- . - - , , - - . - . 4 -
TABLE 4.11-1 (Cont)
. TABLE NOTATION The lower limit of detection (LLD) is defined in Table Notation
- a. 0# Table 4.12-1 of Specification 4.12.1.1.
- b. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is re-presentative of the liquids released.
- c. A batch release is the discharge of liquid wastes of a discrete volume.
Prior to sampling for analysis, each batch shall be isolated and re-circulated to ensure thorough mixing. i
- d. A continuous release is the discharge of liquid waste of a non-discrete volume; e.g. from a volume of system that has an input flow during the continuous release.
- e. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144. This list --
does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. r D. C. COOK - UNIT 2 3/4 11-3 e
RADI0 ACTIVE EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION
! 3.11.1.2 The dose or dose commitment to an individual fran radioactive ' material in liquid effluents released to unrestricted areas (see Figure 5.1-3) shall be limited:
- a. During any cale dar quarter to s1.5 mrem to the total body and to s5 mrem to :ny crgan, and
- b. During any calendar year to s 3 mrem to the total body and to s 10 mrem to any organ.
APPLICABILITY: At all times. ACTION: a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and _ submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause (s) for exceed-ing the limit (s) and defines the corrective actions taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be within the above limits. This Special Report shall also include (1) the results of radiological analyses of the drinking water source, and (2) the radiological impacts on finished drink'ig water supplies with regard to the requirements of 40 CFR 141, Safe Drinking Water Act. (Applicable only if drinking water supply is taken from the receiving water body.)
- b. The provisions of Specification 3.0.3, 3.0[4 and 5.9.1.13 are not apolicable.
l SURVEILLANCE REOUIREMENTS 4.11.1.2 Dose Calculations. Cumulative dose contriettions from liquid l effluents srcil be determined in accordance with the Offsite Dose Calcu-lation Manual (00CM) at least once per 31 days. l L C. COOK - UNIT 2 3/4 11-4 I t I
I Radioactive Effluents . Liquid Waste Treatment ' Limitino Condition For Ooeration 1 3.11.1.3 The liquid radwaste treatment system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the pro-jected doses due to the liquid effluent from the site (see Figure 5.1-3) when averaged over 31 days', would exceed 0.06 mrem to the total body or 0.2 mrem to any organ. . Acolicability? At all times. . Action: '
- a. With radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of any'other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special- _
Report which includes the following information: , i 1. Identification of the inoperable equip. ment or subsystems and the reason for inoperability, _
- 2. Action (s) taken to restore the inoperable equipment to operable '
status, and
- 3. Summary description of action (s) taken to prevent recurrence.
- b. The provisions of Specification 3.0.3, Y.0.4 and 6.9.F.13 a're no;t. .
applicable.
- Surveillance Recuirements ..
4.11.1.3 Doses due to liquid releases to UNRESTRICTED. AREAS shall be pro' jected at least once per 31 days, in accordance with tha 03CM, whenever liquid releases are being made without being processed by.the liquid ra'r6 waste treatment system.
?1
, ~
D. C. COOK - UNIT 2 3/4 11-5
- MO 9
i t Radioactive Effluents ! Liquid Holduo Tanks
- Limitina Condition For Operation 3.11.1.4 The quantity of radioactive material contained in each of the following tanks shall be limited to less than or equal to lQ curies, excluding tritium and dissolved or entrained noble gases,
- a. Outside temporary tarks.
Applicability: At all times.
. Ac tion:
- a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, without delay suspend all additions of
~
s '. radioactive material to the tank and within 48 hours reduce the tank contents to within tne limit.
- b. The provisions of Specification 3.0.3, 3.0.4 and 6.9J .13 are not applicable.
r Surveillance Reauirements 4.11.1.4 .The quantity of radioactive material contained in each of the above 1,isted tanks shall be determined to be within the above limit by analyzing a representative sample of the tank /s contents at least once per 7 days when radioactive materials are being added to the tank. Tanks included in this Specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and sur-- rounding area drains connected to the liquid radwaste treatment system.
+y-D. C. CC0K - UNIT 2 3,4 11-6 O
V t
8 Radioactive Effluents 3/4.11.2 Gaseous Effluents . i Dose Rate Limitino Condition For Ooeration l 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site (See Figure 5.1-3) shall. be limited to the following:
- a. For noble gases: < 500 mrem /yr to the total body and < 3000 mrem /yr _
to the skin, and _
- b. For all radiciodines and fcr all radioactive materials in particulate form and radionuclides (other than noble gases) with half-lives greater than 8 days: < 1500 mrem /yr to any organ. _
Apolicability: At all times. Action: With the dose rate (s) exceeding the above limits, without delay decrease the release rate to within the above limit (s). Surveillance Recuirements - 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM. 4.11.2.1.2 The dose rate due to radioactive materials, other than noble gases, in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the 00CM by ob-taining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2. D. C. COOK - UNIT 2 3/4 11-7
TASLE c.11-2 r RADICACTIVE GASECUS WASTE SA"PLI!G At:0 A?!ALYS:S PRCGRA*t Minimum Ty::e of Lower Limi Analysi s of Ce:ecticn Casecus Releasa Type Frequency Frequency Activity Analysis i (J(ci/ml )* l P P Principal Gamma l Each Tank Each : l a. Wasta Gas Storage Tank Grab Sample Tank e Emi tters
- 1 X 10 "
i I P P Principal Gamma , Emitiers8 1 X 10 ~ Each Purge Each '
. b. Contaiment Purge Grab Sample b Pu rge b l g,3 l7y7ge l ! '
l W
- c. xb Condenser Evacuation Grab Principal Gamma System and G1and Seal Sample b Particulate ,
l' Samole Emi ers e 1 X 10 '
.xhaust
- l i
Mb l H-3 ~
': U" l } Mb j i Charecal i I-131 Sample l1x10-12 Continuous" l Noble v - .,Gas 4 . ,. l Noble Oases l1x10-6 i l n
W- .. I d. Charcoal I-131 1 X 10 -
! Auxiliary Building Vent Continuous d Sample ;
i c W Continuous d Particula a Principal Gamma . Samel e emi + e, s i 1 Y 10 I M Continucus d Canposi te i Particula te 3ross Alcha ' 1 I 10-33 ! Samol e ! d M Continucus lCcm;cstte . H-3 1 X 10-5 , I l
! I Q I, - d ,
l Continucus Cemccsite Sr-89, Sr-90 1 X 10) i Panticulate i ! Sancie !
! } Continuous } $Udi.jas . , lNeoleGases l1X10 ,
i
- As ecuicmen ceccmec cceraticnal l D. C. COOK - UtlIT 2 3/4 '.1-3 i
TABLE 4.11-2 (cont) TABLE NOTATION ,
- a. The lower limit of detection (LLD) is defined in Table Notation a.
of Table 4.12-1 of Specification 4.12.1.1.
- b. Analyses shall also be performed following any operational occurrence which has altered the mixture of radionuclides as indicated by RCS analysis.(ie., start-up.) ,
- c. Samples shall be changed at least once per 7 days and analyses shall be' completed within 48 hours after changing. Analys'es shall also be performed at least once per 24 hours for 7 days following each shutdown, startup or similar operational occurrence which lead to significant increases or decreases in radioiodine in the Reactor Coolant System. When samples collected for 24 hours are analyzed, the corresponding LLD's may be increased by a factor of 10.
- d. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specification 3.11.2.1, 3.11.2.2, and 3.11.2.3.
- e.. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133M, Xe-135 and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measureable and identifiable, together with the above nuclides, shall also be identified and reported.
e D. C. COOK - UNIT 2 3/4 11-9 6 o N
RADIOACTIVE EFFLUENTS DOSE. NOBLE GASES ' - LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose in UNRESTRICTED AREAS due to noble gases released in gaseous effluents _shall be limited to the following:
- a. During any calendar quarter, to < 5 mrad for_ gamma radiation and ;; 10 mrad for beta radiation; ,
- b. During any calendar year, to jE 10 mrad for gamma radiation and j; 20 mrad for beta radiation.
APPLICABILITY: At all times. ACTION:
- a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare-and submit to the Commission within 30 days, pursuant to '
Specification 6.9.2, a Special Report which identifies the __ cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent re-leases will be within the above limits.
- b. The provisions of Specification 3'0.3, 3.0.4 and 6.9.1.13 .
.are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.2 Dose Calculations Cumulative dose contributions for the total time period shall be determined in accordance with the Offsite Dose Calculation Manual (00CM) at least once every 31 days. I 9
- 0. C. COOK - UNIT 2 3/4 11-10 O
RADIOACTIVE EFFLUEITIS - DOSE, RADIOIODINES, RADICACTIV_E MATERIAL IN PARTICULATE FORM, AND RADIONUCLIDES OTHER THAN NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to A FEMBER OF THE PUBLIC f rom radiciodines, radioactive materials in particulate form, and radionuclides other than noble gases with half-lives greater than 8 days in gaseous effluents , released to unrestricted areas shall be limited to the fciloving: - i '
- a. During any calendar quarter to n 7.5 mrem to any organ;
- b. During any calendar year to 6 15 mrem to any organ.
APPLICABILITY: At all times. ACTION:
- a. With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides other than noble gases in gasecus ef fluents exceeding any of the above limits, prepare and submit to the Ccenission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the causc(s) f or exceeding the limit and defines the corrective actions taken to reduce the reletsses and the proposed corrective action to be taken to assure that subsequent releases will be within the above limits.
- b. The provisions of Specification 3.0.3, 3.0.4, and 6.9.1.13 are not applicable.
SURVEILLANCE REOUIREMENTS 4.11.2.3 DOSE CALCULATIONS Cumulative dose contributions for the total time period shall be determined in accordance with the ODCM at least once every 31 days. D. C. COOK - UNIT 2 3/4 11-11
Radioactive Effluents Gaseous Radwaste Treatment Limitine Condition For Oceration 3.11.2.4 The gaseous radwaste treatment system and the ventilation exnaust treatment system shall be used to reduce the radioactive materials in gaseous
, waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases to unrestricted areas (See Figure 5.1.3) when averaged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad i for beta radiation. The ventilation exhaust treatment system shall be used to recuce radioactive materials in gaseous waste prior to their discharge wnen the projected doses due to gaseous effluent releases to unrestricted areas (See Figure 5.1-3) when averaged over 31 days wculd exceed 0.3 mrem to any organ.
Acolicability: At all times. Action:
- a. With gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Ccmmission witnin 30 days, pursuant to Specification 6.9.2, a Special Report which in-cludes the following information:
- 1. Identification of the inocerable equipment or subsystems and the reason for inoperability.
- 2. Action (s) taken to restore the inoperable equipment to operable sta tus.
b. The provisions of Specification 3.0.3, 3.0.4 and 6. .l.13 are not acclicable. Surveillance Recuirements 4.11.2.4 Coses due to gaseous releases to UNRESTRICTED AREAS snall be projectec at least once per 31 days in accordance with tne CCC4, wnenever the gaseous waste trea: ment system or ventilation exhaust treatment system is not coerational. D. C. COOK - UNIT 2 3/2 11-12
RADI0 ACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE - LISITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the waste gas holdup system shall be limited to 1 52 by volume if the hydrogen in the system is > 4% by volume. APPLICABILITY: At all times. ACTION:
- a. With the concentration of oxygen in the waste gas holdup system
> 2% by volume but S 4% by volume and containing > 4% hydrogen, restore the concentration of oxygen to concentration to < 4% within 48 hours. 3 2% or reduce the hydrogen
- b. With the concentration of oxygen in the waste gas holdup system or tank > 45 by volume and > 4% hydrogen by volume without delay suspend all additions of waste gases to the system or tank and reduce the concentration of oxygen to 124 or the concentration of hydrogen to 1S 4 within 48 hours in the system or tank,
- c. The provisions of Specification 3.0.3, 3.0.4 and 6.9.1.13 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.11.2.5 The concentration of oxygen in the waste gas holdup system shall be determined to within the above limits by continuously monitoring the waste gases in the waste gas holdup system with the oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.10. D. C. COOK - UNIT 2 3/4 11-13
l RADICACTIVE EFFLUENTS GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to 43,800 curies noble gas (considered as Xe-133). APPLICABILITY: At all times. ACTION:
- a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, without delay suspend all additions of radioactive tanh contentsmaterial to withintothe thelimit, tank and within 48 hours reduce the
- b. The provisions of Specification 3.0.3, 3.0.4 and 6.9.1.13 are not applicable .
SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas - storage tank shall be determined to be within the above limit at least once per 4 days by analysis of the Reactor Coolant System noble gases. D. C. COOK - UNIT 2 3/4 11-14 1
RADI0 ACTIVE EFFLUENTS 3/4.11.3 SOLID RADI0 ACTIVE WASTE LIMITING CONDITION FOR OPERATION I
~
3.11.3 The solid radwaste system shall be used as applicable in accordance with a PROCESS CONTROL PROGRAM for the SOLIDIFICATION and packaging of radioactive wastes to ensure meeting the requirements of 10 CFR Part 20 and of 10 CFR Part 71 prior to shipment of radioactive wastes from the site. APPLICABILITY: At all times. ACTION:
- a. With the packaging requirements of 10 CFR Part 20 and/or 10 CFR Part 71 not satisfied, suspend shipments of defectively packaged solid radioactive wastes from the site.
- b. With the solid radwaste system inoperable for more than 31 days, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days __
pursuant to Specification 6.9.2 a Special Report which includes the following information:
- 1. Identification of the inoperable equipment or subsystems and the reason for inoperability,
- 2. Action (s) taken to restore the inoperable equipment to operable status,
- 3. A description of the alternative used for SOLIDIFICATION and packaging of radioactive wastes, and 4 Summary description of action (s) taken to prevent a recurrence.
- c. The provisions of Specification 3.0.3, 3.0.4 and 6.9.1.13 are not applicable...
D. C. COOK - UNIT 2 3/4 11-15
l . A _ SOLID RADI0ACTIVZ WASTE SURVEILLANCE REQUIREMENTS 4.11.3.1 The solid radwaste system shall be demonstrated operable at least once per 92 days by:
- a. Operating the solid radwaste system at least once in the previous 92 days in accordance with the PROCESS CONTROL PROGRAM, or b.
Verification of the existence of a valid contract for SOLIDIFICATION to be performed by a contractor in accordance with a PROCESS CONTROL PROGRAM. 4.11.3.2 THE PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive waste (e.g. filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions). ! a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION. - SOLIDIFICATION of the batch . _ may then be resumed using the alternative SOLIDIFICATION parameter.s determined by the PROCESS CONTROL PROGRAM. b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each > consecutive batch of the same type of wet waste untti at least 3 consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.14 to assure SOLIDIFICATION of subsequent batches of waste. t 1 ) D. C. COOK - UNIT 2 3/4 11-16
- RADI0 ACTIVE EFFLUENTS 3/4 11.4 TOTAL DOSE i
L.IMITING CONDITION FOR OPERATION l ' 3.11.4 The dose or dose commitment to a real individual from all uranium fuel cycle sources is limited to < 25 mrem to the total body or any organ (except the thyroid, which is limTted to < 75 mrem) over a period of 12 consecutive months. APPLICABILITY: At all times. ACTION: a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifi-cation 3.11.1.2.a , 3.11.1.2.b, 3.11.2.2.a . 3.11.2.2.b, 3.11.2.3.a or 3.11.2.3.b, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Director, Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, D. C. 20555, within 30 days, which defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of Specification 3.11.4. This Special Report shall include - ~~ an analysis which estimates the radiation exposure (dose) to a member of the public from uranium fuel cycle sources (including all effluent pathways and direct radiation) for a 12 consecutive month period that includes the release (s) covered by this report. If the estimated dose (s exceeds the limits of Specification 3.11.4, and if the release condition) resulting in violation of 40 CFR 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190 and including the specified infor-mation of 5190.11(b). Submittal of the report is considered a timely
- request, and a variance is granted until staff action on the request is complete. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the requirements for dose limitation of 10 CFR Part 20, as addressed in other sections of this Technical Specification.
i b. The provisions of Specification 3.0.3, 3.0.4 and 5.9.1.13 are not applicable. SURVEILLANCE REQUIREMENTS 1 4.11.4 Dose Calculations Cumulative dose contributions from liquid and gasecus effluents shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3 and with the ODCM. D. C. COOK - UNIT 2 3/4 11-17
[ 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION 3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12-1.
; APPLICABILITY: At all times. ' ACTION:
l a. With the radiological environmental monitoring program not being i conducted as specified in Table 3.12-1, prepare and submit to the Commission in the Annual Radiological Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a reccurence. (Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, or to malfunction of automatic sampling equipment. If the latter, every effort shall be made to complete the corrective action prior to the end of the next sampling period.)
- b. With the level of radioactivity in an environmental sampling medium at one or more of the locations specified in Table 3.12-1. exceeding the limits of Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter, a Special Report which includes an evaluation of any release conditions, environmental factors or other aspects which caused the limits of Tahle 3.12-2 to be exceeded.
This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report. When more than one of the radio-nuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted ,if: concentration (1) concentration (2)
+
+...>1 limit level (1) limit level (2)
When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be sub-mitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specification 3.11.1.2, 3.11.2.2 and 3.11.2.3. D. C. COOK - UNIT 2 3/4 12-1 ,
RADIOLOGICAL ENVIR0flMENTAL MONITORING LIMITI!!G C0!!DITIO:1 FOR OPERATIO!! (C0tiTINUED)
- c. With milk or fresh leafy vegetable samples unavailable from any of the sample locations required by Table 3.12-1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause of the unavaila-bility of samples and identifies locations for obtaining replace-ment samples. The locations from which samples were unavailable may then be deleted from Table 3.12-1 provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations, if available.
- d. The provisions of Specification 3.0.3, 3.0.4 and 6.9.1.13 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the locations given in the table 7.nd figures in the ODCM and shall be analyzed pursuant to the requirements of Tables 3.12-1 and 4.12-1. S D. C. COOK - UNIT 2 3/4 12-2
*w - - - - , , - ,-
O TABLE 3.12-1 RADICLOGICAL E:l'/IRCNMENTAL MONITORING 9:CGRAM Ex:csure .:a:hway Sam:i'ng anc Tyce of 7recuency anc/ce 5a.rcias fam:le Locatiens Colla:-den .:recuency of Anaivsis
- 1. Air:cene
- a. Radicicdine al-A6 (Site) Continuous operation Radioicdine canister ano Particulates of samoler with Analy:e New Suffalo, Sample Collection Weekly for I-131 South Bend, as required by Ocwagiac, and Dust Loacing Sut At Particulate samoler-Colema are Least once Per 7 Days Gross Seta Rad-Background icactivity followins Filter Changea , cor posite (by loca-tion) for gamma isotopic quarterly
- 2. Direct Radiation a) T1-T9 (Site) At least once per Gamma Cese. At 92 days Least Once Per b) New Suffalo 92 days South Bend Ccwagiac >
Colcma c) 10 TLD Monitcr Locations in the Five Mile Radiu__s.
- 3. Waterberne ,
- a. Surface L1 , L2 , L3 Cceposite Samole Gamma Isotopic Over One-Mentn Pericd Analysis monthly.
C:mposite For tritium analysis Quarterly.
- b. Ground Wl-W 7 Quarterly Gamma Isc:coic and Tritium analysis Cua rterly .
- c. Drinking St. Josepn Cem:osi tt Sample Gecss Seta and Lake Townsnip Collected over a Gamma Isoteoic New Buffalo Period of < 31 days Analysis of each Cemcosite* Samole c:meosite sam le.
Over a 2 week Tritium Analysis Pericd if I-131 of ccm:csite Analysis is :erformed. Quarterly. I-131 analysis en eacn ::::osi a wnen o Cem:csi:e sam:les shall be : llec:ad :y : ,iiec-ing in alicuct
- ne ccse :alcula ac a: in ervals not excaeding 24 neurs. for ne c:nsume-icn Of tne aater is grea ar nan 1 tram
- er year.
D. C. COOK - UNIT 2 3/4 12-3
r TABLE 3.12-1 (Cont)
- d. Sediment from L2, L3 2/ year Gamma Isotopic Shoreline Analyses Semi-Annually.
- 4. Incestion
- a. Milk Stevensville At least once per Gamma Isotopic Bridgman 15 days when animals Galien and I-131 Analysis are on Pasture. At of Each Sample.
Dowagiac Least Once Per 31 South Bend Days at Other Times.
- b. Fish Plant Site 2/ year Gamma Isotopic {!
' Off-Site AnalysisonEdiblej Portion. j
- c. Food Products Plant Site At time of Harvest Gamma Isotopic t
Off-Site (approx. One Sample of Each Analysis on Edible 20 mi) of the Following Portion. . _ -- Classes of Food Products
- 1. Grapes Plant Site At time of Harvest Gamma Isotopic One sample of Broad Analysis Leaf Vegetation aParticulate sample filters should be analyzed for gross beta 24 hours or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air or water is greater than 10 times the yearly mean of control samples for any medium, gamma isotopic analysis should be performed on the individual samples.
D. C. COOK - UNIT 2 3/4 12-4 -
TAELE 3.12-2 Re;:crting Levels For Radicactivity Concen: rations In Environmentas Samoies i Water Airborne articg ate Analysis (pci/1) Eish Milk Food Proc or Gases pci/m (pc1/Kg, wet) (;ci/1) (pci/kg,wel I ' i 4 I H-3 2 X 10 } Mn-54 1 X 10 3 X 10 Fe-59 4 X 10 1 X 10 Co-58 1 X 10 3 X 10 CO-60 3 X 10 1 X 10 9 Zn-65 4 3 X 10 ' 2 X 10 - Zr-Nb-95 4 X 10 I-131 2 0.9 3 1 X 10 Cs-134 3 30 10 3 1 X 10 60 1 X 10
. Cs-137 3 50 20 3 2 X 10 70 2 X 10 Ba-La-140 2 X 10 2 3 X 10
- 0. C. COOK - UNIT 2 3/4 12-5
TABLE 4.12-1 Maximum Values For The Lower Limits of Detection (LLD)a,c Analysis Airborne Particulate Water or Fish Milk (pci/1) Food Products Sediments Gas 3 (pci/kg) (pci/1) (pci/kg, wet) (pci/kg, dry - (pci/m ) wet b -2 Gross Beta 4 1 X 10 H-3 2000 . Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 In-65 30 260 Zr-95 30 _- Nb-95 15 I-131 1 7 X 10 1 60
~
Cs-134 15 6 X 10 130 15 60 150
~
Cs-137 18 6 X 10 150 18 60 180 Ba-140 60 60 La-140 15 15
.s D. C. COOK - UNIT 2 3/4 12-6
TABLE 4.12-1 (Cont) TABLE NOTATION a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system (which may include radiochemical separation): LLD = 4.66 sb E V 2.22 Y exp (-Aat) where LLD is the "a priori" lower limit of detection as defined above (as pci per unit mass or volume), s h is the standard deviation of the background counting rate or of tMe counting rate of a blank sample as appropriate (as counts per minute), - E is the counting efficiency (as counts per transformation), V is the sample size (in units of mass or volume), 2.22 is the number of transformation per minute per picocurie, Y is the fractional radiochemical yield (when applicable), A is the radioactive decay constant for the particular radionuclide and at is the elapsed time between sample collection (or end of the sample collection period) and time of counting. , The volume of Sb used in the calculation of the LLD for a detection system shall be based on the actual observed vaciance of the background counting rate or of the counting rate of the blank samoles (as appro-priate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide detennined by gamma-ray spec-tremetry, the background shall include the typical contributions of other radionuclides nonnally present in the samples (e.g., potassium - 40 in milk samples). D. C. COOK - UNIT 2 3/4 12-7
C) Table 4.12-1 (Cont) Table Notation Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes,the presence of inter-ferring nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contribution factors will be identified and described in the Annual Radiological Environmental Operating Report.
- b. LLD for drinking water, -
- c. Other peaks which are measurable and identifiable, together with the radionuclides in Table 4.12-1, shall be identified and reported.
D. C. COOK - UNIT 2 3/4 12-8
. -. ._, _ _ , , , , - _ _ _ _ . . _ _ _ - _ _ , _ . . _ , . ~ , _ . ,_ - . , _ _ - . . - . _ _ - ,
~
'~
l 7
'F
, RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION i 3.12.2 A land use census shall be conducted and shall identify the loca- - tion of the nearest milk animal, the nearest residence and the nearest " garden
- of greater than 500 square feet producing fresh leafy vegetables e in each of tne 9 land covering meterological sectors withi.n a distance of five miles.
APPLICABILITY: At all times. ACTION:
~
- a. With a land use census identifying a location (s) which yields a calculated dose or dose commitment greater than the values -
currently being calculated in Specification 4.11.2.3, pre- ' pare and submit to the Commission within 30 days, pursuant to - Specification 6.9.2, a Special Report which identifies the new location (s). g
- b. With a land use census identifying a location (s) which yields a calculated dose or dose commitment (via the same exoosure -
patnway) 20 percent oreater than at a location from which samples . are currently being obtained in accordance with Specification 3.12.1,
~
prepare and submit to the Commission within 30 days, pursuant to - Specification 6.9.2, a Special Report which identifies the new location. The new location shall be added.to the radiological ; environmental monitoring program within 30 days, if possible. The sampling location having the lowest calculated dose or cose cannitment (via the same exposure pathway) may be deleted from this monitoring program after (October 31) of the year in which this land use census was conducted.
- c. The orovisions of Specification 3.0.3, 3. 0.4 and 6.9. l .13 are no t applicable.
SURVEILLANCE REOUIREMENTS 4.12.2. The land use census shall be concucted at least once per 12 months between the dates of June 1 and October 1, by door-to-door survey, aerial survey, or by consulting local agriculture authorities. Broao leaf vegetation sanoling may be performed at the site boundary in the direction sector with the highest D/Q in lieu of the garden census. D. C. COOK - UNIT 2 3/4 12-9
l - Radiological Environmental ffonitoring 3/412.3 Interlaboratory Comoarison Program Limiting Condition For Operation i 3.12.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the Commission. Acclicability: At all times. Action:
- a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report,
- b. The provisions of Specification 3.0.3, 3.0.4 ano 6.9.1.13 are not applicable.
Surveillance Requirements 4.12.3 A summary of the results obtained as part of the above required interlaboratory Comparison Program and in accordance with the 00CM (or participants in the EPA crosscheck program shall provide the EPA program code designation for the unit) shall be included in the Annual Radiological Environmental Operating Report. D. C. COOK - Uti1T 2 3/4 12-10
.. )
3/4.3 INSTRUMENTATION BASES 3/4.3.3.8 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall f acility fire protection program. In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY. Use of containment temperature monitoring is allowed once per hour if containment fire detection in inoperable. 3/4.3.3.9 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases. The alarm / trip setpoints f or these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design _ Criteria specified in Section 11.3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant. 3/4.3.3.10 RADIOACTIVE GASEOUS EFFLUENT INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases. The alarm / trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring the concentrations of potentially explosive gas mixtures in the waste gas holdup system. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria specified in Section 11.3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant. 3/4.3.4 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from exceesive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety related components, equipment or structures. D . C . COOK - UNIT 2 B 3/4 3-3
3/4.11 RADICACTIVE EFFLUENTS BASES 3/4.11.1 LICUID EFFLUE.'ITS 3/4.11.1.1 CONCENTRATICN. This s:ecification is provided to ensure that the concentration of radioactive materials released in liquid wasta effluents from the site to unrestricted areas will be less than the concentration levels speci-fied in 10 CFR Part 20, Appendix 3, Table II. This limitation provides additional assurance that the levels of radicactive materials in bodies of water cutside the site will not result in expcsures within (1) the Section II.A design objectives of Aopendix I,10 CFR Part 50, to an individual and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling raofoisotcpe and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in tne International Ccmmission on Radiological Protec-tion (ICRP) Publication 2. 3/4.11.1.2 COSE. This specification is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix I,10 CFR Part 50. The Limiting Condition for Operaton implenents the guides set forth in Section II.A of Apcendix I. The ACTION statenents provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactivematerial in liquid effluents will be kept "as icw as is ~ reasonably achievable." Also, for fresh water sites with drinking ws:er supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentratiens in :ne finished crinking water that are in excess of the requirements of 40 CFR 141. The dose calculations in the ODCM implement the recuirements in Secticn III.A of Appendix ! that conformance with the guides of Apcendix ! be snown by calculational precedures based on medels anc data such tnat the actual ex:csure of an individual thr0 ugh appr0priate pathways is unlikely to be substantially uncerestimated. The equatiens specified in the CDCM for calculating :ne doses cue to the actual release rates of radioactive materials in licuid effluents will be consistent with the methodolcgy provided in Regulat ry Guide 1.109,
" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Ccmoliance with 10 CFR Par: 50, Accendix I,"
Revisien 1, October 1977, anc Regulat:ry Guide 1.113, " Estimating Acuatic Discersicn of Effluents frem Accidental anc R0utine Reac; r Releases for the
?ur;ose of Implementing Acpendix I," Acril 1977. NUREG-0133 Oravices me:nces for dose calcula:1cns consistent with Regulatory Guide 1.109 and 1.113.
D. C. COOK - UNIT 2 3 3/4 11 1
l . RA0f0 ACTIVE EFPJJEMTS BASES l l This scecification applies to the release of liquid effluents frem i each reactor at the site. The liquid effluents frem the shared system are proportioned among the units sharing the system. 3/4.11.1.3 LIQUID WASTE TREATMENT. The operability of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirements that the appropriate portions of this system be used when speci-fied provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This speci-fication implements the requirements of 10 CFR Part 50.36a, General Design,, Criteria Section 11.1 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant, and design objective Section II.0 of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of thePa,rt CFR dose50, design objectives for 1iquid effluents. set forth in Section II.A of Appendix I,10 3/4.11.1.4 LICUID HOLDUP TANKS Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the, event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix S, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area. 3/4.11.2 GASECUS EFFLUENTS 3/4.11.2.1 COSE RATE. This specification is provided to ensure that the dose rate any any time at the SITE BOUNDARY from gaseous effluents frcm all units on the site will ce within the annual dose limits of 10 CFR Part 20 ' for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II. These limits orovide reasonable assurance that radioactive material discharged in gaseous affluents will not resul: in *ne excosure of an indivicual in an unrestricted area, to annual average concentrations exceeding the limits l specified in Appendix S, Table II of 10 CFR Part 20 (10 CFR Part 20.106 (b)}. I For individuals who may at times be within the site boundary, the cccupancy of the indivicual will be sufficiently low to ceccensate for any increase ! in tne a:mosoneric diffusion factor above tnat for the site boundary. The s:ecified release rate limits restrict, at all times, the corre-sponding gamma and beta dose rates above background to an indivicual at or beyond the site bcundary to 1 (500) mrem / year to the total bcdy or to <(2000) mrem / year to the skin. These release rate 1imits also restrict, at all times, the corres;:ending thyroid dose rate above back-ground to an infant via the cow-milk-infant pathway to 1 500 1 mrem / year for the nearest c:w to One plant. i This scecifica:icn acclies to ne release of gasecus effluents from all reac:ces at the site. The gasecus effluents fr:m tne shared system a re procortionec amcng the units snaring :nat system. l t D. C. COOK - UNIT 2 B 3/4 11-2 i
. 0 i
RADIOACTIVE EFFLUENTS iI' i BASES 3/4.11.2.2 DOSE, NOBLE GASES This specification is provided to implement the requirements of Sections II.8, III.A and IV.A of Appendix I,10 CFR i Part 50. The Limiting Condition for Operation implements the guides set
- forth in Section II.B of Appendix I. The ACTION statements provide the i
required operating flexibility and at the same time implement the guides set forth in section IV.A of Appendix I to assure that the releases of radio-active material in gaseous effluents will be kept "as low as is reasonaole achievable". The Surveillance Requirements implement the requirements in i Section III.A of Appendix I that conform with the guides of Appendix I to i be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is i unlikely to be substantially underestimated. The dose calculations ' established in the 00C4 for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1.109, " Calculation.of Annual Doses to Man from Routine Releases of Reactor Effluents for the - i Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revi: ion - - 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmos-pheric Transport and Dispersion of Gaseous Effluents in Routine Releases 4 i from Light-Water-Cooled Reactors," Revision 1, July 1977. The 00CM equations provided for detennining the air doses at the site boundary will i be based upon the historical average atmospherical conditions. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides
; 1.109 and 1.111.
4 i 3/4.11.2.3 DOSE, RADI0 IODINES, RADI0 ACTIVE MATERIAL IN PARTICULATE FORM AND RADIONUCLIDES OTHER THAN N0BLE GASES. This specification is provided i to implement the requirements of Secti.ons II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I. ~ The ACTION statements provide the required operating flexibility and at i the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODC4 calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the
- actual exoosure of an individual through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methods approved by the NRC for calculating the doses due to the actual release rates of the subject materials are required to be consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man frem Routine
- Releases of Reactor Effluents-for the Purpose of Evaluating Compliance with 3
~
10 CFR Part 50, Appendix I, " Revision 1, October 1977 and Regulatory Guide 1.111, "Metheds for Estimating Atmospheric Transport and Dispersion &f
- D. C. COOK - UNIT 2 8 3/4 11-3 4
.v--- . , , - . , , . ..,m.we, ~ . . , - - . - - . , n, , , - - , , - ,, v.--.--m ,n--m., , ,., , - - - -- - .- e- m- - ---,
RADI0ACTfVE EFFLUENTS "" BASES i l Gaseous Effluents in Routine Releases from Light-Water-Cooled l
, Revision 1, July 1977. These equations also provide for detennining the ; actual doses based upon the historical average atmospneric conditions.
The release rate specifications for radiciodines, radioactive material
;in particulate form and radionuclides other than noble gases are dependant ion the existing radionuclide pathways to man, in the unrestricted area.
The pathways which are examined in the development of these calculations
,are:
- 1) individual inhalation of airborne radionuclides, 2) deposition jof radionuclides onto green leafy vegetation with subsequent consumption
'by man, 3) deposition onto grassy areas where milk animals and meat pro-ducing l4) deposition on the ground with subsequent exposure of man.
3/4.11.2.4 GASE0US WASTE TREATMENT The operability of the gaseous radwaste treatment system and the ventila-tion exhaust treatment systems ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the .-- requirements of 10 CFR Part 50.36a, General Design Criterion Section 11.1 of the Final Safety Analysis Report for The Donald C. Cook Nuclear Plant and design objective Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous effluents. 3/4 11.2.5 Exolosive Gas Mixture i This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treat-ment system is maintained below the flamability limits of hydrogen and oxygen mixtures. Maintaining the concentration of hydrogen or oxygen below their flammability limits provides that the releases of radioactive materials will be controlled in confonnance with the requirements of the General Design Criterion specified in Section 11.1 of the Final Safety l Analysis Report for the Donald C. Cook Nuclear Plant. i l 3/4 11.2.6 Gas Storage Tanks l Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release l of the tank's contents, the resulting total body exposure to an in- ! dividual at the nearest site boundary will not exceed 0.5 rem. This is consistant with Standard Review Plan 15.7.1. " Waste Gas System Failure." i D. C. COOK - UNIT 2 33/411-4 -
p- / 4 \ . RADI0 ACTIVE EFFLUENTS i ! BASES
' i' 3/4.11.3 SOLID RADI0 ACTIVE WASTE
.? - , ,
The operability of the solid radwaste system ensures that thb system ' will be available for use whenever solid radwastes require processing and packaging prior to shipment offsite. This specification implements - the requirements of 10 CFR Part 50.36a and General Design Criterton , specified in Section 11.1 of the Final Safety A1alysis Report for the ' , Donald C. Cook Nuclear Plant. The process parameters included jn establishing the PROCESS CONTROL PROGRN1 may include, but are not limit-ed to waste type, waste pH, waste / liquid / solidification agent / catalyst ratios, waste oil content, waste principal chemical. constituents, mixing and curing time. , 3/4.11.4 TOTAL DOSE The specification is provided to meet the dose limitations of 40 CFR 190. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioective effluents exceed twice the design objective doses of Appendix I. Fcr, sites containing up to 4 reacto s, it is highly unlikely that the resultar.t dose to a member of the public will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action which should result in the limitations of dose to a member of the public for 12 consecutive months to within the 40 CFR 190 limits. For the purposes o" the Special Report, it may be assumed that t3e dose ccmmit-ment to the member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be con-sidered. If the dose to any member of the public is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violati]n of 40 CFR 190 have not already been corrected), in accordance with the provision of 40 CFR,190.ll is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staf f action is conpleted. An individual is not considered a member of the public during any period in which he/she is engaged fr carrying out any operation ;ich is part of the nuclear fuel cycle. D. C. COOK - UNIT 2 B 3/4 11-5 esineem
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i. 3/4.12 RA010LOGXCAL ENV1RONMENTAL MONXTORANG a MSES /
. 3/[.12.1 ft0NITORING PROGRAM
}' The radiological monitoring program required by this specification provides measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of individuals resulting from the station operation. This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measurable concentration of radioactiye materials and levels of radiation are not higher than
; expected on~the basis 'of the effluent measurements and modeling of the 3 environmental; exposure pathways. The initially specified monitoring bs program will be effective for at least the first three years of commercial
,, operation. ~ Following this period, program changes may be initiated
[ 2 based on operational experience. i The detection capabilities roquired by Table 4.12-1 are the state-of-i
? the art for routine environmental measurements in industrial laboratories, f
5 It should be recognized that the LLD is defined an an a priori (before the fact) limit representing the capability of a measurement system and 4 not as a posteriori (after the fact) limit for a particular measurement. _ l Analyses shall be performed in such a manner that the stated LLDs will - be achieved under routine conditions. Occasionally background fluctuations, i unavoidably small sample sizes, the presence of interferring nuclides,
, or other uncontrollable circumstances may render these LLDs unachievable.
l In such cases, the contributing factors will be identified and described 1 in the Annual Radiological Environmental Operating Report. l-3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of unrestricted areas -are identified and that modifications to the monitoring
~
program are made if required by the results of this census. This census , satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 500 square feet provides assurance that.significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quar.tity (25 kg/ year) of leafy vegetablesassumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were used,1) that 20% of the garden was used for growing broad leaf vegetation (i.e. similiar to lettuce and cabbage), and 2) a vegetation field of 2 t kg/ square meter. 3/4.12.3 INTERLABORATORY COMPARISON- PROGRAM The requirement for participation in an Interlaboratory Comparison Program is orovided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the qJality assurance progCam for environmental monitoring in order to demonstrate that the results are reasonably valid. D. 'C. COOK - UNIT 2 8 3/4 12-1
5.0 DESIGN FEATURES 5.1 SITE Exclusion Area 5.1.1 The exclusion area shall be shown in Figure 5.1-1. Low Population Zone 5.1.2 The low population zone shall be shown in Figure 5.1-2. Site Boundary For Gaseous and Liquid Effluents 5.1.3 The site boundary for gaseous and liquid effluents shall be shown in Figure 5.1-3. 5.2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:
- a. Nominal inside diameter = 115 feet,
- b. Nominal inside height = 160 feet.
- c. Minimum thickness of concrete walls = 3' 6".
- d. Minimum thickness of concrete roof = 2' 6".
- e. Minimum thickness of concrete floor pad = 10 feet.
f. Nominal thickness of steel liner = 3/8 inches.
- g. 6 Net free volume = 1.24 x 10 cubic feet.
DESIGN PRESSURE AND TEMPERATURE
- 5. 2. 'i The reactor containnent building is designed and snall be maintained in accordance with the original design provisions contained in Section 5.2.2 of the FSAR.
D. C. COOK - UNIT 2 5-1
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6.0 ADMINISTRATIVE CONTROLS l 6.3 FACILITY STAFF QUALIFICATIONS i ! 6.3,1 Each member of the facility scaff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Radiation Protection Supervisor who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and (2) the Shif t Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in Plant design, and response and analysis of the Plant for transients and accidents. 6.3.2 Until the newly appointed Operations Superintendent obtains a Senior Reactor Operator's License, all of his licensed functions will be performed by a full time assistant who holds a current Senior Reactor Operator's License. 6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Coordinator and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55. 6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Plant Manager and shall meet or exceed the requirements of Section 27 of the NFPA Code-1976. D. C. COOK - UNIT 2 6-5
ADMINISTRATIVE CONTROLS 6.5 REVIEW AND AUDIT 6.5.1 PLANT NUCLEAR SAFETY REVIEW COMMITTEE (PNSRC) FUNCTION 6.5.1.1 The PNSRC shall function to advise the Plant Manager on ail matters related to nuclear safety. COMPOSITION l 6.5.1.2 The PNSRC shall be composed of the: Chairman: Plant Manager or designated alternate Member: Assistant Plant Managers Member: Operations Superintendent Member: Technical Superintendent Member: Maintenance Superintendent Member: Control and Instrument Supervisor Member: Nuclear / Computer Engineering Supervisor Member: Plant Chemical Supervisor Member: Performance Supervising Engineer -- -- Member: Plant Radiation Protection Supervisor i Member: Shift Supervisor Member: Environmental Coordinator ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the PNSRC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in PNSRC activities at any one time. MEETINC FREQUENCY, 6.5.1.4 The PNSRC shall meet at least once per calendar month and as convened by the PNSRC Chairman or his designated alternate. OUORUM i 6.5.1.5 A quorum of the PNSRC shall consist of the Chairman or his designated alternate and four members including alternates.. D. C. COOK - UNIT 2 6-6
ADMINISTRATIVE CONTROLS RESPONSIBILITIES 6.5.1.6 The PNSRC shall be responsible for: 1
- a. Review of (1) all procedures required by Specification 6.8 and changes thereto, (2) any other proposed procedures or changes thereto as determined by the Plant Manager to affect nuclear safety.
- b. Review of all proposed tests and experiments that affect nuclear safety.
- c. Review of all proposed changes to Appendix "A" Technical Specifications.
- d. Review of all proposed changes or modifications to Plant systems or equipment that affect nuclear saf ety.
- e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Chairman of the NSDRC.
- f. Review of events requiring 24 hours written notification to the Commission.
- g. Review of facility operation to detect potential nuclear safety hazards,
- h. Performance of special reviews, investigations or analysis and reports thereon as requested by the Chairman of the NSDRC.
- 1. Review of the Plant Security Plan and implementing procedures and shall submit recocmended changes to the Chairman of the N SDRC .
- j. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Chairman of the NSDRC.
- k. Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the NSDRC.
- 1. Review of changes to the PROCESS CONTROL PROGRAM, OFFSITE DOSE CALCULATION MANUAL, and radwaste treatment systems.
D . C . COOK - UNIT 2 6-7
ADMINISTRATIVE CONTROLS - AUTHORITY 6.5.1.7 ThePdSRCshall: *
- a. Recommend to the Plant Manager written approval or disapproval of items considered under 6.5.1.6 (a) through (d) above.
- b. Render determinations in writing with regard to whether or not each item considered under 6.5.1.6 (a) through (e) above constitutes an unreviewed safety question.
- c. Provide written notification within 24 hours to the NSDRC of disagreement between the PNSRC and the Plant Manager; however, the Plant Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
RECORDS 6.5.1.8 The PNSRC shall maintain written minutes of each meeting and copies shall be provided to the Chairman of the NSDRC. 6.5.2 NUCLEAR SAFETY AND DESIGN REVIEW COMMITTEE (NSDRC) _, FUNCTION l 6.5.2.1 The NSDRC shall function to provide independent review and i audit of designated activities in the areas of: 1
- a. nuclear power plant operations l
l b. nuclear engineering
- c. chemistry and radiochemistry
- d. metallurgy
- l
- e. instrumentation and control
- f. radiological safety
- g. mechanical and electrical engineering
- h. quality assurance practices D. C. COOK - UNIT 2 6-8 -
4
ACMitlIST3AT!VE CCttTROLS CCMPOSITI0t1 6.5.2.2 The flSCRC shall be c:mposed of the: Chair: nan: Assistant Vice President, Nuclear Engineering Me:sber: Vice Chairman, Engineering and Construction Member: President and Chief Operating Officer of I&MECo Member: Executive Vice President, Construction and New York Engineering Member: Vice President, Mechanical Engineering Member: Vice President, Electrical Engineering Member: Vice President, Engineering Mministration Member: Assistant Vice President, Design Division Member: Assistant Vice President, Environmental Engineering Division Member: Plant Manager, D. C. Cook Plant Member: Manager, Nuclear Safety and Licensing Section Alternate: Assistant Chief Mechanical Engineer Alternate: Assistant Plant Manager, D. C. Cook Plant Alternate: Executive Assistant to the President of I&MECo Alternate: Assistant Division Manager, Nuclear Engineering ALTERelATES 6.5.2.3 All alternate memcers snall be accointed in writine by tne t:SCRC Chairman to serve on a temocracy basis; however, no more than two alternates shall participate as voting memcers in ::50RC activities at any cne time. , CC?tSUL7;t:75 6.5.2.4 Consultants snall be utili:ec as de:erminec Oy :ne ?!50RC Chain.an
- o provice ex:er; acvice to the t150RC.
MEET *'G FRECUEt!CY 6.5.2.5 The '50RC shall mee; at least once rer ca:encar cuarter durine :ne initial year of facility aceration folicw1ng fuel leacing anc a: leas:~ nce er s*x mon:ns :neraafter. C. C. CCCK - Ut'IT 2 53
ACHINISTRATIVE C0tlTROLS l 1 AUDITS 6.5.2.8 Audits of facility activities shall be performed under the i cognizance of the NSDRC. These audits shall enccmpass:
- a. The confomance of facility operation to provisions contained within the Technical Specification and applicable license conditions at least once per 12 months.
- b. The perfomance, training and qualifications of the entire facility staff at least once per 12 months.
- c. The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least
, once per 6 months. ,
- d. The perfomance of activities recuired by the Ouality Assurance' Procram to meet the criteria of Aapencix "3",10 CFR 50, at least once per 24 montns.
- e. The Facility Emergency Plan and implementing procedures at least once per 24 months.
- f. The Facility Security Plan and implementing procedures at least once per 24 months.
- g. Any other area of facility operation considered appropriate by the N$0nc. .
i
- h. The Facility Fire Protection program and implementing procedures at least once per 24 months.
f. An independent fire protection and loss prevention pr: gram inspection and audit shall be perfomed at least once per 12 months' utilizing either qualified offsite licensee perscnnel or an cutside fire protecticn firm.
- j. An inspection and audit of the fire protection and loss pre-vention program shall be perfomed by a qualified outside fire consultant at least once per 36 mcnths.
k The radiological environmental menitoring pr: gram and the results thereof at least once per 12 months.
- 1. The OFFSITE 00SE CALCULATION MANUAL and implementing precedures at least once per 24 months.
D. C. COOK - UNIT 2 5-11 \
i ADMINISTRATIVE CONTROLS
- m. The PROCESS CONTROL PROGRAM and implementing procedures for solidification of radioactive wastes at leiast'once per 24 months.
- n. The performance of activities required by the Quality Assurance Program to meet the criteria of Regulatory Guide 1.21 Rev. 1, June 1974 and Regulatory Guide 4.1, Rev.1, April 1975 at least once per 12 months.
AUTHORITY . 6.5.2.9 The NSDRC shall report to and advise the Vice Chairman, Engineering and Construction, AEPSC, on those areas of responsibility specified in Section 6.5.2.7 and 6.5.2.8. RECORDS 6.5.2.10 Records of the NSDRC activities shall be prepared, approved and distributed as indicated below:
- a. Minutes of each NSDRC meeting shall be prepared, approved and forwarded to the Vice Chairman, Engineering and Construction, AEPSC, within 14 days following each meeting.
- b. Reports of reviews encompassed by Section 6.5.2.7 above, shall be prepared, approved and f orwarded to the Vice Chairman,
- Engineering and Construction, AEPSC, within 14 days following completion of the review. .
- c. Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the Vice Chairman, Engineering and Construction, AEPSC, and to the management positions responsible for the areas audited within 30 days af ter completion of the audit.
6.6 REPORTABLE OCCURRENCE ACTION 6.6.1 The following actions shall be taken for REPORTABLE OCCURRENCES: a The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.
- b. Each REPORTABLE OCCURRENCE requirir.g 24 hour notification to the Co= mission, shall be reviewed by PNSRC and submitted to the NSDRC Chairman.
G D. C. COOK - UNIT 2 6- 10
.- , , _, - - -- . . - - - - . _ . - .. -_. m,
ADMINISTRATIVE CONTROLS 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
- a. The f acility shall be placed in at least HOT STANDBY within one hour.
- b. The Safety Limit violation shall be reported to the Commissica and to the Chairman of the NSDRC within 24 hours,
- c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PNSRC. This report shall describe (1) the applicable circumstances preceding the violation, (2) effects of the violation upon f acility components, systems or structures, and (3) corrective action taken to prevent recurrence.
- d. The Safety Limit Violation Report shall be submitted to the Commission, the Chairman of the NSDRC, and the Executive Vice President-Construction and New York Engineering AEPSC within 14 days of the violation.
6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:
- a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, November 1972.
- b. Refueling operations.
- c. Surveillance and test activities of safety related equipment.
- d. Security Plan Lnplementation.
- e. Emergency Pir.n implementation.
- f. Fire Protection Program implementation.
- g. PROCESS CONTROL PROGRAM implementation.
- h. OFFSITE DOSE CALCULATION MANUAL implementation.
- 1. Quality Assurance Program f or effluent and environmental monitoring using the guidance in Regulatory Guide 1.21, Rev.1, June 1974 and Regulatory Guide 4.1, Rev.1, April 1975.
6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be reviewed by the PNSRC and approved by the Pisnt Manager prior to implementation and reviewed periodically as set forth in administrative procedures. D. C. COOK - UNIT 2 6-13 __ ~ _ _ - - - _ - _ - - . - . _ . _ . , _ --
ADMINISTRATIVE CONTROLS
, ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 3/
6.9.l.6 Routine radiological environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. 6.9.1.7 The annual radiological environmental operating reports shall include summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveil-lance reports and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of the land use censuses required by Specification 3.12.2. If harmful effects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem. The annual radiological environmental operating reports shall include - summarized and tabulated results in the format of 3.12-2 of all -- ~ radiological environmental samples taken during the report period. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining tne reasons for the missing results. The missing data shall be submitted as soon as - possible in a supplementary report. The reports shall also include the following: a summary description of the radiological environmental monitoring program including sampling methods for each sample type, size and physical characteristics of each sample type, sample preparation methods, analytical methods, and measuring equip-ment used; a map of all sample locations keyed to a table giving distances and directions from one reactor; the result of land use census required by the Specification 3.12.2; and the results of participation in the Interlaboratory Comparison Program required by Specification 3.12.3. 3/ A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specifiy the releases of radioactive material for each unit. D. C. COOK - UNIT 2 6-16
ADMINISTRATIVE CONTROLS I i SEMI-ANNUAL RADI0 ACTIVE EFFLUENT RELEASE rep 0RT I I 6.9.1.8 Routine radioactive effluent release reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. 6.9.1.9 The radioactive effluent release reports shall include a sumary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the units as outlined in Regulatory Guide 1.21,
" Measuring, Evaluating and Reporting in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water Cooled Nuclear Power Plants," with data sumarized on a quarterly basis following the fomat of Appendix B, thereof.
The radioactive effluent release report to be submitted 60 days after January 1 and July 1 of each year shall include a quarterly sumary of hourly meteoregical data collected during the reporting period. This sumary may be either in the form of an hour-by-hour listing of wind speed, wind direction, atmospheric stability, and precipitation (if measured) on magnetic tape, or in the fom of joint frequency distributions of wind . . speed, wind direction and atmospheric stability. The report submitted 60 days after January 1 shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. These reports shall include as assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary (Figure 5.1-3) during the reporting period. All assumptions used in making these assessments (i.e., specific activity, exposure time and location) shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in effluents (as determined by sampling frequency and measurement) gaseousshall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the Offsite Dose Calculation Manual (00CM). 3/ A single submittal may be made for a multiple unit station. The submittal should combine those sections that are comon to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. D. C. COOK - UNIT 2 6-17
ADMINSTRATIVE CONTROLS The radioactive effluent release report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show confomance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev.1. The radioactive effluent release report shall include the following infomation for each type of solid waste shipped offsite during the report period:
- a. Volume (cubic meters),
- b. Total curie quantity (specify whether detemined by measurement or estimate),
- c. Principal radionuclides (specify whether detemined by measurement or estimate),
- d. Type of waste (e.g'. , spent resin, compacted dry waste, evaporator bottoms) , _-
- e. Type of container (e.g., LSA, Type A, Type B, large Quantity), '
and .
- f. Solidification agent (e.g. , cement).
The radioactive effluent release report shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluent on a quarterly basis. The radioactive effluent release reports shall include any change to the PROCESS CONTROL PROGRAM (PCP) and the OFFSITE DOSE CALCULATION MANUA (ODCM) made during the reporting period. MONTHLY REACTOR OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office Of Management and Program Analysis, U.S. Nuclear Regulatory Comission, Washington, D.C. 20555, with a copy to the Regional Office of Inspection and Enforcement, no later than the 15th of each month following the calendar month covered by the report. REPORTABLE OCCURRENCES 6.9.1.11 The REPORTABLE OCCURRENCES of Specification 6.9.1.12 and 6.9.1.13 below, including corrective actions and measures to prevent recurrence, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date. D. C. COOK - UNIT 2 6-18
ADMIMISTRATIVE CONTROLS
. PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP i :6.9.1.12 The types of events listed below shall be reported within 24 hours j by telephone and confirmed by telegraph, mailgram or facsimile transmission to the Director of the Regional Office, or his designate no later than the ifirst working day following the event, with a written followup report within 14 days. ;
The written followup report shall include, as a minimum a completed copy of a licensee event report form. Information provided on the licensee ; event report form shall be supplemented, as needed, by additional narrative naterial to provide complete explanation of the circumstances surrounding the event. a. Failure of the reactor protection system or other systems, subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reaches the setpoint specified as the limiting safety system setting in the technical specifications or failure to complete the required protective function. b. Operation of the unit or affected systems when any parameter or operation subject to a limiting condition for operation is less - - conservative than the least conservative aspect of the limiting condition for operation established in the technical specifications. -- --
- c. Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment.
- d. 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 conservative than specified in the technical specification; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if subcritical, an unplanned reactivity insertion of more tnan 0.5%ak/k; or occurrence of any unplanned criticality,
- e. 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.
f. Personnel error or procedural inadequancy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the SAR. 4 D. C. COOK - UNIT 2 6-19 i
~
ADMINISTRATIVE CONTROLS
- g. Conditions arising from natural or man-made events that, as a direct result of the event require unit shutdown, operation of safety systems, or other protective measures required by technical specifications.
- h. Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the technical specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.
i. 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 analysis in the safety analysis report or technical specification bases; or discovery during unit life of conditions not specifically considered in the safety analysis report or technical specifications that require remedial action or corrective measures to prevent the existence or development of an unsafe condition. - j. Offsite releases of radioactive material in liquid and gaseous effluents which exceed the limits of Specification 3.11.1.1 or - - __ 3.11.2.1.
- k. Exceeding the limits in Specification 3.11.1.4 or 3.11.2.6 for the storage of radioactive materials in the listed tanks. The written follow-up report shall include a schedule and a description of activities planned and/or taken to reduce the contents to with-in the specified limits.
D. C. COOK - UNIT 2 6-20 *
. 1 ADMINISTRATIVE CONTROLS i l l THIRTY DAY WRITTEN rep 0RTS 6.9.1.13 The types of events listed below shall be the subject of written reports to the Director of the Regional Office within thirty days of occurence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explana-tion of the circumstances surrounding the event.
- a. Reactor protection system or engineered safety feature in-strument settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems.
- b. Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown re-quired by a limiting condition for operation.
- c. Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems.
- d. Abnormal degradation of systems other than those specified in 6.9.1.12.c above designed to contain radioactive' material resulting from the fission process.
- e. An unplanned offsite release of 1) more than 1 curie of radio-active material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radiciodine in gaseous effluents. The report of an unplanned offsite release of radioactive material shall include the following infonnation:
- 1. A description of the event and equipment involved.
- 2. Causes(s) for the unplanned release.
- 3. Actions taken to prevent recurrence.
- 4. Consequences of the unplanned release,
- f. Measured levels of radioactivity in an environmental sampling medium detennined to exceed the reporting level values of Table 3.12-2 when averaged over any calendar quarter sampling period. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the co'ndition shall be reported and described in the Annual Radiological Environmental Operating Report.
D. C. COOK - UNIT 2 6-21
ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director of the 3ffice of Inspection and Enforcement Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of
,the applicable reference specification:
- a. ECCS Actuation, Specifications 3.5.2 and 3.5.3.
- b. Inoperable Seismic Monitoring Instrumentation, Unit No. 1, Specification 3.3.3.3.
- c. Inoperable Meteorological Monitoring Instrumentation, Unit No.
1, Specification 3.3.3.4.
- d. Fire Detection Instrumentation, Specification 3.3.3.8.
- e. Fire Suppression Systems, Specifications, 3.7.9.1, 3.7.9.2, 3.7.9.3 and 3.7.9.4.
- f. Seismic Event Analysis, Specification 4.3.3.3.2.
D. C. COOK - UNIT 2 6-22
ADMINISTRATIVE CONTROLS 6.10 RECORD RETENTION 6.10.1 The fc11oving records shall be retained for at least five
- years
- a. Records and logs of unit operation covering time interval at each power level.
- b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety,
- c. ALL REPORTABLE OCCURRENCES submitted to the Commission.
- d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
- e. Records of changes made to Operating Procedures.
- f. Records of sealed source and fission detection leak tests and results.
- g. Records of annual physical inventory of all sealed source material on record.
6.10.2 The following records shall be retained for the duration of the Facility Operating License:
- a. Records and drawing changes reflecting unit design l modifications made to systems and equipment described in the Final Safety Analysis Report.
,l b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
- c. Records of radiation exposure for all individuals entering radiation control areas.
- d. Records of gaseous and liquid radioactive material released to the environs.
~
- e. Records of transient or operational cycles for those f acility components identified in Table 5.7-1.
! f. Records of reactor tests and experiments. ! g. Records of training and qualification for current members of the Plant staff.
- h. Records of in-service inspections performed pursuant to these Technical Specifications.
- i. Records of Quality Assurance activities required by the QA Manual,
- j. Records of reviews performed for changes made to procedures or equipment or review of tests and experiments pursuant to 10
+ CFR 50.59.
- k. Records of meetings of the PNSRC and the NSDRC.
- 1. Records for Environmental Qualification which are covered j under the provisions of paragraph 6.13.
m. Records of radioactive shipments. l D. C. COOK - UNIT 2 6-23 1 2
ADMINISTRATIVE CONTROLS 6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure. 6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is 1000 mram/hr or less shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit *. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.
- c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device.
This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiation Work Permit. 6.12.2 The requirements of 6.12.1, above, shall also apply to each l high radiation area in which the intensity of radiation is greater than 1000 mram/hr. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or the Plant Henith Physicist. l
- Health Physics personnel shall be exsapt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas.
I D. C. COOK - UNIT 2 6-24
ADMINISTRATIVE CONTROLS 6.13 ENVIRONMENTAL QUALIFICATION 6.13.1 By no later than June 30, 1982 all safety-related electrical equipment in the f acility shall be qualified in accordance with the provisions of: Division of Operating Reactors " Guidelines for Evaluating Environmental Qualification of Class lE Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588 " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment". December 1979. Copies of these documents are attached to Order for Modification of License No. DPR-58 dated October 24, 1980. 6.13.2 By no later than December 1, 1980, complete and auditable records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained as current as equipment is replaced, further tested, or otherwise further qualified. I 1, i i i l l l l t i i D. C. COOK - UNIT 2 6-25 1
ADMINTSTRATTVE CONTROLS 6.14 PROCESS CONTROL PROGRAM (PCP) f i6.14.1 The PCP shall be approved by the Ccmmission prior to implementation.
- 6.14.2 Licensee initiated changes to the PCP:
- 1.
j Shall be submitted to the Commission in the semi-annual Radioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:
- a. Sufficiently detailed infomation to totally support the rationale for the change without benefit of additional or supplemental infomation;
- b. A determination that the change did not reduce the overall confomance of the solidified waste product to existing criteria for solid wastes; and
- c. Documentation of the fact that the change has been reviewed and found acceptable by the PNSRC.
- 2. Shall become effective upon review and acceptance by the PNSRC. --
6.15 0FFSITE DOSE CALCULATION MANUAL (00CM) 6.15.1 The ODCM shall be approved by the Ccmission prior to implementation. 6.14.2 Licensee initiated changes to the ODCM: 1. Shall be submitted to the Commission in the Semi-Annual Radioactive Effluent Release Report in the next report after the report period the change (s) was made effective. This submittal shall contain: a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change (s); b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and
- c. Documentation of the fact that the change has been reviewed and found acceptable by the PNSRC.
D. C. COOK - UNIT 2 6-26
ADMINISTRATIVE CONTROLS h
- 2. Shall become effective upon review and acceptance by the PNSRC.
6.15.3 Commission initiated changes:
- 1. Shall be determined by the PNSRC to be applicable to the facility after consideration of facility design.
- 2. The licensee shall provide the Commission with written noti-fication of-their determination of applicability including any necessary revisions to reflect facility design.
6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous and Solid) 6.16.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid): 1. Shall be reported to the Commission in the Annual Operating Report for the period in which the evaluation was reviewed by the (PNSRC) The discussions of each change shall contain:
- a. A summary of the evaluation that led to the determination that -
~~
the change could be made in accordance with 10 CFR 50.59;
- b. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; s
- c. A detailed description of the equipment, components and processes involved and the interfaces with other plant systems;
- d. An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted,in the license application and amendments thereto;
- e. An evaluation of the change which shows the expected maximum exposure to individuals in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto; f.
A comparison of the predicted releases of radioactive materials, in liquid and caseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made; D. C. COOK - UNIT 2 6-27 s -
~
ADMINISTRATIVE CONTROLS
- g. An estimate of the exposure to plant operating personnel as a result of the change; and
- h. Documentation of the fact that the change was reviewed and found acceptable by the PNSRC.
- 2. Shall become effective upon review and acceptance by the PNSRC.
6.16.2 Comission initiated changes:
- 1. The applicability of the change to the facility shall be determined by the (PNSRC) after consideration of the facility design.
- 2. The licensee shall provide the Comission with written notification of its determination of applicability including any necessary revisions to reflect facility design.
D. C. COOK - UNIT 2 6-28
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