ML17334A963
| ML17334A963 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/19/1986 |
| From: | INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | |
| Shared Package | |
| ML17334A962 | List: |
| References | |
| GL-83-37, NUDOCS 8605230211 | |
| Download: ML17334A963 (33) | |
Text
REACTOR COOLANT, SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE: 'a~One of the containment atmosphere particulate radioactivity monitoring channels (ERS-1301 or ERS-1401), b.The containment, sump flow monitoring system, and c~Either the containment humidity monitor o" one of the containment atmosphere gaseous radioactivity monitoring channels (ERS-1305 or ERS-1405).APPLICABILITY:
MODES 1, 2, 3 and 4.ACTION: With only two of the above required leakage detection systems OPERABLE, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the recuired gaseous and/cr particulate radioactivity monitoring channels are inoperable; otherwise, be in at least HOT STANDBY within the ne'xt 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.SURVEILLANCE RE UIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by: 'a~Containment atmosphere particulate and gaseous (if being used)monitoring system-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, b.Containment sump flow monitoring system-performance of CHANNEL CALIBRATION at least once per 18 months, c~Containment humidity monitor (if being used)-performance of CHANNEL CALIBRATION at least once per 18 months.D.C.COOK-UNIT 1 3/4 4-14~Amendment No.
I C l 1 REACTOR COOLANT SYSTEM This page intenzicnally left blank.D.C.'OOK-UNIT 1 3/4 4-15 Amendment No.
ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator 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~b.c~Inservice Inspection Program Review, Specification 4.4.10.ECCS Actuation, Specifications 3.5.2 and 3.5.3.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 in excess of limits, Specification 4.7.7.1.3.
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.i.Containment Sump Level instrumentation, Table 3.3-11.D.C.COOK-UNIT 1 6-19 Amendment No.
INSTRUMENTATION BASES 3/4.3.3.7 FIRE DETECTION INSTRUMENTATION OPERABILITY of the f're 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 facility fire protection program.In the event that a portion of the fire detection instrumentation is inoperable, the establishment o'f 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 ccntainment fire detection is inoperable.
3/4.3.3.8 POST-ACCiDENT INSTRUMENTATION The OPERABILITY of the post-accident instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident.The containment water level CHANNEL CHECK is a visual inspection of parallel channels which should indicate within acceptable instrument drift that the water level is below the range of the ccntainment water level instrumentation.
Acceptable instrument drift for ccntainment water level instrumentation is presently considered 25%of full scale.The containment sump level channels should indicate the same level on both channels within acceptable instrument drift, which is presently considered a 25%of full scale difference between the two parallel channels.If the channels do not indicate the same level, the containment sump pump actuation and shut-off can be used to indicate if either channel is correct.The drift for both instrumentation systems is attributed to air accumulation in the capillaries.
Provided that the drift is less than 25%, it should not prevent the instrument frcm tracking changes in level.Equipment changes may change the acceptable drift.Such a change will not constitute violation of this T/S, provided appropriate evidenc'e exists to justify the change.D.C.COOK-UNIT 1 B 3/4 3-4 Amendment No.
POS-CCIDENT I STRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The post-accident monitoring instrumentation channels shown in Table 3.3-11 shall be OPERABLE.APPLICABILITY:
MODES 1, 2, and 3.ACTION: With the number of OPERABLE post-accident monitoring channels less than required by Table 3.3-11, either restore the inoperable channel to OPERABLE status within 30 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except where noted in Table 3.3-11.b.The provisions of Specifications 3.0.4 are not applicable.
SURVEILLANCE RE UIREMENTS 4.3.3.8 Each post-accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7.D.C.COOK-UNIT 1 3/4 3-54 Amendment No.
TABLE 3.3-ll POST-ACCIDENT MONITORING INSTRUMENTATION INSTRUMENT HINIHUH CHANNELS OPERABLE 1.Containment Pressure 2.3.4.Reactor Coolant Outlet Temperature
-T (Wide Range)HOT Reactor Coolant Inlet Temperature
-T (Wide Range)COLD Reactor Coolant Pressure-Wide Range 5.Pressurizer Water Level 6.Steam Line Pressure 7.Steam Generator Water Ievel-Na'rrow Range 8.Refueling Water Storage Tank Water Level 9.Boric Acid Tank Solution Level 10.Auxiliar Feedwater Flow Rate Y ll.Reactor Coolant System Subcooling Hargin Monitor 12.PORV Position Indicator-Limit Switches***
13.PORV Block Valve Position Indicator-Limit Switches 14.Safety Valve Position Indicator-Acoustic Monitor 15.Containment Sump Level 16.Containment Water Level 2/Steam Generator 1/Steam Generator 1/Steam Generator*
1**1/Valve 1/Valve 1/Valve 1¹Steam Generator Water Level Channels can be used as a substitute for the corresponding auxiliary feedwater flow rate channel instrument.
PRODAC 250 subcooling margin readout can be used as a substitute for the subcooling monitor instrument.
¹With less than the minimum number of channels OpERABLE restore the system to OPERABLE status within 30 days or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 14 days outlining available backup equipment, the cause of the inoperability and the plans and schedule for restoring..the system to OPERABLE status.
TABLE 4.3-7 POST-ACCIDENT MONITORING INSTRUHENTATION SURVEILLANCE RE UIREHENTS INSTRUMENT CHANNEL CHECK CHANNEL CALIBRATION 1.Containment Pressure 2.Reactor Coolant Outlet Temperature
-T (Wide Range)HOT 3., Reactor Coolant Inlet Temperature
-T (Wide Range)COLD 4.Reactor Coolant Pressure-Wide Range 5.Pressurizer Water Level 6.Steam Line Pressure 7.Steam Generator Water Level-Narrow Range 8.RWST Water Level 9.Boric Acid Tank Solution Level 10.Auxiliary Feedwater Flow Rate 11.Reactor Coolant System Subcooling Margin Monitor 12.PORV Position Indicator-Limit Switches 13.PORV Block Valve Position Indicator-Limit Switches 14.Safety Valve Position Indicator-Acoustic Monitor 15.Containment Sump Level 16.Containment Water Level INS E 0 POST-CCIDENT INSTRUMENTATION LIMITI G CO DITION FOR OPERAT ON 3.3.3,6 The post-accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.APPLICABILITY:
MODES 1, 2, and 3.~ACT ON: With the number of OPERABLE post-accident monitoring channels less than required by Table 3.3-10, either restore the inoperable channel to OPERABLE status within 30 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except where noted in Table 3.3-10'.The provisions of Specifications 3.0.4 are not applicable.
SURVEILLANCE RE UIREME TS 4.3.3.6 Each post-accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-10.D.C.COOK-UNIT 2 3t4 3-45 Amendment No.
TABLE 3.3-10 POST-ACCIDENT MONITORING INSTRUMENTATION INSTRUMENT MINIMUM CHANNELS OPERABLE 1.Containment Pressure 2.3.4 5.Reactor Coolant Outlet Temperature
-T (Wide Range)Reactor Coolant Inlet Temperature
-T (Wide Range)Reactor Coolant Pressure-Wide Range Pressurizer Water Level'.Steam Line Pressure 7.Steam Generator Water Level-Narrow Range 8.Refueling Water Storage Tank Water Level 9.Boric Acid Tank Solution Level 10.Auxiliary Feedwater Flow Rate 11.Reactor Coolant System Subcooling Margin Monitor 12.PORV Position Indicator-Limit Switches"-**
13.PORV Block Valve Position Indicator-Limit Switches 14.Safety Valve Position Indicator-Acoustic Monitor 15.Containment Sump Level 16.Containment Water Level 2/Steam Generator 1/Steam Generator 1/Steam Generator*
1/Valve 1/Valve 1/Valve*Steam Generator Water Level Channels can be used as a substitute for the corresponding auxiliary feedwater flow rate channel instrument.
.**pRODAC 250 subcooling margin readout can be used as a substitute for the subcooling monitor instrument.
With less than the minimum number of channels OPERABLE restore the system to OPERABLE status within 30 days or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 14 days outlining available backup equipment, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
TABLE 4.3-10 POST-ACCIDENT HONITORING INSTRUMENTATION SURVEILLANCE RE UIREHENTS INSTRUMENT CHANNEL CHECK CHANNEL CALIBRATION 1.Containment Pressure 2.3.5." Reactor Coolant Outlet Temperature
-T (Wide Range)HOT Reactor Coolant Inlet Temperature
-T (Wide Range)COLD Reactor Coolant Pressure-Wide Range Pressurizer Water Level R 6.Steam Line Pressure 7.Steam Generator Water Level-Narrow Range 0.RWST Water Level 9.Boric Acid Tank Solution Level 1 0.Auxzlzary Feedwater Flow Rate ll.Reactor Coolant System Subcooling Margin Monitor 12.PORV Position Indicator-Limit Switches 13.PORV Block Valve-Position Indicator-Limit Switches" 14.Safety Valve Position Indicator-Acoustic Monitor 15.Containment Sump Level 16.Containment Water Level*The provisions of Specification 4.0.6 are applicable.
REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1"The following Feactor Coolant System leakage detection systems shall be OPERABLE: a.One of the containment atmosphere particulate radioactivity monitoring channels (ERS-2301 or ERS-2401), b.The containment sump flow monitoring system, and c.Either the containment humidity monitor or one of the containment atmosphere gaseous radioactivity monitoring channels~(ERS-2305 or ERS-2405).
APPLICABILITY:
MODES 1, 2, 3 and 4.ACTXON: With only two of the abcve recuired leakage detection systems OPERABLE, operation may continue for up tc 30 days provided crab samples of the containment atmosphere a e obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required gasecus and/cr particulate radioactivity monitoring channels are inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.SURVEILLANCE RE UIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by: a~Containment atmosphere particulate and gaseous (if being used)monitoring system-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, b.Containment sump flow monitoring system-performance of CHANNEL CALIBRATION at least once per 18 months,*C~Containment humidity monitor (if being used)-performance of CHANNEL CALIBRATION at least once per 18 months.*The provisions of Specification 4.0.6 are applicable.
D.C.COOK-UNIT 2 3/4 4-14 Amendment No.
ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator 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 0 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.Seismic Event Analysis, Specification 4.3.3.3.2.
g.Sealed Source leakage in excess of limits, Specification 4.7.8.1.3.
h.Containment Sump Level instrumentation, Table 3.3-10.D.C.COOK-UNIT 2 6-19 Amendment No.
1 3/4~3 INSTRUMENTATZ BASES 3/4.3>3~2 MOVABLE ZNCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained for use of this system accurately represent the spatial neutron flux distribution of the reactor core.The OPERABILITY of this system is demonstrated by irradiating each detector used and normalizing its respective output.3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event, and evaluate the response of those features important to safety.This capability is required to permit comparison of the measured response to that used in the design basis for the facility.3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere.
This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public.3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATZON The OPERABILITY of the remote'hutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room.This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.3/4.3.3.6 POST-ACCIDENT INSTRUMENTATION The OPERABILITY of the post-accident instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident.The containment water level CHANNEL CHECK is a visual inspection of parallel channels which should indicate within acceptable instrument drift that the water level is below the range of the containment water level instrumentation.
Acceptable instrument drift for containment water level instrumentation is presently considered 25%of full scale.The containment sump level channels should indicate the same level on both channels within acceptable instrument drift, which is presently considered a 25%of full scale difference between the two parallel channels.If the channels do not indicate the same level, the containment sump pump actuation and shut-off can be used to indicate if either channel is correct.D.C.COOK-UNIT 2 B 3/4 3-2 Amendment No.
3/4.3 INSTRUMENTATION BASES The drift for both instrumentation systems is attributed to air accumulation in the capillaries.
Provided that the drift is less than 25%, it should not prevent the inst ument rom tracking changes in level.Equipment changes may change the acceptable drift.Such a change will not constitute a violation of this T/S, prov'ded appropriate evidence exists to justify the change.3/4.3.3.7 AXIAL POWER DISTRIBUTION MONITORING SYSTEM (APDMS)OPERABILITY of the APDMS ensu es that sufficient capability is available for the measurement of the neutron flux spatial distribution within the reactor core.This capability is required to 1)monitor the core flux patterns that are representative of the peak core power density and 2)limit the core average axial power profile such that~the total power peaking factor F is maintained within acceptable limits.3/4.3.3.8 FIRE DETECTION ZNSTRUMENTATZON 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 f'res will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program.Zn 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 tc OPERABILITY.
Use of containment temperature monitoring is allowed once per hour if containment fire detection is inoperable.
3/4.3.3.9 PADIOACTIVE LZ UZD 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 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.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 D.C.COOK-UNIT 2 B 3/4 3-3 Amendment No.
1 I~~~~3/4.3 INSTRUMENTATION BASES 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'ection 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 rom excessive 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 3-4 Amendment No.
u~~(3~Q orloin yst lllfDIANA&NICH~GAIV ElECTBIC CO FORT WAYNE, INDIANA AMOUNT cHEcK No.08 3~0273 DATE 03 24 8b S l 50.004~ln full seNIement of statement accompenytng this che U S NUCLEAR REGULATORY COMM I S S ION Tothe WASHINGTON DC 20555 Order of 25b43024 DISBURSEMENT ACCOUNT LINCOLN NATIONAL BANK AND TRUST COMPANY , FORT WAY N E, INDIANA II'0 29 L I 3II.0 7 t 900 2 7 5I: I 3S 7 5 20 5ll'EMITTANCE ADVICE-DETACH BEFORE DEPOSITING INDIANA L MICHIGAN ELECTRIC CO.P.O.BOX 60 FORT WAYNE, INDIANA 46801 cHEGK No.06 3 0273 OUR REFERENCE DATE 036 030'I6 02-2b-86 DATE 03 24 8b VENDORS REFERENCE ,AEP NRC or5'I 2 pf AMOUNT GROSS 150 00$150 00 DEDUCT TRY 32 REV.I 86 EXPLANATION CD-CASH DISCOUNT RT-RETAINED PR PRIOR PAYMENT OF DEDUCTIONS TR-TRANSPORTATION TX TAX WITHHELD AJ ADJUSTMENT Attachment 3 to AEP:NRC:0856T Regulatory Guide 1.97, Rev.3, Pertinent Sections
~O'">>U.S.NUCLIR REGULATORY COMMI88lot
!RE LAT RY 0FRCE OF NUCLEAR REGULATORY RESEARCH Resea~aa'D-REGULATORY GUIDE'l.97 INSTRUMENTATION FOR LIGHT-WATEROOLED NUCLEAR POWER PLANTS TO ASSESS PLANT AND ENVIRONS CONDITIONS DURING AND FOLLOWING AN ACCIDENT A.INTROOUCTION B.OISCUSSION Criterion 13,"Instrumentation and Control," of Appen-dix A,"General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50,"Domestic Licensing of Production and Utilization Facilities," includes a requirement that instru-mentation be provided to monitor variables and systems over their anticipated ranges for accident conditions as appropriate to ensure adequate safety.Criterion 19,"Control Roofn," of Appendix A to 10 CFR Part 50 includes a requirement that a control room be pro-, vided from which actions can be taken to maintain the nuclear power unit in a safe condition under accident conditions, including losswfwoolant accidents, and that equipment, including the necessary instrumentation, at appropri'ate locations outside the control room be.provided with a design capability for prompt hot shutdown of the reactor.Criterion 64,"Monitoring Radioactivity Releases," of Appendix A to 10 CFR Part 50 includes a requirement that means be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of lo~fwoolant accident fluid, effluent discharge paths, and the plant environs for radioactivity that may be released from postulated accidents.
Indications of plant variables are required by the control room operating personnel during accident situations to (1)provide information required to permit the operator to take preplanned manual actions to accomplish safe plant shut-down;(2)determine whether the reactor trip, engineered-safety-feature systems, and manually initiated safety systems and other systems important to safety are performing their intended functions (i.ereactivity control, core.cooling, maintaining reactor coolant system integrity, and maintaining containment integrity);
and (3)provide informa-tion to the operators that will enable them to determine the potential for causing a gross breach of the barriers to radioactivity release (i.e., fuel chdding, reactor coohnt pressure boundary, and containment) and to determine if a gross breach of a barrier has occurred.In addition to the above, indications of plant variables that provide informa-tion on operation of plant safety systems and other systems important to safety are required by the control room operating personnel during an accident to (I)furnish data regarding the operation of plant systems in order that the operator can make appropriate decisions as to their use and (2)provide information regarding the release of radioactive materials to allow for early indication of the need to initiate action necessary to protect the public and for an estimate of the magnitude of any impending threat.This guide describes a method acceptable to the NRC staff for complying with the Commission's regulations to provide instrumentation to monitor plant variables and systems during and following an accident in a light-water-cooled nuclear power plant.'Ihe Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position.Any guidance in this document related to information collection activities has been cleared under OMB Clearance No.31504011.At the start of an accident, it may be difficult for the operator to determine immediately what accident has occurred or is occurring and therefore to determine the appropriate response, For this reason, reactor trip and certain other safety actions (e.g., emergency core cooling actuation, containment isolation, or depressurization) have been designed to be performed automatically during the initial stages of an accident.Instrumentation is also provided to indicate information about phnt variables required to enable the operation of manually initiated safety systems and other appropriate operator actions involving systems important to safety.USNRC REGULATORY GUIDES Ragulatary Guides ace Issued to describe and make available to the public cnethods acceptable to the NRc staff of Implementing speclflc parts of the Commlsslon's regulations, to delineate tach-nlouas used by the staff In evaluating speclflc problems or postu-lated accidents, or to provide guidance to applicants.
Regulatory Guides sre noc substitutes for regulations, and compliance with them Is not required.Methods and solutions different from those sat out In the guides will be acceptable If they provide a basis for tha flndlngs raaulslte to tho Issuance or continuance of a permit or license by the Commlsslon.
This guide was Issued after conslderatlon of comments received fcom the public.Cocnments snd suggestions for Improvements In these guides are'ncouraged at all times, and guides will be revised, as appropriate, to accommodate comments and to reflect new Informa-tion or eccpecloncs.
Comments should be sent to the Secretary of ths Commlsslor.
U.S.Nuclear Regulatory Commlsslonh Washington, O.C.20555,, Attantlonc Docketing and Service Branch.The guides sra Issued in the following ten broad dlvlslonsc 1.Power Reactors 6.Products 2, Research and Test Reactors 7.Transportation 3.Fuels snd Materials Facllltles 8.Occupational Health 4.Environmental and Sltlng g.Antitrust and Flnanclal Review 5.Materials and Plant Protection 10.General Copies of Issued guides may bo purchased st the current Government pclntlng office price.A subscctptlon service for futuro guides In spo.elf lc dlvlslons Is avallabls through tho Government prlntlng Office.Information on the subscrlptlan socvlce and current Gpo prices may be obtained by wrltlng tho U.s.Nuclear Regulatory commlsslon, washington, D.c.20555, Attontlonc publlcatlons sales Manager.
I~i n 1I I+I TABLE 1 DESIGN AND QUALIFICATION CRlTERlA FOR INSTRUNIENTATION Category 1 l.Equipment Qualification The instrumentation should be qualiTied in accordance with Regulatory Guide 1.89,"Qualification of Class lE Equipment for Nuclear Power Plants," and the method-ology described in NUREG4588,"Interim Staff Posi-tion on Environmental Qualification of Safety-Related Electrical Equipment." Category 2 1.Equipment Qualification Same as Category 1 Category 3 l.Equipment Qualification No specific provision Instrumentation whose ranges are required to extend beyond those ranges calculated in the most severe design.basis accident event for a given variable should be quali-fied using the guidance provided in paragraph 6.3.6 of ANSI'.5.Same as Cat<<g>>ry 1 No sp<<<<ilic pn)vis')on Qualification applies to the complete instrumentation channel from sensor to display where the display is a direct-indicating meter or recording device.If the instru-mentation channel signal is to be used in a computer-based display, recording, or diagnostic program, qualifi-cation applies from the sensor up to and including the channel isolation device.Same as Category 1 No spccifi<<provision The seismic portion of qualification should be in accor-dance with Regulatory Guide 1.100,"Seismic Qualifica-tion of Electric Equipment for Nuclear Power Plants." Instrumentation should continue to read within the required accuracy following, but not necessarily during, a safe shutdown earthquake.
No specil'i<<provisi>>n No specilic provision 2.Redundancy No single failure within either the accident-monitoring instrumentation, its auxiliary supporting features, or its power sources concurrent with the failures that are 2.Redundancy No specific provision 2.Redundancy No specific provision 4 Coploa aro avallabla from the NRC/GPO Sataa Program, U S.Nuclear Regulatory Commlraion, Waahlnaton, D.C.20555.
2.(Continued)
Category 1 TABLE 1 (Continued)
Category 2 Category 3 a condition or result of a speciTic accident should prevent the operators from being presented the informa-tion necessary for them to determine the safety status of the plant and to bring the plant to and maintain it in a safe condition following that accident.Where failure of one accident-monitoring channel results in informa-tion ambiguity (that is, the redundant displays disagree)that could lead operators to defeat or fail to accomplish a required safety function, additional information should be provided to allow the operators to deduce the actual conditions in the plant.This may bc accomplished by providing additional independent clrannels of information of the same variable (addition of an identical channel)or by providing an independent channel to monitor a different variable that bears a known relationship to the multiple channels (addition of a diverse channel).Redun-dant or diverse channels should be electrically independ-ent and physicaHy separated from each other and from equipment not classified important to safety in accor-dance with Regulatory Guide 1.75,"Ph'ysical Independ-ence of Electric Systems," up to and'including any isola-!tion device.Within each redundant division of a safety systeni, redundant monitoring channels are not needed except for steam generator level instrumentation in two-loop plants.3.Power Source 3.Power Source 3.Power Source The instrumentation should be energized from station standby power sources as provided in Regulatory Guide 1.32,"Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants," and should be backed up by batteries where momentary interruption is not tolerable.
The instrumentation should be energized from a high-reliability power source, not necessarily standby power, and should be backed up by bat teries where momentary interruption is not tolerable.
No specific provision Category 1 TABLE 1 (Continued)
Category 2 Category 3 4.Channel Availability The instrumentation channel should be available prior to!an accident except as provided in paragraph 4.11,"Excep-tion," as defined in IEEE Std 279-1971,"Criteria for Pro-tection Systems for Nuclear Power Generating Stations," or as specified in the technical specifications.
4.Channel Availability The out-of-service interval should be based on normal technical specification requirements on out of service for the system it serves where applicable or where specified by other requirements.
4.Channel Availability No specific provision 5.Quality Assurance Regulatory Guide 1.28"Quality Assurance Prograin Requirements (Design and Construction)" Regulatory Guide 1.30 (Safety Guide 30)"Quality Assurance Require-ments for the installation, Inspection, and Testing of Instrumentation and Electric Equipment" The recommendations of the following regulatory guides pertaining to quality assurance should be foUowed: 5.Quality Assurance Same as Category l as>>>odili<<d by th<<following:
Since some instru>nentatio<<
is less ii>>purta>>t to safety than other instrumentation, it may not be necessary to apply the same quality assurance measures to all instrumentation.
The quality assur-ance requirements that are implemented should provide control over activities affecting quality to an extent consistent with the importance to safety of the instrumentation.
These requirements should be determined and documented by personnel knowl-edgeable in the end use of the instrumentation.
5.Quality Assurance l1>e i>>st ru>>>e>>I ation eliou1 d be of lugh-quality co>>>>>i<<rcial gra>le and should be selected to withsta>>d the st>ecitied service environment.
Regulatory Guide 1.38"Quality Assurance Require-ments for Packaging, Shipping, Receiving, Storage, and Han-dling of Items for Water-Cooled Nuclear Power Plants" Regulatory Guide 1.58"Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel" Regulatory Guide 1.64"Quality Assurance Require-ments for the Design of Nuclear Power Plants" Regulatory Guide 1.74"Quality Assurance Terms and Definitions" 5.(Continued)
Category 1 TABLE 1 (Continued)
Category 2 Category 3 Regulatory Guide 1.88"Collection, Storage, and hfain-tenance of Nuclear Power Plant Quality Assurance Records" Regulatory Guide 1.123"Quality Assurance Require-ments for Control of Procure-ment of Items and Services for Nuclear Power Plants" Regulatory Guide 1.144"Auditing of Quality Assurance Programs for Nuclear Power Plants" I OO Regulatory Guide 1.146"Qualification of Quality Assur-ance Program Audit Personnel for Nuclear Power Plants" Reference to the above regulatory guides (except Regula-tory Guides 1.30 and 1.38)is being made pending issuance of a revision to Regulatory Guide 1.28 that is under devel-opment (Task RS 002-5)and that will endorse ANSI/AShIE NQA-I-1 979,"Quality Assurance Program Requirements for Nudear Power Plants." 6.Display and Recording Continuous real-time display should be provided.The indication may be on a dial, digital display, CRT, or stripchart recorder.Recording of instrumentation readout information should be provided for at least one redundant channel.6.Display and Recording The instrumentation signal may be displayed on an individual instrument or it may be processed for display on demand.Signals from effluent radioactivity monitors and area monitors should be recorded.6.Display and Recording Same as Category 2 Signals from effluent radioactivity monitors, area monitors, and meteorology monitors should be recorded.5 Qoptea may he obtained from the American Society of Mechanical Engineers, 345 East 47th Street, New York, New Yo<<teet7.
TABLE 1 (Continued)
Category 2 Category 3 6.(Continued)
If direct and immediate trend or transient information is essential for operator information or action, the recording should be continuously avaBable on redun-dant dedicated recorders.
Otherwise, it may be con-tinuously updated, stored in computer memory, and displayed on demand.Intermittent displays such as data loggers and scanning recorders may be used if no significant transient response information is likely to be lost by such devices.Same as Category 1 Same as Category 1 7.Range If two or more instruments are needed to cover a particular range, overlapping of instrument span should be provided.If the required range of moni-toring instrumentation results in a loss of instru-mentation sensitivity in the normal operating range, separate instruments should be used.7.Range Salile as Category l 7.ltauge Seine as Category 1 8.Equipment Identification Types A, B, and C instruments designated as Cate-gories 1 and 2 should be specifically identified with a common designation on the control panels so that the operator can easily discern that they are intended for use under accident conditions.
8.Equipment Identification Same as Category 1 8.Equipment Identification No specific provision 9.Interfaces The transmission of signals for other use should be through isolation devices that are designated as part of the monitoring instrumentation and that meet the provisions of this document.10.Servicing, Testing, and Calibration Servicing, testing, and calibration programs should be specified to maintain the capability of the monitoring instrumentation.
If the required interval between 9.Interfaces Same as Category I 10.Servicing, Testing, and Calibration Same as Category 1 9.Interfaces No specific provision 10.Servicing, Testing, and Calibration Same as Category I 10.(Continued)
Category 1 TABLE 1 (Continued)
Category 2 testing is less than the normal time interval between plant shutdowns, a capability for testing during power operation should be provided.Whenever means for removing channels from service are included in the design, the design should facilitate administrative control of the access to such removal means.Same as Category 1 Same as Category 1 The design should facilitate administrative control of the access to all setpoint adjustments, module calibra-tion adjustments, and test points.Same as Category 1 Sallle as Category 1 Periodic checking, testing, calibration, and calibration verification should be in accordance with the applicable portions of Regulatory Guide 1.118,"Periodic Testing of Electric Power and Protection Systems," pertaining to testing of instrument channels.(Note: Response time testing not usually needed.)Same as Category I Same as Category 1 The location of the isolation device should be such that it would be accessible for maintenance during accident conditions.
Same as Category 1 No specific provision 11.Human Factors The instrumentation should be designed to facilitate the recognition, location, replacement, repair, or adjustment of malfunctioning components or modules.11.Human Factors Same as Category 1 11.Human Factors Same as Category 1 The monitoring instrumentation design should minimize the development of conditions that would cause meters, annunciators, recorders, alarms, etc., to give anomalous indications potentially confusing to the operator.Human factors analysis should be used in determining type and location of displays.Same as Category 1 Same as Category 1 11.(Continued)
Category 1 TABLE 1 (Continued)
Category 2 Category 3'I To the extent practicable, the same instruments should be used for accident monitoring as are used for the normal operations of the plant to enable the operators to use, during accident situations, instruments with which they are most familiar.Same as Category I Same as Category 1 12.Direct Measurement To the extent practicable, monitoring instrumentation inputs should be from sensors that directly measure the desired variables.
An indirect measurement should be made only when it can be shown by analysis to provide unambiguous information.
12.Direct Measurement Same as Category l 12.Direct Measurement Same as Caltcgofy l TABLE 3 PWR VARIABLES TYPE A Variables:
those variables to be monitored that provide the primary information required to permit the control room operator to take specific manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis accident events.Primary information is informa-tion that is essential for the direct accomplishment of the specified safety functions; it does not include those variables that are associated with contingency actions that may also be identified in written procedures.
A variable included as Type A does not preclude it from being included as Type B, C, D, or E or vice versa.Variable Plant specific Range Plant specific Category (see Regulatory Position 1.4 and Table 1)Information required for operator action TYPE B Variables:
those variables that provide information to indicate whether plant safety functions are being accomplished.
Plant safety functions are (I)reactivity control, (2)core cooling, (3)maintaining reactor coolant system integrity, and (4)maintaining containment integrity (including radioactive effluent control).Variables are listed with designated ranges and category for design and qualification requirements.
Key variables are indicated by design and qualiTication Category l.Reactivity Control Neutron Flux 10%to 1007o full power Function detection; accomplishment of mitigation Control Rod Position RCS Soluble Boron Concen-tration Full in or not full in 0 to 6000 ppm Verification Verification RCS Cold Leg Water Temper-ature 50 Fto 400 F VeriTication Core Cooling RCS Hot Leg Water Temper-ature 50 F to 700 F Function detection; accomplishment of mitigation; verification; long-term surveillance RCP Cold Leg Water Temper-50 F to 700 F ature r Function detection; accomplishment of mitigation; veriTication; long-term surveillance RCS Pressure 0 to 3000 psig (4000 psig for CE plants)200 F to 2300 F 12 33 Function detection; accomplishment of mitigation; veriTication; long-term surveillance VeriTication I Where a variabio is listed for mora than one purpose, the instrumentation raqulrarnants may bo Integrated and only ono maasuramont provided.2 The maxhnum value may be ravisad upward to satisfy ATWS roquiramants.
3 Instrumentation that Is a part of the llnal ICC datactlon systom should meet the design roqulramonts spadflod ln ItamILF.2 of NUREG<737.(Nthan Typo K tharmocouptas bacomo part of tho systam, thay aro considarad to moot the raqulromants.
Howovar, tho romaindar of tho dataction system that is outsido the reactor vassal should moot tho raqulromants spadflod.)
1.97-22 TABLE 3 (Continued)
Range Category (see Regulatory Position 1.4 and Table 1)TYPE C (Continued)
Fuel Cladding (Continued)
Radioactivity Concentration or Radiation Level in Circulating Primary Coolant Analysis of Primary Coohnt (Gamma Spectrum)1/2 Tech Spec limit io 100 times Tech Spec limit 10 pCi/ml to 10 Ci/ml or TID-14844 source term in coolant volume 36 Detection of breach Detail analysis;accomplishment of mitigation; veriTication; long-term surveillance Reactor Coolant Pressure Boundary RCS Pressure I Containment Pressurel 0 to 3000 psig (4000 psig for CE plants)-5 psig to design pressure 4 (-10 psig for subatmospheric containments) 12 Detection of potential for or actual breach;accomplishment of mitiga-tion;long-term surveillance Detection of breach;accomplishment of mitigation; verification; long<erm surveillance Containment Sump Water Leveli Narrow range top to bottom (sump), wide range (plant specific)Detection of breach;accomplishment of mitigation; verification; long-term surveillance Effluent Radioactivity
-Noble Gas Effluent from Condenser Air Removal System Exhaust 10 pCi%c to 10 pCi%c Containment Area Radiation 1 R/hr to 10 R/hr 37,8 39 Detection of breach;verification Detection of breach;verification Containment RCS Pressure I 0 to 3000 psig (4000 psig for CE plants)Detection of potential for breach;accomplishment of mitigation SampUng or monitoring of radioactive Uquids and gases should be performed in a manner that ensures ptocutement of tepesentative samoles.For eases the criteria of ANSI N13.1.1969,"G>>ude to Sampling Aitbotno Radioactive Materials in Nuclear FacUiiies>'hould be appQed.For llquih, provisions should be made for sampling from well.mixed turbulent zones, and sampling lines should be de>>dgned to mlnl-nJo plateout or deposition.
For safe and convenient sampgng, the provisions should indude: a.Shielding to maintain radiation doses ALARA, b.Sample containers with contalnetmmpUng port conneciot compatibility, c.Capabigiy of sampling under primary system pressure and negative pressures, d.HandUng and transport capability, and e.Prearrangement for analysis and mietpteiailon.
7 Mlnhnum of iwo monitors at widely separated locations.
8 Detectors should respond io gamma radiation phoions within any energy range ftom 60 koV to 3 MoV with a dose tato response accuracy within a factor of 2 over the eniito range.9 Monitors should be capable of detecdng and measuring gaseous effluent radioactivity with compositions tanyng from fresh equlUbtlutn noble gas ladon Product mixtures io iordayrold mixtures, with overall system accuracies,within a factor of 2.Effluent radioactivity may bo<<p<<ssod ln terms of concentrations of Xe-133 equivalents, in terms of concentrations of any noble gas nucUdes, or in terms of integrated na MeV pet unIt thno.It is not expected iha!a single monitoring device will have suNcient tango to encompass the oaflte tango provided this regulatory guide and that mulilolo components or systems will be needed.Existing equlpmoat may be used to monitor any portion of ih stated tango witidn the equipment design rating.r 1.97-24 Attachment 4 to AEP:NRC:0856T Reasons and 10 CFR 50.92 Analysis for Change to the Donald C..Cook Nuclear Plant Unit Nos.1 and 2 Technical Specifications Ill r~~\Attachment 4 to A:NRC:0856T Page 1 Containment Water Level Monitor II F 1.5 The guidance given in Generic Letter No.83-37 states that:~"A continuous indication of containment water level should be provided in the control room of each reactor during Power Operation, Startup and Hot Standby modes of operation.
At least one channel for narrow range and two channels for wide range instruments should be operable at all times when the reactor is operating in any of the above modes.Narrow range instruments should cover the range from bottom to the top of the containment sump.Wide range instruments should cover the range from the bottom of the containment to the elevation equivalent to a 600,000 gallon (or less if justified) capacity."Technical Specifications for containment water level monitors should be included with other accident monitoring instrumentation in the present Technical Specifications.
LCOs (including the required Actions)for wide range monitors should include the requirement that the inoperable channel will be restored to operable status within 30 days or the plant will be brought to Hot Shutdown condition as required for other accident monitoring instrumentation.
Typical acceptable LCO and surveillance requirements for accident monitoring instrumentation are included in Enclosure 3." k We are proposing that T/S Tables 3.3-11 and 3.3-10 for Units 1 and 2, respectively, be revised to include the requirement that at least two containment water level channels and one containment sump level channel be operable during Modes 1, 2, and 3.In addition, we are proposing that T/S Tables 4.3-7 and 4.3-10 for Units 1 and 2, respectively, be revised to include the surveillance requirements for these channels.In order to follow the above guidance, and maintain internal consistency with our current Technical Specifications, the 30-day action statement in T/S 3.3'.8 for Unit 1 and 3.3'.6 for Unit 2 is proposed for the containment water level and containment sump level instrumentation.
The format of T/S Tables 3.3-11 and 3.3-10 for Units 1 and 2 varies from the Generic Letter example because our present T/Ss include only one column listing"Minimum Channels Operable." In order to keep the format similar to other accident monitoring instrumentation included in the present T/Ss, the column listing the"Required No.of Channels" is not included.Per 10 CFR 50.92, a proposed amendment will not involve a significant hazArds consideration if the proposed amendment does not: (1)involve'significant increase in the probability or consequences of an accident previously evaluated, (2)create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3)involve a significant reduction in a margin of safety.
lN~P 7 Attachment 4 to A..NRC:0856T Criterion These changes will expand the license requirements for post-accident monitoring instrumentation and assist the operator in recovering from an accident.The changes will not involve a significant increase in the probability or consequences of any previously evaluated accident.Criterion 2 The changes do not affect normal or accident plant operation.
In an accident they will serve to provide data to the operator;therefore, the changes will not create the possibility of a new or different kind of accident from any previously analyzed or evaluated.
Criterion 3 The changes do not involve a significant reduction in the margin of safety, since they will only require that additional data be available to the operator.The Commission has provided guidance concerning the determination of si.gnificant hazards by providing certain examples (48 FR 14870)of amendments considered not likely to involve significant hazards consideration.
The second of these examples refers to changes that impose additional limitations, restrictions, or controls not presently included in the T/Ss.Since the requirement for sump and containment water level monitors constitute a restriction which the current T/Ss do not have, we believe this example is applicable and that the changes involve no significant hazards consideration.
The above T/S changes constitute additional restrictions to the present T/Ss.Therefore, we believe that these changes do not involve a significant hazards consideration as defined in 10 CFR 50.92.It is noted that AEP:NRC:0856I also proposed changes for the axial power distribution monitoring system and several administrative changes.We are limiting this submittal to changes for the containment water level and containment sump level instrumentation.
We request that your staff continue to review the changes regarding the axial power distribution system and the administrative changes as submitted in AEP:NRC:0856I.
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