Regulatory Guide 1.97, "Instrumentation for Light-Water - NRC

Revision 2'December 1980

CU.CCS U.S. NUCLEAR REGULATORY COMMISSION0

pREEGULATORY GUIDEOF OFICE OF STANDARDS DEVELOPMENT

REGULATORY GUIDE 1.97{Task RS 917-4)

INSTRUMENTATION FOR LIGHT-VVATER-COOLED NUCLEAR POWER PLANTSTO ASSESS PLANT AND ENVIRONS CONDITIONS DURING AND FOLLOWING

AN ACCIDENT

A. INTRODUCTION

Criterion 13, "Instrumentatwon and Control," of Appen-dix A, "General Design Criteria for Nuclear Power Plants,"to 10 CFR Part 50, "Domestic Licensing of Production andUtilization Facilities," includes a requirement that instru-mentation be provided to monitor variables and systemsover their anticipated ranges for accident conditions asappropriate to ensure adequale safety.

Criterion 19, "Control Room,", of Appendix A to10 CFR Part 50 includes a requirement that a control roombe provided from which actions can be taken to maintainthe nuclear power unit in a safe condition under accidentconditions, including loss-of-coolant accidents, and thatequipment, including the necessary instrumentation, atappropriate locations outside the control room be providedwith a design capability for prompt hot shutdown of thereactor.

Criterion 64, "Monitoring. Radioactivity Releases," ofAppendix A to 10 CFR Part 50 includes a requirement thatmeans be provided for monitoring the reactor containmentatmosphere, spaces containing components for recirculationof loss-of-coolant accident fluid, effluent discharge paths,and the plant environs for radioactivity that may be releasedfrom postulated accidents.

This guide describes a method acceptable to the NRCstaff for complying with thelCommission's regulations toprovide instrumentation to jnonitor plant variables andsystems during and following, an accident in a light-water-cooled nuclear power plant. The Advisory. Committee onReactor Safeguards has been consulted concerning thisguide and has concurred in the regulatory position.

*.

The substantial number of changes In this revision has made itimpractical to Indicate the changes with lines in the margin.

B. DISCUSSION

Indications of plant variables are required by the controlroom operating personnel during accident situations to (I)provide information required to permit the operator to takepreplanned manual actions to accomplish safe plant shut-down; (2) determine whether the reactor trip, engineered-safety-feature systems, and manually initiated safetysystems and other systems important to safety are performingtheir intended functions (i.e., reactivity control, corecooling, maintaining reactor coolant system integrity, andmaintaining containment integrity); and (3) provide informa-tion to the operators that will enable them to determine thepotential for causing a gross breach of the" barriers toradioactivity release (i.e., fuel cladding, reactor coolantpressure boundary, and containment) and to determine if agross breach of a barrier has occurred. In addition to theabove, indications of plant variables that provide informa-tion on operation of plant safety systems and other systemsimportant to safety are required by the control roomoperating personnel during an accident to (I) furnish dataregarding the operation of plant'systems in order that theoperator can make appropriate decisions as to their use and(2) provide information regarding the release of radioactivematerials to allow for early indication of the need toinitiate action necessary to protect the public and for anestimate of the magnitude of any impending threat.

At the start of an accident, it may be difficult for theoperator to determine immediately what accident hasoccurred or is occurring and therefore to determine theappropriate response. For this reason, reactor trip andcertain other safety actions (e.g., emergency core coolingactuation, containment isolation, or depressurization) havebeen designed to be performed automatically during the;nitial stages of an accident. Instrumentation is also providedto indicate information about plant variables required toenable the operation of manually initiated safety systems

USNRC REGULATORY GUIDES

Regulatory Guides are Issued to describe and make available to thepublic methods acceptable to the NRC staff of Implementingspecific parts of the Commissions regulations, to delineate tech-niques used by the staff In evaluating specific problems or postu-lated accidents or to provide giuidance to applicants. RegulatoryGuides are nol substitutes for regulations, and compliance withthem Is not required. Methods and solutions different from those setout In the guides will be acceptable if they provide a basis for thefindings requisite to the Issuance or continuance of a permit orlicense by the Commission.

Comments and suggestions for Improvements In these guides areencouraged at all times, and guides will be revised, as appropriate,to accommodate comments and to reflect new Information orexperience. This guide was revised as a result of substantive com-ments received from the public and additional staff review.

Comments should be sent to the Secretary of the Commission,U.S. Nuclear Regulatory Commission, Washington, D.C. 20555,Attention: Docketing and Service Branch.

The guides are Issued In the following ten broad divisions:

1. Power Reactors 6. Products2. Research and Test Reactors 7. Transportation3. Fuels and Materials Facilities 8. Occupational Health4. Environmental and Siting 9. Antitrust and Financial Review5. Materials and Plant Protection 10. General

Copies' of Issued guides may be purchased at the current GovernmentPrinting Office price. A subscription service for future guides In spe-cific divisions Is available through the Government Printing Office.Information on the subscription service and current GPO prices maybe obtained by writing the U.S. Nuclear Regulatory Commission,Washington, D.C. 20555, Attention: Publications Sales Manager.

and other appropriate operator actions involving systemsimportant to safety.

Independent of the above tasks, it is important thatoperators be informed if the barriers to the release ofradioactive materials are being challenged. Therefore, it'isessential that instrument ranges be selected so that theinstrument will always be on scale. Narrow-range instrumentsmay not have the necessary range to track the course of theaccident; consequently, multiple instruments with over-lapping ranges may be necessary. (In the past, some instru-ment ranges have been selected based on the setpoint valuefor automatic protection or alarms.) It is essential thatdegraded conditions and their magnitude be identified sothe operators can take actions that are available to mitigatethe consequences. It is not intended that operators beencouraged to prematurely circumvent systems importantto safety but that they be adequately informed in orderthat unplanned actions can be taken when necessary.

Examples of serious events that could threaten safety ifconditions degrade are loss-of-coolant accidents (LOCAs),overpressure transients, anticipated operational occurrer.cesthat become accidents such as anticipated transients withoutscram (ATWS), and reactivity excursions that result inreleases of radioactive materials. Such events require thatthe operators understand, within a short time period, theability of the barriers to limit radioactivity release, i.e., thatthey understand the potential for breach of a barrier orwhether an actual breach of a barrier has occurred becauseof an accident in progress.

It is essential that the required instrumentation becapable of surviving the accident environment in which it islocated for the length of time its function is required. Itcould therefore either be designed to withstand the accidentenvironment or be protected by a local protected environ-ment.

It is desirable that accident-monitoring instrumentationcomponents and their mounts that cannot be located inseismically qualified buildings be designed to continue tofunction, to the extent feasible, following seismic events.An acceptable method for enhancing the seismic resistanceof this instrumentation would be to design it to meet theseismic criteria applicable to like instrumentation installedin seismically qualified locations although a lesser over-all qualification results.

Variables for accident monitoring can be selected toprovide the essential information needed by the operator todetermine if the plant safety functions are being performed.It is essential that the range selections be sufficientlygreat to keep instruments on scale at all times. Further, it isprudent that a limited number of those variables that arefunctionally significant (e.g., containment pressure, primarysystem pressure) be monitored by instruments qualified tomore stringent environmental requirements and with rangesthat extend well beyond that which the selected variablescan attain under limiting conditions; for example, a rangefor the containment pressure monitor extending to the

burst pressure of the containment in order that the operatorswill not be uninformed as to the pressure inside the contain-ment. The availability of such instruments is important sothat responses to corrective actions can be observed and theneed for, and magnitude of, further actions can be deter-mined. It is also necessary to be sure that when a range isextended, the sensitivity and accuracy of the instrument arewithin acceptable limits for monitoring the extended range.

Normal power plant instrumentation remaining functionalfor all accident conditions can provide indication, records,and (with certain types of instruments) time-history responsesfor many variables important to following the course of theaccident. Therefore, it is prudent to select the requiredaccident-monitoring instrumentation from the normalpower plant instrumentation to enable operators to use,during accident situations, instruments with which they aremost familiar. Since some accidents could impose severeoperating requirements on instrumentation components, itmay be necessary to upgrade those normal power plantinstrumentation components to withstand the more severeoperating conditions and to measure greater variations ofmonitored variables that may be associated with an accident.It is essential that instrumentation so upgraded does notdegrade the accuracy and sensitivity required for normaloperation. In some cases, this will necessitate use of over-lapping ranges of instruments to monitor the required rangeof the variable to be monitored, possibly with differentperformance requirements in each range.

ANSI/ANS-4.5-1980,1 "Criteria for Accident MonitoringFunctions in Light-Water-Cooled Reactors," delineates Wcriteria for determining the variables to be monitored bythe control room operator, as required for safety, duringthe course of an accident and during the long-term stableshutdown phase following an accident. ANS-4.5 wasprepared by Working Group 4.5 of Subcommittee ANS-4with two primary objectives: (I ) to address that instrumenta-tion that permits the operators to monitor expected param-eter changes in an accident period and (2) to addressextended-range instrumentation deemed appropriate for thepossibility of encountering previously unforeseen events.ANS-4.5 references a revision to IEEE Standard 497 as thesource for specific instrumentation design criteria. Since therevision to IEEE Standard 497 has not been completed, itsapplicability cannot yet be determined. Hence, specificinstrumentation design criteria have been included in thisregulatory guide.

ANS-4.5 defines three types of variables (definitionsmodified herein) for the purpose of aiding the designer inselecting accident-monitoring instrumentation and applicablecriteria. The types are: Type A, those variables that provideprimary informations needed to permit the control room

Copies may be obtained from the American Nuclear Society,555 North Kensington Avenue, La Grange Park, Illinois 60525.

2Primary information is information that is essential for the Bdirect accomplishment of the specified safety functions; it does notinclude those variables that are associated with contingency actionsthat may also be Identified in written procedures.

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operating personnel to take the specified manually controlledactions for which no automatic control is provided and that

*o are required for safety systems to accomplish their safetyfunctions for design basis accident events; Type B, those

a P variables that provide information to indicate whether plantsafety functions are being accomplished; and Type C, thosevariables that provide information to indicate the potentialfor being breached or the actual breach of the barriers tofission product release, i.e., fuel cladding, primary coolantpressure boundary, and containment (modified to reflectNRC staff position; see regulatory position 1.2). Thesources of potential breach are limited to the energysources within the barrier itself. In addition to the accident-monitoring variables provided in ANS-4.5, variables formonitoring the operation of systems important to safetyand radioactive effluent releases are provided by thisregulatory guide. Two additional variable types are defined:Type D, those variables that proiide information to indicatethe operation of individual safety systems and other systemsimportant to safety, and Type E, those variables to bemonitored as required for use in determining the magnitudeof the release of radioactive materials and for continuouslyassessing such releases.

A minimum set of Type B, C, D, and E variables to bemeasured is listed in this regulatory guide. Type A variableshave not been listed because they are plant specific and willdepend on the operations that the designer chooses forplanned manual action. Types E:, C, D, and E are variablesfor following the course of an accident and are to be used(I) to determine if the plant is responding to the safetymeasures in operation and (2) to inform the operator ofthe necessity for unplanned actions to mitigate the con-sequences of an accident. The five classifications are notmutually exclusive in that a given variable (or instrument)may be applicable to one or more types, as well as fornormal power plant operation or for automatically initiatedsafety actions. A variable included as Type B, C, D, or Edoes not preclude that variable from also being includedas Type A. Where such multiple use occurs, it is essentialthat instrumentation be capable of meeting the morestringent requirements.

The time phases (Phases I and 11) delineated in ANS-4.5are not used in this regulatory guide. These considerationsare plant specific. It is important that the required instru-mentation survive the accident environment and functionas long as the information it provides is needed by thecontrol room operating personnel.

The NRC staff is willing to work with the ANS workinggroup to attempt to resolve the above differences.

Regulatory positions 1.3 and 1.4 of this guide providedesign and qualification criteria for the instrumentationused to measure the various variables listed in Table I (forBWRs) and Table 2 (for PWRs). The criteria are separatedinto three separate groups or categories that provide agraded approach to requirements depending on the impor-S^ tance to safety of the measurement of a specific variable.Category I provides the most stringent requirements and isintended for key variables. Category 2 provides less stringent

requirements and generally applies to instrumentationdesignated for indicating system operating status. Category 3is intended to provide requirements that will ensure thathigh-quality off-the-shelf instrumentation is obtained andapplies to backup and diagnostic instrumentation. It is alsoused where the state of the art will not support requirementsfor higher qualified instrumentation.

In general, the measurement of a single key variable maynot be sufficient to indicate the accomplishment of a givensafety function. Where multiple variables are needed toindicate the accomplishment of a given safety function, it isessential that they each be considered key variables and bemeasured with high-quality instrumentation. Additionally,it is prudent, in some instances, to include the measurementof additional variables for backup information and fordiagnosis. Where these additional measurements are included,the measures applied for design, qualification, and qualityassurance of the instrumentation need not be the same asthat applied for the instrumentation for key variables. Akey variable is that single variable (or minimum number ofvariables) that most directly indicates the accomplishmentof a safety function (in the case of Types B and C) or theoperation of a safety system (in the case of Type D) orradioactive material release (in the case of Type E). It isessential that key variables be qualified to the more stringentdesign and qualification criteria. The design and qualificationcriteria category assigned to each variable indicates whetherthe variable is considered to be a key variable or for systemstatus indication or for backup or diagnosis, i.e., for Types Band C, the key variables are Category I; backup variablesare generally Category 3. For Types D and E, the keyvariables are generally Category 2; backup variables areCategory 3.

The variables are listed, but no mention (beyond redun-dancy requirements) is made of the number of points ofmeasurement of each variable. It is important that thenumber of points of measurement be sufficient to adequatelyindicate the variable value, e.g., containment temperaturemay require spatial location of several points of measure-ment.

This guide provides the minimum number of variables tobe monitored by the control room operating personnelduring and following an accident. These variables are usedby the control room operating personnel to perform theirrole in the emergency plan in,the evaluation, assessment,monitoring, and execution of 'ontrol room functions whenthe other emergency response facilities are not effectivelymanned. Variables are also defined to permit operators toperform their long-term monitoring and execution respon-sibilities after the emergency response facilities are manned.The application of the criteria for the instrumentation islimited to that part of the instrumentation system andits vital supporting features or power sources that providethe direct display of the variables. These provisions are notnecessarily applicable to that part of the instrumentationsystems provided as operator aids for the purpose ofenhancing information presentations for the identificationor diagnosis of disturbances.

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C. REGULATORY POSITION

1. Accident-Monitoring Instrumentation

The criteria and requirements contained in ANSI/ANS-4. 5-1980, "Criteria for Accident Monitoring Functions in Light-Water-Cooled Reactors," are considered by the NRC staff tobe generally acceptable for providing instrumentation tomonitor variables for accident conditions subject to thefollowing:

1.1 Instead of the definition given in Section 3.2.1 ofANS-4.5, the definition of Type A variables should be:Type A, those variables to be monitored that provide theprimary information2 required to permit the control roomoperators to take the specified manually controlled actionsfor which no automatic control is provided and that arerequired for safety systems to accomplish their safetyfunction for design basis accident events.

1.2 In Section 3.2.3 of ANS-4.5, the definition ofType C includes two items, (I) and (2). Item (I) includesthose instruments that indicate the extent to which variablesthat have the potential for causing a breach in the primaryreactor containment have exceeded the design basis values.In conjunction with the variables that indicate the potentialfor causing a breach in the primary reactor containment,the variables that indicate the potential for causing a breachin the fuel cladding (e.g., core exit temperature) and thereactor coolant pressure boundary (e.g., reactor coolantpressure) should also be included. The sources of potentialbreach are limited to the energy sources within the cladding,coolant boundary, or containment. References to Type Cinstruments, and associated parameters to be measured, inANS-4.5 (e.g., Sections 4.2, 5.0, 5.1.3, 5.2, 6.0, 6.3) shouldinclude this expanded definition.

1.3 Section 6.1 of ANS-4.5 pertains to General Criteriafor Types A, B, and.C accident-monitoring variables. In lieuof Section 6.1, the following design and qualificationcriteria categories should be used:

1.3.1 Design and Qualificction Criteria - Category 1

a. The instrumentation should be qualified in accordancewith Regulatory Guide 1.89, "Qualification of Class IEEquipment for Nuclear Power Plants," and the methodologydescribed in NUREG-0588, "Interim Staff Position onEnvironmental Qualification of Safety-Related ElectricalEquipment." Qualification applies to the complete instru-mentation channel from sensor to display where the displayis a direct-indicating meter or recording device. Where theinstrumentation channel signal is to be used in a computer-based display, recording, and/or diagnostic program,qualification applies from the sensor to and includes thechannel isolation device. The location of the isolationdevice should be such that it would be accessible formaintenance during accident conditions. The seismicportion of qualification should be in accordance withRegulatory Guide 1.100, "Seismic Qualification of ElectricEquipment for Nuclear Power Plants." Instrumentationshould continue to read within the required accuracy

following, but not necessarily during, a safe shutdownearthquake. Instrumentation whose ranges are required toextend beyond those ranges calculated in the most severedesign basis accident event for a given variable should bequalified using the guidance provided in paragraph 6.3.6 of _ANS-4.5.

b. No single failure within either the accident-monitoringinstrumentation, its auxiliary supporting features, or itspower sources concurrent with the failures that are acondition or result of a specific accident should preventthe operators from being presented the information neces-sary for them to determine the safety status of the plantand to bring the plant to and maintain it in a safe conditionfollowing that accident. Where failure of one accident-monitoring channel results in information ambiguity (thatis, the redundant displays disagree) that could lead operatorsto defeat or fail to accomplish a required safety function,additional information should be provided to allow theoperators to deduce the actual conditions in the plant. Thismay be accomplished by providing additional independentchannels of information of the same variable (addition ofan identical channel) or by providing an independentchannel to monitor a different variable that bears a knownrelationship to the multiple channels (addition of a diversechannel). Redundant or diverse channels should be electricallyindependent and physically separated from each other andfrom equipment not classified important to safety inaccordance with Regulatory Guide 1.75, "Physical Inde-pendence of Electric Systems," up to and including anyisolation device. At least one channel should be displayedon a direct-indicating or recording device. (Note: Withineach redundant division of a safety system, redundantmonitoring channels are not needed except for steamgenerator level instrumentation in two-loop plants.)

c. The instrumentation should be energized from stationStandby Power sources as provided in Regulatory Guide 1.32,"Criteria for Safety-Related Electric Power Systems forNuclear Power Plants," and should be backed up by batterieswhere momentary interruption is not tolerable.

d. The instrumentation channel should be availableprior to an accident except as provided in paragraph 4.11,"Exemption," as defined in IEEE Standard 279 or asspecified in Technical Specifications.

e. The recommendations of the following regulatoryguides pertaining to quality assurance should be followed:

Regulatory Guide 1.28



"Quality Assurance ProgramRequirements (Design andConstruction)"

"Quality Assurance Require-ments for the Installation,Inspection, and Testing ofInstrumentation and ElectricEquipment"

"Quality Assurance Require- _ments for Packaging, Shipping, W

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Receiving, Storage, and Han-dling of Items for Water-CooledNuclear Power Plants"








"Qualification of Nuclear PcverPlant Inspection, Examination,and Testing Personnel"

"Quality Assurance Require-ments for the Design of Nu-clear Power Plants"

"Quality Assurance Terms andDefinitions"

"Collection, Storage, and Main-tenance of Nuclear Power FlintQuality Assurance Records"

"Quality Assurance Require-ments for Control of Procure-ment of Items and Servicesfor Nuclear Power PHants"

"Auditing of Quality AssurancePrograms for Nuclear PowerPlants"

"Qualification of QualityAssurance Program AuditPersonnel for Nuclear PowerPlants"

the channel signal is to be processed or displayed on demand,qualification applies from the sensor through the isolator/input buffer. The location of the isolation device should besuch that it would be accessible for maintenance duringaccident conditions.

b. The instrumentation should be energized from ahigh-reliability power source, not necessarily StandbyPower, and should be backed up by batteries where momen-tary interruption is not tolerable.

c. The out-of-service interval should be based on normalTechnical Specification requirements on out of service forthe system it serves where applicable or where specified byother requirements.

d. The recommendations of the regulatory guidespertaining to quality assurance listed under paragraph 1.3. Ieof this guide should be followed. Reference to the aboveregulatory guides (except Regulatory Guides 1.30 and 1.38)is being made pending issuance of a regulatory guide(Task RS 002-5) that is under development and will endorseANSI/AShvE NQA-1-1979. Since some instrumentation isless important to safety than other instrumentation, it maynot be necessary to apply the same quality assurance measuresto all instrumentation. The quality assurance requirementsthat are implemented should provide control over activitiesaffecting quality to an extent consistent with the importanceto safety of the instrumentation. These requirements shouldbe determined and documented by personnel knowledgeablein the end use of the instrumentation.

e. The instrumentation signal may be displayed on anindividual instntment or it may be processed for display ondemand by a CRT or by other appropriate means.

f. The method of display may be by dial, digital, CRT,or stripchart recorder indication. Effluent radioactivitymonitors, area radiation monitors, and meteorology monitorsshould be recorded. Where direct and immediate trend ortransient information is essential for operator information oraction, the recording should be continuously available on ded-icated recorders. Otherwise, it may be continuously updated,stored in computer memory, and displayed on demand.

1.3.3 Design and Qualification Criteria - Category 3

a. The instrumentation should be of high-qualitycommercial grade and should be selected to withstand thespecified service environment.

b. The method of display may be by dial, digital, CRT, orstripchart recorder indication. Effluent radioactivity monitors,area radiation monitors, and meteorology monitors should berecorded. Where direct and immediate trend or transientinformation is essential for operator information or action,the recording should be continuously available on dedicatedrecorders. Otherwise, it may be. continuously updated,stored in computer memory, and displayed on demand.

1.4 In addition to the criteria of regulatory position 1.3,the following criteria should apply to Categories I and 2:

Reference to the above regulatory guides (except Regula-tory Guides 1.30 and 1.38) is being made pending issuanceof a regulatory guide (Task RS 002-5) that is under develop-ment and will endorse ANSI/ASME NQA-1-1979, "QualityAssurance Program Requirements for Nuclear PowerPlants."

f. Continuous indication (it may be by recording) displayshould be provided. Where two or more instruments areneeded to cover a particular range, overlapping of instru-ment span should be provided.

g. Recording of instrumentation readout informationshould be provided. Where direct and immediate trend ortransient information is essential for operator informationor action, the recording should be continuously availableon dedicated recorders. Otherwise, it may be continuouslyupdated, stored in computer memory, and displayed ondemand. Intermittent displays such as data loggers andscanning recorders may be used if no significant transientresponse information is likely to be lost by such devices.

1.3.2 Design and Qualification Criteria - Category 2

a. The instrumentation should be qualified in accordancewith Regulatory Guide 1.89 and the methodology describedin NUREG-0588. Seismic qualification according to the pro-visions of Regulatory Guide 1.100 may be needed providedthe instrumentation is part of a safety-related system. Where

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a. Any equipment that is used for either Category I or Cat-egory 2 should be designated as part of accident-monitoringinstrumentation or systems operation and effluent-monitoringinstrumentation. The transmission of signals from suchequipment for other use should be through isolation devicesthat are designated as part of the monitoring instrumentationand that meet the provisions of this document.

b. The instruments designated as Types A, B, and C andCategories I and 2 should be specifically identified on thecontrol panels so that the operator can easily discern thatthey are intended for use under accident conditions.

1.5 In addition to the above criteria, the followingcriteria should apply to Categories L 2, and 3:

a. Servicing, testing, and calibration programs should bespecified to maintain the capability of the monitoringinstrumentation. For those instruments where the requiredinterval between testing will be less than the normal timeinterval between generating station shutdowns, a capabilityfor testing during power operation should be provided.

b. Whenever means for removing channels from serviceare included in the design, the design should facilitateadministrative control of the access to such removal means.

c. The design should facilitate administrative control ofthe access to all setpoint adjustments, module calibrationadjustments, and test points.

d. The monitoring instrumentation design should minimizethe development of conditions that would cause meters, an-nunciators, recorders, alarms, etc., t6 give anomalous indica-tions potentially confusing to the operator. Human factorsanalysis should be used in determining type and location ofdisplays.

e. The instrumentation should be designed to facilitatethe recognition, location, replacement, repair, or adjustmentof malfunctioning components or modules.

f. To the extent practicable, monitoring instrumentationinputs should be from sensors that directly measure thedesired variables. An indirect measurement should be madeonly when it can be shown by analysis to provide unambigu-ous information.

g. To the extent practicable, the same instrumentsshould be used for accident monitoring as are used for thenormal operations of the plant to enable the operators. touse, during accident situations, instruments with whichthey are most familiar. However, where the required rangeof monitoring instrumentation results in a loss of instrumen-tation sensitivity in the normal operating range, separateinstruments should be used.

h. Periodic checking, testing, calibration, and calibrationverification should be in accordance with the applicableportions of Regulatory Guide 1.1 IS, "Periodic Testing ofElectric Power and Protection Systemjs," pertaining to testing

of instrument channels. (Note: Response time testing notusually needed.)

1.6 Sections 6.2.2, 6.2.3, 6.2.4, 6.2.5, 6.2.6, 6.3.2,@26.3.3, 6.3.4, and 6.3.5 of ANS-4.5 pertain to variables andvariable ranges for monitoring Types B and C variables. Inconjunction with the above-listed sections of ANS-4.5,Tables I and 2 of this regulatory guide (which include thosevariables mentioned in these sections) should be consideredas the minimum number of instruments and their respectiveranges for accident-monitoring instrumentation for eachnuclear power plant.

2. Systems Operation Monitoring and Effluent ReleaseMonitoring Instrumentation

2.1 Definitions

a. Type D, those variables that provide information toindicate the operation of individual safety systems andother systems important to safety.

b. Type E, those variables to be monitored as requiredfor use in determining the magnitude of the release ofradioactive materials and in continually assessing suchreleases.

2.2 The plant designer should select variables andinformation display channels required by his design toenable the control room operating personnel to:

a. Ascertain the operating status of each individualsafety system and other systems important to safety to thatextent necessary to determine if each system is operating orcan be placed in operation to help mitigate the consequencesof an accident.

b. Monitor the effluent discharge paths and environswithin the site boundary to ascertain if there have beensignificant releases (planned or unplanned) of radioactivematerials and to continually assess such releases.

c. Obtain required information through a backup ordiagnosis channel where a single channel may be likely togive ambiguous indication.

- 2.3 The process for selecting system operation andeffluent release variables should include the identificationof:

a. For Type D

(1) The plant safety systems and other systemsimportant to safety that should be operating or that couldbe placed in operation to help mitigate the consequences ofan accident; and

(2) The variable or minimum number of variables_that indicate the operating status of each system identifiedin (I) above.

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b. For type E

(1) The planned paths for effluent release;

(2) Plant areas and inside buildings where access isrequired to service equipment necessary to mitigate theconsequences of an accident;

(3) Onsite locations where unplanned releases ofradioactive materials should be detected; and

(4) The variables that should be monitored in eachlocation identified in (1), (2), and (3) above.

2.4 The determination of performance requirements forsystem operation monitoring and effluent release monitoringinformation display channels should include, as a minimum,identification of:

a. The range of the process variable.b. The required accuracy of measurement.c. The required response characteristics.d. The time interval during which the measurement is

needed.e. The local environment(s) in which the information

display channel components must operate.f. Any requirement for rate or trend information.g. Any requirements to group displays of related infor-

mation.h. Any required spatial distribution of sensors.

2.5 The design and qualification criteria for systemoperation monitoring and effluent release monitoring

instrumentation should be taken from the criteria providedin regulatory positions 1.3 and 1.4 of this guide. Tables Iand 2 of this regulatory guide should be considered as theminimum number of instruments and their respectiveranges for systems operation monitoring (Type D) andeffluent release monitoring (Type E) instrumentation foreach nuclear power plant.

D. IMPLEMENTATION

All plants going into operation after June 1983 shouldmeet the provisions of this guide.

Plants currently operating should meet the provisions ofthis guide, except as modified by NUREG-0737 and theCommission Memorandum and Order (CLI-80-21), by June1983.

Plants scheduled to be licensed to operate before June 1,1983, should meet the requirements of NUREG-0737 andthe Commission Memorandum and Order (CLI-80-21) andthe schedules of these documents or pridr to the issuance ofa license to operate, whichever date is later. The balance ofthe provisions of this guide should be completed by June1983.

The difficulties of procuring and installing additions ormodifications to in-place instrumentation have been con-sidered in establishing these schedules.

Exceptions to provisions and schedules will be consideredfor extraordinary circumstances.

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TABLE 1

BWR VARIABLES

TYPE A Variables: those variables to be monitored that provide the primary information required to permit the controlroom operator to take specific manually controlled actions for which no automatic control is provided and that are requiredfor 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 variablesthat 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.

Category (seeRegulatoryPosition 1.3)Variable Range Purpose

Plant specific Plant specific I Information required for operatoraction

TYPE B Variables: those variables that provide information to indicate whether plant safety functions are being accomplished.Plant safety functions are (l) 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 andcategory for design and qualification requirements. Key variables are indicated by design and qualification Category 1.

Reactivity Control

Neutron Flux 10-6 % to 100% full power(SRM, APRM)

3

3

Function detection; accomplishmentof mitigation

VerificationControl Rod Position Full in or not full in

RCS Soluble Boron Concen-tration (Sample)

0 to 1000 ppm 3 Verification

Core Cooling

Coolant Level in Reactor Bottom of core support plate tolesser of top of vessel or center-line of main steam line.

I Function detection; accomplishmentof mitigation; long-term surveillance

BWR Core Thermocouples2 200'F to 2300'F 1 1 To provide diverse indication ofwater level

Maintaining Reactor CoolantSystem Integrity

RCS Pressure2

Drywell Pressure 2

I5 psia to 1500 psig

0 to design pressure3 (psig)

I Function detection; accomplishmentof mitigation; verification

Function detection; accomplishmentof mitigation; verification

IFour thermocouples per quadrant. A minimum of one measurement per quadrant is required for operation.2 Where a variable is listed for inore than one purpose, the Instrumentation requirements may be integrated and only one measurement provided.3 Design pressure is that value corresponding to ASME code values that are obtained at or below code-allowable values for material design

stress.

0*

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TABLE 1 (Continued)


TYPE B (Continued)

Drywell Sump Level2 Bottom to top I Function detection; accomplishmentof mitigation; verification

Maintaining ContainmentIntegrity

Primary Containment Pressure 2

Primary Containment Isola-tion Valve Position (exclud-ing check valves)

10 psia to design pressure3I Function detection; accomplishment

of mitigation; verification

Accomplishment of isolationClosed-not closed I

TYPE C Variables: those variables that provide information to indicate the potential for being breached or the actual breach ofthe barriers to fission product releases. The barriers are (I) fuel cladding, (2) primary coolant pressure boundary, and (3) con-tainment.

Fuel Cladding

Radioactivity Concentration orRadiation Level in CirculatingPrimary Coolant

Analysis of Primary Coolant(Gamma Spectrum)

BWR Core Thermocouples 2

1/2 Tech Spec limit to 100 timesTech Spec limit, R/hr

10 jjCi/gm to 10 Ci/gm orTID-14844 source term incoolant volume

2000 F to 2300'F

l Detection of breach

34 Detail analysis; accomplishment ofmitigation; verification; long-termsurveillance

11 To monitor core cooling

Reactor Coolant PressureBoundary

RCS Pressure2

Primary Containment AreaRadiation2

15 psia to 1500 psig

I R/hr to I05 R/hr

1 5 Detection of potential for or actualbreach; accomplishment of mitiga-tion; long-term surveillance

36,7 Detection of breach; verification

4Sampling or monitoring of radioactive liquids and gases should be performed in a manner that ensures procurement of representative

samples. For gases, the criteria of ANSI N13.1 should be applied. For liquids, provisions should be made for sampling from well-mixed turbu-lent zones, and sampling lines should be designed to minimize platsout or deposition. For safe and convenient sampling, the provisions shouldinclude:

a. Shielding to maintain radiation doses ALARA,b. Sample containers with container-sampling port connector compatibility,c. Capability of sampling tinder primary system pressure and negative pressures,d. Handling and transport capability, ande. Prearrangement for analysis and interpretation.

5 The maximum value may be revised upward to satisfy ATWS requirements.6 Minimum of two monitors at widely separated locations.7 Detectors should respond to gamma radiation photons within any energy range' from 60 keV to 3 MeV with an energy response accuracy

of ±20 percent at any specific photon energy from 0.1 MeV to 3 W~eV. Overall system accuracy should be within a factor of 2 over the entirerange.

1.97-9

TABLE 1 (Continued)


TYPE C (Continued)

Reactor Coolant PressureBoundary (Continued)

TDrywell Drain Sumps Level2

(Identified and UnidentifiedLeakage)

Suppression Pool Water Level

Drywell Pressure2

Bottom to top

Bottom of ECCS suction lineto 5 ft above normal waterlevel


1

I

Detection of breach; accomplishmentof mitigation; verification; long-termsurveillance


Detection of breach; verification

Containment

RCS Pressure2

Primary Containment Pressure2

Containment and DrywellHydrogen Concentration

Containment and DrywellOxygen Concentration (forinerted containment plants)

Containment Effluent2 Radio-activity - Noble Gases (fromidentified release points includ-ing Standby Gas TreatmentSystem Vent)

Radiation Exposure Rate 2 (in-side buildings or areas, e.g.,auxiliary building, fuel hand-ling building, secondary con-tainment, which are in directcontact with primary con-tainment where penetrationsand hatches are located)

15 psia to 1500 psig

10 psia pressure to 3 times designpressure3 for concrete; 4 timesdesign pressure for steel

o to 30% (capability of operatingfrom 12 psia to design pressure3 )

0 to 10% (capability of operatingfrom 12 psia to design pressure3)

I06 ljCi/cc to IO`2 1%i/cc

10-' R/hrto 104 R/hr

1 S Detection of potential for breach;accomplishment of mitigation

I

I

1

38.9

Detection of potential for or actualbreach; accomplishment of mitiga-tion

Detection of potential for breach;accomplishment of mitigation

Detection of potential for breach;.accomplishment of mitigation

Detection of actual breach; accom-plishment of mitigation; verifica-tion

27 Indication of breach

8 Provisions should be made to monitor all identified pathways for release of gaseous radioactive materials to the environs in conformancewith General Design Criterion 64. Monitoring of individual effluent streams Is only required where such streams are released directly into theenvironment. If two or more streams are combined prior to release from a common discharge point, monitoring of the combined stream isconsidered to meet the intent of this regulatory guide provided such monitoring has a range adequate to measure worst-case releases.

9 Monitors should be capable of detecting and measuring radioactive gaseous effluent concentrations with compositions ranging from freshequilibrium noble gas fission product mixtures to 10-day-old mixtures, with overall system accuracies within a factor of 2. Effluent concentraitions may be expressed in terms of Xe-133 equivalents or in terms of any noble gas nuclide(s). It is not expected that a single monitoring devic.will have sufficient range to encompass the entire range provided in this regulatory guide and that multiple components or systems will be e,needed. Existing equipment may be used to monitor any portion of the stated range within the equipment design rating.

1.97-10

TABLE 1 (Continued)


TYPE C (Continued)

Containment (Continued)

Effluent Radioactivity2 - Noble 10-6 ),Ci/cc to 103 PCi/ccGases (from buildings asindicated above)


TYPE D Variables: those variables that provide information to indicate the operation of individual safety systems and othersystems important to safety. These variables are to help the operator make appropriate decisions in using the individual sys-tems important to safety in mitigating the consequences of an accident.

Condensate and FeedwaterSystem

Main Feedwater Flow

Condensate Storage Tank Level

0 to 110% design flow 00 3 Detection of operation; analysis ofcooling

Indication of available water forcooling

Bottom to top 3

Primary Containment-RelatedSystems

Suppression Chamber SprayFlow

Drywell Pressure2

Suppression Pool Water Level

Suppression Pool WaterTemperature

Drywell AtmosphereTemperature

Drywell Spray Flow

0 to 110% design flowl 0

12 psia to 3 psig0 to 110% design pressure3

Top of vent to top of weir well

30'F to 230 0F

40 0F to 440 0F

0 to 110% design flows 0

2

2

2

2

2

2

To monitor operation






Main Steam System

Main Steamline IsolationValves' Leakage ControlSystem Pressure

Primary System Safety ReliefValve Positions, Including ADSor Flow Through or Pressurein Valve Lines

- 0 to 15" of water0 to 5 psid

Closed-not closed or 0 to 50 psig

2

2

To provide indication of pressureboundary maintenance

Detection of accident; boundaryintegrity indication

I Design flow is the maximum flow anticipated in normal operation.

1.97-l1

I

:

TABLE 1 (Continued)

Category (seeRegulatoryPosition 1.3)Variable

TYPE D (Continued)

Safety Systems

Isolation Condenser SystemShell-Side Water Level

Isolation Condenser SystemValve Position

RCIC Flow

HPCI Flow

Core Spray System Flow

LPCI System Flow

SLCS Flow

SLCS Storage Tank Level

Residual Heat Removal (RHR)Systems

RHR System Flow

RIIR Heat Exchanger OutletTemperature

Cooling Water System

Cooling Water Temperature toESF System Components

Cooling Water Flow to ESFSystem Components

Radwaste Systems

High Radioactivity Liquid TankLevel

Ventilation Systems

Emergency Ventilation DamperPosition

Power Supplies

Status of Standby Power andOther Energy Sources Importantto Safety (hydraulic, pneumatic)

Range

Top to bottom 2

Open or closed 2

0 to 110% design flowI°

0 to 110% design flow10

0 to 110% design flow1 0



Bottom to top

O to 1 10% design flowl 0

320 F to 350'F

32°F to 200°F

2

2

2

2

2

2

2

2

.2

Purpose


To monitor status












*To monitor operation

To monitor system status

0 to 110% design flowI 0

Top to bottom 3

Open-closed status 2

Voltages, currents, pressures

0* I Status indication of all Standby Power a.c. buses, d.c. buses, inverter output buses, and pneumatic supplies.

1.597-1 2

TABLE 1 (Continued)

TYPE E Variables: those variables to be monitored as required for use in determining the magnitude of the release of radio-active materials and continually assessing such releases.


Containment Radiation

Primary Containment AreaRadiation - High Range2

Reactor Building or SecondaryContainment Area Radiation2

I R/hr to I0 R/hr

I0-' R/hr to 104 R/hr for Mark Iand 11 containmentsI R/hr to 1 07 R/hr for Mark IIIcontainment

16,7 Detection of significant releases;release assessment; long-termsurveillance; emergency planactuation

29 Detection of significant releases;release assessment; long-term

16,7 surveillance

Area Radiation

Radiation Exposure Rate2

(inside buildings or areas whereaccess is required to serviceequipment important to safety)

10" R/hr to 104 R/hr 27 Detection of significant releases;release assessment; long-termsurveillance

Airborne Radioactive MaterialsReleased from Plant

Noble Gases and Vent Flow Rate

* Drywell Purge, Standby GasTreatment System Purge(for Mark I and 11 plants)and Secondary Contain-ment Purge (for Mark IIIplants)

* Secondary ContainmentPurge (for Mark 1,11, andIII plants)

* Secondary Containment(reactor shield buildingannulus, if in design)

* Auxiliary Building(including any buildingcontaining primary systemgases, e.g., waste gas decaytank)

10-6 pCi/cc to 105 pCi/cc0 to 110% vent design flow 0

(Not needed if effluent dischargesthrough common plant vent)

10-6 pCi/cc to 104 pCi/cc0 to 110% vent design flow1 0


I 06 pCi/cc to 1 04 pCi/(cc0 to 110% vent design flow1 0


10-6 pCi/cc to 103 pCi/cc0 to 110% vent design flow10


2 9 Detection of significant releases;release assessment

29 Detection of significant releases;release assessment


29 Detection of significant releases;release assessment; long-termsurveillance

* Common Plant Vent or Multi- 10-6 IjCi/cc to 103 pCi/cc.purpose Vent Discharging 0 to 110% vent design flow 10

Any of Above Releases (if* drywell or SGTS purge is

included) 10-6 pCi/cc to 104 pCi/cc


1.97-13

TABLE 1 (Continued)


TYPE E (Continued)

Airborne Radioactive MaterialsReleased from Plant (Continued)

Noble Gases and Vent FlowRate (Continued)

* All Other Identified ReleasePoints

1 06 )jCi/Cc to 102 pjCi/cc0 to 1 10% vent design flow 1 0

(Not needed if effluent dischargesthrough other monitored plantvents)


Particulates and Halogens

* All Identified Plant ReleasePoints. Sampling with OnsiteAnalysis Capability

10 3 pCi/cc to 102 lpci/cc0 to 110% vent design flowl 0


Environs Radiation and Radio-activity

Radiation Exposure Meters(continuous indication atfixed locations)

Airborne Radiohalogens andParticulates (portable samplingwith onsite analysis capability)

Plant and Environs Radiation(portable instrumentation)

Plant and Environs Radio-activity (portable instru-mentation)

Range, location, and qualifica-tion criteria to be developed tosatisfy NUREG-0654, SectionHI.H.Sb and 6b requirements for

emergency radiological monitors

1 09 pci/cc to 10-3 pci/cc

10-3 R/hr to 104 R/hr, photons10-3 rads/hr to 104 rads/hr, betaradiations and low-energy photons

Multichannel gamma-rayspectrometer

313

314314

3

Verify significant releases and localmagnitudes V1

Release assessment; analysis



12 To provide information regardingIrelease of radioactive halogens and particulates. Continuous collection of representative samples followedby onsite laboratory measurements of samples for radiohaloger~s and particulates. The design envelope for shielding, handling, and analyticalpurposes should assume 30 minutes of Integrated sampling time at sampler design flow, an average concentration of 102 JCi/cc of radioiodinesin gaseous or vapor form, an average concentration of 10 liCi/cc of particulate radiolodines and particulates other than radioiodines, and anaverage gamma photon energy of 0.5 MeV per disintegration.

1 3 For estimating release rates of radioactive materials released during an accident.

14To monitor radiation and airborne radioactivity concentrations In many areas throughout the facility and the site environs where It isimpractical to install stationary monitor6 capable of covering both normal and accident levels.

1.97-14

TABLE 1 (Continued)


TYPE E (Continued)

Meteorologyl 5

Range Purpose

Wind Direction

Wind Speed

Estimation of Atmos-pheric Stability

O to 3600 (±50 accuracy with adeflection of 150). Starting speed0.45 mps (1.0 mph). Damping ratiobetween 0.4 and 0.6, distance con-stant < 2 meters

() to 30 mps (67 mph) ±0.22 mps(0.5 mph) accuracy for wind speedsless than 11 mps (25 mph) with astarting threshold of less than0.45 mps (1.0 mph)

Based on vertical temperaturedifference from primary system,'50C to I 0C (-90 F to 1 8 0F) and-0.1 5oC accuracy per 50-meterintervals (±0.3 0F accuracy peri64-foot intervals) or analogousrange for alternative stabilityestimates

3 Release assessment



Accident Samplingl 6 Capa-bility (Analysis Capabil-ity On Site)

Primary Coolant and Sump

*Gross Activity. Gamma Spectrum* Boron Content* Chloride Content* Dissolved Hydrogen or

Total Gas' 8* Dissolved Oxygen* pH

Containment Air

* Hydrogen Content

* Oxygen Content* Gamma Spectrum

Grab Sample

Io pCifml to 10 Ci/ml(Isotopic Analysis)0 to 1000 ppm0 to 20 ppm0 to 2000 cc(STP)/kg

34.17 Release assessment; verification;analysis

0 to 20 ppmI to 13

Grab Sample

0 to 1 0%0 to 30% for inerted containments0 to 30%(Isotopic analysis)

*34 Release assessment; verification;analysis

1IsGuidance on meteorolog;ich measurements is being developed in a Proposed Revision I to Regulatory Guide 1.23, "MeteorologicalPrograms in Support of Nuclear Power Plants."

1 6 The time for taking and analyzing samples should be 3 hours or less from the time the decision is made to sample, except for chloridewhich should be within 24 hours.

17An instailed capability should be provided for obtaining containment sump, ECCS pump room sumps, and other similar auxiliarybuilding sump liquid samples.

1 BApplies only to primary coolant, not to sump.

1.97-15

TABLE 2

PWR VARIABLES

TYPE A Variables: those variables to be monitored that provide the primary information required to permit the controlroom operator to take specific manually controlled actions for which no automatic control is provided and that are requiredfor 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 variablesthat 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.


Plant specific Plant specific I Information required for operatoraction

TYPE B Variables: those variables that provide information to indicate whether plant safety functions are being accomplished.Plant safety functions are (1) 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 andcategory for design and qualification requirements. Key variables are indicated by design and qualification Category 1.

Reactivity Control 0Neutron Flux 10-6% to 100% full power I Function detection; accomplishment

of mitigation

Control Rod Position

RCS Soluble Boron Concen-tration

RCS Cold Leg Water Temper-aturei

Full in or not full in

0 to 6000 ppm

500 F to 4000 F

3

3

3

Verification

Verification

Verification

Core Cooling

RCS Hot Leg Water Temper-ature

RCS Cold Leg Water Temper-ature

RCS PressureI

500F to 7500F l Function detection; accomplishmentof mitigation; verification; long-termsurveillance

Function detection; accomplishmentof mitigation; verification; long-termsurveillance

500F to 7500F I

0 to 3000 psig (4000 psig forCE plants)

12. Function detection; accomplishmentof mitigation; verification; long-termsurveillance

0"IWhere a variable is listed for more than one purpose, the instrumentation requirements may be integrated and only one measurement provided.

2 The maximum value may be revised upward to satisfy ATWS requirements.

1.97-16

TABLE: 2 (Continued)


TYPE B (Continued)

Core Cooling (Continued)

Core Exit Temperaturel 200 0F to ?300 0F (for operatingplants- 200F to 16500 F)

33 Verification

Coolant Level in Reactor Bottom of core to top of vessel I(Direct-indicating orrecordingdevice notneeded)

Verification; accomplishment ofmitigation

Degrees of Subcooling 200 0F subcooling to35 0F superheat

2(With con-firmatoryoperatorprocedures)

Verification and analysis of plantconditions

Maintaining Reactor CoolantSystem Integrity

RCS PressureI

Containment Sump WaterLevel'

Containment Pressure 1


Narrow rxnge (sump),Wide ran e (bottom of contain-ment to 00,000-gallon levelequivalent)


12

2

Function detection; accomplishmentof mitigation



I

Maintaining ContainmentIntegrity

* Containment Isolation ValvePosition (excluding check valves)

Containment Pressurel

Closed-not closed I Accomplishment of isolation


10 psia to design pressure4 I

3A minimum of four measurements per quadrapnt is required for operation. Sufficient number should be installed to account for attrition.(Replacement instrumentation should meet the 2300 F range provision.

4Design pressure is that value corresponding to ASME code values that are obtained at or below code-allowable values for material designstress.

1.97-17

TABLE 2 (Continued)

TYPE C Variables: those variables that provide information to indicate the potential for being breached or the actual breach ofthe barriers to fission product releases. The barriers are (1) fuel cladding, (2) primary coolant pressure boundary, and (3) con-tainment.


Fuel Cladding

Core Exit Temperaturel

Radioactivity Concentration orRadiation Level in CirculatingPrimary Coolant

Analysis of Primary Coolant(Gamma Spectrum)

200 OF to 2300 0 F (for operatingplants - 200 0 F to 16500 F)

1/2 Tech Spec limit to 100 timesTech Spec limit, R/hr

10 pCi/gm to 10 Ci/gni orTID-14844 source term incoolant volume

Detection of potential for breach;accomplishment of mitigation; long-term surveillance

I Detection of breach

35 Detail analysis; accomplishment ofmitigation; verification; long-termsurveillance

Reactor Coolant PressureBoundary

RCS Pressurel

Containment Pressurel

Containment Sump WaterLevel'

Containment Area Radiation1

Effluent Radioactivity - NobleGas Effluent from CondenserAir Removal System Exhaust1

0 to 3000 psig (4000 psig for CEplants)

10 psia to design presiure4 psig(5 psia for subatmosphericcontainments)

Narrow range (sump),Wide range (bottom of containmentto 600,000-gal level equivalent)

1 R/hrto 104 R/hr

1 0 6 PCi/cc to 1 02 _Ci/cc

12 Detection of potential for or actualbreach; accomplishment of mitiga-tion; long-term surveillance

I Detection of breach; accomplishmentof mitigation; verification; long-termsurveillance

21

36.7


Detection of breach; verification

38 Detection of breach; verification

SSampling or monitoring of radioactive liquids and gases should be performed in a manner that ensures procurement of representativesamples. For gases, the criteria of ANSI NI 3.1 should be applied. For liquids, provisions should be made for sampling from well-mixed turbulentzones, and sampling lines should be designed to minimize plateout or deposition. For safe and convenient sampling, the provisions should include:

a. Shielding to maintain radiation doses ALARA,b. Sample containers with container-sampling port connector compatibility.c. Capability of sampling under primary system pressure and negative pressures,d. Handling and transport capability, ande. Prearrangement for analysis and interpretation.

6 Minimum of two monitors at widely separated locations.7 Detectors should respond to gamma radiation photons within any energy range from 60 keV to 3 MeV with an energy response accuracy of

±20 percent at any specific photon energy from 0.1 MeV to 3 MeV. Overall system accuracy should be within a factor of 2 over the entire range.

8 Monitors should be capable of detecting and measuring radioactive gaseous effluent concentrations with compositions ranging from freshequilibrium noble gas fission product mixtures to 10day-old mixtures, with overall system accuracies within a factor of 2. Effluent concentra-tions may be expressed in terms of Xe-1 33 equivalents or in terms of any noble gas nuclide(s). It is not expected that a single monitoring devicewill have sufficient range to encompass the entire range proviled in this regulatory guide and that multiple components or systems will beneeded. Existing equipment may be used to monitor any portion of the stated range within the equipment design rating.

1.97-18

TABLE 2 (Continued)


TYPE C (Continued)

Containment

Range Purpose

RCS Pressurel

Containment Hydrogen iConcentration I

Containment Pressure

Containment Effluent Radio'activity - Noble Gases fromIdentified Release Points1

Radiation Exposure Rate (in-side buildings or areas, e.g.,auxiliary building, reactorshield building annulus, fuelhandling building, which arein direct contact with primarycontainment where penetra- 'tions and hatches are located)'

Effluent Radioactivity, - NobleGases (from buildings asindicated above)


0 to 10% (capable of operatingfrom 10opsia to maximum designpressure )0 to 30% for ice-condenser-typecontainment

10 psia pressure to 3 times designpressure4 for concrete; 4 timesdesign pressure for steel (5 psiafor subatmospheric containments)

10-6 pCi/cc to I 02 pCi/cc

I 0_' R/hr to 104 R/hr

10-6 pCi/cc to 103 pCi/cc

12 Detection of potential for breach;accomplishment of mitigation

1

289

Detection of potential for breach;accomplishment of mitigation;long-term surveillance

Detection of potential for or actualbreach; accomplishment of mitiga-tion

Detection of breach; accomplish-ment of mitigation; verification

2 7 Indication of breach


TYPE D Variables: those variables that provide information to indicate the operation of individual safety systems and othersystems important to safety. These variables are to help the operator make appropriate decisions in using the individual sys-tems important to safety in mitigating the consequences of an accident.

Residual Heat Removal (RHR)or Decay Heat Removal System

RHR System Flow

RHR Heat Exchanger OutletTemperature

0 to 110% design flow 10

3320F to 350'F

2 To monitor operation

2 To monitor operation and for analysis

9 Provisions should be made to monitor all identified pathways for release of gaseous radioactive materials to the environs in conformancewith General Design Criterion 64. Monitoring of individual effluent streams is only required where such streams are released directly into theenvironment. If two or more streams are combined prior to release from a common discharge point, monitoring of the combined stream isconsidered to meet the intent of this regulatory guide provided such monitoring has a range adequate to measure worst-case releases.

c0 Design flow is the maximum flow anticipated in normal operation.

1.97-19

TABLE 2 (Continued)


TYPE D (Continued)

Safety Injection Systems

Accumulator TankLevel and Pressure

Accumulator Isolation ValvePosition

Boric Acid Charging Flow

Flow in HPI System

Flow in LPI System

Refueling Water Storage TankLevel

Primary Coolant System

Reactor Coolant Pump Status

Primary System Safety ReliefValve Positions (includingPORV and code valves) orFlow Through or Pressure inRelief Valve Lines

Pressurizer Level

Pressurizer Heater Status

Quench Tank Level

Quench Tank Temperature

Quench Tank Pressure

Secondary System (SteamGenerator)

Steam Generator Level

Steam Generator Pressure

Safety/Relief Valve Positionsor Main Steam Flow

Main Feedwater Flow

Range Purpose

I

10% to 90% volume0 to q50 psig

Closed or Open

2

0 to 110% design flowI 0



Top to bottom

2

2

2

2

2

Motor current

Closed-not closed

3

2


Operation status






Operation status; to monitor forloss of coolant

To ensure proper operation ofpressurizer

To determine operating status








Bottom to top I

Electric current

Top to bottom

SOF to 7500 F

0 to design pressure4

From tube sheet to separators

From atmospheric pressureto 20% above the lowest safetyvalve setting

Closed-not closed

O to 110% design flow1 0

2

3

3

3

1

2

2

3

1.97-20

TABL.E 2 (Continued)

Variable

TYPE D (Continued)

Auxiliary Feedwater or Emer-gency Feedwater System

Auxiliary or Emergency Feed-water Flow

Condensate Storage TankWater Level

Range

0 to 110% design flow 0 0

Plant specific

Category (seeRegulatoryPosition 1.3)

2(I for B&Wplants)

Purpose


To ensure water supply for auxiliaryfeedwater (Can be Category 3 if notprimary source of AFW. Then what-ever is primary source of AFW shouldbe listed and should be Category 1.)

Containment Cooling Systems

Containment Spray Flow

Heat Removal by the Contaihj-ment Fan Heat Removal System

Containment AtmosphereTemperature

Containment Sump WaterTemperature

Chemical and Volume ControlSystem

Makeup Flow- In

Letdown Flow - Out

Volume Control Tank Level

Cooling Water System

Component Cooling WaterTemperature to ESF System

Component Cooling Water Flowto ESF System

Radwaste Systems

High-Level Radioactive LiquidTank Level

Radioactive Gas Holdup TankPressure


Plant specific

40'F to 4000F

2

2

2

50'F to 2500F 2

O to 1 10% design flow IO

O to 110% design flow1 0

Top to bottom

2

2

t 2



To indicate accomplishment of cooling







To indicate storage volume

To indicate storage capacity

320F to 2000F 2

Oto 110% design flowl 0

Top to bottom

0 to 150% design pressure4

2

3

3

1.97-21

TABLE 2 (Continued)


TYPE D (Continued)

Ventilation Systems

Emergency Ventilation Damper Open-closed statusPosition

2 To indicate damper status

Power Supplies

Status of Standby Power andOther Energy Sources Import-ant to Safety (hydraulic,pneumatic)

Voltages, currents, pressures 211 To indicate system status

TYPE E Variables: those variables to be monitored as required for use in determining the magnitude of the release of radio-active materials and continually assessing such releases.

Containment Radiation

Containment Area Radiation -

High Range 1I R/hr to 107 R/hr 16,7 Detection of significant releases;

release assessment; long-termsurveillance; emergency planactuation

Area Radiation

Radiation Exposure Ratel(inside buildings or areas whereaccess is required to serviceequipment important to safety)

I0-' R/hr to 104 R/hr 2 7 Detection of significant releases;release assessment; long-termsurveillance

Airborne Radioactive MaterialsReleased from Plant

Noble Gases and Vent Flow Rate

* Containment or PurgeEffluent 1

* Reactor Shield BuildingAnnulus' (if in design)

* Auxiliary Buildingi(including any buildingcontaining primary systemgases, e.g., waste gas decaytank)

10~6 j~ifcc to 1 05 pCi/cc0 to 1 10% vent design flow 0


IO6 PCi/Cc to 104 pCi/cc0 to 110% vent design flowl 0


10-6 pCi/cc to 103 pCi/cc0 to 110% vent design flowl 0



28

28

Detection of significant releases;release assessment

Detection of significant releases;release assessment; long-termsurveillance

I IStatus indication of all Standby Power n.c. buses, d.c. buse:;, inverter output buses, and pneumatic supplies.

1.97-22

TABL.E 2 (Continued)


Type E (Continued)

Airborne Radioactive MaterialsReleased from Plant (Continued)

Noble Gases and Vent FlowRate (Continued)

Range Purpose

* Condenser Air RemovalSystem Exhaust 1

* Common Plant Vent or Multi-purpose Vent DischargingAny of Above Releases (ifcontainment purge isincluded)

I0 6 p1Ci/cc to 1 05 pCi/cc0 to 11 0o vent design flow1 0


10-6 PCi/cc to 103 pCi/cc0 to 110% vent design flowl 0



i 0-6 pCi/cc to 104'V pi/cc

* Vent From Steam Gen-erator Safety Relief Valvesor Atmospheric DumpValves

* All Other Identified ReleasePoints

1 0' pCi/cc to 1 03 pKi/cc(Duration of releases in secondsand mass of steam per unit time)

I0 6 pCi/cc to 1 02 pV i/ccI) to 110% vent design flowl 0

(Not needed if effluent dischargesthrough other monitored plantvents)



Particulates and Halogens

* All Identified Plant ReleasePoints (except steam gen-erator safety relief valves oratmospheric steam dumpvalves and condenser airremoval system exhaust).Sampling with OnsiteAnalysis Capability

v -3 pCi/cc to 102 lp i/cc0 to 110% vent design flow 10


1 2 Effluent monitors for PWR steam safety valve discharges and atmospheric steam dump valve discharges should be capable of approxi-mately linear response to gamma radiation photons with energies from approximately 0.5 MeV to 3 MeV. Overall system accuracy should bewithin a factor of 2. Calibration sources should fall within the range of approximately 0.5 MeV to 1.5 MeV (e.g., Cs-137, Mn-54, Na-22, andCo-60). Effluent concentrations should be expressed in terms of any gamma-emitting noble gas nuclide within the specified energy range. Calcu-lational methods should be provided for estimating concurrent rsleases of low-energy noble gases that cannot be detected or measured by themethods or techniques employed for monitoring.

1 3 To provide information regirding release of radioactive halogens and particulates. Continuous collection of representative samples followedby onsite laboratory measurements of samples for radiohalogens and particulates. The design envelope for shielding, handling, and analyticalpurposes should assume 30 minutes of integrated sampling time tit sampler design flow, an average concentration of 102 pCi/cc of radiolodinesin gaseous or vapor form, an average concentration of 10 pCi/cc of particulate radioiodines and particulates other than radiolodines, and anaverage gamma photon energy of 0.5 MeV per disintegration.

1.97-23

TABLE 2 (Continued)


TYPE E (Continued)

Environs Radiation and Radio-activity

Radiation Exposure Meters(continuous indication atfixed locations)

Airborne Radiohalogens andParticulates (portable samplingwith onsite analysis capability)

Plant and Environs Radiation(portable instrumentation)

Plant and Environs Radio-activity (portable instru-mentation)

Range, location, and qualifica-tion criteria to be developed tosatisfy NUREG-0654, SectionHI.H.5b and 6b requirements foremergency radiological monitors .

10 9 1 Ci/cc to 10 3 Igi/CC

I03 R/hr to 104 R/hr, photons10-3 rads/hr to 104 rads/hr, betaradiations and low-energy photons

Multichannel gamma-layspectrometer

314

315315

3

Verify significant releases and localmagnitudes

Release assessment; analysis .



Meteorology 16 I.

Wind Direction

Wind Speed

Estimation of Atmos-pheric Stability

0 to 3600 (±50 accuracy with adeflection of 150). Starting speed0.45 mps (1.0 mph). I)amping ratiobetween 0.4 and 0.6, distance con-stant < 2 meters

0 to 30 mps (67 mph) ±0.22 mps(0.5 mph) accuracy for wind speedsless than 11 mps (25 mph) with astarting threshold of less than0.45 mps (1.0 mph)

Based on vertical temperaturedifference from primary system,-50C to 10C (-90 F to 180F) and+0.150 C accuracy per 50-meterintervals (40.3 0 F accuracy per164-foot intervals) or analogousrange for alternative stabilityestimates




For estimating release rates of radioactive materials released during an accident.

1 5To monitor radiation and airborne radioactivity concentrations in many areas throughout the facility and the site environs where It isimpractical to install stationary monitors capable of covering both normal and accident levels.

G6 Guidance on meteorological measurements is being developed in a Proposed Revision I to Regulatory Guide 1.23, "MeteorologicalPrograms in Support of Nuclear Power Plants."

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TABLE 2 (Continued)


TYPE E (Continued)

Accident Samplingl 7 Capa-bility (Analysis Capabil-ity On Site)

Primary Coolant and Sump Grab Sample 3511 8 Release assessment; verification;analysis

- Gross Activity- Gamma Spectrum* Boron Content* Chloride Content* Dissolved Hydrogen

or Total Gas1 9* Dissolved Oxygen 19

* pH

10 liCi/ml to Io Ci/ml(Isotopic Analysis)0 to 6000 ppm0 to 20 ppm0 to 2000 cc(STP)/kg

0 to 20 ppmI to 13

Containment Air

* Hydrogen Content

* Oxygen Content* Gamma Spectrum

Grab Sample

0 to 10%0 to 30% for ice condensers0 to 30%(Isotopic analysis)

35 Release assessment; verification;analysis

1 7 The time for taking and analyzing samples should be 3 hours or less from the time the decision is made to sample, except for chloridewhich should be within 24 hours.

18An installed capability should be provided for obtaining containment sump, ECCS pump room sumps, and other similar auxiliaryO 1AIding sump liquid samples.

41 1 9 Applies only to primary coolant, not to sump.

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VALUE/IMPACT STATEMENT

1. PROPOSED ACTION

1.1 Description

The applicant for a license (or licensee) of a nuclearpower plant is required by the Commission's regulations toprovide instrumentation to ( I) monitor variables andsystems over their anticipated ranges for accident conditionsas appropriate to ensure adequate safety and (2) monitorthe reactor containment atmosphere, spaces containingcomponents for recirculation of loss-of-coolant accidentfluid, effluent discharge paths, and the plant environs forradioactivity that may be released from postulated accidents.This revision to Regulatory Guide 1.97 proposes to improvethe guidance for plant and environs monitoring during andfollowing an accident, including extended ranges for someinstruments to account for consideration of degraded corecooling.

1.2 Need

Regulatory Guide 1.97 was issued as an effective guidein August 1977. At the time the guide was issued, it wasrecognized that more specific guidance than that containedin the guide would be required. However, the difficultyin developing the guide to the point where it could beinitially issued was evidence that experience in using theguide as it then existed was essential before further develop-ment of the guide would be meaningful.

Therefore, in August 1977, the staff initiated TaskAction Plan A-34, "Instruments for Monitoring Radiationand Process Variables During an Accident." The purpose ofthe task action plan was to develop guidance for applicants,licensees, and staff reviewers concerning implementation ofRegulatory Guide 1.97. Such effort would provide a basisfor revising the guide.

When the staff was ready to issue the results of the TaskAction Plan A-34 effort, the accident at TMI-2 occurred.Subsequently, the TMI-2 Lessons Learned Task Force hasissued its "Status Report and Short-Term Recommendations,"NUREG-0578. This report, along with the draft TaskAction Plan A-34 report, Draft 1 of Regulatory Guide 1.97(dated April 12, 1974), and Standard ANS:4.5, providesample basis for revising Regulatory Guide 1.97.

1.3 Value/Impact of Proposed Action

1.3.1 NRC Operations

Since a list of selected variables to be provided withinstrumentation to be monitored by the plant operatorduring and following an accident has not been explicitlyagreed to in the past, the proposed action should result inmore effective effort by the staff in reviewing applicationsfor construction permits and operating licenses. The proposed

action will establish an NRC position by taking advantageof previous staff effort (1) in completing a generic activity(A-34), (2) in evaluating the lessons learned from the TMI-2event (NUREG-0578), and (3) in conjunction with effort indeveloping a national standard (ANS-4.5). For futureplants, the staff review will be simplified with guidancecontained in the endorsed standard developed by a voluntarystandards group and in the regulatory guide, which includesa list of variables for accident monitoring. Efforts by thestaff to implement Revision I to Regulatory Guide 1.97have been fraught with frustration and met with delaysbecause the guide was adjudged by licensees to be vagueand ambiguous. Revision 2 eliminates the problems encoun-tered with Revision I because it provides a minimum set ofvariables to be measured and hence gives more guidance inthe selection of accident-monitoring instrumentation.Consequently, there will be no significant impact on thestaff. There will, however, be effort required to review eachoperating plant and each plant under review to assessConformance with Regulatory Guide 1.97.

1.3.2 Other Government Agencies

Not applicable, unless the government agency is anapplicant.

1.3.3 Industry

The proposed action establishes a more clearly definedNRC position with regard to instrumentation to assess plantand environs conditions during and following an accidentand therefore reduces uncertainty as to what the staffconsiders acceptable in the area of accident monitoring.Most of the impact on industry will be in the area ofproviding instrumentation to indicate the potential breachand the actual breach of the barriers to radioactivityrelease, i.e., fuel cladding, reactor coolant pressure boundary,and containment. Some instruments have extended rangesand others have higher qualification requirements. Therewill be additional impact due to heretofore unspecifiedvariables to be monitored (i.e., water level in reactor forPWRs and radiation level in the primary coolant water forPWRs and BWRs) that have been identified during theevaluation of TMI-2 experience and will require development.

Attempts were made during the comment period todetermine the cost impact on industry for future plants andfor backfitting existing plants. Estimates ranged from

- $4,000,000 to over $20,000,000. The higher estimatesundoubtedly charged all accident-monitoring instrumentationto Revision 2 to Regulatory Guide 1.97. This should not bethe case. The requirement for accident monitoring hasalways been a part of the regulations. Consequently theimpact of Revision 2 to Regulatory Guide 1.97 should onlybe the delta added by Revision 2. A conservative estimateof the increase in requirements are the additions of Type Cmeasurements and the upgrading of some of the Type B

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- -

measurements to higher qualification of the instrumentation.There are 17 unique Type B and C variables to be measuredfor PWRs, less for BWRs. A conservative average cost foreach measurement is $130,000 making a total cost impactof $2,210,000. If this figure were doubled to account foroverhead costs and about a 15 percent contingency added,the cost impact would be about $5,000,000. This costestimate is the same for operating plants as for plants underconstruction and future plants. While it is recognized thatfor operating plants the costs associated with backfittingare generally higher than the costs associated with newplants, some concessions are made in some requirements asa result of existing licensing commitments that bring thecost estimate to about the same value.

1.3.4 Public

The proposed action will improve public safety byensuring that the plant operator will have timely informationto take any necessary action to protect the public.

No impact on the public can be foreseen.

1.4 Decision on Proposed Action

As previously stated, more definitive guidance oninstrumentation to assess plant and environs conditionsduring and following an accident should be given.

2. TECHNICAL APPROACH

This section is not applicable to this value/impactstatement since the proposed action is a revision of anexisting regulatory guide, and there are no alternative-;to providing the plant operator with the required information.

3. PROCEDURAL APPROACH

Previously discussed.

4. STATUTORY CONSIDERATIONS

4.1 NRC Authority

Authority for this guide would be derived from thesafety requirements of the Atomic Energy Act. In particular,Criterion 13, Criterion 19, and Criterion 64 of Appendix Ato 10 CFR Part 50 require, in part, that instrumentation beprovided to monitor variables, systems, and plant environsto ensure adequate safety.

4.2 Need for NEPA Assessment

The proposed action is not a major action as defined inparagraph 51.5(a)(10) of 10 CFR Part 51 and does notrequire an environmental impact statement.

5. RELATIONSHIP TO OTHER EXISTING OR PROPOSEDREGULATIONS OR POLICIES

No conflicts or overlaps with requirements promulgatedby other agencies are foreseen: This guide does include thevariables to be monitored on site by the plant operator inorder to provide necessary information for emergencyplanning. However, information on emergency planning andits relationship to other agencies is provided elsewhere.Implementation of the proposed action is discussed inSection D of this revision.

6. SUMMARY AND CONCLUSIONS

Revision 2 to Regulatory Guide 1.97, "InstrumentationFor Light-Water-Cooled Nuclear Power Plants to AssessPlant and Environs Conditions During and Followingan Accident," should be issued.

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Regulatory Guide 1.97, "Instrumentation for Light-Water - NRC

Documents

Transcript of Regulatory Guide 1.97, "Instrumentation for Light-Water - NRC