Regulatory Guide 1.97: Difference between revisions

Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant & Environs Conditions During & Following Accident
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Issue date:	05/31/1983
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Latest revision as of 02:06, 17 January 2025

Revision 3 U.S. NUCLEAR REGUlATORY COMMISSION

May1983 REGULATORY GUIDE

OFFICE OF NUCLEAR REGUIATORY RESEARCH

REGULATORY GUIDE 1.97 INSTRUMENTATION FOR LIGHT-WATER.COOLED NUCLEAR POWER PLANTS

TO ASSESS PLANT AND ENVIRONS CONDITIONS DURING AND FOLLOWING

.

AN ACCIDENT

.

A. INTRODUCTION

Criterion 13, .. Jnstrumentation and Control, .. of Appen- dix A, "'General Design Criteria for Nuclear Power Plants,"

to 10 CFR Part*SO, ~Domestic Licensing of Production and*

Utilization Facilities," includes a requflement that instru- mentation be provided to monitor variables and systems over their anticipated ranges for accident conditions as appropriate to ensure adequate sa!ety.

Criterion 19, '"Controi Room," of Appendix A to 10 CFR

Part SO includes a requirement that a control room be pro- vided from which actions can be taken to maintain tlienucl~

power unit in a safe condition under accident conditions, including loss-of-coolant accidents, and that equipment, including the necessazy instrumentation, at appropriate 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 SO includes a requirement that means be provided for monitoring the reactor containment atmosphere, spaces conqining components for recirculation of loSH)f-coolant accident fluid, effluent disclwge paths, and the plant environs for radioactivity that may be released from postulated accidents.

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 *. The Advisory Committee on*

Reactor Safeguards has

bett .. consulted concermng this guide and has concurred in th~ regulatory position.

Any auidance in. this document related to information*

collection actmties has been cleared under 0MB Oearance No. 3150-0011.

.

USNqc REGULATORY GUIDES

Rew1,!_,ltory Guides are Issued to describe and make available to th*

publ~ methoas acceptable to the NRC staff of lmptementlnt IP1 eclfk Parts of th* Commlsslon'S r.11ulatlons, to delineate tech* *

n ques used by th* staff In evaluating IPKlflc: problems or postu*

l!t1!!_21Ccldants1 or to provide guidance to applicants, Regulatory u .. .,. are no1 substitutes tor reguhltlons, and compliance with tha1m1 Is not niqulred. Methods and solutions different from those set ou n the guides will lie accaptable If they provide a baSls for th*

f111ndlngs Nqulslte to the Issuance or continuance of a permit or c:ense by the commlssJon.

This guide wu Issued after Consideration *or comments received from the public:, comments and suggestions for Improvements In thase guides are encouniged at all times,. and guides wlll be nvlsed as appropriate, to accommoe1ate comments and to reflect new Inform ..

tlon or experience.

B. DISCUSSION

-

Indicatiqns of plant variables are required by the control room operating personnel during accident situations to (l)

provide information required to permit the operator to take preplanned manwd actions to accomplish ufe plant shut- down; (2) determine whether the :reactor trip, cngineen:d- safety-feature systems, and manually initiated safety systems and other systems important to safety are performing their intended functions (ie., reacti'rity 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 bame.rs to radioactivity release (i.e., fuel cladding, reactor coolant pressure boundary, and containment) and to determine if a gross breach of a barrier has occurred. In aclclition to the above, indications of plant variables that provide informa- tion on operation of plant safety systems and other systems *.

important to afety are required by the control room operating personnel during an accident to (1) fmmsh 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 1he need to initiate action necess:uy to protect the public and for an

estimate of the magnitude of any impending threat.

At the &tut 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 appropriat~ response. For this reason, reactor trip and certain other safely actions (e.g ** emezgency 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 plant nrlables required to enable the operation of manually imtiated safety systems and other appropriate operator actions involving systems important to safety.

Comments lhould be sent to the ~

of the b>>mmls:slon

u.s. Nuclear Regulatory Cofflmlsslon, Washington o c. 20555'

Attention, Docketing anCI Service Branen.

'

TIie guides are Issued In the followtng ten Droad divlSlons:

1. Po_, Reactors

6. Products

. 2. Research and Test Reactors

7. TransPortatlon

. 3. Fuels and MatMlals Facilities

e. Occupational Health

4.5 Environmental and Siting

9. Antitrust and Financial Review

Materlalsand Plant Protection 10. General Co9111s of Issued guides may bepurc:hasedatthecurrent Government Printing Office price. A sublcrlptlon Mntlcc tor future guides In dt,lc: divisions Is avallable through the Government Printing ott.:"

In ormatlon on the rubsc:rlptlon llln,lca ancl current GPO prices ma* *

be obtained by writing tlltl U.S. Nuc:lur Ragulatory CommllSlof Wasttlngton, D,C. 20555, Attention, Publications Sales Manager:

"

Independent of the above tasks, it is important that operators be informed Jf the banicn to the release of radioactive materillls

are being challenged. Therefore, it is es.,ential that instrument ranges be selected so that the instrument will always be on scale. Narrow-range instnunc:lits may not have the necessary range to track the course of the accident; consequently, multiple instnlments With over- lapping ranges may be necessary. (In the past, some instru- ment ranges have been selected based on the setpoint value for automatic pzotec:tion or alarms.) It is essential that degraded conditions and their magnitude be identified so the openton can take actions that are available to mitigate the consequences. It is not intended that operaton be encouraged to prematurely circumvent systems important to safety but that they be adequately informed in order that unplanned actions can be taken when necessary.

Examples of serious events that could threaten safety Jf conditions degrade are lms-of-coo!ant accidents (LOCAi),

oveipressme transients, anticipated operational occ:um,nces that become accidents such as anticipated transients without scram (ATWS). and reactmty exc:msicms that result in releases of radio~ materials. Such events require that the operators understand, withm a short time period, the ability of the bamen to limit radioactivity release, i.e., that they undemand the potent1al for breach of a barrier or.

whether an aetua1 breach of a banicr has occmred because of an accident in progress.

It is essential that. the reqlJired instrumentation be capable of surviving the accident environment in which it is located for the length of time its function is requfrecl. It couid therefore either be designed to withstand the accident enviJOnment or be protected by a local protected en'liron*

meat.

It is desirable that accident-monitoring instrumentation componentl and their mounts that cannot be located in seismically qualified buildfnp be designed to ~tinue to

"function, to the extent feasible, followiq seismic m,nts.

An acceptable method for enhancing the sewnic resistance of this iDstlUmentation would be to dcsip it to meet the seismic criteria applicable to like instrumentation installed in seismically qualffied locations although a lesser ova--

. all qualification resu1ta.

Vanables for ac:cidcmt monitoring can be selected to provide the cssentfal information Deeded by the operator to detenmne if the plant safety funcdom an being performed.

It is mcntial that the range selections b

e. sufficiently

1 great to keep instruments on scale or that one of a set of overlapping mstruments will be on scale at all times.

Further, it is prudent that a limited number of those .

'lariablea that !1%11 functionally sfgmficant (e.g., containment pressure, primary system pressme) be monitored by mstru- mentl qualified to more stringc:nt en'rironmental require- menu and with .ranges that extend well beyond that which the selected varlables cau attain under limitfna* con~;

for example,

range for the conwnment pressuie monitor extending to the burst pressure of the containment iD order that the

operaton will not be uninformed as to the pressuie inside_ the containment. Toe availability of such .instruments is important so that responses to corrective actions can be observed and the need for, and magnitude of, further actions can be determined. It is also necessary to be suze that when a range is extended, the sensitivity and accuracy *

of the instnunent are within acceptabie limits for monitor- *

ing the .extended range.

Normal power plant instrumentaticn remaining functional for all accident conditions can provide indication. records,*

and (with certain typeS of instruments) time-history responses for many variables important to followinJ the course of the *

accident. Therefore, it is prudent to select the required acc::ident-monitoring instriunentation from the normal power plant instramentation to

enable openton to use, during accident situations, instruments with which they an most familiar. Since some accidents could impose ~

operating requirements on instrumentation components, it may be nccesm)' to upgrade those normal power plant instrumentation components to withstand the mme severe .

operating conditions and to measme greater variations of monitored variables that maybe associated withan accid=t.

It Js essential that instrumentation so upgraded does not degrade the accuracy and sensitivity requife4 for normal .

operation. In some cases, this will necessitate use of OTil:'-

lappjq ranges of instruments to monitor the required* ranp.

of. the variable to be monitored, possibly with dli'f'erent performance requirements in each nnge, ANSI/ANs-4,5-1980,1 "Criteria for Accident Monitozina Functions 1D light-Water-Cooled Reactors," delineates criteria for determininJ the val'iablea to be monitored by the control room* operator, aa reqwred for safety, dminl the course of an accident and during the long-tmn stable shutdown . phase followin& an accident. ANM.S .. was prepared by *working Group 4.S of Subcommittee ANs-4 with two primary objectives: (1 )to address that instrumenta- tion that permits the operators to monitor expected param- eter changes in an accident period and (2) to addresl extended-range instrumentation deemed appropriate for the possibility of encounterlDJ ~usly unforeseen events.

ANS4.S references a revision to IEEE Std 497-1977,

"IEEE Standard Critcrla for Accident Monitorina Jnstru.

mentatioJl for Nuclear Power Generating Stations, " 3 31 the source for specific instrumentation design criteria. Since the revision to IEEE Std 497

has not been completed, its applicability cannot yet be detmnin.cd. Hence, specific instrumentation* design criteria han been included in this re~gaide.

ANS-1.S definea three types of variables (definitiom modified herein) for the purpose of aidin& the desigJler in selectinJ accident*monitorlnl instrumenta~cn and applicable criterla. lbe types an: Type A, those varlablea that piovide

1c~ may be obealned from tha Ammfca Nudear So~,

5SS Nmtla ~

AWU111, La Grmp l'lrk. Dl1Doll 60525 *.

2~.m&J be obealned from ihe Immme of Electdcll an4 Electromel En~ Inc.. 345 Eat 47tJa su.t. New York.

NcrwYort 1001'7, *

1.97-2

primary information3 needed to pennit the control room ' * : BWRs) and Table 3 (for PWRs). The criteria are separated openting penonncl to tat: the specified manually controlled .

into three separate groups or categories that provide a actions for which no automatic control ls provided and that p-aded approach to requirements depending on the impor- are tequircd for safety systems to accomplish their safety tance to safety of the measurement of a specific van.able.

functions for design basis accident events; Type B~ those Category 1 provides the most stringent zequimnents and is .

variables that provide infonnation to indicate whether plant intended for key varlables. Category 2 provides less stringent safety functions are being accomplished; and Type C, those tequimnents and generally applies to instrumentation variables that provide information to indicate the potential

designated for indicating system operating status. Category 3 for bemg breached or the actual breach of the bamers to is intended to provide iequircmcnts that will ensure that fission product release. i.e .* fuel cladding, primary coolant hfgh~allty off-the-shelf instrumentation is obtained and pressure boundary. and containment (modified to reflect applies to backup and diagnostic instrumentation. It is also NRC staff position; see regulatory position 1.3). The

used where the state of the art will n~t supporUequirements somccs of potential breach are limited to the energy for higher qualified instrumentation.

sources within the barrier itself. In addition to the ae(?dent- monitorlng variables provided in ANS4.S, miabJes .for

.monitoring the operation of systems important to safety and ndioactive effluent Jelcases are pro'Vided by this

,:egulatory guide. Two additional variable types are defined:

Type D, those Yarlables that provide information-to indicate

. the operation of individual safety systems and other systems important to safety, and Type E, those Ta!iables to be monitored as required for use in determining the magnitude

. of the release of radioactive materials and for continuously assessing wch releases.

A minim.um set of Type B. C, D. and E variables to* be measmed is listed in this :n:gulatory guide. Type A variables have not been listed because they ue plant specific and will depend on the. opmtions that the designer chooses for planned manual action. Types B. C, D, and E are nrlables for following the course of an accident and are to be used

(1) to cletmnine if the plant is responding to the safety measures in operation and (2) to infcmn the operator of the necessity for unplanned actions to mitigate the con*

sequences of an accident. 1he 6.ve dassirications are.not mutually exclusive in that a given wrlable (or instrument)

may be applicable to one or more types, as wdl as for normal power plant opention or for automatically initiated safety actions. A variable inciuded ~ Type B, C, D, or E

does not preclude that Qriable from also being included as Type A. Where mch multiple use occurs, it is essential that 'instrumentation be capable of meeting the more.

stringent requirements.

The time phases (Phases I and *m delineated in ANS4.S

are not used in this regulatory guide. 1bcse considerations are plant specific. It is important that the required instru*

mentation survive the accident.-en'Vironment and function as long as the information it provides

is needed by the control room openting pmonnel.

' 1, ..

The NRC staff is willing to work with the ANS,worl:::ing group to attempt to resolve the above differences.

Regulatory position 1.4 (Table 1) of this guide provides design and qualification cri~ria for the instmmentation used to mcasuie the ftrious variables listed in Table 2 (for*

~

Information Is information that is essential for the

. dfm:t accomplilhment oftbe ~

afetyfmlctiom;lt4oesnot ladude those ffriables am are usoc:lated wlih contingency actions that may also be idelltified in written proccdmcs.

In acneral, the measurement of a single key varlable may not be sufficient to indicate the accomplishment of a given safety function. Where multiple vuiables *are : needed to indicate the accomplis!unent of a given safety function, it is essential. that they each. be

0

considered tey Qriables and be measured with high-quality instrumentation. Additionally, it is prudent, in some instances, to include the measurement

. or additional variables for backup information and for diagnosis. Where these additional measumnentsucincluded, the measures applied for design, qualification, and quality asmnnce of the instrumentation need not be the same as th.at applied for the instnunentation for key. variables. A

key varlable is that single nnable (or minimum number of variables) that most clirect1y .indicates the accomplishment of a safety function (in the case of Types B and C) or the operation of a safety system fm the case of Type D) or radioactive mat~ release (in the case of Type E). It is essential that key variables be qualilied to the more smngent design *and qualification criteria. The design and qualification criteria category assigned to each ftriable .indicates whether the vlliable is considered to be a tey vmable or for system status indication or for backup or diagnosis, Le., for Types B.

and . C, the* tey vwb!es are Category 1: backup qrlables are generally Catego:y 3. For Types D and E, the key . *

Ya!iables are generally Category 2; backup variables are Category 3.

"lhe nr.iables are listed, but no mention (beyond redun- dancy requirements) is made of the number of points of measumnent of each nri.able. It is important that the number of points of measurement be sufficient to adequately indicate the variable nlue. e.g., containment temperature may ~quire spatial location of several points of measure- ment.

'Ibis guide p~"Vides the mimmum number of Ya?iables to be monitored .by the control room operating personnel during and following an aceident, These varlables me used by the eontrol room operating penonnel to perform their role in the emergency plan in the evaluation,* assessment, monitoring, and execution of control room functions when the other emergency response facilities ate not effectively manned. Variables are also defined to permit operators to perl'orm their long-term monitorlng and ~xecution iespon*

abilities after the eme!Ef!ncy response facilities are manned.

The application of the clitetia for the instrumentation is limited to that part of the instrumentation system and

1.97-3

--: **-

its vital supporting features or power sources that provide the direct display of the variablCS:. These provisions are not *

necessarily applicable to that p~ of the instrumentation systems provided as opetator aids for the purpose of enhancing information presentations for the identification or diagnosis of distmbances.

C, REGULATORY POSITION

1. Aceident-Monitorini lnstmmentation The criteria and requuements contained in ANSI/ANs-4.S-

1980, "Criteria for Accident Monitoring Functions in Ught*

Water-COOied Reactors;" are considered by the NRC staff to be generally acceptable for providing instrumentation to monitor variables for accident

conditions subject to the following:

l.l Section I of ANS4.S references IEEE Std 497-_ *

1977.

The specific applicability or acceptability of this standard has not yet been determined.

1.2 Instead of the definition given in Section 3.2.1 of ANS-4.S, the definition of Type A nriables should be:

Type A. those variables to be moJlitored that provide the prlmary information3 ttquired to pennit the control room operators to take the specified m~ually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety function for design basis accident events. *

1.3 In Section 3.2.3 of ANS-4.S, the definition of Type C includes two items, (I) and (2). Item (1) includes those insauments that indicate the extent to which vanables that have the potential for causing a breach in the primary reactor containment have exceeded the design basis values._

In conjunction with the variables that indicate the potential for causing a breach in the primary reactor containment.

the varlables that indicate the potential for causing a breach in the fuel cladding (e.g., core exit temperature) and the reactor coolant pressure boundary (e.g., reactor coolant presmre) should also be included. The sources of potential breach are limited to the energy soun:es within the cladding, coolant boundary, or containment. References *to Type C

instniments, and associated parameten to be measured, in ANS4.5 (e.g., Sections 4.2, s.o, S.1.3, s.2. 6.0. 6.3) should include this expanded definition.

1.4 Section 6.1 of ANS4.S pertain_s to general criteria for Types A, B, and C accident-monitoring variables. In lieu of Section 6.1, the design and qualification cnteria cate- gories in Table 1 should be used for the variables in Tables 2 and 3.

In general. Oitegozy 1 provides for full qualification, redundancy, and continuous real-time display and :requires onsite (standby) power. Category 2 provides for qualifica- tion but is less stringent in that it does not (ofitself)include seismic qualification, redundancy, or continuous display and requires only a high-reliability power source (not necessarily standby power). ~tegoiy 3 is the least strin-1_

gent. It provides for high-quality commercial-grade equip.

ment ~t requires only offsite power.

l.S Sections 6.2.2. 6.2.3, 6.2.4, 6.2.S, 6.2.6, 6.3.2,

6.3.3, 6.3.4. and 6.3.S of ANS-4.S pertain to variables and variable ranges for monitodng Types B and C variables. In

. coztjunction with the above-listed sectiom of ANS4.S.

Tables 2 and 3 of this regulatoiy guide (which include those valiables mentioned hi these sections) should be considered as the minimum number of instruments and their respective ranges for accident-monitoring instrumentation for each nuclear power plant.

.2. Instrumentation for Monitorin1 Systems Operation and Effluent Release

2.1 Type D vmablesare those that providemfonnation to indicate the operation of indbidual safety systems and other systems important to safety. _Type E vmables are those that are to be monitored as requmd* for use in dctennining the magnitude of the release of radioactive materlals and in . continuously assessing such releases.

2.2 The plant designer should select YUiables and information display channels :required*. by Ilia design to enable the control ,:oom operating penonnel to:

a. Ascertam the operating status of each indmdual safety system and other systems important to safety to that extent necessary to determine if each system is operatiq or can be placed in operation to help mitigate the consequences of an accident.

b. Monitor the effluent discharge paths and environs within the site boundary to ~

if there bawi been

significant releases (planned or unplamied) of radioactive matenals and to continuously assess such releases.

c. Obtain requind infoimation through a backup

or diagnosis channel if a single channel may be likely to gm ambiguous indication.

2.3 . The pxocesa for selecting system opemion and effluent release varlables should include the identification

.of:

a. For Typ"e D

(1) The plant safety systems and other systems important to safety that should be operating or that could be placed in operation to help mitigate the consequences of an accident; and

(2) The vanable or minimum number of vmables

1hat indicate the operating status of each system identified in (1) above.

. l.97-4 j

I

.

!

\\ I

' I

) ,.

TABLE1 DESIGN AND OUALIFfCATION CRITERfA FOR INSTRUMENTATION

Categary2 I. Equipment Qaallflcatlon The instrumentation mould be qualified In l!ICCOl'dance with Regulatory Guido 1.89, "Quallff~on of aass IB

Equipment for Nuclear Power Plants," and the met!iod*

. ology described In NUREG-05881 "lnterftn Staff Posi- tion on Bmironmental Qaallflcatlon of Safety~Related Electrical *Equipment. n4 Instrumentation whose nnses ate'l'~Npdred to extend beyond those nnges c:a!culated In the most seffl9 design

. basis accident e.ent for a glyen '1diable shoald be,quaU*

fled uslrlg the guidance prmicfed In pmgraph 6.3.6 of ANS4.5.

-

QuaUftcatlon applies to the complete fnstmmentatlon to

,a channel from sensor to d&play where the dfsplay Is a

"' .

cllnet-lndicatlns meter or recorcUns*dmee. rr the lnstru*

mentatlon channel signal b to~ usod fn a computer-

'

based .display' recording. or diagnostic program, qualifi- cation applies fn>m the sensor up to and fncludlng the channel Isolation dmce.

The 11ebmlc portion of quaUftcatlon should be fn accor- dance with Regulatory Gulde 1.10(), 11Sefsmlc Quallfica*

tlon of Elechfc Equipment for Nuclear Power Pf ants."*

IMtmmentatlon ahoaid continue to read within the required aceancy foDowfn,r, but not necessarily durlnB.

a safe shutd_own earthquake,

2. Redundl!ICJ'

No llngle fafJure within either the accldent~onltodng Jnmumentation,fts auxdfary supporting features,. or Its power sources concurrent with the fallum that are *

I. Equipment Qu.Uficatloit Same as Category* 1 Same as Cate80l'Y I

Same u Category I

No specific prmfslon Z. Redundancy No specific prcmslon

4cop1e1111'11 amtbbh from the NRC/OPO SIies Prosnm. U.S. Nudeat ltesulltOr>' Coinmtalan1 Wuhlnston, D.C. 20555, Cetegory3 I. Equipment Qul!fflcallon No specific provision No specific proYislon No specific provision No specific provision

2. RedundlftCf No specific promfon

Category 1 a condition or reault of a speclflc accident should prevent the operators uom being presented the informa- tion necessuy for them to deteimine th" 11afety status of the plant and to bdng the plant to and maintain il in a safe condition followin& that accident. Where failure of one acddent-monitodng channel results in intoima- tion aml>Jaulty (that Is, the redundant diaplaya di.Agree)

that could lead opeiaton to defeat or fall to accomplish a required safety function, additional info1D1ation should *

be provided to allow the operatoJS to deduce the actual conditions in the planL This may be accomplished by providing additional independent channel& of infonnation of the same variable (addition of"an identical channel)

oi by providing an .independent c:hannel to monitor a different variable that .. bean a.kno'!'ft relalioaabip to the*

-

multiple channela (addition of a diveise channe.l). Redun-

~

dant or diverso channela should be elec.trically independ*

t

ent and physlcally separated from each other and from equipment not c1usified lm~t to llilfety in accor*

~ce with Regulalory Gwde 1.7.S, "Physical lndepend-

. ence of Elecbic Systems," up to and .including any i&ola-

1 tion device. Within each redundant dividoa oh safety

ayate~,.redundant monito~ cbannela ~

not needed except for ateam generator level inatnamentation in two,ioop planta.

.

3. Power Source :

The iuatnun~tation should be enerpzed from s~tion standby PPW'1 iOUROS aa pro¥.ided in Regulatory Qulde

1.32, "Criteria for Safety .. Related Elccldc Power Syatema

. for Nuclear Power Ranta," and ahould be backed up by .

battetica where momentary intenuption is not toienble.

TABLE 1 (Continued)

Catqory2

3. Power Source

1lle.instnan1entation should be e1nerglzed from a hidi-reliability power soun:e, not necessarily standby power, and mould be hKked up by batteriea where momentary interruption is not

tolerable *

Category3

3. Power Source .

No specific provision

. ,.,_, ________________ ........ ______ . ---**

!

I

___ ____J

4. Ctumnel Anlla'bWtr The instmmentatlon channel should be aftllabJe prior to I

an accident except as p!Oflded in pmRflph 4.11, .. Excep.

tion,n is d~fined in JBBB Sfd 279-1971, *ecrtteda*for*Pro- tectlon Systems for Nuclear Power'Generating Stat1ons,n2 or u s~fled fn the technical specifications.

5. QullltJ Ammance The recommendations of the f0Uowm1 regulatory guides pertaining to quality assurance 11fOQ)d 6111 foUowed:

.

.

,.

RegulatOty Guide 1:28 "'Quality Anuran~ Pl'Ogram

. Requirements (Design and Construction)'~

Regulatory Gulde 1.30

(Safety Gulde 30)

"Quality Assurance Reqtlhe- ments for the Installation, (Mpectlon, and .Testing of lftstmmentation and Electdc Equipment" *

Regulatory Gulde 1.38 "Quality Aisurance Require- ments.for Packeifn& Shipping, Recebfn& Stonge, and Han*

cllln1 of Items for Wlltef..Ccoled Nuclear Power Plants" *

Regulatory Gulde 1.58 "Quallfk:ation of Nuclei!lr Power Plant ~on. Examination,. *

and Testfnl Pmonne1"

Regulatcny Gulde 1.64 "'Quality MSUrance Reqahe- ments for the Design of Nuclear Power Plants"

Regulatory Gulde 1.74

1'Quallty Alsunnce Tenns and Definitions"

Categurf 2

4. Channel AftllebUlty The out-of-service interval should be bned on normal technical specification requirements on out of semce for the 11)'9teni 1t nnes where *appllcable or where specified by other nqulrements.

5. Qalllty Aaarance Same as Categor, I u modified by the foDowfng:

Since some lnmumentatlon ls less Important to

111fety than other Instrumentation, it may not be necessny to apply the nme quality 1199Uranee meauft!S to all Instrumentation. The quality 11m1r- ance requirements that ire Implemented should provide control cm,r aeilv!tles affecting quality to an extent consistent with the Importance to safety of the Instrumentation. Thffll requirements should be determined and documented by penonnel knowl- edgeable 1n the end Uffl of the Instrumentation.

Cetegorf 3

4. Channel A..UabUlty .

No specific provision

5. QaaUty Aimnnee Theinstmmentatlonshoutd be of hig

h. quality

commm:ial pde and should be selected to withstand the specified service enmonment.

....

s. (Ceetfnued)

Re11*J1&o,y Gulde 1.88 "Olllectioa. Storap. aacl Mala*

tcDlilco of Nuclcir Povm llaut Ql&lllt; Auuraaee R.ocoru" *

llqula&o,y Gulde 1.123 "Quallty"Aaurwe Raqulre- menta for Conuol of Pr~

men& of ltollll ad Somcu

for Nuclear Power* Ranta"

lleanJlfoJY Gulde 1.144 HAuclltlq of Qlllllty Aliuraace Propuu for NIIOlear Power Banta"

.

Replatoiy G* 1.14' "QuallflcaUon or QuaU.ay Auur- aau:o Propam Aad1t bnonael .

..:. .

for Nuclciu Powe, ftanll"

s do

. Refenaco to tho above maulato,y aulclea. (oxc:ept Regula- tory Guldea 1.30 aad 1.38) u bo1na made pcndlq llluanc:e of a miai~n to Rqulatory Gulde 1.28 that u under elev~

opmoat (Tuk RS 002-5) and that will endone ANSI/ASM.B

NQA*l-15'75', "Quality Auuraac:o Proaram llequJremeata for Nuclear Power flanta. "1

.

. 6. DJlplay_ lad llecoldiq ContlnUOUI roal-tlmo dlaplay ihoulcl be pJovlded. 1be .

iadlcatloa may bo on a dial. dl&ltal dlaplay I CRT. or .

atdpchirt recorder. *

Recording of lmtnameatatlon readout lnfonnatlon should be provided for at leut one redundant channel.

...... ~ ............ ,

6. ~llplay ud llecordiq The lnlwmentatloil sJanal may be d,iaplayed on an iocllvldual inatnameat or it may be proc:cued for .

display on demand.

Signa1a from efOuent radioactivity moniton and area monltora ihould be recorded.

.

c:opa.11111 be oblllald bom tbo AIIMricla Soclat, of Meclwalcal Ea&laNn, 341 laat 41th Strtot. Naw York, NGW York 10017.

Calapy3 Sanie as category 2 Slgnala from efOuent radioactivity moniton.

area moniton, and meteorology monitois should be recorded.

/

---.. **>**--------------------- **--**.

--- .-.. ***** ... --... *--------*-----

-

~

6. * (Contbmed)

.*

If direct and immediate trend or trinslent bitoiinatlon *

Is osmrtlal for operator Information or action, tho recordln1 should bo continuously a'flflable on ndun- .

dant dedicated recorden. *. Otherwise~ It may be. con-

tfnuousty updated, stored In computer menior,, *and

,dfsplayed on demand.* Intermittent dl!pla;i such a

data loam and acannlfl1 recorden may be ueed If no m,nincant tmment tespome lnfonnatlon Is llte1y to

. be Jost by s,ich dmces.

7, Rap ff two or more lnmuments are needed to ccmr a ~-

partlcidn: nilie, oml1ppfn1 of'IM_tmmont IPlft should be prcmded.

ff tlio required nmse of moni- toring Jnmumentatlan ren!ta 1n*a loss of 1nstru.:

mentatlon senslfhfty In the normal operatfn1 mise, separate fnltraments should be *4; .' *

8. Equipment ldentltlcaffan . .

- .. .

. ..

.

.. .

.

Types A, 8~ and C ~struments d*ated as Cate- gories I and 2 *should be specftlcaby Identified with a common desllnatfon on the control pnnels so that tho operator can easn, dbc:em that they m. Intended for.

use under accident ecmdltlons,

. *

. .

,. lnterfleel I

nae tranmib.'lioii of lfpa!s for other use sboold be throush ~latfon dmces that are desl11_1ated as put of the monitoring Instrumentation and that meet the ~ons of thfs document.

10, Semctn,, Testfn1, and callbntlon Senfclna, testing, and calibration proBJ'lms should be specified to maintain the capabDfty of the monitoring lnstmmentatlon, If the requln!d lnteml between

TABLE 1 (Continued)

Cltegary2 Same u Cateaory J *

1, Rase Samo as Catesor, I

8. Equipment ldffltfflcatlon Same as Catel0!7 I

,. lllfflflffll Samo u Category I

10. Senlelnt, Testlq, lftd Calibration Same as Category I

Same as CatOBOl'J I

7. R111111

. Same a Category 1

8. Bqalpment ldentlfleaflon No speclffc pro'flslon

. **~. lnterf-.

No specific prcmslon

10. ~emclnt. Testlnr, and CaUbntlon Same a CateBory I

-

{o ,.. -

~

Ca&agory 1

10. ~CGia&lausd)

tullna 1a leu than the nonnal time inteival between plant llhutdowna, a capablllty for teatln& dUIUII power opc,atlon lhould be proVidod.

Wbenovei meana for removina cbannela &om service ue iDcludod in tho.dalaa, tho dealgn lhould facilllate administrative control of the acc:eu to auc:h removal

.

. .

\\

..

'

meam.

.

'

Tho dealgn sliould faoilltato admlnlatratlvo control of

. the accou to qll setpolnt adjuatmenta, modulo calibn- tlon adjuabnonta, and teit polnta *

. Pedodlc c:heckioa. t~IJN, callbratlon, and callbration

~tlon lhould bo In acco~ with tb,e llJJpllcable ponlOJU of. Re&ulatoiy Guido 1.118.* .. Pedodlc Tosilni of Elcctdc Power and Proteetion Syatema," pe.rtalalna to tNtlns.of lqabument ~ela. (Note: . ~.\\'ome Um" teatiq not UIIJIUy needed.)

lbe locatlon of the ilola&lon dev1co ;houJd be auc:h tbat if would bo accea&lble for malntoaaace dwioa acc:Jdcwt.conditlou.

11. H\\PIWI 'Fldon Tho Jna&riunontatlon lhould be deaiped to facllltate tbo.rocop1tlon; l~n, HP*Cment, ropalr, or *

adjuabnont of nialiianctlonm& compoaenta'or modulea.

.

. .

.

. . .

.

Tho monltodq lmtnunontatlon doil&n shQU!d mln1m1ze tho development of condlUom that would came meten, annu~claton, recorden, alarm.I, etc., to give anomaloua

, incUcatloni potentially conflWIII to tho operator. Human facton analyala lhould \\)e used In detenuin1q type and location of cUaplaya.

TABLE 1 (Continued)

Category 2

' Same u Cateaozy l Same u Cateaoiy 1 Same u Cate&01Y I

~~

.~ Categoiy 1

11. 011111111 Fac&ora Same u cateaoiy 1 *

Same 81 CalefO,Y I

Category3 Same aa Categozy I

Same_ u Cateaozy 1 Same u Catego,y 1

. ~o specific p~vialon Ii. Uunwa.Paclon Same aa Cate1011 I

Same u.Cate&OI)' 1

~ --

Category*,

11. (Contbntecl)

To the extent practtcable, the nme instruments 11houtd be used ror accident monitoring a.s are used for the normal operations or the plant to enable the operaton to uso. during eccfdent sftuatfon

s. lnttmments with

~

they an1 most f'amfllar.

12. Direet_Me11111ement To the extent pnctlcable, monitoring mstmmentatton Inputs 1h01tld be f'mm sensors that ~

meaure the desired fflllbtes. Ari Indirect rqeaarement should be made only when It can buhowni,y anaJJm to ' .

. PIO'f.lde UMmblgaon infomtltlon. .

TABLE 1 (Continued)

Cltegory2 Same as Category I

12. Dlftet Menmement Same II Category I

....

,.

'

.,

.. * -*.

.

~

- -

. ;,

.....

Cmvor,3 Same as Category I

...

.11 Dlftet Meamrement Same as Cat~ I

,. ..

. ..

.. '

b. ForTypeE

(l) 1be planned paths for eff111ent relcasei

. (2) Plant uea1 and inside buildings* *where access is required to. service equip1nent necessary to mitigate the

~uen'7' of an accident;.

(3) Onslte locations where unplanned releases of zadioactive materials should be dctcctedi and

(4) lbo variables that shoulcl be *moll4ored in each

, location identified in (1), (2), and (3) above.

2.4 The dctcnninatfon of performance requirements for qstem operation monitming and effluent release monitorina Information display channela shoulcS include, u a minimum, identification of:

L 1he range of the process Tarlable.

b. 1he required accuracy of measurement.

e. 1be requirecS response characteristics.

d. The time interval dudn1 wbfch the measurement is needed.

e. The local cuviionmcntl m whfc:h the information dlspJay clwmd components must opente.

f. Any requirement for rate or trcncS information.

g. Any requirements to group displays of related information.

h. Any required spatial distribution of senson.

2.4 The *design and qualification criteria for system

. . operation monitoriq and effluent release monitorina instrumentation should be taken from the criteria provided in regulatory position 1.4 of this guide. Tabies 2 and 3 of this regulatory guide should be considered as the minimum number of instruments and their respective ranges for systems operatio~ monitoring (Type D) and effluent release monitoring (Type E) instrumentation for each nudear

. power plant.

D. APPLICABILITY

11m -revision in combination with §S0.49 of 10 CFR

Part 50 provides acceptable guidance for design of new p~tl and for plant .redesign in response to nu-2 Action Pim (NUREG-0737) and its subsequent c:1arifications and generic letten.6

.

. ~7,

"ClriftcadoJl otTMIAcdoD Pim Recru1rementa.*

November 1!180,~obtamecl f.rom.th, NRC/Gl'OSalaPrognm.

U.S. Nadell K

Cormnflllon, WabhiltOn, D.C. 20555.

Sapp!emmd 1 (Game No. a2-33) 11 anilablll ror lmDection m ~.!or* feHttheNRC PmUc Document Rorm, 171711:strnt, NW.,WaJ

pon.D.C.

l.'7-12*

-..:.

I

'j I

TABLE2 BWR VARIABLES

~ .

TYPE A Variables: those variables to be monitcmd that proffl{e the primary information required to permit the control room opentor 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 infonna*

tion that is essential for the direct accomplishment of the specified safety functions; it does not include those nnables that am associated with contingency actions that may also be identified in wrlttcn procedures.

A variable included as Type A does not pRCJude it from being included as Type B. C. D, or E or 'rice nisa.

Variable Plant specific .

Range Plant specific Category (see R.egu)atory Position 1.4 andTableJ)

I

Information requm:d for open.tor action TYPE B Variables: those TUiabics that prov.ide information to indicate whether plant safety functions are being accomplished.

Plant wty flmctions are (l) reactmty control. (2) core CQOling. (3) maintaining reactor c:oolanhystem integrity, and (4)

~taming containment iJitegrity (iqcluding radioactive effluent control). Variables are listed with designated :ranges*and category f'or design and qualification iequimnents. Key variables are indicated by design and qualification Categoxy I.

Reactivity Control Neutron Flux

. Control Rod Position RCS Soluble Boron Conc:en*

tration (Grab Sample)

Core Cooling Coolant Level in Reactor Vessel BWR Core Temperatwel*2 llaintainio,g Reactor Coolant System Integrity RCS Pressure2

10'6~ to 100% run power (SRM,APRM)

Full in or not full in

. 0 to 1000 ppm Bottom of c:oie support plate to lesser of top of vessel or center- line of maia steam line.

200°F to 2300°F

,.!

OtolSOO~.

0 to design l)l'eSSUJe3 (psfg)

.I

3

3 I

1

1 Function *detecti~; accomplishment of mitigation Vaification Verlfication Function d~cction; accompli,,hmcnt of mitigation; long-term sum:illance To provide dhenc indication of water level

. Function detection, accomplishment I

of mitigation; ftrlfication Function detection; accomplishment of mitigation; Yelification

l'rcnblon still bciQg comld=d. mbJect to further dnelopmcm.

*

* .

*

j If a ~le b ~

tJr IIIDR 1111D one pmpoae, the lmtnmematlma requirements may lie ln.te,rated md Cllly one meamemeDt Jlr(M4ed.

3

.

mess. Desfp .~

b 11m nlue correspoadiq to ASME code ftluea that an obtained at or below code--eDowable Yatnea for material~ .

1.97-13

TABLE 2 (Continued)

Variable TYPE B (Contunied)

I DlyweU Sump Leve12 Maintamina Containment Integrity.

Top to Bottom Primuy Containment Pressuie2

-5 psig to design prmure3 Primary Cogtainment Isola- Cosed-not closed tion Valve Position (exclud- ing check valves)

Cate,o11 (see.

R.cpbtory Position l.4 mclTable l)

I

1 Function detection; accomplishment of mitigation; verification Function detection; accomplishment of mitigation; verification Accomptislunent of lSOlatio.n TYlE C Variables: those variables that provide information to indi~te the potential forbeina breached or the actual breach of the barri.m to fission product releases. The barriers are (I) fuel cladding. (2)*primary coolant premzro boundary,and (3)con- tainrnent.

Fuel Cladding Radioactmty Concentration or Radiation Level in Orcalating Primary Coolant Analysis of Primary Coolant (Gamma Spectrum.)

BWR Core Temperatun1 .2 Reactor Coolant Pressure Bounduy I RCS Pressure2 Primary Containment Aiea Radiation2

1/2 Tech Spec limit to 100 times Tech Spec limit

10 µCl/ml to 10 Cl/ml or TID-14844 source term in coolant volume

200°F to 2300°P

0 to 1500 (psig)

I

. 1'

Detection or breach Detail analysis; accomplishment of mitigation; vaific:ation; lon~term suneillance To provide divme indication of

~terlevcl Detection of potential for or actual breach; accompJisbrnent of mitiga- tion; loq,te:rm smveillan.ce

.

Detection of breach;_ verlficatlon

4SamcDn or* momtoma of ndlolctln Hqulds and IUCI llhould lie performed la a maimer thd easmea s,roc:umaad of~

a F"or ~

the c:riteda of ANSI Nl3.l*l96,, "Gulde to Sampliri1 AJltlom9 RadioacUft MatedaJI la lfudear F~ ahouJ4 be Fm llqllldlo j!rOvblom lhould be made for amplln1_ t'tom ffl:11.mbted turhlcsd zona. ad sampllns liDea l!bould be daiped to plateout or aepositicm. For ute and COSSftlllnt rmlllfDro the pnmdcma lllould mc:mde:

.

a. S11feldln1 co lllllntala n&tkia dosea AI.ARA.

a,. Samele i:oJltalnen witJa C011.tain.er-amplm1 port c:omiector compatibmty, c:. Cap&!!~ of samplinl ~~~

aaclneptm preaura.

d. Hmcllln1md~_cap1

,aiad

.

e. Premanpmeld for anal)'llll md

5The mutmmn nJue ma, be rewed upward to utlafJ ATWS zequlremam.

'1.ommum of two IIIOllffon.at wfddJ' aepanted loc:adom.

I

1Detec:ton !lbould retDODd to pmma ndfab photom wW1ba ay *eaau mip flom 60 *uv 10 3 MeV wltla a dose nq rapoma ac:curaq withlll a factor of~ onr th* entire nnp.

.

1.97-14 I

.,

TABLE 2 (Continuedl Vanable

'.l'Yf'E C (Continued)

Reactor Coolant Pressure.

Boundary (Continued)

Drywell Drain Sumps Lcve12 (Identified and Unidentified Leakage)

Suppression. Pool Water LeYe1 Containmqt RCS Pressurc2

~-*

_Top to Bottom Bottom of ECCS suction line to 5 ft abon: normal water level o to design p~s (psig)

0 to 1500 (psig)

Primary Containment PRssure2 * *S psig ~

to 3 times design pressme1 for concrete; 4 times design p:cssun for steel Containment and Drywell Hydrogen Concentration C:ontainment and Dtywell Oxygen Concentration (for inerted containment plants)

0 to 30 \\'Ol~ (capability of o~

from *5 psig to design

~)

o to 10 \\'Ol~ (capability or open~ from *5 psig to design pres.we ) .

Containment Effluent2 Radio-

10°" p;Cl/ec to 10.,.. flCi/ec actiYity - Noble Gases (from

identified Rlease points includ:.

ing Standby Gas Tnatment System Vent)

1.97-15 *

Catqory (see

.Replatory Position 1.4 mdTable I)

l l

l .

l l

I*

,

J>eUctionof breacb;accomplishment *f of mitigation; verification;long-tenn

~cc

. .

Detection of breacb;accomplishment

of mitigation; Terification; long-term IUffCillance Detection of breach; Yerification Detection of potential for breach;

accomplishment or mitigation Detection of potential for or act11al bleach; accomplishment of mitiga- tion Detec:tion of potential for breach; .

accomplishment of mitigation Detection of potential for breach;

accomplishment of mfflption Detection of actual biach; accom- plishment of mitigation; ffrifica*

tion

-::,:."

1 I

TABLE 2 (Continued)

Variable TYPE C (Continued).

Containment (Continued)

. Effluent Radioactivity2 - Noble

10-6 µCi/cc to 103 µCi/cc Gases (from buildings or areas

- where penetrations and hatches an located, e.g.. seconcluy con- tainment and auxiliary buildings and fuel handling buildinp that ue in direct contact with primary contabunent)

Cateaoay (see Regulatory Position 1.4 and Table I)

Indication of breach TYPE D Variables: *those variables that pro-dde information to indicate the operation of individual safety systems and other *

systems important to safety. These variables are to help the operator make appropriate decisions in Wlin1 the individual sys- tems important to safety in mitigating the consequences of an accident.

Condensate and Fccdwatcr System Main Feedwater Flow

0 to 11 K design OowlO

3 Detection of operation; analysis of cooling.

Condensate Storage Tank Level Top to Bottom

3 Indication of available water for COOiin&

Primary Containment-Ilclatecl Systems Suppression Chamber Spray

0 to l 1~ design flow10

2 To monitor operation Flow Drywell Pres.mre2

-5 psig to 3 psig (narrow

2 To monitor operation ruge)andOto 110~

'design pressure3 (wide range) .

Suppression Pool Water Level .

Top of vent to top of weir well

2.

To monitor operation *

Suppnssion Pool Water

40°F to 230°F

2 To monitor operation Temperature Dryweil Atmosphere

40°f to 440°F

2 To m~or operation Temperature Drywell Spray Flow o to l lK design flow10

2 To monitor op

e. ration

. Main Steam, System Main Steamline Isolation O to l S" or water (nanow

2 To pio-dde indication *of pressure Valves" Leakage Control range) anil O to 5 psid boundary maintenance ..

System Pressure (wide .range) .

10nes1p now la tile awdmum flow anticipated Ill normal operation.

1.97-16 I

I

.I

TAB~E 2 (Continued)

Variable TYPE D (Continued)

Main Steam System {Continued)

Primary System Safety Relief Closed-not closed or O to SO psig Valve Positions, Including ADS

or Flow Through or Pressure in Valve Lines Safety Systems Isolation Condenser System Shell-Side Water Level Isolation Condenser System Valve Position.

RCJC Flow HPCI Flow Core Spray System Flow LPCJ System Flow SLCSFlow SLCS Sto~ge Tank Level Residual Heat Remon! {RBR)

Systems RHR System Flow RHR Heat Exchanger Outlet Temperature

. Cooling Water System Top to bottom Open or closed .

0 to 11096 design flow10

. o to 110%.design now10 .

0 to l l<>fo design flow10

0 to 110% desjgn flow10

0 to 11 OJ, design flow"10

Top to Bottom

0 to 1*10% design flowlO

40°F .to 350°F

Cooling Water Temperature to

40°F to 200°F

ESF System Components Cooling Water Flow to ESF

Oto 1109fi design flowlO

System Components Radwaste Systems ltigh Radioactivity liquid TanJc Top to bottom Level

Ventilation Systems Emergency Ventilation D~per . Open-closed status Position

1.97*17 Cate1ory (see Replat~ry Position 1.4 and Table I)

2

3

2 Purpose

. Detection of accident; bounduy integrity indication

To monitor operation To monitor status To monitor operation To monitor operation To monitor operation To monitor operation To monitor operation To monitor operation To monitor operation To monitor operation To monitor operation To monitor operation To monitor operation To monitor operation

Variable TYPE D (Continued)

Power Supplies TAILE 2 (Continued, Ranae Cateaory (see ReauJatory Position 1.4 and Table I)

Status of Standby Power and Plant specific Other Energy Sources Important to Safety (electnc, hydtaulic, pneumatic) (voltages, cUJrents, pressures)

Purpose To monitor 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.

ContainmentRadiauon I

Primary Containment Area Radiation - High llange2 Reactor Building or Second~

Containment Area Radiation Area Radiation l R/hr to 107 R/hr

10*1 R./hr to 104 R./hr for Mark I

and II containments

1 R/hr to 10 7 R/hr for Mark III

containment Radiation Exposure Rate

10*1 RJhr to 1(~4 Rfhr (inside buildings or areas where access is required to service equipment important to safety)

Alrbome RadJoactire Materials Released from Plant Noble Gases and Vent Flow Rate

Drywell Purge, Standby Gas

' Treatment System Purge (for Mark I and n plants)

and Secondary Contain- ment Purge (for Mark W

plants)

- Secondary Containment Purge (for Mark I, II, and Ill plants)

Secondary Containment (reactor shield buildiq annulus, if in design)

10"" µCi/cc to 105 µCi/cc

0 to 110% vent design flow10

(Not needed if effluent dischargt"

through common plant vent)

10"' µCi/cc to 104 µCl/cc

0 to l l 0% vent design flow10

(Not needed if effluent discharges through common plant vent)

10*6 µCi/cc to 104 p.Ci/ce

0 to 110% vent design Oow.1 o (Not nc:eded if effluent discharges through common plant vent)

2'

Detection of significant releases;

release assessmenti long-term sunreillance; emergency plan actuation Detection of significant releases;

release assessment; long.term surveillance Detection of significant releases;

release assessment Detection of significant releases;

release assessment

0 status !ndlcaffo11 of all standby pow. a.c. busei, d.c:. buses, bwerter output b\\lSCI, and pneumatic suppJlel.

1.97-18

,. ,.

i'

i t ;

TABLE 2 (Continued)

Variable TYPE E (Continued)

Auxiliary Building (including any building containing primary system gases, e.g .* waste gas decay tank)

Range

,_

10-6 µCi/cc to 103 p.Ci/cc

0 to 110% vent design now10

(Not needed if efOuent discharges through common plant vent)

Common Plant Vent or Multi- 10"' p.Ci/cc to 103 p.CiJcc purpose Vent. Discharging Oto 110% vent design fiow10

Any or Above Releases (if'

drywell or SGTS purge is included)

10-6 p.Ci/cc to 104 µCi/cc Airbome Radioactive Materials Released from Plant (Continued)

Noble Gases and Vent Flow Rate (Continued)

AD Other Identified Release Points Particulates and Halogens ro"' p.Ci/cc to 102 µCi/cc

0 to 110% vent design flowlO

(Not needed if effluent discharges through other monitored plant vents)

All Identified Plant Release

10*3 p.Ci/cc to 1 o2 µCi/cc .

Points. Sampling with Onsite

~ to 110% vent design flow10

Analysis Capability Environs Radiation and R.acli~

activity13

Airborne Radiohaloge~ and

10-9 p.Ci/cc 1.0 10*3 p.Ci/cc Particulates (portable sampling with onsi.te analysis capability)

Cateaory (see Regulatory Position 1.4 and Table I)

Detection of significant releases;

release assessment; long-term surveillance Detection of significant ~eases;

i'elease assessment; long-term *

siln'.eillance Detection of significant releases;

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

release assessment; long-term surveillance

. Release assenment; analysis

.

12To provide information re,ardinr;nlease ~fndloactive halogens and particulates. Continuous collection of~tative amples followed by onsite laboratory measunments of 11111ples for radiobalor;ens and ~a.

Tile design envelope for lbieldiar;. bm~, and ana!ytical prposes *lhould ISSUme 30 minutes or Int~ amllli=. time 1t 1ampler design Gow. an anrqe concentration of 102 p.a,a: of ndioiodines in pseous or vapor form, ID ~e concentration of IP If Cl/cc of~

ndloiodlacs IDd pt!tieullltes other thm ndiofodlnes ad aa avenae pmma photon eDerJY of 0.5 MeV per disiatgratfon. For the~ of this Item oaJy, .. collectlon of ~tin aamp1esl* means obtaimng the best aamples practicable a;hen the exfgendeli that attend the acc:ldeat emironmeat; line lolses or line deposition lhould be empirically predetermined and appropriate loss COlftction facton lhould be applied.

.

1311

ts ualltely that a few fixed4tadon area monitors could ~e IUfficleDtly reliable lafonnatbl to be of use* la detecting releases

&om uamonltored containment release 1191ats. However, then, may be circumstances ID wblc:h 111ch a system of moaiton may be useful, The decision to Jastall sucb a 111tem Is Jeft to the licensee.

.

14For atimatinr; release ntes oindloactlve materials nlcased. duzini an accident.

.

1.97-19

Variable TYPE E (Continued)

Environs Radiation and Radio- a.ctmty13 (Continued)

Plant and Envkons Radiation

. (portable instrumentation)

Plant and Environs Radio- activity (portable instni*

mentation)

Meteorologyl 7 Wind Direction Wind Speed Estimation of Atmos- pheric Stability TABLE 2 (Continued)

Category (see Regulatory

Position 1.4 and Table l)

10*3 R/hr to 104 )Yhr, photons

10*3 rads/hr to 104 rads/hr, beta radiations and low-energy photons (Isotopic Analysis)

0 to 360° (+S

0 accUracy with a

-

0

deflection of 10 ). Starting speed less than 0.4 mps (1.0 mph).

Damping ratio greater than or equal to 0.4, delay distan~e less than or equal to 2 meters.

0 to 22 mps (SO mph). !().2 mps (O.S mph) accuracy for speeds less 1han 2 mps (S mph), 10% for speeds in excess of 2 mps (S mph),

with a starting threshold of less than 0.4 mps (1.0 mph) and a distance constant not to exceed

2 meters.

Based on vertical temperature difference from primary meteorological system, -S°C

to 10°C (-9°F to 18°F) and i0.1S°C accuracy per SO-meter

-

0

intervals (:+/-().3 F accuracy per

164-foot intervals) or analogous range for alternative stability estimates

3

3 Purpose

. Release assessment; analysis Release assessment; analysis.

Release assessment Release assessment Release assessment

15To*monltor ~oa ~d alrbome radloadhity concentrations In many areas throupout the facility and the site emizonawbere it ii lmpractkal to imtaD statlonar:, monfton capable of covedna both nonnal and ai:cldent ~

16 A ~ble multichannel pmma ray ~metet would Pl'O\\'lde the earliest ~~

fo, scopbl1 the ndlonuc:llde content of the source (see R. C, ~L

D. E. Jones, and G. W, Huclcall*Y..1 "lnsfrumentatlon for Off-site Reactor Plume Studies," In lnttnu,donol Sy,npo- man 011 Envlronmen'ltll M'onltorlll6. IEEE Catalogue No. '75o,\\;H 1004-1 ICESA. lmtltute of Elecuical and Electronics Engineers; 345 East 47th Street, New York, New York JOOl't, 1976).

.

. *

17Gwdance on ~eteorolodc:al measurements In the context of emergency ffSJ>O!I? ia provided ill Replatory Guid~ 1.101, "~ergency

~

and Preparedness for Nuclear Pow.=r Reacton." Guidance on meteorological instrumentation la contained In Replatory Guide 1.23,

"Omit* MeteorolopcaJ Propams." A proposed revision to tlua guide hu been issueil fo, comment u Task ss 926-4.

l.97-20

Variable

. TYPE E (Continued)

A~ident SampUng18 Capa- bility (Analysis Capabil*

ity On Site)

Primary Coolant and Sump

Gross Actmty

Glunma Spectnun

Boron Content

Chloride Content

Dissolved Hycltogen or Tota1Gas20

Dissolved 0xygen20

pH.

ConW;ll!Dent Air

Hydrogen Content

Oxygen Content

Gamma Spectrum TABLE 2 (Continued)

Grab Sample

1 p.G/ml to 10 G/rni (Isotopic Analysis)

0 to 1000 ppm

*

Oto 20ppm

0 to 2000 cc(STP)/]cg Oto 20ppm I to 13 Grab Sample

0 to 10 vol-9&

Category (see Regulato"

Position l.4 and Table l)

0 to 30 voi-,(, for inerted containments Oto30v~

(Isotopic analysis)_

Release assessment; verification;

analysis*

Release assessment; verification;

analysis

I

18.ne time for takm& and UW)'ZUII aamDIIIS lhould be 3 lloun or less from the time the decision ts made to ampl~, except for c:llloride, .

wblc:h mould be witbla. 24 llours on aea or brac:lcish water lites. Plants on flab water lites should perfozm analysis wltlml 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months .

19 An lnstal1etl capabDity lhould be prcmdcd for obtamlns containment 1UJDP, ECCS pmnp room IWD.PI* and other limDllr auziliary lndlding IUIDP liquid samples.

20wittain the &st 30 daYI after an ac:c:ldent, ox~sea. anal~lls need noi be performed until dllmide analya!s la.~ a clllodde coa.c:entra* : * *

tion creater aum 0.15 ppm. Once the chloride eonc:entration exceeds this ftlue, oxnen lhould be determmed witbla. 3 lloms. For this 30-day Nriocl. It ts ac;c,eptable to wrifJ that Umomd o.r.,aen Is less tlam 0.1 ppm If the measured dlssohed llydrotzen nsl.41111 ls 10 cc/kc or leas.

HoweYCr, consistent with mfnhritzms personnel ndiation. espcrmrn (AI.ARA), direct monitoring for dissolved"' oxypa. ls ncommeilded. Tbis .

applies oaly to primarJ' coolant, not to IUID.p.

TABLE3 PWR VARIABLES

TYPE A Variables: those variables to be monitoted 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 inform.a*

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 vem.

Variable Plant specific Ranae Plant specific Cateaory {see Regulatory .

Position 1.4 and Table l)

1 Purpose Information required for operator action TYPE B Variables: those variables that provide information to indicate whether plant safety functions are beJng accomplished.

Plant safety functions are (1) reactivity control, (2) core cooling, (3) maintaining reactor coolant system integrity. and (4)

maintaining containment integrity (including radioactive efftuent control). Variables are Usted with *designated ranges and category for design and qualification requirements. I(ey variables are indicated by design. and _!1-Wllification Category I.

Reactivicy Control Neutron Flux Control Rod Position RCS Soluble Boron Concen*

tration RCS Cold Leg Water Temper- ature1 Cor.e Coolins RCS Hot Leg Water Temper- ature, RCS Cold Leg Water Temper- ature1

RCS Pressure1 J Core Exit Temperature1

1 o-6% to I 00% full power Full in or not full in Oto 6000ppm S0°F to 400°F

S0°F to 700°F

Oto 3000 psic (4000 psig for CE plants)

200°F to 2300°F

I

3

3 I

1 Function detection; accomplishment of mitigation Ve:rification Verification Verification Function detection; accomplishment of mitigation; verification; long-term surveillance Function detection; accomplishment of mitigation; verification; long-term surveillan=

Function detection; accomplishment of mitigatiOD; verification; long-term surveillance Verification

1When a Yariable i. lilted for mote than one purpose, the Instrumentation reqwremema may be Integrated and on11 one measurement pro-nded.

2ne maximum wlue ma1 be reviled upward to satisfy ATWS requirements.

I

3i:n.tnamentatioll that la a part of the final ICC detection system should meet the dcsfp requirements ~d m Item ILF .1 of NUREG,,073'7. (When Typo K thermocouples become part of the system, they are coasiderod to meet the requirements. H-. the remainder of the detection aystem that II outside the reactor~ should meet the iequfrements specified.) *

1.97-22 r ;.

!-

\\.

I

-TABLE 3 (Continued)

Varial,Je TYPE B (Continued)

Core Cooling (Continued)

Coolant Jnventory Degrees of Subeoolmg lfamtaimng Reactor Coolant System Integrity RCS~1 Containment Sump Water Lnell Containment .l'rcsmrc1 M*iDlliDing Contammen.t lntepity Bottom of hot leg to top of"RSSCl4

200°F subc:oolmg to

35°F superheat

0 to 3000.psig (4000 psig for CEplants).

Nanow range (sump).

Wide nnge (plant specific)

o to design pressme5 (psig)

Containment Isolation Valve Closed-not dosed Position (excluding check 'ft!va)

Containment Piasme1

- S psig to design prcssme5

..

Catesory (see Replatory Position L4 and Table I)

1

2'

(With con- firmatory operator procechues)

2

1 l

1 I

Pmpose Verification; accomplishment of mitigation

Verification and analysis of plant c;onditions

. Function detection; accomplishment of mitigation Function detection; accomplishment I

of mitigation; Yerification Function detection; accomplishment ofmit:igation;vaification

Accomplishment of isolation Function *detection; accomplishment I

of mitigation; verification .

TYPE C Variallles: those wmiables that pmvide information to indicate the potential for being brcachcd or the actual breach of the bame:rs to fimon product nleases. 1be barriers are (1) fuel cladding, (2) primary coolant PR&SUre bounduy, and

(3) contmnm.ent.

Fuel Claddins

.

Coze Exit Temperature1

. ! .

200°F to 2300°F

Detection of potential for breach;

acc:omplishment of mitigation; long*

term SUffdllancc

4 A measmemeat to detect the tmad of 'WOids ID the* nactor coolmt J)'ltem with aaetor-~t ~ps rummii lhoald also be p10videci for aD ~

For B&W .reactors.

meamanent lhcndd lie pnmded to detea wfds In the hot kl candy cane wllen seactor coolaJii pumps aeaot~

5Daiga pnl!AUll ls Ulllt ftlm coaapoudhl1 to ASME code 'fthles llaat an ob~ at or below code~ble ft1ues for materiaJ des:ip ma,.

.

. .

.

I

i Variable TYPE C (Continued)

Fuel Cladclins (Continued)

Radioactivity Concentntion or Radiation Level in Circulating Primary Coolant Analysis of Primary Coolant (Gamma Spectrum)

Reactor Coolant* Presmre Boundary RCS Pressure1 ContainmentPrcssurc1 Containment Sump Water Leve11 Containment Area Radiation1 Effluent_ Radioactivity - Noble Gas Effluent from Condenser*

Air Removal System Exhaust1 Containment RCS Pressure1 TABLE 3 (Contlnuedl

1/2 Tech Spec limit to 100 times Tech Spec limit

10 µ.Q/ml to 10 Q/ml or TID-14844 source term in coolant volume

0 to 3000 psig (4000 psig for CE.

plants)

-S psig to design pressure4

(-10 psig for su"atmospheric containments)

Narrow range top to botto.m (sump), wide range (plant specific)

1 R/hr to 104 R/hr

10*6 p.Q/cc to 10*2 p.Q/cc

0 to 3000 psig ( 4000 psig for CE plants)

Category (see Regulatory *

Position 1.4 and Table 1)

1

2

1 Detection of breach Detail analysis; accomplishment of mitigation; verification; long-term SUJVeillance Detection _~r potential for or actual breach; accomplishment of mitiga- tion; long-term surveillance*

Detection of breach; accomplishment

of mitigation; verification; long-term surveillance Detection of breach; accomplishment of mitigation; verification; long-term surveillance Detection of breach; verification Detection of breach; verification

. Detection of potential for breach;

accomplishment of mitigation

6Sampllns or momtorill.s of radioactive Bqufda anll sasea should t,e* performed Ill a manner that eamrea procuromem of ~tatlft samDlea. Far~ the criteria of ANSI NU.1*1969, "Guide to Samp_llns Alrbome Radioactive Matc:daJI Ill Nuclear Facilitlel. shouJd be ap;,llecL For Uq11f41, ~Ill should be .made ror sampl#ls from well-mlsed turbulent zones, and samplfq Hnea should be desipecl to mlni- mzze plateo'llt o, deposition. For sare and convement sampling, 1he pn,risloZll should.f:Ddud,: :

.

a. Shfel4fq to :maintain radiation closca AJ.AAA.

.

. b, Samt:lle contamers with contamer-amplina port connector compatibility, c. Capal>iJlt)' or sampllns und5 system prcssun im,d 11.eptne pressures, cL Handliq and. tnmspart_ cap1

ailcl e. l"rearnn,emem for aDa1JSi1 an.cl terpntatloD.

7Mlnlmum of two momton 1d widely aparatecl locatiom.

'Detecion shouJd rm,mic1 to sam;na radiation photona wfthfD IDJ eneqy ranp from 60

bV to 3 MeV with a dose rate respC>llSIII *

accmaq wlthfll a factor of 2 Oftl thl'i entire ranp.

.

9Monlton lbould be capable of ~

and meamrim: iaseom effluent raclloactMty with composltiQm randns from fresh ecauilihrimn noble ~

fission ~uc:c* mlsturCI to lo-day-old mixtures. with overall antcm acemadcl wftJliD a factor or z. Efiluint raclloa~ may be apzcsiecl ID tcr1111 of conccntradom or Xe-133 equmdalta, Ill tams or coaccntraticml of 1111r 11obl11 pa nudfdcl. ow ID tenm of mte~ted aamma MeV per unit tim

e. It JI not E

that a siade mon1todns de\\'lce will have sufficient nap to encom~ the entira ranp provlde4 ill thJa rcnlatory nlde and that multi e comp_oncnta or system will bl needed. Exlstlna equipment may be used to monitor ay portion or the statecl nnae wfth!n the equipment cslp rating.

.

I.Sl7-24

Variable*

TYPE C (Continued)

Containment (Con~ued)

Containment Hydrogen Concentration Containment Pressure1 Containment Effluent Radio- actiYity - Noble Gases from Identified ttelease Points~

.

Effluent RadioactiYity1 - Noble Gases (from buildings or areas .

where penctratimis and hatches are located, e.g., secondary con- tainment and auxiliary build*

ings and fuel handling: build- ings that are in dJrcct contact with primary containment)

TABLE 3 (Continued)

Category (see

Replatory Position 1.4 and Table 1)

O to 10 vol-% (capable of operating fro~.-5 psig to maximum design pressure5)

0 to 30 vol-% for ice- condenser-type containment

-s psig pressure to 3 times design pressure5 for concm:e;4 times design prcsmre for steel (-10 psig for subatmospheric containments)

10"" p.Ci/cc to 10-2 µCi/cc

10-6 p.Ci/cc to 103 p.Ci/cc I

1

29,lC

Detection of potential for breach;

accomplishment of mitigation;

long-term IUl'ftillance

..

Detection of potential for or actual

breach; accomplishment of mitiga- tion Detection of breach; accomplish- ment of-mitigation; Yerlfication Indication of breach TYPE D Variables: those variables that provide information to indicate *the operation of indbidual safety systems and other systems Important to safety. These variables are to help the operator make appropriate d~cisions in using the individual sys- tems important to safety in mitigating.the consequences of an accident.

Residual Beat Removal (RHR) *

or Decay Heat Removal System RHR System flow RHR Heat Exchanger Outlet Temperature Safety Injection ~)'Stems Accumulat<< Tank Level and Pressuie Accumulato

r. Isolation Valve

Position

0 to 110% design flow11

.;

40°F to 350°F

10% to ,0%,-wolwne

0 to 7SO psig .

.

Closed or Open

2

2 To monitor operation To monitor operation and for analysis I

To monitor operation

()peration status JOPrcmslonl mould 1,e made to monitor Ill identified Dathwa:,s for nleue of m radioa~ materia!s to *'le CJmrODS In conformance with General ~

Criterion 64. M~ of JncUvlduaf effluent IUUms b o~mrecl wbere IIUCh mcams ue nleued directly .into tile enYlronment. If two or mo:re ltrams ue combined prior to release from a common iUldwp point. monitorini of Ce combined lbeam b comldcred to meet the Intent of this nplatory sulde jrcmded auch momioring bis a nn&e adequate to meume wont-cue releases.

UI>algn flow Is the mazlmum flow antic:ipated In normal operation.

1.97-25

TABLE 3 (Continued)

Catesory (see Reaulatory Position 1.4 Variable.

Ranae and Table 1)

_Pmposif *

TYPE D (Continued)

Safety hQection System, (Continued)

Boric Acid Charging Flow

0 to 110% design tlow11 .

2 To monitor operation Flow JD HPI System

0 to 110% deslp flow11

2 To monitor operation Flow in LPI System

0 to 110% design flow1l

2 To monitor operation R.efuellq Water Stonge Tanlc:

Top to bottom

2 To monitor operation Lem Primuy Coolant Syst?l Reactor Coolant Pump Status Motor current

3 To monitor operation Primary $)'stem Safety Relief Oosed-not closed

2 Operation status; to ~onitor for Valve Positions (including loss of coolant *

POR.V and code valves) or Flow Through or Pressure in Relief Valve Unes Pressurizer LeYe1 Top to bottom I

To ensure pmper operation of

. pressurizer Pressudzer Heater Status Electric cum:nt

2 To determine opentin1 status Quench TanJc Level Top to bottom

3 To monitor operation Quench Tank Temperature so°F to 7S0°F

3 To monitor operation

. Quench Tank Presswe o to desjgn pressun:5

3 To monitor operation Secondary System (Steam *

Generator)

Steam Generator Level From tube sheet to separaton I

To monitor operation Steam Generator Pressure From atmospheric pressure *

2

. To monitor operation to 209', above tho lo~ safety valve setting Safety/Relief Valve Positions Oosed-not closed

2 To monitor operation or Main Steam Flow Main Feedwater Flow

0 to llG,r, desip flow11

3 To monitor operation

1.97-26

l TABLE 3 (Conlinued)

Catqory (see Regulatory l'osition I."

Variable Range and Table I)

Purpose TYPE D (Continued)

Auxiliary f eedwater or Emer- gency Feedwater System Auxiliary or Emergency Feed-

0 to 110% design flow11

2, To monitor operation water Flow

(1 forB&W

plants)

..

Condensate Storage Tank Plant specific .

I

T~ ensure water supply for auxiliary Water Level fecdwater (Can be Category 3 if not primary source of AFW. Then what- ever is primary source of AFW should be listed and should be Category * .)

Containment Cooliq Systems Containment Spray Flow

0 to 110% design flow11

2 To monitor operation H~t Removal by the Contain- Plant specific

2 To monitor operation merit Fu Heat Removal System Containment Atmosphere

40°Fto 400°F

2 To indicate accr.cnplisbment d cooling Temperature Containment Sump Water S0°F to 2S0°F

2 To monitor operation Temperature Chem.icll and Volume Control System Makeup Flow

In

0 to 110% d* flow11

2 To monitor operation Letdown Flow* Out

0 to 110% design flowll

2 To monitor operation Volume Control Tank Level *

Top to bottomJ

2 To monitor operation Cooling Water System Component Cooling Water

40°F*to 200°F

2 To monitor operation I

Temperature to ESF System

.

Component C90ling Water Flow 0 to 110% design flowu

2 To monitor operation to ESF System Radwaste Systems High-Level Radioactive Liquid Top to bottom

3 To indicate ~orqc volume Tank Level Radioactive Gas Holdup Tank

0 to 150% design pressme5 Pressure

3 To indicate storage capacity

1.97-27

TABLE 3 (Contlnuedl Variable TYPE D (Continued)

Ventilation Systems Ranae

-

Emergency Ventilation Damper _Ope~osed status Position

. Power Supplies Status of St~dby Power and Other Enersy Sources Impor- tant to Safety (electric, hydraulic,pneuniatic)

(voltages, currents, pressures)

Plant specific Cateaory (see Jteplatory Position l.4 an~Table I)

2 Parpose To indicate damper status To indicate system~

TYPE E Varlablea: tho!!fl variables to be monitored u required for use in determininl the magnitude of the reiease or radio- active materials and continually assessing such releases.

Containment Radiation Containment Area Radiation *

1 Rfhr to 107 R/hr lfjgh Rangel I

Area Radiation Radiation Exposure Rate

10*1 R/hrto 104 R/hr (inside buildings or areas where access is required to senice equipment important to safety)

Airbome ltadioactin Matedab Released from Plant Noble Gases and Vent Flow Rate .

Containment or Purge Effluent1

Reactor Shield Building Annulus1 flf m design)

Auxiliary Buildin11 (including any builclin:I

containinl primary system gases, e.g., waste gas decay tank)

.

1 o-6 µCi.fee to 105 µCi/cc

0 to 110,, vent design .tlowll (not* needed if efftuent discharges through common plant vent)

10*6 p.Ci/cc to 104 µCi/cc

0 to 110,, vent design Oow11 (not needed if effluent discharges through common plant vent)

10*6 p.Ci/cc to 103 p.Q/cc

0 to 11~ vent design flowll (not needed if effluent discharges through common plant vent)

Detection of sfgnlficant releases;

release as.,essment; Iona-term surveillance; emergency plan actuation Detection of significant releases;

release ;messment; Iona-term sum:illance

.

Detection of siplficant releases;

release assessment Detection of* sipificant releases;

release assessment Detection of significant * :releases;

release assessment; long-term surveillance

12Stasu lndicadon of aD studbJ po~ a.c. buses, d.c. buselo Inverter output buses. 11114 pneumatic supp&a.

1.97-28 *

TABLE 3 (Continued)

Variable Type E (Continued)

Airborne Radioactive Mataials Released from Plant (Continued)

Noble Gases and Vent flow Rate (Continued)

Condenser Air Removal System Exhaust1

1 o-6 pO/cc to 105 µCi/cc

0 to 110% vent design flow11 (not needed if effluent discharges through common plant vent)

Common Plant Vent or Multi- 10-6 pCi/cc to 103 p.Ci/cc purpose Vent Dischirging Oto Jl0%vent design flow11

Any of Above Releases (if .

containment pwge is included)

J o-6 µCi/cc to 104 µCi/cc

Vent From Steam Gen- erator Safety Relief Valves or AtmoSPheric Dump Valves

All Other Identified Release Points Particulates and Halogens

All Identified Plant Release Points (except steam gen- erator safety relief valves or atmospheric steam dump valves and condenser air removal system exhaust).

Sampling witb. Onsite Analysis Capability*

10*1 pO/ccto 103 pCi/cc (Duration of releases fn seconds and mass of steam per unit time)

10-6 p.Ci/cc to 102 µCi/cc

0 to 110% vent design flow11 (Not needed if effluent discharges through other monitored plant vents)

10*3 µCi/cc to 102 µ.Ci/ci::

0 to 110% vent design flowll

1.97-29 Category (see R.egu!atory Position 1.,

IDd Table 1)

Purpose

.. Detection of significant releases;

'."release assessment Detection of significant Jeleases;

ietease assessment; long-tenn *

surveillance Detection of mgnificant releases;

release assessment Detection of significant releases;

release assessment; long-tam suneillanc:e Dciection of sipuficmt releases;

release asseament; long-term mrRillance

Variable TYPE E (Continued)*

Environs Radiation and Radio- activi..,.1 s

'

z .

Airborne Radiohalogens and Particulates (portable sampling with onsite analysis capability)

Plant and Environs Radiation (portable instrumentation)

Plant and Environs Radio- activity (portable instru*

mentation)

Meteorology19 Wind Direction Wincl Speed Estimation of Atmos- pheric Stability TABLE 3 CContinued)

Catel()ry (see*

Regnlatory Position 1.4 and Table 1)

10-t IJQ/cc to 10*3 l,lC.i/cc

10*3 R/hr to 104 .R/hr, photons

10*3 nds/hr to 104 rads/hr, beta radiations and low-energy photons (Isotopic Analysis)

0 to 360° (+/-S

0 accuracy with a deflection of 10°). Startin& speed less than 0.4 mps (1.0 mph);

~ping ratio greater than or equal to 0.4, delay distance less than or equal to 2 meters.

0 to 22 mps (SO mph). +/-().2 mps .

(O.S mph) accuracy for speeds less than 2 mps (S mps), I~ for speeds in excess of 2 mps (S mph),

with a starting threshold of less than 0.4 mps (1.0 mph) and a distance constant not to exceed

2meters.

Basedonverticaltcmperatun difference from primary mete- orological system *S° C to

10°C (-9°F to 18iF) and +/-().IS°C

accuracy per SO-meter intervals

(!0.3°F accuracy per 164-foot intervals) or analoaous range *

for alternative stability estimates

3

3*

3 Purpose Release assessment; analysis Release assessment; analysis Release assessment; analysis Rei~ assessment

.Release assessment Release assessment lSlt !I miJibly thzlt a few flxed-mdon area monlton could ~

~

reliable fD.tormatlml to bl of U111 fn detectfnJ releases from unmonilored contmnmeDt ~le POfntl. Howner, there may be c:fn:mmtmlcea Ill which IIZda a S)'ltem of monltozs may lie useful.

Th, dedllcm to instllB SGcla

system.. ft to the llcemee.

.

16For elttrnatfq ~

nta of nd!oaetha materlaJs releuell llurins a acddem.

17To monitor ndJatloD ad airbom1 radloactirity conc:cnimtfom fn ~

area thro!lihout the faciHtJ mll the stt. emfrom when ft fl lmpractieal to lmtall natlolUIQ' monitora capable of coi'erln1 both normal mll acddent ~ *

m A ~ta ~altlc:hmmel~~ IPCCtlometerwouM provide th* earllelt ~ability for scoE!h.!1 ~

ndionuclldl contem of the IOUIC9

.t:m,K.J;e::s.t~~~~~,:~n~1i1:t'rra~*~-,.~~="~J:e~~~~

Nnr York. New York JOlb 7, 1976).

I *

DGuldanc:a ~

meteorolo!dcal measurementa In die contut of emeqenc, ~~

fl prcmdell fn *Reaulator7 °Guldl I.IGl, "Em~q Plamaln1_ amt Preparedness for Nuclear Power Reactors." Guldmc:a on meteorolo~ fmtrimentatlml II c:ontafned In Re,u!atory Guidi 1.23.

"'Omite Meteorological l"J'Ognma." A pioposed revision to thfl sulde hlll been ls.med for comment u Task SS 92M.

1.97-30.

TABLE 3 (Canlinncl)

-*

Category (aee Regulatory POlition 1.4 and Table I)

Variable TYPE E (Continued)

Accident Sampling20 Capa~

bility (Amlysis Capabil- Pmpose

fty On Site) *

..

Primary Coolant and Sump

Gross Actmty

Gamma Spectrum

Boron Content

Cblorlde Content

DJssolved Hydrogen or Total aas21 *

Dissolved 0xygm22

pH

Containment Air

Hydlogm Content

Oxygen Content

.

Gamma Spectrum Grab Sample

1 pa/ml to 10 Ci/ml

. (Isotopic Analysis)

Oto6000ppm Oto 20ppin

0 to 2000 cc(STP)Jkg Oto 20ppm

1 to 13 Grab Sample Oto IOvol-%

0 to 30 yc,i-,, for ice condemm

.0 to 30 vol-%

(Isotopic analysis) *

R.eleuc assessment;Terlfication;

analysis

'*

~ .

Release assessment; verification;

analysis

20ne time for tatmr; and aminms 111mples shoa.ld be 3 hams or Jesa f:rom the time the dedliau is made to ample, except for d&loride, wldch should be witbfD 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for plail.1s tbat use au or brackish water in eaential II.eat czcbaDpn (Le., shutdown coolini) that haft only a ldngle burier from the react~ c:oo!mit. Other plants hne 96 homs to ~onn a dl1od4e 1D11Ja1s. .

.

21 An imtaDed capability lbovld be pnmded far obtammg containment sump, ECCS pump room sumps. ad other llmDllr awdllm)'

building lmDp liquid amples.

.

..!

. ,

I

REGULATORY ANALYSIS

1. STATEMENT OF THE PROBLEM

The applicant for a license (or licensee) of a nuclear power plant fs required by the Commission's regulations to provide instrumentation to (1) monitor variables and systems over their* anticipated ranges for accident con- ditions as appropriate to ensure adequate safety and (2)

monitor the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluid, effluent discharge paths, and the plant env.irons for radioactivity that may be released from postu- lated accidents. This revision to Regulatory Guide 1.97 proposes to modify and update the guidance previously given. The modification is based on the results of studies pertaining to radiation monitors, further evaluation . of meteorological measurements, and initial input from independent evaluation of the overall clarity of the guide, Regulatory Guide 1.97, Revision 2, was issued as an active guide in December 1980. l'be guide was issued with an outstanding question raised by the industry and supported by the AdYisory Committee on Reactor Safe- guards regarding the practicality of deployjng at fixed locations environs radiation moniton capable of detecting *

radioactive material releases from an unidentified breach of the containment. These monitors were listed* in the gi.lide but implementation of these provisions of the guide was

. delayed pending the outcome of a study that was to devel- op guidance as to their number and location. Additionally, shortly after the guide was issued, a research program was initiated with INEL to identify any modifications to the guide that might make the intent more clear.

The study pertaining to the environs radiation moniton has been completed and published in NUREG/CR-2644,

"An Assessment of Offsite, Real-Time Dose Measurement Systems for Emergency Situations."1 The conclusion was that it is unlikely that a few fixed-station area monitors could provide sufficiently reliable information to be of use in detecting releases from unmonitored containment release points. The NRC staff agrees with the conclusion of this study, and the environs radiation monitors have been deleted . from the PWR and BWR tables of variables of the guide.

Another evaluation by the NRC staff concluded that the function of exposure rate ni6niton inside auxiliary buildings and other buildings adjoining the containment (which were intended to measure releases caused by poten- tial breaches in the containment) could be just as effec-:

tively performed by the effluent moniton installed at release points from those buildings. Therefore, the expo- sure rate monitors inside buildings for the pw-pose of detecting containment breach were deleted from. the guide.

Exposure me monitors inside buildings where access is required to semce equipment important to safety have been retained.

1coplea may be obtafiled from the N!~~ Sales Propam, U.S. Nildear Regulatory Commission, W

on, D.C. 20555.

The NRC staff also agreed that the high accuracy speci*

fied in Revision 2 of Regulatory Guide 1.97 for the con- tainment radiation monitors is unnecessary and should be reduced, since conection factors can be applied to com- pensate for the energy spectrum.

An additional change agreed to by the NRCstaffpertains to meteorology. measurements. During the Committee -To Review Generic Requirements (CRGR) review of proposed Revision 1 of Regulatory . Guide 1.23, "Meteorological Programs in Support of Nuclear Power Plants" (Task SS

926-4), the *cRGR noted that several of the instrument range specifications on meteorology variables updated those .

presented in Revision 2 of Regulatory Guide 1.97 and recommended that both guides provide the. same spedfi- cations. Regulatory Guide 1.97 has been modified to agree with Proposed Revision 1 of Regulatory' Guide 1.23.

Of the clarifying modifications that have thus far been identified by the INEL evaluation, those that can be readily agreed to by the NRC staff are also included. These modifi*

cations include (1) listing the provisions for the design and q~cation criteria for Categoiies 1, 2, and 3 in a different format that was recommended as being more understand-

able, (2) changing the "range" provisions in the tables of variables to make them consistent, (3) correcting editing and printing enoi:s, and (4) clarifying the intent of the diseumon and regulatory position of the guide *

The value to NRC operations and industry is that many of the questions regarding radiation monitoring will be

resolved. Additionally, questions on guide .b:!tent frequently asked by industry will be settled by this revision,

.

2. OBJECTIVES

The above-mentioned deletions represent a substantive change in the NRC staff position regarding accident moni- toring that could represent a reduction in cost to the usen of Regulatory* Guide 1.97 with no reduction in safety since the environs radiation moniton were found not to be needed, as discussed above, and the function of the expo- sure rate monitors inside buildings can be effectively performed by effluent moniton. It is desirable that the users of. the guide be notified as soon as pomble to prevent unnecessary costs being applied to meet a provision no

,Onger recommended by the NRC staff. Since the guide is being revised to accomplish the above objectives, it is prudent to also include those modifications that have been identified as being essential to make the guide more under- standable.

Consequently, the guide is being revised to reflect these changes *.

3. ALTERNATIVES

The altemative is to take.no action to revise the guide but to inform licensees on an individual basis as inter- changes between the licensee tnd the staff pertaining to I

accident monitoring. occur.

1.97-32

4. CONSEQUENCES

If J10 action h taken, Ji~ and wendors may continue to incm costs to meet a provision that is no longer a :recom- mendation of the sbff. lime. will be Jost Jn answering questicms that could be uoided by muing a revision. *

5. DECISION RATIONALE

1be R'Vision of the guide should be issued. to .inform its usms of the cum:nt mff position, to clarify the staff posi-

~

and 1Ddimiaateorredueeannec:essarycostsincmred by *

tryiag .to meet PI01mODS that ae no longer recommended.

l, *.

,*!

,. DIPLEIIENTADON

1he implementation for this revision of Regulatory Guide 1.97 does not alter the implementation of Revision

2 as outlined Jn §50.49 of 10 CFR Part 50 and Supple- ment 1 to NUREG-0737. "Clarification of TMI Action Phm Requirements. "1 Since there are no new RCOm- mendaticms, there is no adverse impact on cost or schedule.

2wREGou, may lie obtained from tile NRC/GPO Sales *

~

adw ,..,.,.:Tc Conmrieion, Wllllmpoa, l).C.

1 11 !Mllable or llllsledioli:a er coPJiDg for a fee It tile Docameat Room, I '711 H Slnet,HW .. Wuhhiatoa, D.C.

1.97-33

Regulatory Guide 1.97: Difference between revisions

Latest revision as of 02:06, 17 January 2025

Contents

A. INTRODUCTION

B. DISCUSSION

6. Products

7. TransPortatlon

e. Occupational Health

e. sufficiently

3. Power Source

h. quality

s. lnttmments with

b. ForTypeE

e. ration

e. It JI not E

r. Isolation Valve

2. OBJECTIVES

3. ALTERNATIVES

4. CONSEQUENCES

5. DECISION RATIONALE

Navigation menu

Regulatory Guide 1.97: Difference between revisions

Latest revision as of 02:06, 17 January 2025

A. INTRODUCTION

B. DISCUSSION

6. Products

7. TransPortatlon

e. Occupational Health

e. sufficiently

3. Power Source

h. quality

s. lnttmments with

b. ForTypeE

e. ration

e. It JI not E

r. Isolation Valve

2. OBJECTIVES

3. ALTERNATIVES

4. CONSEQUENCES

5. DECISION RATIONALE

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