ML18041A221

Responds to Re Issues Concerning Reg Guide 1.97, Rev 3, Requirements for Post-Accident Neutron Monitoring Sys. Neutron Flux Monitoring Sys Should Be Environmentally Qualified for Design Basis Accident Spectrum,Per 10CFR50.49

ML18041A221
Person / Time
Issue date:	05/21/1990
From:	Russell W Office of Nuclear Reactor Regulation
To:	Floyd S BWR OWNERS GROUP
References
CON-IIT07-463A-91, RTR-REGGD-01.097 NUREG-1455, NUDOCS 9006050040
Download: ML18041A221 (86)

Text

Mr. Stephen D. Floyd Chairman BWR Owners'roup Carolina Powe~ and Light 411 Fayetteville Street Raleigh, North Carolina 27602 May 21, 1990

SUBJECT:

POSITION ON NRC REGULATORY GUIDE 1.97, REVISION 3 REQUIREMENTS FOR POST-ACCIDENT NEUTRON MONITORING SYSTEM I am responding to your letter of February 21, 1990 in which you raised several issues concerning post-accident neutron flux monitoring systems (NFHS).

The issues were in the areas of environmental qualification, fire and flood conditions, availability, and range.

For a detailed discussion of these issues please see the Enclosure.

The staff agrees that the BWR Owners'roup (BWROG) should develop generic design criteria for post-accident NFHS. It is the staff's position that NFHS should be environmentally qualified for the design basis accident spectrum in accordance with Code of Federal Regulations 10 CFR 50.49 and not other postulated events (e.g., fire).

The staff will consider plant s[ecific justifications for deviations in instrumentation

[ange above 10 percent full power.

However, the staff believes that 10 percent full power can be achieved and is appropriate to monitor shutdown neutron flux.

We look forward to prompt resolution of these

issues, certainly no later than July 1990.

If you have any questions regarding the above information, please contact Barry Marcus, of my staff on 49-20776.

Enclosure:

As stated cc:

A. Udy (EGSG Idaho)

I SEE PREVIOUS CONCURRENCE S incere ly,

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William T. Russell, Associate Director for Inspection 5 Technical Assessment Office of Nuclear Reactor Regulation SICB:DST* SICB:DST* SICB:DST*

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DISTRIBUTION:

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T. Murley F. Hiraglia W. Russell A. Thadani S. Newberry R. Jones F.

Hebdon J. Joyce L. Phillips C. Abbate H. Richings B. Marcus PDR

ENCLOSURE ISSUES RAISED BY BWROG LETTER DATED FEBRUARY 21 1990 Regulatory Guide (R.G.) 1.97 includes a Category 1 neutron flux monitoring system (NFHS) to monitor reactivity control during post-accident situations.

Reference 1 submitted the BWR Owners'roup (BWROG) Licensing Topical Report (LTR) for staff review and approval.

The LTR proposed functional criteria for NFMS as an alternative to the Category 1 criteria identified in R.G. 1.97.

Reference 2 provided the NRC Safety Evaluation Report (SER) which found the LTR to be unacceptable.

Reference 3 raised issues in the areas of environmental qualification, fire and flood conditions, availability, and range such as:

1.

The staff has not identified specific events on which to base qualification requirements or low end minimum range.

2.

The staff's intention is not to require qualification of the NFMS for environments beyond those associated with design basis events.

3..

The staff expects licensees to propose and technically justify specific design criteria to resolve open issues.

It was suggested by the staff that this work could be performed generically in order to avoid multiple reviews by the staff for each licensee.

4.

There appears to be a discrepancy between Reference 2 and Code of Federal Regulations 10 CFR 50.49.

10 CFR 50.49 requires that environmental conditions be established for the most severe design basis accident (DBA) during or following which instrumentation is required to function.

However, Reference 2 takes the position that the post-accident NFHS design should consider events that cannot be anticipated by standard event analyses.

This is a design issue which could raise 10 CFR 50.49 compliance concerns unless it is resolved with the staff prior to individual plant implementation of the regulatory guide.

5.

Reference 2 takes the position, that NFHS should be qualified in accordance with 10 CFR 50.49 and also qualified for events that go beyond 10 CFR 50.49 design basis events.

R.G. 1.97 includes a Category 1

NFMS to monitor reactivity control during post-accident situations.

10 CFR 50.49 requires that certain post-accident monitoring equipment (as provided in Revision 2 of R.G. 1.97) be environmentally qualified under conditions existing during and following DBAs.

This includes the Category 1 NFMS.

As indicated in Reference 3, the scenarios for which the recommended low end of the range (10 percent full power) might be needed to provide an early warning of abnormal reactivity conditions and possible return to criticality following shutdown have not been specifically defined.

The conditions within and surrounding the reactor to be considered for environmental qualification of the NFHS should be those associated with the typical spectrum of design basis events.

Conditions beyond that scope are not required.

The appropriate maximum conditions within that envelope are those associated with the large break LOCA.

Since the NFHS needs to be qualified to DBA environments there is no conflict with 10 CFR 50.49.

Because NFHS are not required in deterministic DBA analyses, does not cause us to conclude that a discrepancy exists between the staff SER and 10 CFR 50.49.

It is our view that a

NFHS qualified for a DBA environment in accordance with 10 CFR 50.49 would be very likely to survive for a spectrum of accidents "beyond" the DBA.

There could be severe accident sequences postulated for which such a system would not survive,

however, we would expect that the need to monitor return to criticality and shutdown margin would not be the primary safety concern for these events.

It is not the staff's intent to require qualification beyond the environment associated with design basis events.

In Reference 3 the BWROG stated that Reference 2 suggested that fire conditions be considered, and that the inclusion of fire conditions in the design of NFHS appears to be outside the scope of R.G. 1.97.

The BWROG also stated that there are other design issues such as flooding that also need to be considered.

The fire conditions referred to in Reference 2 were in reference to conceptual causes for control rod actuations and position information loss and not to environmental conditions used for qualification of the NFHS.

The discussion of flooding was in the context of the environmental conditions associated with DBA conditions.

Reference 3 stated that instrumentation availability time (as referenced in Reference

2) along with related design basis issues need to be established.

The time range for required operability of the NFHS is based on the staff juagment of the necessary and sufficient time frame in which it might usefully assist in diagnosis of possible recriticality problems, and not on specific events.

In the staff's view this is the order of 60 days under conditions which might exist in the DBA environment indicated above.

The staff understands, based on conversations with vendors, that the NFHS being developed are being environmentally qualified for 100 days post-accident for the General Electric system and 6 months post-accident for the Gamma-Hetrics system.

Reference 3 stated that the staff is aware of the potential affficulty for plant-specific implementation constraints with respect to achieving the full r ange specified by R.G. 1.97.

The staff is willing to consider plant specific deviations based on technical justifications and the capabi lity of available equipment.

-6 The sensitivity range of the system, down to 10 percent of full power, was intended to provide the potential for maximum sensitivity to anomalous shutdown reactivity conditions during the time range indicated above.

A normal shutdown (by scram) would typically reach the vicinity of this level (as determined by neutron sources developed during operation) in about 15 to 22 minutes and would remain near that level for an extended period4 A less extended lower range, for example, to 10 percent or possibly 10

percent, would be somewhat less sensitive, but still useful.

The staff would consider, on a plant specific basis, a less extended range for NFMS which cannot meet the 10 percent le~el.

At the present time WNP-2 has installed a system with a range of 10 to 100 percent full power a~d Susquehanna Units 1 and 2

have installed systems with a range of 3.3 x 10 to 100 percent full power.

Reference 3 stated that it may be appropriate for the BWROG to develop generic design criteria for post-accident neutron monitoring which could serve as a

focal point for further discussion with and review by the staff. It is the staff's position that the environmental qualification, fire and flood conditions, time availability and range questions raised in Reference 3 have been answered.

However, these issues appear worthy of some additional review by the BWROG to substantiate our judgment.

NFMS have been installed at BWR plants for which the licensees have certified that the Category I criteria of R.G. 1.97 and 10 CFR 50.49 have been met.

Therefore the installation of NFMS that comply with the Category 1 criteria of R.G. 1.97 and 10 CFR 50.49 is achievable and has been accomplished.

REFERENCES 1)

BWR Owners'roup letter (R. F. Janecek) to NRC (T. E. Murley),

"BWR Owners'roup Licensing Topical Report 'Position on NRC Regulatory Guide 1.97, Revision 3 Requirements for Post-Accident Neutron Monitoring System,'General Electric Report NEDO 31558)," April 1, 1988.

2)

NRC letter (F. J. Miraglia) to BWR Owners'roup (S.

D. Floyd),

"BWR Owners'roup Licensing Topical Report 'Position on NRC Regulatory Guide 1.97, Revision 3 Requirements for Post-Accident Neutron Monitoring System,'General Electric Report NEDO 31558)," January 29, 1990.

3)

BWR Owners'roup letter (S.

D. Floyd) to NRC (F. J. Miraglia), "Position on NRC Regulatory Guide 1.97, Revision 3 Requirements for Post-Accident Neutron Monitoring System (NMS)," February 21, 1990.

0

Page Nc 1

06/07/88 REGULATORY GUIDE 1.97 BWR CATEGORY 1 VARIABLES 12 VARIABLES VARIABLE TYPE CAT SYSTEM PLANT SPECIFIC TYPE A A

CONTAINMENT 5 DRYWELL HYDROGEN CONCENTRA C

CONTAINMENT 5 DRYWELL OXYGEN CONCENTRA C

COOLANT LEVEL IN REACTOR (INVENTORY)

B RYWELL DRAIN SUMP LEVEL C

DRYWELL PRESSURE B,C RYWELL SUMP LEVEL B

NEUTRON FLUX B

PRIMARY CONTAINMENT AREA RADIATION E

PRIMARY CONTAINMENT ISOLATION VALVE POS B

PRIMARY CONTAINMENT PRESSURE B,C RCS PRESSURE B,C SUPPRESSION POOL WATER LEVEL C

1 1

CONTAINMENT 1

CONTAINMENT 1

CORE COOLING 1

REACTOR COOLANT PRESSURE BOUNDRY 1

MAINTAININGRCS INTEGRITY, REACTOR COOLANT PRESS BOUNDRY 1

MAINTAININGRCS INTEGRITY 1

REACTIVITY CONTROL 1

CONTAINMENT RADIATION 1

MAINTAININGCONTAINMENT INTEGRITY 1

MAINTAININGCONTAINMENT INTEGRITY, CONTAINMENT 1

MAINT RCS INTEGRITY, REACTOR COOLANT PRESS

BOUNDRY, CONT 1

REACTOR COOLANT PRESSURE BOUNDRY

Hevlslof1 2 December 1980

-,i REGULATORY GUIDE REGULATORY GUIDE 1.97 (Task RS 9174)

INSTRUMENTATIONFOR LIGHT-WATERCOOLEDNUCLEAR POWER PLANTS TO ASSESS PLANT AND ENVIRONS CONDITIONS DURING AND FOLLOWING AN ACCIDENT A. INTRODUCTION B. DISCUSSION Criterion l3, "instrumentation and Control," of Appen-dix A, "General Design Criteria for Nuclear Power Phnts,"

to IO CFR Part 50, "Domestic Ucensing of Production and Utilization Facilities," includes a requirement that instru-mentation be provided to nlonitor variables and systems over their anticipated ranges for accident conditions as appropriate to ensure adequate safety.

Criterion I 9, "Control Room,"

of Appendix A to IO CFR Part 50 includes a requirement that a control room be provided from which actions can be taken to maintain the nuclear power unit in a safe condition under accident conditions, including losswf.coolant accidents, and that equipment, including the necessary 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 IO CFR Part 50 includes a requirement that means be provided for monitoring the reactor containmcnt atmosphere, spaces containing components for recirculation of losswf-coolant accident fluid, effluent discharge

paths, and the plant environs for radioactivity that may be released from postulated accidents.

Indications of plant variables are required by the control room operating personnel during accident situations to (I) provide information required to permit the operator to take preplanned manual actions to accomplish safe plant shut-down; (2) determine whether the reactor trip, engineered-safety-feature

systems, and manually initiated safety systems and other systems important to safety are performing their intended functions (i.e.,

reactivity control, core cooling, maintaining reactor coolant system integrity, and maintaining containment integrity): and (3) provide informa-tion to thc operators that willenable them to determine the potential for causing a

gross breach of the barriers to radioactivity release (i.e.,

fuel cladding, reactor coolant pressure boundary, and containment) and to determine ifa gross breach of a barrier has occurred.

In addition to the above,.indications of plant variables that provide informa-tion on operation of plant safety systems and other systems important to safety are required by thc control room operating personnel during an accident to (I) furnish data regarding the operation of plant systems in order that the operator can make appropriate decisions as to their use and (2) provide information regarding the release of radioactive materials to allow for early indication of the need to initiate action necessary to protect the public and for an estimate of the magnitude of any impending threat.

This guide describes a method acceptable to the NRC staff for complying with the Commission's regulations to provide instrumentation to monitor plant variables 2nd systems during and following an accident in a light-water-cooled nuclear power plant. The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position.

The substantial number of changes In Ibis revision hss made II Impfscficsl Io indicsle the changes with lines in Ihs msrgin.

At the start of an accident. it may be difficult for the operator Io determine immediately what accident has occlirred or is occurring and thcrcfore to dcterminc the appropriate response.

For this reason.

reactor trip and certain other safety actions (e.g., emergency core cooling actuation, containment isolation, or depressurization) have hccn designed to be performed automatically during the

'nitial stages of an accident. Instrumentation is also provided to indicate information about plant variables required to enable the operation of manually initiated safety systems USNRC REGULATORYOUIDES Rsoulstory Ouides sro Issued to describe snd moke svsllsblo to the Public methods scceptsbl ~ to the NRc slsff of Implementing spo4fic pstts of tho commission's foyulstions, to deiineste tech nieuss used by the staff In evslustiny spe4II>> problems of Qosfu Istsd sc4dents(

or Io Provide evidence to sppllesotL Rsoulstofy ouides sro no( substitutes for ssyvlsflons, sna compiisnco witn them Is not required. Msthoas sna solutions different from those sst oul In tho yuiass will be sccoptsble If they provide s bssis fof the..

Mx%flnalnyslroeulslto to tho.geushee,pr eOntinusnco of.e.permit,or

'ieopso by tho Commission,;~ ~,

. "l

~ or'l',$'...4'~ ',g ~ ",she'l'I comments snd aooestions for Improvements In those yuiass sfo

~hcoursyed st ~Ii limes, snd yuides will bo revised, ss opproprisf ~,

to sceommoaste comments sna Io reflect hsw'Infofmsfion or

~xpsrlsnce.

This oulae wss rwlsed ss s result of subslsnllve corn.

ments received from the public sna saaiflonsl slsff review.

Comments should be sent to tho secretory of tho Commission, U.S.

Nuclear Reyulstory Commission VVsshinyton, O.C. 20SSS, Aitentionl Docketlny snd service bfsndl.

Tho yuldes are issued Ih ths fotiowinyten brood dlvlslonsl

l. Power Resetors d, Roaucts

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. y. Antitrust snd Fihsn4si Revl<<w S. Msterisis sod Rant protection lb. Oenofsi copies of Issuid yuldes msy bo purchased st Iho current Oovorhmsnt pfintlno Office Dries. A subscription service for future yuidss io soe.

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Informstion on tho subscription service snd current OPo prices msy bo obtslnsa by wrltlno the U.s. Nuclear Rsoul ~lory Commission, Vvsshlnofon, O.c.

20SSS, AIIsnfloni publlcstlons Soles Msnsosr.

and other appropriate operator actions involving systems important to safety.

independent of the above tasks, it is important that operators bc informed if thc barriers to the release of radioactive materials are being challenged.

Therefore, it is essential that instrument ranges be selected so that the instrument willalways be on scale. Narrow.range instruments may not have the necessary range to track the course of thc accident; consequently.

nrultiple instruments with over-lapping ranges may be necessary.

(In the past, some instru-ment ranges have been selected based on the setpoint value for automatic protection or alarms.) it is essential that degraded conditions and their magnitude b>> identified so the operators can take actions that are available to mitigate the consequences.

It is not intended that operators 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 if conditions degrade are loss-of-coolant accidents (LOCAs),

ovcrpressure transients, anticipated operational occurrences that become accidentssuch as anticipated transients without sl ram (ATWS)~ and reactivity excursions that result in releases of radioactive materials.

Such events require that the operators rrnderstanJ.

within a short time period, the ability of the barriers to limit radioactivity release, i.e., that they understand the potential for breach of a barrier or whether an actual breach of a barrier has occurred because f an accident in progress.

lt is essential that thc required instrumentation be capable of surviving the accident environment in which it is located for the length of time its function is required. It could therefore either he designed to withstand the accident environment or be protected by a local protected environ-ment.

It is desirable that accident-monitoring instrumentation components and their mounts that cannot be located in seismically qualified buildings bc designed to continue to I'unction, to the extent feasible, following seismic events.

An acceptable method for enhancing the seismic resistance of this instrunrcntation would be to design it to meet the seismic criteria applicable to like instrumentation installed in seismically qualified locations although a lesser over-all qualification results.

Variables for accident monitoring can be selected to provide the essential information needed by the operator to determine ifthc plant safety functions are being performed.

lt is essential that the range selections be sufficiently great to keep instruments on scale at all times. Further, it is prudent that a limited number of those variables that are functionally significant (e,g., containment pressure, primary system pressure) be monitored by Instruments qualified to oro stringent environmental requirements and with ranges hat extend well beyond that wltich tho selocted variables can'attain under limiting conditions; for oxample, a range for,tho containment.pressure monitor extonding to the burst pressure of the containment in order that the operators will not be uninformed as to the pressure inside thc contain.

ment. The 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 bc deter-mined. It is also necessary to be sure that when a range is extended, the sensitivity and accuracy of the instrument arc within acceptable limits for monitoring thc extended range.

Normal power phnt instrumentationremaining functional for all accident conditions can provide indication, records, and (withcertain types ofinstruments) time.history responses for many variables important to following the course of the accident.

Therefore, it is prudent to select the required accident-monitoring instrumentation from the normal power plant instrumentation to enable operators to usc, during accident situations, instruments with which they are most I'amiliar. Since some accidents could impose severe operating requirements on instrumentation components, it may bc necessary to upgrade those normal power plant instrumentation components to withstand the more severe operating conditions and to measure greater variations of monitored variables that may be associated with an accident.

It is essential that instrumentation so upgradcrl docs not degrade the accuracy and sensitivity required for normal operation. In some

cases, this will necessitate use of over.

lapping ranges of instruments to monitor the required range of the variable to be monitored, possibly with different performance requirements in each range.

ANSI/AN&4.5-I980,t "Criteria for Accident Monitoring Functions in Light-Water-Cooled Reactors."

delineates criteria for determining the variables to be monitorea by thc control room operator, as required for safety, during the course of an accident and during the long-term stable shutdown phase following an accident.

ANS<.5 was prepared by Working Group 4.5 ot Subcommittee ANS<

with 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 address extended-range instrumentation deemed appropriate for the possibility of encountering previously unforcsecn events.

ANS<.5 rcfercnces a revision to IEEE Standard 497 as the source I'or specific instrumentation design criteria. Since the revision to IEEE Standard 497 has not been complctcd, its applicability cannot yet be determined.

Hence, specific instrumentation design criteria have been included in this regula tory guide.

ANS<.5 defines three types of variables (definitions modified herein) for the purpose of aiding the designer in selecting accident-monitoring instrumentation and applicable criteria. The types are: Type A, those variables that provide primary information needed to permit the control room Coplee mey be obtehred from the American Nudeer Sodety>

SSS North Kenetnston Avenue, Le Gcsnge Park, Itlinote 60515..

Primery. Irrformettoa ie Informetton that Ie eeeentht for the g:

dtrect aceomptiehment of the epedfted eafety functtone; It does not Indude thrree verteblce that ere eeeodeted wtth conttnseney ecttone that mey eteo be identified In wrttten proceduree.

~ 4 1.97-2

0 C>

operating personnel to take the specified manually controlled actions I'or which no automatic control is provided and that e required for safety systems to accomplish their safety unctions for design basis accident events; Type B, those variables that provide information to indicate whether plant safety functions are being accomplished; and Type C, those variables that provide information to indicate the potential for being breached or the actual breach of the barriers to fission product release, i.e., fuel cladding, primary coolant pressure

boundary, and containment (modified to reflect NRC staff position',

see regulatory position 1.2).

The sources of potential breach are limited to the energy sources within the barrier itself. In addition to the accident-monitoring variables provided in ANS<.5, variables for monitoring the operation of systems important to safety and radioactive effluent releases are provided by this regulatory guide. Two additional variable types are defined:

Type D, those variables that provide information to indicate the operation of individual safety systems and other systems important to safety, and Type E, those variables to be monitored as required for use in determining the magnitude of the release of radioactive materials and for continuously assessing such releases.

A minimum set of Type B, C, D, and E variables to be measured is listed in this regulatory guide. Type A variables have not been listed because they are plant specific and will depend on the operations that the designer chooses for planned manual action. Types B, C, D, and E are variables for foUowing the course of an accident and are to be used I) to determine if the plant is responding to the safety easures in operation and (2) to inform the operator of the necessity for unplanned actions to mitigate the con-sequences of an accident. The five classifications are not mutually exclusive in that a given variable (or instrument) may be applicable to one or more

types, as well as for normal power plant operation or for automatically initiated safety actions.

A variable included as Type B, C, D, or E does not preclude that variable from also being included as Type A. Where such multiple use occurs, it is essential that instrumentation be capable of meeting the more stringent requirements.

requirements and generally appHes to instrumentation designated for indicating system operating status. Category 3 is intended to provide requirements that wiH ensure that highguality oi'f-the-shelf instrumentation is obtained and applies to backup and diagnostic instrumentation. It is also used where the state of the art willnot support requirements for higher qualified instrumentation.

In general, the measurement of a single key variable may not be sufficient to indicate the accompHshment oi' given safety function.

Where multiple variables are needed to indicate the accompHshment of a given safety function, it is essential that they each be considered key variables and be measured with high~uaHty instrumentation. Additionally, it is prudent, in some instances. to include the measurement of additional variables for backup information and for diagnosis. Where these additional measurements are included, the measures appHed for design, qualiflcation, and quality assurance of the instrumentation need not be the same as that applied for the instrumentation for key variables.

A key variable is that single variable (or minimum number of variables) that most directly indicates the accomplishment of a safety function (in the case of Types B and C) or the operation of a safety system (in the case of Type D) or radioactive material release (in the case of Type E). It is essential that key variables be qualiTied to the morc stringent design and qualification criteria. The design and qualification criteria category assigned to each variable indicates whether the variable is considered to be a key variable or for system status indication or for backup or diagnosis, ie., for Types B and C, the key variables are Category I; backup variables are generally Category 3. For Types D and E, the key variables are generally Category 2: backup variables are Category 3.

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

The time phases (Phases I and Il) delineated in ANS<.5 are not used in this regulatory guide. These considerations are plant specific. It is important that the required instru-mentation survive the accident environment and function as long as the information it provides is needed by the control room operating personnel.

\\

The NRC staff is willing to work with the ANS working group to attempt to resolve the above differences, Regulatory positional.3 and 1.4 of this guide provide design and qualification criteria for the hstrumentation used to measure the various variables Hated h Table 1 (for BWRs) and Table 2 (for PWRs). The criteria aro separated hto three separate groups or categories that provide a ded approach to requirements dependhg on tho irnpor-tancc'o safety of tho meaeuzometit of a specifi variable.

Category 1 provides tho moat etringent zoquiremonte'and Ia iintended for key variables. Category 2 provides less stringen This guide provides the minimum number of variables to be monitored by the control room operating personnel during and foHowing an accident. These variables are used by the control room operating personnel to perforin their role in the emergency plan in the evaluation, assessment, monitoring, and execution of control room functions when the other emergency response facilities are not effectively manned.

Variables are also defined to permit operators to perform their long.term monitoring and execution respon-sibilities after the emergency response facIHties are manned.

The application of the criteria for tho instrumentation is Hmited to that part of the hatzumontation system and its vital supporting features or power aourcos that provide the direct display of tho variabloL Theao provisions azo not necessarily appHcablo to that part of tho instrumentation systems provided as operator aide for tho purpose of

.enhanchg hfortnation'presentations for the Identification or dlagztoah of disturbances.

1.97-3

c

C. REGULATORY POSITION ccident-Monitoring Instrumentation

~

~ ~

~

~

he criteria and requirements contained in ANSI/ANS<.5-F 980, "Criteria for Accident Monitoring Functions in Light-Water-Cooled Reactors," are considered by the NRC staff to be generally acceptable for providing instrumentation to monitor variables for accident conditions subject to the following:

I.I Instead of the definition given in Section 3.2.1 of ANS-4.5, the definition of Type A variables should bc:

Type A, those variables to be monitored that provide the primarv information required to permit the control room operators to take the specified manually 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.

1.2 In Section 3.2.3 of ANS<.5, the definition of

~ Type C includes two items, (I) and (2). Item (!) includes those instruments that indicate the extent to which variable that have the potential for causing a breach Ii the primar>

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 variables that indicate the potential for causing a breach in the fuel cladding (e.g., core exit temperatur

) and the reactor coolant pressure boundary (e.g., reactor coolant sure) should also be included. The sources of potential L

ch are limited to the energy sourc s within the cladding, iant boundary, or contain~nent.

Rei'erences;o Type C istrumcnts, and associated paramet rs to be measured, in ANS-4.5 (e.g., Sections 4.2, 5.0, 5.1.3, 5.2, 6.0, 6.3) should include this expanded definition.

1.3 Section 6.1 of ANM.5 pertains to General Criteria for Types A, B, and C accident-monitoring variables. In lieu

~of Section 6.1, the following desigr.

and qualification criteria categories should be used:

1.3.I Design and Quali/icarion Clireria - Caregurv /

a. Thc instrumentation sI>ouid ne qualified in accordance with Regulatory Guide 1.89, "Qualification of Class I E Equipment for Nuclear Power Plants," and the methodology described in NUREG4588, "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment." Qualification applies to the complete instru-mentation channel from sensor to display where the disphy is a direct-indicating meter or recording device. Where the instrumentation channel signal is to be used in a computer-based

display, recording, and/or diagnostic

program, qualification applies from the sensor to and includes the channel isolation device.

The location of the isolation device should be such that it would be accessible for ainten ance during accident conditions.

The seismic ion of qualification should bo in accordance with tlatory Guide 1.100, "Seismic QualiTication of EIoctric

.pment..for.Nuclear PoLver Fiant'" Ittstzumontation would continuo. to road within,tho. sequirod accuracy following, but not necessarily

during, a

safe shutdown earthquake.

Instrumentation whose ranges are required to extend beyond those ranges calculated in the most severe design basis accident event for a given variable should bc qualified using the guidance provided in paragraph 6.3.6 <<

ANS<.5.

b. No single failure withinclthcr the accident. monitoring instrumentation, its auxiliary supporting features.

or its power sources concurrent with thc failures that are a

condition or result of a specific accident should prevent thc operators from being presented thc information neces-sary for the>n to determine the safety status of the plant and to bring the plant to and maintain it in a safe condition I'ollowing that accident.

Where failure of one accident-n:onitoring channel results in ini'ormation ambiguity (that is, the redundant dispiaysdisagree! that could lead operators to defeat or fail to accomplish a required safety function, additional information should be provided to allow the operators to deduce the actual conditions in the plant. This may be accomplished by providing additional independent channels of information of tl'e same variable (addition of an identical channel) or by providing an independent channel to monitor a different variable that bears a known relationship to the tnultiple channels (addition of a divcrsc channel). Redundant or diverse channels should be electrically inaependent and physically separated from each other and from equipment not classified important to safety in accordance with Regulatory Guide 1.75, "Physical Inde-pendence of Electric Systems,"

up to and including any isolation device. At least one channel should be displayed on a direct-indicating or recording device. (Note:

Within each redundant division of a safety

system, redundant monitoring channels are not needed except I'or steam generator level instrumentation in two.loop plants.)

c. The instrumentation should be energized from station Standby Power sources as provided in Regulatory Guide 1.32, "Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants,"and should be backed up by batteries where momentary interruption is not toler ble.

d. The instrumentation channel should be available prior to an accident except as provided in paragraph 4.11, "Exemption,"

as aefined in IEEE Standard 279 or as specified in Technical Specifications.

e. Thc recommendations of the following regulatory guides pertaining to quality assurance should be I'ollowed:

pegulatory Guide 1.28 "Quality Assurance Program Requirements (Design and Construction)"

Regulatory Guide 1.30 "Quality Assurance Require.

ments for the Installation, Inspection, and Testing of Inatsumontatiott and Electric Equipment" Reguiatory Guido 1.38.

"Quality Amntranco Require-

,ments forPackaging, Shipping,g:.'

Receiving, Storage, and Han.

dlingof Items for Water-Cooled Nuclear Power Plants" Regulatory Guide 1.58 "Qualification of Nuclear Power Plant Inspection, Exaniination, and Testing. Personnel" Regulatory Guide 1.64 "Quality Assurance Require ments for the Design of Nu-clear Power Plants" Regulatory Guide l. 74 "Quality Assurance Terms and Definitions" Regulatory Guide 1.88 "Collection. Storage,and Main-tenance of Nuclear Power Plant Quality Assurance Records" Regulatory Guide 1.123 "Quality Assurance Require-ments for Control of Procure-nient of !tems and Ser:icos fnr

."u h."r Pnwc:

isn't;"

Regulatory Guide 1.144 "Auditingnf QL>lityAssurance Programs for Nuclear Pnwcr Phnts" Regulatory Guide 1.146 "Qualification of Qiiality Assurance Program Audit Personnel for Nuclear Power Plants"

f. Continuous indication (it may hc by !ceo!ding) display should bc proviilcd. Where two or innrc instruments are needed to cnvcr a parti<<uhr range, ovcrhpping of instru-ment span should hc provided.

Rcfcrencc tn the above regulatory guides (excep! Regula-tory Guides 1.30 and 1.38) is being made pending. issuance of a regulatory guide (Task RS 002-5) that is under develop-ment and wiH endorse ANSI/ASME NQA-1-1979. "Quality Assurance Program Requirements for Nuclear Power Plants."

the channel signal is to be processed or displayed on demand.

qual!%ication applies from the sensor through thc iscla!rr/

input buffer. The location of the isolation ilcvi<<c shoul>l l>c such that it would be accessible for maintenance during acci dent conditions.

b. The instrumentation should be encigiacd frniu a

high-reliability Power

source, not neccswrily Standby Power, and should be hacked up by batterics where <iionicii-tary interruption is not tolerable.

c. The outwf-service interval should hc based on nnrnial Technical SpeciTication requirenicnts on o:ii of scrvi<<<< fnr the system it serves where applicable or where specified hy other requirements.

d. The recommendations of the regulatory guides pertaining to quality assurance Usted under paragraph 1.3.1c of this guide should bc followed. Reference to the abnv>>

regulatory guides (except Regulatory Guides 1.30 and 1.38) is being made pending issuance nf a rcg>ilator>

gi>ice (Task RS002-5) that is under development a!id '.vill ".nilnrcc A>VSIt>i.'>ME NQA.1.1979.

S..>cc snuic

.'>is'.i>i nant.>!i..>::.

! ss important to sai'ety than other instrumentation. it niay n!>t bc necessary to apply the same quality;sat>rance

>ixusurcs to sll instrumentation. The quality assurance rcqi>iran>cuts that are implemented should provide contr'ol over activities affecting. quality to an extent consistent with the importance to safety of the instrumentation. These requirements should be drtermincd and documented by personnel knowledgeable in the cnd use of the instrumentation.

e. The instrumentation signal may be displayed nn an individual instrument or i! may he processed for Jlst>lay "r.

demand hy a CRT or hy other appropriate mean>.

f. Thc methnd of display niay he by dial. digital. CRT.

or stripchart recorder indication.

EfAuent radioactivity monitors. area radiation monitors, and meteorology nionitnrs shnuld bc recorded.

Where direct and immediate trend or

!ransi<<nt information is essential foroperator informaiiniior aciion. tlic recording should be continuously availahlc nn >le>I~

ica!ed recorders. Othe: wise, it may be contillllnuslyui'.d.>l<<d.

stored in coniputcr memory.

and display<<d nn d<<nixon.

g. Recording of instrumentation readout informatior.

should be provided. Where direct and immediate trend or transient information is essential for operator information or action, the recording should be continuously availahlc on dedicated recorders. Otherwise, it may be continuously

updated, stored in computer memory, and displayed on demand.

Intermittent displays such as data loggers and scanning recorders may be used if no significant transient response information is likely to be. lost by such devices.

1,3,2 Design and guallpaation Criteria - Category 2 a Tho hstrumentation should be qualiTied in accordance

..with Regulatory Guido 1.89 and Iho methodology described

',.h NURBG4588. Seismic qualification <<ccordhg to the pro-visions of Regulatory Guide 1.100 may bo needed provided the hsttumentation Is part of <<safetymlated system, Whero 1.3.3 Design and Qualification Criteria - Carcgor! 3

a. The instrumentation should be nf high quality commercial grade and should bc selected tn withstand ihc specified service environment.

b. The method of display may bcbydial,digital,CRT,or stripchart recorder indication. Effiiantradioactivity monitors, area radiation monitors, and meteorology monitors should be recorded.

Whcro direct and immediate trend or transient information la essential for operator information or action, tho recording should bo conthuously available on dedicated secorde!L Othorwiso, it may be conthuously

updated, stotod in computer memory, and displayed on demand.

1.4. 1n addition to tho criteria ofrogubitory position 1.3, tho foliowhg criteria should apply to Categories 1 and 2:

1.9'7-5

a. Any equipment that is used for either Category I or Cat.

egory 2 should be designated as part of accident-monitoring instrumentation or systems operation and effluent.monitoring instrumentation.

The transmission of signals from such equipment for other use should be through isolation devices that are designated as part of the monitoring instrumentation and that meet the provisions of this document.

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

of instrument channels.

(Note:

Response

time testing not usually needed.}

l.6 Sections 6.2.2, 6.2.3, 6.2.4.

6.2.5, 6.2.6, 6.3.2, 6.3.3, 6.3.4, and 6.3.5 of AN&4.5 pertain to variables and variable ranges for monitoring Types B and C variables.

In conjunction with the above listed sections ol'NSI'.5, Tables I and 2 of this regulatory guide (which include those variables mentioned in these sections) should be considered as the minimum number of instruments and their respective ranges for accident-monitoring instrumentation for each nuclear power plant.

I.S ln addition to the above

criteria, the following criteria should apply to Categories I, 2, and 3:

2. Systems Operation Monitoring and Effluent Release Monitoring Instrumentation

a. Servicing, testing, and caHbration programs should be specified to maintain the capability of the monitoring instrumentation. For those instruments where thc required interval between testing will be less than the normal time interval between generating station shutdowns, a capability for testing during power operation should be provided.

b. Whenever means for removing channels from service are included in the

design, the design should facilitate administrative control of the access to such removal means.

2.1 Definitions

a. Type D, those variables that provide information to indicate the operation of individual safety systems and other systems important to safety.

b. Type E, those variables to be monitored as required for use in determining the magnitude of the release of radioactive materials and in continually assessing such releases.

c. The design should facilitate administrative control of the access to afl setpoint adjustments, module calibration adjustments, and test points.

d. The monitoring instrunx:ntation design should minimize the development of conditions that would cause meters, an-nunciators, recorders, alarms, etc., to give anomalous indica-tions potentially confusing to the operator. Human factors analysis should be used in determining type and location of displays.

e. The instrumentation should be designed to facilitate the recognition. location, replacement, repair, or adjustment of malfunctioning components or modules.

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

a. Ascertain the operating status of each individual safety system and other systems important to safety to that extent necessary to determine ifeach system is operating 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 ascertain if there have been significant rc:leases (planned or unplanned) of radioactive niaterials and to continually assess such releases.

f. To the extent practicable, monitoring instrumentation inputs should be from sensors that directly measure the desired variables. An indirect measurement should be made only when it can be shown by analysis to provide unambigu-ous information.

g. To the extent practicable, the same instruments should be used for accident monitoring as are used for the normal operations of the plant to enable the operators to

usc, during accident situations, instruments with which they are most famiHar. However, where thc required range of monitoring instrumentation results in a loss ofinstrumen-tation sensitivity in the normal operating range, separate

~

~

Instruments should be used.

h. Periodic checking, testing, calibration, and caHbration veriflcation should be in accordauco with tho appHcablo portions of Reguhtory Guide 1.118, "Periodic Testing of Hectric Power and Protection Syatortis,". pertahirigto testing

c. Obtain required inl'ormation through a backup or diagnosis channel where a single channel may be likely to give ambiguous indication.

2.3 The process for selecting system operation and effluent relcasc variables should include the identification of:

a. For Type D (I) The plant safety systems and other systems important to safety that should bo operating or that could bc: placed in oporation to hoip mltigato tho consequc'nces of an accident; and (2) Tho variable or minimum number of variable that indicate tho operating status of oach system IdenNie

'h,(l) abovo.

I.97-6

b. For type E (I) The planned paths for effluent release; (2) Plant areas and inside buildings where access is required to service equipment necessary to mitigate the consequences of an accident; instrumentation. should be taken from the criteria provided in regulatory positions 1.3 and 1.4 of this guide. Tables I and 2 of this regulatory guide should be considered as the minimum number of instruments and their respective ranges for systems operation monitoring (Type D) and effluent release monitoring (Type E) instrumentation for each nuclear power plant.

(3) Onsite locations where unplanned releases of radioactive materials should be detected; and D. IMPLEINENTATION (4) Thc variables that should be monitored in each location identified in ( I ), (2), and (3) above.

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

2.4 Thc determination of performance requirements for system operation monitoring and effluent release monitoring information display channels should include, as a minimum, identification of:

Plants currently operating should meet the provisions of this guide, except as modified by NVREG%737 and the Commission Memorandum and Order (CL1-80-21), by June 1983.

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. Thc 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.

Plants scheduled to be licensed to operate before June I, 1983, should meet the requirements of NUREG4737 and the Commission Memorandum and Order (CLI-80-21) and the schedules of these documents or prior to the issuance of a license to operate, whichever date is later. Thc balance of thc provisions of this guide should be completed by June 1983.

The difficulties of procuring and installing additions or modifications to in-place instrumentation have been con.

sidered in establishing these schedules.

2.5 The design and qualification criteria for system operation monitoring and cfflucnt release monitoring Exceptions to provisions and schedules willbe considered for extraordinary circumstances.

1.97-7

TABLE 1 BWR VARIABLES TYPE A variables: those variables to be monitored that provide the primary information rcquircd to permit the conlriil room operator to take specific manually controlled actions for which no autom ti a

c control is provided and that are required I'or safety systems to accomplish their safety functions for design basis accident cvcnts. Primary information is infori>>a.

tion that is essential for thc direct accomplishment of the specified safety functions: i

's: it does not include those vanulilcs tliat are associated with contingency actions that may also be identified in written procedures.

Yerieble Purpose R~an A variable included as Type A docs not preclude it from being included as Type B, C. D, or E or vice versa.

Category (sec Regulatory Position 1o3)

Plant specific Plant specific Information required for operator action TYPE B Variables: those variables that provide information to indicate whether phnt safety I'unctions are being accomplislicil.

Plant safety functions are (I) reactivity control, (2) core cooling, (3) maintaining reactor coolant system integrity, and (0) niaintaining containmcnt integrity (including radioactive effluent control). Variables are

'isted with designated ranges anil category for design and qualification rcquiremcnts. Kcy variables are indicated by desig an q

n 4

ualification Category I.

Reactivity Control Control Rod Position RCS Soluble Boron Concen-tration (Sample)

IO

% to l00% full power (SRM, APRM)

Full in or not full in 0 to l 000 ppm Function detection; accomplishincnl of mitigation

~

Verification Verification Core Cooling Coolant Level in Reactor Bottom of core support phte to lcsscr of top of vessel or centcr-linc of main stcam line.

Function dctcction; accomplishnicnt of mitigation: long-term surveillance BWR Core Thermocouples 200 F to 2300 F To provide diverse indication of water level Maintaining Reactor Coolant System Integrity RCS Pressure Drywall Pressure 15 psia to I 500 psig 0 to design prcssure (psig)

Function detection; accomplishiucni of mitigation; verification Function detection, accomphsh>>ii>>l ot mitigation, verification Four thcimocouplee per quacrant. A minimum of one measurement per quadrant Is required for operation.

Wh<<re ~ variable ts tiered for nmre than one purpose, the Inetrumentatlon requirements may be integrated md caity one measurement prorlded.

Iiestan ptereure ta that value eoireepondlna to ASMR code values that are obtained at or below cod~owabte values foi material destan a'trees, 1.97-8

0

TABLE 1 (Continued)

Vrrlrblr TYPE 8 (Continued)

Range Category (see Regulatory Position 1r3)

~Pur orr Drywell Sump Level Bottom to top I unction detection'. acconlphshnlc>>t of mitigation; verification Maintaining Contamment Integrity Primary Containment Pressure IO psia to design prcssure 3 I'unction detection'. accomplishment of mitigation', verification Primary Containment Isola*

Closed-not closed

'ion Valve Position (exclud ing check valves)

Accomplishment of isolation TYPE C Variables: those variables that provide information to indicate the potential for being breached or the act>>al brcach of the bamers to fission product releases.

The barriers are ( I) fuel cladding, (2) pril>>ary coolant prcssure boundary.

an J (3)con-tainment.

Fuel Cladding Radioactivity Concentration or Radiation Level in Circulating Primary Coolant I/2 Tech Spec limit to 100 times I

Tech Spec limit. R/hr Detection of breach Analysis of Primary Coolant (Gamma Spectrum) 10 pCi/gm to 10 Ci/gm or TID-14844 source term in coolant volume 34 l)etail analysis: accomplishnlcnt uf mitigation; verification: Iong ter>>i surveillance BWR Core Thermocouples 200'F to 2300 F To monitor core cooling.

Reactor Coolant Pressure Boundary RCS Pressure 15 psia to 1500 psig Detection of potential for or;lct>>ul breach: accomplishl>>cnt nf >>>>tipu-tion'. long-term surveillance Primary Containment Area Radiation I R/hr to 10 R/hr Detection of breach; verification Sampling or monitoring of rsdiosctive liquids snd gases should be performed in e manner that ensures procuremeot or represcnrsdve samples.

For gsses, the criteria of ANSI NI3.t should be applied. For liquids, provisions should be msde for sampling from wcU mixed turbo lent zones, snd ssmpting lines should be designed to minimize plsleout or deposition. For safe snd convenient sampling, the provisions should include:

s. Shielding to maintain radiation doses ALARA,

b. Sample contstnsrs with eontsbler<<smpthlg port connector compstibUity,

c. CspsbUlty ofsampling under primary system pressure snd negative pressures,

d. Hsndtlng snd transport espsbUlty, and

e. Prearrangement for analysis snd interpretation.

gThe maximum value msy be revtged upward to satisfy ATWS requirements.

gttntmum of two monitors at widely separated locations.

Iletectols should respond to gsrams radiation photons wlthln any ensrgl range from 60 keV to 3 Mev with sn energy response accuracy of ago percent at sny specific photon <<nergy from Lt gte Y to 3 itsY. Overall.system accuracy should be within ~ factor of 3 over the entire ranger 1.97-9

TABLE 1 {Continued)

Variable TYPE C {Continued)

Range Category {see Regulatory Position 1.3) brurnore Reactor Coolant Pressure Boundary {Continued)

'Drywell Drain Sumps Level 2

{Identified and Unidentified Leakage)

Bottom to top Detection of brcach: accomplishment of mitigation; verification; long-term surveillance Suppression Pool Water Level Drywell Pressurca Bottom of ECCS suction line to 5 ft above normal water level 0 to design pressure (psig) 3 Detection of brcach'a accomplishment of mitigation; verification; long-term surveillance Detection of breach: verification Containmcnt RCS Prcssure Primary Containment Prcssure Containment an J Drywell Hydrogen Concentration 15 psia to 1500 psig 10 psia pressure to 3 thnes design pressure for concrete; 4 times design pressure for steel 0 to 3(% (capability of operating from 12 psia to design pressure

)

ls Detection of potential for brcach; accomplishtncnt of initigation Detect>on of potent<el for or actual breach; accomplishment of mitiga.

tion Detection of potential for breach; accomplishment of mitigation Containment and Drywcll Oxygen Concentration {for inertcd containcncnt plants)

Containmcnt I'.fffuent Radio-activity - Nnhlc (lascs (from identified release points in<<lud.

ing Standby (bas Treatment System Vent) 0 to 10%(capability of operating from 12 psia to design pressure

)

e 10 pCi/cc to 10 pCi%c 38e9 Detection of potential for brcach; accomplishment of mitigation Detection of actual breach; accom-plishment of mitigation: verifica-tton Radiation Exposure Rate (in-side buildings or areas, c.g.,

auxiliary building, fuel hand-ling building, secondary con.

tainment, which are in direct contact with primary con-tainment where penetrations and hatches are located) 10ot R/hr to 10 R/hr indication of brcach pcovtslons should be made to monitor elt fdcntfAedgsthways fot teteese of Ssseous ta fo<<ttve m d

et~'o the environs fn conformance

~

~

to release from a commo n dfsch pol t e

n 4, fbfonitorfn of fndtvfdu eAluent streams ts only cequtted where such streams ate ease ec n

envfconment. ff two or mote stre ms o

p constdeced to meet the Intent of this ceautatoty Su e pro e

su m

Monttots should be capable of deteetfna and measutln cedfoaetfve us efAuent coneentcatfon t

us efAuent concentrations wtth eomposttfons censfna from fresh

~anllrbdern noble aee Irnroo orodllel nlrerroee lo !Iederorrf olrlllnrea, efrb oeereFerereor oeeoreeree edibrn ~ rearer o a.

Ilare II

..wftt have sufAelent tanae to encompass the entice ranee pto

~

tfons msy be ea pressed fn terms of Xeo Vtd d ln thta Ceau.

tOCy aufde and that mu.tfPte eemoedbente OC Syeteme WN be

~-133 ulvalents ot ln terms of any nobloe pa nuetfde(s). 1t ts not capered that ~

e mon S

cesu.s o

deal cat{n

~ needed. Exfstlna equfpmentcney be used to cnonltot any pottlon of theetated canae wlthln the equipment Sn tat{na.

1.97-10

TABLE 1 (Continued)

Vsriaisl ~

~Ran Category (see Regulatory Position 1.3)

Purpose TYPE C (Continued)

Containment (Continued)

EfAuent Radioactivity

- Noble I 0 pCi/cc to I 0 pCi/<<<<

Ceases (from buildings as indicated above) 29 Indication of breach TYPE D Variables: those variables that provide ini'ormatin;l to indicate thc operation of individual safety systems and other systems ilnportant to safety. These variables arc 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 Fcedwater System Main Fccdwater Flow 0 to I I0% design flnw'etection of operation:

analysis of cooling Condensate Storage Tank Level Bottom to top Indication of available water I'or cooling Primary Containment-Related Systems Suppression Chamber Spray Flow 0 to I I IV'esign Aow To monitor operation Drywell I'ressure I 2 psia to 3 psig 0 to I IO'7o design pressure To monitor operation Suppression Pool Water Temperature 30 F to 230 F Suppression Pool Water Level Top of vent to top of weir well To monitor operation To monitor operation Drywall Atmosphere Tenlpcr:ltufc Drywell Spray Flow 40'F to 440'r:

Oto>>or dcslgnAowto To nlonitor operation To nlonltor operation Main Steam System Main Stcamline Isolation Valves'eakage Control System Pressure 0 to 15" nf water Oto 5psid To provide indication of prcssure boundary maintenance Primary System Safety Relief Valve Positions, Including ADS or Flow Through or Prcssure in Valve Lines Closed-not closed or 0 to 50 psig Detection of accident; boundary integrity indication I.97-11

TABLE 1 (Continued)

Vaiiable TYPE D {Continued)

Safety Systems Isolation Condenser System Shell-Side Water Level

~Ran Top to bottom Category {see Regulatory Position 1.3)

Purpose To monitor operation isolation Condenser System Valve Position RCIC Flow HPCI Flow Core Spray System Flow LPCI System Flow SLCS Flow SLCS Storage Tank Level Open or closed 0 to I I(% design flow 0 to 110% design flow 0 to 110% design flow' to 110% design flow' to I l(yk design flow Bottom to top Tn mnnitor status To mnnitor operation To monitor operation To monitor operation To monitor operation To monitor operation Tn Illollltoropcfatioli Residual Heat Removal {RHR)

Systems RHR System Flow RI IR Iieet Exchanger Outlet Temperature 0 to I I(YYr design flow 32 F to 350I'o monitor operation To monitor operation Cooling Water System Cooling Water Temperature to FSF System Components Cooling Water Flow to I'.SF Systcin Components 32 Fto200 F

0 to I I0% designflow'o monitnr operatinn Tn monitor operation Radwaste Systems High Radioactivity Liquid Tank Level Top to bottom To monitor operation Ventilation Systems Emergency Ventilation Damper Position Opcnwlosed status To monitor operation Power Supplies Status of Standby Power and Voltages, currents, pressures Other Energy Sources1mportant to Safety (hydraulic, pneumatic) 2l l To monitor system status llStatus bidlcatlon of att Standby Rower ac. busaa, dw. buaea, Iovarter output buses, and pneumatic supplies.

1.97-12

TABLE 1 (Continued)

TYPE E Variables: those variables to bc inonitorcd as rcquircd for usc in determining the magnitude of the release of radio-active materials and continually assessing such releases.

Variable Range,.

Category (see Regulatory Position 1e3)

Purpose Containment Radiation Primary Containment Area Radiation

~ High Range 2 Reactor Building or Secondary Containment Area Radiation I R/hr to I 0 R/hr IO'/hr to 10 R/hr for Mark I and H containments I R/hr to 10 R/hr for Mark Hl containment I rae7 q9

) re,7 Dctcction of significant releases:

rcleasc assessment:

long.teri>>

survcilhncc: emergency plan actuation Detection of significant releases:

release assessmcnt:

long.terin surveillance Area Radiation Radiation Exposure Rate (inside buildings or areas where access is required to service equipmcnt important to safety) 10'/hr to IO R/hr Detection of significant rclc:iscs:

release assessment:

Iong-tcrm survcilhnce Airborne Radioactive Materials Relcascd from Plant Noble Gases and Vent Flow Rate

~

Drywell Purge, Standby Gas Treatment System Purge (for Mark I and ll plants) and Secondary Contain-ment Purge (for Mark Ill plants) 10 pCi/cc to 10 pCi/cc 0 to 110% vent design flow (Not needed ifeffluen discharges through common plant vent) q9 Dctcction of signiRbant relca~s:

release aswssmcnt

~

Secondary Containment Purge (for Mark I, II, an J Hl plants)

~

Secondary Containment (reactor shield building annulus, ifin design)

~

Auxiliary Building (including any building containing primary system gases, e.g., waste gas decay tank) 10 pCi/cc to 10 pCi/cc 0 to 110% vent design liowt 0 (Not needed ifcfAuent Jisi barges through common plant v<<nt) 10 pCi/cc to 10 pCi/cc 0 to 110% vent design Anw'Not needed ifcfAuent discharges through common plant vent) 10 pCi/cc to 10 pCi/cc 0 to Ilib vent design flow (Not needed ifeffluent discharges through common plant vent) g9 29 Detection of significant r<<lease

~

release asscssnlcnt Detection of significant releases:

release assessment Detection of significant releases:

rclcase assessment; long-term surve0lance 10~ pCi/cc to 10 pCI/cc 0 to 11'ent design flow'0+

pCI/cc to 10 pCI/cc

~

Common, Plant Vent or Multi-purpose Vent Discharging Any of Above Releases (if drywell or SGTS purge Is included)

Detection of significant releases; rclcase assessment; long.term aurvcilhnce 1.97-13

(

Variable Range Category (see Regulatory Position ).3)

TABLE 1 (Continued)

Purpose TYPE E (Continued)

Airborne Radioactive Materials Released from Plant (Continued)

Noble Cases and Vent Flow Rate (Continued)

AIIOther Identified Release Points IO e pCi/cc to IO p('i/cc 0 to I l0% vent design flow'Not needed ifeff)ucnt discharges through other monitored plant vents) e9 I)etection of significant releases:

release assessment; long-tcrnt surveillance Particulates and Halogens

~

All Identified Phnt Release 10 pCi/cc to I 0 pCi/cc Points. Sampling with Onsite 0 lo I )On vent design AowIo Analysis Capability 3I2 l)etection of significant rck'uses; release assessment:

long.term surveillance Enviions Radiation and Radio-activity Radiation Exposure Meters (continuous indication at

'ixed locations)

Range, location, and qualifica-tion criteria to be developed to sa tisfy N U R EC-0654,Section II.H.Sb and 6b requirements for emergency radiological monitors Verify significant releases and local magnitudes Airborne Radiohalogens and Particulates (portable sampling with onsite analysis capability)

Plant and Environs Radiation (portable instrumentation)

IO pCi/cc to IO pCi/cc IO R/hr to IO R/hr, photons IO rads/hr to IO rads/hr, beta radiations and Iow~nergy photons 3l 3 3t4 3l4 Release assessment:

analysis Release asscssmcnt:

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

Multichannel gamma ray spectrometer Release assessment:

analysis l1To provide information regarding release of cs Jtoactlve halogens and parti

~

f hl ldi tsndting, any analytical a d articulates. Continuous collection of representative samples followed esi n liow n av r S

o ce lc tion of tO~ pCl/ccof csdloiodincs oh o ens snd articulates. The design envelope or s c

ng, )

Cl/

f ttcutate cadloiodlnes and pacticulatesothec chan cadioiodines,and an gaseous or vapor form, an average concentration of lo pcl/cc o part cu ace ca o o n

average gamma photon energy of 0.$ MeY pec disintegration.

For esdmatlng release rates of radioactive materials celeased ductng an scddent.

To monitor radiation and atcb~o

~ csdloacttvlty concentrations tn many areas throughout the fadttty and the site environs where lt ts Impractical to install stationary monlt capable ofcovering both normal and scddent levels.

1.97-I 4

TABLE 1 (Continued)

Variable TYPE E (Continued)

Meteorology' Range Category iree Regulatory Position 1.3)

Purpose Wind Direction 0 to 360

(+Q accuracy with a aellection of 15 ). Starting speed 0.45 mps (1.0 mph). Damping ratio between 0.4 and 0.6, distance con-stant

< 2 meters RclcJsc asscssmcnt Wind Speed 0 to 30 mps (67 mph) +0.22 mps 3

(0.5 mph) accuracy for wind speeds less than 11 mps (25 mph) with a starting threshold of less than 0.45 mps (1.0 mph)

Release assessmcnt Estimation of Atmos.

pheric Stability Based on vertical temperature difference from primary system,

~5'C to 10 C (-9 F to 18 F) and

%.15 C accuracy per 50-meter intervals (%.3 F accuracy per 164-foot intervals) or analogous range for alternative stability estimates Release assessment Accident Sampling'apa-bility (Analysis Capabil-ity On Site)

Primary Coolant and Sump Grab Sample 3 'elease assessmcnt; verification; analysis

~

Gross Activity

~

Gamma Spectrum

~

Boron Content

~

Chloride Content

~

Dissolved llydrogen or Total Gas' Dissolved Oxygen '

pH Containment Air

~

Hydrogen Content

~

Oxygen Content

~

Gamma Spectrum 10 IrCI/ml to 10 Ci/ml (Isotopic Analysis) 0 to 1000 ppm 0 to 20 ppm 0 to 2000 cc(STP)/kg 0 to 20 ppm I to 13 Grab Sample 0 to 10%

0 to 30% for inerted containments 0 to 30%

(isotopic analysis) 34 Release assessm en t; verificat ion; analysis Guidance on meteorotostcaa measurements ls betng developed ln a Proposed Revlslon l to Regulatory Oulde l Qg "Meteorological Programs ln Sup port of Nuclear Power Plants."

ldThe time for tatdnx and anstyztng samples should be S hours or less from the ttme the dectston Is made to sample, except for chlortde whish should be wlthtn Re%ours.

t7An tnstslled capsbglty should bb provided for obtatntng contslnmsnt sump, ECCS pump room sumps, and other stmttst auxNsry buUdlng sump ttqutd samples.

14

'ppttee only to prtmary coolant, pot to s'p 1.97-15

(

TABLE 2 PWR VARIABLES TYPE A Variables: those variables to be monitored that provide thc primary information required to permit the control room operator to take specific manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis accident events.

Primary information is informa-tion that is essential for the direct accomplishment of the specified safety functions: it does not include those variables that are associated with contingency actions that may also be idcntificd in written procidorcs.

A variable included as Type A docs not preclude it from being included as Type B. C, D. or I'. or vice versa.

Verieble Range Category (see Regulatory Position le3)

Purpose Plant specific Plant specific Information required for operator action TYPE B Variables:

those variables that provide information to indicate whether plant safety functions arc being accomplished.

Plant safety functions are (I) reactivity control. (2) core cooling, (3) maintaining reactor coolant system integrity. and (4) maintaining containment integrity (including radioactive efAuent control). Variables are listed with designated ranges and category for design and qualification requirements. Key variables are indicated by design and qualification Category I.

activity Control Neutron Flux IO "A)to I00% full Dow@r I'unction detection: accomplishment of o>itigation Control Rod Position RCS Soluble Boron Concen-tration Full in or not full in 0 to 6000 ppm Verification Verification RCS Cold Leg Water Tcmper-atut'e 50 Fto400 F

Verification Core Cooling RCS Ho! Leg Water Temper-ature 50 F to 750 F I'unction detection'accomplishment of n>itigation; verification: long-term sun eillance RCS Cold Leg Water Temper-50 F to 750 F atureunction detection: accomplishment of mitigation: verification; long-term surveillance RCS Pressur 0 to 3000 psig (4000 pslg for CE plants)

Function detection; accomplishment of mitigation; veriBcatlon', long;term aurve Blanco Where a vartabte Ia tlatod for more than one purpoee, the tnstrumcatatton requtrementa may be Iategratednd anty one measurement prcPPlded.

The maximum value may be revtaed upward to aattafy ATWS requtrementa.

1.97. I6

TABLE 2 (Continued)

Virilbl~

Range Category (see Regulatory Position 1.3)

TYPE 8 (Continued)

Core Cooling {Continued)

Core ExitTemperature'00'F to 2300 F (for operating plants-200 F to l650 F) 33 Verification Coolant Level in Reactor Degrees of Subcooling Bottom of core to top of vessel 200 F subcooling to 35 F superheat I

(Dircct-indicating or recording device not needed)

(With con.

firmatory operator procedures)

Verification; accomplishment of mitigation Verification and analysis nf plant conditions Maintaining Reactor Coolant System Integrity RCS Pressure' to 3000 psig (4000 psig for CE plants)

Function detection: acconlpllsllnlcnt of mitigation Containment Sump Water Level'ontainment Pressure'arrow range (sump),

Wide range (bottom of contain-ment to 600,00~allon level equivalent) 0 to design pressure (psig)

Function detection; accolnplishlncnt of mitigation'; verification Function detection: acconrplishmcnt of mitigation: verification Maintaining Containment Integrity Containment Isolation Valve Closed-not closed Position (excluding check valves)

Accon>plishntent of isolation Containment Pressure 10 psia to design pressure Function detection: accomplishment of mitigation; v<<rification 3A minimum of four measurements per quadrant Is required for operation. SuNclent number should be Installed to account for at trttton.

(Replacement b>strumentatlon should meet the 2300 F range provtalon.)

lDesign pressure ia that value corresponding to ASME code values that an obtatned al or below cod~owabte values for material design stress.

I.97-I 7

TABLE2 (Contlnu~}

lsrcacii of r thc actual tircai'

'icate the potential for be g

boundary,and(3)con.

variables t at P fuel chddiilg, TYPE C Yariables: those h

ovide informatio roduct releases.. The barriers are (I) the bamers to fission pro tainment.

Variable Range Category (sae Regulatory Position 1.3}

Purpere Fuel Cladding ure'ore Exit Temperature Radioactivity Concentration or Radiation Level in Circulating Primary Coolant 200 F to 2300 I'foroperating plants-200 F to 1650 F)

I/2 Tech Spec limit to it to 100 times Tech Spec limit. R/hr 13 otcntial for brcach:

Detection of po c accomplishment nf mitigation', o term survcilhncc Detection ot'reach Analysis of Primary Coolant (Ciamma Spectrum)

Reactor Coolant Pressure Boundary RCS Pressure IO pCi/gm to 10 Ci/gm or TlD-14844 source term in coolant volume for CE 0 to 3000 psig (4000 psig fo plants) 3$

12 Detail ana)ysis; accomp 'lishment of mitigation; t'on verification; long-tcrni surveillance of otcniial for or actual Detection o p

breach; accomplishment o

tion: long-term survcicilhnce Containment Prcssure

ia to design pressure p ig (5 psia for subatmosp containments) t of breach: accornplishmcnt Detectiono:

hshmcnt o mitiga t ation'erification; on

~

surveillance Containment Sump Water Level i iationt Containment Area Radiation Effluent Radioactivity-- Noble Gas Effiucnt from Condenser Air Removal System Exhaust Narrow range (sump),

Wide range o

(b ttom of containmen to 600.00~21 level equivalent)

I R/hr to 10 R/hr 10 pCi%c to 10*

I/Ci/cc 34,'r 38 D

tion of brcach; acc p

'm lishment

'ong-term 0 rill ig e

't ation: verification; rveillance sil f brcach: verification Detection o r

Detection of brcac:i'erification urcmcnt of rcprcscniativc a insane ih i eos cs p o a

d ascs shou c

td be made for sampling f om ons shou thi. ihc prov s o o

of ta os For tlqiitds. provtsl a of ANSt HI3.t shoiit

n. Forsa can be designed io mlnlmtse p a ling linea should be tones,andsamp ng coiitahictmmpting port c Ca ability of sampling Handling and traiisport cage n ement for anaty an D

tespolisc acciiracy 0'.

Pteartangem dct sepata ei d tocattons.

th ao 6D heV to $ MeV&thm enerP ihe entire range 4

of two monitors at bvt y

thin any energy tange from ctoto 2ovcs Minimum o

.t Vto 3 Me stone taiiglng tom f

m free Detectoca should tespon of 2. ENuent conccna cent at any specific p oo d

t t

ctee wtthtiia factor 0 e

a2D parce tecttng and measutln tadioae tIase tutee to 10ctayeot

mixtures, t

braientsortn termsofa yn to encompass the en vcttt have sufhcient tange needed. Existiog equtpmen m

1.97-1 g

(

Veri eble TYPE C (Continued)

Containment TABLE2 (Continued)

Category (see Regulatory Position 1.3)

PU epee ~

RCS Pressure' to 3000 psig (4000 psig for CE plants)

Detection of potential for breach; accomplishment of mitigation Containmcnt Hydrogen Concentration 0 to 10% (capable of operating from 10 psia to maximuin design pressure

)

0 to 30% for ice-condenser-type containment Detection of potential for hrcacll; accomplishment of mitigation; long-tern) surveillance Containment Pressure'0 psia pressure to 3 times design I

pressure for concrete; 4 times design pressure for steel (5 psia for subatmospheric containm<<nts)

Detection of potential for or'actual brcach; accomplishment of mitiga.

tion Containment Effluent Radio-activity - Noble Gases from Identified Release Points Radiation Exposure Rate (

side buildings or areas, e.g.

in-auxiliary budding, reactor shield building annulus, fuel handling building, which are in direct contact with primary containment where penetra-tions and hatches are located)ffluent Radioactivity' Noble Gases (from buildings as indicated above) 10 pCi/cc to 10 pCi/cc 10 R/hr to 10 R/hr 10 pCi/cc to 10 pCi/cc ya,e ga Detection of brcach; accomplish.

ment of mitigation; verification indication of breach indication of breach TYPE D Variables: those variables that provide information to indicate thc operation of inrllvulual safety systcius and otlicr systems important to safety. These variables are to help the operator nuke appropriate decisions in using thc 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 Outlet Temperature 0 to 110% design flowio 32 F to 350 F To monitor operation To monitor operation and foranalysis Ptovtstoas should be made tn mollttot att Idcattfted pathways for telcase of gaseous tedlosctlve matcrtcts to the cnvtroas ln conformance with General Dcslsa crttcttoa di. stoattorlng of Individual eNueat streams Is only requited where such sttcsms ste released directly taro the eavttonmcnt, If two or more stresras are combined prior to tclcsso from a common discharge potnt, moaltortaa of the combined stresm Is coastdcted to meet the Intent of this regulatory guide ptovtded such monitoring hes ~ tease adequate to measure worst~ releases Dost!a ftowIs the maximum ftrrwantid pated In aormst opcrettoa.

1.97-1 9

TABLE 2 (Continued)

Variable Range Category (see Regulatory Position 1e3)

Porpore TYPE 0 (Continued)

Safety Injection Systems Accumulator Tank Level and Prcssure Accumulator Isolation Valve Position l0% to 9(yk volume 0 to 750 psig Closed or Open Tn monitor operation Operation status Boric Acid Charging Flow Flow in HPI System Flow in LPI System Refueling lyatcr Storage Tank Level 0 to l10% design flow' tn I I0% design flow 0 tn I l0% design flow'np to bottom To monitor operation To monitor operation To monitor operation To monitor operation Primary Coolant System Reactor Coolant Pump Status Primary System Safety Relief Valve Positions (including PORV and code valves) or Flow Through or Pressure in Relief Valve Lines Motor current Closed-not closed To monitor operation Operation status; to monitor for loss of coolant Pressurizer Level Bottnm to top To ensure proper operation ni'ressurizer Pressurizer Heater Status Quench Tank Level Quench Tank Temperature Quench Tank Pressure Electric current Tnp tn bottom 50 I:to750 F 0 to design prcssure To determine operating status Tn monitor npcratinn To monitnr npcratinn To monitor operation Secondary System (Steam Generator)

Steam Generator Level Steam Generator Pressure Safety/Relief Valve Positions or Main Steam Flow Main Feedwater Flow From tube sheet to separators From atmospheric prcssure to 20% above the lowest safety valve set ting Closed-not closed 0 to i)Ifdesig flow" 2

To monitor operation To monitor operation To monitor operation To monitor operation I.97-20

TAB!.E 2 (Continued)

Varlabh TYPE 0 (Continued)

Category (see Regulatory Position 1s3)

Purpose Auxiliary Feedwater or Emer-gency Feedwater System Auxiliaryor Emergency Feed-0 to 110% design flow'ater Flow 2

(I for Bh,W plants)

To monitor operation Condensate Storage Tank Water Level Plant specific To ensure water supply for auxiliary feedwater (Can be Category 3 ifnot primary source of AFW. Then what-ever is primary source of AFWshould be listed and should be Category 1.)

Containment Cooling Systems Containment Spray 1'low 0 to 110% design liow'eat Removal by the Contain-

~ Plant specific ment Fan Heat Removal System Tn iiinnitoroperation To ninnitor operation Containment Atmosphere Temperature Containment Sump Water Temperature 40 F to 400 F 50'F to 250 F To indicate acexnplishnicnt ofcooling To monitor operation Chemical and Volume Control System Makeup Flow - ln Letdown Flow - Out Volume Control Tank Level 0 to 110% design flow' to 110% design flow Top tu hottoi>>

To ninnitor operation To monitor operation To nionitor operation Cooling Water System Component Cooling. Water Temperature to ESF System 32 F to 2001'o monitor operation Component Cooling Water Flow 0 to 110% design flow to ESF System To monitor operation Radwaste Systems High-Level Radioactive Liquid Tank Level Top to bottom To indicate storage volume Radioactive Gas Holdup Tank Prcssure 0 to l50% design prcssurce To indicate storage capacity 1.97-21

TABLE 2 {Continued)

Yarisbls Range Category {see Regulatory Position 1.3)

Purpose TYPE D {Continued)

Ventilation Systems Emergency Ventihtion Damper Open~losed status Position To indicate damper status Power Supplies Status of Standby Power and Other Energy Sources Import-ant to Safety (hydraulic, pneumatic)

Voltages, currents, pressures 2ll To indicate system status TYPE E Variables: those variables to be monitored as required for use in determi l

gn'nin'hc ma itude of the release of radio-active materials and continually assessing such releases.

Containment Radiation 7

Containment Arcs Radiation-1 R/hr to 10 R/hr 1 lighRange'6,7 Detection of signiflicant releases; release assessment; long-term:-

surveillance: emergency plan actuation Area Radiation Radiation Exposure Rate I

{inside buildings or areas where access is required to service equipment important to safety) lO t R/hr to 10 R/hr 21 Detection of significant releases; release assessment; long;term surveillance Airborne Radioactive INaterlals Released from Plant Noble Gases and Vent Flow Rate

~

Containment or Purge Effluent' Reactor Shield Building Annulus'if in design) l0 uCi/cc to l 0 uCi/cc 0 to 110,n vent design Anwio (Not needed ifefAuent discharges through common plant vent) l0+ uCi/cc to 10 uCi/cc 0 to ) 10% vent design Aowlo (Not needed ifeffluent discharges through common plant vent) 28 2d Dctec'tion of significant releases:

release assessmcnt Detection of significant releases:

release assessment

~

Auxiliary Buildingt

{includingany building containing primary system gases, c.g., waste gas decay tankl 10+ Qi/cc to 10S tiCi/cc 0 to 110% vent design flowto (Not needed ifeffluen discharges through common phnt vent) 2$

Detection of'ignificant releases; release assessment; long-term surveillance Status tndteatton of att Standby power Lc, buses d~ b~ tnvertot output blN44 and pneoinattc wppttes 1.97-22

TABLE2 (Continued)

(

Vrllmbl~

Type E (Continued)

Range Category (see Regulatory Position 1.3)

Airborne Radioactive Materials Released from Plant (Continued)

Noble Gases and Vent Flow Rate (Continued)

~

Condenser Air Removal System Exhaust 10 e pCi/cc to 105 pCi/cc 0 to 110% vent design flow (Not needed ifeffluent discharges through common plant vent) 25 Detection ol'igniiflicant releases:

release assessment

~

Common Plant Vent or Multi-purpose Vent Discharging Any of Above Releases (if containment purge is included)

~

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

~

AllOther Identified Release Points 10 pCi/cc to 10 pCi%c 0 to 110% vent design flow'0+

pCi/cc to 10 pCi/cc 10 'Ci%c to 103 pCi/cc (Duration of releases in seconds and mass of steam per unit time)

I 0 pCi/cc to 10 pCi/cc I) to 110% vent design flow fNot needed ifeffluent discharges through other monitored plant vents) 212 qg Detection of sigruficant releases; release assessment; long-term surveillance Detection of significant releases:

release asscssmcnt Detection of significant releases; release assessment; long-term surveillance Particulates and Halogens

~

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

Sampling with Onsite Analysis Capability I~ 3 pCi/cc to 10 pCi/cc 0 to 110% vent designflow'I 3

Detection of significant releases; release assessment; long-term surveillance Effluent monitors for PWR steam safety valve discharges and atmospheric steam dump valve dhchargcs should be capable of spproxi ~

msiciy Ihcsr response to gamma tsdhilon phoions with energies from approximately 0.5 McV to 3 McV. 8vcrail system accufscy should be wtrhtn s factor of 2. Calibration sources should fall within the range of approximately 0.$ McV io I.S McV (e.g., Cs-l37, Mn.ss. Ils 22. snd Co<0). Efllucnt conccntations should bc expressed ln terms of any gsmmaomittlng noble gas nuclide within the speclflcd energy ange. Calcu-hilonsl methods should be provided for estimating concurrent releasei of lomncrgy noble gases that cannot be detected or measured by the methods or techniques employed foc monitoring.

To provide Information regarding'release ofadloacttve halogens and pattlcuhtas. Continuo wcottecrton ofcepresentatlve samples followed by onslte hboarory measurements of umplcs for radlohalogens and pact cuhtes The design envelope for shielding. )andUng> and analytical cs should assume 30 minutes of (ntcgtared umptlnq tlm>> at samplet design flow, an average concentrarton of l0 pCI/cc of radlolodlnes gaseow or vapor form, an average concentration of 10 pCI/cc of pattlculate cadloiodlnes and parttcuhtes other than cadlolodlnee, and an aveage gamma photon energy of 0,5 Mnty pec dhlntegratlon.

1.97-23

i TABLE 2 (Continued)

Variable TYPE E (Continued)

Range Category (see Regulatory Position 1.3)

Environs Radiation and Radio.

activity Radiation Exposure Meters (continuous indication at fixed locations)

Airborne Radiohalogens and.

Particulates (portable sampling with onsite analysis capability)

Range, location, and qualifica-tion criteria to be developed to satisfy NUREG%654, Section 11.H.Sb and 6b requirements for emergency radiological monitors 10 pCi/cc to 10 5 pCi/cc 3 le Verify significant releases and local magnitudes Release assessmbnt; analysis Plant and Environs Radiation (portable instrumentation) 10 R/hr to 10 R/hr, photons 10 rads/hr to 10 rads/hr. beta radiations and lowwnergy phntnns 1t 5 Release assessment,

~raly:i::

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

Meteorology I d Multichannel gamma-ray spectrometer Release assessment; analysis Wind Direction 0 to 360

(+D accuracy with a deflection of 15 ). Starting spe<<d OAS mps (1.0 mph). Damping ratio between 0.4 and 0.6, distance con-stant < 2 meters Release assessment Wind Speed 0 to 30 mps (67 mph) +0.22 mps (0.5 mph) accuracy for wind cpecds less than 11 mps (25 mph) with a starting threshold of less than 0.45 mps (1.0 mph)

Release assessment Estimation of Atmos-pheric Stability Based on vertical temperature difference from primary system.

-5 Cto 10 C(9 F to 18 F) and

%.15 C accuracy per 50-metcr intervals (+0.3'F accuracy per 164-foot intervals) or analogous range for alternative stability estimates Release assessment 14For eethcetha release rotoe of redloeettvo materials released durha an ecetdeat.

To monitor redtetton and etrborno redloecttvlty eoncentratlone h many crees throuahout the focNIty and tho alto ecLvlrone where It Is 15

~precttcet to Inetett etettoaery monitors capable ofcoverhg both normal andpectdent Ievote.

ldauMenee on meteorolostcal meeeuremento Ie behg developed h a Propoood Revteha 1 to Regulatory Outdo 1.i3, "M<<toocolostcol Programs h Su pport of Nucteer Power Ptente."

hl t/, ~

> ~

~ >+Le 'I+i pic 4y '

~

'i I'a

>I i4

'.97-24

TABLE 2 (Continued)

Variable TYPE E (Continued)

Range Category (see Regulatory Poaition 1e3)

Purpose Accident Sampling'apa bility (Analysis Capahil.

ity On Site)

Primary Coolant and Sulllp Grab Sample ps. I a Release asscssmcnt: vcrificatlon.

analysis

~

Gross Activity

~

Gamma Spectrum

~

Boron Content

~

Chloride Content

~

Dissolved Hydrogen or Total Gas' Dissolved Oxygen

'H Cnntainmcnt Air

~

Hydrogen Content

~

Oxygen Content

~

Gamma Spectrum 10 ttCi/ml to l0 Ci/ml (isotopic Analysis) 0 to 6000 ppm 0 to 20 ppm 0 to 2000 cc(STP)/kg 0 to 20 ppm l to 13 Grab Sample 0 tn 10~so 0 to 30,e for ice condensers 0 to 30%

(iso topic analysis) 3s Release asm ssmen t: vcr i fibation:

analysis Thetl cf t td and al dn The tlmc for tatdng and anatyatna aamplea should be 3 houra or leaa from the time the decision la made to aamptep except for chtortde taAn InataUed capabQlly ahould be rovtded for obtalnln con p

for obtalntna contatnment aump. ECCS pump room aumpa, and other atmQar auatQary l9Appllea only to primary coolant, not to sump.

).97-25

VALUE/IMPACTSTATEMENT

1. PROPOSED ACTIOV.:-

\\

1.1 Description The applicant for a license (or licensee) of a nuclear power plant is required by the Commission's regulations to provide instrumentation to (I) monitor variables and systems over their anticipated ranges for accident conditions as appropriate to ensure adequate safety and (2) monitor the reactor containment atmosphere, spaces containing components for recirculation of Io~f-coolant accident fluid, effluent discharge

paths, and the plant environs for radioactivity that may be released from postulated accidents.

This revision to Regulatory Guide 1.97 proposes to improve the guidance for plant and environs monitoring during and following an accident, including extended ranges for some instruments to account for consideration of degraded core cooling.

1.2 Need action will establish an NRC position by taking advantage of previous staff effort (I) in completing a generic activity (A-34), (2) in evaluating the lessons learned from the TMI.2 event (NUREG-0578), and (3) in conjunction with effort in developing a

national standard (AN&4.5). For future

plants, the staff review will bc simplified with guidance contained in the endorsed standard developed hy a voluntary standards group and in the regulatory guide, which includes a list of variables for accident monitoring. Efforts by thc staff to implement Revision I to Regulatory Guide 1.97 have been fraught with frustration and mct with delays because the guide was adjudged by licensees to bc vague'nd ambiguous. Revision 2 eliminates the problems encoun-tered with Revhion I because it provides a minimum set of variables to be measured and hence gives more guidance in the selection of accident-monitoring instrumentation.

Consequently, there will be no significant impact on the staff. There will,however, be effort required to review each operating plant and each plant under rcvicw to assess

onformance with Regulatory Guide 1.97.

Regulatory Guide 1.97 was issued-as an effective guide in August 1977. At the time the guide was issued, it was recognized that more specific guidance than that contained

'n the guide would be required.

However, the difficulty developing the guide to the point where it could be atitially issued was evidence that experience in using the guide as it then existed was essential before further develop-ment of the guide would be meaningful.

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

When the staff was ready to issue the results of the Task Action Plan A-34 effort, the accident at TMI-2 occurred.

Subsequently, the TMI-2 Lessons Learned Task Force has issued its "Status Report and Short-Term Recommendations,"

NUREG<578.

This report, along with the draft Task Action Plan A-34 report, Draft I of Regulatory Guide 1.97 (dated April 12, 1974), and Standard AN&4.5, provides ample basis for revising Regulatory Guide 1.97.

1.3 Value/Impact of Proposed Action I.3.1 NRC Operations Since a Hst of aelocted variable to be provided with txumentatlon to bo monitored by the plant operator luring and following an occident has aot been oxpHcitly agreed to h tho.past, the proposed action should zemtlt In moro effective effort by the stiffin reviewIng appHcations forconstruction permits and operating Hcetges. Tho proposed I.3.2 Other Gorerntncnt Agencies Not applicable, unless the government agimcy is an applicant.

I.3.3 Industry <

The proposed action establishes a more clearly defined NRC position with regard to instrumentation to assess plant and environs conditions during and following an accident and therefore reduces uncertainty as to what the staff considers acceptable in the area of accident monitoring.

Most of the impact on industry will be in the area of providing instrumentation to indicate the potential breach and the actual breach of the barriers to radioactivity release, i.c., fuel cladding, reactor coolant prcssure boundary.

and containment.

Some instruments have extended ranges and others have higher qualification requirements.

There will be additional impact duc to heretofore unspecified variables to be monitored (i.e., water level in reactor for PWRs and radiation level in the primary coolant water for PWRs and BWRs) that have been identified during the evaluation of TMI-2experience and willrequire development.

Attempts were made during the comment period to determine the cost impact on industry for future plants and for backfitting existing plants.

Estimates ranged from

$4,000,000 to over

$ 20,000,000.

The higher estimates undoubtedly charged all accident-'monitoring instrumeritation to Revision 2 to Regulatory Guide 1.97. This should not be the caso.

The requirement for accident monitorin has always been o part of the sogulationa Consequently the Itnpact of Revision 2 to Regulatory Guido 1.97 should only be tho delta added by Revision 2. A consorvatIvo estimate of tho Incroaso in requirements aro the additions ofTypo C me~ments and tho upgrading of aomo of the Typo B l.97-26

measurements to higher qualification ofthc.instrumentation.

There are 17 unique Type B and C variables to be measured for PWRs, less for BWRs. A conservative average cost for each measure'ment ia $ 130,000 making a total cost impact of $ 2,210,000. lf +ifigure werc doubled to account for overhead costs and /fat a 15 percent contingency added, the cost impact w~d bc about

$5,000,000.

This cost estimate is the sama Tor operating phtnts as for plants under construction and future plants. While it is recognised that for operating plants the costs associated with backfitting are generally higher than the costs associated with new plants, some concessions are made in some requirements as a result of existing licensing commitments that bring the cost estimate to about the same value.

l.3.4 Public The proposed action will improve public safety by ensuring that the plant operator w01 have timely information to 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 on instrumentation to assess plant and environs conditions during and following an accident should be given.

2. TECHN1CAL APPROACH

4. STATUTORY CONS1DERATlONS 4.1 NRC Authority Authority 1'or this guide would be derived from thc safety requirements of the Atomic Energy Act. ln particular, Criterion 13, Criterion 19. and Criterion 64 of Appendix A to 10 CFR Part 50 require, in part, that instrumentation be provided to monitor variables. systems, and plant environs to ensure adequate safety.

4.2 Need for NEPA Assessment The proposed action is not a major action as defined in paragraph 51.5(aXIO) of 10 CFR Part 51 and docs nnt require an environmental impact statement.

5. RELATIONSH1P TOOTHER EXISTlNG OR PROPOSED REGULAT1ONS OR POLlC1ES No conflicts or overlaps with requirements promulgated by other agencies are foreseen. This guide does include the variables to bc monitored on site by the plant operator in order to provide necessary information for emergency planning. However, information on emergency pl~ing and its relationship to other agencies is provided +where.

implementation of the proposed action is disbhsscd in Section D of this revision.

This section is not applicable to this value/impact statement since the proposed action is a revision of an existing regulatory

guide, and there are no alternatives to providing the plant operator with thc required information.

3. PROCEDURAL APPROACH Previously discussed.

6.

SUMMARY

AND CONCLUS1ONS Revision 2 to Regulatory Guide 1.97. "instrumentation For Light-Water<ooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident." should be issued.

~

~

0, 1.97-27

0

CFF (1-'1-88 Edition) h on that warrant r

of the request.

otectton features accepted

.'taff In Fire Protectlpn iation Reports referred to n (b) of this section and to such reports.

other es covered by paragraph

ompleted as soon as prac.

)o later than the compte.

rently specified ln license r technical specifications cHity, or the date detr.

aragraphs (d)(1) through ih section, whichever ts m the Director of Nuclear Matlon determines, upon y the licensee, that there e for extending such date 1 pubHc health and safety sely affected by such ex-enslons of such date shall the dates determined by (c)(1) through (c)(4) of fire protection features revisions of sdminhtra-

, manpower changes, and

.1 be Implemented within 4

r the date of the NRC Protection Evaluation

~ptfn r requiring such

.'Ir tion features in-sl of modificatfons g

approval or plant all be Implemented within

,fter the date of the NRC

'rotectlon Safety Evalua-scceptlng or requiring s.fire protection

features, ternatlve shutdown cspa-ing installation of modffi-irfng plant shutdown shall nted before the startup arlfest of the f~llowing fencing 9 months or more

,.te of the NRC staff Fire Isfety Evaluation Report requiring such features:

,t refueling outage; er planned outage that

.ast 60 days: or planned outage that lasts 20 days.

.'Ire protection features In-mted shutdown capability

w buildings and systems

~or ilogufotory Commission Implemented within 30 s" th of NRC approval. Other modl-ttons requiring NRC approval prior

~H tsffatton shall be implemented within 8 months after NRC approval.

<e) Nuclear power plants licensed to

~rate after January 1.

1979. shall

~~p]ete all fire protection modlffca-tfpns needed to sathfy Criterion 3 of Appendix A to this part fn accordance with the Provtsfons of their licenses.

(48 FR 76610, Nov. 19. 1980]

N 80.49 Environmental quaHficatlon of electric equipment important to safety I'or nuclear power plants.

(s) Each holder of or each applicant for a license to operate a nuclear power plant shall establhh a program fpr qualifying the electric equipment defined In paragraph (b) of thfs sec-tfon.

(b) Electric equipment important to safety covered by thh section is:

(1)

Safety-related electric equip-ment:3 Thh equipment h that relied upon to remain functional during and fpHowlng design basis events to ensure (f) the fntegrtty of the reactor coolant pressure

boundary, (H) the capabHlty to shut down the reactor and malntafn lt In s safe shutdown condition, and (IH) the capabHlty to prevent or miti-gate the consequences of accidents that could result In potential offsite exposures comparable to the 10 CFR Part 100 guidelines.

Desfgn bash events are defined as conditions of normal operation, including anticipat-ed operational occurrences, design basis accidents, external

events, and natural phenomena for which the plant must be designed to ensure func-tions (I) through (Hl) of thh para-graph.

(2) Nonsafety-related electric equip-ment whose faHure under postulated environmental conditions could pre-vent satisfactory accomplishment of safety functions specified In subpara-graphs (I) through (Hl) of paragraph

'Safety. related electric equipment Is re.

ferred to as "Class 1E" equipment in IEEE 323-1974. Copies of this standard may be ob-tained from the Institute of Electrical and Electronics Engineers, Inc. ~ 345 East 4Vth Street, New York. NY 10017.

g S0.49 (b)(l) of this section by the safety-re-lated equipment.

(3) Certain post-accident monitoring equfpment.<

(c) Requirements for (1) dynamic and seismic quallflcatlon of electric equipment important to safety.

(2) protection of electric equipment lm.

portant to safety against other natural phenomena and external events, and (3) environmental qualification of elec-

, tric equipment important to safety lo-cated In a mild environment are not tncluded within the scope of thh sec-tion. A mHd envfronment h an envi-ronment that would at no time be sig-nificantly more severe than the envi-ronment that would occur during normal plant operation, fncludfng an-ticipated operational occurrences.

(d) The appHcant or licensee shall prepare a lht of electric equipment Im-portant to safety covered by this sec-tion. In addition, the applfcant or li-censee shall fnclude the following in-formation for this electric equipment important to safety ln a qualification file:

(1) The performance specfftcations under condltfons existing during and followingdesign basis accidents.

(2) The voltage. frequency. load, and other electrical characteristics for which the performance specified fn ac-cordance with paragraph (d)(l) of this section can be ensured.

(3) The environmental condltfons, including temperature,

pressure, hu-midity, radfatlon. chemicals, and sub-

-mergence at the location where the equipment must perform as specified In accordance with paragraphs (d)(1) and (2) of this section.

(e) The electric equipment quallflca-tion program must include and be based on the following.

~ Specific guidance concerning the types of variables to be monitored Is provided In Revhlon 2 of Regulatory Guide 1.9V, -In.

strumentation for Light-Water@poled Nu-clear Power Plants to hssess Plant and Envl.

rons Conditions During and Following an hccldent." Copies of the Regulatory Guide may be purchased through the UB. Govern-ment Printing Office by calling 202-275-2060 or by writing to the UB. Government Printing Office. P.O. Box 37082.-Washing.

ton. DC 20013-V082.

505

$ 50A9 (1) Tempernture and pressure.

The tim~ependent temperature and pres-sure at the location of the electric equipment Important to safety must be established for the most severe design basis accident during or follow-ing which this equipment ts required to remain functional.

(2)

Humfdffv.

Humidity during design basis accidents must be consid-ered.

(3) Chemfcal effects. The compos1-tlon of chemicals used must be at least as severe as that resulting from the most Itmtttng mode of plant operation (e.ir., containment

spray, emergency core cooling. or recirculation from con-tainment sump). If the composition of the chemical spray can be affected by equipment malfunctions, the most severe chemical spray environment that results from a single failure tn the spray system must be assumed.

(4) Rruffaffon. The radiation envi-ronment must be based on the type of radiation.

the total dose expected during normal operatton over the in-stalled life of the equipment, and the radiation environment associated with the most severe design bash accident during or following which the equip-ment is required to rematn functional, Including the radiation resulting from recirculating fluids for equipment lo-cated near the recirculating lines and tncludtng dose. rate effects.

(5) Agfng. Equipment qualified by test must be preconditioned by natural or artlftctal (accelerated) aging to its end~f-Installed life condition. Consid-eration must be given to all stgntfleant types of degradation which can have an effect on the functional capability of the equipment. If preconditioning to an endwf-Installed life condition Is not practicable. the equipment may be preconditioned to a shorter designated life. The equipment must be replaced or refurbished at the end of this desig-nated Itfe unless ongoing qualification demonstrates that the item has addi-ttonsl Itfe.

(ft) Submergence (if subJect to being submerged).

(7) Synergfsff c cffecfs. Synergtstlc ef-fects must be considered when these effects are believed to have a signifi-cant effect on equipment performance.

10 CFR Or.

1 (I l~ Edition)

(8) Nargfnr. Margins must be ap-plied to account for unquantified un-certainty, such as the effects of pro-duction varlatlons and Inaccuracies in test instruments. These margtrrs are in addition to any conservatlsms applied during.the derivation of local environ-mental conditions of the equipment unless these conservattsms can be quantified and shown to contain ap-propriate margtrN.

(f) Each item of electric equipment Important to safety must be qualified by one of the followingmethods:

(1) Testing an identical Item of equipment under Identical conditions or under similar conditions with s sup-porting analysts to show that the equipment to be qualified is accepta-ble.

(2) Testtng a similar item of equip-ment with a supporting analysis to show that the equipment to be quali-fied Is acceptable.

(3) Experience with identical or simi-lar equipment under similar conditions with a supporting analysis to show that the equipment to be qualified is acceptable.

(4) Analysis in combination with partial type test, data that supports

'he analytical assumptions and conclu-sions.

(g) Each holder of an operating li-cense issued prior to February 22, 1983. shall, by May 20, 1983. Identify the electric equipment Important to safety within the scope of this section already qualified and submit a sched-ule for either the qualification to the provisions of this section or for the re-placement of the rematrrtng electric equipment Important to safety within the scope of this section. This sched-ule must establish a goal of final envi-ronmental qualtficatton of the electric equipment wIthtn the scope of this section by the end of the second refu-eling outage after March 31, 1982 or by March 31, 1985, whichever Is earli-er. The Director of the Office of Nu-clear Reactor Regulation may grant requests for. extensions of this dead-line to a date no later than November

30. 1985, for specific pieces of equip-ment if these requests are filed on a ttmely basis and demonstrate good cause for the extension. such as pro-curement lead

time, test compltca-(:

a t,

s tr t:

c g

p f<

p rl rr si ti h

qt cr fl ti er br se nr m

ss tb ul er br cl (d

506

0

( ~

~ Edition~

nust be ap.

.'0.

quantified un.

'he effects of pro.

and Inaccuracies In I'hese margins are In vnservathms applied

,fon of local environ-

of the equipment mervathms can be

>own to contain ap-f electric equipment ty must be qualified wing methods:

Identical Item of identical conditions indftfons with a sup-to show that the qualified is accepta-7IQar item of equfp-portfng analysis to Ifpment to be qualf-Ith Identical or sfmf-er sfmQar conditions g analysh to show at to be qualified is combination with data that supports mptions and conclu-erating lf-

>r bruary 22, Ly

.983. Identify

>ment important to

~ope of this section Ind submit a sched-qualification to the ection or for the re-remafnfng electric Int to safety within section. This sched-a goal of final envi.

ation of the electric the scope of tbh of the second refu-March 31. 1982 or

~ whichever fs earlf-f the Office of Nu-n)latfon msy grant

~fons of thh dead.

iter than November ffc pieces of equfp-Iests are fQed on a demonstrate good nsfon, such as pro-

71e, test complfca-

~~~r Regulatory Commfsaion I ns and installation problems. In ex-ceptional cases. the Commhsfon Itself

~sy consider and grant extensions yond November 30, 1985. for comple-tion of environmental qualfffcatfon.

The schedule in this paragraph super-sedes the June 30, 1982, deadline, or any other previously imposed date, for environmental qualification of electric equipment contained in certain nucle-ar power operating licenses.

(h) Each lfcense shall notify the Commhsfon as specfffed fn 550.4 of any significant equipment qualifica-tion problem that may require exten-sion of the completion date provided fn accordance with paragraph (g) of this section within 60 days of fts dis-covery.

(I) Applicants for operatfng licenses granted after February 22.

1983, but prfor to November 30, 1985. shall per-form an analysis to ensure that the plant can be safely operated pending completion of equipment qualfffcation required by thh section. Thh analysis must be submitted, as specified fn

)50.4. for consideration prior to the granting of an operating license and must include, where appropriate, con-sideration of:

(1) Accomplhhfng the safety func-tion by some designated alternative equipment ff the principal equipment has not been demonstrated to be fully quailffed.

(2) The validity of partial test data In support of the original qualfffcs-tfon.

(3) Limited use of administrative controls over equipment that has not been demonstrated to be fully quali-fied.(i) Completion of the safety func-tion prior to exposure to the accident environment resulting from a design bash event and ensuring that the sub-sequent failure of the equipment does not degrade any safety function or mislead the operator.

(5) No significant degradatfon of any safety function or misleading Informa-tion to the operator as a result of faQ-ure of equipment under the accident environment resulting from a design bash event.

(J) A record of the quallflcatfon, fn-cludfng documentatfon In paragraph (d) of thh section, must be maintained In an sudftable form for the entire period during which the covered item Is installed fn the nuclear power plant or Is stored for future use to permit verification that each Item of electric equipment important to safety covered by thh section:

(1) Is qualified for its applfcation; and (2) Meets Its specified performance requirements when ft h subJected to the conditions predicted to be present when It must perform fts safety func-tion up to the end of Its qualified life.

(k) Applicants for and holders of op-erating licenses are not required to re-qualffy electric equipment Important to safety in accordance with the provi-sions of this section ff the Commission has previously required qualfffcatfon of that equipment In accordance with "Guidelines for Evaluating Environ-mental Qualfffcatfon of Class 1E Elec-trical Equipment fn Operating Reac-tors," November 1979 (DOR Guide-lines), or NUREG-0588 (For Comment version), "Interim Staff Posftfon on Environmental Qualification of Safety-Related Electrfcal Equfpment."

(I) Replacement equipment must be qualified In accordance with the provi-sions of this sectfon unless there are sound reasons to the contrary.

(48 FR 2733. Jan. 21, 1983, as amended at 49 FR 45578, Nov. 19. 1984: 51 FR 40308. Nov.

8, 198S: 51 FR 43709, Dee. 3, 198S; 52 FR 31811. bus. 21, 19871 IssvaIrcE, LIMITATIONS, hwv COIIM-TIolfs oF LIcENsEs Ar(D CorrsTRvc-TIoN PEsMITs 05050 Issuance of licenses ar>d construc-tion permita Vpon determination that an applica-tion for a license meets the standards and requirements of the act and regu-lations, and that notifications, If any, to other agencies or bodies have been duly made, the Commission wQI hsue a license, or If appropriate s construc-tion permit, in such form and contain-ing such condftfons and limitations In-cluding technical specifications, as It deems appropriate and necessary.

507

~ ~

rr Fodern) Register ( VOI. 48. No. 25 / Friday. January

22. 1983 '( Rules 'and Regulatinns

~<<ra<<eav~eeee i

i in regulated fresh channels ls 1

cgr.

=)ed to raquire about 55 percent af t:.,c, volume.'%e remaining 44 percent irauid bc avi Qable fot utQJsattoa in

c4N and prccesstng ouUets. The ccrc t4J)tcc indicates that volume and siss

. Uon of the crop ofaavel am 4 that mote than ample suppl ofth'ote dcsirshlelatgersices w'e ave I ble to ca Usfy ths demand in ed channels. The committee rcport that whenmore than ampl suppli oflarger~ ate avaQa e far QJpmr. t, dlsposJUon ofthe stsis

)den l would b clitbineled by this tion csn bc 4 abnplishtd only at a sub;tan ti I price discount and 4 tends ta drpres(thc market for all six Navel orang.c fa ling to meet such tequiwtr.

cauld be shipped

&ash e~,",ett ets. left on trees attain f0~Act.

or uUIIsed h sing.

la hase circ rnstsnccs. 'tion of srbsil an those ed is op>>wpilate the Interest'roducers 44 conscun it Js further cnd that it imobc&mbl~

d am o the public wctetcst tc give relimlna

notice, eosgc Jn publi erne and psstpbb ~ tsccff ds unUI309sys

~ftc.- publicuUon the F I Re)F'stet (5 US.C. 533), b use 0 insufficient ticcc bctwccb thc

'te tn 97fotmaUon bc~me avaQable hich this tcgulsUob is baM nd e cffccUvo dktr neccssery io sie tha dccisrtd putposcs o

! e acL Interested pctsurm watt given a opportun)ty to azbmJ JafonnatJon a

vitws on the tcgulaUon at an op ecting. Handlers bsve heea apprised uchptavls)ons acd ths affaire da Dst of Sub(ed.s In Patt trtty Merlcctiag eats otdcts.

California. Atison.

cs (Navel).

PART QOTWA$A DEDj Therefore. f ~ is a ed to teed

~ 4 fcQow: (four st) expires arch I4, 1S83, and wiQ n t be puhIJs ln the annual cods of cmetal Regs

!Jane):

f$07800 otenye NL (a) During period 21.1Ot)3.

through M W lOLLno shaQ handle aay vel oranges in the ptoducbon 4 which ate of a sateQsr iha ~ inches in diam a:

Prorided, t not to acceed 5 pe t,

by count, o tha oranges in any container ay measure smalla th X52 inch ia diameter.

(b) As ln this secUoa, "h et",

handle" d "ptaducUon area" ms the same ss defined Jn the metic order. D'ar shaQ mean the large measutemeat at a right angle to a traight line running from the stem to the

~ad ofthe Suii.

0 fL4a Stet. St, u ccsebdah t LS.C,

~4) tect Jabcwy it.3SC3.

a Dt t)rDlIeccor.ftultamt Vegtcrrblc Dir ion.AgricularrcrlHctkrdbgService pa awW~accm) ceca &1&4&%

Acl)ICY USDA.

ACllorc OIO r

assI'n Callfornfa an Uon ot Hancf))n tarsi Mar%a Sctvta rule au)Nstctn:

tegulaUon blishcs the quanUty ffresh Calif

-htioaa Iemons ths!

y be shipp to mattct during the January 2Q, 1O53.

Such acean Is ceded to vide for orderly af free lamons for the od due t the ma eUng situaUoa aUng the oa in ustty.

4strcCtttrl aalu ua 23, 1O83.

Fob sc)trrtcxtc

.William J. Doyle, e

FLV,hMS. IJSDA.

202$ ). telephone

,attseLascsjtrAtcv tus oN coÃrkct; Fruit Stanch, shington. DZ.

7~

Anorc TMs nasl rulc has baca swed under Scctrrtaty'4 Memo!

um 2512-1 and EcecuUve Order and hu been designated 4 noa s

t 'tule WUIJam T.Manity~

trator, hgricultural M eUag ce. has determined that s acU n wQIaot have

~ significant omic ect oa a aubstanUaI of sm eaUUss.

This acUon Js sigrced to tumota orderly matlce af ihe a-htfsoaa Ismo mop forthe~! of Ptaducets.a.

willnot subganthIIy affect costs the directly ted handlers.

Thish rale Js issued MakeUng erNo,t)10,as ended(y CFR Patt t): 47Ht N)IOt)).

ting tht han oflemoas growa Califotni andhtfxona The ard Is eiftcUv dcr!hs hgrfcultutal Agrceat hci of1

~as amend P USA fO1~4). T2ce acUoa Is bas upoa tecommcadaUoas ajar info tioa suhmJtted by tha Lemot) hdmi traUve Committee and upon o

avaQsble laformatioa. Itis hereby that this action wJQ tend to

~ffckuate the declared policy af the hcL acUcm Is consistent with!ha m

bating pa)icy for1O))~T2)e atkeUag poIJcy was tccommcbdcd by anmnittee CoQowfsg discassioa at a Pu IicmccUagoa)uIyLlOBZ.Thc ittec mct again publicly oa Jan 1tL 3O53. at Ios hagehe.

ia, to cotradsr the current and Ptas cUve candlUoos ofsupply and d and recorded a quaff)y f lem dcctturd advisable to be cd eh)ting c sirecQIsd wade. The commit c reports the dtcnabd for hmons nd<<ues easier.

itls er faced that ft Js hnpracU leandcoattutytoth ubIJc Jatetclt to vs Fttlimiaatyao I

engage in p blictulemaking.

Postpone cPecUve date 30 days after pub')J os in ihc F Rrryivtat (5 US.C. 553),

accuse af t

time between

.'a date wh

" ocmsUon became avaQs'ge upon whi this regula Uon Isb'bd cct)va date nccessat;:

efftctus

!he declartd putp~~ of tha Inta~lcd persons were g aa o tyto submif Jnform&>>

ews oi) the tegulation at an tUng.his necessary to effcctu t

the decimated purposes of 'the~ t cnalca &Hre regulatory IBOvtskin fftCtbAPcs specified.end r

nave(vs'pprised 0!such 's aa<< its F<<

effective Cmt.

Iht ofSubjects y CFR at!91!)

Metlceting e stets otdirs.

CaBfotnla.

04. Iamo PART92~

c3ED)

SecUoa Oi is acMcd oQovrtc INtLgeg eoa Wt)stet)on The qua tity oflemons h

CaMorni aad htfzoaa wh)ch y bc handled cafng the Petfod J 1O))3.

)anuaty 2L 2Ot)3. fs estaMJ t at SOAXS cattonL (gccL tace'fascia.ascseccdschy C.'e JabuajySO, 39a3.

IL ut)lice)d.

tyDirect'at aeI Vtyct45lc Dl

'gticultutalhtackeccrrSSerncu.

Ac Io 5%4FQal~tQw~

llVCLEAR REOULATCRY COMMISSION

'N CRt Part 50 Envlranmental Oual)f)caUon of E)ectric Equip)nant Important to Safety for Nctciaar Power Planta AolrCr:Nuclear Regulatory.

Cmnmissioa.

2733

~ e Federal Register I Viol. 4g, Nzx TS l Fsiday. 3azzaazy 24 19Lf / )traces azsd Raga)atfons I

ac-+ac Baal tufa.

P e

5 txzzsauvrn The Commfssfon fs amehdfug fis regala5ans

'cable to audear power plants to aad stzengttun the criteria for environmental qua55catfoa of electric equfpmezzt fmportent to safety. Specf5c qualification mithuds currently contamed inastfoual standards.

regulatory guideL aud certetn NBC blicetfons for eqmpment quelt5catfoa ve been givea dfffezent fntezpzetstfans'nd have not hsd the )egal farce ofen agency regulation. This amendmaut cxuffties the eavfzuamezztal quaB5catfon methods and criierfa that aust the

~ Comudssfoa's zzututremenfs tn dA atua Isyxctzva tzaYR Febnzazy 22. 2954

~ slzllfllkAussoklhlYuszt cceRsC2'atfsh Y AggetwsL Omce ofNuc)ear Iteg~ozy Resesizch. KS. Nuclear Reffulstory Cammfssfou. Washington, ILC.20555 Telephone (301) 443 594L S~tXXRAYARYSee CZSSLLY7CSC Previous Ifotfce Oia Jsuuezy 20. 29ff4 NRC pubhshad fzz the Fodersi Rcrpsia a notice of proposed zufemafung m euvinuuneatal qualiBcstfoa o! dectrfc eqzd pmeat for

'udear power plants (47 FR 2NO). The comment period expfzed March 22. 2NLX.

h total of 69 comment letters raising 1ff major bsues were teceivedby Aprfi8.

2962. An additional 10 couuneat letters vieze received by April21. 2982, but zxi new issues were raised. The zuajor Issues are discussed below.

Nature aad Scape ofthe Rulama)dag Nudear power pleat Important io safety anat shia to pezfozm iis safety functions thzoug)uzut fts fostaffed fffiLThis requfzeuLeet fs embodied fn Geoecsl Design Criteri 2.

2. 4, and 23 ofAppeadix A. "Gainers)

Design Czfteri forNudeaz Power HaatL" to 10 CFR Pazt 50, Domestic Liceushg ofProduction and UtfBxatfaa

~

Facffftfes": Ln Criterion ILL Design ContzoL" and Criterfon XL Test CoatzoL ofAppendix B.

"Qu~'ssurance Criteria forNudese Power Pfants aud Fuel Repzocessfug Pleats." to

~ 10 CFR Pszt 50; and fn paragraph.

ffL55a(h) of20 CFRht M. wbLch incorporates by reference IEEE 2m-2Ã2,'"Criteria for) zotsctfoa Systezus for Nudear Power Generating Siatioas.

This tequfzemeut fs applicable to equipmeot located inside as welf as outside ihe coutafnmea4 e Zeaeeyareclee by e'feseees esveeee AgO>>

DLect<e ot Zbe OZSee crMeeez kepis>> m fe>>>>eV 4 ZSS4 Copse may be Obieb>>4 bece 0>> lacunae et Nec&cd elf ZtecsreaJcs eeee~ tee S4S 1eee 47th Sevel Neee Yaek, Zt.Y.Sour.

The NQC haa used a variety of

&eQLods to Lsete that these general tequfzemenis are mei for ahzczric equfpmcnt hnpurtant io safety. Prior to ggn. quaB5catfon was b on the fact 4hat the electri'compoaents werc of high fndustrfal quality. For audear plants Bcensed to opetate after 29y2, quaB5catfon wes J'edged oa the basis of-XEBKg2g-20yLPor ptents whose Safety Rvaluatfon Reports forcmatrucffoa penufts were issued since July 1, 2%4, the Ceeezueeioa has essd Regulatory Cafde 489, "QtuB5catfoa af Class 1E Equipment forLight-Wa oled Nudaar Power Plants," which endorses IEEE 225-29y4.s'TEER Standard for QuaBfiyfug Class IE EqufpeLeet for Nudear Pewee Geaeoatfae Stations,"

subi to supplementary Ctzzzentty, tha Commfss has under wey a progzaza to teevaluate the quafiBcatftm ofefactric equipment !au)I opezatxzg audase power planta As a part ofthis program. more da6aftfvs criteriforeavfzona~ta1 qaa)iacatfaa ofelectric equipmaui fmportan! to safety have beau by tba NRC.A doczxuanj "GuideBuas far Evaluating Eavftcuuneutal QaaliQcatfoa of Class 2E Efactrfcsl Equfpmsut fn Operating Reactors" (DOR GufdeBnas) was fssued tn November 29y9. In addition. the NRC has issued NUHZG-058L "hzterim Staff Pasttfau an Environments) Qaalilcatfon ofSafety-Related Electric)Equfpmeat.

which contains two eats ofcriteri<< the ttzst for plants ceigfnaBy reviewed fu accordance with LEEE 225-am and the eecced for.

L faute reviewed fa ecconfance with X5 2%4.

By fta Memorenduza end Cheer CD-

~ 4aM Maya. 2980. tbe Cecaazfssfoo directed the staff to proceed witha seskaiialdug <<s envemmeatal qesffftcatf<<zifsefety-ze)ated eifefpzuent aad to address the ofbectzlt. The coaumsaoa also that the DOR Gufdeffase end NUKE~f~wm tbe basis fortbe zeqefnmeats iicen~ and appBcants

~~untffu rulem hhqfh s~

completed.'Ihis ru)s fs based on the DOR Geadeiieee and NURK&055LThe Cammissfon zecognfzszs the quaBtfcatfoa efforts ofthe industry as a result ofCD-OD4LThmkzre. ths rule provides that zeqvalftfcetion ofelectric ~pmcnt will ao! be by appf)caoo for and holders operating licenses fornudear power plants previously requited by NRC to qeaiffy equfpmeat ta accordance withDOR GuideBnes or NUREG-0588

~~ Ier)I). C~l

~

~

V

~me le ebhhal twa I~ sl leech>>l eel ZtoCnaks ~seas@,~ aeee

~ Seees. Seer Yeeh. EELY.RmZ.

zequfmme~ ef NflRR&45gLwhich apply to equfpmeut qeafftsed fn acc~facae with fEEE 225-2Ã'4. apply lo

'zzudear poever pleats tor which the coastnsctfoa peradt safety eva)oatfoa zcpect wae fsseed after Jufy 4 2fty4.

IIrequirements. which the secommendatiom afaad apply to eqaf fxneot quaff5e4 fa accordance sefth KEE223-2tty4 apply to tmdeer power p)eats forwhich tfm.

ooastzuctfoa perzsft safety nedeatfon seport~ issued prior as Jaiy 429F4.

hi CD-IIDRt.the Coauzzfs<<cm stated that unlese there were sound resins to the, zepfeceeuat perte should be to fhe stazutazda set forth ta Category IotNURE~e or IEEE225-Sty4. The Commfsefca reaffirms that position ia this tufemakiag. Such

~ queff5catioa osoetttutes campbance wQh the provfsi~ ofpazugreph LfL49(2).The Cocazafssfoa's posftfonis designed to promote tha pehcy of upgradiug the eavtzonmeataf quamfcatioa and zelfebiBty offastened efeciric equi pmeat. Situations msy arise, howewx. fa which each upgradfugariB not be fcasfb)e crccaufu5ble wftb overall pleat safety. Licensee aust review each siteatfoa oa a caseky~se basis to detcnufne that "sound ~zszs to ths coal~ ds exist to justifyea

~xceptfoa &am upgradiag, Examples of acceptable "sound to the-contrazy" willbe fncluded ia Regzhahuy The deice specified fu thfe.zzdefor compfetfoa ofeavizoumeotd quali6catfoa afe)sctzfc equfpmeat fmpoztent to safety apply to alflicensees aad epplfcsnts and superscide any data y fmpoe<<L No changes to or technical specf5catfoas aze necessary to satfect these new'ompletfoa dates.

The scope of the Ias) zale covers that poztfon ofequipment important to safety cammouly zefened to as "aafety-reletcd" (which ths Comusfsefoa.

fntezpzets as essentially "Ciass 2E"

~qufpmeat defined tn LEEK225-and nonsafetyceh ted electric

~qufpment failure under:

paste)otad environmental coodfttaus could prevent the satisfactory accompBshmeat ofrequired safety functions by safetymlated eqefpmcoL,

'af~ted structures. systems. aad czxnpoaeots ate those that aze zeffed apoa to rcznafa functional during aud fallowing design basfs events to caeuze g the fntegrity af Ihe reactor coolant pressure boundary. (ff)tbe capshffftyto shat down the reactor aod maiaiaia ftfn a safe shutdown condition. aod (iB) the ceyabfffty to pcevent ar mitigate ths

~ ~

Fpferal Register / Uo). 48, No. 15 / Friday. January 21, I983 / Ru)cs and RcgulsUo!Ls gp3g consequences olaccidents that could result ia polcaUal oEsfle exposmes ccmparsblc to thc gufdclincs of10 CFR Part 100. Dcsiga basis events are dc5aed as coadiUons ofnormal opcraUoa,'ncluding saUcfpsfcd opcraUoaal occunenccs: dcsiga basis accidents; external events: andnatural phenomena forwhich thc plant must be designed to ensure funcUoas (I) through tiB) above.

Also covered in the scope ol QLS 5asI tule fs ccrlah postsccf dent mouftorfag equfpmcat spcci5cd as "Catcgoty1 and 2."h Revision 2 ofRegulatory Cofd~

LN;"InstrmaentaUon for t-Water.

Cooled Nuclear Power Plants to Assess Plant and Environs CoadfUons Dmhg and Followhg an Aafdcat."

included h the 5nal rule are epecf5C tcchaJaf requfremcats pertainiag to (a) qualL5caUoa paraaetcrs.

(b) ali5caUon methods and (c) ocucLcata Uoa. Qaali5caUoa parameters include tcmpcratunL.

pressure, huaddily. radfaUoa. Cbemfals, and submsrgcaa. Qualf5a Uoa methods hcfude ta) tasthg as fhe princfpsl means ofqusB5aUoa aod Q) analysfs ia combinatfon withpartial type test data or operathg LcqLcrience.

Thc 5llsl nLIS rcqLLLres thai the qusli5a tion program Include syacrgisUC ffecls. radfatfoa.

<<avfroacLcatal condiUons aad margh corLsidcrsUons. Also. a record of qusli5csUon must be mahtahcd.

'Proposed Rcvfsioa 1 to Regulatory

.Guide MO, which hss bcca issued for pubBC comment. describes methods acceptable to the NRC formeeting the provisions ofthfs rulc and includes ~ Bst of typfcal equipmeat covered by ff.

Revisf on 1 fo Regulatory Guide %89 w91 bc issued after tcsoluUoa ofpubBC

ÃRC wfBgenerally not accept analysis alone ia lieu of testfag.

Experienc has shown that quaB5catioa ofcqufpmcat wftffout test data may not be adequate to demonstrate funcUoaal opcrabfBty during design basis event coaditioLLS. Paragraph SL49tf) provides fourmclhods for qualihcstfon. Tcsthg wfBbe preferred. To cnsme htcgrfty of a tcsUag program, QLS Commfssfoa eLcpects that the same piece of

~ equipmcat wfBbe used throughout the complete test uquanca The 5asl rule mpdres that each

+older of an operating Bccase provide a Ist ofelectri equfpmeat important to ufcfywithh the scope ofthis ntle previously quaB5cd based on t<<sthg.

analysis, or a combhaUon thereof, and

~ ILS't of equfpmcnt fhat has not been quaB5cd. These Bats and Qle schcdole for completion ofquaB5aUoa ofcfcctrfc equi pm<<nt must be subadttsd by Msy M19e3 The gcncraI cats forscfsmfc and dynamic caUoa for electric equfpmcat are contained h the Cencral Design Crffcria and are not Included wfthh thc scope of Qds rul>>. Fmther gufdsnce Js provfdcdh tory GuJdc 1.1N, "Seismic Qu tfoa bf Efcctric EquJ pmeat forNuclear Power Pfants." 04vfsfoa 1) and NUREG4bN, Standan} Review Pfan." NRC Js'oasfdcrfag future rulemslihg I ccmccrnhg ants for the quaB5aUon ofelcctric r

~qufpmsnt important to safety aad the cequfremeats for schmLC and dynamic quaB5caUon ofelectric equfpmeat.

'omaaats On Tba Repoeed Rule The Commission nLcefved and coasfdered the comments on the Proposedrule contained La fhe 891ettsre received from the publLc by AprilL 1QLI.Copies ofthose Iattare and a staff response to each comment are avaQabld fotpublh fnspeclfon and copyhg fora tee at the Commfssfoa's PubBC Document Room at sly H Street NW VAshhgtoa.D.C.

The ma for Issues raised by the comments aad NRC sfaNresponscs are

. as foBows:

P) Sekmic and Dynmnic

. Quailfi'cokicu~rgmph 8R4p(c)

Issue: Seismic and dyaaadc quaB5aUoas are an integral part of environmental quaBfiaUoa. Ilfs therefore happroprfste to codify these niquiremenfs separately.

Respoas'e: Electric cqufpmeat at opera Uag nuclear power planta was gcncraBy quaB5ed fot environmental and scisadc stresses separately. La by ashg separate prototypes for

~nvfroamcafal and seismic quali5caUcm tests. The Commission has dicfdcLL after considerable dsliberatha. to Pursue the Issue ofsefsmh and dynanda quaB5aUon separately at a futme data h future sefsmhrule may not require rates thg for.environmental stresses because a shgle prototype was not used chLrhg the orfgfnclquali5atfoa. ALSO.

Protectfaa ofelectric equipmcat fmportant to safety agaiast olha natural Phenomcaa and external events shouM not be wfQda the scope of thLS nde.

pgScop~ld8hufdowa

~~aag~hanag)

Issue: &erule fatroducas a new rettufrement to "equfpmcat needed to complete oas path of

~chfevhg aad mahtahhg a cold ahutdowa conditfon." h change ofQds nLagaitude. at this advanced stage ofthe s quaB5cstioa cEort. most y haodLLccs sfgLBcaat acw

'osts and obBgatfons wLth no demonstrated fmprovc=cat fn safety.

RespoasL'egulatory ~remcnts fa affect at the thae ofBccafag ofthc mafmily ofopcrathg reactors did not sequfre that aB elcarfc cqtdpmcnt aad systems necessary to br~ fhe reactor to cold shutdown be cfassL5ed as safety relateiL However. Slccaic equfpaLcat and systems accessary to shut down the reactor and maLntain fth a safe shutdown condiUoa are required to be CISSSIScd as safely related and therefore are covered by the nde.

Tbc Commfssioa fs CLL~tlystLufyiag the requirements for shutdown decay heat removal under Uaresolved Safety Issue gJSI) h~ The~ puqese of h~ Ls to evaluate the adequacy of current Bcenshg reltufre cats to ensure that failure to remove shutdown decay heat does not pose an Lmacccptshle sf' Uadcr~a comprebcasfveand consistent sct ofshutdown coolhg requirements forexfsthg and future Pleats fs behg devchpcd. Tbc 5aal tcchnical resolutfoa ofh~ fs prescaQy Scheduled for October 1QIL4'.

The Comndssfan bcBcvce Itwould bc Pnanature at this Umc to Impose thc eequflemcnt to envfnnLmcataly qualil'y electric equfpment and systems necessary fo achfcve arLd 'cAfntah coM ahutdown prior to the 5naf resolutfon of h-Lf.Therefore. this requL~cnt Ls not included in the 5nal nde.

p)Scop~uipauurf fa o Mld Envlnuuacat-Paropoph JC4$ fbg hsuc: Tha ruIe ma%Its no dfsUncUoa between equfpment hscstcd in a harsh or mild envfroamcnt. The stresses for equfpmcaf in a m08 environment are less severe than for those ia a harsh environment Response: Tbc 5nalnLIc does not cover QLS electric equipmc=t loafed In a mM environment. The Co~sfonhas ooacludcd that QLS general quaBty aad surveillance requirements appBcabls to electric equfpmcnt as a result ofother Commfssfoa regulatfons. Including 19 CFR Part 50, hppcndfLc B (sce for escample. R tmy Cuidc~

ty Assurance Profpaa Requirements (Opere Uoa). Revfshn g) are suf5cfcnt to ensure adequate performance ofelcccfc cq pmcnt Important to safety located fn mm environments. Sha Ilhss been concluded that no furler cavfronmcata1 quamaUom requirements are needed fotsuch equfpmeat dcd they fully SSUsfy all other appliablc ~SUons.

the Commission has deter~cd that no

~ddfUonal requirements are necessary

~

~

~

~ ~,J 2n2 Fofiersi Register / VOL 48, No.'15 / Frhiay, January 21.'19&3 / Rules and Reffa)ations witn rss pect to elsctzfc equfpmant important to safety )ocatad fa mf)d environments fn order forlicensees to satisfy, with respect to such equipment.

axhting )fcense coadftfons or techafcal specf5cs5ons caffhg for quali5ca5on af safety-related efsctnc equipment La accordance with DQR Guufsliaes or NURE~ItL Lit)S~prrvr'osis Qaahpcotiozr Efforts-hmagmpb 844$(b)

Issu>> The rule does aot recognize that plants have comp)etsd qua)i5ca5on of e.qufpment to the DOR Gufdafhes or hUREG458& Without such reeognftha.

hdustzy efforta, maa power. and bf))fons ofdo))ars wi)lgo down the drah.

Respons>> The Saal rule has been expanded to alleviate this concern. See Paragraph SL4Q(k).

(g)Humldi p-Rzngrrzph RM+r)P)

Issue: The effects of5mWeyendsnt varia5ons ofreh5ve huznfdfty durhg norma) opera tioa caaoot be considered for all equfpmenL Tbere are no detailed standards tor how thh type oftes5ng

~hou) d be performed.

Respons>> The Cotnmfssfon agnea.

Humidity variation during norma) opera5na an dffzfca)t to predict. Ithas zsot been demonstrated that the 5me-dependeat variation fn humidity will produce any differences ln degradaticm ofe)ectric equipment. The words "Tlme-dependent variation ofrelative have been deleted from Paragraph SLL49(e)(2).

fl~

I'Jfff Issu>> The nqufremeat that ongohg quafI5ca5cas be done ushg "prototype eql7fpinent aatuaBy aged" h ovazfy restric5ve. Use of acct) ezutad aghg to defms a quali5sd life fs not techmca))y feasible.

Res pons>> Pzeeoaditfonhg by

~ccelerated aging h teehnfcaOy feasfMa for sfmpfe e)ectri6 equfpmeat forpleat nfe and for complex electzic aquipmsnt for a shorter designated Iffe.Tha Commfsaioa recognfses thats~~

technology willbe u51hed fn any aghg yrogram. Reference to quaff5sd fifehas bean deleted from paragraph 5(L4Q(e)(5).

-JJ~

fd'f hsu>> The marghs eyplied fn adll5oa to known ccaservatilns lead to excessive stnss thet could )aad to fsf)uns ofequipmaat fa azufaalh5c quali5ea tfon tests.

Response: The Coaunhaioa agrees.

This nqufrement could have caused excessive margins. T)>>paragraph haa been modf5ed to recoffnfse conservatfsms that caa be qualified.

(tt)Amr1ysis aad partial test dato-PhrrztPoPh SR4PN(4) bsu>> ifpartfsl type test data that

~dequately support the analy5cal ussump5oas aad conefusfons are avef)ab)e, thefr ana)ysfs should be allowed to cxtzspolate or hterpolate these results for equfpmsnt. regardless atyurchase data.

Respor>>>> The Commfsaon agrees.

Reference to piuchase date" has baca da)eted.

P) Requfrensent fara central/flr-Jhmgmpk tftL~)

Issu>> The requirement for a centra) fifeshould be deleted shee ft fs not cost effective and has no safety bena5t.

Respoiss>> The Commhsion agrees.

Thh requirement has been subject to

'fffenust hteryrets5lxzs. h record of qua)L5ca5oa must be mahtahed fn an "audftab)e farm but not necessari)y in

~ centra) 51e for tha entfze period durfDg

<<hfch 8>> covend Ltam h fastaI)ed h a eucfear power plant. Recordk zequfranent of10 CFR Part h B

must be zueL Certah ncords can be Rcpt at the vendor s shop ff0)/asti/lcctson ofcontinuedoprrrrtlorI jbrcfparatutgpkznts.

Issu>> Tlze requfnment to aubmit fus55ca5on forthe cantfnued operation ofopezathg planta shou)d be de)ated shee thh hfonna5on has been Previous)y submitted to NRC.

~Res pons>> Ms requirement has beea Qa5sfsctorf)y mat and Paragraph NL4Q(J) ofthe proposed zu)s has been deleted h Lts entirety hem 5>> laal rule.

In addition. Paragraph 5L4Q(g) ofthe zu)e has been ds)etad from the rule shee Lth too yrascriptfva. It wfffbe facfudad h Regu)story Guide LSL

~ectr've Ilute Thh rule rep)aces the "htsefm nde" publfshad fa the Fadars)

Regfst>> oa Jane 3LL 1QQ2 (42 FR 263tO).

Tha "fatarira nde" suspended

~avfrorunenta) quali5ca5on deadlhes contafned h )fcense conditions or techrdca) speci5catfons ofopezathg y)ants, Oa the e8ec5va date of thh zu)e (s*a above)y the "htarim rule" fa superseded and the schedule for

. enviroameatal quali5catfoa contahad La thh ra)e takes effect forall planta.

Paperwork Reduc5onhc2 The Snal ru)e con tahs hfonnatioa ca)) ec5OD requirements that have been approved by the OL5ca ofManagement and Budget: OhtB approval number h

$15~)1L Regu)story HexihQlty Statement Ia accordance with tha Regulatory

~bf)ftyma1Qea.5 '.e 5050>>

the Commhsfon hcnby caztf5es that thfs rule wf)Laot have ~ afgnilcant

~aomfc impact oa a sabstaa5al number ofsmall entf5es. This final nde sdfects the method of quaff5cs5on of electric equfpmeat by utf)f5es. U5115es do not fallwftkh ths definition ofa small buaineas hand in Section 3 ofthc Small Bushass hcL 15 U~ 532.

.In addi tfoa. u5ff5ss are nqizfred by Lhe Commfssfnn's Memarandum and Order CUM~dated May 23. 1QSL to e>>et the requirements contained h tha DOR "GuLdelines for Evutuathg Eavfraamentt) Quali6cation of C)sss 1E E)ectriea) Equipment fn Operating Reactors," (November 1QyQ) and

. NUREG-055tL "Iaterim Staff Position au

'Environmental Quali5catfoa af Safety-Re)ated Zleczrical Equfpment." which Conn the basis of thh zu)a.

Consequently. thh ru)e codfles axhtfng requirements aad imposes ao new costs or ob)fga5ons cm a5lf5as.

Qa of S bjects Le 20 CFR Part 50 ha5tzust. Chasi5ed hfonnation. Hre yravaatfoa. Intergovernmental n)atfona.

Nuc)eaz power plants and zeactors, Penalty, Radhtfon protectfon. Reactor sftfrLLcriteria. Reporthg requf~enta.

Pursuant to the htomic Energy hct of 1054. as amended. the Energy Raorffsnfza5onhct of 1QZL as ameaderL and section 553 of5t)e S of the United States Coda. 10 CFR Part 50 h amended as fo)low>>

PART ~AMENDEDI g.'B>> authozfty citation forPazt 50 conthues to retd as fo))owe hu6ecfty: Sea. 10L 10L ZQL1L. ZSL 1SL 2SLL SC SlsL taA Qzr. Q4L QSL QSL QSL SSL as amended. see. 23L QS SteL124L as amemfsd (42 VS.C 2ZSL 2ZSL 22ot. 2222. 22'2ZIL 223tL 22SZ): sees. 2OL 2OL 2NL aa SlsL 1242.

224L 1244. as snead<<f (42 lLLC,SS4L SS42.

sees). unless otherwise aotecL Seetfeo SLZ also fs<<wd sixie puh. LQS 4OL see. 1L Qz Stet. 2QSt (41V~SSSt).

Section SLZS afso fess<<L umfer see. 12L SS Slat. QZS (42 tLS.C. 2152). Seetferm SLS04LSZ also fssu<<L unda see.zSL SS Slat QSL as amended (42 QRC 2234). Sections SLZtO SL102 also fess<<L andes see. 1' SIaL QSS (42 VS.C 2234).

For the parposee ofsee. 22L OS Stet tnL as amended (42 vs.c 22z2) iisL1o(e), (b).

and (e). SL44, SL4L SL4L SLSL end SLSO(e) are fsse<<f ader see. lslb. ss Slat 044, as amended (42 lALG2zot(b)): iisL10 (b) esd (e) erMLSLS4 are fess<<f voder sae.zrrzL SS StsL Qss. as amend<<f (42 VS.C 22OZ(l)): aod IISLSS(e). SLSO(b). SLZL SLZLSLZ2 and NLZQare fssu<<L uader see. 1mo. 48 SLat. QSL as amemf<<f (42 tLSA 2'(o)).

R. Section Stl4Q h revised ta read as Mow>>

~

~

~ Federal Register f Vol. 48. No. 15 / Friday. Jazzuary

21. 1983 / Rules and Regulations gr33

i. Earls favhcottettctd ttueattctt50A ot c bt ~m~~ bttOottsttt io setttty tor

~ power plettts

(.) achhoid*r or~em~~

for 4 BtpJL1c t5 operate 4 nuckazpower plant shsQ tstshiish a program for qu-Q~ th citciric equi eat Ctfictd ia pars~ph (b) of this aecUan.

(b) E!~c cquipinant iiupartaat to aaft ty covered by this secUoa itc (1) SttfttyMttted electric cquipmetcts This equipment is that zeBed upoa to remain functional daring und foQowing design basis events to eniuze (i) the fnttpityoftht reactor cnobtnt pressure bound sly. (ii)th>>

capability to shut down ih ztsctat and maintah itin 4 cafe sbutdawa coadiiioa. aad (iQ) the capsbiBty to prevent cr miUgate tha conrtqusscts ofacddents ihet cotdd zesull iu poteaUal offsite cxposuzes comparab4 to ths 10 CFR Part 50) guidtBnts. Design basis ev~ are Cehnsd az ccuditioas ofnormal operaUon. iududing aaUdpated opezatiaaal occurrences. design basis aa8dtats, external cveatL andnatural phtnomtna for which the plantmust be cftsigned to ensure fuacUuas fi) ihzaugh (tQ) of this p ~3r ph.

(2) Nottsafctymiaied electric equiptamt whose faQuze tmdsr pos tule tea environmental cuztd15ons

~

could prevent satisfactory accompBshmtat ofsafety funcUons sptdfitd in subparsgraphs (i) through (Ui)ofparagraph {b){1)af this secUon by

.tht safety.r lated tquipmsat.

(3) Cutsia pastmcddeat monitoring cquiptt'taut.a (c) Requirements for(i) dynamic aad sthmic quafificaUon ofdeliria equipmsnt imponant to safety. (Q) protecUon ofcltctric equipmsnt important to safety against other natural phenomena aod external events. cnd

[1Q) environmental qua51icsUon of electri tquipmtct t haportent to safety located ia a mfid <<nvnonmeat see trot

$5dudtd within the.scope af this secUoa.

Amfid environment is aa envizoameczt that would at no thne be signi6cantly mora severs than ths environment that wouid occur Curing normal plant opera Uoa, hdudiag anUdpated operaUoaal occuznaices.

a Safctraeteted &cote ettettcesnt tc tetbtted w ee Ctua iE attettcoaet le tEEE ~let@ Ceptae

~rtbtt otaedanl star be cbtaieed eeet tbe laatttote OlBeCOtCal aed CaCOeetCe EatteaWe. hC Sil Ecat Crth SbwtC New l'ett. NYuair.

~Specta c Stddaste ceocetnbta the trpea aC vetttb4s te bc toctitoted lo proruhl te aettstee S

~ittttubtetr Celde tetr, ttta~wttcttee ht tisht Watct Coasts Nactau &neer Plcota to~ Plate eod Eerlrans Ccedl seoa l4rkoe aed FeUctwles ae hcddent." Coziae et tbakcstdatcty Guide cce be obtabted ttete Nod>> aasalctetr Cooeciaaiee.

Oocct~ ktaeaetto~ tbaeck We bLisstctL DC uiptaaat lfpztcuudiUDnicg to 4n cad

~hstaQad )ifa cootQUcn is w cticabie. the equiptcent my be ndiUoatd to a shorw designated

t. Tht equipment nmsi be ~laced or tuzbished at the cad cf this designated e unless ongcing quaLGCS~

traits that the item has Uoaal Bfe.

8) Subzoccgem>> (ifscbjet mbtiag maimed).

P') Syztezmzstk~ S~SUc eats must be costi~ ween these ecis aze beBevcd to hevt ~ sigalficaat ect cm equipment pafazmmce.

(8) Motgtits. Margins must be applied accatmt for enquantieed ~inty.

as the effects afproc'~

Uoas and haccunLdes in test tnnuerts. These ma~ a~ in diUoa to any constrv~ applied the dsrivaUoa of local vizonmeatal conditions af ~~

mtzzt unless these ~~Usms quantified and shown to contain pziatc margins.

(i) Etch item ofebctr~t ~ctit poztuat to safety must be qmlboodby ofthe foQowingmeti~

(1) TCSUag an idtaUcal

!tern nf uipmtnt nader ideaUcal cocdi~ or der similar conditions with a ppoiUng ana)yiis to show Mthe pment to be qualified is eccsptsb)s.

(2) TcsUng 4 simQar item oftquipmtnt th 4 suppuzUng analysis to mw that s equipment to be qua Bfied is table.

(3) Expcrience with idcaUcsl ccr equipment unders~

DdlUons with 4 euppo~ analysis to w that the'eqtdpmcnt to be qaaBfied acctpiabl4 (4) Analysis in tnmbicaUon with type test data that a~its the yUcaf asstlinpUons 4nd ~hisitms (g) Each holder ofan~~ lictase ucd prior to February 22. 1863. ShalL.

May 2L L83. ideaUfy the decal:

pmeat t to safety wMiu e scope of stcUtm alzecdy sBfied and submit a sch~~

for ther tbt qualSCSUoa to tht provisions this secUoa ar for the replacement of remaining electric equipmect portent to safety within the scope of stcUoa. This schedule adust bQsh a gott) of5nal cnvtrx:ucatal aBfication af thse~ ~ ptneat thia the scope ofthis secUCO by the.

d of thc second iefueBag ctz~ after 3L 1982 orby March 3L 1%5.

ever is eczBsr. The9~ ofthe Ece ofNuclear Rasctcrr Regulatory y gran! requests forextra~ af this dfizla to 4 d4te 00 later~

ovtmbcr 30, 195. fur spado pieces of pment ifthese requests a~ filedon (d) The appBcaat or Bccnsec shaQ prepare 4 liat of electric equipmtat fmpoztant to safe ty coveted by this

~ecUoa. ln'eddiUuue the appficsnt c>>

Bcenset shaQ include ths fofiowiag tafonaaUoa forthis clectzic equipment ze important to safety h a quttBficaUtxt Bf 51L demons (1) The pezfocmitace sped5cstions adtB ender condiUons cxisthg daring aad foQowing design basis acddcnts.

cu (2) Thtvoitaet. frequency. ictad. 4ad oihtr electrical chazscttristtcs for.which 46 tha perfonnance epedfied fn accunfaztce elf with paragraph {d)(1)ofthis section caa cff be ecietized.

(3) The cavizccumntal cnndiUona, to fndudiag tactperettue, pressure.

Such humidity. zadiaUoa chtcnicals, cad vane aubmezgtutce at the locaUaa whezs tha ias equipmtnt must perform es spedhtd in ed accordance with paragraphs (d)(l)aad dazing (2) ofthis section.

~ ca (e) The electric equipmsnt

~qtd qualificaUoa mmust fndude cnd can be based on (1) Tern~rotate and Pcsstu>>. The pressure at tha locsthn af ihs electric ons equipmeat important to safety must ba estabBshtd for the most severe design eq basis accident during or foQowiag which en this equipmeat is required to remain au fuacUonaL

~ equi (2) Humidity. Humidityduring design bills accidents must be considered. ~

wi (3) Cbetttico/Eg~. Tbe compositioa th ofchemicals need must be at least as acccp

~evert as that rtsulUng fram the most BmiUng mode ofplant operation (ag aimthtr contaiamsnt spray. emtzgtacy care co oooliag. er zedrculaUon from aho containment sump). lfthe composiUon is ofthe chtunical spnry can be affectedby equipment malfancUons. the most 'azUal severe chemical spray cavizoztmtat that anal results frcnn ~ single fafiuze in the spray system must be essumecL hs (c) Boifiotfoa.The radiatioa by cnvfzotunent must be band cm the type equi ofzadiaUon. the total doss expected th during normal opsntUoa over the qu fnstaQed Bfe of the equipmeat. and the ei zediaUoa environment assodated with of the most seveze dssiga basis acddent the chuiag or foQowiag which the equipment im h rtquhed to remain ftmcUonaL this fndutBng the redieUon zesulUng fram cata zedzculaUng fiufds for cqufpmtnt qa located asar the zectzculsUng Bnee and wi fndutBng te effsctL ea (5)Aguy,Equipmtat bytest Mazch must be pzecoudtUaaed y nctucsl cr which artQidal (accelerated) aging to its cad.

0 OWnstSQedlife caatDUon. ConsidsntUon zaa must be given to aQ significant types of dea dtgradaUoa which can have an effect oa N

the fimcUonal capabiBty ofQ>> 'qui

~

~

~4 Fpdcral Register f Vol. 48. No. 25 / Friday, lanuaty 21, 19LJ / Rules and Rcgulationa a Umsfy basis aad demonstrate good cau>>a for the extea>>Jon; such a ~

ptocu~ent lead time, test comf:ficsUons. and ha tallathn problems. In exec pUonal cases, the Commis>>fon it>>clfmsy consider and grant cxtcn>>ious beyond November 30, 196$. for complctfon ofcnvftenmcataf qualffcaUoa.

th) Each licensee shall noUfy the Commission olany sfgnfScaat cqufpm>>at qaDIScaUon

'problem that msy tequfw exteasioa of the compfetfoa date provided fn accetcfanca with para praph jg) of this sectha withh 60

~ days of J ts discovery.

ti) hppUcants for opetathg licenses that are to be granted on or atter Fchtuaty 22, 1M3. buf priar to November 3L 198$. >>hall perform aa analysis tq ensure that the plant can be safely operated pindfng completioa of cquJpmcat qualiScaUon requited by this

~ecUca. This analysis must be submitted to the Director of the OfSce ofNuclear Reactor Rcguladoa forcensfdcraUcm r to th>> gtaaUng af aa operating cence and must fncfucfc. where

~pptcotiste, consf~Uoa ok P) Accompffshfag the safety fcmcUaa by some designated af tcmaUve equfpmcat ffthe priadpaf equipment hcs not been demonstrated to be fully quslfSac'2) Iha validity ofpartial test data in support ofthe orighal quaifScaUoa.

t3) Ifmf ted usc ofadmJnistra Uve controb over equipment that has not baca demonstrated to be fullyqualiScd.

(4) Compl>>Uoa ofthe safety functha prior to exposure to the accfcfcnt environment tesulUag from a cfcsfgn

'asis evtat end ensuring that the sub>>equmt fsilute of the equipmcnt does not degrade any safety fuacUon or mfslcsd thc operator.

f5) No sigaJS cant degradatha ofany safety functha or mf>>feeding fnfotmaUoa to the opctatat as a result of failure of equfp1acnt under ihe accident environment resulting from a design basis evea')h record of the qualiScaUon.

facfudhg documcats Uon fn paragraph (d) af this!ection. must be maintahed fa an auditahle form forthe entire period during whJcb the coveted item is installed in the nucfcar power plant or fs stored fot future ase to permit vctiScsUon that each ftcm ofelectri equJpment important to safety coveted by this section (I)Ia qualiS ed forfts appffcaUotn aad (2) Meets its specfled perfonnance tequfremcnts whee ftfs subjected to the condJUoa>> predicted to be present when it must perform its safety hactfon up to the end ofits qualfSed Jffe.

0) hpph tsf~m~a epctaUng ffccn>>es ate not requited to tequafffy electric cqufpment fmpoctant to safety fn accotcfance wfth thc provisions ofthis sccUon ffthe Commission has previously requited quaffScaUoa ofthat equipmant h

~ccotdance with "Guidclhes for Evaluathg Envbenmcatal QualUicaUon of Class 1E Eteetricaf Equipment fa OpetaUag Reactors.

November Iffy g)OR GuJdelhcs), or NURE~SSf Por Comment version), Interim Staff PosfUon cm Eavbeamcatal QualfScaUon ofSafety-Related Eiecaical Equipmcat."

g) Replacement equipmcat must ba qualified in accetdance with the

~

ptevfsfons of thfs secUcm ualesa theta ate sound reasons to the ccmttaty..

Dated at Wsahfagtaa. ILC.thf>> 1tthcfay of

)anuatye 19S3.

For thc Necfaat Regul>>tocy Ccxamf>>>>foo.

Same>>f ). Chfffc.

8eaeccuyojcbe Cacus'>>iaL

~a Dea~ rani~ ae a>>f CO OOOV FUIIJRES TRAOJNO CO ON

%7CFR 140 ancf 145 eadquatteta CNJce and Westettt and western Regional ONcea; Chang fAdcfteaa Aaattcv: Cammodi Futures I'tadhg Commis>>foe.

• CtletcFinal rale oust>>sattv: The Commodf

~diag Coaunissfon fs am tegvlaUons fn an attempt to that both the physical location and mailing address ofthe Commi i

's headquarters ofSce ate oae the same forall ptacUcal putpo Ia addi tfoa. the Commfssfoa amendhg fts regula Uons to include w addresses fot fts recently relocs eitcm and Southwestern regfonal ccs. Tha Western Regfonal ofS has beaa snoved ftem Saa F to Los hngeles. California e Southwestern Rcgfonal once. l tcd fa Kansas City.

hHssoutL has m to ~ different suite ofofSccs ia the buffdhg.

gtFscttvg oa january 21.1QcL poN hatt tnlcDc acHttsATtolt ccNTAcle Donald LT dick hcUng ExccuUve

Directar, ty Futures Tradhg Commfssf 2033 XStreet NW Was a.D.C.205ttI. @02) aensL

~

Co sion rcgufaUon f 14L1 cunently pm cs a separate physfeal locatha address fot tha Commission'a headquarters ofScc. Th Commission fs amcading tegulaUan f 14L1 to cfarify that there is no distfncUon between fts physica location and mafffag s ss.

Thc sole address oi the Commis>>

o's he dq~~offb a.ofl~

1963 wfffbc 2033 KStreet.N.

Washington.D.C. 20581.

The Commission fs tagulaUcm14L2to reflect feet that the Wcstcm Reghnal ofS of the Commission has moved San Francisco to It850 Wils Boufcvatch Suite 51L Los hngsles onda f0024.

Thc telephone numb far general fafotmaUoa fs (2LJ) additfon, regula f 14L2 is behg amended to note c Southwestern Regfoaal of6cc>> moved &em Room 208 to SuJte 400 t 4JO1 Maia Street.

Kansas City.

64112. Thc telephone n for general fafotmaUcn (tttb) 3744QL Cattah provisions ofthe Commfss n'a regulaUons ecmtah tefetan to ar adcftes>>cs ofthe Commf sfoa's Western and South astern Regional onfecs. The a

tiate changes have been mad>> to teff thc new addresses fa each of th e provfsfoaL on thc foregoing, pursccant fo fts Ihoritycontahcd fn sccthn 2taQl) of e Conunodity Exchange hct. 7 US.C.

4am (IF/6), thc Commission hereby amends Parts 140 and 145 ofthe Code of Federal Regula thus as follows:

PART 1~AMENDEO)

1. SecUon14Li fs tevfsed to read as foQows:

f14L1 Hasd tusttacs OtaoL (a) CcncteL Ihc headquarters oKcc ofthe Commission fs faceted at 2033 K Street. NW Washhgtoa. ILL2058L

2. Sectfon14L2 fs amended by revising paragraphs (c) aad (d) to read folfowa f

Segfoaotttoe~eftlonaf

~ ~

~

~

(c)

Wcstcm Regional ofSca fs located a 06M Wflsfcfte BoulavatcL Suite SIL hag>>les. California 00034.

and fs respo hi>>for cnfotecmcat of the hct and a tratiea ofthe ptegrams ofthe sion fn the States ofhfasfca.

Coaa. Cafffotnfs.

HawaL idaho. Mon

Nevada, Oregon. Utah. W oa. aad Wycanhg.

(d) The Southwestern onal ofScc fs located at 4JO1 Msh Stre Suite 400.

Kanaaa City, Missour 54112.

fs saspoasfbfc for eafoteemcn1 ot e hct