ML18031A321
| ML18031A321 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/23/1979 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Curtis N PENNSYLVANIA POWER & LIGHT CO. |
| Shared Package | |
| ML18031A323 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 NUDOCS 7912060563 | |
| Download: ML18031A321 (65) | |
Text
Docket f';os.:
50-3G7/3QS DISTRIBUTION:
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1.79 Docket, File@~JBuc 'arian NRC PDR PStoddart Local PDR EWenzi nger TIC AHintze LWR d3 File ACRS (1)
SMiner VBenaroya Fr. i~orrian W. Curtis.
Vice President Engineering and Construction Pennsylvania Power ard Li~.'t t Caspar<y 2 i<orth Ninth Street Allentown, Pennsylvania 18101
Dear ";r. Curtis:
SVarga FWi lliams-OParr HRushbrook DCrutchfield IE (3)
ACRS (16)
GArlotta
SUBJECT:
PROPOSED REVISION 2 TO REGULATOPY GUIDE 1,97 "I.'ST:":I.t<ENTATIOt,'OR LIGHT-WATER-CGGLED NUCLEAR POWER PLi>ITS TO ASSESS i~LA.; A;D E VIROI<
CONDITIONS DUR I"IG AND FOLLOls ING Al~ ACCIDFNT
'he Pegulatory staff is revising Regulatory Guide 1.97 to provide r..ore specific guidance.than, that contained in the guide now being used.
A draft of proposed revision " of this guide is enclosed (Enclosure
- 1).
The Advisory Ccrrnittee on Reactor Sa,eguards has revieved proposed Revision 2 of the guide and
.'.as agreed to its issuance or corz:ent.
Enclosure 1 is based on the results of Task Action Plan A-3!, " Instru-,en-tation for '-'cnitoring Radiation anu Process Variabl..s During n Accident"
- which:;as initiated in June 1977 to develcp r;:ore specific guidance concerning ip~pIe.",entation of Regulatory Guide 1.:.'7.
A r.ajar addition to the current Regulatory Cuide 1.97 is the identification of
.". pecific paraneters to be
'.".easured, the rant'e of the ~.".easure?.".ents ard design criteria for the instr@:ents.
The enclosed dra t guide also incorporates the applicable recor'rendatiors in fiUREG-v'578 "T.I-2 Lessons Learned Task Force Status Report nd 'hor&Terr:
Re<<o;~:endations."
The prapcsed Pevisicn 2 of this guice will have a signi ic<int lr?pact on the cesign of plants such as yours that are currentl: under review fcr an operati <<:
licerise.
Therefore, we are recuesting early corr?ents on Proposed
<".evision fror. applicants for these plants.
l'.e have arranged;.:eetirgs in Qethesca,
.'<aryl and to discuss these cot;.,.ents.
~e will.",;eet with nine applicants for pressuri ed water reac ors on Decerber 13, 1979 and wi.h five ap"licants for boii ing water reactors on."ece;:ber 1<':,
1';79 ~
F e request that you attend the i;eeti rg as indicated in the enclosed
< eet i n<
- iotice (Enclosure
"', to discuss your cc..refts
<:n the feasibili " ('
.esi<.nine arc installin<? instr!a.ents?.eetin<<
the re<< irc;,.ents c.
ore< nso.,'.evision
>912060
Hr. l{orman M. Curtfs o Pegulatory Cuide 1.97 fn your plant.
An advance copy of major corments and the associated rationale should be given to V. Benaroya, Chief, Auxiliary Systems
- Branch, DSS, i'.RC prior to the meeting.
If there are any questions regarding the meeting, call L. L.
Kfntner (301) 492-8344.
Sincerely,
Enclosures:
1.
Draft of Proposed Revision 2
to Regulatory Gufde 1.97 2.
Meetfng Notice cc w/Encls:
See next page S., A. 'larga, Acting Assistant Director for Light !Pater Reactors Dfvfsfon of Project management CUlaNAMCW OATC+,...,...ltd.
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Mr. Norman W. Curtis cc:
Mr. Ea'rle M. Mead Project Engineering Manager Pennsylvania Power 8 Light Company 2 North Ninth Street Al 1 entown, Pennsyl vania 18101 Jay Silberg, Esq.
Shaw, Pittman, Potts 8
Trowbridge 1800 M Street, N.
W.
Washington, D. C.
20036 Mr. William E. Barberich, Nuclear Licensing Group Supervisor Pennsylvania Power 8 Light Company 2 North Ninth Street Allentor, Pennsyl vania 18101 Edward M. Nagel, Esquire General Counsel and Secretary Pennsylvania Power 8 Light Company 2 North Ninth Street Al 1 entown, Pennsyl vani a 18101 Bryan Snapp, Esq.
Pennsylvania Power 8 Light Conipany 901 Hamilton Street Al 1 entown, Pennsyl vania 18101 Robert fl. Gallo Resident Inspector P. 0.
Box 52 Shickshinny, Pennsyl vani a 18655 Susquehanna Environmental Advocates c/o Gerald Schultz, Esq.
500 South River Street Wilkes-Barre, PA 18702 John L. Anderson Oak Ridge National Laboratory Union Carbide Corporatln>>
Bldg. 3500, P. 0.
Box X
Oak Ridge, Tennessee 37830 Mr. Robert J.
Shovl in Project Manager Pennsylvania Power and Light Co.
2 North Ninth Street Al lentown, Pennsyl vania 1810 I Matias F. Travieso-Diaz, Esq.
Shaw, Pittman, Potts h
Trowbridge 1800 M Street, N.
W.
Washington, D. C.
20036 Dr. Judith H. Johnsrud Co-Direc or Environmental Coalition on Nuclear Power 433 Orlando Avenue State College, PA 16801 fir. Thomas M. Gerusky, Di rectory Bureau of Radiation Protection Department of Environmental Resources Commonwealth of Pennsylvania P. 0.
Box 2063 Harr isburg, PA 17120 Ms. Colleen Marsh Box 538A, RD84 Mountain Top, PA 18707 Mrs. Irene Lemanowicz, Chairoerson The Citizens Against tluclear Dangers P. 0.
Box 377 RDP1
- Berwick, PA 18503
Draft. 1 October 15, 1979 PROPOSED REVISION 2 TO REGULATORY GUIDE 1.97 INSTRUMENTATION FOR LIGHT-WATER"COOLED NUCLEAR POWER PLANTS TO ASSESS PLANT AND ENVIRONS CONDITIONS DURING AND FOLLOWING AN ACCIDENT A.
INTRODUCTION Criterion 13, "Instrumentation and Control," of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Licensing of Production and Utilization Facilities," includes a requirement that instrumentation be provided to monitor variables and systems for accident conditions as appropriate to ensure adequate safety.
Criterion 19, "Control, Room," of Appendix A to 10 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 loss-of-coolant accidents and that equipment at appropriate locations outside the control room be provided with, a design capability for prompt hot shutdown of the reactor including necessary instrumentation.
Criterion 64, "Monitoring Radioactivity Releases,"
of Appendix A to 10 CFR Part 50 includes a requirement that means be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluid, effluent discharge
- paths, and the plant environs for radioactivity that may be released from postulated accidents.
This guide describes a method acceptable to the NRC staff for complying with the Commission's regulations to provide instrumentation to monitor plant variables and systems during and following an accident in a light-water-cooled nuclear power plant.
B.
DISCUSSION Indications of plant variables and status of systems important to safety are required by the plant operator (licensee) during accident situations to (1) provide information required to permit the operator to take pre-planned manual actions to accomplish safe plant shutdown; (2) determine whether the reactor trip, engineered-safety-feature
- systems, and manually initiated systems are performing their intended functions, i.e., reactivity control, core cooling, maintaining reactor coolant system integrity, and maintaining containment integrity; (3) provide information to the operator that will enable him to determine the potential for causing a breach of the barriers to radioactivity release (i.e.,
fuel cladding, reactor coolant pressure boundry and containment) and if a barrier has been breached; (4) furnish data for deciding on the need to take unplanned action if an automatic or manually initiated safety system is not functioning properly or the plant is not responding properly to the safety systems in operation; (5) allow for early indication of the need to initiate action necessary to protect the public and for an estimate of the magnitude of the impending threat.
At the start of an accident, it may be difficult for the operator to deter-mine immediately what accident has occurred or is occurring and, therefore, determine the appropriate response.
For this reason, reactor trip and certain other safety actions (e. g.,
emergency core cooling actuation, containment isola-tion, or depressurization) have been designed to be performed automatically during the initial stages of an accident.
Instrumentation is also provided to indicate information about plant parameters required to enable the operation of manually initiated safety systems and other appropriate operator actions involv-ing systems important to safety.
1.97"2
Instrumentation is also needed to provide information about some plant parameters that will alert the operator to conditions that have degraded beyond those postulated in the accident analysis so that the operator can take actions that are available to mitigate the consequences.
It is not intended that the operator be encouraged to circumvent systems important to safety prematurely, but that he be adequately informed in order that unplanned actions can be taken when necessary.
Examples of serious events that could threaten safety if conditions degrade beyond those assumed in the Final Safety Analysis Report are loss-of-coolant accidents (LOCAs), overpressure transients, ATMSs reactivity excursions, and releases of radioactive materials.
Such events require that the operator under-
- stand, in a short time period, the ability of the barriers to limit radioactivity release, i.e., the potential for breach of a barrier, or an actual breach of a barrier by an accident in progress.
It is essential that the required instrumentation be capable of surviving the accident environment in which it is located for the length of time its func-tion is required as defined by ANS-4.5, Section
- 3. 0. It could therefore either be designed to withstand the accident environment or be protected by a local protected environment.
If the environment surrounding an instrument component is the same for accident and normal operating conditions (e.g.,
some instrumen-tation components outside of containment or those in the main control room powered by a Class 1E source),
the instrumentation components need no special environmental qualification.
It is important that accident-monitoring instrumentation components and their mounts that cannot be located in other than non-Seismic Category I build-ings be conservatively designed for the intended service.
- l. 97"3
Parameters selected for accident monitoring can be selected so as to permit relatively few instruments to provide the essential information needed by the operator for postaccident monitoring.
Further, it is prudent that a limited number of those parameter s (e. g., containment pressure, primary system pressure) be monitored by instruments qualified to more stringent environmental r equire-ments and with ranges that extend well beyond that which the selected parameters
/
can attain under limiting conditions.
1t is essential that the range selections not be arbitrary but sufficiently high that the instruments will always be on-scale; for example, a range for the containment pressure monitor extending to the burst pressure of the containment in order that the operator will not be blind as to the level of containment pressure.
Provisions of such instruments are important so that responses to corrective actions can be observed and the need for, and magnitude of, further actions determined.
On the other hand, we should also make sure that when a range is extended, the sensitivity and accu-racy of the instrument are within acceptable limits.
Normal power plant instrumentation remaining functional for all accident conditions can provide indication, records, and (with certain types of instru-ments) time-history responses for many parameters 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 the operator to use, during accident situations, instruments with which he is most familiar.
Since some accidents impose severe operating requirements on instrumen-tation components., it may be necessary to upgrade those instrumentation components to withstand the more severe operating conditions and to measure greater variations of monitored variables that may be associated with the acciden if they are to be
- 1. 97-4
used for both accident and normal operation.
However, it is essential that instru-mentation so upgraded does not compromise the accuracy and sensitivity reauired for normal operation.
In some cases this will necessitate use of overlapping ranges of instruments to monitor the required range of the parameter to be monitored.
Draft Standard ANS-4.5, "functional Requirements for Post Accident Monitoring Capability for the Control Room Operator of a Nuclear Power Generating Station,"
dated September 1979, delineates criteria for determining the variables to be monitored by the control room operator, as required for safety, during the course of an accident and during the long-term stable shutdown phase followng an accident.
Draft Standard ANS-4.5 was prepared by ANS 4 Morking Group 4.5 with two primary objectives, (1) to address that instrumentation which permits the operator to monitor expected parameter changes in an accident period, and (2) to address extended range instrumentation deemed appropriate for the possibility of encounter-ing previously unforeseen events.
The standard defines four classifications af variable types for the purpose of aiding the designer in his selection of accident monitoring instrumentation and applicable criteria.
(A fifth type (Type E) has been added by this regula-tory guide.)
The types are, (1) Type A " those variables that provide informa-tion needed for pre-planned operator actions, (2) Type 8 - those variables that provide information to indicate whether plant safety functions are being accom-
- plished, (3) Type C - those variables that provide information to indicate he 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, (4) Type 0 - those variables that provide information to indicate the performance of individual safety systems, and (5) Type E - those variables to be monitored as required to provide defense-in-depth and for diagnosis and
- 1. 97-5
other useful purposes.
Type A variables have not been included in the listings of variables to be measured because they are plant specific and will depend upon the operations that the designer chooses for pre-planned manual action.
The five classifications are not mutually exclusive in that a given variable (or instrument) may be included in one or more types, as well as for normal power plant operation.
Mhere such multiple listing occurs, it is essential that instrumentation be capable of meeting the most stringent requirements.
1 The time phases (Phases I, II, 5 III) delineated in ANS-4.5 are not speci-fied for each variable in this regulatory guide.
These considerations are plant specific.
It is important that the required instrumentation survive the accident environment and function as long as the information it provides is needed by the plant operator.
C.
REGULATORY POSITION The criteria, requirements, and recommendations contained in Oraft Standard ANS-4.5, "Functional Requirements for Post Accident Monitoring Capability for the Control Room Operator of a Nuclear Power Generating Station," dated September
- 1979, are considered by the NRC staff to be generally acceptable for providing instrumentation to monitor variables and systems for accident condi-tions and for monitoring the reactor containment, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge
- paths, and the plant environs for radioactivity that may be released during and follow-ing an accident from a nuclear power plant; subject to the following:
(1)
Section Z. 0 of ANSI-4.5, defines the scope of the standard as contain-ing criteria for determining the variables to be monitored by the control room operator during and following an accident. that will need some operator action.
- 1. 97"6
Consideration should be given to the additional requirements (e.g.,
emergency planning) of variables to be monitored by the plant operator (licensee) during and following an accident.
Instrumentation selected for use by the plant opera-tor for monitoring conditions of the plant are useful in an emergency situation 4
and for other purposes and therefore should be factored into the emergency plans action level criteria.
(2)
In Section 3.0 of ANS-4.5, the definition of "Type C" includes two items, (1) and (2).
Item (1) includes those instruments that indicate the extent to which parameters, which indicate the potential for a breach in the containment, have exceeded the design basis values.
In conjunction with the parameters that indicate the potential for a breach in the containment, the parameters that have the potential for causing a breach in the fuel cladding (e. g., core exit temperature) and the reactor coolant pressure boundary (e. g.,
reactor coolant pressure) should also be included.
References to Type C instru-
- ments, and associated parameters to be measured, in Oraft Standard ANS-4.5 should include this expanded definition, e.g.,
Section 4.2, Sec~ion 5.0c, Section
- 5. 1.3, Section 5.2.2, Section 6.3.
(3)
Section 3.0 of ANS-4.5 defines design basis accident events.
In conjunction with the design basis accident events delineated in the standard, those events which are expected to occur one or more times during the life of a nuclear power unit and include but are not limited to loss of power to all recirculating pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power, should be included.
(4)
Section 4.2 of ANS-4.5, discusses the various types of variables.
Mith regard to the discussion of Type 0 variables, Type 0 variables and instru-ments are within the scope of Accident i~lonitoring Instrumentation, although
- l. 97-7
they are not addressed in Draft Standard ANS-4.5.
They are,
- however, along with an additional type, Type E, included in this regulatory guide.
(See Tables 1, 2 and 3)
(5)
Section
- 5. 2. 1(5) of ANS-4. 5 pertains to the delineation of the local environment in which instruments must operate.
Section
- 5. 2. 1(5) should be understood to require identification of the range of the local physical and electrical environments (e. g., normal,
- abnormal, accident.,
and post-accident) in which all of the various instrumentation components are required to operate (e.g.,
- sensors, cables, signal conditioning equipment, indicators).
(6)
Section 5.2.2 of ANS-4.5 pertains to the performance requirements for Type C instrumentation.
In conjunction with Section 5.2.2, there should be:
(1)
Identification of the range of the process variable.
(Note-the range selected should extend well beyond that which the variable value can attain under limiting conditions)
(2)
Identification of the required accuracy of measurement (3)
Identification of the required response characteristics (4)
Identification of the time interval beginning with initiation of an accident to as long as the measurement is needed (5)
Identification of the local environment (including energy supply) in which the various instrumentation components are required to operate.
(7)
Section
- 6. 1. 1 of ANS-4.5, pertains to seismic qualification criteria.
In conjunction with Section
- 6. 1.1, those instrumentation components which should be seismically qualified are identified in Table 1 of this regulatory guide.
(8)
Section
- 6. 1. 1 of ANS-4.5, pertains, in part, to the consider ation of vibrational loads.
In conjunc ion with Section
- 6. 1. 1, those instrumentation
- 1. 97-8
components which are subjected to vibrational loads that occur as a result of plant system operation during any phase for which the instrumentation is required should be qualified to function during and/or following such vibrational loads.
(9)
Section 6.1.2 of ANS-4.5, pertains to the duration that instrumentation is qualified to function.
In conjunction with Section
- 6. 1.2, Phase II instrumen-tation should be qualified to function for not less than 200 days unless a shorter time, based on need or component accessability for replacement or repair, can be justified.
(10)
Section
- 6. 1.6 of ANS-4.5 pertains to instrumentation location and identification.
In conjunction with Section
- 6. l. 6, accident monitoring instru-mentation displays should be located in direct view of the plant operator and be distinguished from other displays.
Other accident monitoring instrumentation components should be accessible to the plant operator for maintenance and repair although this may not be possible for some components in some accident conditions.
(11)
Section 6.2. 1 of ANS-4.5 pertains to general requirements for Type B
instruments.
In conjunction with Section
- 6. 2. 1, Type B instruments are essen-tial to meeting the requirements of Criterion 13 and Criterion 64 of Appendix A
to 10 CFR Part 50 and are not considered to be an "extra set of instruments which result in an additional layer of protection."
Type B instruments are essential to the monitoring of variables and systems during accident conditions and in following the course of an accident.
(12)
Section 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 ANS-4.5 pertain to variables and variable ranges for monitoring.
In conjunction with the above sections, Tables 1, 2 and 3 of this regulatory guide (which includes those parameters mentioned in the above sections} should be used in developing the minimum set of instruments and tl eir respective ranges for accident monitoring instrumentation for each nuclear power plant.
(13)
Sections 6.3.2.3, 6.3.3.3, 6.3.3.4, 6.3.4.3, 6.3.4.4, 6.3.5.2, 6.3
~ 5.3, and 6.3'.4 of ANS-4.5 pertain, in part, to instrument transient response and relate his to compatability with recorder capabilities.
In conjunction with the above sections, the transient response requirements of each measurement should be determined on a case-by-case basis by analysis of the event and operator response capabilities.
(14)
Sections 6.3.3.3, 6.3.3.4, and 6.3.4.3 of ANS-4.5, pertain, in part, to measurement accuracy.
In conjunction with the above sections, the accuracy of each measurement should be consistent with the requirements as established by analysis of the event being monitored.
(15)
Section 6.3.6. 1. 1 ANS-4.5 pertains, in part, to the qualification of Type C instrumentation components.
In conjunction with Section
- 6. 3. 6. l. 1, the environmental envelope for qualification should be the extreme value of each environmental parameter, except the variable being monitored, as determined by the accident analysis for all accidents evaluated in the safety analysis of the pl ant.
(16)
Table 6.4.1 of ANS-4.5 pertains to design criteria for accident monitoring instrumentation.
In conjunction with Table 6.4.1, the provisions as indicated in Table 1 of this regulatory guide should be used.
0.
IMPLEMENTATION This proposed revision has been released to encourage public participation in its development.
Except in those cases in which an applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, ihe method to be described in the active guide reflect-ing public comments will be used in the evaluation of the following applica-tions that are docketed after the implementation date to be specified in the guide:
- 1. 97-10
1.
Preliminary Design Approval (POA) applications and Preliminary Duplicate Design Approval (POOA) applications.
2.
Final Design Approval, Type 2 (FDA-2), applications and Final Duplicate Design Approval, Type 2 (FODA-2), applications.
3.
Manufacturing License (ML) applications.
4, Construction Permit (CP) applications except for those portions of CP applications that reference standard designs (i.e.,
- PDA, FOA-1, FOA-2,
- POOA, FOOA-1, FDOA-2, or ML) or that reference qualified base plant designs under the replication option.
In addition, the NRC staff intends to implement part or all of this guide for all operating plants, plants under construction, all POA's and FOA's, all POOA's and all FODA's which may involve additions, elimination, or modification of structures,
- systems, or components of the facility after the construction permit, or design approval has been issued.
All backfitting decisions in accordance with the positions stated in this guide will be determined by the staff on a case-by-case basis.
The implementation date of this guide will in no case be earlier than
CRITERIA TABLE 1 - DESIGN CRITERIA)
INSTRUMENTATION TYPESz
- 1. Seismic gualification per Reg Guide 1.100
- 2. Single Failure Criteria per Reg Guide 1.53
- 3. Environmental tlualification per Reg Guide 1.89
- 4. Consider loss of off-site power 5.
Power source I
- 6. Out of service interval before accident
- 7. Portable
- 8. guality assurance level
- 9. Oisplay type
- 10. Display method
- 11. Unique identification
- 12. Periodic testing
- per, Reg Guide 1.118 yes yes yes yes Emrs no Con'3 Rec>>
yes yes B
yes yes yes yes CB~
no Con "3 Rec)6 yes yes yes yes yes3 yes CBe no'0 Con>>
Rec'6 yes yes no no yes yes Emrs no'0 oo>>
Ind) 7 no yes no no no4 yes Emrs no'0 0O)4 Ind)7,18 no no HOTFS for Table 1:
())
Unless different specifications are given in this regulatory
- guide, the specifications in ANSI N320-1979, "Performance Specifications for Reactor Emergency Radiological Monitoring Instrumentation,"
apply to the high-range containment area monitors, area exposure rate monitors in other buildings, effluent and environmental
- monitors, and portable instruments for measuring radiation or radioactivity.
(z)
Type A - Those instruments which provide information required to take pre-planned manual actions.
Type B - Those instruments which provide information to monitor the process of accomplishing critical safety functions.
Type C - Those instruments that indicate ihe potential or breach-ing or the actual breach of the barriers to fission pro-duct release.
Type 0 - Those instruments that indicate the performance of in-dividual safety systems.
Type E - Those instruments that provide information for defense-in-depth and for diagnosis or other useful purposes.
- 1. 97-12
807ES for Tail 1 continued:
(s).See Paragraph 6.3.6 of Oraft Standard Al'IS-4.5.
(~)
gualified to the conditions of its ooeration.
(s)
Emergency power source.
(6)
Critical Instrument 8uss - Class 1E Power.
(7)
IEEE 279-1971 Paragraph
- 4. 11, "Exemption".
(8) 8ased on normal tech spec requirements on out-of-service for the safety system it serves.
(9)
Hot necessary to include in tech specs.
(lo)
Radiation monitoring outside containment may be portable if as designated.
(11)
Level of quality assurance per 10 CFR Part 50, Appendix 8.
( 12)
Continuous indication or recording displays a
given variable at all times; intermittent indi-cation or recording displays a given variable periodically; on demand indication or recording displays a given variable only when requested.
(ls)
Continuous display.
(lv) indication on demand.
(ls)
Mhere trend or transient information is essential to planned operator actions.
(1s)
Recording.
(17)
Oial or digital indication.
(~a}
Effluent release monitors require recording, in-cluding ef luent radioactivity monitors, environs exposure rate monitors, and meteorology monitors.
1.97-j3
TABLE 2 -
PUR 'IARIASLES 0 i~teasured Variab1e Range Typ e Purpose CORE:
Core Exit Temperature 150'F to,2300'F B,C AXS-4. 5, Section 6.2. 3 Provide incore temperature measurements to identify localized hot areas.
(Approximately 50 measurements)
Control Rod Position Full in or not full in D
Provide positive indication that the con-trol rods are fuLly inserted.
(Minimum 5 days after accident) neutron Flux 1 c/s to 1% power (at least one fission counter)
ANS-4. 5, Section
- 6. 2. 2 For indication of approach to criticality, REACTOR COOLANT SYSTEi~1:
RCS Hot Leg Temper-ature 150'F to 750 F
ANS-4.5$ Section 6.2.3 To aid-in determining reactor system sub-cooling and to provide indication of natural circulation.
RCS Cold Leg Temper-ature 150'F to 750'F Aihs-4. 5, Section 6. 2. 3 To provide ind'cation or" natural circula-tion; to provide input for heat balance calculations; for direct indication of ECCS injection.
RCS Pressure 15 psia to 4000 psig B$ C ANS-4.5, Sections 6.2.3 and 6.2.4 For indication of an accident and to in-dicate that actions must oe taken to mitigate an event.
Pressurizer Level Bottom tangent to top tangent B,D A'fS-4.5, Section 5.2.3 Level indication is required to assure proper operation of the pressurizer and to assure safe operation of heaters.
is also used in conjunction with changes in reactor oressure to determine leak and void sizes.
Degree of Subcool'ng 200'F subcooling to 35 F suoerheat For indication oi margin 'n core coong and the need for emergency coolant add'-
tions or reductions as the margin changes and ro obviate the necessity to consu' steam ta'oles.
1.97-14
TABLE 2 -
PMR VARIABLES continued-Measured Variable Range Type Puroose REACTOR COOLANT SYSTEi~l CONTINUED:
Reactor Coolant Loop Flow 0 to 120%))
-20% to 20%]) design flowi B,D To provide indication that the coze is being cooled.
Primary System Safet)
Relief Valve Posi-tions or 'Flow Throug&
or Pressure Xn Relic Valve Lines Closed-not closed B~D By these measuz'ements the operator knows if there is a path open for loss of cool-ant and that an event may be.in progress.
Radiation Level in Primary Coolant Wate 10 pCi/g to 10 Ci/g C
ASS-4.5, Section 6.3.2 Foz early indication of fuel cladding failure and estimate of extent of damage.
CONTAINMENT:
Containment Pressure 10 psia pressure to 3 times design press-ure2 for concrete; 4
times design pressure for steel B,C A'fS-4.5, Sections 6.2.5, 6.3.3, 6.3.4, and 6.3.5 For indication of the integrity of the primary or secondary system pressure boundaries.
To indicate the potential for leakage from the containment; to indicate integrity oz the containment.
Containment Atmos-phere Temoerature 40 r to 400 F
For indication of the performance of the containment cooling system and adequate mixing.
Containment Hydrogen Concentration 0 to 10%
(capable of operating from 10 psia to maximum design press-ure B,C A'iS-4.5, Sections 6.2.5 and 6.3.5 For indication of the need, and to meas-ure the performance oz the containment hydrogen recombiner.
Containment iso3.a-tion Valve Position Closed-not closed B,D ANS-4.5, Section 6.2.5 To indicate the status oz containment isolation and to provide information on the status oz valves in process lines which could carry radioactive mater'als out of containment.
1,97-15
TABLE 2 -
PMR YAR~ABLES cintinued-Heasured Yariail e Ranae Type Purpose CONTAINMENT CONTINUED:
Containment Sump
'<later Level Harrow range (sump)
Vide range (bottom ox containment to 600,000 gall.on level equiva-lent)
B,C For indication of leakage within the containment and to assure adequate in-ventory for performance of the ECCS.
High Range Contain-ment Area Radiation 1 to 107 R/hr (60 keU to 3 HeV photons with +20% accuracy fox photons of 0.1 to 3
iieV)[107 R/hr for pho-tons is approzimately equivalent to 10 rads per hour for betas,and photons]
B,C For implementation of GDC 64 and to help identify if an accident has degraded be-yond calculated values and to indicate its magnitude to determine action to protect the public..
SECONDARY SYSTEMS:
Steam Generator Pressure From pressure for safety valve setting to plus 20% of safety valve setting For indication of integrity of the sec-ondary system, and an indication of cap-ability zor decay heat removal.
Steam Generator Level From tube sheet to separators For indication of. integrity of the sec-ondary system, and an indication ox cap-ability for decay heat removal.,
Auxiliary Feedwater Flow 0 to 110/ design zlow To indicate an adequate source of water to each steam generator upon loss or main feedwater.
in Feedwater Flow 0 to 110% design flowi To indicate an adequate source of water to each steam generator.
Sazaty/Relief Valve Positions ox'~n Steam Flow Closed-not closed To indicate integrity oz secondary system (vis-a-vis pipe break).
Radiation 'n Condensex'i" Removal System Radioacxiv'ty in Efzlu-ent from Steam Gener-ator Safexy Raliex Valves or Atmosoneric Dump Valves 10 to 10 iiCi/cc 10 7 to 10S WCi/cc B,C To indicate leakage from the primary to the secondary system and measure of noble gas release rate to atmosohera.
An indication oi release xrom the seconaary system and measure oz nob'e gas release rata to atmosphere.
j.97-16
TABLE 2 -
O'AR VARIABLES continued-Measured Variail e Range Type Puroose AUXILIARY SYSTEi'iS:
Containment Spray Flow 0 to 110% design flow For indication of system oper at 'n.
Flow in HPI System Flow in LPI System 0 to 110% design flow 0 to 110% design flow For indication of system ooeration.
For indication of system operation.
Emergency
- Coolant, Water Storage Tank Level Top to bottom To determine the amount of water dis-charged by the ECCS.
This provides in-dication of the nature of the accident, indication of the performance of the
- ECCS, and indication of the necessity for operator action.
Accumulator Tank Level Top to bottom To indicate whether the tanks have in-jected to the reactor coolant system.
Accumulator Isolation Valve Positions Closed-not closed To indicate state of the isolation valves.
(Per Regulatory Guide 1.47)
RHR System Flow 0 to 110% design flow D
For indication of system operation.
RHR Heat Exchanger Out Temperature 32'F to 350'F For indication of system operation.
Component Cooling Water Temperature 32'F to 200'F For indicat'on o
syste~ operation.
Component Cooling Water Flow 0 to 110% design flow D
For indication of system ope ation.
Flow in UHS Loon 0 to 110% design flow For'ndication of system oo'eration.
Temperature in Ulti-mate Heat Sink Loon 30'F to 150'F For indication oz system operat'on.
Ultimate Heat Sink Level Plant specific To ensure adequate source of cooling water.
Heat Removal by the Containment Fan Coolers Plant specific to 'ndicate system operat'on Boric Acid Charging Flow 0 to 110% cesign flowl To provide incication of reactor coo ing and inventory control and maintain ace'-
auate concentration for shutdown margin.
1.97-17
TABLE 2 -
PWR VARIABLES continued-Heasured Variable Range Type Puroose AUXILIARY SYSTEMS CONTI i'IUED:
Letdown Flow 0 to 110% design flow For indication of reactor coolant inven-tory control and boron concentration control.
Sump Level in Spaces of iquipment Required for Safety To corresponding level of safety equipment failure To monitor environmental conditions of equipment in closed spaces.
RADWASTE SYSTEi~lS:
High Level Radioactiv Top to bottom Liquid Tank Level Available volume to store primary coolant Radioactive Gas Hold-up Tank Pressure 0 to 150% of design pressure~
Available capacity to store waste gases.
VENTILATION SYSTEi~lS:
inergency Ventilation Damper Position Open-closed status To ensure pxoper ventilation unde" accident conditions.
Temperature of Soace in Vicinity of iquip-ment Required zor Sazety 30'F to 180'F To monitor environmental conditions of equipment in closed spaces.
POWER SUPPLIES:
Status oz Class li Power Supplies and Sys tens Voltages and currents To ensure an adequate source of electric power for sazety systems.
Status of Don-Class li Power Suppl'as and Systems Voltages and currents It 'ndicate an acequata source of elec-tric powers 1.97-18
TABL: PMR VARIABLES continued
-'easured VariabIe Range Type Puroose RADIATION EXPOSURE RATES INSIDE BUILDINGS OR AREAS MHERE ACCESS IS REQUIRED TO SERVICE SAFETY RELATED E UIPMENT Radiation Exposure Rates 10"- to 10" R/hr for photons (permanently install-ed monitors)
,, For measurement of high-range radiation exposure rates at various locations.
AIRBORNE RADIOACTIVE i~/AT FRIALS RELEASED FROtl THE PLANT:
Effluent Radioactiv-ity - Noble Gases
. Containment
.Secondary Contain-ment Auxiliary Building including building containing primary system gases, e.g.
waste gas decay tank
.Other Release Points [including fuel handling area if separage from auxiliary building (Normal plus accident range foz noble gas) 10 7 to 10S pCi/cc Xe~l33 calibration 10"7 to 10" pCi/cc Ze-133 calibration 10 7 to 10 pCi/cc 10" to 10 pCi/cc (permanently ins tail-ed monitors)
ANS-4.5 Section 6.2.6 To provide operator with information regarding release of radioactive noble gases on a continuous basis.
Effluent Rad'oactiv-ity - High Range Raa"ohalogens and Pazticulates
.Untreated Effluents
.HEPA Filters, min-imum of 2" of TEDA impregnated chaz-
- coal, non-ESF sys-tems
.HEPA Filters, min-um oz 4" of TEDA impregn ted char-
- coal, ESF systems 10 to 10 pCi/cc 10 S to 10 pCi/cc 10-3 to 1 pCi/cc (permanently install-ed monitors)
To proviae the operator with information regarding release oz radioact've nalogen>>
and particulates.
Continuous collection of representative samples followed by monitoring (measurements) oz samples
- foz, zaaiohalogens and fo" particulates.
- 1. 97-19
TABLE 2 -
O'AR VARIABLES con. nued-MeasUred Variable Range Type Purpose AIRBORi'IE RADIOACTIVE MAT ERIALS RELEASED FROM THE PLNT CONTIi'IUED:
Znvirons Radioactiv-ity - High Range Fzposure Rata 10 3 to 10~ R/hr (60 keV to 3 MeV)
(permanently ins tail-ed monitors)
For estimating release rates of radio-active materials released during an accident rom unidentified release paths (not covered by effluent monitors) continuous readout capabil'y, approxi-mately 16 to 20 locations - site de-pendent.
Environs Radioactiv-ity - Radiohalogens and Particulates 10 to 10
'pCi/cc for both radiohalogens and particulates (permanently install-ed monitors)
For estimating releases rates of radio active materials released during an accident from unidentified release paths (not covered by effluent monitors).
Continuous collection of representative samples followed by monitoring (measure-ments) of the samples.
(Approximately 16 to 20 locations)
AIRBORNE RADIOACTIVE i%TERIALS RELEASES FROM THE PLA)lT CONTINUED:
l Plant and Znvi ons Racioactivity (portable ins t~ent~
Hormal Rance 0.1 to 10" mR/h" photons 10 e to 10 " pCi/cc particulates 10
" to 10 4 pCi/cc iodine Dur ng and following an accident, to m:
itor radiation and a'rborne rad'oact
"='oncentrations in many areas throughout the facility whe.e is ~practical tc install stationary -onitors capabl of covering both orbal and accident leva',
Ki h Ranee 0.1 to 10 R/hr pho tons O.l to 10" rads/h betas and low ene gy photons 100-channel gem-ray spectrometer During and following an accident to ra"-
idly scope the composition of ga a-emit ting sources
~
1,.97-20
TABLE 2 -
PMR 'IARIASLES conunued-Measured Variab1e Range Type Purpose
'OST-ACC IOEi'i'T SAMPLING CAPABILITY:
Primary Coolant Sumps Coatainment Air POST-ACCI DDT Ai'lAlYSIS CAPABILITY OPS ITE):
As required based an Reg Guide 1.4 guide-lines 1.
gamma-ray spectrum 2.
pH
- 3. hydrogen
- 4. oxygen
- 5. boron Z/A Provide means for safe and coaveaient sampling.
These provisions should include:
- 1. shielding to.maintain zadiation doses ~,
- 2. sample containers with cantainer-sampling port cannectoz compatibilit~
- 3. capability or sampling under pri~
system pressure and negative pressuri
- 4. handling and transpozt capability,
- 5. pre-arrangement for analysis aad interpretat'on.
METEOROLOGY:
Wind Direction 0 to 360'>>5'ccux'-
'cy with a deflection of 15'. Start.'ng speed 0.45 mps (1 mph)
Fox'etermining axfluent transoozt direc-tion for emergency
- planning, dose assess-
- ment, and source estimates.
Wind Speed 0 to 30 mps (67 mph)
(-0.22 mps '(0.5 mph) accuracy for wind spec less that 11 mps (25 m
ph), with a startiag threshold oi less than 0.45 mps (1 mph)
For determining effluent travel speed anc dilution for emergency planning,. doses assessments and source estimates.
Vertical Temperature Difference
-9': to +O'F (-0.3'F accuracy per 164 "oot intervals For determining effluent diffusion rates for emergency planniag, dose assessmeats aad sou"ce estates.
recipitation Recording rain gage with range suxficient to assure accuracy of total accumulat'on within 10/ ox recar&
ed value - 0.01" resolutioa For determining effluent transpor" and ground deposition for emergency pla.wing.
Hates xoz Table 2-(i ) Design flow the ma~am flaw anticipated ia normal ope ation.
(2) Design pressure that value correspond'ag to ~i"- cade values which aze oataiaed at or below code allowable material design stress values.
1.97-21
TABLE 3 -
BMR VAR!ABLES f'leasured Variable Range Tape Purpose CORE:
Control Rod Position Full in or not full in Prov'ide positive indication that the control rods a
e fully inserted.
(Minimum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> azter accident)
Neutron Flux 1 c/s to 1% power (at least one fission counter)
AlfS-4.5, Section 6.2.2 For indication of approach to cr'icality REACTOR COOLANT SYSTEM:
RCS Pressure 15 psia to 2000 psig B,C AHS-4.5, Sections 6.2.3, 6.2.4, 6.3.3 and 6.3.5 For indication of an accident and to in-dicate that actions must be taken to mitigate an event.
Coolant Level in the Reactor Bottom oz core suppor plate to above top of discharge plenum ASS-4.5, Section 6.2.3 For indication of fuel submergency for a
LOCA, event.
Hain Steamline Flow 0 to 120% design flow To provide an indication o'z -he integrity of the pressure boundary.
i~win Steamline isola-tion Valves 'eakage Control System Press-ure 0 to 15" of water 0 to 5 psid To provide an indication oz tne pressure boundary and containment integrity.
Primary System Safety Reliez Valve Posi-tions includ'ng ADS or Flow Througn or Pressure in Valve Lines Closed-not c'osed or 0 to 50 ps'g Bv these measurements the operator knows if there is a path open zor loss oz cool-ant and that an event may be in progress.
Radiation Level in Coolant 10 pCi/g to 10 Ci/g Abls-4.5, Section 6.3.2 For early indication of zuel c'add'ng failure and estimate of extent of damage.
- 1. 97-22
TABLE 3 -
BWR VARIABLES continued-Measured Variable Range Type Purpose CONTAINMENT:
Primary Containment Pressure 10 psia pressure to 3 times design press-ure for concrete; 4
times design pressure for steel B,C ASS-4.5, Sections 6.2.5, 6.3.3, 6.3.4, and 6.3.5 For indication of the intergrity of the primary containment pressure boundary; to indicate the potential for leakage from the containment.
Containment and Dry-well Hydrog en Concentration 0 to 10%
(capability of ooer-ating from 12 psia to maximum design press-ure B,C
, AHS-4.5, Sections 6.2.5, and 6.3.5 For indication of the need for, and a measurement of the performance of the containment hydrogen recombiner and to verify the operation of the mixing system Containment isolation Valve Position Closed-not closed B,D A.'IS-4.5, Section 6.2.5 To ind'cate the status of contairment isolation and to provide infozmation on the status of valves in process lines which could carry radioactive materials out of containment.
Suppression Pool Water Level Top'f vent to too of weir well ASS-4.5, Section 6.3.3 Suppression
?ool Water Temperature 50'F to 250'F To assure proper temperature. for IPSH of ECCS.
To verity the operation oz the makup system.
Drywell Pressure 12 psia to 3 psig 0 to 110% design pressure A'.lS-4.5. Sect'on 6.3.3 Diagnosis of impact of accident on dry-wall structure.
Drywell Drain Sumps Level (identizied and Unidentified Leakage)
Bo ttom to top B,C AHS-4.5, Section 6.3.3 High Range Contain-ment Area Radiation 1 to 107 R/hr (60 keV to 3
i~!eV pho-tons with =20% acc-uracy for photons of 0.1 to 3 MeV)[
10'/hr for pnotons is approximately equiva-lert to 10a r ds/hr for bet s and photons]
B,C To help identify 'f an accident has de-graded beyond calculated values and in-dicate 'ts magn'ude and to determ'ne act'on to protect the public.
1.97-23
TABLE 3 -
BWR VARIABLES continued-Measured Variable Range TJ'pe Puroose POMER CONVERSION SYSTEMS Main Peedwater Plow 0 to 110% design flow~
To indicate an adequate source of water to the reactor'.
Condensate Storage Tank Level Bottom to top To indicate available water for core cooling.
AUXILIARY SYSTEMS:
Containment Spray Plow 0 to il0% design flow~
D For indication of system operation.
Steam Flow to RCIC 0 to 110% design flow~
E To verify that adequate steam 's avail-able for the system to perform its function.
RCEC Plow RHR System Flow 0 to 110% design flow~
0 to 110% design flow-'
Por indication of system operation.
"."or indication oz system operation.
RHR Heat Exchanger Outlet Temperature 32P to 350P For indication of system ooeration.
Service Cooling Water Temperature 32'F to 200'P Por indication oz system ooerat'on.
Serv'ce Cooling Mater Plow Plow in UHS Loop 0 to 110% design flow~
0 to 110% design flow-'
Por indicat'on of system operation.
Poz indication of system ope ation.
Temperature in Ulti-mate Heat Sink Loop
-30'P to 150 F
Por indication of system opezation.
Ultimate Heat Sink Level Plant specific To ensure adequate source oz cooling water.
SLCS Storage Tank Level Bo'ttom to too To orovide indication of inventory for boron in]ection.or shutdown.
Sump Level in Soaces of Equipment Reauired for Safety To correspond'ng level of safety eouipment ailura To monitor potential for failure of equipment 'n closed spaces due to flooding.
1.97-24
TABLE 3 -.BWR VARIABLES continued-Measured Variable Range Tyoe Purpose RADWASTE SYSTEi~)S High Radioactivity Liquid Tank Level Top to bottom Available volume to store primary coolant.
Charcoal Delay Gas System Gas Flow or Radioactivity Level As required To monitor performance oz system.
VENTILATION SYSTEi~1S
'mergency Ventilatio Damper Position Open-closed status To ensure proper ventilation under accident conditions.
Temperature of Space in Vicinity of Equi-pment Required for Safety 30'P to 130'F To monitor environmental conditions of equipment in closed spaces.
POWER SUPPLiES:
Status of Class lE Power Supplies and Systems Voltages and, currents D
To ensure an adequate source of electric power for safety systems Status oz Non-Class 1E Power Supplies and Systems Voltages and currents E
To indicate an adequage source of electric power.
RADIATION EXPOSURE RATES INSIDE BUILDINGS OR AREAS WHERE ACCESS-IS REQUIRED TO SERVICF SAFETY RELATED EQUIP-MEi'IT:
Radiation Exposure Rates 10 to 10 R/hr 'or ohotons (pe~nently instal'-
ed monitors)
For measurement oz.high-range radiation',
exposure rates at various ocat'ons.
1.97-25
TABLE 3 -
BMR VARIABLES continued-Measured Variable Range Type Purpose AIRBORNE RAOIOACTIVE i4lATERIALS RELEASES FRC1 THE PLANT:
Effluent Radioactiv-ity - Noble Gases
.Containment Exhaus Vent and S>andby Gas Treatment Sys-tem Vent
.Other Release Points [including fuel handling bui-lding, auxiliary building, and tur-bine building]
(Normal plus accident range for noble gas) 10 to 10 uCi/cc Xe-133 calibration 10 to 10 pCi/cc Xe-133 calibration (permanently install-ed monitors)
A"fS-4.5, Section 6.2.6 To provide operator with information re-garding release of radioactive noble gases on a continuous basis.
Effluent Radioactiv-ity - High Range Radiohalogens and Particulates
.Untreated Effluent
.HEPA Filters, min-imum of 2" of TEDA impregnated char-
- coal, non-ESF sys-tems
.HEPA F'lte s,'in-imum of 4" of TEDA impregnated char-
- coal, ESF systems Environs Radioactiv-ity High Range Exposure Rata 10-'o 102 >Ci/cc 10 a to 10 uCi/cc 1.0 a to 1 VCf./cc (permanently install-ed monitors) 10 'o 302 R/hr (60 keV to 3 HeV)
(permanently ins tail-ed monitors)
To provide the operator with information regarding release of radioactive halogens and particulates.
Continuous collection.
of representat've samples followed by monitoring (measurements) of samples for radiohalogens and for particulates.
For estimating release rates or radio-active materials released dur'ng an accident from unidentified release paths (not covered by effluent monitors) continuous readout capability, approx'ately 16 to 20 locations site de-pendent.
Environs Rad'activ-ity - Radiohalogens and Particulates 10 to 10 p Ci/cc for both radiohalogen and particulates (permanently install-ed monitors)
E For estimating releases rates of radio-active materials released dur'ng an accident from unidentified release paths (not covered by effluent mon'ors).
Continuous collection of representative
'amples followed by monitoring (measure-ents) of the samples.
(Approximate'y to 20 locations)
I.c7-26
Measured Variable Range Type Puroose AIRBORNE RADIOACTIVE MATERIALS RELEASES FROM THE PLANT CONTINUED:
P'ant and Environs Radioactivity (portable instruments 0.1 to 10" mR/hr photons 10
~ to 10 " pCi/cc particulates 10
~ to 10 " uCi/cc
'odine Hi h Ran e
- 0. 1 to 10 R/hr photons 0.1 to 104 rads/hr betas and low energy photons 100-channel gamma-ray spectrometer During and following an accident, to mor; itor radiation and airborne radioactivity concentrations in many areas throughout the facility where is impractical to install stationary monitors capable of covering both normal and accident levels.
During and following an accident to rap-idly scope the composition of gamma-emitting sources.
POST-ACCIDENT SAMPLING CAPABILITY:
Primary Coolant Suppression Pool Containment Air POST-ACCIDENT ANALYSIS CAPABILITY ONS ITE As required based on Reg Guide 1.3 guide-lines 1.
gamma-ray spectrum 2.
pH
- 3. hydrogen 4.
ozygen
.f/A Provide means for safe and convenient sampling.
These provisions should include:
- 1. shielding to maintain radiation doses AL~,
- 2. sample contaners with container-sampling port connector compatibility
- 3. capability of sampling under primary system pressure and negative pressure
- 4. handling and transport capability, an
- 5. pre-arrangement for analysis and interpretation.
'ETEOROLOGY:
lfind Direction 0 to 360'-5 accur-acy with a deflection of 15'. Starting speed 0.45 mps (1 mph)
""or determining azfluent t=ansport direc-tion for emergency
- planning, dose assess-
- ment, and source estates.
hind Speed 0 to 30 mos (67 mph)
(-0.22 mps '(0.5 mph) accuracy for wind spec less that 11 mps (25 m ph), with a starting threshold or less than 0.45 mps (1 mph)
Por dete~ning effluent travel speed and dilution for emergency planning,. doses assessments and source est~~ates.
I.97-27
l TABLE. 3 -
BNR VARIABLES inued-Measured '/ariabIe Range I/pe Puroose t 1ETEOROLCG'(
COi'ITIi'lUED:
Vertical Temperature Difference
-9'2 to -'.9'2 (-0.3'7 accuracy per 164 foo t intervals 2or determining effluent diffusion rate=
for emergency planning, dose assessments and source estimates.
2recipitation Recording rain gage with range sufficient to assure accuracy oz total accumulation within 10/ of record-ed value - 0.01" resolution 2or determining ezfluent transport and ground deposition for emergency piannin"-
Notes for Table 3
('-) Design flow - the max~mum flow anticipated 'a no wl operation.
(2) Design pressure - that value correspond~kg to ASM code values wh'cn are ootai ed at or below code allowabl material design stress values.
l.97-28
ANS-4.5 Oraft 3 September, 1979 ORAFT CAUTION NOTICE:
This Standard is being prepared or reviewed and has not been approved by AHS. It is subject to revision or withdrawal befor issue.
ORAFT American National Standard FUNCTIONAL REQUIRBIENTS FOR POST ACCIOEHT MONITORING CAPABILITY FOR THE CONTROL ROOM OPERATOR OF A NUCLEAR POWER GENERATING STATION Assigned Correspondent T. F. Timnons Mestinghouse Electric Corporation Power Systems Company
'.O.
Box 355, MNC-410 Pittsburgh, PA 15230 L
Writing Group ANS 4.5 Standards Cormni ttee NUPPSCO Secretariat ANS
TABLE OF CONTENTS FOREAORO 1.0 Introduction 2.0 Scope 3.0 Oefinition 4.0 Oiscussion 5.0 Oesign Basis 6.0 Oesign Cr it ria - Phase I and Phase II
I'
FORPAORD ANS 4 established Working Group 4.5 in late July 1979 to prepare a draft standard on Accident Monitoring Instrumentation which would complement other standards, but be broader in nature by including economic consid-eraiions.
Two primary objectives were 1) to address that instrumenta-tion which permits the operator to monitor expected parameter changes in the accident per iod, and 2) to address extended range instrumentation deemed appropriate for the possibility of encounte'ring previously unforeseen events.
ANS 4.5 began work on July 30th and met for 13 working days in a seven week period.
In addition, a Design Criteria subgroup met for two days in this same period.
As presented, this draft standard provides:
1.
a list of functions to be performed (design basis section 5.0) 2.
a framework.o determine those variables to be monitored (design basis section 5.0) 3.
an identification of thre time periods of interest (definitions 3.0) 4.
an identification of our variable types (definitions 3.0) 5.
a delineation of applicable design criteria for the variables to be monitored (design criter ia sec ion 6.0)
Ho identification of specific Type A monitored variables is provided in this standard.
Reccrmendations for Type 3 and Type C monitored vari-ables are provided in Section 6.0.
The significant issues in the development of this standard have been:
the scope of the document in terms of applicability to the control room operator or the plant operator (licensee).
The work group chose a control room operator scope.
2.
the pre-planned operator actions designated by the accident analyses in Chapter 15 of a plant's FSAR and the not previously planned operator action that may be required during unforeseen events.
The Morking Group established Type A instrumentation for the
- ormer, and Type 8 or C instrumentation for the latter.
3.
The monitoring of actual fission product barrier integrity and the potential for breach of a given barrier.
The work group chose monitoring of actual breach for the uel, reactor coolant syst m,
and containment barrier, bu only the potential breach of the con-tainment barrier.
4.
the degr ee of alignment of accident monitoring instrumentation with IEEE Class lE (AHS Class EC-3) and whether an intermediate class is needed between 1E and non-1E.
whether a list of variables should be included as an appendix to the s.andard:
a.
a list of only Type C parameters b.
a list of Type A, 8, C and 0 parameters 6.
the definition of ins.rument tyoes 8 and 0 and whether these types should be included in the s andard.
The membership of the 'Aorking Group is as follows:
L. Stanley, Chairman 7.
Timmons, 'lice Chairman and Correspondent
0.
Scmmers E.
Menzinger'.
Lambert R. Bauerle J.
Castanes M. Molpert H. itumford X. Pol anski E. Oowling Additional input has been provided to the Morking Group by industry, university, and government participants throughout the me tings.
The Mork Group is very appreciative of this assistance.
1.0 Introduction The Code of Federal Regulations requires that instrumentation be provided to monitor variables and systems over their anticipated ranges for acci-dent conditions as aporopriat to assur adequate safety.
The purpose of this standard is to establish criteria or the selection or that instru-mentation.
These criteria are based on the sequence and duration of the phases through which an accident progresses.
The control room operator may have different information requirements for each ohase of an accident.
This standard presents criteria for monitoring the response of the plant to design basis events.
It also presents cri.eria for monitoring the integrity of fission product barriers under conditions wnich have degraded beyond the design bases.
This fission product barrier monitoring is con-sideredd to be an extra set of instrumentation beyond that required for satisfac.orily monitoring accident scenarios postulated in the plant safety analysis.
Throughout these criteria, thro e verbs have been used to indicate the degree of rigor intended by the specific criterion.
The word "shall" is used to denote a requi"rement; the word "should" to denote a recommenda-tion; andthe word "may" to denote permiss'on, nei.her a requirement nor a
recommendation.
This standard contains criteria for determining the variables to be moni-tored by the control rocm operator, as required for safety, during the course of an accident and during the long-term stable shutdown phase fol-lowing the accident.
Also included are criteria for determining the requirements for the equipment used to monitor those variables.
The scope of the standard is limited to onsite environment and process monitoring.
emergency preparedness planning is, or will be, covered by other standards.
3.0 OEFINITIOilS Phase I
That portion of the accident extending from the initiation of the accident to that point at which the plant is in a con-trolled condition.
Phase II That portion of the accident extending from the time at which the plant is in a controlled condition to that point at wnich personnel access to the location of the accident is possible.
Phase III That portion of the accident extending from the time at which personnel access to the location of the accident is possible to the time at which the plant has returned to operating status or been deccmmissioned.
Type A Instruments - Those ins.ruments which provide the information required to permit the control room operator to take the ore-planned manual actions to acccmplish safe plant shutdown for design basis accident events and to maintain long term plant stability.
Type B
inose instruments which provide to the control room operator information to monitor the process of accomplishing critical safety unctions, i.e., reactivity control, core cooling, maintaining reactor coolant system integrity, maintaining containment integrity nd radioactive effluent control.
Type C
Those ins.ruments that indicate in the control room (I) the extent to which parameters, which have the potential for causing a breach of he final fission produc. barrier (i.e.,
the containment),
have exceeded the design basis values, or (2) that a fission product barr ier (i.e., fuel clad, reac.or coolant pressure boundary or the containment) has been breached.
Type,0 Those instruments which indicate to the control room operator the performance of individual safety systems.
Oesion Basis Accident Fvents Those events postulated in the plant safety analyses, any one of which may occur during the lifetime of a particular plant, excluding those events which are expected to occur during a calendar year. for a par'ticular plant; and those events that are not expected to occur but are pos.ulated in the plant safety analyses because their consequences would include the pot n-tial for the rele se of significant amounts of radioac ive material.
4.0 DISCUSSION It is the ohilosophy of this Standard that instrumentation is requir d
to monitor plant performance during and after an accident.,
The purposes of the accident monitoring instrumentation are'numerated in Section 5.0, Oesign 8asis.
This Standard specifies the plant safety functions to be performed and the criteria to be used by tho designer in selecting the variables to be monitored.
Certain concepts have been established to aid the system designer in the seleciion of variables to monitor the course of an accident and io arrive at appropriate design criteria for instruments to monitor these variables.
4.1 Planned Versus Unplanned Operator Actions I
The plant safety analysis defines the accident scenarios from which the safety system design bases and the planned or anticipated operator actions are derived.
Accident mani ioring instrumentation is provided to permit the operator to take required actions to address ihese anaIyzed.
situations.
However, instr.mentation must also be provided for unplanned situations, (i.e., io ensure that, should plant conditions evolve differently than predicted by the safeiy analysis, the oaerator has sufficient information io monitor the course of the event).
Irsiru-mentation mus. also be provided to indicate to the oper ator if fission product barrier integrity has degraded beyond the prescribed limits of the Safety Analysis.
.2 Four classifications o7 variables have been iaentified.
Operator manual actions during accidents included in the plant safety analysis are anticipated or pr -planned.
Those variables thai provide information needed by the ooer ator to perform these manual ac ions are designated
Type A.
Those variables ne ded to assess that the plant saf ty functions are being accomplished, as identified in the olant safety analysis, are designated Type 8.
Variables used to monitor for the actual gross breach of one of the fission product barriers or the potential breach of the final fission product barrier (containment) are designated Type C.
Type C variables used to monitor the potential'reach of containment have an arbitrarily-determined, extended range.
Tne fourth classification, Type 0, consists of those variables monitored to ascertain that the safety systems are performing as designed.
Type 0 variables are less important than Types A,
B and C for accident monitoring since safety system per-formance only infers safety function accomplishment.
Type 0 variables and instruments are not considered to be within the scope of Accident i4lonitoring Instrumentation.
Guidance on the selection of Type 0 vari-ables and he specification of appropriate design criteria are not given in this standard.
This guidanc is provided in standards for design of safety syst ms (e.g.
IEEE-603, ANSI H18.2, etc).
The four classifica-.
tions are not mutually exclusive in that a given variable (or instrument) may be included in one or more types.
This differentiation by variable type is intended only to guide the designer in his selection of accident monitoring variables and applicable criteria.
4.3 Accident Phases Tne typical accident sequence has been subdivided into thre phases:
Phase I covers the initial portion of the accident, Phase II covers the stable long-term cooling portion of the accident up to the time where personnel access is possible,'nd Phase.III addr sses the period ollow-ing personnel access to the accident area.
This sub-division has been made so that variable selection and design criteria apolic"ion can reflect the differing conditions which characterize these three phases.
For example, Phase I can be anticipated to be of relatively shor dura-tion,. having relatively severe plant condi ions, and allowing no person-nel access to the accident area.
Phase II is expected to be of longer duration, to require a significan.
number of operator actions, under milder plant conditions, but with sti'il no personnel access to the acci-dent area.
Phase III is expected to be of even longer duration where
personneI acc ss is possib1e.
Oi==erent design crit ria are then appro.-.
priate for each of the three phases.
In this Standard, gUidance and criteria are provided =or Phases I and II.
The plant designer shall per-"orm and document an analysis to select acci-dent monitoring instruments.
He shall identify instruments required by his design to enable the control room operator to:
A.
Perform pre-planned manual actions.
8.
Ascertain the per=ormance o7:
(1)
Reactivity control (2)
Reactor core cool ing (3)
Reactor coolant system integrity (4)
Containment integrity (5)
Radioactive ef=luent control C.
Ascer.ain the extent to which parameters, which have the potential ror causing a breach o7 the containment, have exc ded the design basis values and to ascertain that a fission product barrie. (i.e. fuel clad, reactor coolant system pressure boundary or the containment) has be n breached.
- 5. 1 Variable Selection or Phases I and II The proc ss for selection of the Accident llonitoring Ins.rumentation vari-ables shall include:
5.1.1 For Type A
')
Identification o7 the postulated accidents for which manual action is required.
2)
Identification of planned operator actions 3)
Identi7ication of the monitored va. iables ne d d -or planned operator actions.
5.1.2 For Type 8
1)
Identification of the monitored variables that provide the most direct indication ne ded to assess the accomplish-ments of:
a.
Reactivity Control b.
Reactor Core Cooling c.
Reactor Coolant System Integrity d.
Containment Integrity e.
Radioactive Effluent Control Guidance on the selection of these variables is provided in Section 6.0.
5.1.3 For Type C
1)
Identification of the monitored variables that provide the most direct indication of a gross breach of a fission product barrier or of an approach to breach of the con-tainment.
These instruments may have extended ranaes.
Guidance on the selection of these variables is provided in Section 6.0.
5.1.4 Phase III Acc ss Prior to he termina ion of Phase II, the ability o gain access o the location of the accident must be determined.
Ins.rumentation that indicates when conditions are accep able for personnel access shall be identified.
5.2
'PERFORMANCE REQUIREMENTS FOR PHASES I ANO II The process for determining performance requirements of Accident Moni-toring Ins.rumentation shall
- include, as a minimum, the following con-sidera ions:
5.2.1 For Types A and S
1)
Identification of the 2)
Identification of the 3)
Identification of the 4)
Identification of the ment is needed.
5)
Identification of the ment must operate.
expected range of the process variable.
required accuracy of measurement.
required response characteristics.
time interval during which the me sure-local environment in which the instru-5.2.2 For Type C
The performance requirements for these instruments are arbitrary.
Guidance on these requirements is provided in Section 6.0.
10
6.0 OESIGN CRITERIA
'.1 GENERAL OESIGN CRITERIA 6.1.1 SEISMIC QUALIFICATIONS Accident monitoring ins.rumentation that is to be seismically qual ified shall be qualified according to IEEE Standard 344-1975.
The instiumenta-tion shall be qualified to continue to function within the required accuracy following, but not necessarily
- during, a safe shutdown earth-quake.
Vibration loads which occur as a result of plant syst m operation during any phase for which the instrument is required shall be considered.
- 6. 1. 2 OURATION Accident monitor ing ins.rumentation shall be qualified for the length of time its function is required.
Unless other times can be justified, Phase II instrumentation shall be qualified to function for not less than 100 days.
A shorter time may be acceptable if instrumentation equipment replacement or repair c
n be accomplished within an accept ble out-of-service time, taking into consideration the environment wher e.he equip-ment is located.
6.1.3 DIRECT MEASUREMENT To the extent practical, accident monitoring instrumentation inputs shall be from sensors that dire tly measure the desired variables.
- 6. 1..0 MINIMIZINGMEASUREMENTS To the ex ent practical, the same instruments shall be used for accident monitoring as are used for the normal operations of the plant to enable the operator to use, during an accident situation, instruments with which he is most familiar.
- However, where the required range o
acciden moni-toring instrumentation results in a loss. of instrumentation sensitivity in the normal operating
- range, separate instruments shall be used.
6.l. 5 INSTALLATION Permanently installed instrument equipment is required =or those instru-ments required to function during Phase I.
Permanently installed instru-mentation systems need not be provided for those functions required only foW Phases II and III providing it can be demonstrated that the instru-ment components can be installed when required, considering the local environment.
6.1.6 INSTRUi4IENTATION LOCATION ANO IDENTIFICATION Accident monitoring instrumentation shall be located acc ssible to the ope. ator and be distinguishable from other displays so that in an acci-dent situation, the operator can rapidly identify the accident monitoring instrumentation.
6.1.7 EQUIPMENT REPAIR The accident monitoring instrumentation shall be designed to facilitate timely recognition, location, replacement, and repair or adjus~nent of malfunctioning equipment.
6.1.8 TEST ANO CALIBRATION
- 6. 1.8. i Tes Capability shall be provided for testing, with a high degree of confi-
- dence, the opera ional availability of each instrument channel during plant operation.
This may be accomplished in various ways, for example:
1.
By observing the effect of perturbing the monitored variable.
2.
By observing the effect of introducing and varying, as aporopriate a
substitute input to the sensor of the same nature as the measured variable.
3.
Sy cross-checking between channels that bear a known relationship to.
each other.
Where testing during reac.or operation is not possible, it must be shown that there is no practical way of impIementing such a requir ment without adversely affec.ing plant safety or operability.
In addition, it must be shown that the probability of a failure of the ccmponent which is not periodically tested is acceptably low and tha such testing can be rou-tinely performed when the reactor is shut down.
6.1.8. 2 Cal ibrat ion Capability shall be provided for calibration of each instrument channel during normal plant operation or during shutdown as determined by the required interval between calibrations.
Equipment that does not require periodic calibration is exempt from this requirement.
- 6. 1. 9 OI'/ERS i-TY Oiversity i's preferred in fufilling redundancy requirements.
6 ~ 1.1G REOUNOANT REAOOUT Ai4!BIGUETY Where a disagreement between redundant displays could lead the ope. ator to defeat or fail to accomplish a required safety unction, additional information shall be provided to allow the operator to deduce the actual conditions that are required
=or nim to perform his role.
This may be accomplished by providing an independent channel which monitors a dif-ferent variable bearing a known relationship to the redundant channel or by providing an additional independent channel of ins-'. umentation of the same variable or by providing the capability for the operator to perturb the measured variable and determine by observation of the res onse ~hich instrumentation display has failed.
6.2 TYPE 8 INSTRUMENTS 6.2.1 GENERAL RE(UIREMENTS The number of instruments used shall be only that minimum set needed to adequately monitor the accomplishment of the following functions:
a.
Reactivity Control b.
Reactor Core Cooling c.
Reactor Coolant System Integrity d.
Containment Integrity e.
Radioactiv Effluent Control Type 8 ins.ruments provide control room indication beyond that which may be required for any preplanned operator action and as such constitut an extra set of instrumentation which results in an additional layer of protection.
6.2.2 VARIA8LES FOR REACTI'/ITY CONTROL MONITORING Tne measured variable shall indicate the accomplishment of control of reactivity in the core.
The measured variable should be neutron flux.
Tne range of measurement should extend frcm one count pe.
second on the source range instrument to the intermediate range instrument value cor-respondingg to ~.".. of full reactor power.
This range is intended to encompass all neutron flux levels at which the core can be subcritical.
6.2.3
'lARIABLES FOR CORE COOLING MONITORING The measured variables shall indicate the accomplishment of core cool-ing.
For the PMR, the measured variables should be TH, TC, pres surizer level, and pressurizer pressure.
For the 8NR, the measured variable should be reactor vessel water level.
Incor e thermocouple monitoring should be considered
=or inclusion as a desireable variable to ascertain cooling.
6.2.4
'IARIABLES FOR REACTOR COOLANT SYStci4l INTEGRITY The measured variable shall indicate the accomplishment of RCS Inte-grity.
The measured variable should be primary system pressure.
6.2.5 VARIABLES FOR CONTAINi4IENT INTEGRITY The measured variables shall indicate the accomplishment of containment integrity.
The measured variables should be containment hydrogen con-centration, containment pressure and containment isolation valve posi-tions ..
6.2.6 VARIABLES FOR RADIOACTI'/E EF LUENT CONTROL The measured variables shall indicate the acccmplishment of radioactive effluent control.
The measured variables shoul,d be noble gas monitoring of the identified plant release points.
6.3 TYPE C
INSTRUMENTS 6.3.1 Type C ihstruments shall meet the following cri.aria:
6.3.i. I The number of instruments used shall be only that minimum set needed to adequately monitor the three barriers; 6.3.1.2 Each measurement shall be as direct as possible; 6.3.1.3 Any chosen measurement shall detect a gross breach of one or more barriers (i.e.,
> I pe. cent fuel clad failure, a RCS pressure boundary breach producing a loss of reactor coolant inventory in excess of the normal makeup capabi li.y, a con-tainment breach capable of producing radiation rele sas in excess of 10 CFR 100 at the site boundary using TIO-14844 source terms);
the ranges es ablished for Type C instruments are not mechanistically related to
- a. postulated accident sc nario.
6.3.1.4 Ouring the period of ne d for Type C instruments, no othe".
failures shall be assumed in the analysis beyond the assumed bre ch of a barrier coincident
~1th loss of off-si e po~er; 6.3.2 Fuel Clad Sarrier Monitor in 6.3.2.1 The measured variable shall detect and alarm the breach of the fuel clad barrier (i.e.,
> 1 percent fuel clad,ailure);
6.3.2.2 Operator sampling of reactor coolant'hall be used as the means to verify the measured variable alarm.
6.3.2.3 The measured variable should be reactor coolant system radia-tion.
Tne instrument range should be equivalent to the fuel clad gap activity corresponding to 0.5~ to 5" failed uel.
A narrow accuracy band for this measured variable is not signi-ficant in achieving ihis function; for examole,
~0'to -.'~00,.
accuracy of reading should be accep able.
instrument tran-sient response should be compa.ible with its recorder.
6.3.3 Reactor Coolant Svsiem Pressure 8oundar blonitorin Qe3o3
~
The measured variable(s) shall detect and alarm a breach of ihe reactor coolant system that produc s
a loss of coolant inventory in excess of normal makeup capability.
The spectrum of RCS oressur boundary breaches extends uo to and includes the largest double-ended pipe break.
6.3.3.2 Tne means used to detec.
RCS pressure boundary breach should include one RCS pressure boundary variable and one containment variable over the full spectrum of break sizes.
6.3.3.3 The measured PriR variables should be RCS pressure and contain-ment pressure.
The instrument range should be the design pressure plus a specified marg n (< ~C~}.
tiormaI instrume..t accuracy is acceptable for ihese moni iors.
instrument iran-sient response should be c:moatible uith its recorder.
16
6.3.3.4 The measured SNR variables should be drywell pressure and containment sump level.
The instrument range should, be design values pIus a specified margin (< 10").
Normal instrument accuracy is acceptable for these monitors.
Instrument tran-sient response should be compatible with its recorder.
6.3.4 Containment Pressure Soundar Monitorin 6.3.4.1 The measured variable(s) shall detect and alarm a breach of the containment pressure boundary that is capable of producing radiation releases in excess of 10 CFR 100 at the site boundary using TIO-14844 source terms.
6.3.4.2 The means used to detect containment pressure boundary breach should include containment pressure (8~BR and PMR), environs radiation monitoring for gross gamma (PriR),
and secondary containment air space radiation monitoring for gross gama (SMR).
6.3.4.3 The instrument range for containment pressur should be design pressure pIus a specified margin (<
10>o)
Normal instr ment accuracy is acceptable for this monitor.
instrument transient response should be compatible with its recorder.
6.3.4.4 The instrument range for environs radiation monitoring should be 10-3 to 102 R/hr.
The instrument range for secondary containment air space radiation monitoring should corr spond to the 10 CFR 100 value for off-site doses.
instrument accuracy should be
~ 1/2 decade (100 'Kev-3 Mev).
instrument transi nt response should be compatible with its recorder.
6.3.5 Potential Sreach of the Final Fission Product Sarrier 6.3.5.1 The measured variables should be containment
- pressure, con-
,tainment hydrogen concentration, and RCS pressure for indicating the potential for causing a breach of the final fission product barrier (i.e., containment).
6.3.5.2 An arbi irary range of 3 times design pressur for concrete and 4 iimes design pressure for steel should be used for contain--
ment pressure.
Instrument accuracy should be
+ 10" of span.
Instrument transient response should be compatible with its recorder.
6.3'.5.3 An arbitrary r ange of 0-10 volume percent hydrogen should be used for containment hydrogen concentration.
Instrument accuracy should be +
10'~ of span.
Instrument transieni response should be compatiable with its recorder.
6.3.5.4 An arbitrary range of 1.5 times design pressur should be used for RCS pressure.
instrument accuracy should be
+ 10~ of span.
Instrument transient response should be compatible with its recorder.
6.3.6 ITS iRUMEi<T QUALIFICATION 6.3.6.1 Type C insiruments shall be qualified in the same manner as Type A instruments except:
6.3.6.1.1 For purposes o7 equipment qualification, the assumed maximum value of the monitored parameter shall be ihe value equal to the maximum range for the instrument.
The monitored parameter shall be assumed to approach this peak by extrapolating the most severe initial ramo associated with the Oesign Basis Accidents.
The decay for this paramet r shall be considered proportional to the decay for this parameter associated with the Oesign Basis Accidents.
No additional qualification maroin ne ds to be added to the extended r ange parameter.
See figur 6.3-1.
A'll environmental envelopes except thai per-taining to the parameter measured by the instrumeni shall be ihose associated with the Oesign Basis Accidents.
.4 SPECIFIC DESIGN CRITERIA Oesign Criteria specIfic to particular accident phases and variable types are pr sented in Table 6.4-1.
TAOLE c.n.I DESIGN CRITERIA CR ITER IOtl PIIASE 1
VARIABLE TYPE 0
PIIASE II NRIABLE TYPE 8
I.. fi<<alify seismically to IELE 344-75 (operate after SSE)
Yes Yes No Yes tlo No 2.
Heel. single failure per ICEE 379-77 Yes Yes tlo Yes Yes tlo 3.
$<<a 1ify env ironmen-tally to IEEE 323-74 Yes Yes Yes(')
Yes Yes Yes Consiiler loss of off-site power 'es Yes Yes Yes No No 5.
Power source Emergency oner g.
Emerg.
Emerg.
tlorma 1 Normal 6.
O<<t of service interval (2)
- prior to accident (2)
(2)
(2)
<72 llrs(
~
7.
0<<t of service inter-Hone va 1 - il<<ri>>g acc i ~le>> t tlone
<2 llr (2)
(2)
<2 llrs
TABLE 6. 0-1 (Cont in(ted)
DES ICiN CR ITEAIA CA ITEllION PilASE 1
NRIAOLE TYPE 0
PifASE II VAAIMLE TYPE 0
0.
i'ortable instrumenta-No tion No Yes Yes Yes 9.
Level of quality assurance ANSI N45.2 ANSI e5.2 ANSI W5.2 ANSI ~W5.2 NISI e5.2 ANSI N~5.2 10, Display type (41 Continuous Continuous Continuous Continuous Continuous On demand
- 11. Display method (5)
Recording According Indicator Recording' Indicator indicator (5) 12.
Ident. ifIcation as accident nenitoring type Yes
. Yes Yes Yes
- 13. Periodic Test per IEEE-330-1977 Yes Yes Yes Yes Yes Yes NOlES:
(1)
See Paragraph 6.3.6 of this Standard.
(2)
IEEE 279-1971 Paragraph l. 11 Exemption
NOTES TO TABLE 6.4-1 (Continued)
- 3) Based on normal lech spec requirements on out-of-service safety systems.
0f Continuous indication or recording displays a given variab)e at all times; intermittent indication or recording displays a given variable periodically; on demand indication or recording displays a given variable only ~ihen requested.
(5 ilhere trend or transient information is essential to planned operator actions.
6 tray be manually connected to emergency buss I
Radiation monitoring outside containment may be portable.
Figure 6.3-1.
Typical Environmental Rualification Envelope for Type C Instruments Pa rame Ler
)
I II/lI Assumed Parameter Environment 1
Design Oasis Accident Time