ML19338E719
| ML19338E719 | |
| Person / Time | |
|---|---|
| Issue date: | 09/17/1980 |
| From: | Hintze A NRC OFFICE OF STANDARDS DEVELOPMENT |
| To: | NRC OFFICE OF STANDARDS DEVELOPMENT |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097, TASK-OS, TASK-RS-917-4 NUDOCS 8010060047 | |
| Download: ML19338E719 (38) | |
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NUCLEAR REGULATORY COMMisslON k
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WASHINGTON. D. C. 20555 1
"l September 17, 1980 MEMORANDUM T0:~ Distribution
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FROM:
A. S. Hintze Reactor Systems Standards Branch Division of Engineering Standards-Office of Standards Development
SUBJECT:
REVISION 2 TO REGULATORY GUIDE.1.97
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This memorandum is ' summary report of the meeting held September 5,1980 a
between industry and NRC staff personnel in which the scope of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and _ Environs Conditions During and Following an Accident," was discussed. Those on the distribution list sere in attendance. After consider-
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able discussion, the following conclusions were runrded from the meeting:
- 1. The regulatory guide should address requirements of the control room operating personnel.
It was concluded that the development of the emergency response facilities (SPDS, TSC, EOF, NDL) were not sufficiently along to include identification of specific variables in Regulatory Guide '
1.97 as being used by the emergency response facilities. Additionally,
?he emergency response facilities are to be designed by the licensee using general criteria provided by NRC, which is inconsistent with NRC specifying the variables to be used by those facilities. Bill Coley of Duke Power Company supplied some words for consideration. (See Attachment A).
- 2. The definition of Design Basis Accident Events should be as defined in ANS-4,5. Regulatory Guide 1.97 should not take exception by including anticipated operational occurrences in the guide definition.
- 3. The definition of Type A variables should be modified to more specifically identify the ope, tor actions for Type A consideration.
John Gallagher of Westinghouse supplied some words for consideration.(See Attachment B).
- 4. It should be made clear that the potential for breach of the fission product barriers be limited to the energy sources within the barrier itself.
- 5. The Type D & E variables should be made a separate part or section of the regulatory guide.
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- 6. The discussion of the guide should explain how the design and qualification criteria category (Category 1, 2 or 3) was selected for each variable.
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To: Distribution 2-September 17, 1980 The discussion of the meeting was limited (by time) to the scope of the guide.
Time did not pennit any discussion of the criteria which should be included in Categories 1, 2 and 3, or the list of variables in Tables 2 and 3, or the assignment of a cat.egory to a given variable.
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Distribution List NAME Organization Dieter Fischer General Electric Donald W. Miller Ohio State University Loren Stanley Quadrex/ Nuclear Services (ANS 4.5)
David Cain [
ERPI/NSAC Jim Watt NRC/ DST / RYSCB Leo Beltracchi NRC/DHS/HFEB David Sommersh Consumers Power Company Phil Stoddart NRC/DSI/ETSB Tom Dunning NRC/DSI/ICSB D. F. Sullivan NRC/RSSB/OSD E. F. Dowling Babcock & Wilcox Rod Satterfield NRC/ICSB Warren Minners NRC/ DST Steve Ramos NRC/EPDB Jan Reston-Smith Lawrence Livermore National Lab L. Rolf Peterson Lawrence Livermore National Lab W. V. Johnston NRC/DSI/CPB Bill Coley,-
Duke Power Company John '.iallagher Westinghouse i
PatMck Higgins' HIF j
Alan S. Hintze NRC/OSD E. C. Wenzinger NRC/0SD Henry Martin Nuclear Safety Associates, Inc.
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1-Attachment B j
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Inacoropriateness of_
Class IE Emergency Response Facilities The Regulatory Position in the Regulatory Guide references ANS 4.5's definition of type A, B, and C variables and the associated general criteria.
As currently defined, both types A and B cover many functions that are performed by the Emergency Response Facilities (Technical-Support Center, etc.) and consequently can lead to the application of requirements in section 6.0 of ANS 4.5 and Table 1 of draft Regulatory Guide 1.97 Revision 2 to the Emergency Response Facilities.
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requirements would impose inappropriate Class 1E qualification and design r,riteria on these facil,ities.
In addition the definition of type A variables can lead to the applica-tio'n of these requirements to any instrumentation circuits which provide infonnation to the operator that are identified in written procedures (pre-planned manual actions), independent of whether the action is required for safety purposes.
We"believe that these potential problems can be corrected by the follow-ing modifications:
m
- Mod'ify the definition of type A variables to read: -
a.
Type A variables' are those variables to be monitored that provide
- the primary information required to permit the control room operator to take the specified manually controlled actions for which no auto-s matic control is provided and which are required for safety systems
(
to accomplish their safety functions for design basis accident i
events. -
Primary information is that which is essential for the direct accom-plishment of the specified safety functions and does not include those variables which are associated with contingency actions that may also be identified in written procedures.
.b. ' Change the scope of draft Regulatory Guide 1.97 Revision 2 to limit the application of the requirements for equipment to that part of the instrumentation system and its' vital supporting features or power sources which provide the direct display of the process vari-ables._ TebT1-should-contain a. note 1. hat these requirements are not applicable to instrumentation systems provided as operator aids for the putrase o enhancement of information presentations for the identification or diagnosis of disturbances.
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Contact:
A. S. Hintze, (301) 443-5913
[PROPOSEB] 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, " Domestic Licensing of Production and Utilization Facilities," includes a requirement that instrumen-tation be provided to monitor variables and systems over their anticipated ranges 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, 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 10 CFR Part 50 includes a requirement that means be provided for monitoring the reactor containment atmosphere, spaces containing components for reciiculation 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.
8.
DISCUSSION Indications of plant variables and status of systems important to safety are required by the plant operating organization (licensee) during accident situations to (1) provide information required to permit the operator to take i
preplanned manual actions to accomplish safe plant shutdown; (2) determine i
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 l
will enable him to determine the potential for causing a gross breach of the l
barriers to radioactivity release (i.e., fuel cladding, reactor coolant pressure boundary, and containment) and if a gross breach of a barrier has occurred; (4) furnish data regarding the operation of plant safety systems in order that the operating organization can make appropriate dicisions as to their use; and l
(5) provide infomation 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 the impending threat.
As an aid to the plant operatino oraanization to accomplish the above in carrying out its role to mitigate the consequences of an accident and to protect l
the health and safety of the oublic, <four emeroency response facilities have been identified as beino essential. These facilities are: (1) the' Safety Parameter Display System (SPDS), (2) the onsite Technical Support Center (TSC), (3) the Emergency Operations Facility (EOF), and (4) the Ntelear Data Link (NDL). The primary function of the SPDS is to help operating personnel in the control room make cuid assessments of plant safety status.
It will helo determine whether plant safety functions are beino performed. The crimary function of the TSC is to orovide overview monitorina of plant safety carameters to ensure that actions are taken oromotly to correct or miticate any abnormalities.
The primary function of the EOF is to provide coordination of radiological assess-c l
ments, manage overall emergency response and recovery operations, and provide assistance in the decision making process to protect the safety of the public.
The primary function of the NOL is to transmit data to the NRC in order that The NRC staff can have information to make an independent evaluation of the situation as an aid to carrying out its role in advisina the licensee and in-feminofofficials and the general public on all pertinent aspects of an acci-d4nt.
2
f 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, to 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 variables required to enable the operation of manually initiated safety systems and other appropriate operator actions involving systems important to safety.
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Independent of th,e above tasks, it is important that the operator be informed if the barriers to radioactive materials release are being chal-1enged beyond what was anticipated in the safety analysis..
Therefore.it is t
essential that, instrument ranges.be selected such that the instrument will always be.on-scale.
Narrow-range instruments may not have the necessary range to track the. course of the accident, consequently,. multiple instruments with overlappinq ranges may be necessary.
(In the past, some instrument ranges j
have been selected based on the set point value for automatic protection or alarms.)
It is essential that degraded conditions an.d their magnitude be identified so that the operator can'take actions that are available to mitigate the consequences.
It is not intended that the operator be encouraged to prematurely circumvent systems important to safety but that he be adequately informed in order that unplanned actions can be taken when necessary.
j 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), overpreisure transients, anticipated transients without 3
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e scram (ATWS), and reactivity excursicns witich result in releases of radioactive materials. Such events require that the operator understand, within a short time period, the ability of the barriers to limit radioactivity release, i.e.,
the potential for breach of a barrier, or the actual breach of a barrier by an accident in progress.
j 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.
It could therefore either be designed to withstand the acci-dent environment or be protected by a local protected environment.
It is important that accident-monitoring instrumentation components and l
their mounts that cannot be located in Seismic Category I building
- N designed i
to continue to function, to the extent feasible, during seismic events.
Con-sequently, it is essential that they be designed to resist the effects of seismic excitation. An acceptable method for demonstrating the adequacy of the seismic resistance of this instrumentation would be to qualify it to meet the seismic criteria applicable to instrumentation installed at other loca.tions in the plant.
Variables selected for accident monitoring can be selected to provide the essential information needed by the operator to determine if the plant safety functions are being perfomed.
It is essential that the range selections be sufficiently great that the instruments will always be on scale.
Further, it is prudent that a limited number of those variables which are functionally significant (e.g., containment pressure, primary system pressure) be monitored by instruments qualified to more stringent environmental requirements with ranges that extend well beyond that which the selected variables can attain under limiting conditions; 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 un-aware as to the pressure inside containment.
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.
It is also necessary to be sure'that when a range is extended, the sensitivity and accuracy of the instru-ment are within acceptable limits.
Normal power plant instrumentation remaining function ' for all accident conditions can provide indication, records, and (with certain types of instru-ments) time-history responses fur many variables important to following the course of the accident. Therefore, it is prudent to select the required accident-monitoring instrumentation from the nomal power plant instrumentation to enable 4
the operator to use, during accident situations, instruments with which he-is most familiar. Since some accidents impose severe operating requirements on instrumentation components, it may be necessary to upgrade those instrumentation components to withstand the more sever operating conditions and to measure greater variations of monitored variables that may be associated with an accident.
It is essential that instrumentation so upgraded does not compromise the accuracy and sensitivity required for nomal operation.
In some cases, this will neces-sitate use of overlapping ranges of instruments to monitor the required range of the variable to be monitored.
Standard ANS-4.5,* " Criteria for Accident Monitoring Functions in a Light-Water-Cooled Nuclear Power Generating Station," dated 1980, delineates criteria for detemining the variables to be monitored by the control room oper-ator, as required for safety, during the course of an accident and during the long-tem stable shutdown phase following an accident.
Standard ANS-4.5 was prepared by Working Group 4.5 of Subcommittee ANS-4 with two primary objectives:
(_l) to address that instrumentation that permits the operator to monitor ex-pected parameter changes in an accident period and (2) to address extended range instrumentation deemed appropriate for the possibility of encountering previously unforeseen events.
ANS-4.5 references a revision to IEEE S?.d 497 as the source of specific instrumentation design criteria.
Since the revision to IEEE Std 497 has not yet been completed, its applicability can not yet be detemined.
- Hence, specific. instrumentation design and qualification criteria have been included in this regulatory guide.
The standard defines three variable types (definitions modified herein) for the purpose of aiding the designer in his selection of accident-monitoring instru-mentation and applicable criteria.
(A fourth and fifth type (Type D and Type E) have been added by this regulatory guide.) The types are: Type A - those var-iables that provide infomation needed for preplanned operator actions, Type B -
those variables that provide inftrmation to indicate whether plant safety func-tions are being accomplished, Type C - those variables that provide infomation to indicate the potential for being breached or the actual breach of the barriers to fission product release, i.e., fuel caldding, primary coolant pressure boundary, and containment (modified to reflect NRC staff position; see Position C.2),
- Copies may ~be obtained from the American Nuclear Society, 555 North. Kensington Avenue, La Grange Park, Illinois 60525.
5
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Type D - those variables that provide information to ' indicate'the operation of individual safety systems, and Typa E - those variables to be monitored as re-quired for us in detennining the magnitude of the release of radioactive, mat-rials and for continuously assessing such releases.
A minimum set of Types B, C, D, and E variables to be measured is li ed j
in this regulatory guide. Type A variables have not been listed because they l
are plant specific and will depend on the oper4'ons that the designer ' chooses
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for planned manual action.' Types B, C, D, and E are variables for following the course of an accident and are to be used (a) to detennine if the plant is i
l responding to the safety measures in operation, (b) to infonn the operator of l
the necessity for unplanned actions to mitigate the consequences of an accident, and (c) to implement emergency procedures. 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 auto-matica11y initiated safety actions. A variable included as Type B, C, D, or E does not preclude that variable from being included as Type A also. Where such l
multiple use occurs, it is essential that instrumentation be capable of meeting l
the most stringent requirements.
Since the emergency response facilities (i.e.,
SPDS, TSC, EOF, NDL) are for the purpose of providing support and backsg_fjLr detecting abnormalities, mitigating their consequences, and protectino the safety of the public, the variables to be monitored by these emergency facili+.ies are taken from and/or included in the list of variables provided for Types B, C, D, and E.
This will avoid duplication of requirements and help assure that l
infonnation on power plant conditions is derived from the same data base, thus enhancing correlation of data concerning systems status and appropriate operator actions.
(See NUREG-0696)
The time phases (Phases I and II) delineated in ANS-4.5 are not used in this regulatory guide. These considerations are plant specific.
It is import-ant that the requirad instrumentation survive the accident environment and function as long as the infonnation it providas is needed by the plant operating organization.
Regulatory Position C.7 of this guide provides design and qualification cHteria for the instrumentation used to measure the various variables listed in Table 2 (for PWRs) and Table 3 (for BWRs). The criteria is grouped into three separate groups or categories which provide a graded approach to require.
i ments depending on the importance to safety of a variable being measured.
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i.
4 In general, the measurement of a single key variable may not be sufficient to indicate the accomplishment of a given safety function. Where multiple variables are needed to indicate the accomplishment of a given safety function, it is essential that they be considered key variables and measured with high-quality 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 applied for design, qualification, and quality r,ssurance of the instrumenta'; ion need not be the same as that applied for the instrunentation for key variables.
The design and qualification criteria category assigned to each variable in-dicates whether the variable is considered to be a key variable or for backup
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or diagnosis.
Variables are listed but no mention (beyond redundancy 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 measurement.
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a, 7
i C.
REGULATORY POSITION The criteria, and requirements, contained in Standard ANS-4.5," Criteria for Accident Monitoring Functions in a Light-Water-Cooled' Nuclear Power Generating Station," dated 1980, are considered by the NRC staff to be generally l
acceptable for providing instrumentation to monitor variables and systems for accident conditions and for monitoring the reactor containment, spaces containing j
co'aponents for recirculation of loss-of-coolant accident fluids, effluent discharge i
paths, and the plant environs for radioactivity that may be released during and following an accident from a nuclear power. plant subject to the following:
1.
Section 2.0 of ANS-4.5 defines the scope of the standard as contain-i ing criteria for determining the variables to be monitored by the control room j
operator of a light water reactor, as required for safety during the course of an accident.
Consideration should be given to the additional requirements (e.g.,
emergency planning) of variables to be monitored by the plant operating organiza-tion during an accident in order that it can perform its role in protecting the health and safety of the public.
1 2.
In Section 3.2.3 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 pars.neters which have the potential for causing a breach in the primary reactor containment have exceeded the design basis values.
In conjunction with the parameters that indicate the potential for causing a breach in the primary reactor 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 instruments, and associated parameters to be measured, in Standard ANS-4.5 (e.g., Sections 4.2,5.0,5.1.3,5.2,6.0,6.3) should include this expanded definition.
)
3.
Section 3.3 of ANS-4.5 defines design basis accident events.
In addi-tion to the design basis accident events delineated in the standard, those events t
defined as " anticipated operational occurrences'" in 10 CFR 50 (which are excluded in.the ANS-4.5 definition) should be also included.
-4.
Section 5.0, " Design Basis," of ANS-4.5 pertains to selection of accident monitoring variable and information display channels.
In conjunction I
'8
with Items A, B, and C given in Section 5.0, the plant designer should identify information display channels required by his design to enable the operator to:
D.
Ascertain the' operating status of each individual safety system to that extent necessary to determine if each system is operating or can be placed in operation to help mitigate the consequences of an accident.
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E.
Monitor the effluent discharge paths and environs within the site boundary to ascertain if there has been significant releases (planned or unp1*anned) of radioactive materials and for continuously assessing such releases.
F.
Obtain required information through a backtp or diagnosis channel i
where a single channel may be likely to give ambiguous indication.
i j
5.
Section 5.1, " Variable Selection," pertains to the process for selection of accident monitoring variables.
In conjuction with the selection of variables for Type A, Type B and Type C in Section 5.1, identification should be made of:
i For Type D 1) the plant safety systems which are designed to help mitigate or
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which could be placed in operation to help mitigate the consequences of an accident;
)
2) the variable or minimum list of variables that in"icate the operating status of each system identified in (1) above.
For Type E 1)
.the planned paths for effluent release; i
2) plant areas or inside buildings where access is required to service j
equipment necessary to mitigate the consequences of an accident; j
3) onsite locations where unplanned releases of radioactive mate-rials will be detected; j
4) the variables that should be monitored in each location identi-j fied in 1), 2) and 3) above and additional variables for defense in depth and for diagnosis.
6.
Section 5.2, "Perf% re Requirement," of ANS 4.5 pertains to the determination of performanc t ;q, wants for. accident monP.'oring information display channels.
In conjuration wii.n Section 5.2, determination of performance requirements for Type D and Type E accident monitoring information display channels should be the same as for. Type B.
i 9
e 7.
Section 6.1 of ANS-4.5 pertains to General Criteria for Types A, B, and t; accident monitoring variables.
Ir. lieu of Section 6.1, the following design and qualification criteria categories are established and should be used:
a.
Category 1 (1) The instrumentation should be environmentally qualified in accord-an;x with Regulatory Guide 1.89 (NUREG-0588).
Qualification applies to the complete instrumentation channel (end to end) from sensor to display or recording device, including isolation devices.
The seismic portion of environmental qualification should be in accordance with Regulatory Guide 1.100.
Instrumentation should continue to read within the required 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 be qualified using the guidance provided in paragraph 6.3.6 of ANS-4.5.
(2) No single failure within either the accident-monitoring instrumenta-tion, its auxiliary supporting features or its power sources concurrent with the failures that are a condition or result of a specific accident, should prevent the operator from obtaining the information necessary for him to perform his role in knowing the safety status of the plant and in bringing the plant to and maintaining it in a safe condition following that accident. Where failure of one accident monitoring channel results in the information ambiguity (that is, the redundant displays disagree) which could lead the operator to defeat or fail to accomplish a required safety function, additional information should be provided to allow the operator to deduce the actual conditions that are required for him to perform his role.
This may be accomplished by providing additicnal independent channels of infomation of the same variable (addition of an identical channel), or by providing sn independent channel which monitors a different variable which bears a known relationship to the multiple channels (addition of a diverse channel), or by providing tN capability, if sufficient time is available, for the operator to perturb the measured variable and deter-mine which channel hasafailed by observation of the response on each instrumenta-tion channel.
Redundant channels should be electrically independent, energized from station Class 1E power, and physically separated in accordance with Regula-tory Guide 1.75.
(NOTE: Within each redundant division of a safety system, redundant moni.toring channe,ls are not required.)
10
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k (3) The instrumentation should be energized from station Class 1E power.
(4) An instrumentation channel should be available prior to an accident except as provided in Paragraph 4.11, " Exemption", as defined in IEEE Std 279 or as specified in Technical Specifications.
(5) The recommendations of the following regulatory guides pertaining to quality assurance should be following:
" Quality Assurance Program Requirements (Design
& Construction)"
" Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment" Regulatory Guide 1.38
" Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants" Regulatory Guide 1.58
" Qualification of Nuclear Power Plant Insptection, Examination, and Testing Personnel" Regulatory Guide 1.64
" Quality Assurance Requirements for the Design of Nuclear Power Plants" Regulatory. Guide 1.74
" Quality Assurance Terms and Definitions" Regulatory Guide 1.88
" Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records" Regulatory Guide 1.123
" Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants" Regulatory Guide 1.144
" Auditing of Quality Assurance Programs for Nuclear Power Plants" Task RS 810-5
" Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants" (Guide number to be inserted.)
Reference to the above regulatory guides (except Regulatory Guides 1.30, and 1.38) are being made pending issuance of a regulatory guide endorsing NQA-1 (Task RS 002-5) which is in progress.
(6) Continuous indication (it may be by recording) display should be provided at all times. Where two or more instruments are needed to cover a required range, overlapping of instrument span should be provided.
11
1 (7) Recording of instrumentation readout information should be provided.
Where trend or transient information is essential for operator information or action, the recording should be analog stripchart.
Intermittent displays, such as data loggers and scanning recorders, may be used if no significant transient response can occur inside the recording interval.
(8) The instruments should be specifically identified on the control panels so that the operator can easily discern that they are intended foe use under accident conditions.
b.
Category 2 (1) The instrumentation should be environmentally qualified in accord-ance with Regulatory Guide 1.89 (NUREG-0588).
Qualification applies from the sensor through the isolator / input buffer if the channel signal is'to be processed for display on demand.
i (2) The instrumentation should be energized from a high reliability non-Class 1E power source, batte y backed where momentary interruption is not tolerable.
(3) The out-of-service interval should be based on normal Technical Specification requirements on out-of-service for the system it serves where applicable or where specified by other requirements.
(4) The recommendations of the following regulatory guides pertaining to quality assurance should be followed:
~
" Quality Assurance Program Requirements (Design
& Construction)"
" Quality Assurance Requirements for the Installation, Inspe.: tion, and Testing of Instrumentation and Electric Equipment" Regulatory Guide 1.38
" Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants" Regulatory Guide 1.58
" Qualification of Nuclear Power Plant Insptection, Examination, and Testing Personnel" Regulatory Guide 1.64
" Quality Assurance Requirements for the Design.
of Nuclear Power Plants" l
12 i
" Quality Assurance Terms and Definitions" Regulatory Guide 1.88
" Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records" Regulatory Guide 1.123
" Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants" Regulatory Guide 1.144
" Auditing of Quality Assurance Programs for Nuclear Power Plants" Task RS 810-5
" Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants" (Guide number to be inserted.)
Reference to the above regulatory guides (except Regulatory Guides 1.30, and 1.38) are being made pending issuance of a regulatory guide endorsing NQA-1 (Task RS 002-5) which is in progress.
Since some instrumentation is less important to safety than other instrumentation, it may not be necessary to apply the same quality assurance measures to all instrumentation.
The quality assurance requirements, which are implemented, should provide control over activities affecting quality to an extent consistent with the importance to safety of the instrumentation.
These requirements should be determined and documented by personnel knowledgeable in the end use of the instrumentation.
(5) The instrumentation signal may be displayed on an individual instrument or it may be processed for display on demand by a CRT or other appro-priate means.
(6) The method of display may be dial, digital, CRT or stripchart recorder indication.
Effluent release monitors require recording, including effluent radioactivity monitors, environs exposure rate monitors, and meteorology monitors.
Where trend or transient information is essential for operator information or action, the recording should be analog stripchart.
C.
Category 3 - high quality commercial grade instrumentation select to withstand service environment.
8.
In addition to the criteria of Position C.7, the following criteria i
should apply:
a.
Any equipment that is used for both accident monitoring and non-safety functions should be classified ss part of the accident-monitoring instru-mentation.
The transmission of signals from accident-monitoring equipment for l
13
i e
nonsafety system use should be through isolation devices that are classified as part of the accident-monitoring instrumentation and that meet the provisions of the document.
b.
Means should be provided for checking, with a high degree of confidence.
The operational availability of each accident-monitoring channel, including its input sensor, during reactor operation.
This may be accomplished in,,various ways, for example:
(1) By perturbing the monitored variable; (2) By introducing and varying, as appropriate, a substitute input to the sensor of the same nature as the measured variable; or (3) By cross-checking between channels that bear a known relation-ship to each other and that have readouts available.
c.
Servicing, testing, and calibration programs should be specified to maintain the capability of the accident-monitoring instrumentation.
For those instruments where the 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.
d.
Whenever means for bypassing channels are included in the design, the design should facilitate administrative control of the access to such bypass means.
e.
The design should facilitate administrative control of the access to all setpoint adju-tments, module calibration adjustments, and test points.
f.
The accident monitoring instrumentation design should minimize the development of conditions that would cause meters, annunciators, recorders, alarms, etc., to give anomalous indications potentially confusing to the operator.
g.
The instrumentation should be designed to facilitate the recognition, location, replacement, repair, or adjustment of malfunctioning components or
- modules, h.
To the extent practical, accident monitoring instrumentation inputs should be from sensors that directly measure the desired variables.
i.
To the extent practical, the same instruments should be used for i accident monitoring as are used for the normal operations of the plant to enable l
the operator to use, during accident situations, instruments with which he is l-most familiar.
However, where the required range of accident-monitoring instru-mentation results in a loss of instrumentation sensitivity in the normal operating range, separate instruments. should be used.
U
j.
Periodic testing should be in accordance with the applicable portions of Pagulatory Guide 1.118 pertaining to testing of instruments channels.
9.
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 ANS-4.5 pertain to variables and variable ranges for monitoring.
In conjunction with the above sections, Tables 2, and 3 of this regulatory guide (which include those variables mentioned in the above sections) should be used in-developing the minimum set of instruments and their respective ranges for accident monitoring instrumentation for each nuclear power plant.
D.
IMPLEMENTATION All plants going into operation afte; June 1982 should meet the provisions of this guide.
Plants currently operating or scheduled to be licensed to operate before June 1, 1982 should meet the requirements of NUREG-0578 and NRR letters dated September 13, and October 30, 1979.
The provisions of this guide as specified in Tables 2, and 3 for operating plants are compatible with these documents which are to be completed by January 1,1981.
The balance of provisions of the guide are to be completed by June 1983.
The difficulties of procuring and installing additions or modifications to in place instrumentation have been considered in establishing these schedules.
Exceptions to requirements and schedules will be considered for extraordinary circumstances.
1 1
y p
15
TABLE 2 (Revissd PWR VARIABt.ES Design & Qualif. cation Variable Rance Criteria Categt,'y Purpose D
D TYPE A - Variables for Pre-Plant specific 1
Lo plann:d Manual Actions o' D l9 A
TYPE B - Variables Indicating b!O.b
'A3 Critical Safety Functions R activity Control Control R Position Full in or not 3 (for 2 hr full in minimum)
Neutron Flux 1 c/s *.o 3 power 1
SPDS Soluble Boron 0 to 6000 ppm 3
Content (continuous in-dication)
Boric Acid Charging 0 to 11C% design 3-Flow flow' j
Core Cooling Steam Genarator Level From tune sheet to 1 (_2 for B&W separators a@
RCS Hot Leg Teger-50*F to 750*F 1
SPDS 4ture Coolant Level in 1
Reactor aCs Cold Les temper-50*F to 750*F 1
SPDS ature 0 to 120% ]
1 Reactor Coolant Loop
-12% to i s )) design Flow flow &
l Cegree of Subcooling 200*F subcooling to 2
l 35'F supernest Condensate Storage Plant specific 1 O if not primary T*ak L*I source of AFW.'Then l
whatever is pri=ary
'Cmen flow - the saataus flew anticipated in nor.a1 operation.
source of AFW should l
be listed and should 16 be Category 1)
TABLE 2 (con ' ued) (Revised Category Purposa Variable Ranc7 TYPE B - (continued)
M.*2intaining Reactor Coolant System Integrity RCS Pressure 15 psia to 3000 13 SPDS psig (for CE plancs, 4000 psig)
Pressurizer Level Bottom to top 1
SPDS Privaary System Safety closed-not cir. sed 1
R311ef Valve Positions (including PORY and code valves) or Flow Through or Pressure in Relief Valve I.ines Containment Sump Marrow range (supp).
1 SPDS Vater Level Wide range (bottom of containment to 600,000 ga11on
..,,,_,,,, level equivalent)
Maintaining Containment Integrity Cantainment Pressure 10 psia pressure 1
3pg3 to 3 times design pressurea for concrete; 4 times design pressure for steel Containment Isola-Closed-not closed tion Velve Position 1
m p',J (excluding check valves)
D D
voI mmm~
r Containment Hydrogen 0 to los 1
(1 D I u
Concentration (capaele of y
- n t.
.L J
(
operating from v
JL
(
l 10 psia to maximum design l
pressure )
a O to 30% for ice condenser type containment l
3 Design pressure - that value corresponding to ASME code values that are obtained at or below cade-allowanle.
meterial design stress values.
17
TABLE 2 (continu:d)
(Revis;d Purpose Variable
- Ranc, Category
. TYPE C - Variables for Potential
! for or Breach of Barriers Fuol Cladding Core Exit Temperature 150*F to 2300*F 1"
SPDS Radioactivity Concentration Normal to 10 Ci/gm 3
or Radiation Level in Cir-i culating Primary Coolat.t Reactor Coolant Pressure Boundary 5 19 Containment High-1 to lof R/hr 1
SPDS Range Area Radiation Effluent Radioactivity -
Nsblo Gas Effluent.from
_104 to los uct/cc 2
SPDS Cond2nser Air Removal Syct n Exhaust Staca Turbine Driven Aux-10-6 to 103 pCi/cc 2 16 18 iliary Feedwater Pump Vent i
Containment Effluent Radioactiv-10-8 5
6 16 le to 10 pci/cc -'
2 icy - Noble Cases Environs Radioactiv-10-6 to 10 R/hr 27 IS icy - Exposure Rate p b
~
18
_ _ _ _ _ _ _ _ _ TABLE 2 (cancisuad) -(R vised)
Variablo Ranc7 Category Purposa DIv D
l TYPE D - Variable Inc'icating Oper-1
- ation of Individual Safety Systems D
l '9~ I
~
Sicondary Systems vI S 1
_a (heloweM steam cenerator Free ata=osoneric 2
SPOS Pressure pressure to 20%
aeovejsarety value setting i
e Emer auxiliary ^dency ee.acer 0 to n0% desi a 2 (1 for B&W SPOS 9
Flow flow' plants)
Nain Feedwater Flow 0t UO5 desig" 3
SPDS flows Safety /Re11er Valve Closed-not closed Positions or Main 2
Steam Flow Radioactivity in Efflu-10 1 to 103 Ci/cc 217 3
eat from steam cener-ator safety Relief "stves or Atmospheric Quas valves Auxiliary Systems Sump Water. Temperature 50*F to 250*F 2
Containment Spray 0 to 110% des'ign 2
Flow flow 2 Containment Atmos-40*F g 4aa.F 3
phere Temperature l
Heat Removal ey the Plant specific 2
l Containment Pun Heat "nlar2, Removal System t.
19
TABLE Z (c==:i=uad) (R vised)
Va ria ble Ranc7 Category Purpose TYPE D'- (continued)
Auxiliary Systems (continued)
D P*%
Flow in HPI Systes 0 to 110% design 2
6w1 flowl g]
D
. 9 T.
,j_[A Flow in LP! System o to 110% design 2
g flowl
. Refueling EmergencygCoolant Top to bottos 2
Water Storage Tant Level 10% to 90% volume 2
Ae m iaur Tan, Level or Pressure or 0 to 750 psi
)
Accumulator Isolation Closed-not closed 2
Valve Positions RHR Systes Flow 0 to 110% design g
flow 1 RHR Heat Exchanger 32*F to 350*F g
Out Temperature Component Cooling 32*F to 200*F 2,
Water Tercerature -
Component Cooling 0 to 110% design 2
Water Flow flow 1 Flow in Ultimate O to 110% design 2
Heat Sink Loop flow 1 Taraperature in Ulti-30*F to 150*F 2
mate Heat Stot Loop Ultimace Heat Sink Water Plant specific 2
Level l
2 Letdown Flow - In 0 to 110% design flow l
2 Letdown Flow - Out 0 to 110% design flow Sump Level in Spaces To corresponding 3
af Equipeent Requiree level of safety feet Safety equipment failure Steam Flow to Aux-O to 110% design 2
iliary Feedwater flowl Pumps 20 1
r/D55 75T65NEELE91N Variable Ranc7 Catacorv Purposo
- TYPE D - (. continued) oa' D
o1 e40 waste system 3b
'1'D'32L1b Hign-l.evel Radioactive Top to sottos 3
5 u r" '-* " '
Radfoa[tiveGasHold-0 to 1505 of 3
up Tank Pressure design pressure
- VENTTt.ATTON SYSTEMS Emergency Ventilation Open-closed 2
Damper Position status Teeperature of Space 30*F to IS0*F 3
in Vicinity of Equip-ment Recutred for Safety D M g mpetYes 8
Status of Class 1E voltages 2
Power ~ -
and currents l
Systems Sources pressures 9
Status of Non-f, lass Voltages 3
1E Power J-emi h currents ed Systems 'jources Pressures TYPE E - Variables Which Indicate Magnitude and Direction of Disper-sion of Released Radioactive Materials
~
A 245 W r1E ACCES$ IS eEcoloED T0 SE4vICE ~
SAs:TY-4 ELATE 0 Ecu(P**ENT R/h 13*
Radiation Exposure lo-a to 104 R/hr Rites
.F 21
IT8LE Z (c=ntisued) (Revis:d)
Variable Ranc7 Category Purposo TYPE E - (continued)
ATR80RNE RA0!CACTTvf
-er:arus act:4s:n O
D D
Fe0N NE PtJNT Gaseous Effluent Vol-O to 110% design Flow 2
a umetric Floit Rate flowl f
Effluent Radioactiv-V Q
ity - Noble Gases 1s 18 10*8 o 104, Cf/cc 2
t
. Secondary p
Containment (Reactor shield building annulus)
. Auxiliary Butiding 10- t's los pCf/cc 2
including bui1 dings c6ntaining primary system gases, e.g.,
waste gas decay tant 1s 18
.ather Release
,10 6 to 108 pCf/cc 2
Points (including fuel handling area L
if separate from auxiliary building) 1 Effluent Radioactiv-lo-a to 108 pCi/cc 2
fty "'; '- ;
RIdichalogens,and FCrticulates..
Environs Radioactiv-104 to 10-8 pCf/cc 2
ity - Radionalogens far bota radio-and Particulates halogens and particu,1ates j
l l
Plan
- and Environs Hioh Rame 12 Radfeactivity & Radiation 0.1 to la+ R/hr 3
(portaale instruments) pnotons 0.1 to 10* rads /hr botas and low-energy photons multi-channel 3
gamma-ray spectr. meter l
22 l
TABLES 2 (continu;d) (R;vistd)
Variable Range Category Purpose TYPE E - (continued)
POSTACCIDENT SAMPLING
. CAPABILITY (Analysis Cap"biliry Onsite) 13 14 21 Primary Coolant & Sump Grab Sample 3
Gross Activity 10 pC1/ml to 10 Ci/ml Gamma Spectrum (Isotopic Analysis)
B:ron Content 0 to 6000 ppm Chloride Content 0 to 20 ppa Disolved Oxygen 0 to 20 ppa Dicolved Hydrogen 0 to 1000 cc/kg pH 1 to 13 13 21 Ctuteinment Air Grab Sample 3
Hydrogen Content 0 to 10%
0 to 30% for ice condensors oxygen Content 0 to 30%
Gamma Spectrum (Noble gas analysis) 1 METEOROLOGY 15 I
Wind Direction 0 to 360* (tS* accuracy with 2
a defliction of 15*. Starting speed 0.45 mps (1.0 mph) Dam-ping ratio between 0.4 and 0.6, distance constant $2 meters)
Wind Speed 0 to 30 =ps (67 =ph)(10.22 mps 2~
(0.5 mph) accuracy for wind speed less than 11 mps (25 mph), with a starting threshold of less than 0.45 mps (10 mph))
Ectimation of Atmospheric Based on vertical temperature 2
Sttbility difference from primary system
-5'C to 10*C (-9'F to 18'F) and j
- 0.15'c accuracy per 50 meter in-tervals (i0.3*F accuracy per 164-foot intervals) or analogous range for backup system.
- Time. for taking and analysing samples should be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less from the time d: cision is made to sample, except chloride which should be wit *.in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
l l
l l
23 i
TABLE 2 (continued)
NOTES continued -
3The =mvinum value may be revised upward to satisfy ATWS requirements.
~
4A cinimum of 4 measurements.per quadrant is required for operation.
Sufficient number chtuld be installed to account for attrition.
5 Minimum of two monitors at wid'ely separated locations.
6 rovisions should be made to menitor all identified pathways for release of gaseous P
radioac,tive materials to the environs in confor. nance with General Design Criterion 64.
Monitoring of individual effluent streams only is required where such streams are re-leased directly to the environment.
If two or more streams are combined prior to re-Icase from a common discharge point, monitoring of the combined stream ?.s considered to meet the intent of this guide provided such monitoring has a range adequate to meas-4
~
ure worst-case releases.
Ear estimating release rates of radioactive materials released during an accident 7
from unidentified release paths (not covered by effluent monitors) - continuous j
rcrdout capability.
(Approximately 16 to 20 locations - site dependent.)
i 8 entus indication of all Class 1E A-C buses, D-C buses, inverter output buses and f
S pneumatic supplies.
95tatus indication of all non-Class lE inverter output buses, D-C buses and pneumatic supplies.
IC o provide information regarding release of radioactive halogens and particulates.
TContinuous collection of representative samples followed by onsite laboratory measure-mants of samples for radiohalogens and particulates. The design envelope for shielding, handling, and analytical purposes should assume 30 minutes of integrated sampling time at sampler design flow, an average concentration of 102 pCi/cc of radioiodine in gaseous 2
or vapor form, an average concentration of 10 pCi/cc of particulate radioiodines and particulates other than radiciodines, and an average gamma photon energy of 0.5 Mev per disintegration.
f 11 or estimating release races of radioactive materials released during an accident F
from unidentified release paths (not covered by effluent monitors).
Continous collection of representative samples followed by laboratory measurements of the camples (Approximately 16 to 20 locations - site dependent.)
12 a monitor radiation and airborne var".ioactivity concentrations in many areas T
throughout the facility and the site environs where it is impractical to install ctationary monitors capatie of covering both normal and accident levels.
13 o provide means for safe and convenient sampling, These provisions should include:
T i
- 1. Shielding to maintain radiation doses ALARt.,
- 2. Sample containers with container-sampling port connector compatability, 3
l-
- 3. Capability of sampling under primary system pressure and neg..c. e pressure, l
- 4. Handling ud transport capability, and
(
- 5. Pre-arrangement for analysis and interpretation.
24 1
TABLE 2 (continued)
NOTES cont.nued -
14An installed capability should be provided for obtaining containment sump, ECCS pump roca sumps, and other similar auxiliary building sump liquid samples.
15Mateorological measurements should conform to the provisions of the forthcoming revision
to Regulatory Guide 1.23, "Onsita Meteorological Programs".
16 Monitors should be capable of detecting and measuring radioactive gaseous effluent con-centrati,ons with compositions ranging from fresh equilibrium noble gas fission product mixtures to 10-day old mixtures, vich overall system accuracies of
- 1/2 deccde. Cal-ibration should be performed using radiation sources representative of both low and high cuargy portions of the emission spectrum. Fow low-energy gamma photon calibration, ccurce emission energies should fall within the range of approximately 60 kev to 150 kev (examples - Am-241, cd-109, Tm-171, and co-57). For high-energy gamma photon calibration, ocurce emission energies should fall within the range of approximately 500 kev to 1.5 MeV (examples - Cs-137, Mn-54, and Co-60). Effluent concentrations may be expressed in terms of Xe-133 equivalents or in terms of th.: equivalent of any noble gas nuclide(s).
17 Effluent for PWR steam safety valve discharges and atmospheric steam dump valve dis-charges should be capable of approximately linear response to gamma radiation photons with energies from approximately 0.5 MeV to 3 MeV. Overall system accuracy should be within 1/2 order of magnitude. Calibration sources should fall within the range of cpproximately 0.5 MeV to 1.5 MeV (examples: Cs-137, Mn-54, Na-22, and co-60). Effluent csucentrations should be expressed in terms of any gamma-emitting noble gas nuclide within the specified energy range. Calculational methods should be provided for est-imating concurrent releases of low-energy noble gases which cannot be detected or iaeasured by the methods or techniques employed for monitoring.
18It is not expected that a single monitoring device will have sufficient range to en-compass the entire range provided in this guide and that multiple components or systems will be needed. Existing equipment may be utilized to monitor any portion of the orated range within the equipment design rating. Additional extended range instrument-ction should overlap the range of existing instrumentation by at least a factor of 2.
19 etectors should respond to gamma radiation photons within any energy range from D
60 kev to 3 MeV with an accuracy of $20% at any specific photon energy from 0.1 MeV to 3 MeV. Overall system accuracy should be within t1/2 decade over the entire range.
20 sasurement should be made of the gross gamma radiation emanating from circulating Mprimary coolant, with instrument calibration permitting conversion of readout to r dioactivity concentrations in terms of either curies / gram or curies / unit-volume.
System accuracy should be 1/2 order of magnitude. The point of measurement should
- b3 external to a circulating primary coolant line or loop, such as a hot leg, and thould not be a line or loop subject to isolation, e.g., PWR letdown line or BWR main steam line. While such an instrument may not be currently available off-the-chelf, the staff considers that thernecessary components are available commercially cnd have been employed and demonstdted under adverse environmental conditions in high-level hot cell operations for many years.
Sampling or monitoring of radioactive liquids and gases should be performed in a manner j
21 which assures procurement of representative samples. For gases, the criteria of ANSI l
N13.1 should be applied.
For liquids, provisions should be made for sampling from well-mixed turbulent zones and sampling lines should be designed to ninimize plateout or deposition.
25 e
TABLE 3 EWR VARIA81.ES Design & Qualificatidn Variable Rance Criteria Catecorv Purpose
~
TYPE A - Variables for Pre-Plant specific 1
, plann;d Manual Actions D
]D TYPE B - Variables Indicating v c Iu Critical Safety Functions
'9~IO 11D JD
. i ddo n.
~
R! activity Control Cantrol Red Position Full fg or not 3 (.for 1 hr full in minimum) l n tron nux 1 c/s to a pe er 1
SPDS i
Core Cooling e.eiant i e in the settas of co m 1
SPDS sert plata Reacter to above too of discharge plenus n in steamitne new 0 to 1205 d**isa 1
fws Maintaining Reactor Cooling System Integrity Res Pmesure 1
SPDS 15 0 pi Main Steam 1tne Isela,
O to 15" of water 1
tien valves' Leenage O to 5 paid 1
Control System Pressure 1
Primary System Safety Closed net closed Reifer Valve Post-or tiens, including 0 to 50 psig ABS er Flow Threven or Pressure in Valve
- Lines,
.'Ce si gn flow - the taalma fM *ati"198t*d I"
""#"*I '**I*"*
26
TABLE 3 - (c::ntinued)
Catego 7 Purpose
- __ Var. iable Rance TYPE B - (continued)
Maintaining Containment Integrity Primary contain= eat la peta pressor.
1 SPDS Pressure (Mrywell) to 3 tfees desigst pressures for can.
crete; 4 tfees design pressure o
for steel D
D 6
cantalveewe and Gryw.11 o ta 30%
1 i
Hytrogna Canmtration (capab " '.7 of D
e,.r.t.ng fr
\\
22 psfa ta samf '
y
.(
O res y
1 containment.ind crywell o ta 20 Caygen Cancerstration (capanflity of (for those piants operattog free with fnerted 22 psfa to containments),
design pressure )
a Primary containment closed-nes closed 1
Isetation Position (valve-excluding check vs.1ves)
Suppression Chamber 30'F to 230*F 1
Air Temperature Drywell Temperature 40*F to 440*F 1
TYPE C - Variables for Potential for or areach of Barriers.
l l
Fuel Cladding care saft re eeratur.
uav to noov 1
.g.
11 R-dicactivity Concentration Normal to 10 Ci/gm 3
SPDS cr Radiation Level in 'Cir-culcting Primary Coo. ant
)
l30 sten pressurt. - that value corresponding to ASME code values tnat are catained at or below code-alicwable materfC1 design stress values.
27
TABLE 3-(c:ntinued) i Variable Rance Category Purpose TYPE C - (continued) i Reactor Coolant Pressure Boundary Cont =4="-"t Eish-I to la' R/hr 33, i
Ranga Area Radiation 1
(for Mark III containments, two redundant monitors are required for pri-mary containment &
l reactor building)
OW II Drain Sumos Bottae to tas 1
SPDS tevel Geenstries.
and unfcentseten Leakage)
Containment Standby Gas Treatment 106 I8'17 to 105 pCi/cc 2
SPDS Syctem Vent Effluent Radioactiv-10-6 s
go lo pCi/cc 2 6 18 17 SPDS icy - Noble Gases Environs Radioactiv-10" to 10 R/hr 2 7 la icy - Exposure Rata TYPE D - Variables Indicating Oper-ation of Individual Safety Systems Power Conversion Systems 0,,to 11 5 design 3
Main Feoometer Flow eyy7g candensate Sterage Bottaa to too 3
Tant Level o'
- g
.S.k
_a 28
TABLE 3-(c=ntinued)
Variable Rance Category Purpose TYPE D - (. continued) ong d ov1 Containment Systems Containment sersy 0 to uns desta 2
D
~ 9
~ l-m ci i
new fie=5
. dj A.
_a Drywell Pressure 12 psia to 3 psig 2
SPDS 0 to 110% design 2
pressure wresefoe Chamber rep of vene u 2
SPDS h ter I.svol tap of wofr well s.oecesafea Chamber 30*F to 230*F 2
SPDS wter Temperature A_uxiliary Systems Control Rod Drive O to 110% design 2
System Return' Floe flowl 5 tees Flow to RCIC 0 to uns desip 2
flows BPCI Flow. -
0 to 110% design 2
..fl.8' 1
l ACIC Mew 0ta uc5deeip flows 2
hreSprayFlow 0 to 110% design 2
flowl RMft Systee Flow (LPCI) o ta nos dest e news 2
mat Heat E== neap r 32 F t 3so*F Outlet ressersture 2
$PDS (LPCI)
Service CoeIinf 3 *F to 200*fr
%ter reaperature g
Service CoelinW wt.c new G ta M05 desip no,a 2
Flow fa Ultfesta 0 to uc5 desty Meet $1nk Loop rio s 2
s Teasereture in Ultf*
34*F to 150*F esta Noet,5fnt Lee,
2 Ultfasta Heat $fnt Plant spectfic 2
Level 29 l
TABLE 3 - (c::ntinued)
W ri M e hnca Category Purpose TYPE D - (continued)
]D D
Auxiliary Systems (continued) g o Ju e
+DJ sLes store, Tana
- 3 sette. to a, 1
jl_ b k
_a s
L ei in so.cee er fonfament naquired To co m e,enefn,
fee safety levet of safety 3
equfpment fatture SLCS Flow 0 to 110% design 3
flowl RtcwASTt SYSTmf leign Radioactivity Top to bettes Lfquid Tank Level 3
t i
l l
VGTTtATION SYSTOS Emergency Ventflatfen Damper Posftien Coen-1:lesed status
- 2 Temperature of soace 34*F to 180*F fa Vicinity of Equip =
3
- st negutred for saf;ty 90 win surputs statue of Class 1E Pe.ec e u e ane Voltages currens 8
systems sources 2
, ressures p
status of lean-Class Voltages 2'/r,e"t e soo,...
ams..
3 6
6 30
TABLE 3 - (continued)
Variable Rance Category Purposa TYPE E - Varialbes~Whidh Indicate Magnitude and Direction of Disper-sion of Released Radioactive Materials D
=
RAOAT'ON EXPOSURf 14TM5 M5sc.J ul t.JINGS 0 _9_T i
CR Aafah wi l
<255
'lAF 8Tv-4ELA TEll E0ulP=
g
! s sFeu 'RfD Ti.. sfavic!
M R/hr 18 Radiatten Espesure 10-hts 10* R/hr 2
Ratas Af#9CRNE 8A0f0ACTTyt warERI ALS 4Et.iA5ED Cacn TuE PLANT Effluent Radioactiv-Ity - Meole Gases Reactor Bldg or 10-6 to 10 taci/cc 2
SPDS 4
16 17
. Secondary Contain-ment 16 17
.0tner Release 10-6 se los pct /cc 2
Points (including fust he d1ing buildtag, aunt 11 -
any building, and turnine bul141ng)
Gaseous Effluent Vol-O to 110% design 2
umetric Flow Rate flowl Effluent Radteactiv-10-s to los pC1/cc fl0 fty ^', '
Radionalogens and Particulates c'
11 Emirens Radteactiv-IIN'to10-s u f/cc 2
fty - Radionalogene for oeta radionale-and Particulates
, gens and partic e fatas PkantandEnvirens Mon Rance 12 Radioactivity & Radiation 0.1 to 10* R/hr 3
(portaele instruments) phetane 0.1 ta 10*
reds /hr betas and low-energy phetens multi-channel 3
gassa-ray spectreesitar 31
TABLES 3 (continued)
. Variables Range Category Purpose TYPE E - (continued)
POSTACCIDENT SAMPLING CAPABILITT (Analysis Cap bility Onsite) 13 14 20 Primary Coolant & Sump Grab Sample 3
Grecs Activity 10 pCi/ml to 10 Ci/ml Camma Spectrum (Isotopic Analysis)
Borta Content 0 to 1000 ppm j
Chlerida Content 0 to 20 ppa Dico1ved Oxygen 0 to 20 ppm Disolved Hydrogen 0 to 1000 cc/kg pH 1 to 12 13 20 Cont.cinment Air Grab Sample 3
Hydrogen Content 0 to 30%
Oxygen Content 0 to 30%
~
Cacma Spectrum (Noble gas analysis)
METEOPOLOGY15 Wind Direction 0 to 360* (iS* accuracy with 2
a defliction of 15*. Starting
{
speed 0.45 mps (1.0 mph) Dam-ping ratio between 0.4 and 0.6, distance constant s2 meters)
Wind Speed 0 to 30 mps (67 mph)(to.22 mps 2
(0.5 mph) accuracy for wind speed less than 11 mps (25 mph), with a starting threshold of less than 0.45.mps (,1.0 mph))
Estimation of Atmospheric Based on vertical temperature 2
Stability difference from primary system
-5*C to 10*C (-9'F to 18'F) and
- 0.15*C accuracy per 50 meter in-tervals (f0.3*F accurac.y per 164-foot intervals) or analogous range for backup system.
~
l
- Time fcr taking and analysing samples should be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less from the time decicion is made to sample, except chloride which should be within'24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
32
TABLE 3 (continu:d)
=.
NdTEScontinurd-3Th3 maximum value may be revised upward to satisfy ATWS requirements.
"Approximately 50 thermocouples should be available, the exact number needed will depend en thermocouple location and other characteristics.
In the absence of core spray the charnocouples should detec.t 5 to 10% core area cross sectional blockage, with high con-fid2nce.
Sufficient numbers should be installed to account for attrition.
3 Minimum of two monitors at wid'ely separated locations.
6 rovisions should be made to monitor all identifSd pathways for release of gaseous Prcdioactive materials to the environs in conformance with General Design Criterion 64.
Monitor'Ing of individual effluent streams only is required where such streams are re-Icased directly to the environment.
If two or more streams are conbined prior to re-lance from a common discharge point, monitoring of the combined stream is considered to meet the intent,of this guide provided such monitoring has a range adequate to meas-
~
urs worst-case releases.
7For estimating release rates of radioactive materials released during an accident from unidentified reluase paths (not covered by effluent monitors) - continuous rstdout capability.
(Approximately 16 to 20 locations - site dependent.)
8 Status indication of all Class 1E A-C buses, D-C buses, inverter output buses and s
pasumatic supplies.
SStatus indication of all non-Class 1E inverter output buses, D-C buses and pneu=atic cupplies.
1C o provide information regarding release of radioactive halogens and particulates.
T Continuous canection of representative samples followed by onsite laboratory measure-nents of samples for radiohalogens and particulates. The design envelope for shielding, l
handling, and analytical purposes should assume 30 minutes of integrated sampling time cc campler design flow, an average concentration of 102 pCi/cc of radioiodine in gaseous er vapor form, an average concentration of 102 pCi/cc of particulate radiciodines and Particulates other than radioiodines, and an average gamma photon energy of 0.5 Mev per i
dinintegration.
l 11For esti nating release rates of radioactive materinis released during an accident from unidentified release pt ths (not covered by affluent monitors). Continous collection of representative samples followed by laboratory r.easurements of the j
camples (Approximately 16 to 20 locations - site dependent.)
l 12.s monitor radiation and airborne radioactivity concentrations in many areas T
throughout the facility and the site environs where it is impractical to install stationary monitors capable of covering both normal and accident levels.
~
13 o provide means for safe and convenient sampling.
These provisions should include:
T
- 1. Shielding to mainiain radiation doses ALARA, i
- 2. Sample containers with container-sampling port connector compatability.
l
- 3. Capability of sampling under primary system pressure and negative pressure, l
- 4. Handling and transport capability, and
- 5. Pre-arrangement for analysis and interpretation.
i l
]
l 9
33
TA8t.E 3 (continusd)
NOTES continued -
1 l'*An installed capability should be provided for obtaining containment sump, ECCS, pump rcom sumps, and other sim1. tar anv414=ry building sump liquid samples.
15 ateorological measurements should conform to the provisions of the forchecming revision M
en Regulatory Guide 1.23, "Onsite Meteorological Programs".
16 Monitors should be capable of defcecting and measuring radioactive gaseous affluent con-ccntrations with compositions ranging from fresh equilibrium noble gas fission product mixtures to 10-day old mixtures, with overall system accuracies of 1/2 decade. Cal-ibration should be performed using radiation sources representative of both low and high energy portions of the emission spectrum.
Fow low-energy gamma photon calibration, sturce emission energies should fall vidin the range of approximately 60 kev to 150 kev (czamples -- An-241, Cd-109, Tm-171, o.t C.o-57). For high-energy ga=ma photon calibration, scurce emission ene.rgies should fs*; aithin the range of approximately 500 kev to 1.5 MeV (cxamples - Cs-137, Mc.-54, and Ce 60). Effluent concentrations may be expressed in terms of Xe-133 equivalents or in terr.; of the equivalent of any noble gas nuclide(s).
17It is not expected that a single monitoring device will have sufficient range to en-compass the entire range provided in this guide and that multiple components or systems will be needed. Existing equip =ent may be utilized to monitor any portion of the ctated range within the equipment design rating. Additional extended range instrument-ccion should overlap the range of existing instrumentation by at least a factor of 2, leDatectors should respond to gamma radiation photons within any energy range frca 60 kev to 3 MeV with an accuracy of 20% at any specific photon energy frem 0.1 MeV to 3 MeV. Overall system accuracy should be within 1/2 decade over the entire range.
19 Measurement should be made of the gross gn=ma radiation e=anating from circulating pr* mary coolant, with instrument calibration persitting conversion of readout to adicactivity concentrations in term:s of either curies / gram or curies /unic-volume.
System accuracy should be 1/2 order of eagnicude. The point of =easure=ent should be external to a circulating primary co< lant line or loop, such as a hot leg, 'and chould not be a line or loop subject to isolatfan, e.g., PWit letdown line or BWR.
cain steam line. While such an instrument may not be currently available off-the-chelf, the staff considers that the necessary co=ponents are available commercially cud have been employed and demonstrated under adverse environmental conditions in high-level hot call operations for many years.
20 Sampling or monitoring of radioactive liquids and gases should be performed in a manner which assures procurement of representative sa=ples. For gases, the criteria of ANSI N13.1 should be applied. For liquids, provisions should be made for rampling from well-mixed turbulent zones and sampling lines should be designed to minimi:.; plateout or depositiA.
l l
34
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