ML19338F815
| ML19338F815 | |
| Person / Time | |
|---|---|
| Issue date: | 10/08/1980 |
| From: | Hintze A NRC OFFICE OF STANDARDS DEVELOPMENT |
| To: | |
| Shared Package | |
| ML19338F812 | List: |
| References | |
| REGGD-01.097, REGGD-1.097, NUDOCS 8010270328 | |
| Download: ML19338F815 (57) | |
Text
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1 Oraft 3 2
October 8, 1980 3
Division 1 4
Task RS 917-4 5
Contact:
A. S. Hintze, (301) 443-5913 6
[PR6P6SEB] REVISION 2 TO REGULATORY GUIDE 1.97 7
INSTRUMENTATICN FOR LIGHT-WATER-CColiD NUCLEAR PCWER PLANTS 8
TO ASSESS PLANT AND ENVIRONS CONDITICNS DURING AND FOLLOWING AN ACCIDENT 9
A.
INTRCOUCTICN 10 Criterion 13, " Instrumentation and Control," of Appendix A, " General Design 11 Criteria for Nuclear Power Plants," to 10 CFR Part 50, " Domestic Licensing of 12 Production and Utilization Facilities," includes a requirement that instrumen-13 tation be provided to monitor variables and systems over their anticipated ranges 14 1." accident conditions as appropriate to ensure adequate safety.
15 -
Criterion 19, " Control Roca," of Appendix A to 10 CFR Part 50 includes a j
16 requirement that a control room be provided free whicn actions can be taken to 17 maintain the nuclear power unit in a safe condition under accident conditions, 18 including loss-of-coolant accidents, and that equipment, including the necassary 19 instrumentation, at appropriate locations outside the control roce be provided 20 with a design capability for prompt hot shutdown of the reactor.
21 Criterion 64, " Monitoring Radioactivity Releases," of Appendix A to 10 CFR j
22 Part 50 includes a requirement that means be provided for monitoring the reactor 23 containment atmosphere, spaces containing components for recirculation of loss-24 of-coolant accident fluid, effluent discharge paths, and the plant environs 25 for radioactivity that any be released free postulated accidents.
26 Titis guide describes a method acceptable to the NRC staff for complying 27 with the Commission's regulations to orovide instrumentation to monitor plant 28 variables and systems during and following an accident in a light-water-cooled
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29 nuclear power plant.
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DISCUSSION
'2 Indications of plant variables are required by the control room operating
- 3 personnel during accident situations to (1) provide information required to 4
persit the operator to take preplanned aanual actions to accomplish safe plant 5
shutdown; (2) detarmine whether the reactor trip, engineered-safety-feature 6
systems, and manually initiatad safety systems and other systems important to 7
safety are performing their intanded functions (i.e., reactivity control, core 8
cooling, maintaining reactor coolant system integrity, and maintaining contain-9 ment integrity); and (3) provide inforsation to the operator that will enable 10 him to detarsine the potential for causing a gross breach of the barriers to 11 radioactivity release (i.e., fuel cladding, reactor coolant pressure boundary,
'12 and containment) and if a gross breach of a barrier has occurred. In addition 13 to the above, indications of plant variables which provide information on opera-
- 14 tion of plant safety systans and other systems important to safety are required 15 by the control room operating personnel during an accident to (1) furnish data 8
16 regarding the operation of plant systans in order that the operator can make 17 appropriata cecisions as to their use; and (2) provide information regarding the 18 release of radioactive matarials to allow for early indication of the need to 19 initiata action necessary to protect the public and for an esticate of the 20 magnitude of any impending threat.
21 At the start of an accident, it may be difficult for the operator to deter-22 aine immediately what accident has occurred or is occurring and, therefore, to 23 detarsine the appropriata response. For this reason, reactor trip and certain 24 other safety actions (e.g., emergency core cooling actuation, containment isola-25 tion, or depressuri:ation) have been designed to be performed autcmatically 26 during the initial stages of an accident. Instrumentation is also provided to 27 indicata inforsation about plant variables required to enable the operation of 28 aanually initiated safety systans and other appropriate operator actions involving 29 systans important to safety.
30
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8 Independent of the above tasks, it is important that the operator be informed 9
if the barriers to radioactive materials release are being challenged. Therefore, 10 it is essential that instrument ranges be selected such that the instrument will 11 always be on scale. Narrow-range instruments may not have the necessary range to 12 track the course of the accident, consequently, multiple instruments with over-13 lapping ranges may be necessary.
(In the past, some instrument ranges have been 14 selected based on the set point value for automatic protection or alarms.)
It is 15 essential that degraded conditions and their magnitude be identified so that the 16 operator can take actions that are available to mitigate the consequences.
It is 17 not intended that the operator be encouraged to prematurely circumvent systems 18 important to safety but that he be adequately informed in order that unplanned 19 actions can be taken when necessary.
20 Examples of serious events that could threaten safety if conditions degrade 21 are loss-of-coolant accidents (LOCAs), overpressure transients, anticicated 22 coerational occurrences which beccme accidents such as anticipated transients 23 without scram (ATWS), reactivity excursions which resuit in releases of radio-24 active materials.
Such events require that the operator understand, within a 25 short time period, the ability of the barriers to limit radioactivity release, 26 1.e., the potential for breach of a barrier, or an actual breach of a barrier by 27 an accident in progress.
28 It is essential that the required instrumentation be capable of surviving 29 the accident environment in which it is located for the length of time its func-30 tion is required.
It could therefore either be designed to withstand the accident 31 environment or be protected by a !ccal protected environment.
32 It is important that accident-monitoring instrumentation ccmponents and i
33 their mounts that cannot be located in Seismic Category I buildings be desit;ned l
34 to continue to function, to the extent feasible, during seismic events.
Con-l 35 sequently, it.it is essential that they be designed to resist the effects of 1
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1 seismic excitation. An acceptable method for demonstrating the adequacy of l
2 the seimsic resistance of this instrumentation would be to qualify it to meet 3
the seismic criteria applicable to instrumentation installed at other locations 4
in the plant.
5 Variables selected for accident acnitoring can be selected to provide the 6
essential information needed by the operator to determine if the plant safety 7
functions are being performed. It is essential that the range selections be 8
sufficiently great that the instruments will always be on scale. Further, it 9
is prudent that a limited number of those variables which are functionally 10 significant (e.g., containment pressure, primary system pressure) be sonitored 11 by instruments qualified to more stringent environmental requirements and with 12 ranges that extend well beyond that which the selected variables can attain 13 under limiting conditions; for example, a range for tha containment pressure 14 aonitar extending to the burst pressure of the containment in order that the 15 operator will not be unaware as to the pressure inside containment. Provisions 16 of such instruments are important so that responses to corrective actions can 17 be observed and the need for, and magnitude of, further actions determined.
18 It is also necessary to be sure that when a range is extended, the sensitivity 19 and accuracy of the instrument are within acceptable limits for monitoring the 20 extended range.
1 21 Normal power plant instrumentation remaining functional for all accident 22 conditions can provide indication, records, and (with certain types of instru-23 ments) time-history responses for many variables important to following the 24 course of the accident. Therefore, it is prudent to select the required 25 accident-sonitoring instrumentation from the normal power plant instrumentation 25 to enable the operator to use, during accident situations, instruments with 27 which he i's most familiar. Since some accidents could impose severe operating 28 requirements on instrumentation components, it may be necessary to upgrade 29 those normal power plant instrumentation components to withstand the more 30 severe operating conditions and to measure greater variations of sanitored 31 variables that may be associated with an accident.
It is essential that 32 instrumentation so upgraded does not compromise the accuracy and sensitivity 33 required for normal operation. In some cases, this will necessitate use of 34 overlapping ranges of instruments to monitor the required range of the variable 35 to be monitored, possibly with different performanca requirements in each 36 range.
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1 Standard ANS-4.5,* "Critaria for Accident monitoring Functions in a Light-2 Water-Ccoled Nuclear Power Generating Station," dated 1980, delineates 3
critaria for detarsining the variables to be monitored by the control room 4
operator, as required for safety, during the course of an accident and during 5
the long-tars stable shutdown phase followng an accident. Standard ANS-4.5 6
was prepared by Working Group 4.5 of Subcosmittee ANS-4 with two primary 7
objectives: (1) to address that instrumentation that pemits the operator to 8
monitor expected parameter changes in an accident period and (2) to address 9
extended range instrumentation deemed appropriata for the possibility of 10 encountering previously unforeseen events. ANS-4.5 referencas a revision to 11 IEEE Std 497 as the source for specific instrumentation design critaria. Since 12 the revision to IEEE Std 497 has not yet been completad, its applicability cannot 3
13 yet be determined. Hence, specific instrumentation design critaria have been 14 included in this regulatory guide.
15 The ANS standard defines three variable types (definitions modified herein) l16 for the purpose of aiding the designer in his selection of accident-monitoring j l7 instrumentation and applicable critaria. The types are: Type A - those variables,
- 18 that provide primary ** inforsation needed to pemit the control room operating 19 personnel to take the specified sanually controlled actions for which no automatic l 20 control is provided and which are required for safety systans to ac
- omplish
- 21 their safety functions for design basis accident events. Type 3 - those variables 22 that provide information to indicate whether plant safety functions are being 23 accomplished, and Type C - those variables that provide inforsation to indicate 24 the potential for being breached or the actual breach of the barriers to fission 25 product release, i.e., fuel cladding, primary coolant pressure boundary, and j 26 containment (modified to reflect NRC staff positicn; see Position C.1.2).
The i 27 sources of potential breach are limitad to the energy sources within the barrier 23 29
" Copies say be obtained from the American Nuclear Society, 555 North Kensington 30 Avenue, LaGrange Park, Illinois 60525.
31
- Primary inforsation is that which is essential for the direct accomplishment 32 of the specified safety functions and does not include those variables which 33 are associated with contingency actions that may also be identified in written 34 procedures.
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Il itself. In addition to the accident monitoring variables proviced in ANS-4.5 1
2 standard, variacles for sonitoring the operation of systems important to safety
,3 and radioactive effluent releases are provided by this regulatory guide. Two 4
additional variable types are defined. They are: Type 0 - those variables 5
that provide information to indicate the operation of individual safety systems 6
and other systems important to safety, and Type E - those variables *a be 7
acnitored as required for use in determining the sagnitude of the release of I 8 radioactive saterials and for continuously assessing such releases, 9
A minimum set of Types 3, C, 0, and E variables to be seasured is listad
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10 in this regulatory guide. Type A variables have not been listed because they i
1 11 are plant specific and will depend on the operations that the designer choosas 12 for planned manual action. Types B, C, 0, and E are variables for following 13 the course of an accident and are to be used (a) to determine if the plant is 14 responding to the safety seasures in operation, (b) to inform the opera +4r of l
15 the necessity for unplanned actions to sitigate the consequences of an accident.
16 The five clauifications are not autually exclusive in that a given variable 17 (or instrument) may be applicable to one or acre types, as well as for normal 18 power plant operation or for automatically initiated safety actions. A variable 19 included as Type 3, C, 0, or E does not preclude that variable from being j
20 included as Type A also. Where such multiple use occurs, it is essential that 21 instrumentation be capable of meeting the most stringent requirements.
22 The time phases (Phases I, and II) delineated in ANS-4.5 are not used in 23 this regulatory guide. These considerations are plant specific. It is important 24 that the required instrumentation survive the accident environment and function 25 as long as the information it provides is needed by the control room operating 25 personnel.
27 Regulatory Positions C.1.3 and C.1.4 of this guide provide design and 28 qualification criteria for the instrumentation used to acasure the various i
! 29 variables listed in Table 1 (for SWR) and Table 2 (for PWR). The criteria are 30 separated into three separate groups or categories which provide a graded
- 31 approach to requirements depending on the importance to safety of a variable j
32 being seasured. Category 1 provides the most stringent requirements and is 33 intended for key variables. Category 2 requires less stringent requirements 34 and generally applies to instrumentation designated for indicating system
- 35 operating status. Category 3 is intended to provide requirements which will l
! 36 assure that high quality off-the-shelf instrumentation is obtained and applies 6
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l1 to backup and diagnostic instrumentation.
It is also used where state-of-the-art 2
will not support requirements for higher qualified instrumentation.
3 In general, the asasurement of a single key variable may not be sufficient 4
to indicate the accomplishment of a given safety function. Where multiple 5
variables are needed to indicate the accomplishment of a given safety function, c
6 it is essential that they each be considered key variables and seasured with 7
high quality instrumentation. Additionally, it is prudent, in scae instances, 8
to include the seasurement of additional variables for backup inforsation and 9
for diagnosis. Where these additional seasurements are included, the seasures 10 applied for design, qualification, and quality assurance of the instrumentation 11 need not be the same as that applied for the instrumentation for key variables.
{12 A key variable.is that single variable (or minimum nuncer of variables) that l13 aost directly indicate the accomplishment of a safety function (in the case of 14 Types B & C) or the operation of a system safety (in the case of Type 0) or
!15 radioactive saterials release (in the case of Type E).
It is essential that 16 key variables be qualified to the more strir. gent design and qualification
'17 criteria. The design and qualification critaria category assigned to each 18 variables, indicates whether the variable is considered to be a key variable
!19 or for system status indication or for backup or diagnosis, i.e., for Types B 20 and C, the key variables are Category 1; backup variables are generally Cate-
'21 gory 3.
For Types 0 and E, the key variables are generally ategory 2, backup l22 variables are Category 3.
23 The variables are listed but no mention (beyond redundancy requirements) 24 is made of the number of points of seasurement of each variable.
It is important 25 that the number of points of seasurement be sufficient to adequataly indicate 26 the variable value, e.g., containment temperature say require spatial location 27 of several points of seasurement.
28 This guide provides the sinimum variables to be acnitored by the control 29 roce operating personnel during and following an accident. These variables 30 are used by the control roca operating personnel to perform their role in the 31 emergency plan in the evaluation, assessment, monitoring, and execution of
, 32 control room functions when the other emergency response facilities are not
! 33 effectively sanned. Variables are also defined to permit the operator to l 34 perform his long-ters monitoring and execution responsibilities after the j
i 35 emergency response facilities are sanned. The applicaticq of the critaria for 7
S.
,I the instrumentation is limited to that part of the instrumentation system and i 2 its vital supporting features or power sources which provide the direct display 3
of the variables. These provisions are not necessarily applicable to that 4
part of the intrumentation systans provided as operator aids for the purpose 5
of enhancement of inforsation presentations for the identification or diagnosis
- 6 of disturbances.
7 C.
REGULATORY POSITICN 8
1.
ACCIDENT MONITORING INSTRUMENTATICN 9
The critaria, and requirements, contained in Standard ANS-4.5,"Critaria 10 for Accident Monitoring Functions in a Light-Watar-Cooled Nuclear Power 11 Generating Station," dated 1950, are considered by the NRC staff to 12 be generally acceptable for providing instrumentation to monitor variables for 13 accident conditions subject to the following:
14 1.1 In Section 3.2.1 of ANS-4.5, the definition of Type A variables should
. I5 ' be modified to be as follows: Type A - those variables to be sonitored that I 16 provide the primary information required to permit the control room operator l 17 to take the specified aanually controlled actions for which no automatic control l 13 is provided and which are required for safety systems to accomplish their safety
.f 19 function for design basis accident events. (Nota: Primary information is that
' 20 which is essential for the direct accomplishment of the specified safety function 21 and does not include those variables which are associated with contingency actions j
l 22 that say also be identified in written procedures.)
23 L2 In Section 3.2.3 of ANS-4.5, the definition of " Type C" includes two 24 itans, (1) and (2). Itan (1) includes those instruments that indicate the extant 25 to which parameters which have the potential for causing a breach in the primary 26 reactor containment have exceeded the design basis values.
In conjunction with 27 the parameters that indicate the potential for causing a breach in the primary 28 reactor containment, tne parameters that indicate the potential for causing a 29 breach in the fuel cladding (e.g., core exit tamperature) and the reactor coolant 8
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1 pressure boundary (e.g., reactor coolant pressure) should also be included.
2 References to Type C instruments, and associatad parameters to be measured, in 3
Standard ANS-4.5 (e.g., Sections 4.2, 5.0, 5.1.3, 5.2, 6.0, 6.3) should include 4
this expanded definition.
c 5
- 1. 3 Section 6.1 of ANS-4.5 pertains to General Criteria for Types A, 3, 6
and C accident monitoring variables.
In lieu of Section 6.1, the following 7
design and qualification criteria categories should be used:
i ii!8 1.3.1 Desicn and Qualification Criteria - Catacory 1 i
!9 (1) The instrumentation should be qualified in. accordance with
!10 Regulatory Guide 1.89 (NUREG-0588). Qualification applies to the completa
' ll instrumentation channel from sensor to display where the display is a direct-12 indicating meter or recording device. Where the instrumentation channel signal 13 is to be used in a computar-based display, recording and/or diagnostic progras, 14 qualification applies to and including the channel isolation device. The 15 location of the isolation device should be such tha't it would be accessible 16 for saintenance during accident conditions. The seismic portion of qualification 3
{17 should be in accordance with Regulatory Guide 1.100.
Instrumentation should
!18 continue to read within the required accuracy following, but not necassarily l19 during, a safe shutdown earthqtake.
Instrumentation, whose ranges are required f20 to extend beyond those ranges calculated in the sost severe design basis accident j 21 event for a given variable, should be qualified using the guidance provided in
'22 paragraph 6.3.6 of ANS-4.5.
i 23 (2) No single failure within either the accident-sonitoring instrumenta-
' 24 tion, its auxiliary supporting teatures or its power sources concurrent with 25 the failures that are a condition or result of a specific accident, should prevent
' 26 the operator free being presented the inforsation necessary for his to detersining 27 the safety status of the plant and to bring the plant to and maintain it in a 28 safe condition following that accident. Where failure of one accident-monitoring 29 channel results in information ambiguity (that is, the redundant displays disagree)
- 30. which could lead the operator to defeat or fail to accomplish a required safaty
, 31 function, additional information should be provided to allow the operator to 3
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I deduce the actual conditions in the plant. This say be accomplished by providing 2
additional independent channels of information of the same variable (addition of l3 an identical channel), or by providing an incependent channel which monitors a j4 different variabla which bears a known rolatienship to the sultipie channels l5 (addition of a diverse channel), or by providing the capability, if sufficient
!6 time is available, for the operator to perturb the seasured variable and deter-l7 aine which channel has failed by observation of the response on each instrumenta-8 tion channel. Redundant or diverse channels should be electrically independent 9
and physicaily separated in accordance with Regulatory Guide 1.75 up to and
!10 including any isolation device. At least one channel should be displayed on a f11 direct-indicating or recording device. (NOTE: Within each reduitdant division l 12 of a safety system, redundant acnitoring channels are not needed.)
13 (3) The instrumentation should be energi:ed from station Standby 14 Power sources.
! 15 (4) An instrumentation channel should be available prior to an i 16 accident except as provided in Paragraph 4.11, " Exemption", as defined in IEEE l17 Std 279 or as specified in Technical Specifications.
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! 18 (5) The recommendations of the folicwing regulatory guides
! 19 pertaining to quality assurance should be followed:
" Quality Assurance Program Requirements (Design 21
& Construction)"
22 Regulatorf Guide 1.30
" Quality Assurance Requirements for the Installation,
! 23 Inspection, and Testing of Instrumentation and i 24 Electric Equipment" 25 Regulatory Guide 1.38
" Quality Assurance Requirements for Packaging, 26 Shipping, Receiving, Storage, and Handling of 27
!tems for Water-Cooled Nuclear Power Plants" 28 Regulatory Guide 1.58
" Qualification of Nuclear Power Plant Insptaction, 29 Examination, and Testing Personnel" 30 Regulatory Guide 1.64
" Quality Assurance Requirements for the Design
, 31 of Nuclear Power Plants" 10
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" Quality Assurance Terms and Definitions" s,.
" Collection, Storage, and Maintenance of Nuclear 3
Power Plant Quality Assurance Records" 4
" Quality Assurance Requirements for Control of 5
Procurement of Items and Services for Nuclear 6
Power Plants" 7
Regulatory Guide 1.144 "Aediting of Quality Assurance Programs for Nuclear 8
Power Plants" 9
Task RS 810-5
" Qualification of Quality Assurance Program Audit 10 Personnel for Nuclear Power Plants" (Guide nummer 11 to be inserted.)
12 Reference to the above regulatory guides (except Regulatory Guides 1.30, and 13 1.38) are being made pending issuance of a regulatory guide endorsing NQA-1 14 (Task RS 002-5) which is in progress.
15 (6) Continuous indication (it may be by recording) display should 16 be provided.' Where two or more instruments are needed to cover a particular 17 range, overlapping of instrument span should be provided.
18 (7) Recording of instrumentation readout information should be pro-19 vided. Where direct and immediate trend or transient information is essential 20 for operator information or action, the recording should be analog stripchart.
21 Otherwise, it say be continuously updated, computer memory stored, and displayed 22 on demand.
Intermittent displays, such as data loggers and scanning recorders, 23 may be used if no significant transient response information is likely 'A be 24 lost by such devices.
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- 25 1.3.2 Design and Qualification Criteria - Category 2 26 (1) The instrumentation should be qualified in accordance with Regula-l 27 tory Guide 1.89 (NUREG-0538). Where the channel signal is to be processed or
' 28 displayed on demand, qualification applies from the sensor through the isolator /
i 29 input buffer. The location of the isolation device should be such that it would 30 be accessible for saintenance during accident conditions.
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(2) The instrumentation should be energized from a high reliability 2
power source, not necessarily Standby Power, battary backed where momentary interrup-3 tion is not tolerable.
4 (3) The out-of-service interval should be based on normal Technical
- 5 - Specification requirements on out-of-service for the system it serves where i6 applicable or where specified by other requirements.
I 7
(4) The recommendations of the following regulatory guides j
'8 pertaining to quality assurance should ce followed:
" Quality Assurance Program Requirements (Design
- 10
& Construction)"
" Quality Assurance Requirements for the Installation, 12 Inspection, and Testing of Instrumentation and 13 Electric Equipment" 14 Regulatory Guide 1.38
" Quality Assurance Requirements for Packaging,
- 15 Shipping, Receiving, Storage, and Handling of 16 Items for Watar-Cooled Nuclear Power Plar.ts" l 17 Regulatory Guide 1.58
" Qualification of Nuclear Power Plant Insptection,
! 18 Examination, and Testing Personnel" l 19 Regulatory Guide 1.64
" Quality Assurance Requirements for the Design i 20 of Nuclear Power Plants" i
21 Regulatorf Guide 1.74
" Quality Assurance Terms and Definitions" i 22 Regulatory Guide 1.88
- Collection, Storage, and Maintenance of Nuclear 23 Power Plant Quality Assurance Records" l 24 Regulatory Guide 1.123
" Quality Assurance Requirements for Control of
- 25 Procurement of Items and Services for Nuclear I
- 25 Power Plants" l
" Auditing of Quality Assurance Programs for Nuclear
- - - 28 Power Plants" 29 Task RS 810-5
" Qualification of Quality Assurance Pmgram Audit 30 Personnel for Nuclear Power Plants" (Guide number 31 to be inserted.)
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1 Reference to the above regulatory guides (except Regulatory Guides 1.30, and 2
1.38) are being made pending issuance of a regulatory guide endorsing NQA-1 3
(Task RS 002-5) which is in progress. Since some instrumentation is less 4
important to safety than other instrumentation, it say not be necessary to apply 5
the same quality assurance seasures to all instrumentation. The quality assurance 6
requirements, which are implemented, should provide control over activities 7
affecting quality to an extent consistent with the incertance to safety of the l8 instrumentation. These requirements should be detarsined and documented by l9 personnel knowledgeable in the end use of the instrumentation.
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- 10 (5) The instrumentation signal may be displayed on an individual 11 instrument or it say be processed for display on demand by a CRT or other appro-12 priate means.
13 (6) The method of display may be dial, digital, CRT or stripchart 14 recorder indication. Effluent release monitors should be recorded, including 15 effluent radioactivity sonitors, environs exposure rate acnitors, and meteorology j 16 monitors. Where direct and immediate trend or transient information is essential l17 fcr operator information or action, the recording should be analog stripchart.
i 18 Othenvise, it may be continuously updated, computer temory stored, and displayed l 19 on demand.
20 1.3.3 Design and Qualification Criteria - Category 3
' 21 (1) liigh quality cosmiercial grade instrumentation selected to with-22 stand the specified service environment.
23 1.4 In addition to the criteria of Position C.1.3, the following criteria should 24 apply to Categories 1 and 2:
25 1.4.1 Any equipment that is used for either Category 1 or Categor/ 2 26 should be designated as part of accident monitoring or systems operation 27 and effluent monitoring instrumentation. The transmission of signals from 2S such equipment for other use should be through isolation devices that are 13 l
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1 designated as part of monitoring instrumentation and that meet the provisions
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2 of this document.
3 1.4.2 The instruments designated as Types A, B and C and Categories 1 and 4
2 should be specifically identified on the control panels so that the operator 5
can easily discern that they are intended for use under accident conditions.
6 1.5 In addition to the above criteria, the following should apply to Categories
/
1, 2 and 3.
8 1.5.1 Means should be provided for checking, with a high degree of confidence 9
the operational availability of each monitoring channel, including its input 10 sensor, during reactor operation. This may be accomplished in various ways, 11 for example:
12 (1) By perturbing the sonitored variable 13 (2) By introducing and varying, as appropriate, a substitute input 14 to the sensor of the same nature as the measured variable; or.
15 (3) By cross-checking between channels that bear a known relation-16 ship to each other and that have readouts available.
17 1.5.2 Servicing, testing, and calibration programs shoulit be specified 18 to maintain the capability of the monitoring instrumentation.
For those 19 instruments where the required interval between testing will be less than the 20 nonaal time interval between generating station shutdowns, a capability for 21 testing during power operation should be provided.
22 1.5.3 Whenever means for removing channels from service are included in
- 23 the design, the design should facilitate administrative control of the access 24 to such removal means.
25 1.5.4 The design should facilitate administrative control of the access 26 to all setpoint adjustments, module calibration adjustments, and test points.
14
1 1.5.5 The monitoring instrumentation design should minimize the development 2
of conditions that would cause meters, annunciators, recorders, alarms, etc.,
3 to give anomalous indications potentially confusing to the operator.
4 1.5.6 The instrumentation should be designed to facilitate the recogni-5 tion, location, replacement, repair, or adjustment of malfunctioning components 6
or modules.
7 1.5.7 To the extent [ptsetiesi] possible, monitoring instrumentation inputs 13 should be from sensors that directly measure the desired variables.
9 1.5.8 To the extent practical, the same instruments should be used for 10 accident monitoring as are used for the normal operations of the plant to enable 11 the operator to use, during accident situations, instruments with which he is 12 most familiar. However, where the required range of monitoring instrumentation 13 results in a loss of instrumentation sensitivity in the normal operating range, 14 separat'e instruments should be used.
15 1.5.9 Periodic testing should be in accordance with the applicable portions.
16 of Regulatory C 118 pertaining to testi-f instruments channels.
(Note-17 Response time testing not usually needed.)
18 1.6 Sections 6.2.2, 6.2.3, 6.2.4, 6.2.5, 6.2.6, 6.3.2, 6.3.3, 6.3.4, and 19 6.3.5 of ANS-4.5 pertain to variables and variable ranges for monitoring Types B 20 and C variables.
In conjunction with the above sections, Tables.1, and 2 of 21 this regulatory guide (which include those variables mentioned in the above 22 sections) should be used as the minimum set of instruments and their respective 23 ranges for accident-monitoring instrumentation for each nuclear power plant.
12 4 2.
SYSTEMS OPERATION MONITORING AND EFFLUENT RELEASE MONITORING INSTRUMENTATION 25 2.1 Definitions
, 26 2.1.1 Type 0 - those variables that provide information to indicate 27 the operation of individual safety systems and other systems important to safety.
15
.t f
6 -
1
~
1 2.1.2 Type E - those variables to be monitored as required for use in 2
deternining the magnitude of the release of radioactive saterials and continually 3
assessing such releases.
4 2.2 The plant designer should select variables and information display 5
channels required by his design to enable the control roca operating personnel 6
to:
7 2.2.1 Ascertain the operating status of each individual safety system 8
and other systems important to safety to that extent necessary to determine if j
9 each system is operating or can be placed in operation to help sitigate the 10 consequences of an accident.
11 2.2.2 Monitor the effluent discharge paths and environs within the i
i 12 site boundary to ascertain if there has been significant releases (planned or l13 unplanned) of radioactive saterials and for continually assessing such releases.
l14 2.2.3 Obtain required information through a backup or diagnosis
__. }15 _ channel.where a single channel say be likely to give ambiguous indication.
j i
1jl6 2.3 The process for selecting system operation and effluent release I 17 variables should include the identification of:
18 2.3.1 For Type 0 19 (1) the plant safety systems and other systems important to safety 20 which should be operating or which could be placed in operation to help zitigate 21 the consequences of an accident; 22 (2) the variable or ainimum list of variables that indicate the 23 operating status of each system identified in (1) above.
24 2.3.3 For Type E 25 (1) the planned paths for affluent release; 16
.- T...
- e. '
j 1
(2) plant areas and inside buildings where access is required to 2
service equipment necessary to mitigate the consequences of an accident; 3
(3) onsite locations where unplanned releases of radioactive 4
satarials should be detected; 5
(4) the variables that should be monitored in each location l
6 identified in (1), (2), and (3) above.
)
7
- 2. 4 The detarmination of performance requirements for system operation j8 monitoring and effluent release monitoring inforsation display channels should
!9 include, as a sinf aum, identification of:
j 10 (1) the range of the process variable.
fil (2) the required accuracy of measurement.
- 12 (3) the required response characteristics.
l13 (4) the time intarval during which the measurement is needed.
l14 (5) the local environment (s) in which the information display
- 15 channel components aust operata.
l16 (6) any requirement for rate or trend information.
,17 (7) any requirements to group displays of related information.
!18 (8) any required spatial distribution of sensors.
I l19 2.5 The design and qualification critaria for system operation monitoring
[20 and effluent release monitoring instrumentation should be taken from the criteria j
!21 provided in Positions C.1.3 and C.1.4 of this guide. Tables 1 and 2 of this 22 regulatory guide should be should be used as a ainimum set of instruments and 23 their respectives ranges for systems operation monitoring (Type 0) and effluent
,24 release sonitoring (Type E) instrumentation for each nuclear power plant.
25 07fMPLEMENTATION 26 All plants going into operation after June 1982 should meet the provisions 27 of this guide.
l 17 l
I
~.
b
- L-1 Plants currently operating or seneduled to be licensed to cperate before 2
June 1, 1982 should meet the requirements of N(JREG-0578 and NRR letters dated 3
Septancer 13, and October 30, 1979. The provisions of this guide as specified 4
in Tables 1, and 2 for operating plants are compatible with these documents 5
which are to be completed by January 1,1981. The balance of provisions of 6
the guide are to be completed by June 1983.
7 The difficulties of procuring and installing additions or modifications 8
to in place instrumentation have been considered in es*4 11shing these schedules.
9 Exceptions to requirements and schedules will be considered for extraordinary 10 circumstances.
18
TABLE 1 BWR VARIABLES Type A - Those variables to be monitored that provide the primary information required to permit the control room operator to take specific manually controlled actions for which no automatic control is provided and which are required for safety systems to accomplish their safety function for design basis accident events. Primary information is that which is essential for the direct accomplish-ment of the specified safety function and does not include those variables wnich are associated with contingency actions that may also be identified in written procedures.
A variable included as Type A does not preclude it from being included as Type B, C, 0, or E, or vice versa.
Category (see Variable Range Posttton C.I.3)
Purtose Plant specific plant specific 1
Information required for operator action l
10
TABLE 1 SWR VARIABLES (continued)
TfPE B Yariables - Those variables that provide informtion to indicate whether plant safety functions are being accomolished. Plant safety functions are (1) reactivity control, (2) core cooling, (3) maintaining reactor coolant system integrity, (4) containment integrity (which includes radioactive effluent centrol).
Variables are listed with designated ranges and category for design and qualification requirements. Key variables are indicated by design and qualification Category 1.
Category (see Variable Range Position C.1.3)
Purecse TYPE B VARIABLES Reactivity Control Neutron Flux 10-sto 5% full power 1
Function detection; Accomplishment of mitigation Control Rod Position Full in or not 3
7erification full in RCS Soluble Baron 0 to 1000 ppm 3
Verification Concentration Core Cooling Coolant Level in Bottom of core support 1
Function datection; the Reactor plate to above the top Accomplishment of of discharge plenum mitigation; Lcng-term surveillance 31:R Core Ther:noccuples Unresolved s To monitor core cooling if water level is low, spray is lost, or channels restricted.
20
- ~
e TABLE 1 (continued)
Category (.see
~
Variable Range Position C.I.3)
Purcose TYPE B - continued
~
Maintaining Reactor Cool-ant System Integri ty RCS Pressure 15 psia to 2000 psig 1"
Function detection; Accomplishment of mitigation; Verification l
2 1
Function detection;'
Drywell Pressure O to design pressure (psig)
Accomplishment of mitigation; Verification l
Drywell Sump Level 3ottom to top 2
Function detection; Accomplishment of aitigation; Verification Maintaining Containment Integrity Primary Contai ment 10 psia to design 1
Function detection; 2
Pressure (Drywell)1 pressure Accomplishment of aitigation; Verification Primary Containment Closed - not closed 1
Accomplishment of Isolation Valve Pos-isolation icion (excluding check valves) 21
TABLE 1 (continnd)
BWR VARIABLES (continued)
TYPE C Variables - Those variables that provide infornation to indicate the potential for being breached or the actual breach of the barriers to fission product releases.
The barriers are: (1) fuel cladding, (2) primary coolant pressure boundarf, and (3) containment.
Categorf' (see Variable Range Position C.1.3)
Purcose TYPE C VARIABLES Fuel Cladding 6
Radioactivity Concen-31 Tech Spec limit to 1
Detectica of breach tration or Radiation 100 Times Tech Spec Level in Circulating limit R/hr Pri: nary Coolant 17 Accident Sampling and 10 pCi/gm to 10 C1/gm 3
Detail analysis; Analysis of Primary or TID-14844 source term Accomplishment of Coolant in coolant volume mitigation; Gross Activity Verification; Gansna Spectrum Long-ters surveillance 3WR Core Thermocouples Unresolveds To monitor core cool-ing if water level is low, spray is lost, or channels restricted Reactor Coolant Pressure Bouncary RCS Pressurel 15 psia to 1500 psig 1"
Detection of potential for or actual breach; Accomplishment of mitigation; Long-term surveillance 7 11 Primary Contai:nnent 1 R/hr to 10s R/hr 3
Detection of breach; Area Radiatiar.11 Verification l
Drywell Drain Sumps Bottom to top 2
Detection of breach; Level (Identified and Accomplishment of Unidistified Laakage) mitigation; Verification; Long-term surveillance Suppression Pool Water Bottom of ECCS suction 1
as Meheh Level (for operating line to Sft above normal plants) water level 22
a
^
TABLE 1 (continu:d)
Category (see Variable Range Position C.I.3)
Purcose TYPE C - continued Reactor Coolant Pressure Bouncary (continued) l 2
Dryvell Pressure O to design pressure 1
Detection of breach; (psig)
Verification Containment RCS Pressurel 15 psia to 1500 psig i
Detection of potential for breach; Accomplishment of sitigation l
Primary Containment 10 psia pressure to 3 1
Detection of potential 2
for or actual breach; Pressure (Drywell) times design pressure for concrete; 4 times Accomplishment of design pressure for steel mitigation Containment and Dry-O to 30 (capability of 1
Detection of potential well Hydrogen Con-operating from 12 psia to for breach; 2
centration design pressurs )
Accomplishment of nitigation Containment and Dry-O to 10*: (capability of 1
Detection of potential well Oxygen Concen-operating from 12 asia for breach; tration (for inerted to design pressure-)
Accomplishment of cuntainment plants) mitigation U
Containment Effluenel 10-6 to 10-2 uCi/cc 3
Detection of actual Radioactivity - Notle breach; Gases (from identified Accomplishment of release points includ-mitigation; ing Standby Gas Treat-Verification ment System Vent)
Environs Radioactiv-10-8tc 10 R/hr 2
Detection of breach; l
icy - Exposure Raca Accomplishment of mitigation; Verification 23
~
.s TABLE 1 (continued) 3WR vpRIABLES (continued)
TYPE D Variables - those variabits that provide information to indicate the oper-ation of individual safety systems and other systems important to safety. These va"iables are to help the operator make appropriate decisions in using the indi-viccal systens important to safety in mitigating the consequences of an accident.
Category (see Range Position C.I.3)
Purcose Variable TYPE O VARIABLES Condensate and Feed-water System 3
Main Feedwater Flow 0 to 110% design flow 3
Detection of operation; Analysis of cooling Condensate Storags Bottom to top 3
Indication of avail-Tank Lorcl able water for cool-ing Primary Containment-Related Systems Suppression Chamber 0 to 110t design flow 3 2
To monitor operation j
Spray Flow Drywell Pressure 12 psia to 3 psig 2
To monitor operation 2
l 0 to 110% design pressure Suppression Pool Top of vent to top of 2
To monitor operation Water Level weir well i
Suppression Pool 30*F to 230*7 2
To monitor operation Water Temperature Drywell Atmosphere 40*? to 440*F 2
To monitor operation Temperature 1
24
TABLE 1 (continued)
Category (see Variable Range Position C.1.3)
Purcose TYPE D - continued Main Steam System 3
Main Steamline Flow 0 to 120% design flov 1
To :nonitor operation Main Steamline Isola-O to 15" of water 1
To provide indication tion Valves' Leakage O to 5 psid of pressure boundary Control System Pressure saintenance Primary System Safety Closed-not closed or 1
Detection of accident; Relief Valve Positions, O to 50 psig boundary integrity in-Lacluding ADS or Flow dication Through or Pressure in 7alve lines 25
TABLE 1 (continued)
Category (see Variable Rance Pasitien C.1.3)
Purcose TYPE O - continued Safety Systems 3
2 To monitor operation RCIC Flow 0 to 110% design flow 3
2 To monitor operation HPCI Flow 0 to 110% design flew 3
2 To monitor operation Core Spray Flow 0 to 110% design flow 3
2 To monitor operation RHR System Flow 0 to 110% design flov (LPCI)
RER Heat Exchanger 32*F to 350*F 2
To monitor operation Outlet Temperature (LPCI) 3 SLCS Flow 0 to 110% design flov 3
To monitor operation SLCS Storage Tank Bottom to top 3
To monitor operacica Level 26
^ :.
~..
^
e TABLE 1 (continued)
Category (see Variable Range Position 0.1.3)
Purcose TYPE D - continued i
Cooling Water System ESF System Component 32*F to 200*F 2
To monitor operation i
Cooling Water Temper-acura 1
ESF System Component 0 to 110% design flow 3 2
To :nonitor operation Cooling Water Flow i
4 Radwaste Systems High Radioactivity Top to bottom 3
To monitor operation Liquid Tank Level i
Ventilation Systems Emergency Ventilation Open-closed status 2
To monitor operation Damper Position i
Power Supplies Status of Otandby Pow-Voltages, currents, 2
To :nonitor operation er & Other energy pressures Sources Important to Safety 27
TABLE 1 (continued)
BWR VARIABLES ( continued )
TYPE E Variables - Those variables to be monitored as required for use in detennin-ing the magnitude of the release of radfoactive materials and continually assessing such releases.
Category (see Variable Range Position C.1.3)
Purcose TYPE E VARIABLES Containment Radiation 7 11 7 R/hr 1
Detection of signif-Primary Containment 1 R/hr to 10 Area Radiation -
icant releases; l
High Range Release assessment; Long-term surveillance!
Emergency plan actuation 4
10 Reactor Bldg or Sec-10-6 to 10 uC1/cc 2
Detection of signif-ondary Cont =f - ne icant releases; Area Radiation Release assessment Lcng-term surveillance Area Radiation Radiation Exposure 10-1 R/hr to 10 R/hr 2"
Detection of signif-4 Race (Inside b1dgs or icant releases; areas where access is Release assessment; required to service Long-term surveillance equipment important to safety)
Airborne Radioactive Materials Released frem the Plant Noble Gases and Vent Flow Rate o Drywell Purge, Stand-10-6 to 105 uC1/cc 2"
Detection of signif-by Gas Treatment Sys-O to 110% vent design icant releases; tem Purge (for Mark flow 3 Release assessment I, II, III plants) &
(Not needed if effluent Secondary Containment discharges thru common Purge (for Mark I plants) plant vent) o Secondary Containment 10-8 to 104 pC1/cc 2
Detection of signif-Purge (for Mark I, II, III O to 110% vent design icant releases; plants) flow 3 Release assessment (Not needed if effluent discharges thru cotanon plant vent) 28
~
~
e
p TABLE 1 (continued)
Category (see Variable Range Position C.I.3)
Purcose TYPE E - continued Airborne Radioactive Materials Released from the Plant Noble Gases and Vent Flow Hate (continued) 4 10 o Secondary Contain-10 6 to 10 uC1/cc 2
Detection of signif-ment (reactor shield 0 to 110% vent design icant releases; 3
bldg annulus, if in _
flow Release assessment design)
(Not needed if effluent dis-caarges thru common plant vent) i 4
10 Auxiliary Building 10 8 to 10 uCi/cc 2
Detection of signif-o (including any b1dg 0 to 110% vent design icant releases; 3
containing primary flow Release assessment; system gases, e.g.,
(Not needed if affluent dis-Long-term surveillance waste gas decay tank) charges thru common plant vent) 3 10 o Common Plant Vent or 10-6 to 10 uC1/cc 2
Detection of signif-Multi-purpose Vent 0 to 110% vent design icant releases; Discharging Any of flow 3 Release assessment; the Above Releases Long-term surveillance 10 o All Other Identified 10-6 to 102 uCi/cc 2
Detection of signif-Release Points 0 to 110% vent design icant releases; i
flow 3 Release assessment!
(Not needed if effluent dis-Long-term surveillance i
charges thru other monitored plant vents)
Particulates and Halogens 2
13 o All Identified Plant 10-3 to 10 uCi/cc 3
Detection of signif-Release Points.
O to 110% vent design icant releases; Sampling, with Onsite flow 3 Release assessment; Analysis Capability Long-term surveillance i
29
l TABLE 1 (con 11nued)
Catescr/ (see
'la ri abl e Range Position C.I.3)
Purtose TYPE E - continued 1
Env' irons Radiation and lacicactivi ty 11 Radiation Exposure 10-0 R/hr to 10 R/hr 2
Detection of signif-Racal icant releases, (Installed instrument-7erification; ation)
Release assessment; Long-term surre111ance Airborne Radichalogens 10-9 to 10-3 uC1/cc 3
Release assessment; and Particulates Analysis (Sampling, with on-site analysis cap-ability) 15
?lant and Environs 0.1 to 10*- R/hr, photons 3
Release assessment; 33 Radiation 0.1 to 10* rads /tr, beta 3
Analysis (Portable lostrument-radiations aod low-energy ation) photons Plant and Environs h iti-ch m al Gamma-Ray 3
Raleases assessment; Radioactivity spectrometer Analysis (Portable Instrument-4 tion) l l
r t
l 30
TABLE 1 (c::ntinued)
Categor/ (see Variable Range Pesition C.I.3)
Purtose TYPE E - Centinued 18 METEOROLOGY Wind Direction 0 to 360* (:5* accuracy 3
Release assessment with a deflection of 15*.
Starting speed 0.45 aps (1.0 mph). Damping ratio betwen 0.4 and 0.6, dis-cance constant 12 meters.
Wind Speed 0 to 30 sps (67 sph) 20.22 3
Release assessment sps (0.5 mph) accuracy for wind speeds less than 11
- nps (25 mph), wi.h a start-ing threshold of less than 0.45 mps (1.0 mph).
Estimation of Atmos-Based on vertical tager-3 Release assessment phric Stability ature difference from pri-mary system, -5'C to 10*C
(-9*7 to 13*7) and 20.15'C accuracy per 50 matar int-ervals (20.3*7 accuracy per 164 foot intervals) or analogous range for back-up system.
31
ATABLE:1 (continu d)
Catescry (see
'la ri abl e Range Position C.1.3)
Purcose TYPE E - (continued)
ACCIDENT SAMPLING CAP
- ABILITf ( Analysis Cap-aoility Onsite)
Primary Coolant & Sump Grab Sample Y"
Release assessment; Verification; o Gross Actiit:y 10 uC1/mi :o 10 C1/al g7,3 o Gamea Spectrum (Isotopic Analysis) o 3eren Centent 0 to 1000 ppa o Chloride Centent 0 to 20 ppa o Disolved Oxygen 0 to 20 ppm o pH 1 to 13 Courminment Air Grab Sample i7 Release assessment; Verif1:stica; o Hydrogen Centen:
0 to 10%
hlysh 0 to 30% for inerted containments o Oxygen Content 0 to 30%
o Gamma Spectrum (Noble gas analysis) l l
- The time for enking and analysing samples should he 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 the decision is made to sample, except chloride which should be wi:hin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
32 l
- -\\
y
. v..,
TABLE 1 (continued)
NOTES IWhere a variable is listed for more than one purpose, the instrumentation requirements l
may be integrated and only one measurement provided.
Design pressure is that value corresponding to ASME code values that are obtained at or j
2 below code-allowable material design stress values.
3Design flow is the maximum flow anticipated in normal operation.
l "The maximum value may be revised upward to satisfy ATWS requirements.
1 5
l The number.of thermocouples,.their range and: Locatten es Se determined.
4 1
6Measurement should be made of the gross gama radiation emanating from circulating pri-i i
mary coolant, with instrument calibration permitting conversion of readout to radioac-tivity concentrations in terms of either curies / gram or curie;/ unit-volume.- System i
accuracy should be t\\ order of magnitude. The point of measurament should be external to a circulating primary coolant line or loop, and should'not be a line or loop subject 4
to isolation, e.g., main steam line. While such an instrument may noe. be currently available off-the-shelf, the staff considers that the necessary components are avail-able comercial17 and have been employed and demonstrated under adverse envirerr antal l
conditions in high-level hot cell operations for many years.
\\
7Minimum of two monitors at widely separated locations.
87or estimating release ra'tes of radioactive materials released during an accident from unidentified release paths (not covered by effluent monitors) - continuous readout capability.
(Approximately 16 to 20 locations - site dependent.)
' Provisions should be made to monitor all identified pathways for release of gaseous i
radioactive materials to the environs in conformance with General Design Griterion 64.
Monitoring of individual effluent streams only is required where such streams are re-leased directly into the environment. If two or more streams are combined prior to release 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 measure worst-case releases.
"Honitors should be capable of detecting and measuring radioactive gaseous effluent con-centrations with compositions ranging from fresh equilibrium noble gas fission product mixtures to 10-day old mixtures, with overall system accuracies of t decade. Effluent i
~
concentrations may be expressed in terms of Xe-133 equivalents or in terms of any noble gas nuclide(s). Itisnotexpectedthatasinpamonitoringdevicewillhavesufficient
' range to encompass the entire range prcvided in this guide and that multiple components or systems will be needed. Existing equipment may be utilized to monitor any portion of the stated range within the equipment design rating. Additional extended range instru-mentation should overlap the range of existing instrumentation by a least a factor of 2.
]
11 Detectors should respond to gamma radiation photons within any energy range from 60 kev
- to 3 MeV with an accuracy of 220*. at any specific photon energy from 0.1 MeV to 3 MeV.
Overall system accuracy should be within t\\ decade ever the entire range. -
t 33 1 -
~ -
TABLE 1 (continued) l NOTES - continued 123g,g,' indication of all Standby Power A-C busess, D-C buses, inverter output buses and poumatic supplies.
13To provide information regarding release of radioactive halogens and particulates.
Continous collection of representative samples followed by onsite laboratory measure-ments of samples for radiohalogens and particulates. The design envelope for shield-ing, handling, and analytical purposes should assume 30 minutes of integrated sampling 2 uCi/cc of radiciodines time at sampler design flow, an average concentrations of 10 2
in gaseous or vapor form, an average concentration of 10 uCi/cc of particulate radio-iodines and particulates other than radiciodines, and an average gamma photon energy of 0.5 MeV per disintegration.
t 14For estimating release rates of radioactive materials released during an accident.from unidentified release paths (not covered by affluent monitors).
Continous collection of representative samples followed by laboratory measurements of the samples.
(Approximately 16 to 20 locations - site dependent.)
ISTo monitor radiation and airborne radioactivity concentrations in many areas through-out the facility and the site environs where it is impractical to install stationary monitors capable of covering both normal and accident levels.
16 Meteorological measurements should conform to the' provisions of the forthcoming rev-ision of Regulatory Guide 1.23, "Onsite Meteorological Programs".
17Sampling or monitoring of radioactive liquids and gases should be performed in a manner which assures procurement of representative samples. For gases, the criteria of.CSI N13.1 should be applied. For liquida, provisions should be made for sampling from well-mixed turbulent zones and sampling lines should be designed to minimi:e plateout or deposition. For safe and convenient sampling, the' provisions should include:
a.
Shielding to maintain radiation deses ALARA, n.
Sample containers with container-sampling port connector compatability, c.
Capability of sampling under pr; Unary system pressure and negative pressures, d.
Handling and transport capability, and e.
Pre-arrangement for analysis and interpretation.
18 n installed capability should be provided for obtaining containment sump, ECCS pump A
room sumps, aad other similar auxiliary building sump liquid samples.
34
TABLE 2 PWR VARIABLES Type A - Those variables to be monitored that provide the primary in' formation required to permit the control room operator to take specific manually controlled actions for which no automatic control is provided and which are required for safety systems to accomplish their safety function for design basis accident events. Primary information is that which h essential for the direct accomplish-ment of the specified safety function and does not include those variables which are associated with contingency actions that may also be identified in written procedures.
A variable included as Type A does not preclude it from being included as Type B, C, 0, or E, or sice versa.
Category (C.I.3) see Variable Range Position Purcose Plant specific plant specific 1
Information required for operator action e
b 4 e
G 35
TABLE 2 PWR VARIABLES (continued)
Type B Variables - Those variables that provide information to indicate whether plant safety functions are being acccmplished. Plant safety functions are (1) reactivity control. (2) core cooling,(w(3) maintaining reactor ecolant system integrity, (4) containment integrity hich includes radioactive effluent control).
Variables are listed with designated ranges and category for design and qualification requi rements. Key variables are indicated by design and qualification Category 1.
Category (see Variable Range Position C.1.3)
Purcose TYPE B VARIABLES Reactivity Control Neutron Flux 10-6 to 5% full power 1
Function detection; Accomplishment of mitigation.
Control Rod Position Full in or not 3
Verification full in RCS Soluble acron 0 to 6000 ppm 3
Verification Concentration RCS. Cold Leg Temper-50*F to 400*F 3
Verification l
a ture Core Cooling RCS Hot Leg Temper-50*F to 750*F 1
Function detection; ature Accomplishment of mitigation; Verification; Long-tr.m surniilmce RCS Cold Leg Temper-50*F to 750*F 1
Function detection; l
ature Accomplishment of mitigation; Verification; Long-term surveillance O to 3000 psig 1"
Function detection; RCS Pressural
(,4000 psig for CE Accomplishmant of plants) mitigation; Verification; Long-term surveillance 36
- ~.
TAaLE 2 (.continuedl Category (see Variable Range Position C.1.3)
Purocse TYPE B - centinued Core Cooling (continued)
Core Exic Temperature 150*F to 2300*F 3'
Verification l
(for operating plants -
150*F to 1650*F)
Coolant Level in Bottom of core to 1 (Direct Verification Reactor top of vessel indicating or re-cording device not needed)
Degrees of Subcooling 200*F subcooling to 1
Verification and 35*F superheat (for operating analysis of plant planca - 2, with conditions confirmatory oper-ator procedures)
Maintaining Reactor Coolant, System Integrity RCS Pressurel 0 to 3000 psig i
Function detection; (4000 psig for CE Accomplishment of plants) nicigation Containment Sump Water Narrow rar.ge (sump),
3 Function detection; Level Wide range (bottom of 1
Accomplishment of containment to 600,000-sitigation; gallon level equivalent)
Verification Containment Pressure!
O to design pressure 1
Function detection; 2
(psig)
Accomplishment of altigation; Verification 37
TABLE 2 (continued)
Category (see Variable Rance Posttfon C.I.3) ~
Purcose TYPE B - continued Maintaining Containment Integri ty Containment Isolation Closed-not closed 1
Accomplishment of Valve Position (exclud-isolation ing check valves) 4 38 7
a TABLE 2 (continued)
PWR VARIABLES (continued)
TYPE C Variables - Those variables that provide infor: nation to indicate the potential for being breached or the actual breach of the barriers to fission product releases.
The barriers are: (1) fuel cladding, (2) primary coolant pressure boundary, and (3) containment.
Category (see Variable Range Position ~C.1.3)
Purcose TYPE C VARIABLES Fuel Cladding s
l 150*F to 2300*F i
Detection of potential Core Exit Temperatura (for operating plants -
for breach; 150*F to 1650*F)
Accomplishment of mitigation; Long-term surveillance Radioactivity Concen-15 Tech Spec limit to 1
Detection of breach tration or Radiation 100 times Tech Spec Level in Circulating limit R/hr Primary Coolant Accident Sampling and 10 uC1/gm to 10 Ci/gm 3
Detail analysis; Analysis of Primary or TID-14844 source term Accomplishment of Coolant in coolant volume mitigation; Gross Activity Verification;
- Cama Spectrum Long-term surveillance Reactor Coolant Pressure Souncary 4
l O to 3000 psig 1
Detection of potential RCS Pressure (4000 psig for CE for or actual breach; plants)
Accomplishment of mitigation; Long-term surveillance Containment Pressurel 10 psia to design 1
Detection of breach; 2
pressure psig Accomplishment of (5 psia for sub-atmos-mitigation; pheric containments)
Ve:1fication; Long-term surveillance 39
TABLE 2 (continued)
Category (see Variable Range Posttien C.1.3)
Purtose TYPE C - continued Reactor Coolant Pressure Boundary (continued)
Containment Sump Water Narrow range (. sump),
3 Detection of breach; Levell Wide range Gottom of 1
Accomplishment of containment to 600,000-mitigation; gallon level equivalent)
Verification; Long-term surveillance 7 11 Contain==nt. Area 1 to 10" R/hr 3
Detection of breach; i
k Radiation Verification Effluent Radioactivity 6 to 10-2 uCi/cc 3
Detection of breach; Noble Gas Effluent from Verification Condenser Air Removal l
System Exhaust i
Centainment RCS Pressural O to 3000 psig 1"
. Detection of potential (4000 psig for CE for breach; plants)
Accomplishment of mitigation Containment Hydrogen 0 to 10% (capable of oper-1 Detection of potencial Concentration ating from 10 psia to max-for breach; 2
imum design pressure )
Accomplishment of
- 3*'
0 to 30% for ice condenser Long-term..<e111ance e
,g l
Containment Pressure 10 psia pressure to 3 times 1 Detection of potencial 2
design pressure for concrete; for or actual breach; 4 times design pressure for Accomplishment of steel mitigation 40 n
TABLE 2 (continued) 1 Category (see
~
Va riable Range Position C.I.3)
Purcose TYPE C - continued Containment (continued)
I contain= ant Effluent 10-6 to 10-2 pCi/cc 2' "
Detection of breach; Radioactivity - Noble Accomplishment of Gases from Identified mitigacion; i
Release Pointal Verification Environs Radioactiv-10 to 10 R/hr 2
Detection of breach; 4
l Accomplishment of icy - Exposure Raca mitigation; Verificacion j
l 41
TABLE 2 (continu2d)
PWR VARIABLES (cuntinuad)
TYPE D Variables - Those variables that provide information to indicate the oper-ation of individual safety systems and other systems important to safety. These variables ara to help the operator make appropriate decisions in using the indi-vidual systems important to safety in mitigating the consequences of an accident.
Category (see Variable Range Position C.1.3)
Purcose TYPE D VARIABLES Residual Heat Removal or Decay Heat Removal System 3
2 Te monitor operation RHR System Flow 0 to 110% design flow RER Heat Exchanger 32*F to 350'F 2
To monitor soperation Out Temperature and for analysis Safety Infection Systems Accumulator Tank Level 10% to 90% volume 2
To monitor operation Level or Pressure O to 750 psig Accumulator Isola-Closed or Open 2
Operation status tion Valve Position Boric Acid Charging 0 to 110% design flow 3 3
To monitor operation Flow Flow in EFI System 0 to 110% design flow 3 2
To monitor operation Flow in LPI System 0 to 110% design flow 3 2
To monitor operation Refueling Water Top to bottom 2
To monitor operation Storage Tank Level Primary Coolant Systen Reactor Ceolant Pump Motor current 3
To monitor operation Status 42
TABLE 2 (continued)
Category (see Variable Range Position C.1.3)
Purcose TYPE D - continued Primary Coolant System -
(con tinued)
Primary System Safety Closed-not closed 2
Operation status; to Relief valve Positions monitor for loss of (including PORY and coolant code valves) or Flow Through or Pressure in Relief Valve Lines Pressurizer Level Bottom to top 1
To assure proper oper-ation of pressurtzer Pressurizer Heater Electric current 3
To determine operating S ta tus status Quench Tank Level Top to bottom 3
To monitor operation Quench Tank Temp-50*F to 750*F 3
To monitor operation erature Quench Tank Pressure O to design pressure 2 3
To monitor operation Secondary System (Steam Generator)
Steam Generator Level From tube sheet to 2 (Category 1 To monitor operation separators for 2-loop planrs}
Steam Generator From acnospheric pressure 2
To monitor operation Pressure to 20% above the lowest safety valve setting Safety / Relief Valve Closed - not closed 2
To monitor operation Positions or Main Stama Flow 3
Main Feedvater Flow 0 to 110% design flow 3
To monitor operation 43
TABLE 2 (continued)
Categorf (see Va riable Rance Position C.1.3)
Purcosa TYPE D - continued Auxiliary Feedwater or Emergency Feedwater System 3
Auxiliary or Emergency 0 to 110% design flow 2
To monitor operation Feedwater Flow
(.1 for B&W plants)
Condensate Storage Plant specific 1
To ensure water supply Tank Water Level for auxiliary feedwater (Can be Category 3 if not primary source of AFW. Then whatever is primary source of AT4 should be listed and should be Category 1)
Containment Cooling Systems 3
Containment Spray 0 to 110% design flov 2
To monitor operation Flow Eeat Removal By the Plant specific 2
To monitor operation Containment Fan Heat Re:noval System Containment Atmos-40*F to 400*F 3
To indicate accomplish-phare Temperature ment of cooling containment Sump 50*F to 250*F 2
To monitor operation Water Temperature 44 4
^
TABLE 2 (continued)
Category (see Variable Range Position C.1.3)
Pu rcose TYPE D - continued Chemical and Volume Control System 3
2 To monitor operation Makaup Flow - In 0 to 110% design flow 3
2 To monitor operation Letdown Flow - Out 0 to 110% design flow 7olume Control Top to bottom 2
To monitor operation Tank Level Cooling Water System Component Coe'.ing 32*F cc 200*F 2
To monitor operation Water Temperature to ESF System Components 3
2 To monitor operatica Component Cooling 0 to 110% design flow Water Flow to ESF System Components Radwaste Systems Eigh-Level Radioactive Top to bottom 3
To indicate Liquid Tank Level storage volume.
Radioactive Gas Hold-O to 150% design 3
To indicate 2
up Tank Pressure pressure storage capacity Ventilation Systens Emergency Vent 114-Open-closed status 2
To indicate damper tion Damper Position status Pcwer Supplies 13 Status of Standby Voltages, currents, 2
To indicate system Power & Otner Energy pressures status Sources Imporcant to Safety 45
TABLE 2 (continued)
PWR VARIABLES (continued)
TYPE E Variables - Those variables to be monitored as required for use in deternin-ing the magnitude of the release of radioactive materials and continually assessing such releases.
Categorf (see Variable Range Position C.1.3)
Purcose TYPE E VARIABLES Containment Radiation 7
Containment Area 1 R/hr to 10 R/hr 1
Detection of signif-l Radiation - Hi Range icant releases; Release assessment; Long-term surveillance; Emergency plan actuation Ar_ea Radiation 11 Radiation Exposure 10-1 R/hr to 10" R/hr 2
Detection of signif-Rate (Inside b1dgs or icant releases; areas where access is Release assessment; required to service Long-term surveillance equipment important to safety)
Airborne Radioactive Materials Released from the Plant Noble Gases and Vent Flow Rate s
10 o Containment or Purge 10-6 to lo uC1/cc 2
Detection of signif-Effluencl O to 110% vent desiga icant releases; flow 3
. Release assessment (Not needed if effluent dis-charges thru ccamon plant vent) 4 10 o Secondary Contain-10-6 to 10 uC1/cc 2
Detection of signif-ment (reactor shield 0 to 110% vent design icant releases; 3
bidg annulus, if in flow Release assessment design)
(Not needed if effluent dis-charges thru conunon plant vent) 4 10 o Auxiliary Suilding 10-i to 10 uCi/cc 2
Detection of signif-(including any b1dg 0 to 110% vent design icant releases; containing primary flow 3 b;1 ease assessment; system gases, e.g.,
(not needed if affluent dis-
.mg-term surveillance waste gas decay tank) charges thru common plant vent) 46
x.
TABLE 2 (continued)
Category (see Variable Rance Positien C.I.3)
~Purcose TYPE E - continued Airborne Radioactive Materials Release from the Plant (continued)
Noble Cases and Vent Flow Race (continued) 10 o Condensor Air Removal 10-6 to 105 uCi/cc 2
Detection of signif-l System Exhaust 0 to 110% vent design icant releases; flow 3 Release assessment (Not needed if offluent dis-charges thru common plant vent)
IO o Common Plant Vent or 10-6 to 103 uCi/cc 2
Detection of signif-Multi-purpose Vent 0 to 110% vent design icant releases; Discharging Any of flov3 Release assessment; the Above Releases Long-term surveillance o Vent From Steam Gen-10-1 to 10 uC1/cc 2
Detection of signif-3 erator Safer:y Relief (puration of releases in icant releases; Valves or Atmospheric. seconds, and mass of steam Release assessment Dump Valves per unit time) 10 o All Other Identified 10-6 en 10 uC1/cc 2
Detection of signif-Release Points 0 to 110% vent design icant releases; flow Release assessment; (Not needed if effluent dis-Long-term surveillance charges thru other monitored plant vents)
Parriculates and Halogens o All Identified Plant 10-3 to 102 uCi/cc 3
Detection of signif-Release Points (ex-O to 110% vent design icant releases; cape Stesa Generator flow 3 Release assessment; Safety Relief Valves
_Long-ters surveillance or Atmospheric Steam Dump Valves and Con-dessor Air Removal S stem Exhaust) f Sampling, With on-site Analysis Cap-j ability 47-
TABLE 2 (continu;d)
Category (see
'lariable Rance Position C.1.3)
Purpcse TYPE E - centinued Environs Radiation and Racicactivi ty 11 Radiation Exposure 10-6 R/hr to 10 R/hr 2
Detection of signif-l Rate icant releases; (Installed instrument-Verification; ation)
Release assessment; Long-ters surveillance Airborne Radiohalogens 10-9 to 10-3 uCi/cc 3
Release assessment; and Particulates Analysis (Sampling, with on-site analysis cap-ability) 16 Plant and Environs 0.1 to 10" R/hr, photons 3
Release assessment; 16 Radiation 0.1 to 10" rads /hr, beta-3 Analysis (Portable Instrument-radiations and low-energy ation) photons Plant and Environs Multi-channel Gamma-Ray 3
Releases assessment; Radioactivity spectrometer Analysis (Portable Instrument-ation) 48
~
c.
TABLE 2 (continued)
Category (see Variable Range Pesition C.1.3)
Purcose TYPE E - Continued METEOROLOGY 17
. Wind Directim 0 to 360* (25* accuracy 3
Release assessment with.a deflection of 15*.
Starting speed 0.45 mps (1.0 mph). Damping ratio between 0.4 and 0.6, dis-tance constant 52 meters.
Wind Speed 0 to 30 mps (67 aph) 10.22 3
Release assessment mps (0.5 mph) accuracy for wind speeds less than 11 mps (25 mph), with a start-ing threshold of less than 0.45 mps (1.0 mph).
Estimation of Atmos-Based on vertical temper-3 Releast assessment phric Stability ature difference from pri-mary system, -5'C to 10*C
(-9*7 to 18*F) and 20.15*c accuracy per 50 meter int-ervals ( 0.3*F accuracy per 164 foot intervals) or analogous range for back-up system.
49
TABLE 2 (continued)
Catecorj (see Variable Rance Position C.I.3)
Purcose TYPE E - (continued)
ACCIDENT SAMPLING CAP
- ABILITf (Analysis Cap-ability Onsite) 18 I' Primary Coolant & Sump Grab Sample 3
Release assessment; "I
o Gross Actisicy 10 pC1/ml to 10 C1/ml y,
o Gamma Spectrum (Isotopfe Analysis) o 3eron Content 0 to 6000 ppa o Chloride Content 0 to 20 ppm o Disolved Oxygen 0 to 20 ppm o pH 1 to 13 8
Containment Air Grab Sample 3
Release assessment; V**i I"**i "I o Hydrogen Content 0 to 10 78 O to 30% for ice condensors o Oxygen Concent 0 to 30%
o Gamma Spectrum (Noble gas analysis)
- The 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 the decision is Tade 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 />.
50
TABLE 2 (continutd)
NOTES-bhere a variable is listed for more than one purpose, the instrumentation requirements may be integrated and only one measurement provided.
2Design pressure is that value corresponding to ASME code values that are obtained at or below code-allowable material design stress values.
3Design flow is the =rinann flow anticipated in normal operation.
"The = rim'an value may be revised upward to satisfy ATWS requirements.
5A minimum of 4 measurements per quadrant is required for operation. Sufficient number should be installed to account for attrition.
6Measurement should be made of the gross gamma radiation emanating from circulating pri-mary coolant, with instrument calibration permitting conversion of readout to radioac-tivity concentrations in terms of either curies / gram or curies / unit-volume. System accuracy should be i order of magnitude. The point of measurement should be external to a circulating primary coolant line or loop, such as a hot leg, and should not be a line or loop subject to isolation, e.g.,
letdown line. While such an instrument may not be currently available off-the-shelf, the staff considers that the necessary com-ponents are available commercially and have been employed and demonstrated under ad-verse environmental conditions in high-level hot cell operations for many years.
bin 4- = of two monitors at widely separated locations.
3For estimating release races of radioactive materials released during an accident from unidentified release paths (,not coverad by effluent monitors) - continuous readout capability. (Approximately 16 to 20 locations - site dependent.)
9 Provisions should be made to monitor all identified pathways for release of gaseous radioactive materials to the environs in conformance with General Design Criterion 64.
Monitoring of individual effluent streams only is required where such streams are re-leased directly into the environment. If two or more streams are combined prior to release from a common discharge point, monitoring of the combined stream is considered I
to meet the intent of this guida provided such monitoring has a range adequate to measure worst-case releases.
MMonitors should be capable of detecting and measuring radioactive gaseous effluent con-centrations with compositions ranging from fresh equilibrium noble gas fission product mixtures to 10-day old mixtures, with overall system accuracies of decade. Effluent concentrations may be expressed in terms of Xe-133 equivalents or in terms of any noble gas nuclide(s). It is not expected that a single monitoring device will have sufficient range to encompass the entire range provided in this guide and that multiple components or syst,9 will be needed. Existing equipment may be utilized to monitor any portion of the stated range within the equipment design rating. Additional extended range instru-mentation should overlap the range of existing instrumentation by a least a factor of 2.
11 Detectors should respond to gasmaa radiation photons within any energy range from 60 kev to 3 MeV with an accuracy of t20" at any specific photon energy from 0.1 MeV to 3 MeV.
1 Overall system accuracy shodd ba within : decade over the entire range.
51 t
^
^
A*
~.
TABLE 2 (continued)
~
NOTES - continued 12 Effluent for PWR steam safety valve discharges and at:nospheric 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 % order of magnitude. Calibration sources should fall within the range of approximately 0.5 MeV to 1.5 MeV (examples: Cs-137, Mn-54, Na-2?, and Co-60).
Effluent concentrations should be expressed in terms of any gamma-emitting noble gas nuclide within the specified energy range. C41culational methods should be provided for est-imating concurrent releases of low-energy noble gases which cannot be detected or measured by the methads or techniques employed for monitoring.
13Status indication of all Standby Fewer A-C buses, D-C buses, inverter output buses and pneumatic supplies, lTo provide information : egarding release of radioactive halogens and particulates.
Continous collection of representative samples followed by onsite laboratory measure-ments of samples for radiohalogens and particulates. The design envelope for shield-ing, handling, and analytical purposes should assume 30 minutes of integrated sampling 2 uC1/cc of radiciodines time at sampler design flow, an average concentrations of 10 2
in gaseous or vapor form, an average concentration of 10 uCi/cc of particulate radio-iodines and particulates other than radiciodines, and an average gamma photon energy of 0.5 MeV per disintegration.
15For estimating release rates of radioactive materials released during an accident.from unidentified release paths (not covered by effluent monitors).
Continous collection of representative samples followed by laboratory measurements of the sanges.
i (Approximately 16~ to 20 locations - site dependent.)
16To monitor radiation and airborne radioactivity concentrations in many areas through-out the facility and the site environs where it is impractical to install stationary monitors capable of covering both normal and accident levels.
)
17 Meteorological measurements should conform to the provisions of the forthcoming rev-i ision of Regulatory Guide 1.23, "Onsite Meteorological Programs".
13 Sampling or monitoring of radioactive liquids and gases should be performed in a manner which assures procurement of representative samples. For gases, the criteria of ANSI N13.1 should be applied. For liquids, provisions should be made for sampling from well-mixed turbulent zones and sampling lines should be designed ;;o minimize plateout or deposition. For safe and convenient sampling, the provisions should include:
a.
Shielding to maintain radiation deses ALARA, b.
Sample containers with container-eampling port connector compacability, c.
Capability of sampling under primary system pressure and negative pressures, d.
Handling and transport capability, and 4
e.
Pre-arrangement for analysis and interpretation.
19 n installed capability should be provided for obtaining containment sump, ECCS pump A
room sumps, and other sf=flmr auxiliary building sump liquid samples.
52
.__.s
g, e.
A.
VALUE/ IMPACT STATEMENT 1.
PROPOSED ACTION 1.1 Description The applicant (licensee) of a nuclear power plant is required by the Com-mission's regulations to provide instrumentation to (1) monitor variables and systems over their anticipated ranges for accident conditions as appropriate to ensure adequate safety and (2) monitor the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluid, effluent discharge paths, and the plant environs for radioactivity that may be released frca postulated accidents.
This revision to Regulatory Guide 1.97 proposes to improve the guidance for plant and environs monitoring during and following an accident, including extended ranges for some instruments to account for consideration of degraded core cooling.
- 1. 2 Need
~
Regulatory Guide 1.97 was issued as an effective guide in August 1977. At the time the guide was issued, it was recognized that more specific guidance than that contained in the guide would be required. However, the difficulty in developing the guide to the point where it could be initially issued was evidence that experience in using the guide as it then existed was essential before further development of the guide would be meaningful.
Therefore, in August 1977, the staff initiated Task Action Plan A-34,
" Instruments for Monitoring Radiation and Process Variables During an Accident."
The purpose of the task action plan was to develop guidance for applicants, licensees, and staff reviewers concerning implementation of Regulatory Guide 1.97.
Such effort would provide a basis for revising the guide.
When the staff was ready to issue the results of the Task Action Plan A-34 effort, the accident at TMI-2 occurred. Subsequently, the TMI-2 Lessons Learned
-Task Force has issued its " Status Report and Short-Term Recommendations," NUREG-0578.
i l
l 53 l
e.
~.
This report, along with the draft Task Action Plan A-34 report; Draft 1 of Regula-tory Guide 1.97, dated April 12, 1974; and Standard ANS-4.5, provides ample basis for revising Regulatory Guide 1.97.
1.3 Value/ Impact of the Procosed Action t
l 1.3.1 NRC Ooerations Since a list of selected variables to be provided with instrumentation to be monitored by the plant operator during and following an accident has not.
been explicitly agreed to in the past, the proposed action shoulif result in more effective effort by the staff in reviewing applications for construction permits and operating licenses. The proposed action will establish an NRC position by taking advantage of previous staff effort (1) in completion of a generic activity (A-34), (2) in evaluating the lessons learned from the TMI-2 event (NUREG-0578), and (3) in conjunction with effort in developing a nicional standard (ANS-4.5). For future plants, the staff review will be siIntified with guidance contained in the endorsed standard developed by a voluntary standards.,
1 group and the regulatory guide, which includes a list of varicles for accident' monitoring.
Efforts by the staff to implement Revision 1 to Regulatory Guide 1.97 has been fraYght with frustration and met ~with delays because the guide was adjudged by licensees to be vague and ambiguous.
Revision 2 eliminates the problems encountered with Revision 1 because it provides a minimum set of vari-ables to be measured and hence gives more guidance in the selection of accident monitoring instrumentation.
Consequently, there will be no significant impact on the staff. There will, however, be effort required to review each operating plant and plants under construction to assure compliance with Regulatory Guide 1.97.
1.3.2 Other Government Agencies Not applicable, unless the government agency is an applicant.
1.3.3 Industry I
The proposed action establishes a more clearly defined NRC position with I
j regard to instrumentation to assess plant and environs conditions during and 54
-y,-
o a
l
~
l following an accident and, therefore, reduces uncertainty as to what the staff
~;
, considers acceptable in the area,of accide,nt monitoring. Most of the impact on industry will be in the area of providing instrumentation to indicate the poten-tial breach and the actual breach of the barriers to radioactivity release, i.e.,
fuel cladding, reactor coolant pressure boundary, and containment.
These instru-ments have extended ranges and there are others with qualification requirements not previously imposed. There will be additional impact due to a heretofore unspecified variables to be monitored (i.e., water level in reactor for FWRs and radiation level in the primary coolant water for PWRs and SWRs) that have been identified during the evaluation of TMI-2 experience and will require development.
Attempts were made during the comment period to determine the cost impact on industry for future plants and for backfitting existing plants.
Estimates ranged from $4,000,000 to over $20,000,000. The higher estimates undoubtedly charged all accident monitoring instrumentation to Revision 2 to Regulatory Guide 1.97,which should not be the case. The requirement for accident monitoring has always been a part of the regulations.
Consequently the impact of Revision 2 to Regulatory Guide 1.97 should only be the delta added by Revision 2.
A conser-vative estimate of the increase in r'equirements are the additions of Type C measurements and the upgrading of some of the Type 8 measurements to higher qualification of the instrumentation. There are 17 unique Type B & C variables to be measured for PWRs, less for BWRs. A conservative average cost for each measurement is $130,000 making a total cost impact of $2,210,000.
If this figure were doubled to account for overhead costs and about a 151 contingency added, the cost impact would be about $5,000,000.
This cost estimate is the same for operating plants as for plants under construction and future plants.
While it is recognized that for operating plants the costs associated with backfitting are generally higher than the costs associated with new plants, there are some concessions made in some of the requirements due to existing licensing committments which brings the cost estimate to about the same value.
1.3.4 Public The proposed action will impreve public safety by ensuring that the plant operator will have timely information to take any necessary action to protect the public.
55 i
4 --
-r
u.
No impact on the public can be foreseen.
~
- 1. 4 Decision on Procosed Action As previously stated, more definitive guidance on instrumentation to assess plant and environs conditions during and following an accident should be given.
2.
TECHNICAL APPROACH This section is not applicable to this value/ impact statement since the proposed action is a revision of an existing regulatory guide, and there are no alternatives to provi' ding the plant operator with the required information.
3.
PROCEDURAL APPROACH Previously discussed.
4.
STATUTORY CONSIDERATIONS x
4.1 NRC Authority nuthority for this gu,ide,would be derived from the safety requirements of the Atomic Energy Act through the Commission's regulations, in particular, Criterion 13, Criterion 19, and Criterion 64 of Appendix A to 10 CFR Part 50, which require, in part, that instrumentation be provided to monitor variables, systems, and plant environs to ensure adequate safety.
4.2 Need for NEPA Assessment The preposed action is not a major action as defined in paragraph 51.5(a)(10) of 10 CFR Part 51 and does not require an environmental impact statement.
5.
RELATIONSHIP TO OTHER EXISTING OR PROF 0 SED REGULATIONS OR POLICIES No conflicts or overlaps with requirements prcmulgated by other agencies are foreseen.
This guide does include the variables to be monitored on site by 56
3 2-e the plant operator in order to provide necessary information for emergency planning.
However, emergency planning and its relationship to other agencies is provided by other means.
Implementation of the proposed action is discussed in Section 0 af the proposed revision.
J 6.
SLWtARY AND CONCLUSIONS The revision to Regulatory Guide 1.97, " Instrumentation For Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," should be issued and implemented according to existing schedules.
e 0
4 57
~
--.,1 a.
s.
VALUE/ IMPACT STATEMENT 1
1.
PROPOSED ACTION 1.1 Description The applicant (licensee) of a nuclear power plant is required by the Com-mission's regulations to provide instrumentation to (1) monitor variables and systems over their anticipated ranges for accident conditions as appropriate to ensure adequate safety and (2) monitor the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluid, effluent discharge paths, and the plant environs for radioactivity that may be released from postulated accidents.
This revision to Regulatory Guide 1.97 proposes to improve the guidance for plant and environs monitoring during and following an accident, including extended ranges for some instruments to account for consideration of degraded core cooling.
- 1. 2 Need Regulatory Guide 1.97 was issued as an effective guide in August 1977. At the time the guide was issued, it was recognized that more specific guidance than that contained in the guide would be required. However, the difficulty in daveloping the guide to the point where it could be initially issued was evidence that experience in using the guide as it then existed was essential before further development of the guide would be meaningful.
Therefore, in August 1977, the staff initiated Task Action Plan A-34,
" Instruments for Monitoring Radiation and Process Variables During an Accident."
The purpose of the task action plan was to develop guidance for applicants, licensees, and staff reviewers concerning implementation of Regulatory Guide 1.97.
Such effort would provide a basis for revising the guide.
When the staff was ready to issue the results of the Task Action Plan A-34 effort, the accident at TMI-2 occurred.
Subsequently, the TMI-2 Lessons Learned Task Force has issued its " Status Report and Short-Term Recommendations," NUREG-0578.
53
w This report, along with the draft Task Action Plan A-34 report; Oraft 1 of Regula-tory Guide 1.97, dated April 12, 1974; and Standard ANS-4.5, provides ample basis for revising Regulatory Guide 1.97.
- 1. 3 Value/ Impact of the Proposed Action 1.3.1 NRC Operations Since a list of selected variables to be provided with instrumentation to be monitored by the plant operator during and following an accident has not.
been explicitly agreed to in the past, the proposed action should result in more effective effort by the staif in reviewing applications for construction permits and operating licenses. The proposed action will establish an NRC position by taking advantage of previous staff effort (1) in completion of a generic activity (A-34), (2) in evaluating the lessons learned from the TMI-2 event (NUREG-0578), and (3) in conjunction with effort in developing a national standard (ANS-4.5).
For future plants, the staff review will be simplified with guidance contained in the endorsed standard developed by a voluntary staadards,,
group and the regulatory guide, which includes a list of variables for accident' monitoring.
Efforts by the staff to implement Revision 1 to Regulatory Gaide 1.97 has been fraught with frustration and met"with delays because the guide was adjudged by licensees to be vague and ambiguous.
Revision 2 eliminates the problems encountered with Revision 1 because it provides a minimum set of vari-ables to be measured and hence gives more guidance in the selection of accident monitoring instrumentation.
Consequently, there will be no significant impact on the staff.
There will, however, be effort required to review each operating plant and plants under construction to assure compliance with Regulatory Guide 1.97.
1.3.2 Other Government Agencies Not applicable, unless the government agency is an applicant.
1.3.3 Industry The proposed action establishes a more cles.rly defined NRC position with regard to instrumentation to assess plant and environs conditions during and 54
L
+
following an accident and, therefore, reduces uncertainty as to what the staff
~
considers acceptable in the area;of accident monitoring. Most of the impact on industry will be in the area of providing instrumentation to indicata the poten-tial breach and the actual breach of the barriers to radioactivity release, i.e.,
fuel cladding, reactor coolant pressure boundary, and containment. These instru-ments have extended ranges and there are others with qualification requirements not previously imposed. There will be additional impact due to a heretofore unspecified variables to be monitored (i.e., water level in reactor for PWRs and radiation level in the primary coolant water for PWRs and BWRs) that have been identified during the evaluation of TMI-2 experience and will require development.
Attempts were made during the comment period to determine the cost impact on industry for future plants and for backfitting existing plants.
Estimates ranged from $4,000,000 to over $20,000,000. The higher estimates undoubtedly charged all accident monitoring instrumentation to Revision 2 to Regulatory Guide 1.97,which should not be the case. The requirement for accident monitoring has always been a part of the regulations. Consequently the impact of Revision 2 4
to Regulatory Guide 1.97 should only be tne delta added by Revision 2.
A conser-vative estimate of the' increase in requirements are the additions of Type C measurements and the upgrading of some of the Type 8 measurements to higher qualification of the instrumentation. There are 17 unique Type B & C variables to be measured for PWRs, less for BWRs. A conservative average cost for each measurement is $130,000 making a total cost impact of $2,210,000.
If this figure were doubled to account for overhead costs and about a 15% contingency added, the cost impact would be about $5,000,000. This cost estimata is the same for operating plants as for plants under construction and future plants.
While it is recognized that for operating plants the costs associated with backfitting are generally higher than the costs associated with new plants, there are scme concessions made in some of the requirements due to existing licensing committments which brings the cost estimate to about the same value.
1.3.4 Public The proposed action will improve public safety by ensuring that the plant operator will have timely information to take any necessary action to. protect the public.
1 55
O
~
d.
No impact on the public can be foreseen.
~
- 1. 4 Decision on Procosed Action As previously stated, more definitive guidance on instrumentation to assess plant and environs conditions during and following an accident should be given.
2.
TECHNICAL APPROACH This section is not applicable to this value/ impact statement since the proposed action is a revision of an existing regulatory guide, and there are no alternatives to providing the plant operator with the required information.
3.
PROCEDURAL APPROACH Previously discussed.
4.
STATUTORY CONSIDERATIONS
~
4.1 NRC Authority Authority for this guide would be derived from the safety requirements of the Atomic Energy Act through the Commission's regulations, in particular, Criterion 13, Criterion 19, and Criterion 64 of Appendix A to 10 CFR Part 50, which require, in part, that instrumentation be provided to monitor variables, systems, and plant environs to ensure adequate safety.
- 4. 2 Need for NEPA Assessment The proposed action is not a major action as defined in paragraph 51.5(a)(10) of 10 CFR Part 51 and does not require an environmental impact statement.
5.
RELATIONSHIP TO OTHER EXISTING OR PROPOSED REGULATIONS OR POLICIES No conflicts or overlaps with requirements promulgated by other agencies are foreseen. This guide does' include thi variables to be monitored on site by 56
g g _* -
.w.
o the plant operator in order to provide necessary information for emergency planning.
However, emergency planning and its relationship to other agencies is provided by other means.
Implementation of the proposed action is discussed in Section D of the proposed revision.
J 6.
SUMMARY
AND CONCLUSIONS The revision to Regulatory Guide 1.97, "Instrumenta+ ion For Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," should be issued and implemented according to existing schedules.
57