ML19343A621

Modified Draft 3 of Revision 2 to Reg Guide 1.97, Instrumentation for Light-Water-Cooled Nuclear Plants to Assess Plant & Environs Conditions During & Following Accident

ML19343A621
Person / Time
Issue date:	11/06/1980
From:	Hintze A NRC OFFICE OF STANDARDS DEVELOPMENT
To:
References
TASK-OS, TASK-RS-917-4 REGGD-01.097, REGGD-1.097, NUDOCS 8011190199
Download: ML19343A621 (58)

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A. S. Hintze, (301) 443-5913 6 [PROPOSEB] RE ] N 2 TO REGULATORY GUIDE 1.97 / 7 INSTRUMENTATION FOR LI -WATER-COOLED NUCLEAR POWER PLANTS 8' TO ASSESS PLANT AND ENVIRONS 0 ITIONS DURING AND FOLLOWING AN ACCIDENT / 9 A. . NTRO -TION 10 Criterion 13, " Instrument tion and Contro17 of Appendix A, " General Design 11 Criteria for Nuclear Pcwer Plan ,", to 10 CFR Part 50, " Domestic Licensing of 12 Production and Utilization Facil s"includesarequirementthatinstrumen-13 tation be provided to monitor va ab\\esan shstemsovertheiranticipatedranges for accident conditions as apr$ophiate o en,h re adequate safety. 14 15 Criterion 19, " Control Ri c of Appe ,A to 10 CFR Part 50 includes a 16 requirement that a control roc e provide from which actions can be taken to 17 maintain the nuclear power unit i a safe condition under accident conditions, 18 including loss-of-coc. taccidepts,andtflatequipment,includingthenecessary instrumentation,athpropiate ocationsfoutsidethecontrolroombeprovided 19 20 with a design capabili for

pt hot' shutdown of the reactor.

r 21 Criterion 64, "Moni ring Radioacti ity Releases," of Appendix A to 10 CFR 22 Part 5 ncludes a quirer. nt that mears be provided for monitoring the reactor 23 cor nment atmosphere, cpac containi g components for recirculation of loss-24 of to lant accident fluid, ef uent discharge paths, and the plant environs 'vityth$,may rleasedffrompostulatedaccidents. i 25 f r radioa 26 This guid desf'ibeQ3 hcd acceptable to the NRC staff for complying 27 wit the Coatnissi 's regula ans to rovide instrumentation to monitor plant 4 28 variabTt+ a oysten, durir and foll wing an accident in,a light-water cooled 29 nuclear power plant. 60111 D 0@ 1 MGGb

e 1 B. 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-permit the operator to take preplanned manual actions to accomplish safe plant 5 shutdown; (2) determine whether the reactor trip, engineered-safety-feature G systems, and manually initiated safety systems and other syste'ms important to 7 safety are performing their intended functions (i.e., reactivity control, core 8 cooling, maintaining reactor coolant system integrity, and maintaining contain-9 ment integrity); and (3) provide information to the operator that will enable 10 him to determine the potential for causing a gross breach ';f 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 systems and other systems important to safety are required 15 by the control room operating personnel during an accident to (1) furnish data 16 regarding the operation of plant systems in order that the operator can make 17 appropriate decisions as to their use; and (2) provide information regarding the 18 release of radioactive materials to allow for early indication of the need to 19 initiate action necessary to protect the public and for an estimate of the 20 magnitude of any impending threat. 21 At the start of an accident, it may be di.f ficult for the operator to deter-22 mine immediately what accident has occurred or is occurring and, therefore, to 23 determine the appropriate response. For this reason, reactor trip and certain 24 other safety actions (e.g., emergency core cooling actuation, containment isola-25 tion, or depressurization) have been designed to be performed automatically 26 during the initial stages of an accident. Instrumentation is also provided to 27 indicate information about plant variables required to enable the operation of 28 manually initiated safety systems and other appropriate operator actions involving 29 systems important to safety. 30 [ instrumentation-iswaiso-needed-to provide-information-about-some plant-31 parameters-that-is-currently not-availabic-using presert-technology wiii aiert-32 the-operat--to conditions-that-have-degraded-boyend-those postulated-in-the-33 accident- < sis---In particular--it-is-important-that-the-operator-be-informed 2

i o 1 regarding that-status-of-coolant-level-in-the-reactor vessei or-the-existenae 2 of-core-voiding-thus providing-indication of potential-degraded-core cooling 3 and-imminent-fuei-damage --Birect-indication of-cooiant-level-in-the reactor 4 vessel-is not currently-availabie-in pressurized-water-reactors:--Hewever--it-is 5 imperative-that-this capability-be-feveloped-within-a-reasonabic-time-in-order 6 to provide-the-operator with-this-vital-information-in-a positive--unambiguous 7 manner:] 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 t!.at 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 intend'ed 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, anticipated 22 operational occurrences which become accidents such as anticipated transients 23 without scram (ATWS), reactivity excursions which result 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 i.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 local protected environment. 32 It is important that accident monitoring instrumentation components and 33 their mounts that cannot be located in Seismic Category I buildings be designed 34 to continue to function, to the extent feasible, during seismic events. Con-35 sequently, it it is essential that they be designed to resist the effects of 3

1 seismic excitation. An acceptable method for demonstrating the adequacy of 2 the seimsic resistance of this instrumentation would be to qualify it to meet 3 the seismic criteria applicable to instrumentat. ion instal}ed at other locati/ns 4 in the plant. 5 Variables selected for accident monitoring 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 monitored 11 by instruments qualified to more stringent environmental requirements and with I 12 ranges that extend well beyond that which the selected variables can attain under limiting conditions; for exampic, a range for the containment pressure 13 14 monitor 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. 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-historyresponsesformanyvariabiesimportanttofollowingthe 24 course of the accident. Therefore, it is prudent to select the required 25 accident-monitoring instrumentation from the normal power plant instrumentation 26 to enable the operator to use, during accident situations, instruments with 27 which he is 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 monitored 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 performance requirements in each 36 range. 4

1 Standard ANS-4.5,* " Criteria for Accident monitoring Functions in a Light-2 Water-Cooled Nuclear Power Generating Station," dated 1980, delineates 3 criteria for determining the variables to be monitored by the controi room 4 operator, as required for safety, during the course of an accident and during 5 the long-term stable shutdown phase folicwng an accident. Standard ANS-4.5 6 was prepared by Working Group 4.5 of Subcommittee ANS-4 with two primary 7 objectives: (1) to address that instrumentation that permits the operator to 8 monitor expected parameter changes in an accident period and (2) to address 9 extended range instrumentation deemed appropriate for the possibility of 10 encountering previously unforeseen events. ANS-4.5 references a revision to 11 IEEE Std 497 as the source for specific instrumentation design criteria. Since 12 the revision to IEEE Std 497 has not yet been completed, its applicability cannot 13 yet be determined. Hence, specific instrumentation design criteria have been 14 included in this regulatory guide. 15 The ANS standard defines three variable types (definitions modified herein) 16 for the purpose of aiding the designer in his selection of accident-monitoring 17 instrumentation and applicable criteria. The types are: Type A - those variables 18 that provide primary ** information needed to permit the control room operating 19 personnel to take the specified manually controlled actions for which no automatic 20 control is provided and which are required for safety systems'to accomplish 21 their safety functions for design basis accident events. Type B - those variables 22 that provide information to indicate whether plant safety functions are being 23 accomplished, and Type C - those variables that provide information 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 26 containment (modified to reflect NRC staff position; see Position C.1.2). The 27 sources of potential breach are limited '.o the energy sources within the barrier 28 29 ^ Copies may be obtained from the American Nuclear Society, 555 North Kensington 30 Avenue, LaGrange Park, Illinois 60525. 31-Primary information 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. 5 y v m~-

1-itself. In addition to the accident monitoring variables provided in ANS-4.5 2 standard, variables for monitoring 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 D - 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 to be 7 nonitored as required for use in determining the magnitude of the release of 8 radioactive materials and for continuously assessing such releases, 9 A minimum set of Types B, C, D, and E variables to be measured is listed 10 in this regulatory guide. Type A variables have not been listed because they 11 are plant specific and will depend on the operations that the designer chooses 12 for planned manual action. Types B, C, D, 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 measures in operation, (b) to inform the operator of 15 the necessity for unplanned actions to mitigate the consequences of an accident. 16 The five classifications are not mutually exclusive in that a given variable 17 (or instrument) may be applicable to one or more types, as well as for normal 18 power plant operation or for automatically initiated safety actions. A variable 19 iv:luded as Type B, C, D, or E does not preclude that variable from being 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 ir.strumentation survive the accident environment and function 25 as long as the information it provides is needed by the control room operating 26 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 measure the various 29 variables listed in Table 1 (for BWR) and Table 2 (for PWR). The criteria are 30 separated into three separate groups or cat'egories which provide a graded 31 approach to requirements depending on the importance to safety of a variable 32 being measured. 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 36 assure that high quality off-the-shelf instrumentation is obtained and applies 6

I 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 measurement of a single key variable nay 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, 6 it is essential that they each be considered key variables and measured with 7 high quality instrumentation. Additionally, it'is prudent, in some instances, 8 to include the measurement of additional variables for backup information and 9 for diagnosis. Where these additional measurements are included, the measures 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 number of variables) that 13 most 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 D) or 15 radioactive materials release (in the case of Type E). It is essential that 16 key variables be qualified to the more stringent design and qualification 17 criteria. The design and qualification criteria 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 D and E, the key variables are generally Category 2, backup 22 variables are Category 3. 23 The variables are listed but no mention (beyond redundancy requirements) 24 is made of the number of points of measurement of each variable. It is important 25 that the number of points of measurement be sufficient to adequately indicate 26 the variable value, e.g., containment temperature may require spatial location 27 of several points of measurement. 28 This guide provides the minimum variables to be monitored by the control 29 room operating personnel during and following an accident. These variables 30 are used by the control room operating personnel 1 rerform 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 manned. Variables are also defined to permit the operator to 34 perform his long-term monitoring and execution responsibilities after the 35 emergency response facilities are manned. The application of the criteria for 7

1 the instrumentation is limited to that part of the instrumentation system and 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 systems provided as operator aids for the purpose 5-of enhancement of information presentations for-the identification or diagnosis 6 of disturbances. 7 C. REGULATORY POSITION 8 1. ACCIDE'NT MONITORING INSTRUMENTATION i 9 The criteria, and requirements, contained in Standard ANS-4.5," Criteria 10 for Accident Monitoring Functions in a Light-Water-Cooled Nuclear Power 11 Generating Station," dated 1980, 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 15 be modified to be as follows: Type A - those variables to be monitored that 16 provide the primary information required to permit the control room operator 4 17 to take the specified manually controlled actions for which no automatic control 18 is provided and which are required for safety' systems.to accomplish their safety 19 function for design basis accident events. (Note: 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 coritingency actions 22 that may also be identified in vritten procedures.) 23 1.2 In Section 3.2.3 of ANS-4.5, the definition of " Type C" includes two 24 items, (1) and (2). Item (1) includes those instruments that indicate the extent 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 j 28 reactor containment, the parameters that indicate the potential for causing a 29 breach in the fuel cladding (e.g., core exit temperature) and the reactor coolant l [ 8 1 .~.

1-pressure boundary (e.g., reactor coolant pressure) should also be included. t 2 References to Type C instruments, and associated parameters to be measured, in Stan'ard ANS-4.5 (e.g., Sections 4.2,5.0,5.1.3,5.2,6.0,6.3) should include 3 d 4 this expanded definition. 5 1.3 Section 6.1 of ANS-4.5 pertains to General Criteria for Types A, B, 6 and C accident monitoring variables. In lieu of Section 6.1, the following 7 design and qualification criteria categories should be used: 8 1.3.1 Design and Qualification Criteria - Category 1 9 (1) The instrumentation should be qualified in accordance with 10 Regulatory Guide 1.89 & NUREG 0588. Qualification applies to the complete 11 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 computer-based display, recording and/or diagnostic program, 14 qualification appli'es to and including the channel isolation device. The 15 location of the isolation device should be such that it would be accessible 16 for maintenance during accident conditions. The seismic portion of qualification 17 should be in accordance with Regulatory Guide 1.100. Instrumentation should 18 continue to read within the required accuracy following, but not necessarily 19 during, a safe shutdown earthquake. Instrumentation, whose ranges are required } 20 to extend beyond those ranges calculated in the most 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. l 23 (2) No single failure within either the accident monitoring instrumenta-24 tion, its auxiliary supporting features or' its power sources concurrent with 25 the failures that are a condition or result of a specific accident, should prevent l l 26 the operator from being presented the information necessary for him to determining ) 27 the safety status of the plant and to brino 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 safety 31 function, additional information should be provided to allow the operator to 9 i -l -,-.. a

1 deduce the actual conditions in the plant. This may be accomplished by providing 2 additional independent channels of information of the same variable (addition of 3 an identical channel), or by providing an independent channel which monitors a 4 _different variable which bears a known relationship to the multiple channels 5-(addition of a diverse channel), or by providing the capability, if sufficient 6 time is available, for the operator to perturb the measured variable and deter-7 mine 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 physically 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 11 direct-indicating or recording device. (NOTE: Within each redundant division 12 of a safety system, redundant monitoring channels are not needed except for steam generator level instrumentation in two-loop plants _._) 13 (3) The instrumentation should be energized from station Standby 14 Power sources as provided in Regulatory Guide 1.32, battery backed where momentary interruption is not tolerable. 15 (4) The instrumentation channel should be available prior to an 16 accident except as provided in Paragraph 4.11, " Exemption", as defined in IEEE 17 Std 279 or as specified in Technical Specifications. l a 18 (5) The recommendations of the following regulatory guides j 19 pertaining to qt,ality assurance should be followed: 1 20 Regulatory Guide 1.28 " Quality Assurance Program Requirements (Design 21 & Construction)" 22 Regulatory Guide 1.30 " Quality Assurance Requirements for the Installation, 23 Inspection, and Testing of Instrumentation and 24 Electric Equipment" 25 Regulatory Guide 1.38 " Quality Assurance Requirements for Packaging, 26 Shipping, Receiving, Storage, and Handling of 27 Items for Wat'er-Cooled Nuclear Power Plants" 28 Regulatory Guide 1.58 " Qualification of Nuclear Power Plant Insptection, 29 Examination, and Testing Personnel" 30 Regulatory Guide 1.64 " Quality Assurance Requirements for the Design 31 of Nuclear Power Plants" 10

1 Regulatory Guide 1.74 " Quality Assurance Terms and Definitions" 2 Regulatory Guide 1.88 " Collection, Storage, and Maintenance of Nuclear 3 , Power Plant Quality Assurance Records".. 4 Regulatory Guide 1.123 " Quality Assurance Requirements for Control of 5 Procurement of Items and Services for Nuclear 6 Power Plants" 7 Regulatory Guide 1.144 " Auditing of Quality Assurance Programs for' Nuclear 8 Power Plants" 9 Regulatory Guide 1.146 " Qualification of Quality Assurance Program Audit 10 Personnel for Nuclear Power Plants" (Guide number 11 to be inserted.) 12 Reference to the above regulatory guides (except Regulatorf 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 spans 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 may 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 to be 24 lost by such devices. 25 1.3.2 Design and Qualification Criteria - Category 2 26 (1) The instrumentation.should be qualified in accordance with Regula-27 tory Guide 1.89 & NUREG 0588. Where the channel signal is to be processed or 28 displayed on demand, qualification applies from the sensor through the isolator / 29 input buffer. The location of the isolation device should be such that it would 30 be accessible for maintenance during accident conditions. 11

~ a 1 (2) The instrumentation should be energized from a high reliability 2 power source, not necessarily Standby Power, battery 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 6 applicable or where specified by other requirements. 7 (4) The recommendations of the following regulatory guides 8 pertaining to quality assurance should be followed: 9 Regulatory Guide 1.28 " Quality Assurance Program Requirements (Design 10 & Construction)" 11 Regulatory Guide 1.30 " 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, R6ceiving, Storage, and Handling of 16 Items for Water-Cooled Nuclear Power Plants" 17 Regulatory Guide 1.58 " Qualification of Nuclear Power Plant Insptection, 18 Examination, and Testing Personnel" 19 Regulatory Guide 1.64 " Quality Assurance Requirements for the Design 20 of Nuclear Power Plants" 21 Regulatory Guide 1.74 " Quality Assurance Terms and Definitions" 22 Regulatory Guide 1.88 " Collection, Storage,andIdaintenanceofNuclear 23 Powee Plant Quality Assurance Records" 24 Regulatory Guide 1.123 " Quality Assurance Requirements for Control of 25 Procurement.of Items and Services for Nuclear 26 Power Plants" 27 Regulatory Guide 1.144 " Auditing of Quality Assurance Programs for Nuclear 28 Power Plants" 29 Regulatory Guide 1.146 " Qualification of Quality Assurance Program Audit 30 Personnel for Nuclear Power Plant.s" (Guide number 31 to be-inserted.) l d' F 12 1

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 may not be necessary to apply 5-the same quality assurance measures 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 importance to safety of the 8 instrumentation. These requirements should be determined and documented by 9 personnel knowledgeable in the end use of the instrumentation. 10 (5) The instrumer.l.ation signal may be displayed on an individual 11 instrument or it may 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 monitors, environs exposure rate monitors, and meteorology 16 monitors. Where direct and immediate trend or transient information is essential 17 for operator information or action, the recording should be analog stripchart. 18 Otherwise, it may be continuously updated, computer memory stored, and displayed 19 on demand. 20 1.3.3 Design and Qualification Criteria - Category 3 21 (1) High quality commercial 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 Cate0ories 1 and 2: 25 1.4.1 Any equipment that is used for either Category 1 or Category 2 26 should be designcted as part of accident monitoring or systems operation 27 and effluent monitoring instrumentation. The transmission of signals from 28 .uch equipment for other use should be through isolation devices that are 13

1 designated as part of monitoring instrumentation and that meet the provisions 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 casily discern that they are intended for use under accident conditions. t 6

1. 5~

In addition to the above criteria, the following should apply to Categories 7 1, 2 and 3. l 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 monitored 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. j 17 1.5.2 Servicing, testing, and calibration programs should be specified I 18 to maintain the capability of the monitoring instrumentation. For those 19 instruments where the required interval between testing will be less than the i 20 normal time interval between generating station shutdowns, a capability for 21 testing during power operation should be provided. 4 22 1.5.3 Whenever means for removing channels from service are included in 23 the design, the design siiculd facilitate administrative control of the access 24 to such removal means. i ~ 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

e 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 bo oesigned to facilitate the recogni-5 tion, location, replacement, repair, or ad ustment of malfunctioning components J 6 or modules. 7 1.5.7 To the extent [practica)) possible, monitoring instrumentation inputs 8 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 us are used for the normal operations of the plant to enable 11 the operator to use, during accident situr'!ans, instruments with which he is 12 most familiar. However, where ti.e ~ rr 'uired range of monitoring instrumentation 13 results in a loss of instrumentation sensitivity in the normal operating range, 14 separate instruments should be used. 15 .1. 5. 9 Periodic testing should be in accordance with the applicable portions 16 of Regulatory Guide 1.118 pertaining to testing of 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 4 20 and C variables. In conjunction with the above sections, Tables 1, and 2 of 21 this regulatory guide (which include those yariables 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. 24 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 l 27 the operation of individual safety systems and other systems important to safety. 15

1 2.1.2 Type E - those variables to be monitored as required for use in 2 determining the magnitude of the release of radioactive materials and continually 3 assessing such releases. 4 2.2 The plant designer should select variables and information display j 5 channels required by his design to enable the control room operating personnel j 6 to: 1 i 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 9 each system is operating or can be placed in operation to help mitigate the 10 consequences of an accident. 11 2.2.2 Monitor the effluent discharge paths and environs within the 12 site boundary to ascertain if there have been significant releases (planned or l 13 unplanned) of radioactive materials and for continually assessing such releases. 14 2.2.3 Obtain required information through a backup or diagnosis 15 channel where a single channel may be likely to give anbiguous indication. i 16 2.3 The process for selecting system operation and effluent release 17 variables should include the identification of: 4 i 18 2.3.1 For Type D t 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 mitigate 21 the consequences of an accident; i 22 (2) the variable or minimum l'ist' of variables that indicate the 23 operating status of each system identified in (1) above.. 24 2.3.3 For Type E i 25 (1) the planned paths for offluent release; 16

e i 1 (2) plant areas and inside buildings where access is required to ] 2 service equipment necessary to mitigate the consequences of an accidant; 3 (3) onsite locations where unplanned releases of radioactive 4 materials should be detected; 5 (4) the variables that should be monitored in each location 6 identified in (1), (2), and (3) above. 7 2.4 The determination of performance requirements for syste:a operation 8 monitoring and effluent release monitoring information display channels should 9 include, as a minimum, identification of: l 10 (1) the range of the process variable. 11 (2).the required accuracy of measurement. 12 (3) the required response characteristics. 13 (4) the time interval during which the measurement is needed. j 14 (5) the local environment (s) in which the information display j 15 channel components must operate. 16 (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. 19 2.5 The design and qualification criteria for system operation monitoring 20 and effluent release monitoring instrumentation should be taken from the criteria 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 used as a minimum set of instruments and 23 their respectives ranges for systems operation monitoring (Type D) and effluent 24 release monitoring (Type E) instrumentation for each nuclear power plant. 25 D. IMPLEMENTATION i 26 All plants going into operation after June 1982 should meet the provisions 27 of this guide. 17 v, , -. _. _. - ~

1 Plants currently openting or scheduled to be licensed to operate before 2 June 1,1982 should meet the requirements of flVREG-0578 and flRR letters dated 3 September 13, and October 30, 1979. The provisions of this guide as specified 4 in Tables 1, and 2 for operating plants are comnatible with these documents. 5 wh i ch'-a re - te -be -eempl e te d-by-Jan u a ry - b-1981 - Implementation schedules for 6 these items are given in flVREG-0578 and subsequent flRR letters. The balance 1 7 of provisions of the guide are to be completed by June 1983. 8 The difficulties of procuring and installing additions or modifications 9 to in-place instrumentation have been considered in establishing these schedules. 10 Exceptions to requirements and schedules will be considered for extraordinary 11 circumstances. i s i I i ( i 18 - ~ e w e e - - ~, ,--n-w we s~'-

TABLE 1 + BWR VARIABLES Type A - Those variables to be monitored that provide the primary infomation 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 which are associated with contingency actions that may also be identified in written procedures. A variable included as Type A dor.s not preclude it from being included as Type B, C, D, or E, or vice versa. Category (see Variable Range Position C.1.3) Purpose l Plant specific ' plant specific 1 Information required for operator action 4 4 s J l e ) 19

~ ~ ~~ e TABLE 1 BWR VARIABLES (continued) TYPE B Variables - Those variables that pervide information to indicate whether plant safety functions are being accomplished. Plant safety functions are (1) reactivity control, (2) core cooling, (3) maintaining reactor coolant system integrity, (4) containment integrity (which includes radioactive effluent control). 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). Purpose TYPE B VARIABLES Reactivity Control Neutron Flux 10 to 5% full power 1 Function detection; ^ Accomplishment of mitigation Control Rad Position Full in or not 3 Verification full in RCS Soluble Boron 0 to 1000 ppm 3 Verification Concentration Core Cooling i Coolant Level in Bottom of core support 1 Function detection; the Reactor plate to above the top Accomplishment of of discharge plenum mitigation; Long-term survsillance HWR Core Thermocouples Unresolved s To monitor core cooling i if water level is low, spray is lost, or channels restricted. 20 i

1 TABLE 1 (co5..inued) 4 ' Category (see i Variable Range Position C.1.3) Purpose TYPE B - continued Maintaining Reactor Cool-ant System Integrity I 1500 N Function detection; i . RCS Pressure 15 psia to 4MMMP psig I Accomplishment of I mitigation; Verification 'Drywell Prcosure 0 to design pressure 1 Function detection;' l 2 (psig) Accomplishment of mitigation; Verification i j Drywell Sump Level Botton to top 1)[ Function detection; l 1 Accomplishment of 4 mitigation; Verification i i Maintaining Containment Integri ty Primary Containment 10 psia to design' 1 Function detection; ) Pressure (Drywell)I pressure Accomplishment of 2 mitigation; Verification l Primary Contaimment Closed - not closed 1 Accomplishment of Isolation Valve Pos-isolation l ition (excluding check valves) 4 e 1 21

TABLE 1 (continued) BWR var.IABLES (continued) TYPE C Variables - Those variables that provide information 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.I.3) Purpose TYPE C VARIABLES Fuel Cladding Radioactivity Concen-Tech Spec limit to 1 Detection of breach tration or Radiation 100 Times Tech Spec Level in Circulating limit R/hr Primary Coolant I Accident Sampling and 10 pCi/gm to 10 Ci/gm 3 Detail analysis; Analysis of Primary or TlD-14844 source term Accomplishment of Coolant in coolant volume mitigation; Gross Activity Verification; Gamma Spectrum Long-term surveillance BWR Core Thermocouples Unresolved 5 To monitor core cool-ing if water level is low, spray is lost, or channels restricted i Reactor Coolant Pressure Boundary 2000 RCS Pressurel 15 psia to 1500rpsig 1" Detection of potential for or actual breach; Accomplishment of mitigation; Long-term surveillance 7 31 Primary Containment 1 R/hr to 105 R/hr 3 Detection of breach; 1 Area Radiation Verification l Drywell Drain Sumps Bottom to top 2 Detection of breach; Level (Identified and Accomplish =ent of Unidintified Leakage) mitigation; Verification; Long-term surveillance Suppression Pool Water Bottom of ECCS suction 1 Same as immediately above Level {.or-opcraeter line to Sf t above normal @nt4 water level 22 l

e e g TABLE 1 (continued) Category (see Variable Range Position C.1.3) Purpose l TYPE C - continued I Reactor Coolant Pressure Boundary (continued) l 2 i Drywell Pressure O to design pressure 1 Detection of breach; (psig) Verification I I .1 Containment 15 psia to N psig RCS Pressurel 1 Detection of potential for breach; i Accomplishment of mitigation l Primary Containment 10 psia pressure to 3 1 Detection of potential 2 Pressure (Drywell) times design pressure for or actual breach; for concrete; 4 times Accomplishment of j design pressure for steel mitigation 0 to ]Q3 for inerted contain-1 4 i O to 30% (capability of ments Detection of potential operatin[gfrom12psiato Containment and Dry-for breach; well llydrogen Con-2 centration design pressure ) Accomplishment of mitigation 3 Containment and Dry-O to 10% (capability of 1 Detection of potential j well _0xygen Concen-operatingfrom12gsia for breach; j tration (for inerted to design pressur'e ) Accomplishment of containment plants) mitigation Containment Effluenti 10-6 to 10-2 pCi/cc 3 Detection of actual Radioactivity - Noble -breach; l Cases (from identified Accomplishment of release points includ-mitigation; ing Standby Gas Treat-Verification ment System Vent) Environs Radioactiv-10-Ito10R/hr 2 Detection of breach; l ity - Exposure Rate Accomplishment of mitigation; Verification 23

TABLE 1 (continued) BWR VARIABLES (continued) 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 are 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) Purpose TYPE D VARIABLES Condensate and Feed-water System 3 Main Feedwater Flow 0 to 110% design flow 3 Detectioa of operation; Analysis of cooling Condensate Storage Bottom to top 3 Indication of avail-Tank Level able water for cool-ing Primary Containment-Related Systems Suppression Chamber 0 to 110% design flow 3 2 To monitor operation Spray Flow 1 Drywell Pressure 12 psia to 3 psig 2 To monitor operation 2 0 to 110% design pressurc Suppression Pool Top of vent to top of 2 To monitor operation Water Level weir well Suppression Pool 30*F to 230*F 2 To monitor operation Water Temperature Drywell Atmosphere 40*F to 440*F 2 To monitor operation Temperature 3 Drywell Spray Flow 0 to 110% desir,n flow 2 To monitor operation 24 c.

TABLE 1(continued) Category (see Variable Range Position C.1.3) Purpose TYPE D - continued Main Steam System J 1 Te eeniter :; ::ti Mein St:- 'ine rir; e :: 120! decign fict Main Steamline Isola-O to 15" of water 2 )( To provide indication tion Valves' Leakage O to 5 psid of pressure boundary Control System Pressure maintenance Primary System Safety Closed-not closed or 2)( Detection of accident; Relief Valve Positions, O to 50 psig boundary integrity in-incl'. ding ADS or Flow dication Through or Pressure in Valve lines 4 1 l 1 25 I

~ TABLE 1 (continued)- Category (see Variable Range Position C.I.3) Purpose TYPE D - continued i Safety Systems 3 2 To monitor operation RCIC Flow 0 to 110% design flow r 3 2 To monitor operation HPC1 Flow 0 to 110% design flow 3 2 To monitor operation Core Spray Flow 0 to 110% design flow LPCI 3 RHa System Flow 0 to 110% design flow 2 To monitor operation A OrPCdt RMR.41aat, Exchanger 33* F-to-350

• F 2

To-moni-tor--opeeat lett-Qu elee-Temperatur-e (!ffi-)" SLCS Flow 0 to 110% desigt. flow 3 3 To monitor operation SLCS Storage Tank Bottom to top 3 To monitor operation 3 Level i Residual lleat Removal Systems 3 RilR System Flow 0 to 110% design flow , 2__ To monitor operation RHR 11ent Exchanper 32*F to 350*F 1 To monitor operation Outlet Temperature 0 d 26

TABLE 1(continutd) Category (see ~ Variable Range Position C.1.3) Purpose TYPE D - continued Cooling Water System ESF System Component 32*F to 200*F 2 To monitor operation Cooling Water Temper-ature ESF System Component 0 to 110% design flow 3 2 To monitor operation Cooling Water Flow i Radwaste Systems High Radioactivity Top to bottom 3 To monitor operation Liquid Tank Level 1 Ventilation Systems Emergency Ventilation Open-closed status 2 To monitor operation Damper Position 1 W Power Supplies 2 Status of Standby Pow-Voltages, currents, 2 To monitor operation er & Other energy pressures Sources Important to Safety f \\ 27

\\ TABLE 1 (continu:d) l l BWR VARIABLES ( continued ) l l l TYPE E Variables - Those variables to be monitored as required for use in determin-ing the magnitude of the release of radioactive materials and continually assessing such releases. Category (see Variable Range Position C.1.3) Purpose TYPE E VARIABLES Containment Radiation 7 11 Primary Containment 1 R/hr to 107 R/hr 1 Detection of signif-i Area Radiation - icant releases; l High Range Release assessment; Long-term surveillance! Emergency plan actuation 10-1 R/hr to 104 R/hr 10 Reactor Bldg or Sec-10-0-t -104 ;Oi/co-- 2 Detection of signif-ondary Containment icant releases; Area Radiation Release assessment Long-term surveillance Area Radiation 4 II Radiation Exposure 10-I R/hr to 10 R/hr 2 Detection of signif-Rate (Inside bldgs or icant releases; areas where access is Release assessment; required to service Long-term surveillance equipment important to safety) Airbo'rne Radioactive Materials Released from the Plant Noble Cases and Vent Flow Rate 0 o Drywell Purge, Stand-10-6 to 10s pCi/cc 2 Detection of signif-by Cas 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 >brk I plants) plant vent) o Secondary Containment 10-6 to 104 pCi/cc 2 Detection of signif-Purge (for Ibrk I, II, III O to 110% vent design. icant releases; plants)- flow 3 Release assessment (Not needed if cffluent discharges thru coc=on plant vent) 28

TABLE 1 (continutd) Category (see Variable Range Position C.I.3) Puroose TYPE E - continued Airborne Radioactive Materials Released 7 rom-the Plant Noble Cases and Vent Flow Rate (continued) IO o Secondary Contain-10 6 to 10" pCi/cc 2 Detection of signif-ment (reactor shield 0 to 110% vent design icant releases; b1dg annulus, if in flow 3 Release assessment l design) (Not needed if effluent dis-charges thru common plant vent) 4 10 Auxiliary Building 10 6 to 10 pCi/cc 2 Detection of signif-o (including any bldg 0 to 110% vent design icant releases; 3 containing primary flow Release assessment; system gases, e.g., (Not needed if effluent dis-Long-term surveillance waste gas decay tank) charges thru common plant vent) 3 10 o Common Plant Vent or 10-6 to 10 pCi/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 2 10 o All Other Identified 10-6 to 10 pC1/cc 2 Detection of signif-Relear.e Points 0 to 110% vent design icant releases; flow 3 Release assessment! (Not needed if effluent dis-Long-term surveillance charges thru othe.r monitored plant vents) i. 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 flov3 Release assessment; Analysis Capability Long-term surveillance e t 29 l

TABLE 1 (continued) Category (see Variable Range Position C.I.3I Purcose TYPE E - continued Environs Radiation and Radioactivity 8 11 Radiation Exposure 10-6 R/hr to 10 R/hr 2 Detection of signif-Ratel icant releases; (Installed instrument- , Verification; ation) Release assessment; Locg-term surveil', ance I Airborne Radiohalogens 10-9 to 10-3 pCi/cc 3 Release assessment; and Particulates Analysis (Sampling, with on-site analysis cap-ability) 5 Plant and Environs 0.1 to 10" R/hr, photons 3 Release assessment; 15 Radiation 0.1 to 10" rads /hr, beta 3 Analysis (Portable Instrument-radiations and low-energy ation) photons Plant and Environs Multi-channel Gac:ma-Ray 3 Releases assess 6ent; Radioactivity spectromoter Analysis (Portable Instrument-ation) l 4 30

TABLE 1 (continued) Category (see Variable Range Position C.1.3) Purcose TYPE E - Continued 16 i METEOROLOGY Wind Direction 0 to 360* (tS* accuracy 3 Release assessment with a deflection of 15*. Starting speed 0.45 eps (1.0 mph). Damping ratio between 0.4 and 0.6, dis-tance constant $2 meters. i Wind Speed 0 to 30 mps (67 mph) 10.22 ? Release assessment mps (0.5 mph) accuracy for wind rpeeds 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 Release assessment phric Stability ature differenca from pri-mary system. -5'C to 10*C (-9'F to 18'F) and 10.15*C accuracy per 50 meter int-ervals (10.3*F accuracy per 164 foot Latervals) or analogous range for back-up system. + e I 31 )

TABLE 1 (continusd) Category (see Variable - Range Position C.1.3) Purcose TYPE E - (cor.tinued) 1 ACCIDENT SAMPLING CAP

• ABILITY (Analysis Cap-ability Onsite:1 Primary Coolaat & Sump Grab Sample 33 8

Release assassment; o Gross Activity 10 pCi/mi to 10 C1/mi A alysi o Gamma Spectrum (Isotopic Analysis) o Boron Content 0 col 000 ppm o Chloride Content 0 to 20 rpm o Disolved Oxygen 0 to 20 ppm o pH 1 to 13 A Containment Air Grab Sample Y7 Release assessment; V*#ifi"**i "I o Hydrogen Content 0 to 10%. Analysis O to 30% for inerted containments o Oxygen Content 0 to 30% o Gam =a Spectrum (Noble gas analysis) i i i

• The time for taking and analysing sa=ples should be 3. hours or less from the l

time the decision is made to sample, except chloride which should be within 24 hours. l 32 e

TABLE 1 (continued) NOTES IWhere a variable is listed for more than one purpose, the instrumentation requirements may be integrated and only one measure =ent provided. b aign 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 maximum flow anticipated in normal operation. "The maximum value may be revised upward to satisfy ATUS requirements. % e number.of thermocouples,.their range and lecation to be. determined. 6Measurement should be made of the gross gamma radiation emanating from circulating pri-mary coolant, with instru=ent calibration permitting conversion of readout to radioac-tivity concentrations in terms of either curies / gram or curies / unit-volume. System accuracy should be ih order of magnitude. The point of measurement should be external to a circulating primary coolant line or loop, and should'not be a line or loop subject to isolation, e.g., main steam line. While such an instrument may not be currently available off-the-shelf, the staff considers that the necessary components are avail-able commercially and have been employed and demonstrated under adverse environmental conditions in high-level hot cell operations for many years. 7111nimum of two monitors at widely separated locations. 8For 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 locat$ons - site dependent.) 9Provisions should be made to monitor all identified pathways for relesse 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-Icasad directly into the environment. If two or more streams are co=bined prior to release from a common discharge point, monitoring of the combined atream is considered to meet the intent of this guide provided such monitoring has a range adequate to measure worst-case releases. 10Monitors should be capabic of detecting and measuring radioactive gaseous effluent con-centrations with compositions ranging f rom fresh equilibrium noble gas fission product mixtures to 10-day old mixturcs, with overall system accuracies of h decade. Effluent concentrations may be expressed in terms of Xc-133 equivalents or in terms of any noble gas nuclide(s). It is not expected that a singic monitoring device will have sufficient range to encompass the entire range provided in this guide and that multiple components or systems will be needed. Existing equipment may be utilized to monitor any portion of the stated range within the equipment design rating. Addi+iel--enended-ange-4nstro-mentacion-should-oveclap-tha-rango--oC-existing-,inscrumantat4on-.by-a--least-a-foo or-of 2, 11Detcetors sho'ild respond to gamma radiation photons within any energy range from 60 kev to 3 McV with an accuracy of a20% at any specific photon energy from 0.1 MeV to 3 McV. Overall system accuracy should be within ih decade over the entire range. 33 e

TABLE 1 (continued) ~ NOTES - continued l 12Status indication of all Standby Power A-C busee. D-C-buses, inverter output buses and pneumatic supplies. 13To provide information regarding release of radioactive halogens and particulates. Continous collection of representative samples followed by onsite laboratory measure-ments of sampics for radiohalogens and particulates. The design envelope for shield-ing, handling, and analytical purposes should assume 30 minutes of integrated sampling 2 pCi/cc of radioiodines time at sampler design flow, an average concentrations of 10 2 in gaseous or vapor form, an avera;e concentration of 10 pCi/cc of particulate radio-iodines and particulates other than radiciodines, and an average gamma photon energy of 0.5 MeV per disintegration. 14 or estimating release rates of radioactive materials released during an accident.from F ~ unidentified release paths (not covered by effluent monitors). Continous collection of representative sampics followed by laboratory measurements of the samples. (Approximately 16 to 20 locations - site dependent.) ISTo monitor radiation and airborne radioactivity concentrations in many arcas through- .out the facility and the site environs where it is impractical to install stationary monitors capable of covering both normal and accident levels. 16Hetcorological measurements should conform to the' provisions of tbc 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 sanples. 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 to minimise plateout or deposition. For safe and convenient sampling, the provisions should include: a. Chielding to maintain radiation deses ALARA, b. Sample containers with container-sampling port connector compatability, c. Capability of sampling under primary system pressure and negative pressures, d. Ilandling and transport capability, and c. Pre-arrangement for analysis and interpretation, 36An installed capability should be provided for obtaining containment sump, ECCS pump room sumps, and other cimilar auxiliary building sump liquid samples. 4 34

- = i TABLE 2 PWR VARIABLES I Type A - Those variables to be monitored that provide the primary information i 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-1 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,' D, or E, or vice versa. 1 Catekory see ) Purpose __ Variable Range .1.3 Posi ion ~ Plant specific plant specific 1 Information required for operator action J r I 4 Aw a e l 1 35 1 -

TABLE 2 PWR VARIABLES (continued) 1 Type B Jariables - Those variables that provide infomation to indicate whether plant sa'ety functions are being accomplished. Plant safety functions are (1) reactivity control, (2) core cooling, (3) maintaining reactor coolant system integrity, (4) containment integrity (which includes radioactive effluent control). i Vari 2bles are listed with designated ranges and category for design and qualification i requi rements. Key va. ables are indicated by design and qualification Category 1. Category (see Variable Range Position C.I.3) Purpose TYPE B VARIABLES Reactivity Control Neutron Flux 10-6 to 57. full power 1 Function detection; j Accomplishment of mitigation. i 4 Control Rod Posf tion Ful'. in or not 3 Verification i full in i RCS Soluble aoron 0 to 6000 ppm 3 Verification Concentration li RCS. Cold Leg Temper-50*F to 400*F 3 ~ Verification l ature Core Cooling i RCS !!ot Leg Temper-50*F to 750*F 1 Function detection; ature Accomplishmer.L of j mitigation; Verification; Long-term surveillance j I i i RCS Cold Leg Temper-50*F to 750*F 1 Function detection; l i ature Accomplishment of mitigation; Verification; Long-term surveillance RCS Pressurel 0 to 3000 psig 1" Function detection; (4000 psig for CE Accomplishment of-plants) mitigation; Verification; Long-term surveillance l l 36 lL...,

TAELE 2 (.continuedl ~' Category (see Variable Range Position C.1.3) Purpose TYPE B - continued ' Core Cooling (continued) 200* l Core Exit Temperature t96#F to 2300*F 3 Verification (for operating plants - 200* t$0aF to 1650*F) t Coolant Level in Bottom of core to 1 (Direct Verification, accom-Reactor top of vessel indicating or re-plishment of mitigation cording device not i needed) Degrees of Subcooling 200*F subcooling to 1 Verification and 35'F superheat (for operating analysis of plant plants - 2, with conditions confirmatory oper-ator procedures) Maintaining Reactor Coolant System Integrity l RCS Pressure 0 ra 3000 psig 1 Function detection; (4000 psig for CE Accomplishment of plants) mitigation i l Containment Sump Water Narrow range (sump), 3' Function detection; i Level Wide range (bottom of 1 Accomplishment of containment to 600,000-mitigation; gallon level equivalent) Verification l 2 Containment Pressure 0 to design pressure 1 Function detection; (psig) Accomplishment of mitigation; Verification O e 4 37 s

TABLE 2(continued) Category (see i Variable Range Position C.I.3) Purpose l TYPE B - continued Maintaining Containment Integ ri ty i i Contain=cnt Isolation Closed-not closed 1 Accomplishment of Valve Position (exclud-isolation in', check valves) l Containment Pressurel 30 pain to design pressure 2 1 Function _ detection Accomolishment of f nitJnation_ Verification i I 1 I j l 4 38 e y. -.,__.,.p 4 s

g. N e

TABLE 2 (continued) PWR VARIABLES (. continued) TYPE C Variables - Those variables that provide infomation-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.I.3) Purpose TYPE C VARIABLES Fuel Cladding 200* 5 Core Exit Temperatur:I M F F to 2300*F 1 lietection of potential (for operating plants - for breach; 200* H02F to 1650*F) Accomplishment of mitigation; Long-term surveillance Radioactivity Concen-is Tech Spec limit to 1 Detection of breach t tration or Radiation 100 times Tech Spec Level in Circulating limit R/hr Primary Coolant 18 Accident Sampling and 10 pCi/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; ' Gamma Spectrum Long-term surveillance Reactor Coolant Pressure Boundary l 4 RCS Pressure 0 to 3000 psig 1 Detection of potential (4000 psig for CE for or actual breach; plants) Accomplishment of mitigation; Long-term surveillance ^ l Containment Pressura 10 psia to design 1 Detection of breach; 2 pressure psig Accomplishment of (5 psia for sub-atmos-mitigation; pheric containments) Verification; i Long-term surveillance 39

TABLE 2 (continued) ~ Category (see Variable Range Position C.1.3) Purpose TYPE C - continued 4 Reactor Coolant Pressure Boundary (continued) Containment Sump Water Narrow range (sump), 3 Detection of breach; Levell Wide range (bottom of 1 . Accomplishment of containment to 600,000-mitigation; gallon level equivalent) Verification; Long-term surveillance i 7 11 Containment Area 1 to 10" R/hr 3 Detection of breach; i Radiation Verification Effluent Radioactivity 6 to 10-2 tici/cc 3 Detection of breach; Noble Gas Effluent from Verification Condenser Air Removal System ExhaustI { l l Containment RCS Pressure O to 3000 psig 1" Detection of potential l (4000 psig for CE. for breach; plants) Accomplishment of mitigation Containment Hydrogen 0 to 10% (capable of oper-1 Detection of potential Concentration ating from 10 psia to max-for breach; 2 imum design pressure ) Accomplishment of E y ont e l l Contain:nent Pressure 10 psia pressure to 3 times 1 Detection of potential l 2 design pr 'ure for concrete; for or actual breach; Jn pressure for Accomplishment of 4 times ) steel (5 p..a for sub-atmospheric mitigation containments) I 40 )

t TABLE 2 (continued) Category (see Variable Range Position C.I.3) Purpose l TYPE C - continued Containment (continued) 9 0 Containment Effluent 10-6 to 10-2 pCi/cc 2 Detection of breach; Radioactivity - Noble Accomplishment of Cases from Identified mitigation; Release Pointal Verification 8 11 'l Environs Radioactiv-10-4 to 10 R/hr 2 Detection of breach; ity - Exposure Rate Accomplishment of l mitigation; Verification r l j l 1 a l 1 l a i i k I ~ l 1 41 \\

TABLE 2 (continusd) PWR VARIABLES (continued) 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 are 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 i Variable Range Position C.I.3) Purpose 1YPE D VARIABLES 4 Residual Heat Removal or Decay Heat Removal System 3 2 To monitor operation RHR System Flow 0 to 110% design flow 1 RHR Heat Exchanger 32*F to 350*F 2 To monitor soperation Out Temperature and for analysis 4 4 i . Safety injection Systems Accumulator Tank Level 10% to 90% volume 2 To monitor operation Level or Pressure 0 to 750 psig Accumulator Isola-Closed or Open 2 Operation status tion Valve Position Boric Acid Charging 0 to 110% design [ low 3 3 To monitor operation Flow l Flow in HPI System 0 to 110% design flow 3 2 To' monitor operation 1 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 System Reactor Coolant Pump Motor current 3 To monitor operation. Status I 1 42

~ o TABLE 2(continued) Category (see Variable Range Position C.1.3) Purpose TYPE D - continued Primary Coolant System - (continued) Primary System Safety Closed-not closed 2 Operation status; to Relief Valve Positions monitor for loss of (including PORV and coolant code valves) or Flow Through or Pressure in Relief Valve Lines P.essurizer Level Bottom to top 1 To assure proper oper-ation of pressurizer Pressurizer Heater Electric current 3 To determine operating Status status Quench Tank Level Top to bottom 3 To monitor operation Quench Tank Temp-50*F to 750*F 3 To monitor operation crature 2 3 To monitor operation Quench Tank Pressure O to design pressure . Secondary System (Steam Generator) Steam Concrator Level From tube sheet to 4-{Sabegoge 1 To monitor operation separators 6es. 2 4 cop-plants) Steam Generator From ntmospheric pressure 2 To monitor operation Pressure to 20% above the lowcst safety valve setting Safety /Rolief Valve Closed - not closed 2 To monitor operation Positions or Main Steam Flow 3 Main Feedwater Flow 0 to 110% design flow To monitor operation ^ 43 O

TABLE 2(continued) Category (see ~ Variable Range Position C.1.3) Purpose , TYPE O - continued Auxiliary Feedwater or Emergency Feedwater System ~ 3 2 To monitor operation Auxiliary or Emergency 0 to 110% design flow Feedwater Flow (1 for i B&W plants) Ocndansate 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 AFW should be listed and should be Category 1) l 4 I Containment Cooling Systems 3 2 To monitor operation Containment Spray 0 to 110% design flow Flng Heat Removal by the Plant specific 2 To monitor operation Containment Fan Heat Removal System Containment Atmos-40*F to 400*F 3 To indicate accomplish-phere Temperature ment of cooling Containment Sump 50*F to 250*F 2 To monitor operation Water Temperature J '44 I 9

TABLE 2(continu:d) l. Category (see Variable Range Position C.1.3) Purpose TYPE D - continued Chemical and Volume Control System 3 2 To monitor operation Makeup Flow - In 0 to 110% design flow 3 2 'To monitor operation Letdown Flow - Out 0 to 110% design flow Volume Control Top to bottom 2 To monitor operation Tank Level Cooling Water System Component Cooling 32*F to 200*F -2 To monitor operation Water Temperature to ESF System Components 3 2 To monitor operation Component Cooling 0 to 110% design flow Water Flow to ESF System Components Radwaste Systems High-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 Systems Emergency Ventila-Open-closed' status. 2 To indicate damper tion Damper Position status Power Supplies 13 Status of Standby Voltages, currents, 2 To indicate system Power 6 Other Energy pressures status Sources Important to Safety 45

TABLE 2 (continued) PWR VARIABLES (continued) TYPE E Variables - Those variables to be monitored as required for use in determin-ing the magnitude of the release of radioactive materials and continually assessing such releases. Category (see Variable Range Position C.1.3) Purpose TYPE E VARIABLES Cantainment Radiation Containment Area 1 R/hr to 107 n/hr 1 Detection of signif-l , Radiation - 111 Range icant releases; Release assessment; Long-term surveillance; Emergency plan actuation Area Radiation 4 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-tera surveillance equipment important to safety) Airborne Radioactive Materials Celeased from the Plant Noble Gases and Vent Flow Race 10 o Containment or Purge 10-6.to 10s pCi/cc 2 Detection of signif-Effluentl 0 to 110% vent design icant releases; 3 flow Release assessment (Not needed if effluent dis-charges thru common plant vent) 4 10 o Secondary Contain-10-6 to 10 uCi/cc 2 Detection of signif-ment (reactor shield 0 to 110% vent design icant releases; b1dg annulus, if in flow 3 Release assessment design) (Not needed if effluent dis-charges thru common plant vent) 4 10 o Auxiliary Building 10-6 to 10 uCi/cc 2 Detection of signif-(including any bldg 0 to 110% vent design icant releases; containing primary flow 3 Release assessment; systen gases, e.g., (not needed if effluent dis-Long-tern surveillance waste gas decay tank) charges thru common plant vent) 46

m, q i TABLE 2'(continued) i Category (see Variable Range Position C.1.3) Purpose TYPE E - continued Airborne Radioactive Materials Release from the Plant (continued) Noble Gases and Vent Flow Rate (continued) I0 o Condensor Air Removal 10-6 to 105 pCi/cc 2 Detection of signif-l System Exhaust 0 to-110% vent design icant releases; flow 3 Release assessment (Not needed if effluent dis-charges thru common plant vent) 10 o Commen Plant Vent or 10-6 to 103 pCi/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 strveillance o Vent From Steam Gen-10-1 to 10 3 pCi/cc 2 Detection of signif-erator Safety Relief (puration of releases in icant releases; Valves or Atmospheric seconds, and mass of steam Release assessment-Dump Valves per unit time) .o All Other Identified 10-6 to lo2'pCi/cc 2 Detection of signif-10 Release Points 0 to 110% vent design Icant releases; flow delease assessment; (Not needed if effluent dis-Long-term surveillance charges taru other monitored plant vents) Particulates 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; cept Steam Generator flow 3 Release assessment; Safety Relief Valves Long-term surveillance or Atmospheric Steam Dump Valves and Con-densor Air Removal System Exhaust) Sampling, With On-site Analysis Cap-ability i i 47 l t

TABLE 2 (continued) Category (see Variable Range Position C.I.3) Purpose TYPE E - continued Environs Radiation and Radioactivi ty a 11 Radiation Exposure 10-6 R/hr to 10 R/hr 2 Detection of signif-Ratel icant releases; (Installed instrument-Verification'; ation) ' Release assessment; Long-term surveillance Airborne Radiohalogens 10-9 to 10-3 pCi/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; 4 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 j (Portable Instrument-f j ation) I 1

TABLE-2 (continued) Category (see Variable Range Position C.1.3) Purpose ~' TYPE E - Continued METEOROLOGY 17 Wind Direction 0 to 360' (tS' accuracy 3 Release assessment with it deflection of 15*. Starting speed 0.45 mps (1.0 mph). Damping ratio between 0.4 and 0.6, dis-tance constant $2 meters. Wind Speed 0 to 30 mps (67 mph) 0.22 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 O.45 mps (1.0 mph). Estimation of Atmos-Based on vertical temper-3 Release assessment phric Stability ature difference from pri-mary system, -5'C to 10*C (-9'F to 18'F) and 10.15'C accuracy per 50 meter int-ervalc ( 0.3*F accuracy per 164 foot intervals) or analogous range for back-up system. l l O 1 k lr 49 b

TABLE 2 (continued) Category (see Variable Range Position C.I.3) Purpose TYPE E - (continued) ACCIDQlT SAMPLIrlG CAP * - ABILIli (Analysis Cap-ability Onsite) 9 Primary Coolant & Sump Grab Sample. 3 Release assessment; o Gross Activity 10 pCi/mi to 10 Ci/ml A al s o Camma Spectrum (Isotopic Analysis) o Boron Content 0 to 6000 ppm 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; "I o Hydrogen Content 0 to 10% Analysis O to 30% for ice condensors o Oxygen Content 0 to 30% o Camma Spectrum (Noble gas analysis) I 1 I I i j

• The time for taking and analysing samples should be 3 hours or less from the l

time the decision is made to sample, except chloride ~ which should be within 24 hours. 4 r i l 50 l t

TABLE 2 (continu;d) NOTES IWhere 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 maximum flow anticipated in normal operation. "The maximum value may be revised upward to satisfy ATWS requiremen'ts. 5A minimum of 4 measurements per quadrant is required for operation. Sufficient number should be installed to account for attrition. (Replacement instrumentation should meet de 2300*F rarme provis ion. ) 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 h order of magnitudt.. 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 =ay 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. 7Minimum of two monitors at widely separated locations. 8 For estimating release rates of radioactive materials released'during an accident from u*nidentified release paths (not covered 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 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. 10Monitors should be capable ot detecting and reasuring 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 concentrations may be expressed in terms of Xc-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 systems will be needed. Existing equipment may be utilized to monitor any portion of the stated range within the equipment design rating. -Add i t ional-e x tended-range-ins t ru-me n ta t4on-shou 1 J-ove r-la p-t he-ra n ge-o f-ex is t-in g -ins t r umen t a t ion-by-a--lea a t-e--f ac to r-o f-A 11 Detectors should respond to gamma radiation photons within any energy range from 60 kev to 3 MeV with an accuracy of 120% at any specific photon energy from 0.1 MeV to 3 MeV. Overall system accuracy should be within ib decade over the entire range. 51

TABLE 2 (continued) ~ N9TES - continued monitor _s 12 Effluent3 or PWR steam safety valve discharges and atmospheric steam dump valve dis-f charges should be capable of approximately linear response to gnmmn radiation photons with energies from approxinately 0.5 MeV to 3 MeV. Overall system accuracy should be within I\\ 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-22, and Co-60). Effluent concentrations should be expressed in terms of any gnmmn-emitting noble gas nuclide within the specified energy range. Calculational methods should be provided for est-imating concurrent releases of low-energy noble gases which cannot be detected or measured by the methods or techniques employed for monitoring. 13Status indication of all Standby Power A-C buses, D-C buses, inverter output buses and pneumatic supplies. 4 14To provide information regarding release of radioactive halogens and particulates. Ccntinous collection of representative samples followed by onsite laboratory =easure-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 time at sampler design flow, an average concentrations of 10 pCi/cc of radioiodines 2 in gaseous or vapor form, an average concentration of 10 pCi/cc of particulate radio-iodines and particult.tes other than radioiodines, and an average gamma photon energy of 0.5 McV per disintegration. 15For estimating release rates of radioactive materials released ducing an accident.from unidentified release paths (not covered 'by effluent monitors). Continous collection of representative samples followed by laboratory measurements of the sa=ples. (Approximately 16 to 20 locations - site dependent.) 16To monitor radiation and airborne radicactivity 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 eteorological measurements should conform to the provisions of the forthcoming rev-M ision of Regulatory Guide 1.23, "Onsite Miteorological Prograns". 18 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 to minimize plateout or deposition. For safe and convenient sampling, the provi. ions should include: a. Shielding to maintain radiation deses ALARA, b. Sample containers with container-sampling port connector compatability, c. Capability of sampling under primary system pressure and negative pressures, d. Handling and transport capability, and e. Pre-arrangement for analysis and interpretation. 19 n installed capability should be provided for obtaining containment sump, ECCS pu=p A room sumps, and other similar auxiliary building sump liquid samples. 52. r - n

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 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 considgration of degraded core cooling.

1. 2 Heed m

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-Tera Recc=endations," I'UREG-0578. 53

o 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 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 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 national standard'(ANS-4.5). For future plants, the staff review will be sibplified with guidance contained in the endorsed standard developed by a voluntary standards... group and the regulatory guide, which includes a list of variables for accident monitoring. Efforts by the staff to implement Revision 1 to Regulatory Guide 1.97 has been fra~ught 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 clearly defined NRC position with regard to instrumentation to assess plant and environs conditions during and 54

~ ~ 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 in'strumentation 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 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 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 B 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 I 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 improve public safety by ensuring that the plant operator will have timely information to take any necessary action to protect the public. 55

. + No impact on the public can be foreseen. ~ 1.4 Decision on' Proposed Action ~ As previously stated, more definitive guidance on instrumentation to assess plant and environs conditions during and following an accident should be given. ~ 2. 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 s g 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.ac. tion 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 ) I ~ are foreseen. This guide does include the variables to be monitored on site by i 56 I (

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. 6. SUfVtARY AtlD C0f1CLUSI0f15 The revision to Regulatory Guide 1.97, " Instrumentation For Light-Water-J Cooled fluclear Power Plants to Assess Plant and En,v, irons Conditions During and Following an Accident," should be issued and implemented according to existing schedules. d e g O 4 57 4 Y ~w

^ ./ APPEtIDIX FOR SECTION III' EXHIBIT 37 (continued) ( 8A i g ? ~3 b d d 3, /,ccessionUnit{ 5: !=g g %g i,c : - .o : Roar 050 EEi lips Euildin<> ro From: .. jg g Please place the attached docut.ent in the PDP. usir.g tiie following file ar.1 file \\ points: PDR File Additional Info (Sclect one and enter number) (Enter if appropriate) Proposed Rule (PR) ACRS Minutes fio. Rec. Guide $ A ( M J d Q Relates to Propcsed Riile (PR) Petition (PR.1f Relates to Reg. Guide fe ( Relates to Petition (PPJi)d_6fdfS7 Effective Rule Tei) ( Af:SI Relates to Effective Rule [Pi-i) IAEA SD Task tio. 25 f))-y Federal Register ilotice !!UREG Report Contract lio.

Subject:

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