ML19256G493
| ML19256G493 | |
| Person / Time | |
|---|---|
| Issue date: | 12/31/1979 |
| From: | NRC OFFICE OF STANDARDS DEVELOPMENT |
| To: | |
| References | |
| TASK-OS, TASK-RS-917-4 REGGD-01.097, REGGD-1.097, TAC-48786, NUDOCS 7912310387 | |
| Download: ML19256G493 (32) | |
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[ ',,".., 'o,j NUCLEAR REGULATORY COMMISSION o
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yr Dece:f.c r 4, 1979 REGULATORY GUIDE DISTRIBUTION LIST (DIVISION 1)
Proposed Revisicn 2 to Regulatory r-side 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to As,ess Plant and Environs Conditions During and Following an Accident," is being developed to describe a method acceptable to the NRC staff for complying with the Commission's requirements to provide instrumentation to monitor plant variables and systems curing and following accidents in a nuclear power plant. This proposed revision is being released for comment to encourage public participation in its development.
The proposed revision of Regulatory Guide 1.97 was initiated as a result of the lessons learned from an evaluation of the investigation of TMI-2 and is being developed on a high-priority basis. The scope of proposed Revision 2 has been expanded to include all accident-monitoring instrumentation needed by the plant operator (licensee) to protect the health and safety of the public, including that required for emergency planning.
The guide also includes consideration of degraded core cooling conditions to a greater extent than was considered in Revision 1 to Regulatory Guide 1.97, which was issued in August 1977.
The guide endorses, with certain exceptions, a standard which is still under development, draf t ANS-4.5, " Functional Recuirements for Accident Monitoring in a Nuclear Power Generating Station," Draft 4. dated November 1979.
The draf t standard has not yet been approved as an American National Standard, but per-mission has been granted to use Draft 4 of the standard with proposed Revision 2 of Regulatory Guide 1.97 during the public review and coment period.
Ccements received on the draft standard will be transmitted to the ANS-4 working group that is developing the standard in addition to being given consideration by the staff for inclusion in the reguletory guide.
Since the standard and the guide ere being developed in parallel, ccments can be resolved, as appropriate, by modification to either the standard or the guide.
Concurrent with the public review and comment period, the NRC staf f will arrange meetings with the various owners' groups and/or utilities to obtain input on backfitting recommendations and impact.
It would be helpful if comments on the guide and the standard are accompanied by specific word changes to the guide in order to avoid any misunderstanding as to the thrust of the comment.
Sincerely, 1664 220
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Guy A. Arlotto, Director Division of Engineering Standards Office of Standards Development 7012339 Q
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U.S. NUCLEAR REGULATORY COMMISSION i Y ')/w-*
0FFICE OF STANDARDS DEVELOPMENT o
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DRAFT REGULATORY GUIDE AND VALUE/ IMPACT STATEMENT Division 1
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PROPOSED REVISION 2* TO REGULATORY GUIDE 1.91 INSTRUMENTATION FOR LIGHT-WATER-COOLED NUCLEAR PGWER PLANTS TO ASSESS PLANT AND ENVIRONS CONDITIONS DURING AND FOLLOWING AN ACCIDENT A.
INTRODUCTION Criterion 13, " Instrumentation and Control," of Appendix A, " General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, " Domestic Licensing of Production and Utilization Facilities," includes a requirement'that instrumen-tation be provided to monitor variables and systems for accident conditions as appropriate to ensure adequate safety.
Criterion 19, " Control Room," of Appendix A to 10_CFR Part 50 includes a requirement that a control room be provided from which actions can be taken to maintain the nuclear power unit in a safe condition under accident conditions, including loss-of-coolant accidents, and that equipment,l including the necessary instrumentation, at appropriate locations outside the control room be provided
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a with a design capability for prompt hot shutdoun of the reactor.
Criterion 64, " Monitoring Radioactivity-Releases," of Appendix A to 10 CFR Part 50 include:S a requirement that means be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluid, effluent discharge paths, and the plant environs for radioactivity that may be released from postulated accidents.
This guide describes a method acceptable to the NRC staff for complying with the Commission's regulations to" provide instrumentation to monitor plant variables and systems during and following an accident in a light-water-cooled nuclear power plant.
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"The substantial nu er of changes in this proposed revision has made it imprac-tical to indicate the" changes with lines in the margin.
This regulatory guide arid the associated value/ impact statement are being issued in draf t form to involve the public in the early stages of the development of a regulatory position in this area.
They have not received complete staff, review, have not been reviewed by the HPC Regulatory Requirements Review Committee, and do not represent an.of ficial NRC staf f position.
Pubile comments are being solicited on both drafts, the guide (including any implementation schedule) and D
data.
the value/ impact statement. Comments on the value/ impact statement should be accompanied by supporting Comments on both draf ts should be sent to the Secretary of the Commission,FEB U.S. Nuclear Regulatory Commission, Wash.ngton, D.C. 20555, Attention: Dock ting and Service Branch, by 14 Iggt)
Requests for single copies of draf t guides (which may be reproduced) or for placement on an automatic distribution list for single copies of future draf t guides in specific divisions should be made in writing to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attention: Director, Division of Technical Information and Document Control.
B.
DISCUSSION Indications of plant variables and status of systems important to safety are required by the plant operator (licensee) during accident situations to (1) provide information required to permit the operator to take preplanned manual actions to accomplish safe plant shatdown; (2) determine whether the reactor trip, engineered-safety-feature systems, and manually initiated systems are performing their intended functions (i.e., reactivity control, core co ling, maintaining reactor coolant system integrity, and maintaining containment integrity); (3) provide information to the operator that will enable him to determine the potential for causing a breach of the barriers to radioactivity release (i.e., fuel cladding, reactor coolant pressure boundary, and containment) and if a barrier has been breached; (4) furnish data for deciding on the need to take unplanned action if an automatic or manually initiated safety system is not functioning properly or the plant is not responding properly to the safety systems in operation; and (5) allow for early indication of the need to initiate action necessary to protect the public and for an estimate of the magnitude of the impending threat.
At the start of an accident, it may be difficult for the operator to deter-mine immediately what accident has occurred or is occurring and, therefore, to determine the appropriate response.
For this reason, reactor trip and certain other safety actions (e.g., emergency core cooling actuation, containment isola-tion, or depressurization) have been designed to be performed automatically during the initial stages of an accident.
I..strumentation is also provided to indicate information about plant parameters required to enable the operation of manually initiated safety systems and other appropriate operator actions involv-ing systems important to safety.
Instrumentation is also needed to provide information about some plant parameters.that will alert the operator to conditions that have degraded beyond those postulated in the accident analysis.
In particular, it is important that the operator be informed regarding that status of coolant level in the reactor vessel or the existence of core voiding that would indicate degraded core cooling.
Direct indication of coolant level in the reactor vessel is not currently available in pressurized water reactors.
However, it is imperative that this capability be developed within a reasonable time in order to provide the operator with this vital information in a positive, unambiguous manner.
It is essential that degraded conditions be identified so that the operator can take 1664 222 2
actions what are available to mitigate the consequences.
It is not intended that the operator be encouraned to prematurely circumvent systems important to safety but that he be adequately informed in order that unplanned actions can be taken when necessary.
Examples of serious events that could threaten safety if conditions degrade beyond those assumed in the Final Safety Analysis Report are loss of-coolant accidents (LOCAs), overpressure transients, anticipated transients without scram (ATWS), reactivity excursions, and releases of radioactive materials.
Such events require that the operator understand, within a short time period, the ability of the barriers to limit radioactivity release, i.e.,
the potential for breach of a barrier, or an actual breach of a barrier by an accident in progress.
It is essential that the required instrumentation be capable of surviving the accident environment in which it is located for the length of time its func-tion is required as defined by Section 3.0 of Draft Standard ANS-4.5,* " Functional Requirements for Accident Monitoring in a Nuclear Power Generating Station,"
Draft 4 dated November 1979.
It could therefore either be designed to withstand the accident environment or be protected by a local protected environment.
If the environment surrounding an instrument component is the same for accident and normal operating conditions (e.g., some instrumentation components outside of containment or those in the main control room powered by a Class 1E source.i.
the instrumentation components need no special environmental qualification.
It is important that accident-monitoring instrumentation components and their mounts that cannot be located in other than non-Seismic Category I build-ings be conservatively designed for the intended service.
Furameters selected for accident monitoring can be selected so as to permit relatively few instruments to provide the essential information needed by the operator for postaccident monitoring.
Further, it is prudent that a limited
- Copies may be obtained from the American Nuclear Society, 555 North Kensington Avenue, La Grange Park, Illinois 60525.
Although this standard has been balloted by the responsible subcommittee and reviewed by the responsible con-senpus body, Draft 4 does not reflect the resolution of all comments.
A sub-sequent draft is intended to address the comments that formed the basis of the negative subcommittee ballots.
h 1664 223 3
number of those parameters (e.g., containment pressure, primary system pressure) be monitored by instruments qualified to more stringent environmental require-ments and with ranges that extend well beyond that which the selected parameters can attain under limiting conditions.
It is essential that the range selections not be arbitrary but sufficiently high that the instruments will always be on scale; for example, a range for the containment pressure monitor extending to the burst pressura of the containment in order that the operator will rat be blind as to the level of containment pressure.
Provisions of such instruments are important so that responses to corrective actions can be observed and the need for, and magnitude of, further actions determined.
On the other hand, it is also necessary to make sure that when a range is extended, the sensitivity and accuracy of the instrument are within acceptable limits.
Normal power plant instrumentation remaining functional for all accident conditions can provide indication, records, and (with certain types of instru-ments) time-history responses for many parameters important to following the course of the accident.
Therefore, it is prudent to select the required accident-monitoring instrumentation from the normal power plant instrumentation to enable the operator to use, during accident situations, instruments with which he is most familiar.
Since some accidents impose severe operating requirements on instrumen-tation components, it may be necessary to upgrada those instrumentation componente to withstand the more severe operating conditions and to measure greater variations of monitored variables that may be associated with the accident if they are to be used for both accident and normal operation.
However, it is essential that instrumentation so upgraded does not compromise the accuracy and sensitivity required for normal operation.
In some cases, this will necessitate use of overlapping ranges of instruments to monitor the required range of the parameter to be monitored.
Draft Standard ANS-4.5, Draft 4 dated November 1979, delineates criteria for determining the variables to be monitored by the control room operator, as required for safety, during the course of an accident and during the long-term stable shutdown phase followng an accident.
Draft Standard ANS-4.5 was prepared by Working Group 4.5 of subcommittee ANS-4 with two primary objectives:
(1) to address that instrumentation that permits the operator to monitor expected parameter changes in an accident period and (2) to address extended range instrumentation deemed appropriate for the possibility of encountering previously unforeseen events.
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The standard defines four classifications of variable types for the purpose of aiding the designer in his selection of accident-monitoring instrumentation and applicable criteria. (A fifth type [ Type E] has been added by this regula-tory guide. ) The types are: (1) Type A - those variables that provide informa-tion needed for preplanned operator actions, (2) Type B - those variables that provide information to indicate whether plant safety functions are being accom-plished, (3) Type C - those variables that provide information to indicate the potential for being breached or the actual breach of the barriers to fission product release, i.e., fuel cladding, primary coolant pressure boundary, and containment, (4) Type 0 - those variables that provide information to indicate the performance of individual safety systems, and (5) Type E - those variables to be monitored as required for use in c'atermining the magnitude of the release of radioactive materials and for continuously assessing such releases, for providing defense in depth, and for diagnosis. Type A variables have not been included in the listings of variables to be measured because they are plant specific and will depend on the operations that the designer chooses for pre-plsnned manual action. The five classifications are not mutually exclusive in that a given variable (or instrument) may be included in one or more types, as well as for normal power plant operation or for automatically initiated safety actions. Where such multiple listing or use occurs, it is essential that instrumentation be capable of meeting the most stringent requirements. The time phases (Phases I, II, and III) delineated in ANS-4.5 are not specified for each variable in this regulatory guide. These considerations are plant specific. It is important that the required instrumentation survive the accident environment and function as long as the information it provides is needed by the plant operator. C. REGULATORY POSITION The criteria, requirements, and recommendations (identified as important to safety) contained in Draft Standard ANS-4.5, " Functional Requirements for Accident Monitoring in a Nuclear Power Generating Station," Draft 4 dated November 1979, are considered by the NRC staff to be generally acceptable for providing instrumentation to monitor variables and systems for accident D conditions and for monitoring the reactor containment, spaces containing compo-nents for recirculation of loss-of-coolant accident fluids, effluent discharge 1664 225 5
paths, and the plant environs for radioactivity that may be released during and following an accident from a nuclear power plant subject to the following: 1. Section 2.0 of ANS-4.5 defines the scope of the standard as contain-ing criteria for determining the variables to be monitored by the control room operator during and following an accident that will need some operator action. Consideration should be given to the additional requirements (e.g., emergency planning) of variables to be monitored by the plant operator (licensee) during and following an accident. Instrumentation selected for use by the plant cpera-tor for monitoring conditions of the plant is useful in an emergency situation and for other purposes and therefore should be factored into the emergency plans action level criteria. 2. In Section 3.0 of ANS-4.5, the definition of " Type C" includes two items, (1) and (2). Item (1) includes those instruments that indicate the extent to which parameters that indicate the potential for a breach in the containment have exceeded the design basis values. In conjunction with the parameters that indicate the potential for a breach in the containment, the parameters that have the potential for causing a breach in the fuel clawing (e.g., core exit temperature) and the reactor coolant pressure boundary (e.g., reactor coolant pressure) should also be included. References to Type C instru-ments, and associated parameters to be measured, in Draft Standard ANS-4.5 (e.g., Sections 4.2, 5.0, 5.1.3, 5.2, 6.1, 6.3) should include this expanded definition. 3. Section 3.0 of ANS-4.5 defines design basis accident events. In conjunction with the design basis accident events delineated in the standard, those events that are expected to occur one or more times during the life of a nuclear power unit and include but are not limited to loss of power to all recirculating pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power should be included. 4. Section 4.2 of ANS-4.5 discusses the various types of variables. With regard to the discussion of Type D variables, Type D variables and instru-ments are within the scope of Accident Monitoring Instrumentation although they are not addressed in Draft Standard ANS-4.5. They are, however, along with those of an additional type, Type E, included in this regulatory guide. (See Tables 1, 2, and 3.) 5. Section 6.1 of ANS-4.5 pertains to General Design Criteria for instru-mentateion monitoring Types A, B, and C variables. In conjunction with Section 6.1, l 1664 226 6
instrumentation monitoring Types 0 and E variables should also be included. Noted applicable design criteria are identified in Table 1 of this regulatory guide. 6. Section 6.1.2 of ANS-4.5 pertains to the duration that instrumentation is qualified to function. In conjunction with Section 6.1.2, Phase II instrumen-tation should be qualified to function for not less than 200 days unless a shorter time, based on need or component accessability for replacement or repair, can be justified. 7. Sections 6.2.2, 6.2.3, 6.2.4, 6.2.5, 6.2.6, 6.3.2, 6.3.3, 6.3.4, and 6.3.5 of AtlS-4.5 pertain to variables and variable ranges for monitoring. In conjunction with the above sections, Tables 1, 2, and 3 of this regulatory guide (which include those parameters mentioned in the above sections) should be used in developing the minimum set of instruments and their respective ranges for accident-monitoring instrumentation for each nuclear power plant. 8. Section 6.4 of ANS-4.5 pertains to specific design criteria for accident-monitoring instrumentation. In conjunction with Section 6.4, the provisions as indicated in Table 1 of this regulatory guide should be used. D. IMPLEMENTATION This proposed revision has been released to encourage public,articipa-tion in its development. Except in those cases in which an applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the method to be described in the active guide reflecting public comments will be used in the evaluation of the following applications that are docketed after the implementation date to be specified in the guide: 1. Preliminary Design Approval (PDA) applications and Preliminary Duplicate Design Approval (PDDA) applications. 2. Final Design Approval, Type 2 (FDA-2), applications and Final Duplicate Design Approval, Type 2 (FDDA-2), applications. 3. Manufacturing License (ML) applications. 4. Construction Permit (CP) applications except for those portions of CP applications that reference standard designs (i.e., PDA, FDA-1, FDA-2, PDDA, FDDA-1, FDDA-2, or ML) or that reference } qualified base plant designs under the replication option. 1664 227 7
In addition, the NRC staff intends to implement part or all of this guide for all operating plants, plants under construction, all PDAs and FDAs, all I PDDAs and all FDDAs that may involve additions, elimination, or modification of structures, systems, or components of the facility af ter the construction Permit or design approval has been issued. All backfitting decisions in accordance with the positions stated in this guide will be determined by the staff on a case-by-case basis. The implementation date of this guide will in no case be earlier than April 15, 1980. 1664 228 I 8
Table 1 DESIGN CRITERIA 1 CRITERIA INSTRUMENTATION TYPES 2 A B C D E 1. Seis[nic qualification yes yes yes no no a per Regulatory Guide 1.100 2. Single failure criteria yes yes yes no no per Regulatory Guide 1.53 3. Environmental qualification yes yes yes yes no 4 5 per Regulatory Guide 1.89 4. Power source Emr6 CB7 CB7 Emr6 Emr6 5. Out-of-service interval 8 8 8 9 10 before accident 6. Portable no no no no no tl it 21 7. Quality assurance level 12 12 12 12 12 8. Display type 13 Cont 4 Con 14 Con 14 OD15 ODis 9. Display method Rec 16 Rec 17 Rect 7 Ind18 Ind s,19 i 10. Unique identification yes yes yes no no 11. Periodic testing per yes yes yes yes no 2o Regulatory Guide 1.118 1Unless different specifications are given in this regulatory guide, the specifications in ANSI N320-1979, " Performance Specifications for Reactor Emergency Radiological Monitoring Instrumentation," apply to the high-range containment area monitors, area exposure rate monitors in other buildings, effluent and environmental monitors, and portable instruments for measuring radiation or radioactivity. 2 Type A - Those instruments that provide information required to take preplanned manual actions. Type B - Those instruments that provide information to monitor the process of accomplishing critical safety functions. Type C - Those instruments that indicate the potential for breaching or the the actual breach of the barriers to fission product release. Type D - Those instruments that indicate the performance of individual safety systems. Type E - Those instruments that provide information for use in determining the magnitude of the release of radioactive materials and for continuously assessing such releases, for defense in depth, and for diagnosis. 3 Radiation monitors should meet the requirements of ANSI N320-1979, Section 5.14 and/or Section 9.1.15, as appropriate. 4See paragraph 6.3.6 of Draft Standard ANS-4.5. (Footnotes continued) 1664 229 9
Footnotes continued for Table 1 5 Qualified to the conditions of its operation and, for radiation monitors, ANSI N320-1979. 6 Emergency power source. I 7 Critical Instrument Bus - Class 1E Power. 8 Paragraph 4.11, " Exemption," of IEEE Standard 279-1971. 'JBased on normal Technical Specification require-ents on out-of-service for the safety system it serves. toNot necessary to include in the Technical Specifications unless specified by other requirements. 11 Radiation monitoring outside containment may be portable if so designated in Tables 2 and 3. 12 Level of quality assurance per Appendix B to 10 CFR Part 50. 13 Continuous indication or recording displays a given variable at all timet, intermittent indication or recording displays a given variable per'odically; on-demand indication or recording displays a given variable only wnen requested. 14 Continuous display. 15 Indication on demand. 1GWhere trend or transient information is essential to planned operator actions. 17 Recording. 18 Dial or digital indication. I'JEffluent release monitors require recording, including effluent radioactivity monitors, environs exposure rate monitors, and meteorology monitors. 20 Radiation monitors should be periodically tested in accordance with the requirements of ANSI N320-1979. 1664 230 9 8 10
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e Table 2 (Continued) Measured Variable Range Type Purpose REACTOR COOLANT SYSTEM (Continued) 0 to 120% ] sign B,D To provide indication that the core -12% to 12%))
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is being cooled. Flow Primary System Safety Closed-not closed B,D By these measurements the operator Relief Valve Positions knows if there is a path open for (including PORV and loss of coolant and that an event code valves) or Flow may be in progress. Through or Pressure in Relief Valve Lines Radiation Level in 10 pCf/cc to 10 Ci/cc C ANS-4.5, Section 6.3.2. Primary Coolant Water For early indication of fuel cladding failure and estimate of extent of damage. CONTAINMENT Containment Pressure 10 psia pressure B,C ANS-4.5, Sections 6.2.5, 6.3.3, to 3 times design 6.3.4, and 6.3.5. 2 for To indicate the integrity of the pressure concrete; 4 times primary or secondary system pres-design pressure sure boundaries; to indicate the for steel potential for leakage from the containment; to indicate integrity of the containment. Containment Atmos-40*F to 400*F E For indication of the performance phere Temperature of the containment cooling system and adequate mixing. Containment Hydrogen 0 to 10% B,C ANS-4.5, Sections 6.2.5 and 6.3.5. Concentration (capable of For indication of the need for and operating from to measure the performance of the 10 psia to containment hydrogen recombiner. maximum design 2 pressure ) 10esign flow - the maximum flow anticipated in normal operation. 2 Design pressure - that value corresponding to ASME code values that are obtained at or below code-allowable material design stress values. 1664 232 12
Table 2 (Continued) Measured Variable Range Type Purpose CONTAINMENT (Continued) Containment Isola-Closed-not closed 8,0 ANS-4.5, Section 6.2.5. tion Valve Position To indicate the status of contain-ment isolation and to provide information on the status of valves in process lines that could carry radioactive materials out of containment. Cantainment Sumo Narrow range (sump). 8,C ANS-4.5, Section 6.3.3. Water Level Wide range (bottom For indication of leakage within of containment to the containment and to ensure 600,000 gallon adequate inventory for performance level equivalent) of the ECCS. High-Range Contain-1 to 107 R/hr 8,C To help identify if an accident ment Area Radiation (60 kev to 3 MeV has degraded beyond calculated phctons with 220% values and to indicate its magni-accuracy for tude to determine action to photons of 0.1 to protect the public. 3 MeV)(107 R/hr for photons is approximately equivalent to 108 rads per hour for betas and photons] SECONDARY SYSTEMS Steam Generator From atomospheric D For indication of integrity of the Pressure pressure to 20% secondary system and an indication aDove safety value of capability for decay heat setting removal. Steam Genarator Level From tube sheet to 0 For indication of integrity of the separators secondary system and an indication of capability for decay heat removal. Auxiliary Feedwater 0 to 110% design D To indicate an adequate source of Flow flowl water to each steam generator upon loss of main feedwater. Main Feedwater Flow 0 to 110% design E To indicate an adequate source of flowl water to each steam generator. 1664 233 13
e Table 2 (Continued) Measured Variable Range Type Purpose "fENDARY SYSTEMS (Lontinued) Safety / Relief Valve Closed-not closed B,D To indicate integrity of secondary Positions or Main system (vis-a-vis pipe break). Steam Flow Radioactivity in 10-7 to 105 pCi/cc E To indicate leakage f rom the Condenser Air Xe-133 calibration primary to the secondary system Removal System and measure of noble gas release rate to atmosphere. Radioactivity in Efflu-10-7 to 105 Ci/cc E An indication of release from the ent from Steam Gener-(alternatively, secondary system and measure of ator Safety Relief amoient background noble gas release rate to atmos-Valves or Atmospheric to 2 Ci/sec/MWth) phere. Dump Valves Xe-133 calibration AUXILIARY SYSTEMS Containment Spray 0 to 110% design D For indication of system operatior. Flow flowl Flow in HPI System O to 110*. design D For indication of system operation. flow 1 Flow in LPI System 0 to 110*. design D For indication of system operation. flows Emergency Coolant Top to bott om D To determine the amount of water Water Storage Tank discharged by the ECCS. This Level provides indication of the nature of the accident, indication of the performance of the ECCS, and indi-cation of the necessity for operator action. Condensate Storage Plant specific 8,C (For those plants where the con-Tank Level densate storage tank is the prin-cipal source of auxiliary feedwater.) To ensure water supply for auxi-liary feedwater pumps. Accumulator Tank Top to bottom 0 To indicate whether the tanks have Level injected to the reactor coolant system. 1664 234 14
Table 2 (Continued) Measured Variable Range Type Purpose AUXILIARY SYSTEMS (Continued) Accumulator Isolation Closed-not closed D To indicate state of the isolation Valve Positions valves (per Regulatory Guide 1.47). RHR System Flow 0 to 110% design D For indication of system op(ration, flowa RHR Heat Exchanger 32*F to 350*F D For indication of system operation. Out Temperature Component Cooling 32*F to 200*F D For indication of system operation. Water Temperature Component Cooling 0 to 110% design D For indication of system operation. Water Flow flow 1 Flow in Ultimate 0 to 110% design D For indication of system operation. Heat Sink Loop flow! Temperature in Ulti-30*F to 150*F 0 For irdication of system operation. mate Heat Sink Loop Ultimate Heat Sink Plant specific 0 To ensure adequate source of Level cooling water. Heat Removal by the Plant specific B To indicate system operation. Containment Fan Coolers Boric Acid Charging 0 to 110% design 9 To provide indication of reactor Flow flow 1 ccoling and inventory control in order to maintain adequate concen-tration for shutdown margin. Letdown Flow 0 to 110% design D For indication of reactor coolant flowl inventory control and baron concen-tration control. Sump Level in Spaces To corresponding 0 To monitor environmental conditions of Equipment Required level of safety of equipment in closed spaces. for Safety equipment failure RADWASTE SYSTEMS High-level Radioactive Top to oottom E Available volume to store primary Liquid Tank Level coolant. .t 1664 235 n
9 Table 2 (Continued) Measured Variable Range Type Purpose RADWASTE SYSTEMS (Continued) Radioactive Gas Hold-O to 150% of E Available capacity to store waste 2 up Tank Pressure design pressure gases. VENTILATION SYSTEMS Emergency Ventilation Open-closed D To t ure proper ventilation under Damper Position status accident conditions. Temperature of Space 30*F to 180*F D To monitor environmental conditions in Vicinity of Equip-of equipment in closed spaces. ment Required for Safety POWER SUPPLIES Status of Class 1E Voltages and D To ensure an adequate source of Power Supplies and currents electric power for safety systems. Systems Status of Non-Class Voltages and E To indicate an adequate source of IE Power Supplies currents electric power. and Systems RADIATION EXPOSURE RATES { INSIDE BUILDINGS OR AREAS WHERE ACCESS IS REQUIRED TO SERVICE SAFETY-RELATED EQUIPMENT Radiation Exposure 10-1 to 104 R/hr E For measurement of high-range Rates for photons radiation exposure rates at various locations. (permanently installed monitors) AIRBORNE RADICACTIVE MATERIALS RELEASED FROM THE PLANT Effluent Radioactiv-(Normal plus E ANS-4.5, Section 6.2.6. ity - Noble Gases accident range To provide operator with informa-for noble gas) tion regarding release of radio-active noble gases on continuous . Containment 10-7 to 105 pCi/cc basis. Xe-133 calibration Provisions should be made to monitor all . Secondary 10-7 to 104 pCi/cc potential pathways for release of gaseous Containment Xe-133 calibration radioactive materials to the environs in (Reactor shield conformance with General Design Criterion 64. building annulus) Note: Monitoring of individual effluent streams only is required where such streams are released directly to 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 f has a range adequate to measure worst-case releases. 1664 236
9 Table 2 (Continued) Measured Variable Ranje Type Purpose AIRBORNE RADIOACTIVE MATERIALS RELEASE 0 FRCM THE PLANT (Continued) . Auxiliary Building 10-7 to 103 yCi/cc including buildings Xe-133 calibration containing primary system gases, e.g., waste gas decay tank .Jther Release 10-7 to 102 pCi/cc Points (including Xe-133 calibration fuel handling area if separate from (permanently auxiliary building) installed monitors) Effluent Radioactiv-10-3 to 102 pCi/cc E To provide the operator with ity - High-Range information regarding release of Radiohalogens and (pe rmanently radioactive halogens and particu-Particulates installed lates. Continuous collection monitors) of representative samples followed by monitoring (measurements) of samples for radiohalogens and for particulates. Environs Radioactiv-10-8 to 102 R/hr E ANS-4.5, Section 6.3.4. ity - Exposure Rate (60 kev to 3 MeV) For estimating release rates of radioactive materials released (oermanently during an accident from unidenti-installed fled release paths (not covered monitors) by effluent monitors) continuous readout capability. (Approximately 16 to 20 locations - site depen-dent.) Environs Radioactiv-10** to 10-3 pC1/cc E For estimating release rates of ity - Radiohalogens for both radio-radioactive materials released and Particulates halogens and during an accident from unidenti-particulates fied release paths (not covered by effluent monitors). Continuous (permanently collection of representative installed samples followed by monitoring samplers) (measurements) of the samples. (Approximately 16 to 20 locations - site dependent.) 1664 237 17
e Table 2 (Continued) Measured Variable Range Type Purpose AIRBORNE RADICAC!IVE MATERIALS RELEA5ED FROM THE PLANT (Continued) Plant and Environs High Range E During and following an accident, Radioactivity 0.1 to 104 R/hr to monitor radiation and airborne (pcrtable instruments) pnotons radioactivity concentrations in 0.1 to 104 rads /hr many areas throughout the facility betas and low-and the site environs where it is energy photons impractical to install stationary monitors capable of covering both normal and accident levels. 100-channel E During and following an accident, gamma-ray to rapidly scope the composition spectrometer of gamma emitting sources. POSTACCIDENT SAMPLING CAPABILITY h Primary Coolant As required based E ANS-4.5, Section 6.3.2. Sumps on Regulatory To provide means for safe and Containment Air Guide 1.4 guide-convenient sampling. These lines provisions should include: POSTACCIDENT ANALYSIS 1. gamma-ray E 1. Shielding to maintain radiation CAPABILITY (ON SITE) spectrum doses ALARA, 2. pH 2. Sample containers with container-3. hydrogen sampling port connector compati-4 oxygen
- bility, 5.
boron 3. Capability of sampling under primary system pressure and negative pressure, 4. Handling and transport capabi-lity, and 5. Pre arrangement for analysis and interpretation. METEOROLOGY Wind Direction 0 to,60* (15' E For determining effluent transport accu acy with a direction for emergency planning, deflection of 15* dose assessment, and source esti-Starting speed mates. 0.45 mos (1 mph)) 1664 238 G 18
Table 2 (Continued) Measured Variable Range Type Purpose METEOROLOGY (Continued) Wind Speed 0 to 30 mps (67 E For determining effluent travel mph) (10.22 mps speed and dilution for emergency (0.5 mph) accuracy planning, doses assessments, and for wind speeds source estimates. less that 11 mps (25 mph), with a starting thres-hold of les, than 0.45 mps (1 mph)) Temperature -60*F to 120*F E For determining nature of precipitation (10.9*F accuracy) and ground deposition for emergency planning. Vertical Temperature -9*F to +9*F E For determining ef fluent dif fusion Difference (10.3*F accuracy rates for emergency planning, dose per 16a-foot assessments, and source estimates. intervals) Precipitation Recording rain E For determining effluent transport gage with range and ground depositicn for emergency sufficient to planning. ensure accuracy of total accumu-lation within 10% of recorded value - 0.01" resolution ) I 19
Table 3 BWR VARIABLES Measured Variable Range Type Purpose CORE Core Exit Temperature 150'F to 2300*F 8,C To provide incore temperature measurements to identify localized hot areas. (Approximately 50 measurements) Control Rod Position Full in or not D To provide position indication that the full in control rids are fully inserted. (Minimum of 2 hours after accident) Neutron Flux 1 c/s to L% power 8 ANS-4.5, Section 6.2.2. (at least one For indication of approach to fission counter) criticality. REACTOR COOLANT SYSTEM RCS Pressure 15 osia to B,C ANS-4.5, Sections 6.2.3, 6.2.4, 2000 psig 6.3.3, and 6.3.5. For indication of an accident and to indicate that actions must be taken to mitigate an event. Coolant Level in the Bottom of core 8 ANS-4.5, Section 6.2.3. Reactor support plate For indication of fuel submergency to above top of for a LOCA event. discharge plenum Main Steamline Flow 0 to 120% design B To provide an indication of the ficwt integrity of the pressure boundary. Main Steamline Isola-O to 15" of water 8 To provide an indication of the tion valves' Leakage O to 5 psid pressure boundary and containment Control System Pressure Primary System Safety Closed-not closed 8,0 By these measurements, the operator Relief Valve Posi-or knows if there is a path open for tions, including 0 to 50 psig loss of coolant and if an event ADS or Flow Through may be in progress. or Pressure in Valve Lines 10esign flow - the maximum flow anticipated in normal operation. ) 2 21 s
9 Table 3 (Continued) Measured Variable Range Type Purpose REACTOR COOLANT SYSTEM (Continued) Radiation Level in 10 pCi/cc to C ANS-4.5, Section 6.3.2. Coolant 10 Ci/cc For early indication of fuel cladding failure and estimate of extent of damage. CONTAINMENT Primary Containment 10 psia pressure B,C ANS-4.5, Sectionx 6.2.5, 6.3.3, Pressu*e to 3 times design 6.3.4, and 6.3.5. 2 pressure for con-For indication of the integrity crete; 4 times of the primary containment pressure design pressure boundary; to indicate the potential for steel for leakage from the containment. Containment and Drywell 0 to 10% 8,C ANS-4.5, Sections 6.2.5 and 6.3.5. Hydrogen Concentration (capability of For indication of the need for and operating from a measurement of the performance 12 psia to maxi-of the containment hydrogen recom-mum design biner and to verify the operation of 2 pressure ) the mixing system. Containment and Drywell O to 10% B,C For indication of the need for and a Oxygen Concentration (capability of measurement of the performance of (for those plants operating from the containment oxygen elimination with inerted 12 osia to system. 2 containments) design pressure ) Primary Containment Closed-not closed 8,0 ANS-4.5, Section 6.2.5. Isolation valve To indicate the status of containment Position isolation and to provide information on the status of valves in process lines that could carry radioactive materals out of containment. Suppression Pool Top of vent to B ANS-4.5, Section 6.3.3. Water Level top of weir well Suppression Pool 50'F to 250'F 8 To ensure proper temperature for Water Temperature NPSH of ECCS. To verify the opera-tion of the makeup system. Drywell Pressure 12 psia to 3 psig B ANS-4.5 Section 6.3.3. O to 110% design E Diagnosis of impact of accident on 2 pressure drywell structure. 20esign pressure - that value corresponding to ASME code values that are obtained at or below code-allowable material design stress values. 1664 241 22
Table 3 (Continued) Measured Variable Range Type Purpose CONTAINMENT (Continued) Orywell Drain Sumps Bottom to top B,C ANS-4.5, Section 6.3.3. Level (Identified and Unidentified Leakage) High-Range Contain-1 to 107 R/hr 8,C To help identify if an accider,it ment Area Radiation (60 kev to 3 MeV has degraded beyond calculated photons with 120% values and to indicate its accuracy for pho-magnitude in order to determine tons of 0.1 to action to.otect the public. 3 MeV) [107 R/hr for photons is approximately equivalent to 108 rads /hr for betas and photons] POWER CONVERSION SYSTEMS Main Feedwater Flow 0 to 110% design E To indicate an adequate source of flowl water to the reactor. Condensate Storage Bottom to top E To indicate available water for Tank Level core cooling. AUXILIARY SYSTEMS Containment Spray 0 to 110% design D for indication of system operation. Flow flowl Steam Flow to RCIC 0 to 110% design E To verify that adequate steam is flow 1 available for the system to perform its function. RCIC Flow 0 to 110% design D For indication of system operation. flowl RHR System Flow 0 to 110% design D for indication of system operation. flowl RHR Heat Exchanger 32*F to 350*F D For indication of system operation. Outlet Temoerature Service Cooling 32*F to 200*F 0 For indication of system operation. Water Temperature 1664 242 23
s e Table 3 (Continued) Measured Variable Range Type Purpose AUXILIARY SYSTEMS (Continued) Service Cooling 0 to 110% design D For indication of system aperation. Water Flow flow 1 Flow in Ultimate O to 110% design D For indication of system operation. Heat Sink Loop flowl Temperature in Ulti-30*F to 150*F 0 For indication of system operation. mate Heat Sink Loop Ultimate Heat Sink Plant specific D To ensure adequate source of cool-Level ing water. SLCS Storage Tank Bottom to top E To provide indication of inventory Level for boron injection for sautdown. Sump Level in Spaces To corresponding D To monitor potential for failure of of Equipment Required level of safety equipment in closed spaces due to for Safety equipment failure flooding. RADWASTE SYSTEMS High Radioactivity Top to bottom E A<ailable volume to store primary Liquid Tank Level coolant. Charcoal Delay Gas As required E To monitor performance of system. System Gas Flow or Radioactivity Level VENTILATION SYSTEMS Emergency Ventilation Open-closed status D To ensure proper ventilation under Damper Position accident conditions. Temperature of Space 30'F to 130*F B To monitor environmental conditions in Vicinity of Equip-of equipment in closed soaces. ment Required for Safety POWER SUPPLIES Status of Class 1E Voltages and 0 To ensure an adequate source of Power Supplies and currents electric power for safety systems. Systems Status of Non-Class Voltages and E To indicate an adequate source of IE Power Supplies currents electric power. and Systems 1664 243 a
9 Table 3 (Contir.ued) Measured Variable Range Type Purpose RADIATION EXPOSURE RATES INSIDE BUILDINGS CR AREAS WHERE ACCESS IS REQUIRED TO SERVICE SAFETY-RELATED EQUIP-MENT Radiation Exposure 10-" to 10* R/hr E For measurement of high-range Rates for photons radiation exposure rates at various locations. (permanently installed moni-tors) AIRBORNE RADI0 ACTIVE MATERIALS RELEASED FRCM THE PLANT Effluent Radioactiv-(Normal plus E ANS-4.5, Section 6.2.6 ity - Noble Gases accident range To provide operator with informa-for noble gas) tion regarding release of radio-active noble gas on a continuous . Containment 10-7 to 105 pCi/cc basis. Exhaust Vent and Xe-133 calibration Standby Gas Provisions should be made to monitor all Treatment System potential pathways for release of gaseous Vent radioactive materials to the environs in conformance with General Design Criterion 64. .0ther Release 10-7 to 102 pCi/cc Note: Monitoring of individual effluent Points (including Xe-133 calibration streams only is required where such fuel handling . streams are released directly to building, auxili-(permanently the environment. If two or more ary building, and installed moni-streams are combined prior to turbine building) tors) 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. Effluent Radioactiv-10-3 to 102 pCi/cc E To provide the operator with ity - High-Range information regarding release of Radiohalogens and (permanently radioactive halogens and particu-Particulates installed moni-lates. Continuous collection of tors) representative samples followed by monitoring (measurements) of samples for radiohalogens and for particulates. 1664 244 25
S Table 3 (Continued) Measured Variable Range Type Purpose AIRBORNE RADI0 ACTIVE MATERIALS RELEASED FROM THE PLANT (Continued) Environs Radioactiv-10-8 to 102 R/hr E For estimating release rates of ity - Exposure Rate (60 kev to 3 MeV) radioactive materials released during an accident from unidentified (permanently release paths (not covered by installed moni-effluent monitors) - continuous tors) readout capability. Approximately 16 to 20 locations - site depen-dent.) Environs Radioactiv-10-9 to 10-3 pCf/cc E For estimating release rates of ity - Radiohalogens for both radiohalo-radioactive materials released and Particulates gens and particu-during an accident from unidenti-lates fied release paths (not covered by effluent monitors). Continuous (permanently collection of representative installed moni-samples followed by monitoring tors) (measurements) of the samples. (Approximately 16 to 20 locations.) Plant and Environs High Range E Ouring and follcwing an accident. Radioactivity 0.1 to 104 R/hr to monitor radiation and airborne (portable instruments) photons radioactivity concentrations in 0.1 to 104 many areas throughout the facility rads /hr betas where it is impractical to install and low energy stationary monitors capable of photons covering both normal and accident levels. 100-channel E During and following an accident, gamma-ray to rapidly scope the composition spectrometer of gamma emitting sources. POSTACCICENT SAMPLING CAPABILITY Primary Ccolant As required based E ANS-4.5, Section 6.3.2. Suppression Pool on Regulatory Guide To provide means for safe and Containment Air 1.3 guidelines convenient sampling. These Standby Gas Treatment provisions should include: System (upstream of each area served) 1664 26 I 26
Table 3 (Continued) Measured Variable Range Type Purpose 1. Shielding to maintain radiation doses ALARA, 2. Sample containers with con-POSTACCIDENT ANALYSIS 1. gamma-ray E tainer sampling port connector CAPABILITY (ON SITE) spectrum compatibility, 2. pH 3. Capability of sampling under 3. hydrogen primary system pressure and 4. Oxygen negative pressure, 4. Handling and transport capability, and 5. Pre-arrangement for analysis and interpretation. METECROLOGY Wind Direction 0 to 360* (15* E For determining effluent transport accuracy with a direction for emergency planning, deflection of 15* dose assessment, and source esti-Starting speed mates. 0.45 mps (1 mph)) S Wind Speed 0 to 30 mps E For determining effluent travel (67 mph) (10.22 speed and dilution for emergency mos (0.5 mph) planning, dose assessments, and accuracy for wind source estimates. speeds less than 11 mps (25 mph), with a starting threshold of less than 0.45 mps (1 mph)) Temperature -60*F to I20*F E For determining nature of precipitation (10.9*F ac:uracy) and ground deposition for emergency planning. Vertical Temoerature -9*F to +9*F E For determining effluent diffusion Offference (10.3*F accuracy rates for emergency planning, dose per 164-foot assessments, and source estimates. intervals) Precipitation Recording rain E For determining effluent transport gage with range and ground deposition for emergency sufficient to planning. ensure accuracy of total accumu-lation within 10% of recorded value - 0.01" resolution 1664 246 27 s
DRAFT VALUE/ IMPACT STATEMENT C 1. PROPOSED ACTION 1.1 Description The applicant (licensee) of a nuclear power plant is required by the Commission's regulations to provide instrumentation to (1) monitor variables and systems for accident conditions as appropriate to ensure adequate safety and (2) monitor the react (? 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.
- 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. This report, along with the draft Task Action Plan A-34 report; Draft 1 of Regulatory Guide 1.97, dated April 12, 1974; and Standard ANS-4.5, Draft 4, dated November 1979, provides ample basis for revising Regulatory Guide 1.97. 28 4[
- 1. 3 Value/ Impact of the Proposed Action S
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 draft national standard (ANS-4.5). For future plants, the staff review will be simplified with guidance contained in the endorsed industry standard and the regulatory guide, which includes a list of variables for accident monitoring. Consequently, there will be no significant impact on the staff. There will, however, be effort required to review each operating plant and plant under review to determine the extent of backfitting required. This will be done on a case-by-case basis. 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 following an accident and, therefore, reduces uncertainty as to what the staff considers acceptable in the area of accident monitoring. Most of the impact on industry will be in the area of providing instrumentation to indicate the potential breach and the actual breach of the barriers to radioactivity release, i.e., fuel cladding, reactor coolant pressure boundary, and containment. There will be some impact due to a heretofore unspecified variable to be monitored (i.e., water level in reactor) that has been identified during the evaluation l of TMI-2 experience and will require development. 1664 248 29 i
A cost estimate of the impact on industry for future plants has not yet been made but will be developed by the staff, with industry input, during the comment period. The staff intends to meet with the various owners' groups and determine, on a case-by-case basis, the cost impact on each individual operating plant and plant under review as it determines the extent of backfitting in each case. 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 oublic. 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 I
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, 30 J
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 the variables to be monitored on site by 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.
SUMMARY
AND CONCLUSIONS The proposed revision to Regulatory Guide 1.97, " Instrumentation For Light-Water-Coolea Nuclear Power Plants to Assess Plant and Environs Conditions During and following an Accident," should be issued. 1664 250 l 31
UNIT?D STATES NUCLEAR REGULATORY CouMISSION F l WASHINGTON, D. C 20555 postmotApoFEESPAto OFFICI AL SUhlNESS u5 NULLE AR mgGULATORY PE N ALTY FOR PRIVATC USE. S300 COMusssioN LmJ 120555006386 2 SN_ US hRON SECY PUBritC COCUMENT-ROGM LRANCH CHIEF WASHIhGICN N-DC 20555 s. 1664 251 8 1}}