Regulatory Guide 1.97
| ML061580448 | |
| Person / Time | |
|---|---|
| Issue date: | 06/30/2006 |
| From: | Office of Nuclear Regulatory Research |
| To: | |
| Marcus, B.S., Wilson A.A., Tartal G.M. | |
| References | |
| DG-1128 RG-1.097, Rev 4 | |
| Download: ML061580448 (10) | |
The U.S. Nuclear Regulatory Commission (NRC) issues regulatory guides to describe and make available to the public methods that the NRC staff considers acceptable foruse in implementing specific parts of the agency's regulations, techniques that the staff uses in evaluating specific problems or postulated accidents, and data that the staffneed in reviewing applications for permits and licenses. Regulatory guides are not substitutes for regulations, and compliance with them is not required. Methods andsolutions that differ from those set forth in regulatory guides will be deemed acceptable if they provide a basis for the findings required for the issuance or continuance ofa permit or license by the Commission.This guide was issued after consideration of comments received from the public. The NRC staff encourages and welcomes comments and suggestions in connection withimprovements to published regulatory guides, as well as items for inclusion in regulatory guides that are currently being developed. The NRC staff will revise existing guides,as appropriate, to accommodate comments and to reflect new information or experience. Written comments may be submitted to the Rules and Directives Branch, Officeof Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.Regulatory guides are issued in 10 broad divisions: 1, Power Reactors; 2, Research and Test Reactors; 3, Fuels and Materials Facilities; 4, Environmental and Siting;5, Materials and Plant Protection; 6, Products; 7, Transportation; 8, Occupational Health; 9, Antitrust and Financial Review; and 10, General.Requests for single copies of draft or active regulatory guides (which may be reproduced) should be made to the U.S. Nuclear Regulatory Commission, Washington, DC 20555,Attention: Reproduction and Distribution Services Section, or by fax to (301) 415-2289; or by email to Distribution@nrc.gov. Electronic copies of this guide and other recentlyissued guides are available through the NRC's public Web site under the Regulatory Guides document collection of the NRC's Electronic Reading Room athttp://www.nrc.gov/reading-rm/doc-collections/ and through the NRC's Agencywide Documents Access and Management System (ADAMS) athttp://www.nrc.gov/reading-rm/adams.html, under Accession No. ML061580448.
U.S. NUCLEAR REGULATORY COMMISSION
Revision 4 June 2006 REGULATORY GUIDE
OFFICE OF NUCLEAR REGULATORY RESEARCH
REGULATORY GUIDE 1.97 (Draft was issued as DG-1128, dated June 2005
)CRITERIA FOR ACCIDENT MONITORING INSTRUMENTATION
FOR NUCLEAR POWER PLANTS
A. INTRODUCTION
The U.S. Nuclear Regulatory Commission (NRC) developed this regulatory guide to describe a method that the NRC staff considers acceptable for use in complying with the agency's regulations with respect tosatisfying criteria for accident monitoring instrumentation in nuclear power plants. Specifically, the methoddescribed in this regulatory guide relates to General Design Criteria 13, 19, and 64, as set forth in Appendix A
to Title 10, Part 50, of the Code of Federal Regulations (10 CFR Part 50), "Domestic Licensing of Production and Utilization Facilities":*Criterion 13, "Instrumentation and Control," requires operating reactor licensees to provide instrumentationto monitor variables and systems over their antic ipated ranges for accident conditions as appropriateto ensure adequate safety.*Criterion 19, "Control Room," requires operating reactor licensees to provide a control room from whichactions can be taken to maintain the nuclear pow er unit in a safe condition under accident conditions,including loss-of-coolant accidents (LOCAs). In addition, operating reactor licensees must provide equipment (including the necessary instrumentation), at appropriate locations outside the control room,with a design capability for prompt hot shutdown of the reactor.*Criterion 64, "Monitoring Radioactivity Releases," requires operating reactor licensees to provide the meansfor monitoring the reactor containment atmosphere, spaces containing components to recirculate LOCA
fluids, effluent discharge paths, and the plant environs for radioactivity that may be released as a result of postulated accidents.
1IEEE publications may be purchased from the IEEE Service Cent er, which is located at 445 Hoes Lane, Piscataway, NJ 08855 [
http://www.ieee.org, phone (800) 678-4333].
2The terms "new nuclear power plant" and "new plant" refer to any nuclear power plant for which the licensee obtainedan operating license after the NRC issued Revision 4 of Regulatory Guide 1.97. The terms "current operating reactor"and "current plant" refer to any nuclear power plant for which the licensee obtained an operating license beforethe NRC issued Revision 4 of Regulatory Guide 1.97.
3 Copies are available at current rates from the U.S. Go vernment Printing Office, P.O. Box 37082, Washington, DC20402-9328 [telephone (202) 512-1800], or from the National Technical Information Service (NTIS), 5285 Port Royal Road, Springfield, Virginia 22161 [
http://www.ntis.gov, telephone (703) 487-4650]. Copies are available forinspection or copying for a fee from the NRC's Public Docu ment Room (PDR), which is located at 11555 RockvillePike, Rockville, Maryland; the PDR's mailing address is USNRC PDR, Washington, DC 20555-0001. The PDRcan also be reached by telephone at (301) 415-4737 or (800) 397-4205, by fax at (301) 415-3548, and by email
to PDR@nrc.gov
.RG 1.97, Rev. 4, Page 2In addition, Subsection (2)(xix) of 10 CFR 50.34(f), "Additional TMI-Related Requirements,"
requires operating reactor licensees to provide adequate instrumentation for use in monitoring plantconditions following an accident that includes core damage.This revision of Regulatory Guide 1.97 represents an ongoing evolution in the nuclear industry's thinking and approaches with regard to accident monitoring systems for the Nation's nuclear powerplants. Specifically, this revision endorses (with certain clarifying regulatory positions specified in Section C of this guide) the "IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations," which the Institute of Electrical and Electronics Engineers (IEEE)promulgated as IEEE Std. 497-2002.
1This revised regulatory guide is intended for licensees of new nuclear power plants.
2 Previousrevisions of this regulatory guide remain in effect for licensees of current operating reactors, 2 who areunaffected by this revision. (See the discussion of regulatory position #1 in Section C of this guideregarding the applicability of IEEE Std. 497-2002 for current operating reactors.)In general, information provided by regulatory guides is reflected in the NRC's "StandardReview Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (NUREG-0800).
3 The NRC's Office of Nuclear Reactor Regulation (NRR) uses the Standard Review Plan (SRP) to review
applications to construct and opera te nuclear power plants. Chapter 7, "Instrumentation and Controls,"
and its Branch Technical Position HICB-10, "Guidance on Application of Regulatory Guide 1.97,"of the SRP will require updates for consistency with this revision of Regulatory Guide 1.97.Any information collections mentioned in this regulatory guide are established as requirementsin 10 CFR Part 50, which provides the regulatory basis for this guide. The Office of Managementand Budget (OMB) has approved those information collection requirements under OMB control number3150-0011. The NRC may neither conduct nor sponsor , and a person is not required to respond to,a request for information or an information collection requirement unless the requesting documentdisplays a currently valid OMB control number.
4Copies may be obtained from the American Nuclear Society, which is located at 555 North Kensington Avenue, La Grange Park, Illinois 60525 [
http://www.ans.org, phone (708) 352-6611].
5IEEE publications may be purchased from the IEEE Service Cent er, which is located at 445 Hoes Lane, Piscataway, NJ 08855 [
http://www.ieee.org, phone (800) 678-4333].
6 Copies are available at current rates from the U.S. Go vernment Printing Office, P.O. Box 37082, Washington, DC20402-9328 [telephone (202) 512-1800], or from the National Technical Information Service (NTIS), 5285 Port Royal
Road, Springfield, Virginia 22161 [
http://www.ntis.gov, telephone (703) 487-4650]. Copies are available forinspection or copying for a fee from the NRC's Public Docu ment Room (PDR), which is located at 11555 RockvillePike, Rockville, Maryland; the PDR's mailing address is USNRC PDR, Washington, DC 20555-0001. The PDR canalso be reached by telephone at (301) 415-4737 or (800) 397-4205, by fax at (301) 415-3548, and by email
to PDR@nrc.gov
.RG 1.97, Rev. 4, Page 3
B. DISCUSSION
In the aftermath of the accident at Three Mile Island, Unit 2 (TMI-2), in 1979, the United Statesadopted a more rigorous approach for accident monitoring systems, which resulted in three major sourcesof related requirements:(1)ANSI/ANS-4.5-1980, "Criteria for Accident Monitoring Functions in Light-Water-Cooled Reactors,"
4 delineated criteria for determining the variables that the control room operatorshould monitor to ensure safety during an accident and the subsequent long-term stable shutdownphase. The American National Standards Institute (ANSI) promulgated this standard, which wasdeveloped by the American Nuclear Society (ANS) Standards Committee, Subcommittee ANS-4,Writing Group 4.5. In so doing, ANSI and ANS sought to address (1) instrumentation that permits operators to monitor expected parame ter changes during an accident, and (2) extended-range instrumentation deemed appropriate for previously unforeseen events. As the source forspecific instrumentation design criteria, ANS
I/ANS-4.5-1980 referenced the draft IEEE Std. 497-1977, "IEEE Trial-Use Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations,"
5 which IEEE subsequently issued as IEEE Std. 497-1981.
5(2)IEEE Std. 497-1981, "IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations," provided the relevant instrumentation design criteria.(3)Revision 3 of Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Cond itions During and Following an Accident,"
6 datedMay 1983, prescribed a detailed list of variables to monitor, and specified a comprehensive listof design and qualification criteria to be met.Given its prescriptive nature, Revision 3 of Regulatory Guide 1.97 quickly became the de factostandard for accident monitoring, and both ANSI/
ANS-4.5-1980 and IEEE Std. 497-1981 fell out of useand were subsequently withdrawn as active standards. Nonetheless, Revision 3 of Regulatory Guide 1.97has become outdated, in that it does not provide criteria for advanced instrumentation system designsbased on modern digital technology. Revision 3 also does not address the need for technology-neutral guidance for licensing new plants. In addition, the guidance should be less prescriptive and based onthe accident management functions of the individual variable types.
7IEEE publications may be purchased from the IEEE Service Cent er, which is located at 445 Hoes Lane, Piscataway, NJ 08855 [
http://www.ieee.org, phone (800) 678-4333].
8 Copies are available at current rates from the U.S. Go vernment Printing Office, P.O. Box 37082, Washington, DC20402-9328 [telephone (202) 512-1800], or from the National Technical Information Service (NTIS), 5285 Port Royal Road, Springfield, Virginia 22161 [
http://www.ntis.gov, telephone (703) 487-4650]. Copies are available forinspection or copying for a fee from the NRC's Public Docu ment Room (PDR), which is located at 11555 RockvillePike, Rockville, Maryland; the PDR's mailing address is USNRC PDR, Washington, DC 20555-0001. The PDRcan also be reached by telephone at (301) 415-4737 or (800) 397-4205, by fax at (301) 415-3548, and by email
to PDR@nrc.gov
.RG 1.97, Rev. 4, Page 4With the increased use of digital instrumentation systems in advanced nuclear power plantdesigns, the nuclear industry came to recognize a need to develop a consolidated standard that was moreflexible than Revision 3 of Regulatory Guide 1.97. Instead of prescribing the instrument variables to bemonitored (as was the case in Revision 3), the industry recognized the advantage of providing performance-based criteria for use in selecting variables. Similarly, rather than providing design andqualification criteria for each variable category, the industry sought to standardize the criteria based onthe accident management functions of the given ty pe of variable. These efforts resulted in thedevelopment of IEEE Std. 497-2002, "IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations,"
7 by the IEEE Power Engineering Society, Nuclear PowerEngineering Committee, Subcommittee 6, Working Group 6.1, "Post-Accident Monitoring."
Unlike its predecessor, IEEE Std. 497-2002 establishes flexible, performance-based criteria forthe selection, performance, design, qualification, display and quality assurance of accident monitoringvariables. As such, these variables are the operators' primary sources of accident monitoring information. In some instances, additional variables which provide backup or diagnostic information may exist;
however, these backup and diagnostic va riables, which are not considered primary sources of information, need not be classified in accordance with the variable types in IEEE Std. 497-2002, and they need notmeet the criteria in this guide.
Clause 8.1.2 of IEEE Std. 497-2002 cites several industry codes and standards for human factorscriteria. The NRC provides additional guidance in NUREG-0700, "Human-System Interface Design Review Guideline: Review Methodology and Procedures"
8; NUREG-0711, "Human Factors EngineeringProgram Review Model"
8; and Chapter 18, "Human Factors Engineering,"of the NRC's Standard Review Plan (NUREG-0800).
8
9 Copies are available at current rates from the U.S. Go vernment Printing Office, P.O. Box 37082, Washington, DC20402-9328 [telephone (202) 512-1800], or from the National Technical Information Service (NTIS), 5285 Port Royal Road, Springfield, Virginia 22161 [
http://www.ntis.gov, telephone (703) 487-4650]. Copies are available forinspection or copying for a fee from the NRC's Public Docu ment Room (PDR), which is located at 11555 RockvillePike, Rockville, Maryland; the PDR's mailing address is USNRC PDR, Washington, DC 20555-0001. The PDRcan also be reached by telephone at (301) 415-4737 or (800) 397-4205, by fax at (301) 415-3548, and by email
to PDR@nrc.gov
.RG 1.97, Rev. 4, Page 5 Clause 6.2 of IEEE Std. 497-2002 states, in part, that the use of identical software in redundantinstrumentation channels is acceptable, provided that the licensee conducts an analysis to demonstratedefense-in-depth against common-mode software fa ilure. The NRC provides related guidance in Branch Technical Position HICB-19, "Guidance for Evaluation of Defense-in-Depth and Diversity in DigitalComputer-Based Instrumentation and Control Systems,"
9 as detailed in Chapter 7 of the NRC's Standard Review Plan (NUREG-0800).In addition, IEEE Std. 497-2002 includes two informative annexes:*Annex A provides general guidance regarding "Accident Monitoring Instrument ChannelAccuracy." In that annex, Clause A.2 provides guidance on accuracy requirement groupings according to how control room personnel should use the displayed functions, while Clause A.3provides typical accuracy requirements. Specifically, Clause A.3 states, in part, "Historically, the required accuracy for instrument channels relied upon to monitor containment pressureand hydrogen concentration has been +/-10 percent of full span." However, the NRC staff notesthat this example may not be applicable to all nuclear power plants. Traditionally, the requiredaccuracy of accident monitoring instrument channels is established based on the assignedfunction and the plant's safety analysis and licensing basis.*Annex B, "Bibliography," lists the references cite d in the standard, and provides sufficient detailfor users to obtain further information re garding specific aspects of the standard.
RG 1.97, Rev. 4, Page 6
C. REGULATORY POSITION
This regulatory guide endorses IEEE Std. 497-
2002, "IEEE Standard Criteria for AccidentMonitoring Instrumentation for Nuclear Power Generating Stations," as an acceptable methodfor providing instrumentation to monitor variabl es for accident conditions, subject to the followingregulatory positions:
(1)If a current operating reactor licensee voluntar ily converts to the criteria in Revision 4 of this guide, the licensee should perform the conversion on the plant's entire accident monitoring program to ensure a complete analysis.
If the licensee voluntarily uses the criteria in Revision 4 of this guide to perform m odifications that do not involve a conversion, the licensee should first perform an analysis to dete rmine the complete list of accident monitoring variables and their associated types in accor dance with the selection criteria in Revision 4.Regulatory position #1 clarifies the applicability of IEEE Std. 497-2002 for current operating reactors. Clause 1.1 of IEEE Std. 497-2002 states th at the standard is intended for new plants,although current plants may find its guidance useful in performing design-basis evaluationsor implementing design modifications. Having carefully considered the applicability and usefulness of the new standard, the NRC staff recognizes that current operating reactors could be interested in converting to Revision 4. In this context, conversion means adapting the plant's entire accident monitoring program from a given plant's current licensing basis (namely Revision 2 or 3 of this guide), to the guidance in Revisi on 4 of this guide. This adaptation could includephysical changes (e.g., replacing an instrument), licensing changes (e.g., technical specification changes), or both for each variable. The staff al so recognizes that Revisions 3 and 4 of this guidediffer in several ways, including variable type definitions and associated criteria, removalof design and qualification categories, removal of prescriptive tables of monitored variables,analysis required to produce the necessary design-basis documentation, and related changesin licensing basis and/or commitments. These differences could involve modifications to existinginstrumentation and could have significant cost implications for current operating reactor licensees who decide to convert to the new standard under Revision 4 of this guide.Licensees of current operating reactors could also be interested in voluntarily performingmodifications based on Revision 4 of this guide. For these modifications, the licensee should firstperform an analysis to determine the complete list of variables and their associated types in accordance with the selection criteria in Revision 4. Without such analysis, there is no means to correlate Revision 4 criteria being applied to the modification of variables that have been licensed to the criteria in Revisions 2 or 3.Revision 4 is primarily intended for licensees of new nuclear power plants. However, the NRCstaff sees no technical reason to prohibit a current operating reactor licensee from voluntarily
using the new guidance for conversion or modifications.
(2)Modify the first sentence in the second paragraph of Clause 6.7, as follows:
"Means shall be provided for validating instrument calibration during the accident."Regulatory position #2 modifies the requireme nt of IEEE Std. 497-2002, as it relatesto instrumentation calibration during an accide nt. Clause 6.7 of IEEE Std. 497-2002 requireslicensees to provide the means to calibrate instrumentation during an accident, and Clause 6.11requires licensees to consider the selection and location of instrumentation with respect to potential inaccessibility during an accident. Plants should strategically locate instruments to ensure that they are readily accessible for mainte nance. However, the NRC staff recognizes thatsome instruments (e.g., in-line sensors and area monitors) must be located in areas that are notaccessible during an accident. Furthermore, recalibration is one of the four methods stated in Clause 6.7, but the only method of "maintaining" instrument calibration. In many situations,
10 Copies are available at current rates from the U.S. Go vernment Printing Office, P.O. Box 37082, Washington, DC20402-9328 [telephone (202) 512-1800], or from the National Technical Information Service (NTIS), 5285 Port Royal Road, Springfield, Virginia 22161 [
http://www.ntis.gov, telephone (703) 487-4650]. Copies are available forinspection or copying for a fee from the NRC's Public Docu ment Room (PDR), which is located at 11555 RockvillePike, Rockville, Maryland; the PDR's mailing address is USNRC PDR, Washington, DC 20555-0001. The PDRcan also be reached by telephone at (301) 415-4737 or (800) 397-4205, by fax at (301) 415-3548, and by email
to PDR@nrc.gov
.RG 1.97, Rev. 4, Page 7it is not possible to recalibrate instrumentation during an accident due to environmental conditionsat the instrument location. The other three methods stated in Clause 6.7 cannot be usedto maintain instrument calibration, but rather can be used to verify that the instrument has notexcessively deviated from calibration. Consequently, licensees should provide means for validating instrument calibration during the accident.
(3)The range criteria for Type C variables (paragr aph 2 of Clause 5.1) should include the basis for the expanded ranges as follows:
"The range for Type C variables shall encompass thos e limits that would indicate a breach in a fission product barrier. These variables shall have expanded r anges and a source term that consider a damagedcore (see NUREG-0660). For example, ..."Regulatory position #3 clarifies the requirement to provide expanded ranges for Type C variables,which Clause 4.3 of IEEE Std. 497-2002 describes as those "that provide the most direct indication of the integrity of the three fission product barriers and provide the capability for monitoring beyond the normal operating range."
Clause 5.1 of the standard adds, "the rangefor Type C variables shall encompass, with margin, those limits that would indicate a breach in a fission product barrier." In a related pr ovision in 10 CFR 50.34(f)(2)(xix), the NRC requireslicensees to provide instrumentation to monito r plant conditions following an accident thatincludes core damage. The underlying basis for this regulation, documented in NUREG-0660,"NRC Action Plan Developed as a Result of the TMI-2 Accident,"
10 was that licensees shouldprovide instrumentation "with expanded ranges and a source term that considers a damaged corecapable of surviving the accident environment in which it is located for the length of time its function is required." To include the basis for the expanded range (from NUREG-0660),licensees should modify the range criteria for Type C variables (paragraph 2 of Clause 5.1),
as stated in regulatory position #3.
(4)Modify the last sentence in Clause 4.1 as follows:
"Type A variables include those variables that are a ssociated with contingency actions that are within the plant licensing basis and may be identified in written procedures."
Modify the last sentence in Clause 1.3, as follows:
"This standard also does not apply to instrumentati on required to support plant shutdown from outside the control room."Regulatory position #4 modifies the application of the term "contingency actions," which Clause 3.6 of IEEE Std. 497-2002 defines as "alternative ac tions taken to address unexpected responsesof the plant or conditions beyond its licensing basis (for example, actions taken for multipleequipment failures)." Clause 1.3 uses this term in defining th e application of IEEE Std. 497-2002, while Clause 4.1 uses it in defining selection criteria for Type A variables. The staff agrees with thecriteria in these clauses, except where they exclude contingency actions. Contingency actions wereexcluded from the scope of Revision 3 of this guide, but neither Revi sion 3 nor its endorsedstandard provided a definition of the term "contingency action." NSSS vendors have not usedthis term consistently in EPGs for current plant designs and, therefore, the staff recommendsconsidering contingency actions in accordance with the modified criteria in Clause 4.1. Furthermore, Revision 3 provided a prescriptive list of variables to monitor, whereas this revision RG 1.97, Rev. 4, Page 8provides a non-prescriptive, performance-based appro ach to variable selection. Thus, in thisperformance-based guide, the staff cannot endorse the carte blanche exclusion of contingencyactions from the selection criteria (especially those associated with plant-specific operatingprocedures or guidelines). Rather, the scope of instruments that could potentially be selected for accident monitoring (based on the selection criteria) should initially be as encompassing
as possible. Then, in the process of selecting the actual list of variables to be monitored,licensees could screen out instruments associated with contingency actions that take placebeyond the plant's licensing basis.
(5)The number of measurement points should be suffic ient to adequately indicate the variable value.Regulatory position #5 provides guidance concerning the number of measurement points for each variable, which IEEE Std. 497-2002 does not mention (with the exception of redundancyrequirements). In general, the number of measurement points should be sufficient to adequatelyindicate the variable value (e.g., containment temperature may require spatial distribution of several measurement points).
(6)If the NRC's regulations incorporate an i ndustry code or standard referenced in Clause 2 of IEEE Std. 497-2002, licensees and applicants must comply with that code or standard as set forth in the regulations. Similarly, if the NRC staff has endorsed a referenced codeor standard in a regulatory guide, that code or standard constitu tes an acceptable method for use in meeting the related regulatory requiremen t as described in the regulatory guide(s).
By contrast, if a referenced code or standard has neither been incorporated into the NRC's regulations nor been endorsed in a regul atory guide, licensees and applicants may consider and use the information in the referenced c ode or standard, if appr opriately justified, consistent with current regulatory practice.
(7)Modify paragraph (c) of Clause 5.4, as follows:
"The operating time for Type C variable instrument channels shall be at least 100 days or the duration for which the measured variable is required by the plant's LBD."Regulatory position #7 modifies the required instrument duration for Type C variables from"at least 100 days" to include cases where the plant's LBD defines a different operating time. The plant's LBD provides an appropriate basis for determining the operating time for Type C
variables and is consistent with the required instrument durations for other variable types. Consequently, licensees may specify the Type C variable operating time based on the plant's LBD.
(8)Modify Clause 5.4 to replace the term "pos t-event operating time" with "operating time."The term "post-event operating time" implies that th e plant is in a controlled condition (the eventhas been mitigated) when the instrumentation is first required to function. This is inconsistentwith the criteria for selection of accident monito ring variables, as the variables are derived from actions based on plant procedures (e.g., AOPs, EOPs , and EPGs). The actions described in theseprocedures encompass conditions during accident mitigation, as well as when the plantis in a controlled condition. The operating time for each variable is determined by the plant's LBDand should not imply that they are only requi red during the "post-event" phase of accidentmanagement. Consequently, licensees should consider the plant LBD's operating time for eachvariable when determining the required instrument duration.
RG 1.97, Rev. 4, Page 9
D. IMPLEMENTATION
The purpose of this section is to provide information to applicants and licensees regardingthe NRC staff's plans for using this regulatory guide. No backfitting is intended or approved in connection with the issuance of this guide.
Except in cases in which an applicant or licensee proposes or has previously establishedan acceptable alternative method for complying with specified portions of the NRC's regulations,the methods described in this guide will be used in evaluating (1) submittals in connection withapplications for construction permits, design certifications, operating licenses, and combined licenses,and (2) submittals from operating reactor licensees who voluntarily propose to initiate system modifications if there is a clear nexus between the proposed modifications and the subject for which guidance is provided herein.
RG 1.97, Rev. 4, Page 10
REGULATORY ANALYSISA separate regulatory analysis was not prepared for this regulatory guide. The regulatory analysisprepared for Draft Regulatory Guide DG-1128, "Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants," dated August 2005, also provides the regulatory basis for this regulatory guide. The NRC issued DG-1128 to solicit public comment c oncerning the draft of this fourth revisionof Regulatory Guide 1.97.A copy of the regulatory analysis for DG-1128 is available for inspection and copying for a feeat the NRC's Public Document Room (PDR), wh ich is located at 11555 Rockville Pike, Rockville,Maryland; the PDR's mailing address is USNRC PDR, Washington, DC 20555-0001. The PDR can also be reached by telephone at (301) 415-4737 or (800) 397-4205, by fax at (301) 415-3548, and by email to PDR@nrc.gov. Copies are also available at current rates from the U.S. Government Printing Office atP.O. Box 37082, Washington, DC 20402-9328 or by te lephone at (202) 512-1800. In addition, copiesare available at current rates from the National Technical Information Service at 5285 Port Royal Road, Springfield, VA 22161, on the Internet at http://www.ntis.gov, or by telephone at (703) 487-4650. In addition, the regulatory analysis is available electronically as a part of Draft Regulatory Guide DG-1128 through the NRC's Agencywide Documents Access and Management System (ADAMS)
at http://www.nrc.gov/reading-rm/adams.html , under Accession No. ML052150210.