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 for use 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 staff need in reviewing applications for permits and licenses.  Regulatory guides are not substitutes for regulations, and compliance with them is not required.  Methods and solutions 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 of a 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 with improvements 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, Office of 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, Produc
ts; 7, Transportation; 8, Occupational Health; 9, Antitrust and Financial Review; and 10, General.
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
reactor4A.  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 to satisfying criteria for accident monitoring instrumentation in nuclear power plants.  Spec
ifically, the method described 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 instrumentation
to monitor variables and systems over their anticipated ranges for accident conditions as appropriate
to ensure adequate safety.
•Criterion 19, “Control Room,” requires operating reactor licensees to provide a control room from which actions can be taken to maintain the nuclear power 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 means for 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.
In 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 plant conditions 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 power plants.  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. 497-2002.1
This revised regulatory guide is intended for licensees of new nuclear power plants.2  Previous revisions of this regulatory guide remain in effect for licensees of current operating reactors,2 who are
unaffected by this revision.  (See the discussion of regulatory position #1 in Section C of this guide regarding the applicability of IEEE Std. 497-2002 for current operating reactors.)
In general, information provided by regulatory guides is reflected in the NRC’s “Standard Review 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 operate 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 requirements in 10 CFR Part 50, which provides the regulatory basis for this guide.  The Office of Management
and Budget (OMB) has approved those information collection requirements under OMB control number 3150-0011.  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 document displays a currently valid OMB control number.
1IEEE publications may be purchased from the IEEE Service Center, which is located at 445 Hoes Lane, Piscataway, NJ 08855 [, phone (800) 678-4333].
2The terms “new nuclear power plant” and “new plant” refer to any nuclear power plant for which the licensee obtained an 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 before
the NRC issued Revision 4 of Regulatory Guide 1.97.
3Copies are available at current rates from the U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20402-9328 [telephone (202) 512-1800], or from the National Technical Information Service (NTIS), 5285 Port Royal Road, Springfield, Virginia 22161 [v, telephone (703) 487-4650].  Copies are available for
inspection or copying for a fee from the NRC’s Public Document Room (PDR), which 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
B.  DISCUSSION
In the aftermath of the accident at Three Mile Island, Unit 2 (TMI-2), in 1979, the United States adopted a more rigorous approach for accident monitoring systems, which resulted in three major sources of 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 operator
should monitor to ensure safety during an accident and the subsequent long-term stable shutdown phase.  The American National Standards Institute (ANSI) promulgated this standard, which was
developed 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 parameter changes during an accident, and (2) extended-range instrumentation deemed appropriate for previously unforeseen events.  As the source for
specific instrumentation design criteria, ANSI/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 Conditions During and Following an Accident,”6 dated
May 1983, prescribed a detailed list of variables to monitor, and specified a comprehensive list
of design and qualification criteria to be met.
Given its prescriptive nature, Revision 3 of Regulatory Guide 1.97 quickly became the de facto standard for accident monitoring, and both ANSI/ANS-4.5-1980 and IEEE Std. 497-1981 fell out of use and were subsequently withdrawn as active standards.  Nonetheless, Revision 3 of Regulatory Guide 1.97 has become outdated, in that it does not provide criteria for advanced instrumentation system designs based 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 on the accident management functions of the individual variable types.
4Copies may be obtained from the American Nuclear Society, which is located at 555 North Kensington Avenue, La Grange Park, Illinois 60525 [, phone (708) 352-6611].
5IEEE publications may be purchased from the IEEE Service Center, which is located at 445 Hoes Lane, Piscataway, NJ 08855 [, phone (800) 678-4333].
6Copies are available at current rates from the U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20402-9328 [telephone (202) 512-1800], or from the National Technical Information Service (NTIS), 5285 Port Royal Road, Springfield, Virginia 22161 [v, telephone (703) 487-4650].  Copies are available for
inspection or copying for a fee from the NRC’s Public Document Room (PDR), which 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
With the increased use of digital instrumentation systems in advanced nuclear power plant designs, the nuclear industry came to recognize a need to develop a consolidated standard that was more flexible than Revision 3 of Regulatory Guide 1.97.  Instead of prescribing the instrument variables to be monitored (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 and qualification criteria for each variable category, the industry sought to standardize the criteria based on the accident management functions of the given type of variable.  These efforts resulted in the development 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 Power Engineering Committee, Subcommittee 6, Working Group 6.1, “Post-Accident Monitoring.”
Unlike its predecessor, IEEE Std. 497-2002 establishes flexible, performance-based criteria for the selection, performance, design, qualification, display and quality assurance of accident monitoring vari
ables.  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 variables, 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 not meet the criteria in this guide.
Clause 8.1.2 of IEEE Std. 497-2002 cites several industry codes and standards for human factors criteria.  The NRC provides additional guidance in NUREG-0700, “Human-System Interface Design Review Guideline:  Review Methodology and Procedures”8; NUREG-0711, “Human Factors Engineering Program Review Model”8; and Chapter 18, “Human Factors Engineering,”of the NRC’s Standard Review Plan (NUREG-0800).8
7IEEE publications may be purchased from the IEEE Service Center, which is located at 445 Hoes Lane, Piscataway, NJ 08855 [, phone (800) 678-4333].
8Copies are available at current rates from the U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20402-9328 [telephone (202) 512-1800], or from the National Technical Information Service (NTIS), 5285 Port Royal Road, Springfield, Virginia 22161 [v, telephone (703) 487-4650].  Copies are available for
inspection or copying for a fee from the NRC’s Public Document Room (PDR), which 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
Clause 6.2 of IEEE Std. 497-2002 states, in part, that the use of identical software in redundant instrumentation channels is acceptable, provided that the licensee conducts an analysis to demonstrate defense-in-depth against common-mode software failure.  The NRC provides related guidance in Branch Technical Position HICB-19, “Guidance for Evaluation of Defense-in-Depth and Diversity in Digital Computer-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 Channel Accuracy.”  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.3 provides typical accuracy requirements.  Specifically, Clause A.3 states, in part, “Historically, the required accuracy for instrument channels relied upon to monitor containment pressure
and hydrogen concentration has been ±10 percent of full span.”  However, the NRC staff notes that this example may not be applicable to all nuclear power plants.  Traditionally, the required accuracy of accident monitoring instrument channels is established based on the assigned
function and the plant’s safety analysis and licensing basis.
•Annex B, “Bibliography,” lists the references cited in the standard, and provides sufficient detail for users to obtain further information regarding specific aspects of the standard.
9Copies are available at current rates from the U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20402-9328 [telephone (202) 512-1800], or from the National Technical Information Service (NTIS), 5285 Port Royal Road, Springfield, Virginia 22161 [v, telephone (703) 487-4650].  Copies are available for
inspection or copying for a fee from the NRC’s Public Document Room (PDR), which is located at 1155
5 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

版权声明：本站内容均来自互联网，仅供演示用，请勿用于商业和其他非法用途。如果侵犯了您的权益请与我们联系QQ:729038198，我们将在24小时内删除。