ML18136A762

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Rev. 5, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants
ML18136A762
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Issue date: 04/30/2019
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RG 1.97, Rev. 5
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U.S. NUCLEAR REGULATORY COMMISSION REGULATORY GUIDE 1.97, REVISION 5 Issue Date: April 2019 Technical Lead: Pong C. Chung CRITERIA FOR ACCIDENT MONITORING INSTRUMENTATION FOR NUCLEAR POWER PLANTS A. INTRODUCTION Purpose This regulatory guide (RG) describes a method that is acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) for use in complying with the agencys regulations with respect to satisfying criteria for accident monitoring instrumentation in nuclear power plants. It endorses, with exceptions and clarifications, the Institute of Electrical and Electronic Engineers (IEEE) Standard (Std.) 497-2016, IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations (Ref. 1).

Applicability This RG applies to all holders of, and applicants for, operating licenses for nuclear power reactors under the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Domestic Licensing of Production and Utilization Facilities (Ref. 2), including those who have permanently ceased operations and have certified that fuel has been permanently removed from the reactor vessel. It also applies to holders of, and applicants for, a power reactor design certification or combined license under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants (Ref. 3), including those who have permanently ceased operations and have certified that fuel has been permanently removed from the reactor vessel.

Applicable Orders and Regulations

  • NRC Order EA-12-051, Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, requires, in part, that all licensees have a reliable indication of the water level in associated spent fuel storage pools (Ref. 4).
  • 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities provides regulations for licensing production and utilization facilities.

Written suggestions regarding this guide or development of new guides may be submitted through the NRCs public Web site in the NRC Library at https://nrcweb.nrc.gov/reading-rm/doc-collections/reg-guides/, under Document Collections, in Regulatory Guides, at https://nrcweb.nrc.gov/reading-rm/doc-collections/reg-guides/contactus.html.

Electronic copies of this RG, previous versions of RGs, and other recently issued guides are also available through the NRCs public Web site in the NRC Library at https://nrcweb.nrc.gov/reading-rm/doc-collections/reg-guides/, under Document Collections, in Regulatory Guides. This RG is also available through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under ADAMS Accession Number (No.) ML18136A762. The regulatory analysis may be found in ADAMS under Accession No. ML17083A133. The associated draft guide DG-1335 may be found in ADAMS under Accession No. ML17083A134, and the staff responses to the public comments on DG-1335 may be found under ADAMS Accession No. ML18136A761.

o 10 CFR 50.34(f)(2)(xix) requires applicants and licensees to provide instrumentation adequate for monitoring plant conditions following an accident that includes core damage.

o 10 CFR Part 50, Appendix A, GDC 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.

o 10 CFR Part 50, Appendix A, GDC 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. 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.

o 10 CFR Part 50, Appendix A, GDC 64, Monitoring Radioactivity Releases, requires operating reactor licensees to provide the means for monitoring the reactor containment atmosphere, spaces containing components to recirculate loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released as a result of postulated accidents.

  • 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants, governs the issuance of early site permits, standard design certifications, combined licenses, standard design approvals, and manufacturing licenses for nuclear power facilities.

o 10 CFR 52.47(a)(8), requires a design certification application for light-water reactor (LWR) designs to provide information necessary to demonstrate compliance with any technically relevant portions of the Three Mile Island requirements set forth in 10 CFR 50.34(f), except paragraphs (f)(1)(xii), (f)(2)(ix), and (f)(3)(v).

o 10 CFR 52.47(a)(23), requires a design certification application for LWR designs to present a description and analysis of design features for the prevention and mitigation of severe accidents.

o 10 CFR 52.79(a)(38), requires a combined license application for LWR designs to present a description and analysis of design features for the prevention and mitigation of severe accidents.

o 10 CFR 52.137(a)(23), requires a standard design approval application for LWR designs to present a description and analysis of design features for the prevention and mitigation of severe accidents.

o 10 CFR 52.157(f)(23), requires a manufacturing license application for LWR designs to present a description and analysis of design features for the prevention and mitigation of severe accidents.

RG 1.97, Page 2

Related Guidance

  • NUREG-0700, Human-System Interface Design Reviews Guidelines (Ref. 5), provides human factors engineering guidelines that address the physical and functional characteristics of human-system interfaces.
  • NUREG-0711, Human Factors Engineering Program Review Model (Ref. 6), contains information to assist the NRC staff with the review of the human factors engineering programs.
  • NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (Ref. 7).

o NUREG-0800, Chapter 7, Instrumentation and Controls, Branch Technical Position 7-10, "Guidance on the Application of Regulatory Guide 1.97" provides additional guidelines for reviewing an applicants or licensees accident monitoring instrumentation.

o NUREG-0800, Chapter 19, Severe Accident, identifies guidance for reviewing an applicants deterministic evaluation of design features for the prevention or mitigation of severe accidents.

  • RG 1.32, Criteria for Power Systems for Nuclear Power Plants, contains additional guidance related to the normative references in IEEE Std. 497-2016, Section 2 (Ref. 8).
  • RG 1.53, Application of the Single-Failure Criterion to Safety Systems, contains additional guidance related to the normative references in IEEE Std. 497-2016, Section 2 (Ref. 9).
  • RG 1.75, Physical Independence of Electric Systems, contains additional guidance related to the normative references in IEEE Std. 497-2016, Section 2 (Ref. 10).
  • RG 1.89, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants, contains additional guidance related to the normative references in IEEE Std. 497-2016, Section 2 (Ref. 11).
  • RG 1.100, Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants, contains additional guidance related to the normative references in IEEE Std. 497-2016, Section 2 (Ref. 12).
  • RG 1.152, Criteria for Use of Computers in Safety Systems of Nuclear Power Plants, contains additional guidance related to the normative references in IEEE Std. 497-2016, Section 2 (Ref. 13).
  • RG 1.153, Criteria for Safety System, contains additional guidance related to the normative references in IEEE Std. 497-2016, Section 2 (Ref. 14).

Purpose of Regulatory Guides The NRC issues RGs to describe to the public methods that the staff considers acceptable for use in implementing specific parts of the agencys regulations, to explain techniques that the staff uses in evaluating specific problems or postulated events, and to provide guidance to applicants. Regulatory guides are not substitutes for regulations and compliance with them is not required. Methods and RG 1.97, Page 3

solutions that differ from those set forth in RGs 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.

Paperwork Reduction Act This RG provides voluntary guidance for implementing the mandatory information collections in 10 CFR Parts 50 and 52 that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et.

seq.). These information collections were approved by the Office of Management and Budget (OMB),

approval numbers 3150-0011 and 3150-0151. Send comments regarding this information collection to the Information Services Branch (T6-A10M), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to Infocollects.Resource@nrc.gov, and to the OMB reviewer at: OMB Office of Information and Regulatory Affairs (3150-0011 and 3150-0151), Attn: Desk Officer for the Nuclear Regulatory Commission, 725 17th Street, NW Washington, DC20503; e- mail:

oira_submission@omb.eop.gov.

Public Protection Notification The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless the document requesting or requiring the collection displays a currently valid OMB control number.

RG 1.97, Page 4

B. DISCUSSION Reason for Revision The staff is issuing Revision 5 of RG 1.97 to endorse IEEE Std. 497-2016 with exceptions and clarifications. Revision 5 also expands the applicability of RG 1.97 to holders of, or applicants for, power reactor design certifications or combined licenses under 10 CFR Part 52, which was promulgated after the issuance Revision 4 of RG 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants (Ref. 15), and adds references to the NRCs 10 CFR Part 52 regulations and related NRC guidance documents.

Background

In March 1979, an accident occurred in Unit 2 of the Three Mile Island Nuclear Station. In the aftermath, the nuclear industry and the NRC adopted a more rigorous approach to accident monitoring. In May 1983, the NRC issued Revision 3 of RG 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident (Ref. 16). The RG prescribed a detailed list of variables to be monitored and specified a comprehensive list of design and qualification criteria to be met.

Because of its prescriptive nature, Revision 3 of RG 1.97 quickly became the de facto standard for accident monitoring. However, with the new instrumentation systems in advanced nuclear power plant designs, the nuclear industry recognized the need to develop a consolidated standard that was more flexible. Instead of prescribing the instrument variables to be monitored (as was the case in Revision 3 of RG 1.97), the industry developed performance-based criteria for use in selecting variables. These efforts resulted in the development of IEEE Std. 497-2002, IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations (Ref. 17). IEEE Std. 497-2002 established flexible, performance-based criteria for the selection, performance, design, qualification, display, and quality assurance of accident monitoring variables, which are the operators primary sources of accident monitoring information. The NRC endorsed IEEE Std. 497-2002 in Revision 4 of RG 1.97, subject to eight regulatory positions. The IEEE revised Std. 497 in 2010 (Ref. 18) to incorporate some of the NRC staffs regulatory positions in Revision 4 of RG 1.97 and also to revise some definitions and terminology.

Based on insights from the Fukushima Dai-ichi accident in March 2011, the nuclear industry in the United States recognized the need for instrumentation to monitor plant conditions after fuel damage has occurred. The NRC determined that all power reactor licensees must have a reliable means of remotely monitoring wide-range spent fuel pool levels. The NRC issued an order (NRC Order EA 051) to all licensees requiring spent fuel pool level instrumentation. Subsequently, the IEEE issued IEEE Std. 497-2016, which broadened the scope of the standard to include consideration of the instrumentation potentially required for coping with severe accidents. IEEE Std. 497-2016 also reflected a more technology-neutral approach and an effort to bring the standard in line with related international standards, as referenced in IEEE Std. 497-2016 and the Harmonization with International Standards, section of this Discussion below.

In addition to the Types A, B, C, D, and E variables defined in the previous revisions, IEEE Std. 497-2016 adds a new kind of variable, Type F, which provides primary information to indicate fuel damage and the effects of fuel damage. The regulatory requirement for Type F variables derives from specific design certification rules codified in the appendices of 10 CFR Part 52. As of the date of issuance of RG 1.97, Revision 5, at least one of the design certification rules incorporates, by reference, a design certification document that requires combined license holders to implement severe accident management RG 1.97, Page 5

guidelines (SAMGs). Because the SAMGs cannot be implemented without instrumentation enabling operators to determine what actions are needed during a severe accident, the design certification document requires these licensees to use variables (i.e., Type F variables) to provide information on fuel damage and the effects of fuel damage under severe accident conditions. As of the date of issuance of RG 1.97, Revision 5, every nuclear power plant in the United States has agreed, through a voluntary industry initiative, to perform timely updates of their site-specific SAMGs based on revisions to Owners Group generic SAMGs per a letter from Nuclear Energy Institute to NRC of October 26, 2015 (Ref. 19).

Therefore, this RG may be useful for new plant applicants or for existing plant licensees who choose to add new or convert some of their existing accident monitoring instrumentation to Type F variables.

In a related provision in 10 CFR 50.34(f)(2)(xix), the NRC requires new reactor applicants to provide instrumentation to monitor plant conditions following an accident that includes core damage. The basis for this regulation originated in NUREG-0660, NRC Action Plan Developed as a Result of the TMI-2 Accident, which was followed up by NUREG-0737, Clarification of TMI Action Plan Requirements. These NUREGs state that licensees should provide instrumentation with sufficient range and accuracy to perform the intended function in the environment to which the instruments will exposed during the accidents.

IEEE Std. 497-2016, Clauses 5.3, Response time, 8.1.2, Human factors, and 8.5, Display location, discuss the incorporation of human factors guidelines or accepted human factors principles into the design of displays for accident monitoring instrumentation. The NRC provides additional guidance in Chapter 18, Human Factors Engineering, of the NRCs Standard Review Plan in NUREG-0800, as well as in NUREG-0700, and NUREG-0711.

Harmonization with International Standards The NRC has a goal of harmonizing its guidance with international standards, to the extent practical. The International Atomic Energy Agency (IAEA) and the International Electrotechnical Commission (IEC) have issued a significant number of standards, guidance and technical documents, and recommendations addressing good practices in most aspects of radiation protection and accident monitoring pertinent to the topics covered in this RG, including the following:

  • IAEA Nuclear Energy Series No. NP-T-3.16, Accident Monitoring Systems for Nuclear Power Plants, issued February 2015 (Ref. 20).
  • IAEA Safety Standard Series No. TECDOC-1818, Assessment of Equipment Capability to Perform Reliably under Severe Accident Conditions, issued July 2017 (Ref. 21).
  • IAEA Safety Standards Series No. SSG-39, Design of Instrumentation and Control Systems for Nuclear Power Plants, issued April 2016 (Ref. 22).
  • IEEE 63147-2017 - IEEE/IEC International Standard, Criteria for accident monitoring instrumentation for nuclear power generating stations, December 22, 2017 (Ref. 23).

This RG adopts similar design and performance guidelines and is consistent with the safety principles provided in these IAEA publications.

RG 1.97, Page 6

Documents Discussed in Staff Regulatory Guidance This RG endorses the use of one or more codes or standards developed by external organizations and other third party guidance documents. These codes, standards, and third party guidance documents may contain references to other codes, standards or third party guidance documents (secondary references). If a secondary reference is incorporated by reference into NRC regulations as a requirement, then licensees and applicants must comply with that standard as set forth in the regulation. If the secondary reference is endorsed in a RG as an acceptable approach for meeting an NRC requirement, then the standard constitutes a method acceptable to the NRC staff for meeting that regulatory requirement as described in the specific RG. If the secondary reference is neither incorporated by reference into NRC regulations nor endorsed in a RG, then the secondary reference is neither a legally-binding requirement nor a generic NRC approved acceptable approach for meeting an NRC requirement. However, licensees and applicants may consider and use the information in the secondary reference, if appropriately justified and consistent with applicable NRC requirements and current regulatory practice.

RG 1.97, Page 7

C. STAFF REGULATORY GUIDANCE This RG endorses IEEE Std. 497-2016 as an acceptable method for providing instrumentation to monitor variables for accident conditions, subject to the following regulatory positions:

1. The applicability of IEEE Std. 497-2016 for licensees and applicants already committed to an earlier revision of RG 1.97 is clarified as follows:

A licensee or applicant committed to an earlier revision of RG 1.97 may voluntarily convert its entire accident monitoring program using the criteria in this revision. To do so, the licensee or applicant needs to analyze its accident monitoring program in its entirety.

The NRC staff recognizes that licensees or applicants may be interested in converting to the latest revision of RG 1.97, which would update the plants entire accident monitoring program in its current licensing basis to a less prescriptive program. This conversion would entail conducting a new analysis of every type of variable and may result in physical modifications to the plant and changes to its procedures and technical specifications.

2. Licensees and applicants that have committed to an earlier revision of RG 1.97 may voluntarily add Type F variables, that is, those variables to be monitored while managing a severe accident.

A licensee or applicant committed to an earlier revision of RG 1.97 may use the guidance in this revision to add Type F variables to their plants accident monitoring program while remaining committed to an earlier version of the RG for Types A through E variables. The licensee or applicant should first evaluate the basis underlying its severe accident analyses to determine the variables to be selected as Type F variables.

This revision of RG 1.97 may be used to select only Type F variables and determine their design, performance, qualification, and display criteria. When a licensee elects to adopt this revision for only Type F variables and remains committed to an earlier version of RG 1.97 for variables of Type A, B, C, D, and E, the licensee should document how it applied each revision of RG 1.97.

3. The scope of variables analyzed as Type A should include those variables that are associated with specific, planned, manually-controlled actions for which no automatic control is provided. The second to last sentence of Clause 4.1, Type A variable, of IEEE Std. 497-2016 is clarified as follows:

Type A variables include those variables that are associated with specific, planned, manually-controlled actions for which no automatic control is provided that may also be identified in written procedures.

IEEE Std. 497-2016 defines contingency actions as alternative actions taken to address unexpected responses of the plant or conditions beyond its licensing basis (e.g., actions taken for multiple equipment failures). Clause 4.1 uses this definition to prescreen out potential Type A variables. The staff agrees with this clause, but the staff does not agree that the criteria in this clause should be used during the initial identification process for Type A variables. Nuclear steam supply system vendors have not used the contingency actions term consistently in Emergency Procedures Guidelines for current plant designs and, therefore, the term contingency actions should not be used in accordance with the criteria in Clause 4.1. Furthermore, the variable selection process encompassed a non-prescriptive, performance-based approach in Revision 4 and Revision 5 of RG 1.97. Thus, the staff cannot endorse the complete exclusion of variables based upon the contingency actions term RG 1.97, Page 8

(especially those associated with plant-specific operating procedures or guidelines). Rather, the scope of instruments that could potentially be selected for accident monitoring (based upon the selection criteria) should be initially be as encompassing as possible. Later, in the process of selecting the final list of variables to be monitored, licensees could screen out those instruments associated only with contingency actions designed to address conditions beyond the plants licensing basis.

4. The ranges of instrumentation for Type F variables should include appropriate margins, as noted by the following clarifications to Clause 5.1, Range. To better communicate the intent of Range, the third paragraph of Clause 5.1 is clarified by creating three independent elements: one paragraph describing the range for normal operations and anticipated operational occurrences, a second paragraph for emergency operating procedures (EOPs), and a final paragraph for SAMGs.

To clarify, the range of instrumentation with respect to normal operation and anticipated operational occurrences shall have sufficient range with appropriate margin to cover the predicted limits of the variables and address a source term that considers fuel damage.

The range of instrumentation used to implement EOPs should cover, with appropriate margin, the predicted full range of the variables with the consideration of analytical uncertainties and environment measurement errors under design-basis accident conditions.

If the SAMG option is adopted, the instrumentation used to implement SAMGs should have sufficient range to cover, with appropriate margin, the predicted limits of the variables determined as part of severe accident modeling of the plant. To support SAMG functions, the licensee may use multiple instrument channels with different ranges to cover the desired full range.

5. The accuracy of instrumentation should be derived from the licensing basis. Clause 5.2, Accuracy, is clarified to mean that the instrumentation used to implement EOPs must be evaluated to fulfill accuracy requirements derived from the plant design bases and modeled based on expected plant conditions that could occur during the portion of the accident progression when the affected instrumentation would be needed. During SAMG implementation, although the value of a variable may be needed, it is usually more important to follow the increasing or decreasing trends of the variables used such that the effectiveness of the mitigating strategies applied can be observed. The accuracy requirements specified for each severe accident mitigation strategy are based on the level of accuracy needed for decision making for support of the mitigation strategy or for support of emergency preparedness functions. If redundant instrumentation is provided, or where spatial orientation and distribution is required for measuring the same parameter, the equipment must be sufficiently accurate to preclude ambiguous trend information.
6. IEEE Std. 497-2016 does not mention the number of measurement points for each variable (with the exception of redundancy requirements). In general, the number of measurement points should be sufficient to adequately indicate the variable value (e.g., containment temperature may require spatial distribution of several measurement points).
7. The following portions of IEEE Std. 497-2016 are clarified as follows:

a) IEEE Std. 497-2016, Section 4.7, the description of a Type F variable under the second column Selection criteria for the variable type of Table 1, Summary of accident monitoring variable types/source documents, states:

Monitor the direct effects (e.g. combustible gases concentration, radiation, pressure, or temperature of fuel damage).

RG 1.97, Page 9

The wording temperature of fuel damage can cause confusion and is clarified as follows:

Monitor the direct effects of fuel damage (e.g. combustible gases concentration, radiation, pressure, or temperature).

b) IEEE Std. 497-2016, Section 2, Normative references, should be utilized as described in the corresponding RGs that endorsed those standards. A list of these RGs is included in Section A, Related Guidance, above.

c) IEEE Std. 497-2016, Section 3, Definitions, revises the definition of anticipated operational occurrence. This term was already defined in 10 CFR Part 50, Appendix A, and the definition in previous versions of IEEE Std. 497 was identical to the definition in 10 CFR Part 50, Appendix A. Although it does not appear that the intent was to significantly change the way the standard uses the term, the NRC only uses the definition in 10 CFR Part 50, Appendix A. Thus licensees and applicants should retain the definition of anticipated operational occurrence in 10 CFR Part 50, Appendix A.

8. IEEE Std. 497-2016 includes four informative annexes; however, the NRC does not endorse any annex in IEEE Std. 497-2016 because those annexes are informative only.

RG 1.97, Page 10

D. IMPLEMENTATION The purpose of this section is to provide information on how applicants and licensees1 may use this RG and information regarding the NRCs plans for using this RG. In addition, it describes how the NRC staff complies with 10 CFR 50.109, Backfitting, and any applicable issue finality provisions in 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.

Use by Applicants and Licensees Applicants and licensees may voluntarily2 use the guidance in this document to demonstrate compliance with the underlying NRC regulations. Methods or solutions that differ from those described in this RG may be deemed acceptable if they provide sufficient basis and information for the NRC staff to verify that the proposed alternative demonstrates compliance with the appropriate NRC regulations.

Current licensees may continue to use guidance the NRC found acceptable for complying with the identified regulations as long as their current licensing basis remains unchanged.

Licensees may use the information in this RG for actions that do not require NRC review and approval such as changes to a facility design under 10 CFR 50.59, Changes, Tests, and Experiments.

Licensees may use the information in this RG or applicable parts to resolve regulatory or inspection issues.

Use by NRC Staff The NRC staff does not intend or approve any imposition or backfitting of the guidance in this RG. The NRC staff does not expect any existing licensee to use or commit to using the guidance in this RG, unless the licensee makes a change to its licensing basis. The NRC staff does not expect or plan to request licensees to voluntarily adopt this RG to resolve a generic regulatory issue. The NRC staff does not expect or plan to initiate NRC regulatory action which would require the use of this RG. Examples of such unplanned NRC regulatory actions include issuance of an order requiring the use of the RG, requests for information under 10 CFR 50.54(f) as to whether a licensee intends to commit to use of this RG, generic communication, or promulgation of a rule requiring the use of this RG without further backfit consideration.

During regulatory discussions on plant-specific operational issues, the staff may discuss with licensees various actions consistent with staff positions in this RG, as one acceptable means of meeting the underlying NRC regulatory requirement. Such discussions would not ordinarily be considered backfitting even if prior versions of RG 1.97 are part of the licensing basis of the facility. However, unless this RG is part of the license for a facility, the staff may not represent to the licensee that the licensees failure to comply with the positions in this RG constitutes a violation of underlying NRC regulatory requirements.

If an existing licensee voluntarily seeks a license amendment or change and (1) the NRC staffs consideration of the request involves a regulatory issue directly relevant to this RG and (2) the specific subject matter of this RG is an essential consideration in the staffs determination of the acceptability of 1 In this section, licensees refers to licensees of nuclear power plants under 10 CFR Parts 50 and 52; and applicants, refers to applicants for licenses and permits for (or relating to) nuclear power plants under 10 CFR Parts 50 and 52, and applicants for standard design approvals and standard design certifications under 10 CFR Part 52.

2 In this section, voluntary and voluntarily mean that the licensee is seeking the action of its own accord, without the force of a legally binding requirement or an NRC representation of further licensing or enforcement action.

RG 1.97, Page 11

the licensees request, then the staff may request that the licensee either follow the guidance in this RG or otherwise demonstrate compliance with the underlying NRC regulatory requirements. This is not considered backfitting as defined in 10 CFR 50.109(a)(1). Neither is it a violation of any of the issue finality provisions in 10 CFR Part 52.

Additionally, an existing applicant may be required to comply with new rules, orders, or guidance if 10 CFR 50.109(a)(3) applies.

If a licensee believes that the NRC is either using this RG or requesting or requiring the licensee to implement the methods or processes in this RG in a manner inconsistent with the discussion in this Implementation section, then the licensee may file a backfit appeal with the NRC in accordance with the guidance in NRC Management Directive 8.4, Management of Facility-Specific Backfitting and Information Collection (Ref. 24), and NUREG-1409, Backfit Guidelines (Ref. 25).

RG 1.97, Page 12

REFERENCES 3

1. Institute of Electrical and Electronics Engineers (IEEE), Standard (Std.) 497-2016, IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations, New York, NY, 2016.4
2. U.S. Code of Federal Regulations (CFR), Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter 1, Title 10, Energy.
3. CFR, Licenses, Certifications, and Approvals for Nuclear Power Plants, Part 52, Chapter 1, Title 10, Energy.
4. U.S. Nuclear Regulatory Commission (NRC), Order EA-12-051, Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, Washington, DC, March 12, 2012 (ADAMS Accession No. ML12056A044).
5. NRC, NUREG-0700, Human-System Interface Design Review Guidelines, Washington, DC.
6. NRC, NUREG-0711, Human Factors Engineering Program Review Model, Washington, DC.
7. NRC, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Washington, DC.
8. NRC, Regulatory Guide (RG) 1.32, Criteria for Power Systems for Nuclear Power Plants, Washington, DC.
9. NRC, RG 1.53, Application of the Single-Failure Criterion to Safety Systems, Washington, DC.
10. NRC, RG 1.75, Physical Independence of Electric Systems, Washington, DC.
11. NRC, RG 1.89, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants, Washington, DC.
12. NRC, RG 1.100, Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants, Washington, DC.
13. NRC, RG 1.152, Criteria for Use of Computers in Safety Systems of Nuclear Power Plants, Washington, DC.

3 Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public Web site at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html. The documents can also be viewed online or printed for a fee in the NRCs Public Document Room (PDR) at 11555 Rockville Pike, Rockville, MD. For problems with ADAMS, contact the PDR staff at 301-415-4737 or (800) 397-4209; fax (301) 415-3548; or e-mail pdr.resource@nrc.gov.

4 Copies of Institute of Electrical and Electronics Engineers (IEEE) documents may be purchased from the Institute of Electrical and Electronics Engineers Service Center, 445 Hoes Lane, PO Box 1331, Piscataway, NJ 08855, or through IEEEs public Web site at http://www.ieee.org/publications_standards/index.html. IEC has published IEEE Std. 497-2016 as IEC 63147:2017.

RG 1.97, Page 13

14. NRC, RG 1.153, Criteria for Safety System Washington, DC.
15. NRC, RG 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, Revision 4, Washington, DC, June 2006.
16. NRC, RG 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, Revision 3, Washington, DC, May 1983.
17. IEEE Std. 497-2002, IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations, New York, NY, November 2002.
18. IEEE Std. 497-2010, IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations, New York, NY, November 2010.
19. Nuclear Energy Institute, Industry Initiative to Maintain Severe Accident Management Guidelines, October 26, 2015 (ADAMS Accession No. ML15335A442).
20. International Atomic Energy Agency (IAEA), Nuclear Energy Series No. NP-T-3.16, Accident Monitoring Systems for Nuclear Power Plants, Vienna, Austria, February 2015.5
21. IAEA, Safety Standards Series No. TECDOC-1818, Assessment of Equipment Capability to Perform Reliably under Severe Accident Conditions, Vienna, Austria, July 2017.
22. IAEA, Safety Standards Series No. SSG-39, Design of Instrumentation and Control Systems for Nuclear Power Plants, Vienna, Austria, April 2016.
23. IEEE 63147-2017 - IEEE/IEC International Standard, Criteria for accident monitoring instrumentation for nuclear power generating stations, December 22, 2017.6
24. NRC Management Directive 8.4, Management of Facility-Specific Backfitting and Information Collection, Washington DC.
25. NRC, NUREG-1409, Backfitting Guidelines, Washington DC.

5 Copies of International Atomic Energy Agency (IAEA) documents may be obtained through their Web site:

www.IAEA.Org or by writing the International Atomic Energy Agency, PO Box 100 Wagramer Strasse 5, A-1400 Vienna, Austria. Telephone (+431) 2600-0, Fax (+431) 2600-7, or E-Mail at Official.Mail@IAEA.Org.

6 Copies of International Electrotechnical Commission (IEC) documents may be obtained through their Web site:

http://www.iec.ch/ or by writing the IEC Central Office at P.O. Box 131, 3 Rue de Varembé, 1211 Geneva, Switzerland, Telephone +41 22 919 02 11 RG 1.97, Page 14

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