ML21288A565: Difference between revisions

From kanterella
Jump to navigation Jump to search
(StriderTol Bot insert)
 
(StriderTol Bot change)
Line 18: Line 18:


=Text=
=Text=
{{#Wiki_filter:U.S. NUCLEAR REGULATORY COMMISSION DRAFT REGULATORY GUIDE DG-1393 Proposed New Regulatory Guide 1.248 Preliminary Draft for ACRS Consideration November 2021 Issue Date: February 2022 Technical Lead: Mohammad Sadollah GUIDE FOR ASSESSING, MONITORING, AND MITIGATING AGING EFFECTS ON ELECTRICAL EQUIPMENT USED IN NUCLEAR POWER GENERATING STATIONS AND OTHER NUCLEAR FACILITIES A. INTRODUCTION Purpose This regulatory guide (RG) describes an approach that is acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) to meet regulatory requirements for assessing, monitoring, and mitigating aging effects on electrical equipment used in nuclear power generating stations and other nuclear facilities. It endorses Institute of Electrical and Electronics Engineers (IEEE) Standard (Std.)
{{#Wiki_filter:}}
1205-2014, IEEE Guide for Assessing, Monitoring, and Mitigating Aging Effects on Electrical Equipment Used in Nuclear Power Generating Stations and Other Nuclear Facilities (Ref. 1).
Applicability This RG applies to licensees and applicants subject to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities (Ref. 2);
10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants (Ref. 3);
10 CFR Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants (Ref. 4).
With respect to 10 CFR Part 50, this RG applies to holders of or applicants for an operating license. With respect to 10 CFR Part 52, this RG applies to holders of and applicants for combined licenses, standard design certifications, standard design approvals, and manufacturing licenses. This RG does not apply to nuclear power plants that have submitted licensing termination certifications as required by 10 CFR 50.82(a)(1) and 10 CFR 52.110(a). Though the guidance primarily reflects past reviews of large light-water nuclear power plant applications, this RG may also provide useful guidance to fuel cycle, spent fuel storage, waste disposal facilities, and non-power production or utilization licensees.
This RG is being issued in draft form to involve the public in the development of regulatory guidance in this area. It has not received final staff review or approval and does not represent an NRC final staff position. Public comments are being solicited on this DG and its associated regulatory analysis. Comments should be accompanied by appropriate supporting data. Comments may be submitted through the Federal rulemaking Web site, http://www.regulations.gov, by searching for draft regulatory guide DG-1361. Alternatively, comments may be submitted to the Office of Administration, Mailstop: TWFN 7A-06M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN:
Program Management, Announcements and Editing Staff. Comments must be submitted by the date indicated in the Federal Register notice.
Electronic copies of this DG, previous versions of DGs, and other recently issued guides are available through the NRCs public Web site under the Regulatory Guides document collection of the NRC Library at https://nrcweb.nrc.gov/reading-rm/doc-collections/reg-guides/. The DG is also available through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML21288A115. The regulatory analysis may be found in ADAMS under Accession No. ML21288A112. .
 
Applicable Regulations
* 10 CFR Part 50 requires, among other things, that structures, systems, and components (SSCs) that are important to safety in a nuclear power plant must be designed to accommodate the effects of environmental conditions (i.e., remain functional under postulated design-basis events).
o    10 CFR 50.34(b)(6)(iv) requires each application for an operating license to include a final safety analysis report that contains plans for conducting normal operations, including maintenance, surveillance, and periodic testing of SSCs.
o    10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, requires that holders of or applicants for an operating license issued under 10 CFR Part 50 shall establish a program for the environmental qualification of electric equipment as defined in 10 CFR 50.49.
o    10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, requires that each holder of an operating license for a nuclear power plant under 10 CFR Part 50 and each holder of a combined license under 10 CFR Part 52 (after the Commission makes the finding under 10 CFR 52.103(g)) shall monitor the performance or condition of SSCs against licensee-established goals, in a manner sufficient to provide reasonable assurance that these SSCs are capable of fulfilling their intended functions. These goals shall be established commensurate with safety and, where practical, take into account industrywide operating experience. When the performance or condition of an SSC does not meet established goals, appropriate corrective action shall be taken. For a nuclear power plant for which the licensee has submitted the certifications specified in 10 CFR 50.82(a)(1) or 10 CFR 52.110(a)(1), as applicable, 10 CFR 50.65 shall only apply to the extent that the licensee shall monitor the performance or condition of all SSCs associated with the storage, control, and maintenance of spent fuel in a safe condition, in a manner sufficient to provide reasonable assurance that these SSCs are capable of fulfilling their intended functions.
o    General Design Criterion 4, Environmental and dynamic effects design bases, of Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50 states, in part, that SSCs important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.
* 10 CFR Part 52 requires that SSCs important to safety in a nuclear power plant be designed to accommodate the effects of environmental conditions and that design control measures, such as testing, be used to check the adequacy of the design.
o    Under 10 CFR 52.79(a)(15), an application for a combined license shall contain a final safety analysis report that describes the program, and its implementation, for monitoring the effectiveness of maintenance necessary to meet the requirements of 10 CFR 50.65.
o    10 CFR 52.137(a)(13) requires that an applicant for a standard design approval must provide the list of electric equipment that is important to safety as defined by 10 CFR 50.49(d).
10 CFR 50.49(a) requires that holders of a combined license or a manufacturing license issued under 10 CFR Part 52 shall establish a program for the environmental qualification of electric equipment as defined in 10 CFR 50.49. For a manufacturing license as defined in 10 CFR 52.157, Contents of applications; technical information in the final safety analysis report, only electric equipment defined in DG-1393, Page 2 Preliminary Draft for ACRS Consideration
 
10 CFR 50.49(b) that is within the scope of the manufactured reactor must be included in the environmental qualification program.
* 10 CFR Part 54 provides the requirements for the issuance of renewed operating licenses and renewed combined licenses for nuclear power plants licensed in accordance with Sections 103 or 104b of the Atomic Energy Act of 1954, as amended (Ref. 5), and Title II of the Energy Reorganization Act of 1974 (Ref, 6).
o  10 CFR 54.21, Contents of applicationtechnical information, provides the requirements for the technical information in a license renewal application or a subsequent license renewal application.
o  10 CFR 54.21(3) states, For each structure and component identified in paragraph (a)(1) of this section, demonstrate that the effects of aging will be adequately managed so that the intended function(s) will be maintained consistent with the [current licensing basis]
for the period of extended operation.
Related Guidance
* RG 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants (Ref. 7),
provides general guidelines for complying with 10 CFR 50.65(a)(1).
* RG 1.218, Condition Monitoring Techniques for Electric Cables Used in Nuclear Power Plants (Ref. 8), describes a programmatic approach to condition monitoring of electric cable systems and their operating environments, as well as acceptable condition monitoring techniques.
* NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear power Plants, Revision 2, issued December 2010 (Ref. 9), provides the criteria used by the NRC staff for reviewing aging management programs for in-scope SSCs for nuclear power plants for the initial renewal of an operating license.
* NUREG-1801, Generic Aging Lessons Learned (GALL) Report, Revision 2, issued December 2010 (Ref. 10), provides recommendations for aging management of in-scope SSCs for the initial renewal of an operating license.
* NUREG-2191, Volumes 1 and 2, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, issued July 2017 (Ref. 11), provides recommendations for aging management of in-scope SSCs for the subsequent renewal of renewed operating licenses.
* NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power PlantsFinal Report, issued July 2017 (Ref. 12), provides guidance to the NRC staff for the review of aging management programs for in-scope SSCs for subsequent license renewal applications that request operation for 60 to 80 years.
Purpose of Regulatory Guides The NRC issues RGs to describe methods that are acceptable to the staff for implementing specific parts of the agencys regulations, to explain techniques that the staff uses in evaluating specific issues or postulated events, and to describe information that the staff needs in its review of applications for permits and licenses. Regulatory guides are not NRC regulations and compliance with them is not required. Methods and solutions that differ from those set forth in RGs are acceptable if supported by a basis for the issuance or continuance of a permit or license by the Commission.
DG-1393, Page 3 Preliminary Draft for ACRS Consideration
 
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, DC 20503; 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.
DG-1393, Page 4 Preliminary Draft for ACRS Consideration
 
B. DISCUSSION Reason for Issuance The NRC is issuing this guide to endorse IEEE Std. 1205-2014 which provides an approach acceptable for meeting regulatory requirements for assessing, monitoring, and mitigating aging effects on electrical equipment used in nuclear power generating stations and other nuclear facilities.
 
===Background===
Nuclear power generating stations and other nuclear facilities licensed under 10 CFR Part 50 and 10 CFR Part 52 are required to monitor the performance or condition of SSCs against licensee-established goals, in a manner sufficient to provide reasonable assurance that these SSCs are capable of fulfilling their intended functions. These goals shall be established commensurate with safety and, where practical, take into account industrywide operating experience. When the performance or condition of SSCs does not meet established goals, appropriate corrective action shall be taken.
The integrity of SSCs is monitored, in part, through periodic surveillance testing and performance monitoring at a system level. However, this may not specifically focus on each component and may not be sufficient to detect all the aging and degradation mechanisms to which a particular piece of equipment or component is susceptible. While these tests can demonstrate the function and performance of the equipment and components under test conditions, the tests do not verify the continued successful performance of equipment and components when they are called upon to operate under their worst case design loading conditions for extended periods, as they would under anticipated normal service operating conditions or under design-basis accident conditions. Therefore, periodic surveillance testing of associated equipment may not provide specific information on the status of aging degradation processes or physical integrity of the aggregate parts and components. Consequently, a component with undetected damage or degraded condition could pass a system functional test but still fail unexpectedly when called upon to operate under anticipated environmental conditions or the more severe stresses encountered in emergency operation during a design-basis event (e.g., fully loaded equipment, more extreme environmental conditions, extended operation in a heavily loaded state). As an example, conventional test programs do not detect insulation degradation of a cable under normal service environments. However, operating experience indicates the inservice failures of several power cables (see, for example, Generic Letter 2007-01, Inaccessible or Underground Power Cable Failures that Disable Accident Mitigation Systems or Cause Plant Transients, dated February 7, 2007 (Ref. 13)).
Nuclear power generating stations with renewed licenses under 10 CFR Part 54 are required to demonstrate that the effects of aging will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. The scope of the SSCs under this rule includes, for the most part, passive long-lived components such as electrical cables, metal enclosed buses, electrical connections, transmission conductors, and switchyard buses.
This RG provides guidelines and methods acceptable to the NRC staff for assessing, monitoring, and mitigating aging effects on electrical equipment in nuclear generating stations and other nuclear facilities.
Consideration of International Standards The International Atomic Energy Agency (IAEA) works with member states and other partners to promote the safe, secure, and peaceful use of nuclear technologies. The IAEA develops safety standards and safety guides for protecting people and the environment from the harmful effects of ionizing DG-1393, Page 5 Preliminary Draft for ACRS Consideration
 
radiation. These standards and guides provide a system of safety fundamentals, safety requirements, and safety guides reflecting an international perspective on what constitutes a high level of safety. In developing or updating RGs, the NRC has considered IAEA safety standards and safety guides in order to benefit from the international perspectives.
The NRC staff has actively participated with the IAEA in developing and updating the IAEA International Generic Aging Lessons Learned (IGALL) program. The development of IGALL started with NRC guidance for license renewal from the GALL Report, Revision 2, and the latest updates to IGALL used GALL-SLR technical information as the basis for many of the updates. The NRC continues to review information from IGALL program counterparts and to evaluate the need to make changes to the GALL Report and the GALL-SLR. The staff notes that IAEA Safety Reports Series No. 82, Ageing Management for Nuclear Power Plants: International Generic Ageing Lessons Learned (IGALL), issued 2015 (Ref. 14), may provide additional useful information.
Documents Discussed in Staff Regulatory Guidance This RG endorses in part, 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 has itself been 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 has been 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 has neither been 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, consistent with current regulatory practice, and consistent with applicable NRC requirements.
DG-1393, Page 6 Preliminary Draft for ACRS Consideration
 
C. STAFF REGULATORY GUIDANCE The NRC staff considers conformance with the requirements in IEEE Standard 1205-2014 to constitute an acceptable method for use in satisfying the Commissions regulations with respect to maintenance and aging management of applicable SSCs subject to aging stressors, aging mechanisms, and aging effects to ensure facility safety throughout the period of initial license operation, extended operation, and subsequent extended operation.
DG-1393, Page 7 Preliminary Draft for ACRS Consideration
 
D. IMPLEMENTATION The NRC staff may use this regulatory guide as a reference in its regulatory processes, such as licensing, inspection, or enforcement. However, the NRC staff does not intend to use the guidance in this regulatory guide to support NRC staff actions in a manner that would constitute backfitting as that term is defined in 10 CFR 50.109, Backfitting, and as described in NRC Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests, (Ref. 15), nor does the NRC staff intend to use the guidance to affect the issue finality of an approval under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants. The staff also does not intend to use the guidance to support NRC staff actions in a manner that constitutes forward fitting as that term is defined and described in Management Directive 8.4. If a licensee believes that the NRC is using this regulatory guide in a manner inconsistent with the discussion in this Implementation section, then the licensee may file a backfitting or forward fitting appeal with the NRC in accordance with the process in Management Directive 8.4.
DG-1393, Page 8 Preliminary Draft for ACRS Consideration
 
REFERENCES 1
: 1.      Institute for Electrical and Electronics Engineers (IEEE) Std. 1205-2014, IEEE Guide for Assessing, Monitoring, and Mitigating Aging Effects on Electrical Equipment Used in Nuclear Power Generating Stations and Other Nuclear Facilities, Piscataway, NJ.2
: 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.      CFR, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, Part 54, Chapter 1, Title 10, Energy.
: 5.      Atomic Energy Act of 1954, as amended, Section 42, United States Code (U.S.C.) § 2161, et seq.
: 6.      Energy Reorganization Act of 1974, Public Law 93-438, 88 Stat. 1233.
: 7.      U.S. Nuclear Regulatory Commission (NRC), Regulatory Guide 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Washington DC.
: 8.      NRC, Regulatory Guide 1.218, Condition Monitoring Techniques for Electric Cables Used in Nuclear Power Plants, Washington DC.
: 9.      NRC, NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, Revision 2, Washington DC, December 2010. (ADAMS Accession No. ML103490036)
: 10.      NRC, NUREG-1801, Generic Aging Lessons Learned (GALL) Report, Revision 2, Washington, DC, December 2010. (ADAMS Accession No. ML103490041)
: 11.      NRC, NUREG-2191, Volumes 1 and 2, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, Washington, DC, July 2017. (ADAMS Accession Nos. ML17187A031 and ML17187A204)
: 12.      NRC, NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power PlantsFinal Report, Washington, DC, July 2017. (ADAMS Accession No. 17188A158).
: 13.      NRC, Generic Letter 2007-01, Inaccessible or Underground Power Cable Failures that Disable Accident Mitigation Systems or Cause Plant Transients, Washington, DC, February 7, 2007.
1  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 1  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 the IEEEs public Web site at http://www.ieee.org/publications_standards/index.html DG-1393, Page 9 Preliminary Draft for ACRS Consideration
: 14. International Atomic Energy Agency, Safety Reports Series No. 82, Ageing Management for Nuclear Power Plants: International Generic Ageing Lessons Learned (IGALL), Vienna, Austria, 20153
: 15. NRC, Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests, Washington, DC, September 20, 2019.
3  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, P.O. Box 100 Wagramer Strasse 5, A-1400 Vienna, Austria.
DG-1393, Page 10 Preliminary Draft for ACRS Consideration}}

Revision as of 18:44, 16 January 2022

ACRS Version DG-1393 (Rg 1.248 Rev. 0) (Advance Copy for ACRS Use)
ML21288A565
Person / Time
Issue date: 11/09/2021
From: Mohammad Sadollah
NRC/RES/DE
To:
Jenkins R
Shared Package
ML21271A013 List:
References
DG-1393 RG-1.248
Download: ML21288A565 (10)
v • d • e


Text

Retrieved from "https://kanterella.com/w/index.php?title=ML21288A565&oldid=3390216"